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Sample records for argonne high flux reactor

  1. High Flux Isotope Reactor (HFIR)

    Data.gov (United States)

    Federal Laboratory Consortium — The HFIR at Oak Ridge National Laboratory is a light-water cooled and moderated reactor that is the United States’ highest flux reactor-based neutron source. HFIR...

  2. British high flux beam reactor.

    Science.gov (United States)

    Egelstaff, P A

    1970-10-24

    The neutron scattering technique has become an accepted method for the study of condensed matter. Because of the great scientific and technical value of neutron experiments and the growing body of users, several proposals have been made during the past decade for a nuclear reactor devoted primarily to this technique. This article reviews the reasons for and history behind these proposals.

  3. Decommissioning of the high flux beam reactor at Brookhaven Lab

    Energy Technology Data Exchange (ETDEWEB)

    Hu, J.P. [National Synchrotron Light Source, Brookhaven Laboratory, Upton, NY 11973 (United States); Reciniello, R.N. [Radiological Control Div., Brookhaven Laboratory, Upton, NY 11973 (United States); Holden, N.E. [National Nuclear Data Center, Brookhaven Laboratory, Upton, NY 11973 (United States)

    2011-07-01

    The high-flux beam reactor (HFBR) at the Brookhaven National Laboratory was a heavy water cooled and moderated reactor that achieved criticality on Oct. 31, 1965. It operated at a power level of 40 megawatts. An equipment upgrade in 1982 allowed operations at 60 megawatts. After a 1989 reactor shutdown to reanalyze safety impact of a hypothetical loss of coolant accident, the reactor was restarted in 1991 at 30 megawatts. The HFBR was shut down in December 1996 for routine maintenance and refueling. At that time, a leak of tritiated water was identified by routine sampling of groundwater from wells located adjacent to the reactor's spent fuel pool. The reactor remained shut down for almost three years for safety and environmental reviews. In November 1999 the United States Dept. of Energy decided to permanently shut down the HFBR. The decontamination and decommissioning of the HFBR complex, consisting of multiple structures and systems to operate and maintain the reactor, were complete in 2009 after removing and shipping off all the control rod blades. The emptied and cleaned HFBR dome, which still contains the irradiated reactor vessel, is presently under 24/7 surveillance for safety. Detailed dosimetry performed for the HFBR decommissioning during 1996-2009 is described in the paper. (authors)

  4. Decommissioning of the High Flux Beam Reactor at Brookhaven Lab

    Energy Technology Data Exchange (ETDEWEB)

    Hu, J. P. [Brookhaven National Lab. (BNL), Upton, NY (United States); Reciniello, R. N. [Brookhaven National Lab. (BNL), Upton, NY (United States); Holden, N. E. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2011-05-27

    The High Flux Beam Reactor at the Brookhaven National Laboratory was a heavy water cooled and moderated reactor that achieved criticality on October 31, 1965. It operated at a power level of 40 mega-watts. An equipment upgrade in 1982 allowed operations at 60 mega-watts. After a 1989 reactor shutdown to reanalyze safety impact of a hypothetical loss of coolant accident, the reactor was restarted in 1991 at 30 mega-watts. The HFBR was shutdown in December 1996 for routine maintenance and refueling. At that time, a leak of tritiated water was identified by routine sampling of ground water from wells located adjacent to the reactor’s spent fuel pool. The reactor remained shutdown for almost three years for safety and environmental reviews. In November 1999 the United States Department of Energy decided to permanently shutdown the HFBR. The decontamination and decommissioning of the HFBR complex, consisting of multiple structures and systems to operate and maintain the reactor, were complete in 2009 after removing and shipping off all the control rod blades. The emptied and cleaned HFBR dome which still contains the irradiated reactor vessel is presently under 24/7 surveillance for safety. Details of the HFBR cleanup conducted during 1999-2009 will be described in the paper.

  5. RADIATION DOSIMETRY AT THE BNL HIGH FLUX BEAM REACTOR AND MEDICAL RESEARCH REACTOR.

    Energy Technology Data Exchange (ETDEWEB)

    HOLDEN,N.E.

    1999-09-10

    RADIATION DOSIMETRY MEASUREMENTS HAVE BEEN PERFORMED OVER A PERIOD OF MANY YEARS AT THE HIGH FLUX BEAM REACTOR (HFBR) AND THE MEDICAL RESEARCH REACTOR (BMRR) AT BROOKHAVEN NATIONAL LABORATORY TO PROVIDE INFORMATION ON THE ENERGY DISTRIBUTION OF THE NEUTRON FLUX, NEUTRON DOSE RATES, GAMMA-RAY FLUXES AND GAMMA-RAY DOSE RATES. THE MCNP PARTICLE TRANSPORT CODE PROVIDED MONTE CARLO RESULTS TO COMPARE WITH VARIOUS DOSIMETRY MEASUREMENTS PERFORMED AT THE EXPERIMENTAL PORTS, AT THE TREATMENT ROOMS AND IN THE THIMBLES AT BOTH HFBR AND BMRR.

  6. High Flux Metallic Membranes for Hydrogen Recovery and Membrane Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Buxbaum, Robert

    2010-06-30

    We made and tested over 250 new alloys for use as lower cost, higher flux hydrogen extraction membrane materials. Most of these were intermetallic, or contained significant intermetallic content, particularly based on B2 alloy compositions with at least one refractory component; B2 intermetallics resemble BCC alloys, in structure, but the atoms have relatively fixed positions, with one atom at the corners of the cube, the other at the centers. The target materals we were looking for would contain little or no expensive elements, no strongly toxic or radioactive elements, would have high flux to hydrogen, while being fabricable, brazable, and relatively immune to hydrogen embrittlement and corrosion in operation. The best combination of properties of the membrane materials we developed was, in my opinion, a Pd-coated membrane consisting of V -9 atomic % Pd. This material was relatively cheap, had 5 times the flux of Pd under the same pressure differential, was reasonably easy to fabricate and braze, and not bad in terms of embrittlement. Based on all these factors we project, about 1/3 the cost of Pd, on an area basis for a membrane designed to last 20 years, or 1/15 the cost on a flux basis. Alternatives to this membrane replaced significant fractions of the Pd with Ni and or Co. The cost for these membranes was lower, but so was the flux. We produced successful brazed products from the membrane materials, and made them into flat sheets. We tested, unsuccessfully, several means of fabricating thematerials into tubes, and eventually built a membrane reactor using a new, flat-plate design: a disc and doughnut arrangement, a design that seems well- suited to clean hydrogen production from coal. The membranes and reactor were tested successfully at Western Research. A larger equipment company (Chart Industries) produced similar results using a different flat-plate reactor design. Cost projections of the membrane are shown to be attractive.

  7. Fabrication of control rods for the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sease, J.D.

    1998-03-01

    The High Flux Isotope Reactor (HFIR) is a research-type nuclear reactor that was designed and built in the early 1960s and has been in continuous operation since its initial criticality in 1965. Under current plans, the HFIR is expected to continue in operation until 2035. This report updates ORNL/TM-9365, Fabrication Procedure for HFIR Control Plates, which was mainly prepared in the early 1970's but was not issued until 1984, and reflects process changes, lessons learned in the latest control rod fabrication campaign, and suggested process improvements to be considered in future campaigns. Most of the personnel involved with the initial development of the processes and in part campaigns have retired or will retire soon. Because their unlikely availability in future campaigns, emphasis has been placed on providing some explanation of why the processes were selected and some discussions about the importance of controlling critical process parameters. Contained in this report is a description of the function of control rods in the reactor, the brief history of the development of control rod fabrication processes, and a description of procedures used in the fabrication of control rods. A listing of the controlled documents and procedures used in the last fabrication campaigns is referenced in Appendix A.

  8. High Flux Isotope Reactor system RELAP5 input model

    Energy Technology Data Exchange (ETDEWEB)

    Morris, D.G.; Wendel, M.W.

    1993-01-01

    A thermal-hydraulic computational model of the High Flux Isotope Reactor (HFIR) has been developed using the RELAP5 program. The purpose of the model is to provide a state-of-the art thermal-hydraulic simulation tool for analyzing selected hypothetical accident scenarios for a revised HFIR Safety Analysis Report (SAR). The model includes (1) a detailed representation of the reactor core and other vessel components, (2) three heat exchanger/pump cells, (3) pressurizing pumps and letdown valves, and (4) secondary coolant system (with less detail than the primary system). Data from HFIR operation, component tests, tests in facility mockups and the HFIR, HFIR specific experiments, and other pertinent experiments performed independent of HFIR were used to construct the model and validate it to the extent permitted by the data. The detailed version of the model has been used to simulate loss-of-coolant accidents (LOCAs), while the abbreviated version has been developed for the operational transients that allow use of a less detailed nodalization. Analysis of station blackout with core long-term decay heat removal via natural convection has been performed using the core and vessel portions of the detailed model.

  9. High Flux Isotope Reactor cold neutron source reference design concept

    Energy Technology Data Exchange (ETDEWEB)

    Selby, D.L.; Lucas, A.T.; Hyman, C.R. [and others

    1998-05-01

    In February 1995, Oak Ridge National Laboratory`s (ORNL`s) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH{sub 2}) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH{sub 2} cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept.

  10. High flux isotope reactor cold source preconceptual design study report

    Energy Technology Data Exchange (ETDEWEB)

    Selby, D.L.; Bucholz, J.A.; Burnette, S.E. [and others

    1995-12-01

    In February 1995, the deputy director of Oak Ridge National Laboratory (ORNL) formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced Neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. The anticipated cold source will consist of a cryogenic LH{sub 2} moderator plug, a cryogenic pump system, a refrigerator that uses helium gas as a refrigerant, a heat exchanger to interface the refrigerant with the hydrogen loop, liquid hydrogen transfer lines, a gas handling system that includes vacuum lines, and an instrumentation and control system to provide constant system status monitoring and to maintain system stability. The scope of this project includes the development, design, safety analysis, procurement/fabrication, testing, and installation of all of the components necessary to produce a working cold source within an existing HFIR beam tube. This project will also include those activities necessary to transport the cold neutron beam to the front face of the present HFIR beam room. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and research and development (R and D), (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the preconceptual phase and establishes the concept feasibility. The information presented includes the project scope, the preliminary design requirements, the preliminary cost and schedule, the preliminary performance data, and an outline of the various plans for completing the project.

  11. Level 1 Tornado PRA for the High Flux Beam Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bozoki, G.E.; Conrad, C.S.

    1994-05-01

    This report describes a risk analysis primarily directed at providing an estimate for the frequency of tornado induced damage to the core of the High Flux Beam Reactor (HFBR), and thus it constitutes a Level 1 Probabilistic Risk Assessment (PRA) covering tornado induced accident sequences. The basic methodology of the risk analysis was to develop a ``tornado specific`` plant logic model that integrates the internal random hardware failures with failures caused externally by the tornado strike and includes operator errors worsened by the tornado modified environment. The tornado hazard frequency, as well as earlier prepared structural and equipment fragility data, were used as input data to the model. To keep modeling/calculational complexity as simple as reasonable a ``bounding`` type, slightly conservative, approach was applied. By a thorough screening process a single dominant initiating event was selected as a representative initiator, defined as: ``Tornado Induced Loss of Offsite Power.`` The frequency of this initiator was determined to be 6.37E-5/year. The safety response of the HFBR facility resulted in a total Conditional Core Damage Probability of .621. Thus, the point estimate of the HFBR`s Tornado Induced Core Damage Frequency (CDF) was found to be: (CDF){sub Tornado} = 3.96E-5/year. This value represents only 7.8% of the internal CDF and thus is considered to be a small contribution to the overall facility risk expressed in terms of total Core Damage Frequency. In addition to providing the estimate of (CDF){sub Tornado}, the report documents, the relative importance of various tornado induced system, component, and operator failures that contribute most to (CDF){sub Tornado}.

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

    Energy Technology Data Exchange (ETDEWEB)

    Bryant, Rebecca [Bryant Research, LLC; Kszos, Lynn A [ORNL

    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

  13. Operating manual for the High Flux Isotope Reactor. Volume I. Description of the facility

    Energy Technology Data Exchange (ETDEWEB)

    1982-09-01

    This volume contains a comprehensive description of the High Flux Isotope Reactor Facility. Its primary purpose is to supplement the detailed operating procedures, providing the reactor operators with background information on the various HFIR systems. The detailed operating procdures are presented in another report.

  14. Dissolution Flowsheet for High Flux Isotope Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, W. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Karay, N. S [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-09-27

    As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U3O8) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H2. The HFIR fuel cores will be dissolved and the recovered U will be down-blended into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H2) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy, and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H2 and other permanent gases in the dissolution offgas, allowing the development of H2 generation rate profiles. The H2 generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the

  15. Dissolution Flowsheet for High Flux Isotope Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, W. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Karay, N. S [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-09-27

    As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U3O8) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H2. The HFIR fuel cores will be dissolved and the recovered U will be down-blended into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H2) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H2 and other permanent gases in the dissolution offgas allowing the development of H2 generation rate profiles. The H2 generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the calculated lower

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  17. COMPARISON OF COOLING SCHEMES FOR HIGH HEAT FLUX COMPONENTS COOLING IN FUSION REACTORS

    Directory of Open Access Journals (Sweden)

    Phani Kumar Domalapally

    2015-04-01

    Full Text Available Some components of the fusion reactor receives high heat fluxes either during the startup and shutdown or during the operation of the machine. This paper analyzes different ways of enhancing heat transfer using helium and water for cooling of these high heat flux components and then conclusions are drawn to decide the best choice of coolant, for usage in near and long term applications.

  18. High-flux first-wall design for a small reversed-field pinch reactor

    Science.gov (United States)

    Cort, G. E.; Graham, A. L.; Christensen, K. E.

    To achieve the goal of a commercially economical fusion power reactor, small physical size and high power density should be combined with simplicity (minimized use of high technology systems). The Reversed-Field Pinch (RFP) is a magnetic confinement device that promises to meet these requirements with power densities comparable to those in existing fission power plants. To establish feasibility of such an RFP reactor, a practical design for a first wall capable of withstanding high levels of cyclic neutron wall loadings is needed. Associated with the neutron flux in the proposed RFP reactor is a time averaged heat flux of 4.5 MW/sq m with a conservatively estimated transient peak approximately twice the average value. The design for a modular first wall made from a high-strength copper alloy that will meet these requirements of cyclic thermal loading is presented. The heat removal from the wall is by subcooled water flowing in straight tubes at high linear velocities.

  19. Next generation fuel irradiation capability in the High Flux Reactor Petten

    Energy Technology Data Exchange (ETDEWEB)

    Fuetterer, Michael A., E-mail: michael.fuetterer@jrc.n [European Commission, Joint Research Centre, Institute for Energy (JRC-IE), P.O. Box 2, NL-1755 ZG Petten (Netherlands); D' Agata, Elio; Laurie, Mathias; Marmier, Alain; Scaffidi-Argentina, Francesco [European Commission, Joint Research Centre, Institute for Energy (JRC-IE), P.O. Box 2, NL-1755 ZG Petten (Netherlands); Raison, Philippe [European Commission, Joint Research Centre, Institute for Transuranium Elements (JRC-ITU), D-76334 Eggenstein-Leopoldshafen (Germany); Bakker, Klaas; Groot, Sander de; Klaassen, Frodo [Nuclear Research and consultancy Group (NRG), P.O. Box 25, NL-1755 ZG Petten (Netherlands)

    2009-07-15

    This paper describes selected equipment and expertise on fuel irradiation testing at the High Flux Reactor (HFR) in Petten, The Netherlands. The reactor went critical in 1961 and holds an operating license up to at least 2015. While HFR has initially focused on Light Water Reactor fuel and materials, it also played a decisive role since the 1970s in the German High Temperature Reactor (HTR) development program. A variety of tests related to fast reactor development in Europe were carried out for next generation fuel and materials, in particular for Very High Temperature Reactor (V/HTR) fuel, fuel for closed fuel cycles (U-Pu and Th-U fuel cycle) and transmutation, as well as for other innovative fuel types. The HFR constitutes a significant European infrastructure tool for the development of next generation reactors. Experimental facilities addressed include V/HTR fuel tests, a coated particle irradiation rig, and tests on fast reactor, transmutation and thorium fuel. The rationales for these tests are given, results are provided and further work is outlined.

  20. A conceptual high flux reactor design with scope for use in ADS applications

    Science.gov (United States)

    Pal, Usha; Jagannathan, V.

    2007-02-01

    A 100 MWt reactor design has been conceived to support flux level of the order of 1015 n/cm2/s in selected flux trap zones. The physics design considers high enriched metallic alloy fuel in the form of annular plates placed in a D2O moderator tank in a hexagonal lattice arrangement. By choosing a tight lattice pitch in the central region and double the lattice pitch in the outer region, it is possible to have both high fast flux and thermal flux trap zones. By design the flux level in the seed fuel has been kept lower than in the high flux trap zones so that the burning rate of the seed is reduced. Another important objective of the design is to maximize the time interval of refueling. As against a typical refueling interval of a few weeks in such high flux reactor cores, it is desired to maximize this period to as much as six months or even one year. This is possible to achieve by eliminating the conventional control absorbers and replacing them with a suitable amount of fertile material loading in the reactor. Requisite number of seedless thorium-aluminum alloy plates are placed at regular lattice locations vacated by seed fuel in alternate fuel layers. It is seen that these thorium plates are capable of acquiring asymptotic fissile content of 14 g/kg in about 100 days of irradiation at a flux level of 8 x 1014 n/cm2/s. In summary, the core has a relatively higher fast flux in the central region and high thermal flux in the outer region. The present physics design envisages a flat core excess reactivity for the longest possible cycle length of 6 months to one year. It is also possible to modify the design for constant subcriticality for about the same period or longer duration by considering neutron spallation source at the centre and curtailing the power density in the inner core region by shielding it with a layer of thoria fuel loading.

  1. A conceptual high flux reactor design with scope for use in ADS applications

    Indian Academy of Sciences (India)

    Usha Pal; V Jagannathan

    2007-02-01

    A 100 MWt reactor design has been conceived to support flux level of the order of 1015 n/cm2/s in selected flux trap zones. The physics design considers high enriched metallic alloy fuel in the form of annular plates placed in a D2O moderator tank in a hexagonal lattice arrangement. By choosing a tight lattice pitch in the central region and double the lattice pitch in the outer region, it is possible to have both high fast flux and thermal flux trap zones. By design the flux level in the seed fuel has been kept lower than in the high flux trap zones so that the burning rate of the seed is reduced. Another important objective of the design is to maximize the time interval of refueling. As against a typical refueling interval of a few weeks in such high flux reactor cores, it is desired to maximize this period to as much as six months or even one year. This is possible to achieve by eliminating the conventional control absorbers and replacing them with a suitable amount of fertile material loading in the reactor. Requisite number of seedless thorium–aluminum alloy plates are placed at regular lattice locations vacated by seed fuel in alternate fuel layers. It is seen that these thorium plates are capable of acquiring asymptotic fissile content of 14 g/kg in about 100 days of irradiation at a flux level of 8 × 1014 n/cm2 /s. In summary, the core has a relatively higher fast flux in the central region and high thermal flux in the outer region. The present physics design envisages a flat core excess reactivity for the longest possible cycle length of 6 months to one year. It is also possible to modify the design for constant subcriticality for about the same period or longer duration by considering neutron spallation source at the centre and curtailing the power density in the inner core region by shielding it with a layer of thoria fuel loading.

  2. Partial Safety Analysis for a Reduced Uranium Enrichment Core for the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Primm, Trent [ORNL; Gehin, Jess C [ORNL

    2009-04-01

    A computational model of the reactor core of the High Flux Isotope Rector (HFIR) was developed in order to analyze non-destructive accidents caused by transients during reactor operation. The reactor model was built for the latest version of the nuclear analysis software package called Program for the Analysis of Reactor Transients (PARET). Analyses performed with the model constructed were compared with previous data obtained with other tools in order to benchmark the code. Finally, the model was used to analyze the behavior of the reactor under transients using a different nuclear fuel with lower enrichment of uranium (LEU) than the fuel currently used, which has a high enrichment of uranium (HEU). The study shows that the presence of fertile isotopes in LEU fuel, which increases the neutron resonance absorption, reduces the impact of transients on the fuel and enhances the negative reactivity feedback, thus, within the limitations of this study, making LEU fuel appear to be a safe alternative fuel for the reactor core.

  3. Neutron Radiography Facility at IBR-2 High Flux Pulsed Reactor: First Results

    Science.gov (United States)

    Kozlenko, D. P.; Kichanov, S. E.; Lukin, E. V.; Rutkauskas, A. V.; Bokuchava, G. D.; Savenko, B. N.; Pakhnevich, A. V.; Rozanov, A. Yu.

    A neutron radiography and tomography facilityhave been developed recently at the IBR-2 high flux pulsed reactor. The facility is operated with the CCD-camera based detector having maximal field of view of 20x20 cm, and the L/D ratio can be varied in the range 200 - 2000. The first results of the radiography and tomography experiments with industrial materials and products, paleontological and geophysical objects, meteorites, are presented.

  4. Neutron Spectral Brightness of Cold Guide 4 at the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Winn, B L; Robertson, J L; Iverson, E B; Selby, D L, E-mail: winnbl@ornl.gov

    2010-11-01

    The High Flux Isotope Reactor resumed operation in June of 2007 with a supercritical hydrogen cold source in horizontal beam tube 4. Cold guide 4 is a guide system designed to deliver neutrons from this source with a reasonable flux at wavelengths greater than 4 A to several instruments, and includes a 15-m, 96-section, 4-channel bender. A time-of-flight spectrum with calibrated detector was recorded at port C of cold guide 4, and compared to McStas simulations, to generate a brightness spectrum.

  5. Neutron Spectral Brightness of Cold Guide 4 at the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Winn, B. L. [Brookhaven National Lab. (BNL), Upton, NY (United States). Neutron Scattering Group; Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Neutron Facilities Development Div.; Robertson, J. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Neutron Facilities Development Div.; Iverson, Erik B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Neutron Facilities Development Div.; Selby, D. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Neutron Facilities Development Div.

    2009-05-03

    The High Flux Isotope Reactor resumed operation in June of 2007 with a super-critical hydrogen cold source in horizontal beam tube 4. Cold guide 4 is a guide system designed to deliver neutrons from this source at reasonable flux at wavelengths greater than 4 Å to several instruments, and includes a 15-m, 96-section, 4-channel bender. A time-of-flight spectrum with calibrated detector was recorded at port C of cold guide 4, and compared to McStas simulations, to generate a brightness spectrum.

  6. Neutron Spectral Brightness of Cold Guide 4 at the High Flux Isotope Reactor

    Science.gov (United States)

    Winn, B. L.; Robertson, J. L.; Iverson, E. B.; Selby, D. L.

    2010-11-01

    The High Flux Isotope Reactor resumed operation in June of 2007 with a supercritical hydrogen cold source in horizontal beam tube 4. Cold guide 4 is a guide system designed to deliver neutrons from this source with a reasonable flux at wavelengths greater than 4 Å to several instruments, and includes a 15-m, 96-section, 4-channel bender. A time-of-flight spectrum with calibrated detector was recorded at port C of cold guide 4, and compared to McStas simulations, to generate a brightness spectrum.

  7. Advanced Multiphysics Thermal-Hydraulics Models for the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jain, Prashant K [ORNL; Freels, James D [ORNL

    2015-01-01

    Engineering design studies to determine the feasibility of converting the High Flux Isotope Reactor (HFIR) from using highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL). This work is part of an effort sponsored by the US Department of Energy (DOE) Reactor Conversion Program. HFIR is a very high flux pressurized light-water-cooled and moderated flux-trap type research reactor. HFIR s current missions are to support neutron scattering experiments, isotope production, and materials irradiation, including neutron activation analysis. Advanced three-dimensional multiphysics models of HFIR fuel were developed in COMSOL software for safety basis (worst case) operating conditions. Several types of physics including multilayer heat conduction, conjugate heat transfer, turbulent flows (RANS model) and structural mechanics were combined and solved for HFIR s inner and outer fuel elements. Alternate design features of the new LEU fuel were evaluated using these multiphysics models. This work led to a new, preliminary reference LEU design that combines a permanent absorber in the lower unfueled region of all of the fuel plates, a burnable absorber in the inner element side plates, and a relocated and reshaped (but still radially contoured) fuel zone. Preliminary results of estimated thermal safety margins are presented. Fuel design studies and model enhancement continue.

  8. Application of the successive linear programming technique to the optimum design of a high flux reactor using LEU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mo, S.C.

    1991-01-01

    The successive linear programming technique is applied to obtain the optimum thermal flux in the reflector region of a high flux reactor using LEU fuel. The design variables are the reactor power, core radius and coolant channel thickness. The constraints are the cycle length, average heat flux and peak/average power density ratio. The characteristics of the optimum solutions with various constraints are discussed.

  9. Application of the successive linear programming technique to the optimum design of a high flux reactor using LEU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mo, S.C.

    1991-12-31

    The successive linear programming technique is applied to obtain the optimum thermal flux in the reflector region of a high flux reactor using LEU fuel. The design variables are the reactor power, core radius and coolant channel thickness. The constraints are the cycle length, average heat flux and peak/average power density ratio. The characteristics of the optimum solutions with various constraints are discussed.

  10. RADIATION DOSIMETRY OF THE PRESSURE VESSEL INTERNALS OF THE HIGH FLUX BEAM REACTOR.

    Energy Technology Data Exchange (ETDEWEB)

    HOLDEN,N.E.; RECINIELLO,R.N.; HU,J.P.; RORER,D.C.

    2002-08-18

    In preparation for the eventual decommissioning of the High Flux Beam Reactor after the permanent removal of its fuel elements from the Brookhaven National Laboratory, both measurements and calculations of the decay gamma-ray dose rate have been performed for the reactor pressure vessel and vessel internal structures which included the upper and lower thermal shields, the transition plate, and the control rod blades. The measurements were made using Red Perspex{trademark} polymethyl methacrylate high-level film dosimeters, a Radcal ''peanut'' ion chamber, and Eberline's high-range ion chamber. To compare with measured gamma-ray dose rate, the Monte Carlo MCNP code and geometric progressive Microshield code were used to model the gamma transport and dose buildup.

  11. Experimental study of thermal crisis in connection with Tokamak reactor high heat flux components

    Science.gov (United States)

    Gallo, D.; Giardina, M.; Castiglia, F.; Celata, G. P.; Mariani, A.; Zummo, G.; Cumo, M.

    2000-04-01

    The results of an experimental research on high heat flux thermal crisis in forced convective subcooled water flow, under operative conditions of interest to the thermal-hydraulic design of TOKAMAK fusion reactors, are here reported. These experiments, carried out in the framework of a collaboration between the Nuclear Engineering Department of Palermo University and the National Institute of Thermal - Fluid Dynamics of the ENEA - Casaccia (Rome), were performed on the STAF (Scambio Termico Alti Flussi) water loop and consisted, essentially, in a high speed photographic study which enabled focusing several information on bubble characteristics and flow patterns taking place during the burnout phenomenology.

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

    Science.gov (United States)

    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.

  13. Establishing Specifications for Low Enriched Uranium Fuel Operations Conducted Outside the High Flux Isotope Reactor Site

    Energy Technology Data Exchange (ETDEWEB)

    Pinkston, Daniel [ORNL; Primm, Trent [ORNL; Renfro, David G [ORNL; Sease, John D [ORNL

    2010-10-01

    The National Nuclear Security Administration (NNSA) has funded staff at Oak Ridge National Laboratory (ORNL) to study the conversion of the High Flux Isotope Reactor (HFIR) from the current, high enriched uranium fuel to low enriched uranium fuel. The LEU fuel form is a metal alloy that has never been used in HFIR or any HFIR-like reactor. This report provides documentation of a process for the creation of a fuel specification that will meet all applicable regulations and guidelines to which UT-Battelle, LLC (UTB) the operating contractor for ORNL - must adhere. This process will allow UTB to purchase LEU fuel for HFIR and be assured of the quality of the fuel being procured.

  14. Heat Transfer Calculation on Plate-Type Fuel Assembly of High Flux Research Reactor

    Directory of Open Access Journals (Sweden)

    Daxin Gong

    2015-01-01

    Full Text Available Heat transfer characteristics of fuel assemblies for a high flux research reactor with a neutron trap are numerically investigated in this study. Single-phase turbulence flow is calculated by a commercial code, FLUENT, where the computational objective covers standard and control fuel assemblies. The simulation is carried out with an inlet coolant velocity varying from 4.5 m/s to 7.5 m/s in hot assemblies. The results indicate that the cladding temperature is always lower than the saturation temperature in the calculated ranges. The temperature rise in the control fuel assembly is smaller than that of the standard fuel assembly. Additionally, the assembly with a hot spot is specially studied, and the safety of the research reactor is also approved.

  15. Development of a Scale Model for High Flux Isotope Reactor Cycle 400

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Dan [ORNL

    2012-03-01

    The development of a comprehensive SCALE computational model for the High Flux Isotope Reactor (HFIR) is documented and discussed in this report. The SCALE model has equivalent features and functionality as the reference MCNP model for Cycle 400 that has been used extensively for HFIR safety analyses and for HFIR experiment design and analyses. Numerical comparisons of the SCALE and MCNP models for the multiplication constant, power density distribution in the fuel, and neutron fluxes at several locations in HFIR indicate excellent agreement between the results predicted with the two models. The SCALE HFIR model is presented in sufficient detail to provide the users of the model with a tool that can be easily customized for various safety analysis or experiment design requirements.

  16. Status of FeCrAl ODS Irradiations in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fuel Cycle Research and Development (FCRD); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fuel Cycle Research and Development (FCRD)

    2016-08-19

    FeCrAl oxide-dispersion strengthened (ODS) alloys are an attractive sub-set alloy class of the more global FeCrAl material class for nuclear applications due to their high-temperature steam oxidation resistance and hypothesized enhanced radiation tolerance. A need currently exists to determine the radiation tolerance of these newly developed alloys. To address this need, a preliminary study was conducted using the High Flux Isotope Reactor (HFIR) to irradiate an early generation FeCrAl ODS alloy, 125YF. Preliminary post-irradiation examination (PIE) on these irradiated specimens have shown good radiation tolerance at elevated temperatures (≥330°C) but possible radiation-induced hardening and embrittlement at irradiations of 200°C to a damage level of 1.9 displacement per atom (dpa). Building on this experience, a new series of irradiations are currently being conceptualized. This irradiation series called the FCAD irradiation program will irradiate the latest generation FeCrAl ODS and FeCr ODS alloys to significantly higher doses. These experiments will provide the necessary information to determine the mechanical performance of irradiated FeCrAl ODS alloys at light water reactor and fast reactor conditions.

  17. Status of FeCrAl ODS Irradiations in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fuel Cycle Research and Development (FCRD); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fuel Cycle Research and Development (FCRD)

    2016-08-19

    FeCrAl oxide-dispersion strengthened (ODS) alloys are an attractive sub-set alloy class of the more global FeCrAl material class for nuclear applications due to their high temperature steam oxidation resistance and hypothesized enhanced radiation tolerance. A need currently exists to determine the radiation tolerance of these newly developed alloys. To address this need, a preliminary study was conducted using the High Flux Isotope Reactor (HFIR) to irradiate an early generation FeCrAl ODS alloy, 125YF. Preliminary post-irradiation examination (PIE) on these irradiated specimens have shown good radiation tolerance at elevated temperatures (≥330°C) but possible radiation-induced hardening and embrittlement at irradiations of 200°C to a damage level of 1.9 displacement per atom (dpa). Building on this experience, a new series of irradiations are currently being conceptualized. This irradiation series called the FCAD irradiation program will irradiate the latest generation FeCrAl ODS and FeCr ODS alloys to significantly higher doses. These experiments will provide the necessary information to determine the mechanical performance of irradiated FeCrAl ODS alloys at light water reactor and fast reactor conditions.

  18. Countercurrent flow limited (CCFL) heat flux in the high flux isotope reactor (HFIR) fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Ruggles, A.E.

    1990-10-12

    The countercurrent flow (CCF) performance in the fuel element region of the HFIR is examined experimentally and theoretically. The fuel element consists of two concentric annuli filled with aluminum clad fuel plates of 1.27 mm thickness separated by 1.27 mm flow channels. The plates are curved as they go radially outward to accomplish constant flow channel width and constant metal-to-coolant ratio. A full-scale HFIR fuel element mock-up is studied in an adiabatic air-water CCF experiment. A review of CCF models for narrow channels is presented along with the treatment of CCFs in system of parallel channels. The experimental results are related to the existing models and a mechanistic model for the annular'' CCF in a narrow channel is developed that captures the data trends well. The results of the experiment are used to calculate the CCFL heat flux of the HFIR fuel assembly. It was determined that the HFIR fuel assembly can reject 0.62 Mw of thermal power in the CCFL situation. 31 refs., 17 figs.

  19. Modeling and Depletion Simulations for a High Flux Isotope Reactor Cycle with a Representative Experiment Loading

    Energy Technology Data Exchange (ETDEWEB)

    Chandler, David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Betzler, Ben [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Hirtz, Gregory John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Sunny, Eva [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division

    2016-09-01

    The purpose of this report is to document a high-fidelity VESTA/MCNP High Flux Isotope Reactor (HFIR) core model that features a new, representative experiment loading. This model, which represents the current, high-enriched uranium fuel core, will serve as a reference for low-enriched uranium conversion studies, safety-basis calculations, and other research activities. A new experiment loading model was developed to better represent current, typical experiment loadings, in comparison to the experiment loading included in the model for Cycle 400 (operated in 2004). The new experiment loading model for the flux trap target region includes full length 252Cf production targets, 75Se production capsules, 63Ni production capsules, a 188W production capsule, and various materials irradiation targets. Fully loaded 238Pu production targets are modeled in eleven vertical experiment facilities located in the beryllium reflector. Other changes compared to the Cycle 400 model are the high-fidelity modeling of the fuel element side plates and the material composition of the control elements. Results obtained from the depletion simulations with the new model are presented, with a focus on time-dependent isotopic composition of irradiated fuel and single cycle isotope production metrics.

  20. Structural biology facilities at Brookhaven National Laboratory`s high flux beam reactor

    Energy Technology Data Exchange (ETDEWEB)

    Korszun, Z.R.; Saxena, A.M.; Schneider, D.K. [Brookhaven National Laboratory, Upton, NY (United States)

    1994-12-31

    The techniques for determining the structure of biological molecules and larger biological assemblies depend on the extent of order in the particular system. At the High Flux Beam Reactor at the Brookhaven National Laboratory, the Biology Department operates three beam lines dedicated to biological structure studies. These beam lines span the resolution range from approximately 700{Angstrom} to approximately 1.5{Angstrom} and are designed to perform structural studies on a wide range of biological systems. Beam line H3A is dedicated to single crystal diffraction studies of macromolecules, while beam line H3B is designed to study diffraction from partially ordered systems such as biological membranes. Beam line H9B is located on the cold source and is designed for small angle scattering experiments on oligomeric biological systems.

  1. Progress towards boron neutron capture therapy at the High Flux Reactor Petten.

    Science.gov (United States)

    Moss, R L

    1990-01-01

    During 1988 the first positive steps were taken to proceed with the design and construction of a neutron capture therapy facility on the High Flux Reactor (HFR) at Petten. The immediate aim is to realise within a short time (summer 1989), an epithermal neutron beam for radiobiological and filter optimisation studies on one of the 10 small aperture horizontal beam tubes. The following summer, a much larger neutron beam, i.e., in cross section and neutron fluence rate, will be constructed on one of the two large beam tubes that replaced the old thermal column in 1984. This latter beam tube faces one whole side of the reactor vessel, extending from a 50 x 40 cm input aperture to a 35 x 35 cm exit hole. The radiotherapeutic facility will be housed here, with the intention to start clinical trials at the beginning of 1991. This paper describes the present status of the project and includes: a general description of the pertinent characteristics with respect to NCT of the HFR; results of the recently completed preliminary neutron metrology and computer modeling at the input end of the candidate beam tube; the structure and planning of the proposed Work Programme; and the respective direct and indirect participation and collaboration with the Netherlands Cancer Institute and the European Collaboration Group on BNCT.

  2. STATUS OF HIGH FLUX ISOTOPE REACTOR IRRADIATION OF SILICON CARBIDE/SILICON CARBIDE JOINTS

    Energy Technology Data Exchange (ETDEWEB)

    Katoh, Yutai [ORNL; Koyanagi, Takaaki [ORNL; Kiggans, Jim [ORNL; Cetiner, Nesrin [ORNL; McDuffee, Joel [ORNL

    2014-09-01

    Development of silicon carbide (SiC) joints that retain adequate structural and functional properties in the anticipated service conditions is a critical milestone toward establishment of advanced SiC composite technology for the accident-tolerant light water reactor (LWR) fuels and core structures. Neutron irradiation is among the most critical factors that define the harsh service condition of LWR fuel during the normal operation. The overarching goal of the present joining and irradiation studies is to establish technologies for joining SiC-based materials for use as the LWR fuel cladding. The purpose of this work is to fabricate SiC joint specimens, characterize those joints in an unirradiated condition, and prepare rabbit capsules for neutron irradiation study on the fabricated specimens in the High Flux Isotope Reactor (HFIR). Torsional shear test specimens of chemically vapor-deposited SiC were prepared by seven different joining methods either at Oak Ridge National Laboratory or by industrial partners. The joint test specimens were characterized for shear strength and microstructures in an unirradiated condition. Rabbit irradiation capsules were designed and fabricated for neutron irradiation of these joint specimens at an LWR-relevant temperature. These rabbit capsules, already started irradiation in HFIR, are scheduled to complete irradiation to an LWR-relevant dose level in early 2015.

  3. A Level 1+ Probabilistic Safety Assessment of the High Flux Australian Reactor. Vol 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-01-01

    The Department of Industry, Science and Tourism selected PLG, an EQE International Company, to systematically and independently evaluate the safety of the High Flux Australian Reactor (HIFAR), located at Lucas Heights, New South Wales. PLG performed a comprehensive probabilistic safety assessment (PSA) to quantify the risks posed by operation of HIFAR . The PSA identified possible accident scenarios, estimated their likelihood of occurrence, and assigned each scenario to a consequence category; i.e., end state. The accident scenarios developed included the possible release of radioactive material from irradiated nuclear fuel and of tritium releases from reactor coolant. The study team developed a recommended set of safety criteria against which the results of the PSA may be judged. HIFAR was found to exceed one of the two primary safety objectives and two of the five secondary safety objectives. Reactor coolant leaks, earthquakes, and coolant pump trips were the accident initiators that contributed most to scenarios that could result in fuel overheating. Scenarios initiated by earthquakes were the reason the frequency criterion for the one primary safety objective was exceeded. Overall, the plant safety status has been shown to be generally good with no evidence of major safety-related problems from its operation. One design deficiency associated with the emergency core cooling system was identified that should be corrected as soon as possible. Additionally, several analytical issues have been identified that should be investigated further. The results from these additional investigations should be used to determine whether additional plant and procedural changes are required, or if further evaluations of postulated severe accidents are warranted. Supporting information can be found in Appendix A for the seismic analysis and in the Appendix B for selected other external events refs., 139 tabs., 85 figs. Prepared for Department of Industry, Science and Tourism

  4. Utilization of the High Flux Isotope Reactor at Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Selby, Douglas L [ORNL; Bilheux, Hassina Z [ORNL; Meilleur, Flora [ORNL; Jones, Amy [ORNL; Bailey, William Barton [ORNL; Vandergriff, David H [ORNL

    2015-01-01

    This paper addresses several aspects of the scientific utilization of the Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR). Topics to be covered will include: 1) HFIR neutron scattering instruments and the formal instrument user program; 2) Recent upgrades to the neutron scattering instrument stations at the reactor, and 3) eMod a new tool for addressing instrument modifications and providing configuration control and design process for scientific instruments at HFIR and the Spallation Neutron Source (SNS). There are 15 operating neutron instrument stations at HFIR with 12 of them organized into a formal user program. Since the last presentation on HFIR instruments at IGORR we have installed a Single Crystal Quasi-Laue Diffractometer instrument called IMAGINE; and we have made significant upgrades to HFIR neutron scattering instruments including the Cold Triple Axis Instrument, the Wide Angle Neutron Diffractometer, the Powder Diffractometer, and the Neutron Imaging station. In addition, we have initiated upgrades to the Thermal Triple Axis Instrument and the Bio-SANS cold neutron instrument detector system. All of these upgrades are tied to a continuous effort to maintain a high level neutron scattering user program at the HFIR. For the purpose of tracking modifications such as those mentioned and configuration control we have been developing an electronic system for entering instrument modification requests that follows a modification or instrument project through concept development, design, fabrication, installation, and commissioning. This system, which we call eMod, electronically leads the task leader through a series of questions and checklists that then identifies such things as ES&H and radiological issues and then automatically designates specific individuals for the activity review process. The system has been in use for less than a year and we are still working out some of the inefficiencies, but we believe that this will become a very

  5. Modeling and Simulations for the High Flux Isotope Reactor Cycle 400

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). High Flux Isotope Reactor (HFIR); Chandler, David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). High Flux Isotope Reactor (HFIR); Ade, Brian J [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). High Flux Isotope Reactor (HFIR); Sunny, Eva E [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). High Flux Isotope Reactor (HFIR); Betzler, Benjamin R [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). High Flux Isotope Reactor (HFIR); Pinkston, Daniel [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). High Flux Isotope Reactor (HFIR)

    2015-03-01

    A concerted effort over the past few years has been focused on enhancing the core model for the High Flux Isotope Reactor (HFIR), as part of a comprehensive study for HFIR conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel. At this time, the core model used to perform analyses in support of HFIR operation is an MCNP model for the beginning of Cycle 400, which was documented in detail in a 2005 technical report. A HFIR core depletion model that is based on current state-of-the-art methods and nuclear data was needed to serve as reference for the design of an LEU fuel for HFIR. The recent enhancements in modeling and simulations for HFIR that are discussed in the present report include: (1) revision of the 2005 MCNP model for the beginning of Cycle 400 to improve the modeling data and assumptions as necessary based on appropriate primary reference sources HFIR drawings and reports; (2) improvement of the fuel region model, including an explicit representation for the involute fuel plate geometry that is characteristic to HFIR fuel; and (3) revision of the Monte Carlo-based depletion model for HFIR in use since 2009 but never documented in detail, with the development of a new depletion model for the HFIR explicit fuel plate representation. The new HFIR models for Cycle 400 are used to determine various metrics of relevance to reactor performance and safety assessments. The calculated metrics are compared, where possible, with measurement data from preconstruction critical experiments at HFIR, data included in the current HFIR safety analysis report, and/or data from previous calculations performed with different methods or codes. The results of the analyses show that the models presented in this report provide a robust and reliable basis for HFIR analyses.

  6. Low-Enriched Uranium Fuel Design with Two-Dimensional Grading for the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Primm, Trent [ORNL

    2011-05-01

    An engineering design study of the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel is ongoing at Oak Ridge National Laboratory. The computational models developed during fiscal year 2010 to search for an LEU fuel design that would meet the requirements for the conversion and the results obtained with these models are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating HEU fuel core. The results obtained indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations under the assumption that the operating power for the reactor fueled with LEU can be increased from the current value of 85 MW to 100 MW.

  7. Decontamination and dismantlement of the JANUS Reactor at Argonne National Laboratory-East. Project final report

    Energy Technology Data Exchange (ETDEWEB)

    Fellhauer, C.R.; Clark, F.R. [Argonne National Lab., IL (United States). Technology Development Div.; Garlock, G.A. [MOTA Corp., Cayce, SC (United States)

    1997-10-01

    The decontamination and dismantlement of the JANUS Reactor at Argonne National Laboratory-East (ANL-E) was completed in October 1997. Descriptions and evaluations of the activities performed and analyses of the results obtained during the JANUS D and D Project are provided in this Final Report. The following information is included: objective of the JANUS D and D Project; history of the JANUS Reactor facility; description of the ANL-E site and the JANUS Reactor facility; overview of the D and D activities performed; description of the project planning and engineering; description of the D and D operations; summary of the final status of the JANUS Reactor facility based upon the final survey results; description of the health and safety aspects of the project, including personnel exposure and OSHA reporting; summary of the waste minimization techniques utilized and total waste generated by the project; and summary of the final cost and schedule for the JANUS D and D Project.

  8. Simulating High Flux Isotope Reactor Core Thermal-Hydraulics via Interdimensional Model Coupling

    Energy Technology Data Exchange (ETDEWEB)

    Travis, Adam R [ORNL

    2014-05-01

    A coupled interdimensional model is presented for the simulation of the thermal-hydraulic characteristics of the High Flux Isotope Reactor core at Oak Ridge National Laboratory. The model consists of two domains a solid involute fuel plate and the surrounding liquid coolant channel. The fuel plate is modeled explicitly in three-dimensions. The coolant channel is approximated as a twodimensional slice oriented perpendicular to the fuel plate s surface. The two dimensionally-inconsistent domains are linked to one another via interdimensional model coupling mechanisms. The coupled model is presented as a simplified alternative to a fully explicit, fully three-dimensional model. Involute geometries were constructed in SolidWorks. Derivations of the involute construction equations are presented. Geometries were then imported into COMSOL Multiphysics for simulation and modeling. Both models are described in detail so as to highlight their respective attributes in the 3D model, the pursuit of an accurate, reliable, and complete solution; in the coupled model, the intent to simplify the modeling domain as much as possible without affecting significant alterations to the solution. The coupled model was created with the goal of permitting larger portions of the reactor core to be modeled at once without a significant sacrifice to solution integrity. As such, particular care is given to validating incorporated model simplifications. To the greatest extent possible, the decrease in solution time as well as computational cost are quantified versus the effects such gains have on the solution quality. A variant of the coupled model which sufficiently balances these three solution characteristics is presented alongside the more comprehensive 3D model for comparison and validation.

  9. Reactivity Accountability Attributed to Reflector Poisons in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chandler, David [ORNL; Maldonado, G Ivan [ORNL; Primm, Trent [ORNL

    2009-12-01

    The objective of this study is to develop a methodology to predict the reactivity impact as a function of outage time between cycles of 3He, 6Li, and other poisons in the High Flux Isotope Reactor s (HFIR) beryllium reflector. The reactivity worth at startup of the HFIR has been incorrectly predicted in the past after the reactor has been shut-down for long periods of time. The incorrect prediction was postulated to be due to the erroneous calculation of 3He buildup in the beryllium reflector. It is necessary to develop a better estimate of the start-of-cycle symmetric critical control element positions since if the estimated and actual symmetrical critical control element positions differ by more than $1.55 in reactivity (approximately one-half inch in control element startup position), HFIR is to be shutdown and a technical evaluation is performed to resolve the discrepancy prior to restart. 3He is generated and depleted during operation, but during an outage, the depletion of 3He ceases because it is a stable isotope. 3He is born from the radioactive decay of tritium, and thus the concentration of 3He increases during shutdown. SCALE, specifically the TRITON and CSAS5 control modules including the KENO V.A, COUPLE, and ORIGEN functional modules were utilized in this study. An equation relating the down time (td) to the change in symmetric control element position was generated and validated against measurements for approximately 40 HFIR operating cycles. The newly-derived correlation was shown to improve accuracy of predictions for long periods of down time.

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

    Energy Technology Data Exchange (ETDEWEB)

    Boing, L.E.; Henley, D.R. (Argonne National Lab., IL (USA)); Manion, W.J.; Gordon, J.W. (Nuclear Energy Services, Inc., Danbury, CT (USA))

    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.

  11. Delivery of completed irradiation vehicles and the quality assurance document to the High Flux Isotope Reactor for irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Petrie, Christian M. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); McDuffee, Joel Lee [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Katoh, Yutai [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    This report details the initial fabrication and delivery of two Fuel Cycle Research and Development (FCRD) irradiation capsules (ATFSC01 and ATFSC02), with associated quality assurance documentation, to the High Flux Isotope Reactor (HFIR). The capsules and documentation were delivered by September 30, 2015, thus meeting the deadline for milestone M3FT-15OR0202268. These irradiation experiments are testing silicon carbide composite tubes in order to obtain experimental validation of thermo-mechanical models of stress states in SiC cladding irradiated under a prototypic high heat flux. This document contains a copy of the completed capsule fabrication request sheets, which detail all constituent components, pertinent drawings, etc., along with a detailed summary of the capsule assembly process performed by the Thermal Hydraulics and Irradiation Engineering Group (THIEG) in the Reactor and Nuclear Systems Division (RNSD). A complete fabrication package record is maintained by the THIEG and is available upon request.

  12. Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2008

    Energy Technology Data Exchange (ETDEWEB)

    Primm, Trent [ORNL; Chandler, David [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Sease, John D [ORNL; Jolly, Brian C [ORNL

    2009-03-01

    This report documents progress made during FY 2008 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Scoping experiments with various manufacturing methods for forming the LEU alloy profile are presented.

  13. Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual report for FY 2009

    Energy Technology Data Exchange (ETDEWEB)

    Chandler, David [ORNL; Freels, James D [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Primm, Trent [ORNL; Sease, John D [ORNL; Guida, Tracey [University of Pittsburgh; Jolly, Brian C [ORNL

    2010-02-01

    This report documents progress made during FY 2009 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Studies are reported of the application of a silicon coating to surrogates for spheres of uranium-molybdenum alloy. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. A description of the progress in developing a finite element thermal hydraulics model of the LEU core is provided.

  14. Thermal Safety Analyses for the Production of Plutonium-238 at the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hurt, Christopher J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Freels, James D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hobbs, Randy W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jain, Prashant K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Maldonado, G. Ivan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-08-01

    There has been a considerable effort over the previous few years to demonstrate and optimize the production of plutonium-238 (238Pu) at the High Flux Isotope Reactor (HFIR). This effort has involved resources from multiple divisions and facilities at the Oak Ridge National Laboratory (ORNL) to demonstrate the fabrication, irradiation, and chemical processing of targets containing neptunium-237 (237Np) dioxide (NpO2)/aluminum (Al) cermet pellets. A critical preliminary step to irradiation at the HFIR is to demonstrate the safety of the target under irradiation via documented experiment safety analyses. The steady-state thermal safety analyses of the target are simulated in a finite element model with the COMSOL Multiphysics code that determines, among other crucial parameters, the limiting maximum temperature in the target. Safety analysis efforts for this model discussed in the present report include: (1) initial modeling of single and reduced-length pellet capsules in order to generate an experimental knowledge base that incorporate initial non-linear contact heat transfer and fission gas equations, (2) modeling efforts for prototypical designs of partially loaded and fully loaded targets using limited available knowledge of fabrication and irradiation characteristics, and (3) the most recent and comprehensive modeling effort of a fully coupled thermo-mechanical approach over the entire fully loaded target domain incorporating burn-up dependent irradiation behavior and measured target and pellet properties, hereafter referred to as the production model. These models are used to conservatively determine several important steady-state parameters including target stresses and temperatures, the limiting condition of which is the maximum temperature with respect to the melting point. The single pellet model results provide a basis for the safety of the irradiations, followed by parametric analyses in the initial prototypical designs that were necessary due to the

  15. DESIGN STUDY FOR A LOW-ENRICHED URANIUM CORE FOR THE HIGH FLUX ISOTOPE REACTOR, ANNUAL REPORT FOR FY 2010

    Energy Technology Data Exchange (ETDEWEB)

    Cook, David Howard [ORNL; Freels, James D [ORNL; Ilas, Germina [ORNL; Jolly, Brian C [ORNL; Miller, James Henry [ORNL; Primm, Trent [ORNL; Renfro, David G [ORNL; Sease, John D [ORNL; Pinkston, Daniel [ORNL

    2011-02-01

    This report documents progress made during FY 2010 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current level. Studies are reported of support to a thermal hydraulic test loop design, the implementation of finite element, thermal hydraulic analysis capability, and infrastructure tasks at HFIR to upgrade the facility for operation at 100 MW. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. Continuing development in the definition of the fuel fabrication process is described.

  16. Main technical options of the Jules Horowitz reactor project to achieve high flux performances and high safety level

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A.; Bergamaschi, Y.; Bouilloux, Y.; Bravo, X.; Guigon, B.; Rommens, M.; Tremodeux, P. [CEA Cadarache, Dir. de l' Energie Nucleaire DEN, 13 - Saint-Paul-lez-Durance (France)]|[CEA Saclay Dir. de l' Energie Nucleaire DEN, 91 - Gif sur Yvette (France)

    2003-07-01

    Since the shutdown of the SILOE reactor in 1997, the OSIRIS reactor has ensured the needs regarding technological irradiation at CEA including those of its industrial partners and customers. The Jules Horowitz Reactor will replace it and will offer a quite larger experimental field. It has the ambition to provide the necessary nuclear data and to maintain a fission research capability in Europe after 2010. The Jules Horowitz Reactor will represent a significant step in terms of performances and experimental capabilities. This paper will present the main design option resulting from the preliminary studies. The choice of the specific power around 600 kW/I for the reference core configuration is a key decision to ensure the required flux level. Consequently many choices have to be made regarding the materials used in the core and the fuel element design. These involve many specific qualifications including codes validation. The main safety options are based on: - A safety approach based upon the defence-in-depth principle. - A strategy of generic approaches to assess experimental risks in the facility. - Internal events analysis taking into account risks linked to reactor and experiments (e.g., radioactive source-term). - Systematic consideration of external hazards (e.g., earthquake, airplane crash) and internal hazards. - Design of containment to manage and mitigate a severe reactor accident (consideration of 'BORAX' accident, according to french safety practice for MTRs, beyond design basis reactivity insertion accident, involving core melting and core destruction phenomena). (authors)

  17. Design Study for a Low-enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2007

    Energy Technology Data Exchange (ETDEWEB)

    Primm, Trent [ORNL; Ellis, Ronald James [ORNL; Gehin, Jess C [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Sease, John D [ORNL

    2007-11-01

    This report documents progress made during fiscal year 2007 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low enriched uranium fuel (LEU). Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. A high volume fraction U/Mo-in-Al fuel could attain the same neutron flux performance as with the current, HEU fuel but materials considerations appear to preclude production and irradiation of such a fuel. A diffusion barrier would be required if Al is to be retained as the interstitial medium and the additional volume required for this barrier would degrade performance. Attaining the high volume fraction (55 wt. %) of U/Mo assumed in the computational study while maintaining the current fuel plate acceptance level at the fuel manufacturer is unlikely, i.e. no increase in the percentage of plates rejected for non-compliance with the fuel specification. Substitution of a zirconium alloy for Al would significantly increase the weight of the fuel element, the cost of the fuel element, and introduce an as-yet untried manufacturing process. A monolithic U-10Mo foil is the choice of LEU fuel for HFIR. Preliminary calculations indicate that with a modest increase in reactor power, the flux performance of the reactor can be maintained at the current level. A linearly-graded, radial fuel thickness profile is preferred to the arched profile currently used in HEU fuel because the LEU fuel media is a metal alloy foil rather than a powder. Developments in analysis capability and nuclear data processing techniques are underway with the goal of verifying the preliminary calculations of LEU flux performance. A conceptual study of the operational cost of an LEU fuel fabrication facility yielded the conclusion that the annual fuel cost to the HFIR would increase significantly from the current, HEU fuel cycle. Though manufacturing can be accomplished with existing technology

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

    Energy Technology Data Exchange (ETDEWEB)

    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).

  19. High-Speed Neutron and Gamma Flux Sensor for Monitoring Surface Nuclear Reactors Project

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA needs compact nuclear reactors to power future bases on the moon and/or Mars. These reactors require robust automatic control systems using low mass, rapid...

  20. High-Speed Neutron and Gamma Flux Sensor for Monitoring Surface Nuclear Reactors Project

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA needs compact nuclear reactors to power future bases on the moon and Mars. These reactors require robust automatic control systems using low mass, rapid...

  1. Feasibility analyses for HEU to LEU fuel conversion of the LAUE Langivin Institute (ILL) High Flux Reactor (RHF).

    Energy Technology Data Exchange (ETDEWEB)

    Stevens, J.; Tentner. A.; Bergeron, A.; Nuclear Engineering Division

    2010-08-19

    The High Flux Reactor (RHF) of the Laue Langevin Institute (ILL) based in Grenoble, France is a research reactor designed primarily for neutron beam experiments for fundamental science. It delivers one of the most intense neutron fluxes worldwide, with an unperturbed thermal neutron flux of 1.5 x 10{sup 15} n/cm{sup 2}/s in its reflector. The reactor has been conceived to operate at a nuclear power of 57 MW but currently operates at 52 MW. The reactor currently uses a Highly Enriched Uranium (HEU) fuel. In the framework of its 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 worldwide research and test reactors have already started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on a mixture of uranium and molybdenum (UMo) is expected to allow the conversion of compact high performance reactors like the RHF. This report presents the results of reactor design, performance and steady state safety analyses for conversion of the RHF from the use of HEU fuel to the use of UMo LEU fuel. The objective of this work was to show that is feasible, under a set of manufacturing assumptions, to design a new RHF fuel element that could safely replace the HEU element currently used. The new proposed design has been developed to maximize performance, minimize changes and preserve strong safety margins. Neutronics and thermal-hydraulics models of the RHF have been developed and qualified by benchmark against experiments and/or against other codes and models. The models developed were then used to evaluate the RHF performance if LEU UMo were to replace the current HEU fuel 'meat' without any geometric change to the fuel plates. Results of these direct replacement analyses have shown a significant degradation of the RHF performance, in terms of both neutron flux and cycle

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-03-01

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

  3. Preliminary Assessment of the Impact on Reactor Vessel dpa Rates Due to Installation of a Proposed Low Enriched Uranium (LEU) Core in the High Flux Isotope Reactor (HFIR)

    Energy Technology Data Exchange (ETDEWEB)

    Daily, Charles R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    An assessment of the impact on the High Flux Isotope Reactor (HFIR) reactor vessel (RV) displacements-per-atom (dpa) rates due to operations with the proposed low enriched uranium (LEU) core described by Ilas and Primm has been performed and is presented herein. The analyses documented herein support the conclusion that conversion of HFIR to low-enriched uranium (LEU) core operations using the LEU core design of Ilas and Primm will have no negative impact on HFIR RV dpa rates. Since its inception, HFIR has been operated with highly enriched uranium (HEU) cores. As part of an effort sponsored by the National Nuclear Security Administration (NNSA), conversion to LEU cores is being considered for future HFIR operations. The HFIR LEU configurations analyzed are consistent with the LEU core models used by Ilas and Primm and the HEU balance-of-plant models used by Risner and Blakeman in the latest analyses performed to support the HFIR materials surveillance program. The Risner and Blakeman analyses, as well as the studies documented herein, are the first to apply the hybrid transport methods available in the Automated Variance reduction Generator (ADVANTG) code to HFIR RV dpa rate calculations. These calculations have been performed on the Oak Ridge National Laboratory (ORNL) Institutional Cluster (OIC) with version 1.60 of the Monte Carlo N-Particle 5 (MCNP5) computer code.

  4. High Fluency Low Flux Embrittlement Models of LWR Reactor Pressure Vessel Embrittlement and a Supporting Database from the UCSB ATR-2 Irradiation Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Odette, G. Robert [Univ. of California, Santa Barbara, CA (United States)

    2017-01-24

    Reactor pressure vessel embrittlement may limit the lifetime of light water reactors (LWR). Embrittlement is primarily caused by formation of nano-scale precipitates, which cause hardening and a subsequent increase in the ductile-to-brittle transition temperature of the steel. While the effect of Cu has historically been the largest research focus of RPV embrittlement, there is increasing evidence that Mn, Ni and Si are likely to have a large effect at higher fluence, where Mn-Ni-Si precipitates can form, even in the absence of Cu. Therefore, extending RPV lifetimes will require a thorough understanding of both precipitation and embrittlement at higher fluences than have ever been observed in a power reactor. To address this issue, test reactors that irradiate materials at higher neutron fluxes than power reactors are used. These experiments at high neutron flux can reach extended life neutron fluences in only months or several years. The drawback of these test irradiations is that they add additional complexity to interpreting the data, as the irradiation flux also plays a role into both precipitate formation and irradiation hardening and embrittlement. This report focuses on developing a database of both microstructure and mechanical property data to better understand the effect of flux. In addition, a previously developed model that enables the comparison of data taken over a range of neutron flux is discussed.

  5. Conceptual Process for the Manufacture of Low-Enriched Uranium/Molybdenum Fuel for the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sease, J.D.; Primm, R.T. III; Miller, J.H.

    2007-09-30

    The U.S. nonproliferation policy 'to minimize, and to the extent possible, eliminate the use of HEU in civil nuclear programs throughout the world' has resulted in the conversion (or scheduled conversion) of many of the U.S. research reactors from high-enriched uranium (HEU) to low-enriched uranium (LEU). A foil fuel appears to offer the best option for using a LEU fuel in the High Flux Isotope Reactor (HFIR) without degrading the performance of the reactor. The purpose of this document is to outline a proposed conceptual fabrication process flow sheet for a new, foil-type, 19.75%-enriched fuel for HFIR. The preparation of the flow sheet allows a better understanding of the costs of infrastructure modifications, operating costs, and implementation schedule issues associated with the fabrication of LEU fuel for HFIR. Preparation of a reference flow sheet is one of the first planning steps needed in the development of a new manufacturing capacity for low enriched fuels for U.S. research and test reactors. The flow sheet can be used to develop a work breakdown structure (WBS), a critical path schedule, and identify development needs. The reference flow sheet presented in this report is specifically for production of LEU foil fuel for the HFIR. The need for an overall reference flow sheet for production of fuel for all High Performance Research Reactors (HPRR) has been identified by the national program office. This report could provide a starting point for the development of such a reference flow sheet for a foil-based fuel for all HPRRs. The reference flow sheet presented is based on processes currently being developed by the national program for the LEU foil fuel when available, processes used historically in the manufacture of other nuclear fuels and materials, and processes used in other manufacturing industries producing a product configuration similar to the form required in manufacturing a foil fuel. The processes in the reference flow sheet are

  6. Fabrication procedures for manufacturing High Flux Isotope Reactor fuel elements - 2

    Energy Technology Data Exchange (ETDEWEB)

    Knight, R.W.; Morin, R.A.

    1999-12-01

    The original fabrication procedures written in 1968 delineated the manufacturing procedures at that time. Since 1968, there have been a number of procedural changes. This rewrite of the fabrication procedures incorporates these changes. The entire fuel core of this reactor is made up of two fuel elements. Each element consists of one annular array of fuel plates. These annuli are identified as the inner and outer fuel elements, since one fits inside the other. The inner element consists of 171 identical fuel plates, and the outer element contains 369 identical fuel plates differing slightly from those in the inner element. Both sets of fuel plates contain U{sub 3}O{sub 8} powder as the fuel, dispersed in an aluminum powder matrix and clad with aluminum. Procedures for manufacturing and inspection of the fuel elements are described and illustrated.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  8. Neutronics Conversion Analyses of the Laue-Langevin Institute (ILL) High Flux Reactor (RHF)

    Energy Technology Data Exchange (ETDEWEB)

    Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Calzavara, Y. [Inst. Laue-Langevin (ILL), Grenoble (France)

    2014-09-30

    The following report describes the neutronics results obtained with the MCNP model of the RHF U7Mo LEU reference design that has been established in 2010 during the feasibility analysis. This work constitutes a complete and detailed neutronics analysis of that LEU design using models that have been significantly improved since 2010 and the release of the feasibility report. When possible, the credibility of the neutronics model is tested by comparing the HEU model results with experimental data or other codes calculations results. The results obtained with the LEU model are systematically compared to the HEU model. The changes applied to the neutronics model lead to better comparisons with experimental data or improved the calculation efficiency but do not challenge the conclusion of the feasibility analysis. If the U7Mo fuel is commercially available, not cost prohibitive, a back-end solution is established and if it is possible to manufacture the proposed element, neutronics analyses show that the performance of the reactor would not be challenged by the conversion to LEU fuel.

  9. Impact of the High Flux Isotope Reactor HEU to LEU Fuel Conversion on Cold Source Nuclear Heat Generation Rates

    Energy Technology Data Exchange (ETDEWEB)

    Chandler, David [ORNL

    2014-03-01

    Under the sponsorship of the US Department of Energy National Nuclear Security Administration, staff members at the Oak Ridge National Laboratory have been conducting studies to determine whether the High Flux Isotope Reactor (HFIR) can be converted from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. As part of these ongoing studies, an assessment of the impact that the HEU to LEU fuel conversion has on the nuclear heat generation rates in regions of the HFIR cold source system and its moderator vessel was performed and is documented in this report. Silicon production rates in the cold source aluminum regions and few-group neutron fluxes in the cold source moderator were also estimated. Neutronics calculations were performed with the Monte Carlo N-Particle code to determine the nuclear heat generation rates in regions of the HFIR cold source and its vessel for the HEU core operating at a full reactor power (FP) of 85 MW(t) and the reference LEU core operating at an FP of 100 MW(t). Calculations were performed with beginning-of-cycle (BOC) and end-of-cycle (EOC) conditions to bound typical irradiation conditions. Average specific BOC heat generation rates of 12.76 and 12.92 W/g, respectively, were calculated for the hemispherical region of the cold source liquid hydrogen (LH2) for the HEU and LEU cores, and EOC heat generation rates of 13.25 and 12.86 W/g, respectively, were calculated for the HEU and LEU cores. Thus, the greatest heat generation rates were calculated for the EOC HEU core, and it is concluded that the conversion from HEU to LEU fuel and the resulting increase of FP from 85 MW to 100 MW will not impact the ability of the heat removal equipment to remove the heat deposited in the cold source system. Silicon production rates in the cold source aluminum regions are estimated to be about 12.0% greater at BOC and 2.7% greater at EOC for the LEU core in comparison to the HEU core. Silicon is aluminum s major transmutation product and

  10. The CG-1D Neutron Imaging Beamline at the Oak Ridge National Laboratory High Flux Isotope Reactor

    Science.gov (United States)

    Santodonato, Lou; Bilheux, Hassina; Bailey, Barton; Bilheux, Jean; Nguyen, Phong; Tremsin, Anton; Selby, Doug; Walker, Lakeisha

    The Oak Ridge National Laboratory Neutron Sciences Directorate has installed a neutron imaging beamline at the High Flux Isotope Reactor (HFIR) cold guide hall. CG-1D is one of the three instruments that make up the CG1 instrument suite. The beamline optics and detector have recently been upgraded to meet the needs of the neutron imaging community (better "smoothing" of guide system artifacts, higher flux or spatial resolution). These upgrades comprise a new diffuser/aperture system, two new detectors, a He-filled flight tube and silicon (Si) windows. Shielding inside the flight tube, beam scrapers and a beam stop ensure that biological dose is less than 50 μSv/hr outside of the radiation boundary. A set of diffusers and apertures (pinhole geometry) has been installed at the exit of the guide system to allow motorized L/D variation. Samples sit on a translation/rotation stage for alignment and tomography purposes. Detectors for the CG-1D beamline are (1) an ANDOR DW936 charge coupled device (CCD) camera with a field of view of approximately 7 cm x 7 cm and ∼ 80 microns spatial resolution and 1 frame per second time resolution, (2) a new Micro-Channel Plate (MCP) detector with a 2.8 cm x 2.8 cm field of view and 55 microns spatial resolution, and 5 μs timing capability. 6LiF/ZnS scintillators of thickness varying from 50 to 200 microns are being used at this facility. An overview of the beamline upgrade and preliminary data is presented here.

  11. [Project for] a high-flux extracted neutron beam reactor [for physicists]; Un [projet de] reacteur a haut flux et faisceaux sortis [pour physiciens

    Energy Technology Data Exchange (ETDEWEB)

    Ageron, P. [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1964-07-01

    French requirements in neutron beams of different energies extracted from a reactor are briefly described. The well-known importance of cold neutrons (above 4 Angstrom) is emphasized. The main characteristics of a reactor suitable for physicists are outlined: They are: 1 - A flux of about 7. 10{sup 14} thermal neutrons in the heavy water of the reflector, 2 - Maximum flexibility obtained by: - physical separation of the core and the reflector, - independence of the different experiments, - possibility of modifying physical experiments up to - and including - the nature of the used reflector, without any appreciable interruption in the operation of the reactor, - reduction of fixed shields to a minimum; ample use of liquid shields (water) and fluid shields (sands). 3 - Technological continuity as far as possible with French research reactors (Siloe, Pegase, Osiris) already existing or under construction. 4 - Safety of operation arising from simplicity of conception. 5 - Minimised construction costs. Lowering of the operating costs is looked for indirectly in the simplification of the solutions and the reduction of operating staff, rather than directly by reducing the consumption of fuel elements and energy. The recommended solution can be described as a closed-core non-pressurized swimming-pool reactor, highly under-moderated by the cooling light water. Surrounding the reactor are a number of 'beam tubes-loops' each consisting of: - a part of the reflector (heavy water in the example described), - a part of neutron extraction beam tube, - the circuits required for their cooling, - the inlet systems of suitable fluids to the beam tube nose (liquid hydrogen in the example described), - the necessary outlets for measurement and control system. The whole 'beam tubes loops' is immersed in the water of the metallic self-supporting swimming-pool. The shielding outside the swimming-pool is composed for the most part by heavy sand in which is the rest of

  12. Assumptions and Criteria for Performing a Feasability Study of the Conversion of the High Flux Isotope Reactor Core to Use Low-Enriched Uranium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Primm, R.T., III; Ellis, R.J.; Gehin, J.C.; Moses, D.L.; Binder, J.L.; Xoubi, N. (U. of Cincinnati)

    2006-02-01

    A computational study will be initiated during fiscal year 2006 to examine the feasibility of converting the High Flux Isotope Reactor from highly enriched uranium fuel to low-enriched uranium. The study will be limited to steady-state, nominal operation, reactor physics and thermal-hydraulic analyses of a uranium-molybdenum alloy that would be substituted for the current fuel powder--U{sub 3}O{sub 8} mixed with aluminum. The purposes of this document are to (1) define the scope of studies to be conducted, (2) define the methodologies to be used to conduct the studies, (3) define the assumptions that serve as input to the methodologies, (4) provide an efficient means for communication with the Department of Energy and American research reactor operators, and (5) expedite review and commentary by those parties.

  13. Spheromak reactor with poloidal flux-amplifying transformer

    Science.gov (United States)

    Furth, Harold P.; Janos, Alan C.; Uyama, Tadao; Yamada, Masaaki

    1987-01-01

    An inductive transformer in the form of a solenoidal coils aligned along the major axis of a flux core induces poloidal flux along the flux core's axis. The current in the solenoidal coil is then reversed resulting in a poloidal flux swing and the conversion of a portion of the poloidal flux to a toroidal flux in generating a spheromak plasma wherein equilibrium approaches a force-free, minimum Taylor state during plasma formation, independent of the initial conditions or details of the formation. The spheromak plasma is sustained with the Taylor state maintained by oscillating the currents in the poloidal and toroidal field coils within the plasma-forming flux core. The poloidal flux transformer may be used either as an amplifier stage in a moving plasma reactor scenario for initial production of a spheromak plasma or as a method for sustaining a stationary plasma and further heating it. The solenoidal coil embodiment of the poloidal flux transformer can alternately be used in combination with a center conductive cylinder aligned along the length and outside of the solenoidal coil. This poloidal flux-amplifying inductive transformer approach allows for a relaxation of demanding current carrying requirements on the spheromak reactor's flux core, reduces plasma contamination arising from high voltage electrode discharge, and improves the efficiency of poloidal flux injection.

  14. Production of Medical Radioisotopes in the ORNL High Flux Isotope Reactor (HFIR) for Cancer Treatment and Arterial Restenosis Therapy after PTCA

    Science.gov (United States)

    Knapp, F. F. Jr.; Beets, A. L.; Mirzadeh, S.; Alexander, C. W.; Hobbs, R. L.

    1998-06-01

    The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) represents an important resource for the production of a wide variety of medical radioisotopes. In addition to serving as a key production site for californium-252 and other transuranic elements, important examples of therapeutic radioisotopes which are currently routinely produced in the HFIR for distribution include dysprosium-166 (parent of holmium-166), rhenium-186, tin-117m and tungsten-188 (parent of rhenium-188). The nine hydraulic tube (HT) positions in the central high flux region permit the insertion and removal of targets at any time during the operating cycle and have traditionally represented a major site for production of medical radioisotopes. To increase the irradiation capabilities of the HFIR, special target holders have recently been designed and fabricated which will be installed in the six Peripheral Target Positions (PTP), which are also located in the high flux region. These positions are only accessible during reactor refueling and will be used for long-term irradiations, such as required for the production of tin-117m and tungsten-188. Each of the PTP tubes will be capable of housing a maximum of eight HT targets, thus increasing the total maximum number of HT targets from the current nine, to a total of 57. In this paper the therapeutic use of reactor-produced radioisotopes for bone pain palliation and vascular brachytherapy and the therapeutic medical radioisotope production capabilities of the ORNL HFIR are briefly discussed.

  15. Production of medical radioisotopes in the ORNL High Flux Isotope Reactor (HFIR) for cancer treatment and arterial restenosis therapy after PTCA

    Energy Technology Data Exchange (ETDEWEB)

    Knapp, F.F. Jr.; Beets, A.L.; Mirzadeh, S.; Alexander, C.W.; Hobbs, R.L.

    1998-06-01

    The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) represents an important resource for the production of a wide variety of medical radioisotopes. In addition to serving as a key production site for californium-252 and other transuranic elements, important examples of therapeutic radioisotopes which are currently routinely produced in the HFIR for distribution include dysprosium-166 (parent of holmium-166), rhenium-186, tin-117m and tungsten-188 (parent of rhenium-188). The nine hydraulic tube (HT) positions in the central high flux region permit the insertion and removal of targets at any time during the operating cycle and have traditionally represented a major site for production of medical radioisotopes. To increase the irradiation capabilities of the HFIR, special target holders have recently been designed and fabricated which will be installed in the six Peripheral Target Positions (PTP), which are also located in the high flux region. These positions are only accessible during reactor refueling and will be used for long-term irradiations, such as required for the production of tin-117m and tungsten-188. Each of the PTP tubes will be capable of housing a maximum of eight HT targets, thus increasing the total maximum number of HT targets from the current nine, to a total of 57. In this paper the therapeutic use of reactor-produced radioisotopes for bone pain palliation and vascular brachytherapy and the therapeutic medical radioisotope production capabilities of the ORNL HFIR are briefly discussed.

  16. Reactor antineutrino fluxes - status and challenges

    CERN Document Server

    Huber, Patrick

    2016-01-01

    In this contribution we describe the current understanding of reactor antineutrino fluxes and point out some recent developments. This is not intended to be a complete review of this vast topic but merely a selection of observations and remarks, which despite their incompleteness, will highlight the status and the challenges of this field.

  17. Reactor antineutrino fluxes – Status and challenges

    Energy Technology Data Exchange (ETDEWEB)

    Huber, Patrick, E-mail: pahuber@vt.edu

    2016-07-15

    In this contribution we describe the current understanding of reactor antineutrino fluxes and point out some recent developments. This is not intended to be a complete review of this vast topic but merely a selection of observations and remarks, which despite their incompleteness, will highlight the status and the challenges of this field.

  18. Neutron flux spectra and radiation damage parameters for the Russian Bor-60 and SM-2 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karasiov, A.V. [D.V. Efremov Scientific Rresearch Institute of Electrophysical Apparatus, St. Petersburg (Russian Federation); Greenwood, L.R. [Pacific Northwest Laboratory, Richland, WA (United States)

    1995-04-01

    The objective is to compare neutron irradiation conditions in Russian reactors and similar US facilities. Neutron fluence and spectral information and calculated radiation damage parameters are presented for the BOR-60 (Fast Experimental Reactor - 60 MW) and SM-2 reactors in Russia. Their neutron exposure characteristics are comparable with those of the Experimental Breeder Reactor (ERB-II), the Fast Flux Test Facility (FFTF), and the High Flux Isotope Reactor (HFIR) in the United States.

  19. Production of medical radioisotopes in the ORNL high flux isotope reactor (HFIR) for cancer treatment and arterial restenosis therapy after PICA

    Science.gov (United States)

    Knapp, F. F.; Beets, A. L.; Mirzadeh, S.; Alexander, C. W.; Hobbs, R. L.

    1999-01-01

    The High Flux Isotope Reactor ( HFIR) at the Oak Ridge National Laboratory ( ORNL) represents an important resource for the production of a wide variety of medical radioisotopes. First beginning operation in 1965, the high thermal neutron flux (2.5×1015 neutrons/cm2/sec at 85 MW) and versatile target irradiation and handling facilities provide the opportunity for production of a wide variety of neutron-rich medical radioisotopes of current interest for therapy. In addition to serving as a key production site for californium-252 and other transuranic elements, important examples of therapeutic radioisotopes which are currently routinely produced in the HFIR for distribution include dysprosium-166 (parent of holmium-166), rhenium-186, tin-117 m and tungsten-188 (parent of rhenium-188). The nine hydraulic tube ( HT) positions in the central high flux region permit the insertion and removal of targets at any time during the operating cycle (22-24 days) and have traditionally represented a major site for production of medical radioisotopes. To increase the irradiation capabilities of the HFIR, special target holders have recently been designed and fabricated which will be installed in the six Peripheral Target Positions ( PTP), which are also located in the high flux region. These positions are only accessible during reactor refueling and will be used for long-term irradiations, such as required for the production of tin-117 m and tungsten-188. Each of the PTP tubes will be capable of housing a maximum of eight HT targets, thus increasing the total maximum number of HT targets from the current nine, to a total of 57. In this paper the therapeutic use of reactor-produced radioisotopes for bone pain palliation and vascular brachytherapy and the therapeutic medical radioisotope production capabilities of the ORNL HFIR are briefly discussed.

  20. Design of a high-flux epithermal neutron beam using 235U fission plates at the Brookhaven Medical Research Reactor.

    Science.gov (United States)

    Liu, H B; Brugger, R M; Rorer, D C; Tichler, P R; Hu, J P

    1994-10-01

    Beams of epithermal neutrons are being used in the development of boron neutron capture therapy for cancer. This report describes a design study in which 235U fission plates and moderators are used to produce an epithermal neutron beam with higher intensity and better quality than the beam currently in use at the Brookhaven Medical Research Reactor (BMRR). Monte Carlo calculations are used to predict the neutron and gamma fluxes and absorbed doses produced by the proposed design. Neutron flux measurements at the present epithermal treatment facility (ETF) were made to verify and compare with the computed results where feasible. The calculations indicate that an epithermal neutron beam produced by a fission-plate converter could have an epithermal neutron intensity of 1.2 x 10(10) n/cm2.s and a fast neutron dose per epithermal neutron of 2.8 x 10(-11) cGy.cm2/nepi plus being forward directed. This beam would be built into the beam shutter of the ETF at the BMRR. The feasibility of remodeling the facility is discussed.

  1. Experimental Plan and Irradiation Target Design for FeCrAl Embrittlement Screening Tests Conducted Using the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-06-26

    The objective of the FeCrAl embrittlement screening tests being conducted through the use of Oak Ridge National Laboratories (ORNL) High Flux Isotope Reactor is to provide data on the radiation-induced changes in the mechanical properties including radiation-induced hardening and embrittlement through systematic testing and analysis. Data developed on the mechanical properties will be supported by extensive microstructural evaluations to assist in the development of structure-property relationships and provide a sound, fundamental understanding of the performance of FeCrAl alloys in intense neutron radiation fields. Data and analysis developed as part of this effort will be used to assist in the determination of FeCrAl alloys as a viable material for commercial light water reactor (LWR) applications with a primary focus as an accident tolerant cladding.

  2. Current and future capabilities of the neutron reflectometer MIRROR at Oak Ridge National Laboratory's High Flux Isotope Reactor

    Science.gov (United States)

    Hamilton, W. A.; Smith, G. S.; Taylor, G. B.; Larkins, B. M.; Porcar, L.

    2006-11-01

    The peripatetic ORNL HFIR Center for Neutron Scattering reflectometer instrument MIRROR has recently been re-installed in an interim beam line position in the reactor beam room. In 2006 an upgraded version of the instrument will move to a high intensity guide hall position fed by the new HFIR cold source. In this short note, we present some aspects of current instrument operation-particularly with respect to data reduction from the instrument's linear reflection plane detector-with examples of ongoing research and analysis, and a brief outline of the expected capabilities of the fully upgraded guide hall instrument.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  4. Optimization of a partially non-magnetic primary radiation shielding for the triple-axis spectrometer PANDA at the Munich high-flux reactor FRM-II

    CERN Document Server

    Pyka, N M; Rogov, A

    2002-01-01

    Monte Carlo simulations have been used to optimize the monochromator shielding of the polarized cold-neutron triple-axis spectrometer PANDA at the Munich high-flux reactor FRM-II. By using the Monte Carlo program MCNP-4B, the density of the total spectrum of incoming neutrons and gamma radiation from the beam tube SR-2 has been determined during the three-dimensional diffusion process in different types of heavy concrete and other absorbing material. Special attention has been paid to build a compact and highly efficient shielding, partially non-magnetic, with a total biological radiation dose of less than 10 mu Sv/h at its outsides. Especially considered was the construction of an albedo reducer, which serves to reduce the background in the experiment outside the shielding. (orig.)

  5. Generation of annular, high-charge electron beams at the Argonne wakefield accelerator

    Science.gov (United States)

    Wisniewski, E. E.; Li, C.; Gai, W.; Power, J.

    2013-01-01

    We present and discuss the results from the experimental generation of high-charge annular(ring-shaped)electron beams at the Argonne Wakefield Accelerator (AWA). These beams were produced by using laser masks to project annular laser profiles of various inner and outer diameters onto the photocathode of an RF gun. The ring beam is accelerated to 15 MeV, then it is imaged by means of solenoid lenses. Transverse profiles are compared for different solenoid settings. Discussion includes a comparison with Parmela simulations, some applications of high-charge ring beams,and an outline of a planned extension of this study.

  6. Fuel slugs considered for use in the high flux reactor EL3; Elements combustibles envisages pour la pile a haut flux EL 3

    Energy Technology Data Exchange (ETDEWEB)

    Stohr, J.A.; Caillat, R.; Gauthron, M.; Montagne, R. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    EL3 was designed essentially for the study, under irradiation conditions, of materials used in the construction of atomic reactors. The study schedule allocates considerable time and effort to new types of fuel slugs. The present report described the various types of slug being tested or scheduled for tests. After laboratory study, each slug is tested in an experimental cell in the pile. The best are retained and used to charge the reactor (the present charge is purely provisional to permit first criticality and power rise tests)ren. [French] La pile EL3 est essentiellement destinee a l'etude sous irradiation des materiaux utilises dans la construction des reacteurs atomiques. Dans ce programme, une tres large part est reservee a l'etude de nouveaux elements combustibles. Le present rapport decrit les differentes solutions de cartouches dont l'essai est envisage ou en cours. Apres etude en laboratoire, chacune de ces solutions est testee dans une cellule experimentale en pile. Les meilleures seront retenues pour constituer le chargement normal de la pile (le chargement actuel etant essentiellement une solution provisoire qui a permis la divergence de la pile et les premiers essais de montee en puissance). (auteur)

  7. Research and Development of Multiphysics Models in Support of the Conversion of the High Flux Isotope Reactor to Low Enriched Uranium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bodey, Isaac T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Curtis, Franklin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Arimilli, Rao V. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ekici, Kivanc [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Freels, James D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-11-01

    The findings presented in this report are results of a five year effort led by the RRD Division of the ORNL, which is focused on research and development toward the conversion of the High Flux Isotope Reactor (HFIR) fuel from high-enriched uranium (HEU) to low-enriched uranium (LEU). This report focuses on the tasks accomplished by the University of Tennessee Knoxville (UTK) team from the Department of Mechanical, Aerospace, and Biomedical Engineering (MABE) that provided expert support in multiphysics modeling of complex problems associated with the LEU conversion of the HFIR reactor. The COMSOL software was used as the main computational modeling tool, whereas Solidworks was also used in support of computer-aided-design (CAD) modeling of the proposed LEU fuel design. The UTK research has been governed by a statement of work (SOW), which was updated annually to clearly define the specific tasks reported herein. Ph.D. student Isaac T. Bodey has focused on heat transfer and fluid flow modeling issues and has been aided by his major professor Dr. Rao V. Arimilli. Ph.D. student Franklin G. Curtis has been focusing on modeling the fluid-structure interaction (FSI) phenomena caused by the mechanical forces acting on the fuel plates, which in turn affect the fluid flow in between the fuel plates, and ultimately the heat transfer, is also affected by the FSI changes. Franklin Curtis has been aided by his major professor Dr. Kivanc Ekici. M.Sc. student Adam R. Travis has focused two major areas of research: (1) on accurate CAD modeling of the proposed LEU plate design, and (2) reduction of the model complexity and dimensionality through interdimensional coupling of the fluid flow and heat transfer for the HFIR plate geometry. Adam Travis is also aided by his major professor, Dr. Kivanc Ekici. We must note that the UTK team, and particularly the graduate students, have been in very close collaboration with Dr. James D. Freels (ORNL technical monitor and mentor) and have

  8. Research and Development of Multiphysics Models in Support of the Conversion of the High Flux Isotope Reactor to Low Enriched Uranium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bodey, Isaac T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Curtis, Franklin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Arimilli, Rao V. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ekici, Kivanc [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Freels, James D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-11-01

    ABSTRACT The findings presented in this report are results of a five year effort lead by the RRD Division of the ORNL, which is focused on research and development toward the conversion of the High Flux Isotope Reactor (HFIR) fuel from high-enriched uranium (HEU) to low-enriched uranium (LEU). This report focuses on the tasks accomplished by the University of Tennessee Knoxville (UTK) team from the Department of Mechanical, Aerospace, and Biomedical Engineering (MABE) that provided expert support in multiphysics modeling of complex problems associated with the LEU conversion of the HFIR reactor. The COMSOL software was used as the main computational modeling tool, whereas Solidworks was also used in support of computer-aided-design (CAD) modeling of the proposed LEU fuel design. The UTK research has been governed by a statement of work (SOW), which was updated annually to clearly define the specific tasks reported herein. Ph.D. student Isaac T. Bodey has focused on heat transfer and fluid flow modeling issues and has been aided by his major professor Dr. Rao V. Arimilli. Ph.D. student Franklin G. Curtis has been focusing on modeling the fluid-structure interaction (FSI) phenomena caused by the mechanical forces acting on the fuel plates, which in turn affect the fluid flow in between the fuel plates, and ultimately the heat transfer, is also affected by the FSI changes. Franklin Curtis has been aided by his major professor Dr. Kivanc Ekici. M.Sc. student Adam R. Travis has focused two major areas of research: (1) on accurate CAD modeling of the proposed LEU plate design, and (2) reduction of the model complexity and dimensionality through interdimensional coupling of the fluid flow and heat transfer for the HFIR plate geometry. Adam Travis is also aided by his major professor, Dr. Kivanc Ekici. We must note that the UTK team, and particularly the graduate students, have been in very close collaboration with Dr. James D. Freels (ORNL technical monitor and mentor

  9. Reactor Neutrino Flux Uncertainty Suppression on Multiple Detector Experiments

    CERN Document Server

    Cucoanes, Andi; Cabrera, Anatael; Fallot, Muriel; Onillon, Anthony; Obolensky, Michel; Yermia, Frederic

    2015-01-01

    This publication provides a coherent treatment for the reactor neutrino flux uncertainties suppression, specially focussed on the latest $\\theta_{13}$ measurement. The treatment starts with single detector in single reactor site, most relevant for all reactor experiments beyond $\\theta_{13}$. We demonstrate there is no trivial error cancellation, thus the flux systematic error can remain dominant even after the adoption of multi-detector configurations. However, three mechanisms for flux error suppression have been identified and calculated in the context of Double Chooz, Daya Bay and RENO sites. Our analysis computes the error {\\it suppression fraction} using simplified scenarios to maximise relative comparison among experiments. We have validated the only mechanism exploited so far by experiments to improve the precision of the published $\\theta_{13}$. The other two newly identified mechanisms could lead to total error flux cancellation under specific conditions and are expected to have major implications o...

  10. Neutron flux optimization in irradiation channels at NUR research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Meftah, B. [Division Reacteur, Centre de Recherche Nucleaire Draria (CRND), BP 43 Sebala DRARIA, Alger (Algeria)]. E-mail: b_meftah@yahoo.com; Zidi, T. [Division Reacteur, Centre de Recherche Nucleaire Draria (CRND), BP 43 Sebala DRARIA, Alger (Algeria); Bousbia-Salah, A. [Dipartimento di Ingegneria Meccanica, Nucleari e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2 - 56126 Pisa (Italy)

    2006-09-15

    Optimization of neutron fluxes in experimental channels is of great concern in research reactor utilization. The general approach used at the NUR research reactor for neutron flux optimization in irradiation channels is presented. The approach is essentially based upon a judicious optimization of the core configuration combined with the improvement of reflector characteristics. The method allowed to increase the thermal neutron flux for radioisotope production purposes by more than 800%. Increases of up to 60% are also observed in levels of useful fluxes available for neutron diffraction experiments (small angle neutron scattering (SANS), neutron reflectometry, etc.). Such improvements in the neutronic characteristics of the NUR reactor opened new perspectives in terms of its utilization. More particularly, it is now possible to produce at industrial scales major radio-isotopes for medicine and industry and to perform, for the first time, material testing experiments. The cost of the irradiations in the optimized configuration is generally small when compared to those performed in the old configuration and an average reduction factor of about of 10 is expected in the case of production of Molybdenum-99 (isotope required for the manufacturing of Technetium-99 medical kits). In addition to these important results, safety analysis studies showed that the more symmetrical nature of the core geometry leads to a more adequately balanced reactivity control system and contributes quite efficiently to the operational safety of the NUR reactor. Results of comparisons between calculations and measurements for a series of parameters of importance in reactor operation and safety showed good agreement.

  11. Thermal neutron flux distribution in ET-RR-2 reactor thermal column

    Directory of Open Access Journals (Sweden)

    Imam Mahmoud M.

    2002-01-01

    Full Text Available The thermal column in the ET-RR-2 reactor is intended to promote a thermal neutron field of high intensity and purity to be used for following tasks: (a to provide a thermal neutron flux in the neutron transmutation silicon doping, (b to provide a thermal flux in the neutron activation analysis position, and (c to provide a thermal neutron flux of high intensity to the head of one of the beam tubes leading to the room specified for boron thermal neutron capture therapy. It was, therefore, necessary to determine the thermal neutron flux at above mentioned positions. In the present work, the neutron flux in the ET-RR-2 reactor system was calculated by applying the three dimensional diffusion depletion code TRITON. According to these calculations, the reactor system is composed of the core, surrounding external irradiation grid, beryllium block, thermal column and the water reflector in the reactor tank next to the tank wall. As a result of these calculations, the thermal neutron fluxes within the thermal column and at irradiation positions within the thermal column were obtained. Apart from this, the burn up results for the start up core calculated according to the TRITION code were compared with those given by the reactor designer.

  12. OPAL REACTOR: Calculation/Experiment comparison of Neutron Flux Mapping in Flux Coolant Channels

    Energy Technology Data Exchange (ETDEWEB)

    Barbot, L.; Domergue, C.; Villard, J. F.; Destouches, C. [CEA, Paris (France); Braoudakis, G.; Wassink, D.; Sinclair, B.; Osborn, J. C.; Huayou, Wu [ANSTO, Syeney (Australia)

    2013-07-01

    The measurement and calculation of the neutron flux mapping of the OPAL research reactor are presented. Following an investigation of fuel coolant channels using sub-miniature fission chambers to measure thermal neutron flux profiles, neutronic calculations were performed. Comparison between calculation and measurement shows very good agreement.

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  14. Monitoring Akkuyu Nuclear Reactor Using Anti-Neutrino Flux Measurement

    CERN Document Server

    Ozturk, Sertac; Ozcan, V Erkcan; Unel, Gokhan

    2016-01-01

    We present a simulation based study for monitoring Akkuyu Nuclear Power Plant's activity using anti-neutrino flux originating from the reactor core. A water Cherenkov detector has been designed and optimization studies have been performed using Geant4 simulation toolkit. A first study for the design of a monitoring detector facility for Akkuyu Nuclear Power Plant has been discussed in this paper.

  15. Which reactor antineutrino flux may be responsible for the anomaly?

    CERN Document Server

    Giunti, Carlo

    2016-01-01

    We investigate which among the reactor antineutrino fluxes from the decays of the fission products of $^{235}\\text{U}$, $^{238}\\text{U}$, $^{239}\\text{Pu}$, and $^{241}\\text{Pu}$ may be responsible for the reactor antineutrino anomaly. We find that it is the $^{235}\\text{U}$ flux, which contributes to the rates of all reactor neutrino experiments. From the fit of the data we obtain the precise determination $ \\sigma_{^{235}\\text{U}} = ( 6.34 \\pm 0.10 ) \\times 10^{-43} \\, \\text{cm}^2 / \\text{fission} $ of the $^{235}\\text{U}$ cross section per fission, which is more precise than the calculated value and differs from it by $2.0\\sigma$.

  16. Neutron flux reduction programs for reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, C.S. [Korea Atomic Energy Research Inst. KAERI, 150 Deogjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Kim, B.C. [Korea Reactor Integrity Surveillance Technology KRIST, 150 Deogjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2011-07-01

    The objective of this work is to implement various fast neutron flux reduction programs on the belt-line region of the reactor pressure vessel to reduce the increasing rate of reference temperature for pressurized thermal shock (RT PTS) for Korea Nuclear Unit 1. A pressurized thermal shock (PTS) event is an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. A PTS concern arises if one of these transients acts in the belt-line region of a reactor vessel where a reduced fracture resistance exists because of neutron irradiation. Generally, the RT PTS value is continuously increasing according to the fast neutron irradiation during the reactor operation, and it can reach the screening criterion prior to the expiration of the operating license. To reduce the increasing rate of RT PTS, various neutron flux reduction programs can be implemented, which are focused on license renewal. In this paper, neutron flux reduction programs, such as low leakage loading pattern strategy, loading of neutron absorber rods, and dummy fuel assembly loading are considered for Korea Nuclear Unit 1, of which the RT PTS value of the leading material (circumferential weld) is going to reach the screening criterion in the near future. To evaluate the effects of the neutron flux reduction programs, plant and cycle specific forward neutron transport calculations for the various neutron flux reduction programs were carried out. For the analysis, all transport calculations were carried out by using the DORT 3.1 discrete ordinate code and BUGLE-96 cross-section library. (authors)

  17. Reactor Antineutrino Flux and Spectrum Shape from Daya Bay

    Science.gov (United States)

    Napolitano, Jim; Daya Bay Collaboration

    2017-01-01

    The Daya Bay Reactor Neutrino Experiment has collected very large samples of νe p ->e+ n events, where the νe are from the cores of six power plant reactors that undergo regular refueling. With 621 days of data, more than 1.2 million events of this type were detected. The collaboration has analyzed these data in terms of the absolute flux (addressing the ``Reactor Neutrino Anomaly''), the spectrum shape (including the excess in the region of 5 MeV prompt energy), and other effects. This talk will summarize the results from our most recent analyses, and discuss new initiatives aimed at continuing to understand the fine detail of the reactor νe spectrum.

  18. Bayesian calibration of reactor neutron flux spectrum using activation detectors measurements: Application to CALIBAN reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cartier, J. [Commissariat a l' Energie Atomique et aux Energies Alternatives CEA, DAM, DIF, F-91297 Arpajon (France); Casoli, P. [Commissariat a l' Energie Atomique et aux Energies Alternatives CEA, DAM, Valduc, F-21120 Is sur Tille (France); Chappert, F. [Commissariat a l' Energie Atomique et aux Energies Alternatives CEA, DAM, DIF, F-91297 Arpajon (France)

    2013-07-01

    In this paper, we present calibration methods in order to estimate reactor neutron flux spectrum and its uncertainties by using integral activation measurements. These techniques are performed using Bayesian and MCMC framework. These methods are applied to integral activation experiments in the cavity of the CALIBAN reactor. We estimate the neutron flux and its related uncertainties. The originality of this work is that these uncertainties take into account measurements uncertainties, cross-sections uncertainties and model error. In particular, our results give a very good approximation of the total flux and indicate that neutron flux from MCNP simulation for energies above about 5 MeV seems to overestimate the 'real flux'. (authors)

  19. Reactors

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1988-01-01

    This standard applies to the following types of reactors: shunt reactors, current-limiting reactors including neutral-earthing reactors, damping reactors, tuning (filter) reactors, earthing transformers (neutral couplers), arc-suppression reactors, smoothing reactors, with the exception of the following reactors: small reactors with a rating generally less than 2 kvar single-phase and 10 kvar three-phase, reactors for special purposes such as high-frequency line traps or reactors mounted on rolling stock.

  20. Thermal neutron flux distribution in ET-RR-2 reactor thermal column

    OpenAIRE

    Imam Mahmoud M.; Roushdy Hassan

    2002-01-01

    The thermal column in the ET-RR-2 reactor is intended to promote a thermal neutron field of high intensity and purity to be used for following tasks: (a) to provide a thermal neutron flux in the neutron transmutation silicon doping, (b) to provide a thermal flux in the neutron activation analysis position, and (c) to provide a thermal neutron flux of high intensity to the head of one of the beam tubes leading to the room specified for boron thermal neutron capture therapy. It was, therefore, ...

  1. γ-ray fluxes in Oklo natural reactors

    Science.gov (United States)

    Gould, C. R.; Sharapov, E. I.; Sonzogni, A. A.

    2012-11-01

    Background: Uncertainty in the operating temperatures of Oklo reactor zones impacts the precision of bounds derived for time variation of the fine structure constant α. Improved 176Lu/175Lu thermometry has been discussed but its usefulness may be complicated by photoexcitation of the isomeric state 176mLu by 176Lu(γ,γ') fluorescence.Purpose: We calculate prompt, delayed, and equilibrium γ-ray fluxes due to fission of 235U in pulsed mode operation of Oklo zone RZ10.Methods: We use Monte Carlo modeling to calculate the prompt flux. We use improved data libraries to estimate delayed and equilibrium spectra and fluxes.Results: We find γ-ray fluxes as a function of energy and derive values for the coefficients λγ,γ' that describe burn-up of 176Lu through the isomeric 176mLu state.Conclusion: The contribution of the (γ,γ') channel to the 176Lu/175Lu isotopic ratio is negligible in comparison to the neutron burn-up channels. Lutetium thermometry is fully applicable to analyses of Oklo reactor data.

  2. High Flux Calorimetry.

    Science.gov (United States)

    1984-05-05

    These approaches are based on proven principles which have served the thermal test community well for years. Other concepts hold promise of being able to...8217. --......- - ... .... - - The thermal test community has developed instrumentation which is quite suitable for the moderate, and relatively constant, flux...on the maximum phase II system fluence of 400 cal/cm2 . Second, the present thermal test community will have confidence in the performance of an

  3. Calculation with MCNP of capture photon flux in VVER-1000 experimental reactor.

    Science.gov (United States)

    Töre, Candan; Ortego, Pedro

    2005-01-01

    The aim of this study is to obtain by Monte Carlo method the high energy photon flux due to neutron capture in the internals and vessel layers of the experimental reactor LR-0 located in REZ, Czech Republic, and loaded with VVER-1000 fuel. The calclated neutron, photon and photon to neutron flux ratio are compared with experimental measurements performed with a multi-parameter stilbene detector. The results show clear underestimation of photon flux in downcomer and some overestimation at vessel surface and 1/4 thickness but a good fitting for deeper points in vessel.

  4. On flux effects in a low alloy steel from a Swedish reactor pressure vessel

    Science.gov (United States)

    Boåsen, Magnus; Efsing, Pål; Ehrnstén, Ulla

    2017-02-01

    This study aims to investigate the presence of Unstable Matrix Defects in irradiated pressure vessel steel from weldments of the Swedish PWR Ringhals 4 (R4). Hardness tests have been performed on low flux (surveillance material) and high flux (Halden reactor) irradiated material samples in combination with heat treatments at temperatures of 330, 360 and 390 °C in order to reveal eventual recovery of any hardening features induced by irradiation. The experiments carried out in this study could not reveal any hardness recovery related to Unstable Matrix Defects at relevant temperatures. However, a difference in hardness recovery was found between the low and the high flux samples at heat treatments at higher temperatures than expected for the annihilation of Unstable Matrix Defects-the observed recovery is here attributed to differences of the solute clusters formed by the high and low flux irradiations.

  5. Department of Energy's High Flux Beam Reactor (HFBR), September 15--19, 1980: An independent on-site safety review

    Energy Technology Data Exchange (ETDEWEB)

    1981-02-01

    The intent of this on-site safety review was to make a broad management assessment of HFBR operations, rather than conduct a detailed in-depth audit. The result of the review should only be considered as having identified trends or indications. The Team's observations and recommendations for the most part are based upon licensed reactor facility practices used to meet industry standards. These standards form the basis for many of the comments in this report. The Team believes that a uniform minimum standard of performance should be achieved in the operation of DOE reactors. In order to assure that this is accomplished, clear standards are necessary. Consistent with the past AEC and ERDA policy, the team has used the standards of the commercial nuclear power industry. It is recognized that this approach is conservative in that the HFBR reactor has a significantly greater degree of inherent safety (low pressure, temperature, power, etc.) than a licensed reactor.

  6. Gamma-ray fluxes in Oklo natural reactors

    CERN Document Server

    Gould, C R; Sonzogni, A A; 10.1103/PhysRevC.86.054602

    2012-01-01

    Uncertainty in the operating temperatures of Oklo reactor zones impacts the precision of bounds derived for time variation of the fine structure constant $\\alpha$. Improved $^{176}$Lu/$^{175}$Lu thermometry has been discussed but its usefulness may be complicated by photo excitation of the isomeric state $^{176m}$Lu by $^{176}$Lu($\\gamma,\\gamma^\\prime $) fluorescence. We calculate prompt, delayed and equilibrium $\\gamma$-ray fluxes due to fission of $^{235}$U in pulsed mode operation of Oklo zone RZ10. We use Monte Carlo modeling to calculate the prompt flux. We use improved data libraries to estimate delayed and equilibrium spectra and fluxes. We find $\\gamma$-ray fluxes as a function of energy and derive values for the coefficients $\\lambda_{\\gamma,\\gamma^\\prime}$ that describe burn-up of $^{176}$Lu through the isomeric $^{176m}$Lu state. The contribution of the ($\\gamma,\\gamma^\\prime $) channel to the $^{176}$Lu/$^{175}$Lu isotopic ratio is negligible in comparison to the neutron burn-up channels. Lutetium...

  7. Delayed Gamma Measurements in Different Nuclear Research Reactors Bringing Out the Importance of the Delayed Contribution in Gamma Flux Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Fourmentel, D.; Radulovic, V.; Barbot, L.; Villard, J-F. [Alternative Energies and Atomic Energy Commission, CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, 13108 Saint- Paul-Lez-Durance (France); Zerovnik, G.; Snoj, L. [Reactor Physics Department, Jozef Stefan Institute, SI-1000 Ljubljana (Slovenia); Tarchalski, M.; Pytel, K. [National Centre for Nuclear Research A. Soltana 7, 05-400 Swierk (Poland); Malouch, F. [Alternative Energies and Atomic Energy Commission - CEA, DEN, DM2S, Saclay, 91191, Gif-sur-Yvette (France)

    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

  8. 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

    Energy Technology Data Exchange (ETDEWEB)

    Stillman, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Foyto, L [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Kutikkad, K [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; McKibben, J C [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Peters, N. [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Stevens, J. [Argonne National Lab. (ANL), Argonne, IL (United States)

    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.

  9. A Level 1+ Probabilistic Safety Assessment of the high flux Australian reactor. Vol. 2. Appendix C: System analysis models and results

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-01-01

    This section contains the results of the quantitative system/top event analysis. Section C. 1 gives the basic event coding scheme. Section C.2 shows the master frequency file (MFF), which contains the split fraction names, the top events they belong to, the mean values of the uncertainty distribution that is generated by the Monte Carlo quantification in the System Analysis module of RISKMAN, and a brief description of each split fraction. The MFF is organized by the systems modeled, and within each system, the top events associated with the system. Section C.3 contains the fault trees developed for the system/top event models and the RISKMAN reports for each of the system/top event models. The reports are organized under the following system headings: Compressed/Service Air Supply (AIR); Containment Isolation System (CIS); Heavy Water Cooling System (D20); Emergency Core Cooling System (ECCS); Electric Power System (EPS); Light Water Cooling system (H20); Helium Gas System (HE); Mains Water System (MW); Miscellaneous Top Events (MISC); Operator Actions (OPER) Reactor Protection System (RPS); Space Conditioner System (SCS); Condition/Status Switch (SWITCH); RCB Ventilation System (VENT); No. 1 Storage Block Cooling System (SB)

  10. Steel slag carbonation in a flow-through reactor system: the role of fluid-flux.

    Science.gov (United States)

    Berryman, Eleanor J; Williams-Jones, Anthony E; Migdisov, Artashes A

    2015-01-01

    Steel production is currently the largest industrial source of atmospheric CO2. As annual steel production continues to grow, the need for effective methods of reducing its carbon footprint increases correspondingly. The carbonation of the calcium-bearing phases in steel slag generated during basic oxygen furnace (BOF) steel production, in particular its major constituent, larnite {Ca2SiO4}, which is a structural analogue of olivine {(MgFe)2SiO4}, the main mineral subjected to natural carbonation in peridotites, offers the potential to offset some of these emissions. However, the controls on the nature and efficiency of steel slag carbonation are yet to be completely understood. Experiments were conducted exposing steel slag grains to a CO2-H2O mixture in both batch and flow-through reactors to investigate the impact of temperature, fluid flux, and reaction gradient on the dissolution and carbonation of steel slag. The results of these experiments show that dissolution and carbonation of BOF steel slag are more efficient in a flow-through reactor than in the batch reactors used in most previous studies. Moreover, they show that fluid flux needs to be optimized in addition to grain size, pressure, and temperature, in order to maximize the efficiency of carbonation. Based on these results, a two-stage reactor consisting of a high and a low fluid-flux chamber is proposed for CO2 sequestration by steel slag carbonation, allowing dissolution of the slag and precipitation of calcium carbonate to occur within a single flow-through system.

  11. Discussion about modeling the effects of neutron flux exposure for nuclear reactor core analysis

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1986-04-01

    Methods used to calculate the effects of exposure to a neutron flux are described. The modeling of the nuclear-reactor core history presents an analysis challenge. The nuclide chain equations must be solved, and some of the methods in use for this are described. Techniques for treating reactor-core histories are discussed and evaluated.

  12. Do you want to build such a machine? : Designing a high energy proton accelerator for Argonne National Laboratory.

    Energy Technology Data Exchange (ETDEWEB)

    Paris, E.

    2004-04-05

    Argonne National Laboratory's efforts toward researching, proposing and then building a high-energy proton accelerator have been discussed in a handful of studies. In the main, these have concentrated on the intense maneuvering amongst politicians, universities, government agencies, outside corporations, and laboratory officials to obtain (or block) approval and/or funds or to establish who would have control over budgets and research programs. These ''top-down'' studies are very important but they can also serve to divorce such proceedings from the individuals actually involved in the ground-level research which physically served to create theories, designs, machines, and experiments. This can lead to a skewed picture, on the one hand, of a lack of effect that so-called scientific and technological factors exert and, on the other hand, of the apparent separation of the so-called social or political from the concrete practice of doing physics. An exception to this approach can be found in the proceedings of a conference on ''History of the ZGS'' held at Argonne at the time of the Zero Gradient Synchrotron's decommissioning in 1979. These accounts insert the individuals quite literally as they are, for the most part, personal reminiscences of those who took part in these efforts on the ground level. As such, they are invaluable raw material for historical inquiry but generally lack the rigor and perspective expected in a finished historical work. The session on ''Constructing Cold War Physics'' at the 2002 annual History of Science Society Meeting served to highlight new approaches circulating towards history of science and technology in the post-WWII period, especially in the 1950s. There is new attention towards the effects of training large numbers of scientists and engineers as well as the caution not to equate ''national security'' with military preparedness, but rather

  13. Determination flux in the Reactor JEN-1; Medida de flujos de neutrones en el nucleo del Reactor JEN-1

    Energy Technology Data Exchange (ETDEWEB)

    Manas Diaz, L.; Montes Ponce de leon, J.

    1960-07-01

    This report summarized several irradiations that have been made to determine the neutron flux distributions in the core of the JEN-1 reactor. Gold foils of 380 {mu} gr and Mn-Ni (12% de Ni) of 30 mg have been employed. the epithermal flux has been determined by mean of the Cd radio. The resonance integral values given by Macklin and Pomerance have been used. (Author) 9 refs.

  14. Background radiation measurements at high power research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ashenfelter, J. [Wright Laboratory, Department of Physics, Yale University, New Haven, CT 06520 (United States); Balantekin, B. [Department of Physics, University of Wisconsin, Madison, WI 53706 (United States); Baldenegro, C.X. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Band, H.R. [Wright Laboratory, Department of Physics, Yale University, New Haven, CT 06520 (United States); Barclay, G. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Bass, C.D. [Department of Chemistry and Physics, Le Moyne College, Syracuse, NY 13214 (United States); Berish, D. [Department of Physics, Temple University, Philadelphia, PA 19122 (United States); Bowden, N.S., E-mail: nbowden@llnl.gov [Nuclear and Chemical Sciences Division, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Bryan, C.D. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Cherwinka, J.J. [Physical Sciences Laboratory, University of Wisconsin, Madison, WI 53706 (United States); Chu, R. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Department of Physics, University of Tennessee, Knoxville, TN 37996 (United States); Classen, T. [Nuclear and Chemical Sciences Division, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Davee, D. [Department of Physics, College of William and Mary, Williamsburg, VA 23187 (United States); Dean, D.; Deichert, G. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Dolinski, M.J. [Department of Physics, Drexel University, Philadelphia, PA 19104 (United States); Dolph, J. [Physics Department, Brookhaven National Laboratory, Upton, NY 11973 (United States); Dwyer, D.A. [Physics Division, Lawrence Berkeley National Laboratory, Berkeley, CA 94720 (United States); Fan, S. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Department of Physics, University of Tennessee, Knoxville, TN 37996 (United States); and others

    2016-01-11

    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.

  15. Background Radiation Measurements at High Power Research Reactors

    CERN Document Server

    Ashenfelter, J; 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 $\\gamma$-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.

  16. Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    McCulloch, R.W.; MacPherson, R.E.

    1983-03-01

    The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm/sup 2/, 1000/sup 0/C cladding temperature, and (2) 40 h at 40 W/cm/sup 2/, 1200/sup 0/C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 1370/sup 0/C.

  17. Simulation of the SONGS Reactor Antineutrino Flux Using DRAGON

    CERN Document Server

    Jones, C L

    2011-01-01

    For reactor antineutrino experiments, a thorough understanding of the fuel composition and isotopic evolution is of paramount importance for the extraction of $\\theta_{13}$. To accomplish these goals, we employ the deterministic lattice code DRAGON, and analyze the instantaneous antineutrino rate from the San Onofre Nuclear Generating Station (SONGS) Unit 2 reactor in California. DRAGON's ability to predict the rate for two consecutive fuel cycles is examined.

  18. High density operation for reactor-relevant power exhaust

    Science.gov (United States)

    Wischmeier, M.

    2015-08-01

    With increasing size of a tokamak device and associated fusion power gain an increasing power flux density towards the divertor needs to be handled. A solution for handling this power flux is crucial for a safe and economic operation. Using purely geometric arguments in an ITER-like divertor this power flux can be reduced by approximately a factor 100. Based on a conservative extrapolation of current technology for an integrated engineering approach to remove power deposited on plasma facing components a further reduction of the power flux density via volumetric processes in the plasma by up to a factor of 50 is required. Our current ability to interpret existing power exhaust scenarios using numerical transport codes is analyzed and an operational scenario as a potential solution for ITER like divertors under high density and highly radiating reactor-relevant conditions is presented. Alternative concepts for risk mitigation as well as strategies for moving forward are outlined.

  19. Fission reactor flux monitors based on single-crystal CVD diamond films

    Energy Technology Data Exchange (ETDEWEB)

    Almaviva, S.; Marinelli, M.; Prestopino, G.; Tucciarone, A.; Verona, C.; Verona-Rinati, G. [Dipartimento di Ingegneria Meccanica, Universita di Roma ' ' Tor Vergata' ' , Via del Politecnico 1, 00133 Roma (Italy); INFN - Sezione Roma ' ' Tor Vergata' ' (Italy); Milani, E. [INFN - Sezione Roma ' ' Tor Vergata' ' (Italy); Angelone, M.; Lattanzi, D.; Pillon, M. [Associazione EURATOM-ENEA sulla Fusione, Via E. Fermi 45, 00144 Frascati (Roma) (Italy); Rosa, R. [Dipartimento Fusione e Presidio Nucleare ENEA C.R. Casaccia, Via Anguillarese 301, 00123 Roma (Italy)

    2007-09-15

    Diamond based thermal neutron flux monitors have been fabricated using single crystal diamond films, grown by chemical vapour deposition. A 3 {mu}m thick {sup 6}LiF layer was thermally evaporated on the detector surface as a converting material for thermal neutron monitoring via the {sup 6}Li(n, {alpha}) T nuclear reaction. The detectors were tested in a fission nuclear reactor. One of them was positioned 80 cm above the core mid-plane, where the neutron flux is 2.2 x 10{sup 9} neutrons/cm{sup 2}s at 1 MW resulting in a device count rate of about 150000 cps. Good stability and reproducibility of the device output were proved over the whole reactor power range (up to 1 MW). During the irradiation, several pulse height spectra were recorded, in which both products of the {sup 6}Li(n,{alpha})T reaction, e.g. 2.73 MeV tritium and the 2.06 MeV {alpha}, were clearly identified, thus excluding a degradation of the detector response. A comparison with a reference fission chamber monitor pointed out a limitation of the adopted readout electronics at high count rates, due to multiple pile-up processes. However, once this effect is properly accounted for, a good linearity of the diamond flux monitor response is observed as a function of the fission chamber one, as well as an excellent agreement between the temporal behaviour of the two detector response. (copyright 2007 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  20. Argonne National Lab gets Linux network teraflop cluster

    CERN Multimedia

    2003-01-01

    "Linux NetworX, Salt Lake City, Utah, has delivered an Evolocity II (E2) Linux cluster to Argonne National Laboratory that is capable of performing more than one trillion calculations per second (1 teraFLOP). The cluster, named "Jazz" by Argonne, is designed to provide optimum performance for multiple disciplines such as chemistry, physics and reactor engineering and will be used by the entire scientific community at the Lab" (1 page).

  1. OVERALL CONTROL SYSTEM FOR HIGH FLUX PILE

    Science.gov (United States)

    Newson, H.W.; Durham, N.C.; Wigner, E.P.; Princeton, N.J.; Epler, E.P.

    1961-05-23

    A control system is given for a high fiux reactor incorporating an anti- scram control feature whereby a neutron absorbing control rod acts as a fine adjustment while a neutron absorbing shim rod, actuated upon a command received from reactor period and level signals, has substantially greater effect on the neutron level and is moved prior to scram conditions to alter the reactor activity before a scram condition is created. Thus the probability that a scram will have to be initiated is substantially decreased.

  2. Measurement of photon flux with a miniature gas ionization chamber in a Material Testing Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Fourmentel, D., E-mail: damien.fourmentel@cea.fr [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France); Filliatre, P.; Villard, J.F.; Lyoussi, A. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France); Reynard-Carette, C. [Aix-Marseille Université, LISA EA 4672, cedex 20, Marseille 13397 (France); Carcreff, H. [CEA, DEN, DRSN, Saclay, F-91191 Gif-sur-Yvette (France)

    2013-10-01

    Nuclear heating measurements in Material Testing Reactors (MTR) are crucial for the design of the experimental devices and the prediction of the temperature of the hosted samples. Nuclear heating in MTR materials (except fuel) is mainly due to the energy deposition by the photon flux. Therefore, the photon flux is a key input parameter for the computer codes which simulate nuclear heating and temperature reached by samples/devices under irradiation. In the Jules Horowitz MTR under construction at the CEA Cadarache, the maximal expected nuclear heating levels will be about 15 to 18 W g{sup −1} and it will be necessary to assess this parameter with the best accuracy. An experiment was performed at the OSIRIS reactor to combine neutron flux, photon flux and nuclear heating measurements to improve the knowledge of the nuclear heating in MTR. There are few appropriate sensors for selective measurement of the photon flux in MTR even if studies and developments are ongoing. An experiment, called CARMEN-1, was conducted at the OSIRIS MTR and we used in particular a gas ionization chamber based on miniature fission chamber design to measure the photon flux. In this paper, we detail Monte-Carlo simulations to analyze the photon fluxes with ionization chamber measurements and we compare the photon flux calculations to the nuclear heating measurements. These results show a good accordance between photon flux measurements and nuclear heating measurement and allow improving the knowledge of these parameters.

  3. High-intensity power-resolved radiation imaging of an operational nuclear reactor

    Science.gov (United States)

    Beaumont, Jonathan S.; Mellor, Matthew P.; Villa, Mario; Joyce, Malcolm J.

    2015-01-01

    Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors. PMID:26450669

  4. Flux attenuation at NREL's High-Flux Solar Furnace

    Science.gov (United States)

    Bingham, Carl E.; Scholl, Kent L.; Lewandowski, Allan A.

    1994-10-01

    The High-Flux Solar Furnace (HFSF) at the National Renewable Energy Laboratory (NREL) has a faceted primary concentrator and a long focal-length-to-diameter ratio (due to its off-axis design). Each primary facet can be aimed individually to produce different flux distributions at the target plane. Two different types of attenuators are used depending on the flux distribution. A sliding-plate attenuator is used primarily when the facets are aimed at the same target point. The alternate attenuator resembles a venetian blind. Both attenuators are located between the concentrator and the focal point. The venetian-blind attenuator is primarily used to control the levels of sunlight failing on a target when the primary concentrators are not focused to a single point. This paper will demonstrate the problem of using the sliding-plate attenuator with a faceted concentrator when the facets are not aimed at the same target point. We will show that although the alternate attenuator necessarily blocks a certain amount of incoming sunlight, even when fully open, it provides a more even attenuation of the flux for alternate aiming strategies.

  5. Plasma–Surface Interactions Under High Heat and Particle Fluxes

    Directory of Open Access Journals (Sweden)

    Gregory De Temmerman

    2013-01-01

    Full Text Available The plasma-surface interactions expected in the divertor of a future fusion reactor are characterized by extreme heat and particle fluxes interacting with the plasma-facing surfaces. Powerful linear plasma generators are used to reproduce the expected plasma conditions and allow plasma-surface interactions studies under those very harsh conditions. While the ion energies on the divertor surfaces of a fusion device are comparable to those used in various plasma-assited deposition and etching techniques, the ion (and energy fluxes are up to four orders of magnitude higher. This large upscale in particle flux maintains the surface under highly non-equilibrium conditions and bring new effects to light, some of which will be described in this paper.

  6. Improved Measurement of the Reactor Antineutrino Flux and Spectrum at Daya Bay

    CERN Document Server

    An, F P; Band, H R; Bishai, M; Blyth, S; Cao, D; Cao, G F; Cao, J; Cen, W R; Chan, Y L; Chang, J F; Chang, L C; Chang, Y; Chen, H S; Chen, Q Y; Chen, S M; Chen, Y X; Chen, Y; Cheng, J -H; Cheng, J; Cheng, Y P; Cheng, Z K; Cherwinka, J J; Chu, M C; Chukanov, A; Cummings, J P; de Arcos, J; Deng, Z Y; Ding, X F; Ding, Y Y; Diwan, M V; Dolgareva, M; Dove, J; Dwyer, D A; Edwards, W R; Gill, R; Gonchar, M; Gong, G H; Gong, H; Grassi, M; Gu, W Q; Guan, M Y; Guo, L; Guo, R P; Guo, X H; Guo, Z; Hackenburg, R W; Han, R; Hans, S; He, M; Heeger, K M; Heng, Y K; Higuera, A; Hor, Y K; Hsiung, Y B; Hu, B Z; Hu, T; Hu, W; Huang, E C; Huang, H X; Huang, X T; Huber, P; Huo, W; Hussain, G; Jaffe, D E; Jaffke, P; Jen, K L; Jetter, S; Ji, X P; Ji, X L; Jiao, J B; Johnson, R A; Joshi, J; Kang, L; Kettell, S H; Kohn, S; Kramer, M; Kwan, K K; Kwok, M W; Kwok, T; Langford, T J; Lau, K; Lebanowski, L; Lee, J; Lee, J H C; Lei, R T; Leitner, R; Li, C; Li, D J; Li, F; Li, G S; Li, Q J; Li, S; Li, S C; Li, W D; Li, X N; Li, Y F; Li, Z B; Liang, H; Lin, C J; Lin, G L; Lin, S; Lin, S K; Lin, Y -C; Ling, J J; Link, J M; Littenberg, L; Littlejohn, B R; Liu, D W; Liu, J L; Liu, J C; Loh, C W; Lu, C; Lu, H Q; Lu, J S; Luk, K B; Lv, Z; Ma, Q M; Ma, X Y; Ma, X B; Ma, Y Q; Malyshkin, Y; Caicedo, D A Martinez; McDonald, K T; McKeown, R D; Mitchell, I; Mooney, M; Nakajima, Y; Napolitano, J; Naumov, D; Naumova, E; Ngai, H Y; Ning, Z; Ochoa-Ricoux, J P; Olshevskiy, A; Pan, H -R; Park, J; Patton, S; Pec, V; Peng, J C; Pinsky, L; Pun, C S J; Qi, F Z; Qi, M; Qian, X; Raper, N; Ren, J; Rosero, R; Roskovec, B; Ruan, X C; Steiner, H; Sun, G X; Sun, J L; Tang, W; Taychenachev, D; Treskov, K; Tsang, K V; Tull, C E; Viaux, N; Viren, B; Vorobel, V; Wang, C H; Wang, M; Wang, N Y; Wang, R G; Wang, W; Wang, X; Wang, Y F; Wang, Z; Wang, Z; Wang, Z M; Wei, H Y; Wen, L J; Whisnant, K; White, C G; Whitehead, L; Wise, T; Wong, H L H; Wong, S C F; Worcester, E; Wu, C -H; Wu, Q; Wu, W J; Xia, D M; Xia, J K; Xing, Z Z; Xu, J Y; Xu, J L; Xu, Y; Xue, T; Yang, C G; Yang, H; Yang, L; Yang, M S; Yang, M T; Ye, M; Ye, Z; Yeh, M; Young, B L; Yu, Z Y; Zeng, S; Zhan, L; Zhang, C; Zhang, H H; Zhang, J W; Zhang, Q M; Zhang, X T; Zhang, Y M; Zhang, Y X; Zhang, Y M; Zhang, Z J; Zhang, Z Y; Zhang, Z P; Zhao, J; Zhao, Q W; Zhao, Y B; Zhong, W L; Zhou, L; Zhou, N; Zhuang, H L; Zou, J H

    2016-01-01

    A new measurement of the reactor antineutrino flux and energy spectrum by the Daya Bay reactor neutrino experiment is reported. The antineutrinos were generated by six 2.9 GW$_{\\mathrm{th}}$ nuclear reactors and detected by eight antineutrino detectors deployed in two near (510~m and 560~m flux-weighted baselines) and one far (1580~m flux-weighted baseline) underground experimental halls. With 621 days of data, more than 1.2 million inverse beta decay (IBD) candidates were detected. The IBD yield in the eight detectors was measured, and the ratio of measured to predicted flux was found to be $0.946\\pm0.020$ ($0.992\\pm0.021$) for the Huber+Mueller (ILL+Vogel) model. A 2.9 $\\sigma$ deviation was found in the measured IBD positron energy spectrum compared to the predictions. In particular, an excess of events in the region of 4-6~MeV was found in the measured spectrum, with a local significance of 4.4 $\\sigma$. A reactor antineutrino spectrum weighted by the IBD cross section is extracted for model-independent p...

  7. The study of aeroball system for measuring 3D neutron flux distribution in reactor core

    Institute of Scientific and Technical Information of China (English)

    LuoZheng-Pei; LiFu; 等

    1997-01-01

    Aeroball system is attractive in several aspects because it can easily transport the map of neutron flux distribution to be measured from incore to outside of a reactor vessel.However,before the aeroball system is put to practical use in the heating reactor.there are four topics that have to be further studied.They are the stability of the activated positions,enhancement of signal/noise(S/N)ratio,distributed control and data-acquisition system and on-lin nbeutron flux distribution reconstruction.Besides describing the rasons for them,this paper gives out the theory,concept and solution about the first two topics and it is helptul to give the possibility to enhance the reactor-power.

  8. Neutron flux measurements in the side-core region of Hunterston B advanced gas-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Allen, D.A. [Serco, Rutherford House, Quedgeley, Gloucester, GL2 4NF (United Kingdom); Shaw, S.E. [British Energy, Barnett Way, Barnwood, Gloucester, GL4 3RS (United Kingdom); Huggon, A.P.; Steadman, R.J.; Thornton, D.A. [Serco, Rutherford House, Quedgeley, Gloucester, GL2 4NF (United Kingdom); Whiley, G.S. [British Energy, Barnett Way, Barnwood, Gloucester, GL4 3RS (United Kingdom)

    2011-07-01

    The core restraints of advanced gas-cooled reactors are important structural components that are required to maintain the geometric integrity of the cores. A review of neutron dosimetry for the sister stations Hunterston B and Hinkley Point B identified that earlier conservative assessments predicted high thermal neutron dose rates to key components of the restraint structure (the restraint rod welds), with the implication that some of them may be predicted to fail during a seismic event. A revised assessment was therefore undertaken [Thornton, D. A., Allen, D. A., Tyrrell, R. J., Meese, T. C., Huggon, A.P., Whiley, G. S., and Mossop, J. R., 'A Dosimetry Assessment for the Core Restraint of an Advanced Gas Cooled Reactor,' Proceedings of the 13. International Symposium on Reactor Dosimetry (ISRD-13, May 2008), World Scientific, River Edge, NJ, 2009, W. Voorbraak, L. Debarberis, and P. D'hondt, Eds., pp. 679-687] using a detailed 3D model and a Monte Carlo radiation transport program, MCBEND. This reassessment resulted in more realistic fast and thermal neutron dose recommendations, the latter in particular being much lower than had been thought previously. It is now desirable to improve confidence in these predictions by providing direct validation of the MCBEND model through the use of neutron flux measurements. This paper describes the programme of work being undertaken to deploy two neutron flux measurement 'stringers' within the side-core region of one of the Hunterston B reactors for the purpose of validating the MCBEND model. The design of the stringers and the determination of the preferred deployment locations have been informed by the use of detailed MCBEND flux calculations. These computational studies represent a rare opportunity to design a flux measurement beforehand, with the clear intention of minimising the anticipated uncertainties and obtaining measurements that are known to be representative of the neutron fields to which

  9. Chemical research at Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    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.

  10. Scaleable, High Efficiency Microchannel Sabatier Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — A Microchannel Sabatier Reactor System (MSRS) consisting of cross connected arrays of isothermal or graded temperature reactors is proposed. The reactor array...

  11. Steel slag carbonation in a flow-through reactor system:The role of fluid-flux

    Institute of Scientific and Technical Information of China (English)

    Eleanor J.Berryman; Anthony E.Williams-Jones; Artashes A.Migdisov

    2015-01-01

    Steel production is currently the largest industrial source of atmospheric CO2.As annual steel production continues to grow,the need for effective methods of reducing its carbon footprint increases correspondingly.The carbonation of the calcium-bearing phases in steel slag generated during basic oxygen furnace (BOF) steel production,in particular its major constituent,lamite {Ca2SiO4},which is a structural analogue of olivine {(MgFe)2SiO4},the main mineral subjected to natural carbonation in peridotites,offers the potential to offset some of these emissions.However,the controls on the nature and efficiency of steel slag carbonation are yet to be completely understood.Experiments were conducted exposing steel slag grains to a CO2-H2O mixture in both batch and flow-through reactors to investigate the impact of temperature,fluid flux,and reaction gradient on the dissolution and carbonation of steel slag.The results of these experiments show that dissolution and carbonation of BOF steel slag are more efficient in a flow-through reactor than in the batch reactors used in most previous studies.Moreover,they show that fluid flux needs to be optimized in addition to grain size,pressure,and temperature,in order to maximize the efficiency of carbonation.Based on these results,a two-stage reactor consisting of a high and a low fluid-flux chamber is proposed for CO2 sequestration by steel slag carbonation,allowing dissolution of the slag and precipitation of calcium carbonate to occur within a single flow-through system.

  12. Wire chamber degradation at the Argonne ZGS

    Energy Technology Data Exchange (ETDEWEB)

    Haberichter, W.; Spinka, H.

    1986-01-01

    Experience with multiwire proportional chambers at high rates at the Argonne Zero Gradient Synchrotron is described. A buildup of silicon on the sense wires was observed where the beam passed through the chamber. Analysis of the chamber gas indicated that the density of silicon was probably less than 10 ppM.

  13. High flux lithium antineutrino source with variable hard spectrum

    CERN Document Server

    Lyashuk, V I

    2016-01-01

    The high flux antineutrino source with hard antineutrino spectrum based on neutron activation of 7Li and subsequent fast beta-decay (T 1/2 = 0.84 s) of the 8Li isotope with emission of antineutrino with energy up to 13 MeV - is discussed. Creation of the intensive isotope neutrino source of hard spectrum will allow to increase the detection statistics of neutrino interaction and it is especially urgent for oscillation experiments. The scheme of the proposed neutrino source is based on the continuous transport of the created 8Li to the neutrino detector, which moved away from the place of neutron activation. Analytical expressions for lithium antineutrino flux is obtained. The discussed source will ensure to increase the cross section for reactions with deuteron from several times to tens compare to the reactor antineutrino spectrum. An another unique feature of the installation is the possibility to vary smoothly the hardness of the antineutrino spectrum.

  14. Calculation of the inventory and near-field release rates of radioactivity from neutron-activated metal parts discharged from the high flux isotope reactor and emplaced in solid waste storage area 6 at Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Kelmers, A.D.; Hightower, J.R.

    1987-05-01

    Emplacement of contaminated reactor components involves disposal in lined and unlined auger holes in soil above the water table. The radionuclide inventory of disposed components was calculated. Information on the composition and weight of the components, as well as reasonable assumptions for the neutron flux fueling use, the time of neutron exposure, and radioactive decay after discharge, were employed in the inventory calculation. Near-field release rates of /sup 152/Eu, /sup 154/Eu, and /sup 155/Eu from control plates and cylinders were calculated for 50 years after emplacement. Release rates of the europium isotopes were uncertain. Two release-rate-limiting models were considered and a range of reasonable values were assumed for the time-to-failure of the auger-hole linear and aluminum cladding and europium solubility in SWSA-6 groundwater. The bounding europium radionuclide near-field release rates peaked at about 1.3 Ci/year total for /sup 152,154,155/Eu in 1987 for the lower bound, and at about 420 Ci/year in 1992 for the upper bound. The near-field release rates of /sup 55/Fe, /sup 59/Ni, /sup 60/Co, and /sup 63/Ni from stainless steel and cobalt alloy components, as well as of /sup 10/Be, /sup 41/Ca, and /sup 55/Fe from beryllium reflectors, were calculated for the next 100 years, assuming bulk waste corrosion was the release-rate-limiting step. Under the most conservative assumptions for the reflectors, the current (1986) total radionuclide release rate was calculated to be about 1.2 x 10/sup -4/ Ci/year, decreasing by 1992 to a steady release of about 1.5 x 10/sup -5/ Ci/year due primarily to /sup 41/Ca. 50 refs., 13 figs., 8 tabs.

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

    Science.gov (United States)

    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.

  16. Neutron Flux Measurement at TAPIRO Fast Reactor for APD's Irradiation Fluence Evaluation

    CERN Document Server

    Angelone, M; Diemoz, Marcella; Festinesi, Armando; Longo, Egidio; Organtini, Giovanni; Rosi, G

    1998-01-01

    The Avalanche Photodiodes ( APD) were chosen as photon sensors for the region of the CMS electromagnetic calorimeter. The LHC will be a hard environment for what concerns the radiation levels in the detectors. The most relevant damage on APDs is caused by neutrons that produce an increase in the dark current of these devices. In the CMS-ECAL collaboration a big effort was indeed done to understand this damage, but the evaluation of the absolute effect was limited by the knowledge of the neutron flux calibration of the various irradiation facilities. This investigation describes the calibration of the neutron flux of the Tapiro reactor in Rome and the calculation of the Non-Ionizing-Energy-Loss on Silicon for this reactor. The damage parameter alpha for the APDs is evaluated to be about 10-11*10^-17 A/cm/neutron at 18C and 2 days after the irradiation. Some cross-checks with other irradiation facilities are also presented.

  17. Ultra high temperature particle bed reactor design

    Science.gov (United States)

    Lazareth, Otto; Ludewig, Hans; Perkins, K.; Powell, J.

    1990-01-01

    A direct nuclear propulsion engine which could be used for a mission to Mars is designed. The main features of this reactor design are high values for I(sub sp) and very efficient cooling. This particle bed reactor consists of 37 cylindrical fuel elements embedded in a cylinder of beryllium which acts as a moderator and reflector. The fuel consists of a packed bed of spherical fissionable fuel particles. Gaseous H2 passes over the fuel bed, removes the heat, and is exhausted out of the rocket. The design was found to be neutronically critical and to have tolerable heating rates. Therefore, this particle bed reactor design is suitable as a propulsion unit for this mission.

  18. Improved measurement of the reactor antineutrino flux and spectrum at Daya Bay

    Science.gov (United States)

    An, F. P.; Balantekin, A. B.; Band, H. R.; Bishai, M.; Blyth, S.; Cao, D.; Cao, G. F.; Cao, J.; Cen, W. R.; Chan, Y. L.; Chang, J. F.; Chang, L. C.; Chang, Y.; Chen, H. S.; Chen, Q. Y.; Chen, S. M.; Chen, Y. X.; Chen, Y.; Cheng, J.-H.; Cheng, J.; Cheng, Y. P.; Cheng, Z. K.; Cherwinka, J. J.; Chu, M. C.; Chukanov, A.; Cummings, J. P.; de Arcos, J.; Deng, Z. Y.; Ding, X. F.; Ding, Y. Y.; Diwan, M. V.; Dolgareva, M.; Dove, J.; Dwyer, D. A.; Edwards, W. R.; Gill, R.; Gonchar, M.; Gong, G. H.; Gong, H.; Grassi, M.; Gu, W. Q.; Guan, M. Y.; Guo, L.; Guo, R. P.; Guo, X. H.; Guo, Z.; Hackenburg, R. W.; Han, R.; Hans, S.; He, M.; Heeger, K. M.; Heng, Y. K.; Higuera, A.; Hor, Y. K.; Hsiung, Y. B.; Hu, B. Z.; Hu, T.; Hu, W.; Huang, E. C.; Huang, H. X.; Huang, X. T.; Huber, P.; Huo, W.; Hussain, G.; Jaffe, D. E.; Jaffke, P.; Jen, K. L.; Jetter, S.; Ji, X. P.; Ji, X. L.; Jiao, J. B.; Johnson, R. A.; Jones, D.; Joshi, J.; Kang, L.; Kettell, S. H.; Kohn, S.; Kramer, M.; Kwan, K. K.; Kwok, M. W.; Kwok, T.; Langford, T. J.; Lau, K.; Lebanowski, L.; Lee, J.; Lee, J. H. C.; Lei, R. T.; Leitner, R.; Li, C.; Li, D. J.; Li, F.; Li, G. S.; Li, Q. J.; Li, S.; Li, S. C.; Li, W. D.; Li, X. N.; Li, Y. F.; Li, Z. B.; Liang, H.; Lin, C. J.; Lin, G. L.; Lin, S.; Lin, S. K.; Lin, Y.-C.; Ling, J. J.; Link, J. M.; Littenberg, L.; Littlejohn, B. R.; Liu, D. W.; Liu, J. L.; Liu, J. C.; Loh, C. W.; Lu, C.; Lu, H. Q.; Lu, J. S.; Luk, K. B.; Lv, Z.; Ma, Q. M.; Ma, X. Y.; Ma, X. B.; Ma, Y. Q.; Malyshkin, Y.; Martinez Caicedo, D. A.; McDonald, K. T.; McKeown, R. D.; Mitchell, I.; Mooney, M.; Nakajima, Y.; Napolitano, J.; Naumov, D.; Naumova, E.; Ngai, H. Y.; Ning, Z.; Ochoa-Ricoux, J. P.; Olshevskiy, A.; Pan, H.-R.; Park, J.; Patton, S.; Pec, V.; Peng, J. C.; Pinsky, L.; Pun, C. S. J.; Qi, F. Z.; Qi, M.; Qian, X.; Raper, N.; Ren, J.; Rosero, R.; Roskovec, B.; Ruan, X. C.; Steiner, H.; Sun, G. X.; Sun, J. L.; Tang, W.; Taychenachev, D.; Treskov, K.; Tsang, K. V.; Tull, C. E.; Viaux, N.; Viren, B.; Vorobel, V.; Wang, C. H.; Wang, M.; Wang, N. Y.; Wang, R. G.; Wang, W.; Wang, X.; Wang, Y. F.; Wang, Z.; Wang, Z.; Wang, Z. M.; Wei, H. Y.; Wen, L. J.; Whisnant, K.; White, C. G.; Whitehead, L.; Wise, T.; Wong, H. L. H.; Wong, S. C. F.; Worcester, E.; Wu, C.-H.; Wu, Q.; Wu, W. J.; Xia, D. M.; Xia, J. K.; Xing, Z. Z.; Xu, J. Y.; Xu, J. L.; Xu, Y.; Xue, T.; Yang, C. G.; Yang, H.; Yang, L.; Yang, M. S.; Yang, M. T.; Ye, M.; Ye, Z.; Yeh, M.; Young, B. L.; Yu, Z. Y.; Zeng, S.; Zhan, L.; Zhang, C.; Zhang, H. H.; Zhang, J. W.; Zhang, Q. M.; Zhang, X. T.; Zhang, Y. M.; Zhang, Y. X.; Zhang, Y. M.; Zhang, Z. J.; Zhang, Z. Y.; Zhang, Z. P.; Zhao, J.; Zhao, Q. W.; Zhao, Y. B.; Zhong, W. L.; Zhou, L.; Zhou, N.; Zhuang, H. L.; Zou, J. H.; Daya Bay Collaboration

    2017-01-01

    A new measurement of the reactor antineutrino flux and energy spectrum by the Daya Bay reactor neutrino experiment is reported. The antineutrinos were generated by six 2.9 GWth nuclear reactors and detected by eight antineutrino detectors deployed in two near (560 m and 600 m flux-weighted baselines) and one far (1640 m flux-weighted baseline) underground experimental halls. With 621 days of data, more than 1.2 million inverse beta decay (IBD) candidates were detected. The IBD yield in the eight detectors was measured, and the ratio of measured to predicted flux was found to be 0.946±0.020 (0.992±0.021) for the Huber+Mueller (ILL+Vogel) model. A 2.9σ deviation was found in the measured IBD positron energy spectrum compared to the predictions. In particular, an excess of events in the region of 4-6 MeV was found in the measured spectrum, with a local significance of 4.4σ. A reactor antineutrino spectrum weighted by the IBD cross section is extracted for model-independent predictions. Supported in part by the Ministry of Science and Technology of China, the United States Department of Energy, the Chinese Academy of Sciences, the CAS Center for Excellence in Particle Physics, the National Natural Science Foundation of China, the Guangdong provincial government, the Shenzhen municipal government, the China General Nuclear Power Group, the Research Grants Council of the Hong Kong Special Administrative Region of China, the MOST and MOE in Taiwan, the U.S. National Science Foundation, the Ministry of Education, Youth and Sports of the Czech Republic, the Joint Institute of Nuclear Research in Dubna, Russia, the NSFC-RFBR joint research program, the National Commission for Scientific and Technological Research of Chile

  19. Measurement of the Reactor Antineutrino Flux and Spectrum at Daya Bay

    CERN Document Server

    An, F P; Band, H R; Bishai, M; Blyth, S; Butorov, I; Cao, D; Cao, G F; Cao, J; Cen, W R; Chan, Y L; Chang, J F; Chang, L C; Chang, Y; Chen, H S; Chen, Q Y; Chen, S M; Chen, Y X; Chen, Y; Cheng, J H; Cheng, J; Cheng, Y P; Cherwinka, J J; Chu, M C; Cummings, J P; de Arcos, J; Deng, Z Y; Ding, X F; Ding, Y Y; Diwan, M V; Dove, J; Draeger, E; Dwyer, D A; Edwards, W R; Ely, S R; Gill, R; Gonchar, M; Gong, G H; Gong, H; Grassi, M; Gu, W Q; Guan, M Y; Guo, L; Guo, X H; Hackenburg, R W; Han, R; Hans, S; He, M; Heeger, K M; Heng, Y K; Higuera, A; Hor, Y K; Hsiung, Y B; Hu, B Z; Hu, L M; Hu, L J; Hu, T; Hu, W; Huang, E C; Huang, H X; Huang, X T; Huber, P; Hussain, G; Jaffe, D E; Jaffke, P; Jen, K L; Jetter, S; Ji, X P; Ji, X L; Jiao, J B; Johnson, R A; Kang, L; Kettell, S H; Kohn, S; Kramer, M; Kwan, K K; Kwok, M W; Kwok, T; Langford, T J; Lau, K; Lebanowski, L; Lee, J; Lei, R T; Leitner, R; Leung, K Y; Leung, J K C; Lewis, C A; Li, D J; Li, F; Li, G S; Li, Q J; Li, S C; Li, W D; Li, X N; Li, X Q; Li, Y F; Li, Z B; Liang, H; Lin, C J; Lin, G L; Lin, P Y; Lin, S K; Ling, J J; Link, J M; Littenberg, L; Littlejohn, B R; Liu, D W; Liu, H; Liu, J L; Liu, J C; Liu, S S; Lu, C; Lu, H Q; Lu, J S; Luk, K B; Ma, Q M; Ma, X Y; Ma, X B; Ma, Y Q; Caicedo, D A Martinez; McDonald, K T; McKeown, R D; Meng, Y; Mitchell, I; Kebwaro, J Monari; Nakajima, Y; Napolitano, J; Naumov, D; Naumova, E; Ngai, H Y; Ning, Z; Ochoa-Ricoux, J P; Olshevski, A; Pan, H -R; Park, J; Patton, S; Pec, V; Peng, J C; Piilonen, L E; Pinsky, L; Pun, C S J; Qi, F Z; Qi, M; Qian, X; Raper, N; Ren, B; Ren, J; Rosero, R; Roskovec, B; Ruan, X C; Shao, B B; Steiner, H; Sun, G X; Sun, J L; Tang, W; Taychenachev, D; Tsang, K V; Tull, C E; Tung, Y C; Viaux, N; Viren, B; Vorobel, V; Wang, C H; Wang, M; Wang, N Y; Wang, R G; Wang, W; Wang, W W; Wang, X; Wang, Y F; Wang, Z; Wang, Z M; Wei, H Y; Wen, L J; Whisnant, K; White, C G; Whitehead, L; Wise, T; Wong, H L H; Wong, S C F; Worcester, E; Wu, Q; Xia, D M; Xia, J K; Xia, X; Xing, Z Z; Xu, J Y; Xu, J L; Xu, J; Xu, Y; Xue, T; Yan, J; Yang, C G; Yang, L; Yang, M S; Yang, M T; Ye, M; Yeh, M; Young, B L; Yu, G Y; Yu, Z Y; Zang, S L; Zhan, L; Zhang, C; Zhang, H H; Zhang, J W; Zhang, Q M; Zhang, Y M; Zhang, Y X; Zhang, Z J; Zhang, Z Y; Zhang, Z P; Zhao, J; Zhao, Q W; Zhao, Y F; Zhao, Y B; Zheng, L; Zhong, W L; Zhou, L; Zhou, N; Zhuang, H L; Zou, J H

    2015-01-01

    This Letter reports a measurement of the flux and energy spectrum of electron antineutrinos from six 2.9~GW$_{th}$ nuclear reactors with six detectors deployed in two near (effective baselines 512~m and 561~m) and one far (1,579~m) underground experimental halls in the Daya Bay experiment. Using 217 days of data, 296,721 and 41,589 inverse beta decay (IBD) candidates were detected in the near and far halls, respectively. The measured IBD yield is (1.55 $\\pm$ 0.04) $\\times$ 10$^{-18}$~cm$^2$/GW/day or (5.92 $\\pm$ 0.14) $\\times$ 10$^{-43}$~cm$^2$/fission. This flux measurement is consistent with previous short-baseline reactor antineutrino experiments and is $0.946\\pm0.022$ ($0.991\\pm0.023$) relative to the flux predicted with the Huber+Mueller (ILL+Vogel) fissile antineutrino model. The measured IBD positron energy spectrum deviates from both spectral predictions by more than 2$\\sigma$ over the full energy range with a local significance of up to $\\sim$4$\\sigma$ between 4-6 MeV. A reactor antineutrino spectrum...

  20. Measurement of the Reactor Antineutrino Flux and Spectrum at Daya Bay

    Science.gov (United States)

    An, F. P.; Balantekin, A. B.; Band, H. R.; Bishai, M.; Blyth, S.; Butorov, I.; Cao, D.; Cao, G. F.; Cao, J.; Cen, W. R.; Chan, Y. L.; Chang, J. F.; Chang, L. C.; Chang, Y.; Chen, H. S.; Chen, Q. Y.; Chen, S. M.; Chen, Y. X.; Chen, Y.; Cheng, J. H.; Cheng, J.; Cheng, Y. P.; Cherwinka, J. J.; Chu, M. C.; Cummings, J. P.; de Arcos, J.; Deng, Z. Y.; Ding, X. F.; Ding, Y. Y.; Diwan, M. V.; Dove, J.; Draeger, E.; Dwyer, D. A.; Edwards, W. R.; Ely, S. R.; Gill, R.; Gonchar, M.; Gong, G. H.; Gong, H.; Grassi, M.; Gu, W. Q.; Guan, M. Y.; Guo, L.; Guo, X. H.; Hackenburg, R. W.; Han, R.; Hans, S.; He, M.; Heeger, K. M.; Heng, Y. K.; Higuera, A.; Hor, Y. K.; Hsiung, Y. B.; Hu, B. Z.; Hu, L. M.; Hu, L. J.; Hu, T.; Hu, W.; Huang, E. C.; Huang, H. X.; Huang, X. T.; Huber, P.; Hussain, G.; Jaffe, D. E.; Jaffke, P.; Jen, K. L.; Jetter, S.; Ji, X. P.; Ji, X. L.; Jiao, J. B.; Johnson, R. A.; Kang, L.; Kettell, S. H.; Kohn, S.; Kramer, M.; Kwan, K. K.; Kwok, M. W.; Kwok, T.; Langford, T. J.; Lau, K.; Lebanowski, L.; Lee, J.; Lei, R. T.; Leitner, R.; Leung, K. Y.; Leung, J. K. C.; Lewis, C. A.; Li, D. J.; Li, F.; Li, G. S.; Li, Q. J.; Li, S. C.; Li, W. D.; Li, X. N.; Li, X. Q.; Li, Y. F.; Li, Z. B.; Liang, H.; Lin, C. J.; Lin, G. L.; Lin, P. Y.; Lin, S. K.; Ling, J. J.; Link, J. M.; Littenberg, L.; Littlejohn, B. R.; Liu, D. W.; Liu, H.; Liu, J. L.; Liu, J. C.; Liu, S. S.; Lu, C.; Lu, H. Q.; Lu, J. S.; Luk, K. B.; Ma, Q. M.; Ma, X. Y.; Ma, X. B.; Ma, Y. Q.; Martinez Caicedo, D. A.; McDonald, K. T.; McKeown, R. D.; Meng, Y.; Mitchell, I.; Monari Kebwaro, J.; Nakajima, Y.; Napolitano, J.; Naumov, D.; Naumova, E.; Ngai, H. Y.; Ning, Z.; Ochoa-Ricoux, J. P.; Olshevski, A.; Pan, H.-R.; Park, J.; Patton, S.; Pec, V.; Peng, J. C.; Piilonen, L. E.; Pinsky, L.; Pun, C. S. J.; Qi, F. Z.; Qi, M.; Qian, X.; Raper, N.; Ren, B.; Ren, J.; Rosero, R.; Roskovec, B.; Ruan, X. C.; Shao, B. B.; Steiner, H.; Sun, G. X.; Sun, J. L.; Tang, W.; Taychenachev, D.; Tsang, K. V.; Tull, C. E.; Tung, Y. C.; Viaux, N.; Viren, B.; Vorobel, V.; Wang, C. H.; Wang, M.; Wang, N. Y.; Wang, R. G.; Wang, W.; Wang, W. W.; Wang, X.; Wang, Y. F.; Wang, Z.; Wang, Z.; Wang, Z. M.; Wei, H. Y.; Wen, L. J.; Whisnant, K.; White, C. G.; Whitehead, L.; Wise, T.; Wong, H. L. H.; Wong, S. C. F.; Worcester, E.; Wu, Q.; Xia, D. M.; Xia, J. K.; Xia, X.; Xing, Z. Z.; Xu, J. Y.; Xu, J. L.; Xu, J.; Xu, Y.; Xue, T.; Yan, J.; Yang, C. G.; Yang, L.; Yang, M. S.; Yang, M. T.; Ye, M.; Yeh, M.; Young, B. L.; Yu, G. Y.; Yu, Z. Y.; Zang, S. L.; Zhan, L.; Zhang, C.; Zhang, H. H.; Zhang, J. W.; Zhang, Q. M.; Zhang, Y. M.; Zhang, Y. X.; Zhang, Y. M.; Zhang, Z. J.; Zhang, Z. Y.; Zhang, Z. P.; Zhao, J.; Zhao, Q. W.; Zhao, Y. F.; Zhao, Y. B.; Zheng, L.; Zhong, W. L.; Zhou, L.; Zhou, N.; Zhuang, H. L.; Zou, J. H.; Daya Bay Collaboration

    2016-02-01

    This Letter reports a measurement of the flux and energy spectrum of electron antineutrinos from six 2.9 GWt h nuclear reactors with six detectors deployed in two near (effective baselines 512 and 561 m) and one far (1579 m) underground experimental halls in the Daya Bay experiment. Using 217 days of data, 296 721 and 41 589 inverse β decay (IBD) candidates were detected in the near and far halls, respectively. The measured IBD yield is (1.55 ±0.04 ) ×10-18 cm2 GW-1 day-1 or (5.92 ±0.14 ) ×10-43 cm2 fission-1 . This flux measurement is consistent with previous short-baseline reactor antineutrino experiments and is 0.946 ±0.022 (0.991 ±0.023 ) relative to the flux predicted with the Huber -Mueller (ILL -Vogel ) fissile antineutrino model. The measured IBD positron energy spectrum deviates from both spectral predictions by more than 2 σ over the full energy range with a local significance of up to ˜4 σ between 4-6 MeV. A reactor antineutrino spectrum of IBD reactions is extracted from the measured positron energy spectrum for model-independent predictions.

  1. Measurement of the Reactor Antineutrino Flux and Spectrum at Daya Bay.

    Science.gov (United States)

    An, F P; Balantekin, A B; Band, H R; Bishai, M; Blyth, S; Butorov, I; Cao, D; Cao, G F; Cao, J; Cen, W R; Chan, Y L; Chang, J F; Chang, L C; Chang, Y; Chen, H S; Chen, Q Y; Chen, S M; Chen, Y X; Chen, Y; Cheng, J H; Cheng, J; Cheng, Y P; Cherwinka, J J; Chu, M C; Cummings, J P; de Arcos, J; Deng, Z Y; Ding, X F; Ding, Y Y; Diwan, M V; Dove, J; Draeger, E; Dwyer, D A; Edwards, W R; Ely, S R; Gill, R; Gonchar, M; Gong, G H; Gong, H; Grassi, M; Gu, W Q; Guan, M Y; Guo, L; Guo, X H; Hackenburg, R W; Han, R; Hans, S; He, M; Heeger, K M; Heng, Y K; Higuera, A; Hor, Y K; Hsiung, Y B; Hu, B Z; Hu, L M; Hu, L J; Hu, T; Hu, W; Huang, E C; Huang, H X; Huang, X T; Huber, P; Hussain, G; Jaffe, D E; Jaffke, P; Jen, K L; Jetter, S; Ji, X P; Ji, X L; Jiao, J B; Johnson, R A; Kang, L; Kettell, S H; Kohn, S; Kramer, M; Kwan, K K; Kwok, M W; Kwok, T; Langford, T J; Lau, K; Lebanowski, L; Lee, J; Lei, R T; Leitner, R; Leung, K Y; Leung, J K C; Lewis, C A; Li, D J; Li, F; Li, G S; Li, Q J; Li, S C; Li, W D; Li, X N; Li, X Q; Li, Y F; Li, Z B; Liang, H; Lin, C J; Lin, G L; Lin, P Y; Lin, S K; Ling, J J; Link, J M; Littenberg, L; Littlejohn, B R; Liu, D W; Liu, H; Liu, J L; Liu, J C; Liu, S S; Lu, C; Lu, H Q; Lu, J S; Luk, K B; Ma, Q M; Ma, X Y; Ma, X B; Ma, Y Q; Martinez Caicedo, D A; McDonald, K T; McKeown, R D; Meng, Y; Mitchell, I; Monari Kebwaro, J; Nakajima, Y; Napolitano, J; Naumov, D; Naumova, E; Ngai, H Y; Ning, Z; Ochoa-Ricoux, J P; Olshevski, A; Pan, H-R; Park, J; Patton, S; Pec, V; Peng, J C; Piilonen, L E; Pinsky, L; Pun, C S J; Qi, F Z; Qi, M; Qian, X; Raper, N; Ren, B; Ren, J; Rosero, R; Roskovec, B; Ruan, X C; Shao, B B; Steiner, H; Sun, G X; Sun, J L; Tang, W; Taychenachev, D; Tsang, K V; Tull, C E; Tung, Y C; Viaux, N; Viren, B; Vorobel, V; Wang, C H; Wang, M; Wang, N Y; Wang, R G; Wang, W; Wang, W W; Wang, X; Wang, Y F; Wang, Z; Wang, Z; Wang, Z M; Wei, H Y; Wen, L J; Whisnant, K; White, C G; Whitehead, L; Wise, T; Wong, H L H; Wong, S C F; Worcester, E; Wu, Q; Xia, D M; Xia, J K; Xia, X; Xing, Z Z; Xu, J Y; Xu, J L; Xu, J; Xu, Y; Xue, T; Yan, J; Yang, C G; Yang, L; Yang, M S; Yang, M T; Ye, M; Yeh, M; Young, B L; Yu, G Y; Yu, Z Y; Zang, S L; Zhan, L; Zhang, C; Zhang, H H; Zhang, J W; Zhang, Q M; Zhang, Y M; Zhang, Y X; Zhang, Y M; Zhang, Z J; Zhang, Z Y; Zhang, Z P; Zhao, J; Zhao, Q W; Zhao, Y F; Zhao, Y B; Zheng, L; Zhong, W L; Zhou, L; Zhou, N; Zhuang, H L; Zou, J H

    2016-02-12

    This Letter reports a measurement of the flux and energy spectrum of electron antineutrinos from six 2.9 GWth nuclear reactors with six detectors deployed in two near (effective baselines 512 and 561 m) and one far (1579 m) underground experimental halls in the Daya Bay experiment. Using 217 days of data, 296 721 and 41 589 inverse β decay (IBD) candidates were detected in the near and far halls, respectively. The measured IBD yield is (1.55±0.04) ×10(-18)  cm(2) GW(-1) day(-1) or (5.92±0.14) ×10(-43)  cm(2) fission(-1). This flux measurement is consistent with previous short-baseline reactor antineutrino experiments and is 0.946±0.022 (0.991±0.023) relative to the flux predicted with the Huber-Mueller (ILL-Vogel) fissile antineutrino model. The measured IBD positron energy spectrum deviates from both spectral predictions by more than 2σ over the full energy range with a local significance of up to ∼4σ between 4-6 MeV. A reactor antineutrino spectrum of IBD reactions is extracted from the measured positron energy spectrum for model-independent predictions.

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

    Science.gov (United States)

    Mozley, A

    1971-10-01

    , William B. Cannon, who is vice president of programs and projects of the University of Chicago, and a small selection of staff members believe that the Laboratory is going through a natural and inevitable process of change consonant with altered missions and objectives in an atomic energy laboratory. The general mood, however, demonstrates the Jeffersonian insight, as relevant in science as in politics, that only democratic governance provides salutary checks and balances when things go wrong. The point deserves close scrutiny when Argonne's tripartite contract comes up for renegotiation in October 1971. Fundamentally Argonne's relations with its sponsoring agency remain at the center of its progress and future plans. Despite administrative and management changes, there is little doubt that he who pays the piper calls the tune. In common with other federal contract research and development adjuncts, Argonne has undoubtedly undergone tightening and winnowing away of flexibility in the past 6 years. In the nuclear reactor program the consequences have been strongly felt, and stringent national budgets have widened the tendency in the research domain. The impact of these changes and of AEC's attitude to basic research raise large questions for the future of the national laboratories. Few doubt that these "major national assets," with their outstanding scientific and technical personnel and equipment, fulfill a unique function and are here to stay, though their missions may undergo some change; the question of their most effective direction and handling, however, remains crucial for those concerned with priorities and decision-making for science. A recent review of 40 national federal adjuncts (30,31) has indicated that the primary sponsoring agency obtains better performance from a center that has a relatively high degree of independence than from one that is tightly controlled. The point is confirmed at Argonne where the present tendency (particularly on the nuclear reactor

  3. Neutron flux measurements at the TRIGA reactor in Vienna for the prediction of the activation of the biological shield

    Energy Technology Data Exchange (ETDEWEB)

    Merz, Stefan [Vienna University of Technology, Atominstitut, Stadionallee 2, 1020 Vienna (Austria); Djuricic, Mile [Vienna University of Technology, Atominstitut, Stadionallee 2, 1020 Vienna (Austria); Nuclear Engineering Seibersdorf, 2444 Seibersdorf (Austria); Villa, Mario; Boeck, Helmuth [Vienna University of Technology, Atominstitut, Stadionallee 2, 1020 Vienna (Austria); Steinhauser, Georg, E-mail: georg.steinhauser@ati.ac.at [Vienna University of Technology, Atominstitut, Stadionallee 2, 1020 Vienna (Austria)

    2011-11-15

    The activation of the biological shield is an important process for waste management considerations of nuclear facilities. The final activity can be estimated by modeling using the neutron flux density rather than the radiometric approach of activity measurements. Measurement series at the TRIGA reactor Vienna reveal that the flux density next to the biological shield is in the order of 10{sup 9} cm{sup -2} s{sup -1} at maximum power; but it is strongly influenced by reactor installations. The data allow the estimation of the final waste categorization of the concrete according to the Austrian legislation. - Highlights: > Neutron activation is an important process for the waste management of nuclear facilities. > Biological shield of the TRIGA reactor Vienna has been topic of investigation. > Flux values allow a categorization of the concrete concerning radiation protection legislation. > Reactor installations are of great importance as neutron sources into the biological shield. > Every installation shows distinguishable flux profiles.

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

    Energy Technology Data Exchange (ETDEWEB)

    Stillman, J. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jaluvka, D. [Argonne National Lab. (ANL), Argonne, IL (United States); Wilson, E. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Foyto, L. P. [Univ. of Missouri, Columbia, MO (United States); Kutikkad, K. [Univ. of Missouri, Columbia, MO (United States); McKibben, J. C. [Univ. of Missouri, Columbia, MO (United States); Peters, N. J. [Univ. of Missouri, Columbia, MO (United States)

    2017-02-01

    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 in the Research and Test Reactor Department at the Argonne National Laboratory (ANL) and the MURR Facility. MURR LEU conversion is part of an overall effort to develop and qualify high-density fuel within the U.S. High Performance Research Reactor Conversion (USHPRR) program conducted by the U.S. Department of Energy National Nuclear Security Administration’s Office of Material Management and Minimization (M3).

  5. Neutron flux optimization in irradiation facilities at Peruvian research reactor RP-10

    Energy Technology Data Exchange (ETDEWEB)

    Vela, M.; Arrieta, R.; Salazar, A.; Urcia, A.; Canaza, D.; Felix, J; Veramendi, E.; Ovalle, E.; Giol, R.; Zapata, L.; Ramos, F.; Tordocillo, J. [Instituto Peruano de Energia Nuclear (IPEN), Lima (Peru). Direccion de Instalaciones. Dept. de Reactores]. E-mail: mvela@ipen.gob.pe; rarrieta@ipen.gob.pe

    2005-07-01

    In this work we show the values distribution of the neutron flux at Peruvian Research Reactor RP-10, determined under two different safety and control rods configurations. The method applied was to irradiate small gold foils in irradiation facilities of the core to carry out the nuclear reaction {sup 197}Au(n, {gamma}){sup 198}Au; then using a gamma spectrometry system and the Westcott formalism we obtained the neutron flux. The results confirm the favorable effect of such configurations, increasing the neutron flux, both thermal and epithermal. These results have consistency with the weekly activity reports of radioisotopes lots given by the Radioisotopes Production Plant and Neutron Activation Analysis Group. (author)

  6. Predicting ion flux uniformity at the ion extraction plate in a 3D ICP reactor

    Science.gov (United States)

    Roy, Abhra; Bhoj, Ananth

    2016-09-01

    In order to achieve better control in processing the wafer surface, the ion fluxes in a remote plasma system are often focused through one or more ion extraction plates between the main plasma chamber and the downstream wafer plane. The ion extraction plates are typically of showerhead pattern with multiple holes. The focus of this particular study is to predict the ion flux uniformity over the ion extraction plate for a full 3D inductively coupled discharge reactor model using Argon chemistry. We will use the commercial modeling tool, CFD-ACE +, which can address such a process involving gas flow, heat transfer, plasma physics, reaction chemistry and electromagnetics in a coupled fashion. The plasma characteristics in the chamber and uniformity of the ion fluxes at ion extraction plate are discussed. Parametric studies varying the geometrical dimensions and process conditions to determine the effect on ion flux uniformity are presented. The showerhead-like ion extraction plate will be modeled as a porous media with a specified porosity. Further, a spatially varying porosity of the ion extraction plate is used to simulate ion recombination in order to reduce the ion flux non-uniformity. The goal is to optimize the system maximizing the ion flux while maintaining the uniformity.

  7. High Performance Fuel Desing for Next Generation Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-31

    The use of internally and externally cooled annular fule rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and econmic assessment. The investigation was donducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperatre. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasiblity issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density.

  8. Study of efficient high-power, high-energy neutral beams for the Reference Mirror Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Fink, J.H.; Barr, W.L.; Hamilton, G.W.

    1976-11-11

    An injector design for the Reference Mirror Reactor is described which uses negative ions created by charge-exchange in a cesium vapor cell and neutralized by photodetachment. Some of the innovations discussed include a continuously operating cathode for an LBL/LLL ion source, a negative ion beam line with cooled grids, a high voltage accelerator configuration with insulators shielded from the neutron and gamma flux, and cryopanels which continuously cycle between pumping and outgassing modes.

  9. Localized fast neutron flux enhancement for damage experiments in a research reactor; Accroissement local du flux rapide pour des experiences de dommages dans un reacteur de recherche

    Energy Technology Data Exchange (ETDEWEB)

    Malouch, F

    2003-06-01

    In irradiation experiments on materials in the core of the Osiris reactor (CEA-Saclay) we seek to increase damage in irradiated samples and to reduce the duration of their stay in the core. Damage is essentially caused by fast neutrons (E {>=} 1 MeV); we have therefore pursued the possibility of a localized increase of their level in an irradiation experiment by using a flux converter device made up of fissile material arranged according to a suitable geometry that allows the converter to receive experiments. We have studied several parameters that are influential in the increase of fast neutron flux within the converter. We have also considered the problem of the converter's cooling in the core and its effect on the operation of the reactor. We have carried out a specific neutron calculation scheme based on the modular 2D-transport code APOLLO2 using a two-level transport method. Experimental validation of the flux calculation scheme was carried out in the ISIS reactor, the mock-up of OSIRIS, by optimizing the loading of fuel elements in the core. The experimental results show that the neutron calculation scheme computes the fluxes in close agreement with the measurements especially the fast flux. This study allows us to master the essential physical parameters needed for the design of a flux converter in an MTR reactor. (author)

  10. Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2010-03-01

    The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these

  11. High flux film and transition boiling

    Science.gov (United States)

    Witte, L. C.

    1993-02-01

    An investigation was conducted on the potential for altering the boiling curve through effects of high velocity and high subcooling. Experiments using water and Freon-113 flowing over cylindrical electrical heaters in crossflow were made to see how velocity and subcooling affect the boiling curve, especially the film and transition boiling regions. We sought subcooling levels down to near the freezing points of these two liquids to prove the concept that the critical heat flux and the minimum heat flux could be brought together, thereby averting the transition region altogether. Another emphasis was to gain insight into how the various boiling regions could be represented mathematically on various parts of the heating surface. Motivation for the research grew out of a realization that the effects of very high subcooling and velocity might be to avert the transition boiling altogether so that the unstable part of the boiling curve would not limit the application of high flux devices to temperatures less than the burnout temperatures. Summaries of results from the study are described. It shows that the potential for averting the transition region is good and points the way to further research that is needed to demonstrate the potential.

  12. Request for Naval Reactors Comment on Proposed Prometheus Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to JPL

    Energy Technology Data Exchange (ETDEWEB)

    D. Kokkinos

    2005-04-28

    The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory.

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

    Energy Technology Data Exchange (ETDEWEB)

    Stillman, J. A. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Div., Research and Test Reactor Dept.; Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Div., Research and Test Reactor Dept.; Wilson, E. H. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Div., Research and Test Reactor Dept.; Foyto, L. P. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; Kutikkad, K. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; McKibben, J. C. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; Peters, N. J. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; Cowherd, W. M. [Univ. of Missouri, Columbia, MO (United States). College of Engineering, Nuclear Engineering Program; Rickman, B. [Univ. of Missouri, Columbia, MO (United States). College of Engineering, Nuclear Engineering Program

    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

  14. Thermal Energetic Reactor with High Reproduction of Fission Materials

    Directory of Open Access Journals (Sweden)

    Vladimir M. Kotov

    2012-01-01

    On the base of thermal reactors with high fission materials reproduction world atomic power engineering development supplying higher power and requiring smaller speed of raw uranium mining, than in the variant with fast reactors, is possible.

  15. HIGH STRENGTH CONTROL RODS FOR NEUTRONIC REACTORS

    Science.gov (United States)

    Lustman, B.; Losco, E.F.; Cohen, I.

    1961-07-11

    Nuclear reactor control rods comprised of highly compressed and sintered finely divided metal alloy panticles and fine metal oxide panticles substantially uniformly distributed theretbrough are described. The metal alloy consists essentially of silver, indium, cadmium, tin, and aluminum, the amount of each being present in centain percentages by weight. The oxide particles are metal oxides of the metal alloy composition, the amount of oxygen being present in certain percentages by weight and all the oxygen present being substantially in the form of metal oxide. This control rod is characterized by its high strength and resistance to creep at elevated temperatures.

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

    Energy Technology Data Exchange (ETDEWEB)

    Yavar, A.R. [School of Applied Physics, Faculty of Science and Technology, National University of Malaysia (UKM), 43600 Bangi, Selangor (Malaysia); Sarmani, S.B. [School of Chemical Sciences and Food Technology, Faculty of Science and Technology, National University of Malaysia (UKM), 43600 Bangi, Selangor (Malaysia); Wood, A.K. [Analytical Chemistry Application Group, Industrial Technology Division, Malaysian Nuclear Agency (MNA), Bangi, 43000 Kajang, Selangor (Malaysia); Fadzil, S.M. [School of Applied Physics, Faculty of Science and Technology, National University of Malaysia (UKM), 43600 Bangi, Selangor (Malaysia); Radir, M.H. [Analytical Chemistry Application Group, Industrial Technology Division, Malaysian Nuclear Agency (MNA), Bangi, 43000 Kajang, Selangor (Malaysia); Khoo, K.S., E-mail: khoo@ukm.m [School of Applied Physics, Faculty of Science and Technology, National University of Malaysia (UKM), 43600 Bangi, Selangor (Malaysia)

    2011-05-15

    Determination of thermal to fast neutron flux ratio (f{sub fast}) and fast neutron flux ({phi}{sub fast}) is required for fast neutron reactions, fast neutron activation analysis, and for correcting interference reactions. The f{sub fast} and subsequently {phi}{sub fast} were determined using the absolute method. The f{sub fast} ranged from 48 to 155, and the {phi}{sub fast} was found in the range 1.03x10{sup 10}-4.89x10{sup 10} n cm{sup -2} s{sup -1}. These values indicate an acceptable conformity and applicable for installation of the fast neutron facility at the MNA research reactor.

  17. Determination of the neutron flux in the reactor zones with the strong neutron absorption and leakage

    Directory of Open Access Journals (Sweden)

    Ljubenov Vladan

    2004-01-01

    Full Text Available The procedures for the numerical and experimental determination of the neutron flux in the zones with the strong neutron absorption and leakage are described in this paper. Numerical procedure is based on the application of the SCALE-4.4a code system where the Dancoff factors are determined by the VEGA2DAN code. Two main parts of the experimental methodology are measurement of the activity of irradiated foils and determination of the averaged neutron absorption cross-section in the foils by the SCALE-4.4a calculation procedure. The proposed procedures have been applied for the determination of the neutron flux in the internal neutron converter used with the RB reactor core configuration number 114.

  18. High heat flux loop heat pipes

    Science.gov (United States)

    North, Mark T.; Sarraf, David B.; Rosenfeld, John H.; Maidanik, Yuri F.; Vershinin, Sergey

    1997-01-01

    Loop Heat Pipes (LHPs) can transport very large thermal power loads, over long distances, through flexible, small diameter tubes and against high gravitational heads. While recent LHPs have transported as much as 1500 W, the peak heat flux through a LHP's evaporator has been limited to about 0.07 MW/m2. This limitation is due to the arrangement of vapor passages next to the heat load which is one of the conditions necessary to ensure self priming of the device. This paper describes work aimed at raising this limit by threefold to tenfold. Two approaches were pursued. One optimized the vapor passage geometry for the high heat flux conditions. The geometry improved the heat flow into the wick and working fluid. This approach also employed a finer pored wick to support higher vapor flow losses. The second approach used a bidisperse wick material within the circumferential vapor passages. The bidisperse material increased the thermal conductivity and the evaporative surface area in the region of highest heat flux, while providing a flow path for the vapor. Proof-of-concept devices were fabricated and tested for each approach. Both devices operated as designed and both demonstrated operation at a heat flux of 0.70 MW/m2. This performance exceeded the known state of the art by a factor of more than six for both conventional heat pipes and for loop heat pipes using ammonia. In addition, the bidisperse-wick device demonstrated boiling heat transfer coefficients up to 100,000 W/m2.K, and the fine pored device demonstrated an orientation independence with its performance essentially unaffected by whether its evaporator was positioned above, below or level with the condenser.

  19. Neutron flux parameters for k{sub 0}-NAA method at the Malaysian nuclear agency research reactor after core reconfiguration

    Energy Technology Data Exchange (ETDEWEB)

    Yavar, A.R. [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Sarmani, S. [School of Chemical Sciences and Food Technology, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Wood, A.K. [Analytical Chemistry Application Group, Industrial Technology Division, Malaysian Nuclear Agency (MNA), Bangi, Kajang, Selangor 43000 (Malaysia); Fadzil, S.M. [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Masood, Z. [Analytical Chemistry Application Group, Industrial Technology Division, Malaysian Nuclear Agency (MNA), Bangi, Kajang, Selangor 43000 (Malaysia); Khoo, K.S., E-mail: khoo@ukm.m [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia)

    2011-02-15

    The Malaysian Nuclear Agency (MNA) research reactor, commissioned in 1982, is a TRIGA Mark II swimming pool type reactor. When the core configuration changed in June 2009, it became essential to re-determine such neutron flux parameters as thermal to epithermal neutron flux ratio (f), epithermal neutron flux shape factor ({alpha}), thermal neutron flux ({phi}{sub th}) and epithermal neutron flux ({phi}{sub epi}) in the irradiation positions of MNA research reactor in order to guarantee accuracy in the application of k{sub 0}-neutron activation analysis (k{sub 0}-NAA).The f and {alpha} were determined using the bare bi-isotopic monitor and bare triple monitor methods, respectively; Au and Zr monitors were utilized in present study. The results for four irradiation positions are presented and discussed in the present work. The calculated values of f and {alpha} ranged from 33.49 to 47.33 and -0.07 to -0.14, respectively. The {phi}{sub th} and the {phi}{sub epi} were measured as 2.03 x 10{sup 12} (cm{sup -2} s{sup -1}) and 6.05 x 10{sup 10} (cm{sup -2} s{sup -1}) respectively. These results were compared to those of previous studies at this reactor as well as to those of reactors in other countries. The results indicate a good conformity with other findings.

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Youker, Amanda J., E-mail: youker@anl.gov [Chemical Sciences and Engineering Division, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States); Stepinski, Dominique C.; Maggos, Laura E.; Bakel, Allen J.; Vandegrift, George F. [Chemical Sciences and Engineering Division, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer GTRI program supports conversion from HEU to LEU. Black-Right-Pointing-Pointer High performance research reactors require a dense LEU fuel such as U-10Mo foils. Black-Right-Pointing-Pointer Dissolution conditions for U-10Mo foils in acidic media have been optimized. Black-Right-Pointing-Pointer Solvent-extraction processing can be used to recover U lost in fuel fabrication. Black-Right-Pointing-Pointer Flowsheets were developed using Argonne-design contactors but other contactors can be used as well. - Abstract: 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(NO{sub 3}){sub 3}. HNO{sub 3} and HF concentrations have been optimized for timely dissolution of the fuel scrap and prevention of the formation of the U-Zr{sub 2} intermetallic, explosive complex, while

  2. Argonne Tandem Linac Accelerator System (ATLAS)

    Data.gov (United States)

    Federal Laboratory Consortium — ATLAS is a national user facility at Argonne National Laboratory in Argonne, Illinois. The ATLAS facility is a leading facility for nuclear structure research in the...

  3. Fusion reactors-high temperature electrolysis (HTE)

    Energy Technology Data Exchange (ETDEWEB)

    Fillo, J.A. (ed.)

    1978-01-01

    Results of a study to identify and develop a reference design for synfuel production based on fusion reactors are given. The most promising option for hydrogen production was high-temperature electrolysis (HTE). The main findings of this study are: 1. HTE has the highest potential efficiency for production of synfuels from fusion; a fusion to hydrogen energy efficiency of about 70% appears possible with 1800/sup 0/C HTE units and 60% power cycle efficiency; an efficiency of about 50% possible with 1400/sup 0/C HTE units and 40% power cycle efficiency. 2. Relative to thermochemical or direct decomposition methods HTE technology is in a more advanced state of development, 3. Thermochemical or direct decomposition methods must have lower unit process or capital costs if they are to be more attractive than HTE. 4. While design efforts are required, HTE units offer the potential to be quickly run in reverse as fuel cells to produce electricity for restart of Tokamaks and/or provide spinning reserve for a grid system. 5. Because of the short timescale of the study, no detailed economic evaluation could be carried out.A comparison of costs could be made by employing certain assumptions. For example, if the fusion reactor-electrolyzer capital installation is $400/(KW(T) ($1000/KW(E) equivalent), the H/sub 2/ energy production cost for a high efficiency (about 70 %) fusion-HTE system is on the same order of magnitude as a coal based SNG plant based on 1976 dollars. 6. The present reference design indicates that a 2000 MW(th) fusion reactor could produce as much at 364 x 10/sup 6/ scf/day of hydrogen which is equivalent in heating value to 20,000 barrels/day of gasoline. This would fuel about 500,000 autos based on average driving patterns. 7. A factor of three reduction in coal feed (tons/day) could be achieved for syngas production if hydrogen from a fusion-HTE system were used to gasify coal, as compared to a conventional syngas plant using coal-derived hydrogen.

  4. High Performance Photocatalytic Oxidation Reactor System Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Pioneer Astronautics proposes a technology program for the development of an innovative photocatalytic oxidation reactor for the removal and mineralization of...

  5. Neutrino mass hierarchy and precision physics with medium-baseline reactors: Impact of energy-scale and flux-shape uncertainties

    Science.gov (United States)

    Capozzi, F.; Lisi, E.; Marrone, A.

    2015-11-01

    Nuclear reactors provide intense sources of electron antineutrinos, characterized by few-MeV energy E and unoscillated spectral shape Φ (E ). High-statistics observations of reactor neutrino oscillations over medium-baseline distances L ˜O (50 ) km would provide unprecedented opportunities to probe both the long-wavelength mass-mixing parameters (δ m2 and θ12) and the short-wavelength ones (Δ mee 2 and θ13), together with the subtle interference effects associated with the neutrino mass hierarchy (either normal or inverted). In a given experimental setting—here taken as in the JUNO project for definiteness—the achievable hierarchy sensitivity and parameter accuracy depend not only on the accumulated statistics but also on systematic uncertainties, which include (but are not limited to) the mass-mixing priors and the normalizations of signals and backgrounds. We examine, in addition, the effect of introducing smooth deformations of the detector energy scale, E →E'(E ), and of the reactor flux shape, Φ (E )→Φ'(E ), within reasonable error bands inspired by state-of-the-art estimates. It turns out that energy-scale and flux-shape systematics can noticeably affect the performance of a JUNO-like experiment, both on the hierarchy discrimination and on precision oscillation physics. It is shown that a significant reduction of the assumed energy-scale and flux-shape uncertainties (by, say, a factor of 2) would be highly beneficial to the physics program of medium-baseline reactor projects. Our results also shed some light on the role of the inverse-beta decay threshold, of geoneutrino backgrounds, and of matter effects in the analysis of future reactor oscillation data.

  6. Neutron flux mapping of Argonauta reactor in the new configuration of its reactor core; Mapeamento do fluxo de neutrons do reator Argonauta na nova configuracao do seu nucleo

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Maria Ines Silvani; Furieri, Rosanne Cefaly de Aranda Amado [Instituto de Engenharia Nuclear (IEN), Rio de Janeiro, RJ (Brazil)

    2000-07-01

    Whenever tasks involving the use of a nuclear reactor are carried out, e.g., radioisotope production, activation analysis, neutrongraphy, etc., it is necessary to know the magnitude of the associate neutron flux. The Argonauta reactor operating in Rio de Janeiro, at Instituto de Engenharia Nuclear - IEN/CNEN, was submitted to some modifications in its core, which made necessary to measure again its new neutronic characteristics, not only in the core itself, but also at the irradiation pads. In this type of research reactor, the neutrons are energetically distributed from values below 1 eV, to values reaching the magnitude of MeV. Therefore, depending on the kind of experiment to be conducted, it may become necessary to know the integrated neutron flux within certain energy ranges. In this work, the neutron flux for thermal and epithermal regions were determined by using the foil activation method. To accomplish this goal, two different techniques were applied. In the first technique {beta}-{gamma} gamma coincidence measurements were performed using a proportional 4{pi}{beta} gaseous detector and a NaI(Tl) scintillation detector, while in the second one, gamma spectroscopy was carried out using Hp-Ge and NaI(Tl) detectors. In both cases, the flux was computed using the FLUXO software, specially developed for this purpose. (author)

  7. Decision no. 2011-DC-0216 of the French nuclear safety authority from May 5, 2011, ordering the Laue Langevin Institute to proceed to a complementary safety evaluation of its basic nuclear facility (high flux reactor - INB no. 67) in the eyes of the Fukushima Daiichi nuclear power plant accident; Decision no. 2011-DC-0216 de l'Autorite de surete nucleaire du 5 mai 2011 prescrivant a l'Institut Laue Langevin (ILL) de proceder a une evaluation complementaire de la surete de son installation nucleaire de base (Reacteur a Haut Flux - INB n.67) au regard de l'accident survenu a la centrale nucleaire de Fukushima Daiichi

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-07-01

    As a consequence of the accident of the Fukushima Daiichi nuclear power plant (Japan), the French Prime Minister entrusted the French nuclear safety authority (ASN) with the mission to carry out a safety analysis re-evaluation of the French nuclear facilities, and in particular the nuclear power plants. A decision has been addressed by the ASN to each nuclear operator with the specifications of this safety re-evaluation analysis and the list of facilities in concern. This document is the decision addressed to the Laue Langevin Institute, operator of the high flux research reactor (RHF) of Grenoble (France). (J.S.)

  8. Critical Heat Flux Phenomena at HighPressure & Low Mass Fluxes: NEUP Final Report Part I: Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, Michael [Univ. of Wisconsin, Madison, WI (United States); Wu, Qiao [Oregon State Univ., Corvallis, OR (United States)

    2015-04-30

    This report is a preliminary document presenting an overview of the Critical Heat Flux (CHF) phenomenon, the High Pressure Critical Heat Flux facility (HPCHF), preliminary CHF data acquired, and the future direction of the research. The HPCHF facility has been designed and built to study CHF at high pressure and low mass flux ranges in a rod bundle prototypical of conceptual Small Modular Reactor (SMR) designs. The rod bundle is comprised of four electrically heated rods in a 2x2 square rod bundle with a prototypic chopped-cosine axial power profile and equipped with thermocouples at various axial and circumferential positions embedded in each rod for CHF detection. Experimental test parameters for CHF detection range from pressures of ~80 – 160 bar, mass fluxes of ~400 – 1500 kg/m2s, and inlet water subcooling from ~30 – 70°C. The preliminary data base established will be further extended in the future along with comparisons to existing CHF correlations, models, etc. whose application ranges may be applicable to the conditions of SMRs.

  9. Critical heat flux in natural convection cooled TRIGA reactors with hexagonal bundle

    Energy Technology Data Exchange (ETDEWEB)

    Yang, J.; Avery, M.; De Angelis, M.; Anderson, M.; Corradini, M. [Univ. of Wisconsin-Madison, 1500 Engineering Drive, Madison, WI 53706 (United States); Feldman, E. E.; Dunn, F. E.; Matos, J. E. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

    2012-07-01

    A three-rod bundle Critical Heat Flux (CHF) study at low flow, low pressure, and natural convection condition has been conducted, simulating TRIGA reactors with the hexagonally configured core. The test section is a custom-made trefoil shape tube with three identical fuel pin heater rods located symmetrically inside. The full scale fuel rod is electrically heated with a chopped-cosine axial power profile. CHF experiments were carried out with the following conditions: inlet water subcooling from 30 K to 95 K; pressure from 110 kPa to 230 kPa; mass flux up to 150 kg/m{sup 2}s. About 50 CHF data points were collected and compared with a few existing CHF correlations whose application ranges are close to the testing conditions. Some tests were performed with the forced convection to identify the potential difference between the CHF under the natural convection and forced convection. The relevance of the CHF to test parameters is investigated. (authors)

  10. High-energy fluxes of atmospheric neutrinos

    CERN Document Server

    Sinegovskaya, T S; Sinegovsky, S I

    2013-01-01

    High-energy neutrinos from decays of mesons, produced in collisions of cosmic ray particles with air nuclei, form unavoidable background for detection of astrophysical neutrinos. More precise calculations of the high-energy neutrino spectrum are required since measurements in the IceCube experiment reach the intriguing energy region where a contribution of the prompt neutrinos and/or astrophysical ones should be discovered. Basing on the referent hadronic models QGSJET II-03, SIBYLL 2.1, we calculate high-energy spectra, both of the muon and electron atmospheric neutrinos, averaged over zenith-angles. The computation is made using three parameterizations of cosmic ray spectra which include the knee region. All calculations are compared with the atmospheric neutrino measurements by Frejus and IceCube. The prompt neutrino flux predictions obtained with thequark-gluon string model (QGSM) for the charm production by Kaidalov & Piskunova do not contradict to the IceCube measurements and upper limit on the astr...

  11. Gammasphere activities at Argonne

    Energy Technology Data Exchange (ETDEWEB)

    Khoo, T.L.; Carpenter, M.; Ahmad, I. [and others

    1995-08-01

    A powerful third-generation national gamma-ray facility consisting of 110 Ge detectors with BGO Compton suppressors is being constructed at LBL. After 18 months of operation there it will move to another site. This detector system combines calorimetric and multiplicity information with the excellent energy resolution, large efficiency, and high granularity of the Ge detectors. The large number of Ge detectors is essential for high- ({>=} 3) fold coincidences. Since each additional fold results in roughly an order-of-magnitude improvement in selectivity, this feature makes it possible to isolate cleanly weak structures, where new physics will undoubtedly lie. Since Gammasphere represents a national facility, we have made substantial contributions in its construction. In addition, T. L. Khoo is the Chairman of the Gammasphere Scientific Advisory Committee (formerly Steering Committee) which follows, and provides advice on, the construction of Gammasphere, while R.V. F. Janssens is Chairman of the Users Executive Committee.

  12. High heat flux engineering in solar energy applications

    Energy Technology Data Exchange (ETDEWEB)

    Cameron, C.P.

    1993-07-01

    Solar thermal energy systems can produce heat fluxes in excess of 10,000 kW/m{sup 2}. This paper provides an introduction to the solar concentrators that produce high heat flux, the receivers that convert the flux into usable thermal energy, and the instrumentation systems used to measure flux in the solar environment. References are incorporated to direct the reader to detailed technical information.

  13. Anisotropic flux pinning in high Tc superconductors

    Science.gov (United States)

    Koleśnik, S.; Igalson, J.; Skośkiewicz, T.; Szymczak, R.; Baran, M.; Pytel, K.; Pytel, B.

    1995-02-01

    In this paper we present a comparison of the results of FC magnetization measurements on several PbSr(Y,Ca)CuO crystals representing various levels of flux pinning. The pinning centers in our crystals have been set up during the crystal growth process or introduced by neutron irradiation. Some possible explanations of the observed effects, including surface barrier, flux-center distribution and sample-shape effects, are discussed.

  14. Core Physics of Pebble Bed High Temperature Nuclear Reactors

    NARCIS (Netherlands)

    Auwerda, G.J.

    2014-01-01

    To more accurately predict the temperature distribution inside the reactor core of pebble bed type high temperature reactors, in this thesis we investigated the stochastic properties of randomly stacked beds and the effects of the non-homogeneity of these beds on the neutronics and thermal-hydraulic

  15. Tritium permeation behavior through pyrolytic carbon in tritium production using high-temperature gas-cooled reactor for fusion reactors

    Directory of Open Access Journals (Sweden)

    H. Ushida

    2016-12-01

    Full Text Available Under tritium production method using a high-temperature gas-cooled reactor loaded Li compound, Li compound has to be coated by ceramic materials in order to suppress the spreading of tritium to the whole reactor. Pyrolytic carbon (PyC is a candidate of the coating material because of its high resistance for gas permeation. In this study, hydrogen permeation experiments using a PyC-coated isotropic graphite tube were conducted and hydrogen diffusivity, solubility and permeability were evaluated. Tritium permeation behavior through PyC-coated Li compound particles was simulated by using obtained data. Hydrogen permeation flux through PyC in a steady state is proportional to the hydrogen pressure and is larger than that through Al2O3 which is also candidate coating material. However, total tritium leak within the supposed reactor operation period through the PyC-coated Li compound particles is lower than that through the Al2O3-coated ones because the hydrogen absorption capacity in PyC is considerably larger than that in Al2O3.

  16. The High Energy Neutrino Nuisance at a Medium Baseline Reactor Experiment

    CERN Document Server

    Ciuffoli, Emilio; Zhang, Xinmin

    2012-01-01

    10 years from now medium baseline reactor experiments will attempt to determine the neutrino mass hierarchy from the differences (RL+PV) between the extrema of the Fourier transformed neutrino spectra. Recently Qian et al. have claimed that this goal may be impeded by the strong dependence of the difference parameter RL+PV on the reactor neutrino flux and on slight variations of Delta M^2_32. We demonstrate that this effect results from a spurious dependence of the difference parameter on the very high energy (8+ MeV) tail of the reactor neutrino spectrum. This dependence is spurious because the high energy tail depends upon decays of exotic isotopes and is insensitive to the mass hierarchy. An energy-dependent weight in the Fourier transform not only eliminates this spurious dependence but in fact increases the chance of correctly determining the hierarchy.

  17. Status of RF superconductivity at Argonne

    Energy Technology Data Exchange (ETDEWEB)

    Shepard, K.W.

    1989-01-01

    Development of a superconducting (SC) slow-wave structures began at Argonne National Laboratory (ANL) in 1971, and led to the first SC heavy-ion linac (ATLAS - the Argonne Tandem-Linac Accelerating System), which began regularly scheduled operation in 1978. To date, more than 40,000 hours of bean-on target operating time has been accumulated with ATLAS. The Physics Division at ANL has continued to develop SC RF technology for accelerating heavy-ions, with the result that the SC linac has, up to the present, has been in an almost continuous process of upgrade and expansion. It should be noted that this has been accomplished while at the same time maintaining a vigorous operating schedule in support of the nuclear and atomic physics research programs of the division. In 1987, the Engineering Physics Division at ANL began development of SC RF components for the acceleration of high-brightness proton and deuterium beams. This work has included the evaluation of RF properties of high-{Tc} oxide superconductors, both for the above and for other applications. The two divisions collaborated while they worked on several applications of RF SC, and also worked to develop the technology generally. 11 refs., 6 figs.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    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.

  20. Final safeguards analysis, high temperature lattice test reactor. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Hanthorn, H.E.; Brown, W.W.; Clark, R.G.; Heineman, R.E.; Humes, R.M.

    1966-01-01

    The PMACS `reactor-normal` signal signifies that important process variables do not exceed their set points, that various interlocks are properly set, that functional tests of the computer operation are satisfactory, and that the reactor flux level and period derived from two additional, independent, and dissimilar channels are within set limits. This safety circuit combines the features of redundancy, dissimilar components, and frequent testing which are required for best reliability. The experimental equipment auxiliary to the reactor includes two oscillator mechanisms, one to move the test cell or the adjoining cell into and out of position, the other to move small specimens in the test cell or adjoining cells. They have cooling chambers for the removal of specimens from the test cell without the necessity of cooling the reactor. A neutron chopper and time-of-flight spectrometer are provided; the neutron detectors, at the end of a 25-meter flight tube, are in an adjoining small building. Test cores may be assembled on a core dolly have a load capacity of 14,000 lb. Two wire traverse mechanisms are provided for measurements of flux distribution.

  1. Argonne National Laboratory 1985 publications

    Energy Technology Data Exchange (ETDEWEB)

    Kopta, J.A. (ED.); Hale, M.R. (comp.)

    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.

  2. Flux pinning characteristics and irreversibility line in high temperature superconductors

    Science.gov (United States)

    Matsushita, T.; Ihara, N.; Kiuchi, M.

    1995-01-01

    The flux pinning properties in high temperature superconductors are strongly influenced by thermally activated flux motion. The scaling relation of the pinning force density and the irreversibility line in various high temperature superconductors are numerically analyzed in terms of the flux creep model. The effect of two factors, i.e., the flux pinning strength and the dimensionality of the material, on these properties are investigated. It is speculated that the irreversibility line in Bi-2212 superconductors is one order of magnitude smaller than that in Y-123, even if the flux pinning strength in Bi-2212 is improved up to the level of Y-123. It is concluded that these two factors are equally important in determination of the flux pinning characteristics at high temperatures.

  3. High power density reactors based on direct cooled particle beds

    Science.gov (United States)

    Powell, J. R.; Horn, F. L.

    Reactors based on direct cooled High Temperature Gas Cooled Reactor (HTGR) type particle fuel are described. The small diameter particle fuel is packed between concentric porous cylinders to make annular fuel elements, with the inlet coolant gas flowing inwards. Hot exit gas flows out along the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBRs) are extremely high resulting in compact, lightweight systems. Coolant exit temperatures are high, because of the ceramic fuel temperature capabilities, and the reactors can be ramped to full power and temperature very rapidly. PBR systems can generate very high burst power levels using open cycle hydrogen coolant, or high continuous powers using closed cycle helium coolant. PBR technology is described and development requirements assessed.

  4. 2015 Annual Report - Argonne Leadership Computing Facility

    Energy Technology Data Exchange (ETDEWEB)

    Collins, James R. [Argonne National Lab. (ANL), Argonne, IL (United States); Papka, Michael E. [Argonne National Lab. (ANL), Argonne, IL (United States); Cerny, Beth A. [Argonne National Lab. (ANL), Argonne, IL (United States); Coffey, Richard M. [Argonne National Lab. (ANL), Argonne, IL (United States)

    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.

  5. 2014 Annual Report - Argonne Leadership Computing Facility

    Energy Technology Data Exchange (ETDEWEB)

    Collins, James R. [Argonne National Lab. (ANL), Argonne, IL (United States); Papka, Michael E. [Argonne National Lab. (ANL), Argonne, IL (United States); Cerny, Beth A. [Argonne National Lab. (ANL), Argonne, IL (United States); Coffey, Richard M. [Argonne National Lab. (ANL), Argonne, IL (United States)

    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.

  6. High-temperature reactor developments in the Netherlands

    Energy Technology Data Exchange (ETDEWEB)

    Schram, R.P.C.; Cordfunke, E.H.P.; Heek, A.I. van

    1996-01-01

    The high-temperature reactor development in the Netherland is embedded in the WHITE reactor program, in which several Dutch research institutes and engineering companies participate. The activities within the WHITE program are focused on the development of a small scale HTS for combined heat and power generation. In 1995, design choices for a pebble bed reactor were made at ECN. The first concept HTR will gave a closed cycle helium turbine and a power level of 40 MWth. It is intended to make the market introduction of a commercially competitive HTR feasible. The design will be an optimization of the Peu-a-Peu (PAP) concept of KFA Juelich. Computer codes necessary for the evaluation of reactor physics aspects of this reactor are developed in cooperation with international partners. An evaluation of a 20 MWth PAP concept showed that the maximum fuel termmperature after depressurization does not exceed 1300 C. (orig.).

  7. Measurement of in-phantom neutron flux and gamma dose in Tehran research reactor boron neutron capture therapy beam line

    OpenAIRE

    Elham Bavarnegin; Alireza Sadremomtaz; Hossein Khalafi; Yaser Kasesaz

    2016-01-01

    Aim: Determination of in-phantom quality factors of Tehran research reactor (TRR) boron neutron capture therapy (BNCT) beam. Materials and Methods: The doses from thermal neutron reactions with 14N and 10B are calculated by kinetic energy released per unit mass approach, after measuring thermal neutron flux using neutron activation technique. Gamma dose is measured using TLD-700 dosimeter. Results: Different dose components have been measured in a head phantom which has been designed an...

  8. Periods of High Intensity Solar Proton Flux

    Science.gov (United States)

    Xapsos, Michael A.; Stauffer, Craig A.; Jordan, Thomas M.; Adams, James H.; Dietrich, William F.

    2012-01-01

    Analysis is presented for times during a space mission that specified solar proton flux levels are exceeded. This includes both total time and continuous time periods during missions. Results for the solar maximum and solar minimum phases of the solar cycle are presented and compared for a broad range of proton energies and shielding levels. This type of approach is more amenable to reliability analysis for spacecraft systems and instrumentation than standard statistical models.

  9. FASTER test reactor preconceptual design report summary

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, Steven [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-02-29

    The FASTER reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  10. FASTER Test Reactor Preconceptual Design Report

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, S. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-03-31

    The FASTER test reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  11. Thermal evaluation of uranium silicide miniplates irradiated at high heat flux

    Energy Technology Data Exchange (ETDEWEB)

    Post Guillen, Donna, E-mail: Donna.Guillen@inl.gov [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3710 (United States)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Best estimate of thermal conditions during irradiation experiment. Black-Right-Pointing-Pointer Thermal evaluation of 25% enriched, high-density U{sub 3}Si{sub 2}/Al dispersion fuel miniplates. Black-Right-Pointing-Pointer Predictions of heat flux and temperature for as-run, high heat flux conditions. Black-Right-Pointing-Pointer Finite-element analysis uses measured values of hydroxide layer thickness. - Abstract: The Gas Test Loop (GTL)-1 irradiation experiment was conducted in the Advanced Test Reactor (ATR) to assess corrosion performance of proposed booster fuel at heat flux levels {approx}30% above the design operating condition. Sixteen miniplates fabricated from 25% enriched, high-density (4.8 g U/cm{sup 3}) U{sub 3}Si{sub 2}/Al dispersion fuel with 6061 aluminum cladding were subjected to peak beginning of cycle (BOC) heat fluxes ranging from 411 to 593 W/cm{sup 2}. No adverse impacts to the miniplates were observed at these high heat flux levels. A detailed finite element model was constructed to calculate temperatures and heat flux for an as-run cycle average effective ATR south lobe power of 25.4 MW(t). Miniplate heat flux levels and fuel, cladding, hydroxide, and coolant-hydroxide interface temperatures were calculated using the average hydroxide thickness on each miniplate measured during post-irradiation examination. The purpose of this study was to obtain a best estimate of the as-run experiment temperatures to aid in establishing acceptable heat flux levels and designing fuel qualification experiments for this fuel type.

  12. Thermal Hydraulics of the Very High Temperature Gas Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh; Eung Kim; Richard Schultz; Mike Patterson; Davie Petti

    2009-10-01

    The U.S Department of Energy (DOE) is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R&D) that will be critical to the success of the NGNP, primarily in the areas of: • High temperature gas reactor fuels behavior • High temperature materials qualification • Design methods development and validation • Hydrogen production technologies • Energy conversion. This paper presents current R&D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs.

  13. The Jules Horowitz Reactor - A new High Performance European Material Testing Reactor open to International Users Present Status and Objectives

    Energy Technology Data Exchange (ETDEWEB)

    Iracane, Daniel; Bignan, Gilles [CEA Atomic Energy Commission Saclay Batiment 121- 91191 Gif Sur Yvette (France); Lindbaeck, Jan-Erik; Blomgren, Jan [VATTENFALL AB Nuclear Power Jaemtlandsgatan 99 SE-16287 Stockholm (Sweden)

    2010-07-01

    The development of sustainable nuclear energy requires R and D on fuel and material behaviour under irradiation with a high level of performance in order to meet the needs and challenges for the benefit of industry, research and public bodies. These stakes require a sustainable and secured access to an up-to-date high performance Material Testing Reactor. Following a broad survey within the European Research Area, the international community agreed that the need for Material Test Reactors in support of nuclear power plant safety and operation will continue in the context of sustainable nuclear energy. The Jules Horowitz Reactor project (JHR) copes with this context. JHR is designed as a user facility addressing the needs of the international community. This means: - flexibility with irradiation loops able to reproduce a large variation in operation conditions of different power reactor technologies, - high flux capacity to address Generations II, III, and IV needs. JHR is designed, built and operated as an international user facility because: - Given the maturity and globalization of the industry, domestic tools have no more the required level of economic and technical efficiency. Meanwhile, countries with nuclear energy need an access to high performance irradiation experimental capabilities to support technical skill and guarantee the competitiveness and safety of nuclear energy. - Many research items related to safety or public policy (waste management, etc.) require international cooperation to share costs and benefits of resulting consensus. JHR design is optimised for offering high performance material and fuel irradiation capability for the coming decades. This project is driven and funded by an international consortium gathering vendors, utilities and public stakeholders. This consortium has been set up in March 2007 when the construction began. The construction is in progress and the start of operation is scheduled for 2014. The JHR is a research

  14. Argonne National Laboratory 1986 publications

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. Total absorption spectroscopy study of $^{92}$Rb decay: a major contributor to reactor antineutrino flux

    CERN Document Server

    Zakari-Issoufou, A -A; Porta, A; Algora, A; Tain, J L; Valencia, E; Rice, S; Bui, V M; Cormon, S; Estienne, M; Agramunt, J; Äystö, J; Bowry, M; Briz, J A; Caballero-Folch, R; Cano-Ott, D; Cucoanes, A; Elomaa, V -V; Eronen, T; Estévez, E; Farrelly, G F; Garcia, A R; Gelletly, W; Gomez-Hornillos, M B; Gorlychev, V; Hakala, J; Jokinen, A; Jordan, M D; Kankainen, A; Karvonen, P; Kolhinen, V S; Kondev, F G; Martinez, T; Mendoza, E; Molina, F; Moore, I; Perez, A; Podolyák, Zs; Penttilä, H; Regan, P H; Reponen, M; Rissanen, J; Rubio, B; Shiba, T; Sonzogni, A A; Weber, C

    2015-01-01

    The antineutrino spectra measured in recent experiments at reactors are inconsistent with calculations based on the conversion of integral beta spectra recorded at the ILL reactor. $^{92}$Rb makes the dominant contribution to the reactor spectrum in the 5-8 MeV range but its decay properties are in question. We have studied $^{92}$Rb decay with total absorption spectroscopy. Previously unobserved beta feeding was seen in the 4.5-5.5 region and the GS to GS feeding was found to be 87.5(25)%. The impact on the reactor antineutrino spectra calculated with the summation method is shown and discussed.

  16. High-Temperature Gas-Cooled Test Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Laboratory; Bayless, Paul David [Idaho National Laboratory; Nelson, Lee Orville [Idaho National Laboratory; Gougar, Hans David [Idaho National Laboratory; Kinsey, James Carl [Idaho National Laboratory; Strydom, Gerhard [Idaho National Laboratory; Kumar, Akansha [Idaho National Laboratory

    2016-04-01

    A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

  17. Proliferation resistance assessment of high temperature gas reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chikamatsu N, M. A. [Instituto Tecnologico y de Estudios Superiores de Monterrey, Campus Santa Fe, Av. Carlos Lazo No. 100, Santa Fe, 01389 Mexico D. F. (Mexico); Puente E, F., E-mail: midori.chika@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    The Generation IV International Forum has established different objectives for the new generation of reactors to accomplish. These objectives are focused on sustain ability, safety, economics and proliferation resistance. This paper is focused on how the proliferation resistance of the High Temperature Gas Reactors (HTGR) is assessed and the advantages that these reactors present currently. In this paper, the focus will be on explaining why such reactors, HTGR, can achieve the goals established by the GIF and can present a viable option in terms of proliferation resistance, which is an issue of great importance in the field of nuclear energy generation. The reason why the HTGR are being targeted in this writing is that these reactors are versatile, and present different options from modular reactors to reactors with the same size as the ones that are being operated today. Besides their versatility, the HTGR has designed features that might improve on the overall sustain ability of the nuclear reactors. This is because the type of safety features and materials that are used open up options for industrial processes to be carried out; cogeneration for instance. There is a small section that mentions how HTGR s are being developed in the international sector in order to present the current world view in this type of technology and the further developments that are being sought. For the proliferation resistance section, the focus is on both the intrinsic and the extrinsic features of the nuclear systems. The paper presents a comparison between the features of Light Water Reactors (LWR) and the HTGR in order to be able to properly compare the most used technology today and one that is gaining international interest. (Author)

  18. Measurement of neutron flux and beam divergence at the cold neutron guide system of the new Munich research reactor FRM-II

    Science.gov (United States)

    Zeitelhack, K.; Schanzer, C.; Kastenmüller, A.; Röhrmoser, A.; Daniel, C.; Franke, J.; Gutsmiedl, E.; Kudryashov, V.; Maier, D.; Päthe, D.; Petry, W.; Schöffel, T.; Schreckenbach, K.; Urban, A.; Wildgruber, U.

    2006-05-01

    A sophisticated neutron guide system has been installed at the new Munich neutron source FRM-II to transport neutrons from the D 2 cold neutron source to several instruments, which are situated in a separate neutron guide hall. The guide system takes advantage of supermirror coatings and includes a worldwide unique "twisted" guide for a desired phase space transformation of the neutron beam. During the initial reactor commissioning in summer 2004, the integral and differential neutron flux as well as the distribution of beam divergence at the exit of two representative and the twisted neutron guide were measured using time-of-flight spectroscopy and gold-foil activation. The experimental results can be compared to extensive simulation calculations based on MCNP and McStas. The investigated guides fulfill the expectations of providing high neutron fluxes and reveal good quality with respect to the reflective coatings and the installation precision.

  19. Measurement of neutron flux and beam divergence at the cold neutron guide system of the new Munich research reactor FRM-II

    Energy Technology Data Exchange (ETDEWEB)

    Zeitelhack, K. [ZWE FRM-II, TU Muenchen, D-85747 Garching (Germany)]. E-mail: karl.zeitelhack@frm2.tum.de; Schanzer, C. [Physik-Department E21, TU Muenchen, D-85747 Garching (Germany); Kastenmueller, A. [ZWE FRM-II, TU Muenchen, D-85747 Garching (Germany); Roehrmoser, A. [ZWE FRM-II, TU Muenchen, D-85747 Garching (Germany); Daniel, C. [Physik-Department E22, TU Muenchen, D-85747 Garching (Germany); Franke, J. [Max-Planck-Institut fuer Metallforschung, D-70569 Stuttgart (Germany); Gutsmiedl, E. [ZWE FRM-II, TU Muenchen, D-85747 Garching (Germany); Kudryashov, V. [GKSS Forschungszentrum GmbH, D-21502 Geesthacht (Germany); Maier, D. [ZWE FRM-II, TU Muenchen, D-85747 Garching (Germany); Paethe, D. [ZWE FRM-II, TU Muenchen, D-85747 Garching (Germany); Petry, W. [ZWE FRM-II, TU Muenchen, D-85747 Garching (Germany); Schoeffel, T. [ZWE FRM-II, TU Muenchen, D-85747 Garching (Germany); Schreckenbach, K. [ZWE FRM-II, TU Muenchen, D-85747 Garching (Germany); Urban, A. [ZWE FRM-II, TU Muenchen, D-85747 Garching (Germany); Wildgruber, U. [Max-Planck-Institut fuer Metallforschung, D-70569 Stuttgart (Germany)

    2006-05-10

    A sophisticated neutron guide system has been installed at the new Munich neutron source FRM-II to transport neutrons from the D{sub 2} cold neutron source to several instruments, which are situated in a separate neutron guide hall. The guide system takes advantage of supermirror coatings and includes a worldwide unique 'twisted' guide for a desired phase space transformation of the neutron beam. During the initial reactor commissioning in summer 2004, the integral and differential neutron flux as well as the distribution of beam divergence at the exit of two representative and the twisted neutron guide were measured using time-of-flight spectroscopy and gold-foil activation. The experimental results can be compared to extensive simulation calculations based on MCNP and McStas. The investigated guides fulfill the expectations of providing high neutron fluxes and reveal good quality with respect to the reflective coatings and the installation precision.

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

    Energy Technology Data Exchange (ETDEWEB)

    Heeger, Karsten M. [Yale Univ., New Haven, CT (United States)

    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.

  1. Reactor target from metal chromium for "pure" high-intensive artificial neutrino source

    Science.gov (United States)

    Gavrin, V. N.; Kozlova, Yu. P.; Veretenkin, E. P.; Logachev, A. V.; Logacheva, A. I.; Lednev, I. S.; Okunkova, A. A.

    2017-01-01

    The paper presents the first results of development of manufacturing technology of metallic chromium targets from highly enriched isotope 50Cr for irradiation in a high flux nuclear reactor to obtain a compact high intensity neutrino source with low content of radionuclide impurities and minimum losses of enriched isotope. The main technological stages are the hydrolysis of chromyl fluoride, the electrochemical reduction of metallic chromium, the hot isostatic pressing of chromium powder and the electrical discharge machining of chromium bars. The technological stages of hot isostatic pressing of chromium powder and of electrical discharge machining of Cr rods have been tested.

  2. Development and modelling of fission chambers designed for high neutron fluxes: applications at the HFR reactor (ILL) and the MEGAPIE target (PSI); Developpement et modelisation de chambres a fission pour les hauts flux, mise en application au RHF (ILL) et a MEGAPIE (PSI)

    Energy Technology Data Exchange (ETDEWEB)

    Chabod, S

    2006-11-15

    The international project MEGAPIE (MEGAwatt PIlot Experiment) at the Paul Scherrer Institute aims to build and operate the first 1 MW liquid lead-bismuth spallation target. This work is dedicated to the characterization of the neutron flux and the actinide incineration potential of the target. This mission has required the development of an innovating neutron detector (DNM) made of 8 micro fission chambers, installed inside the central rod of the MEGAPIE target. The combination of uranium chambers with chambers without deposit allows an efficient compensation of the gamma radiation background. The optimisation and development work on the MEGAPIE chambers have enabled us to measure the {sigma}{sub f} * {phi} product at each level of the DNM with an uncertainty of less than 3 per cent. We have inferred from these data the value of the epithermal neutron flux (E > 1 eV) at 37 cm away from the window: 3.4*10{sup 13} n.cm{sup -2}.s{sup -1}, and the values of the neutron flux at 50, 60 and 74 cm: 1.2*10{sup 13}, 7.9*10{sup 12} and 3.9*10{sup 12} n.cm{sup -2}.s{sup -1} respectively. All these values are notably less important than those obtained from MCNPX simulations. Thermocouples installed in DMN have enabled us to know the temperature distribution inside the target. For a beam intensity of 1.2 mA, the temperature ranges from 360 to 420 Celsius degrees in the low part of the central rod. The thermal inertia of the system composed of the central rod and DNM has been assessed for brutal changes of the beam intensity and is worth about 60 s. (A.C.)

  3. Argonne Laboratory Computing Resource Center - FY2004 Report.

    Energy Technology Data Exchange (ETDEWEB)

    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.

  4. High Torque Density Transverse Flux Machine without the Need to Use SMC Material for 3D Flux Paths

    DEFF Research Database (Denmark)

    Lu, Kaiyuan; Wu, Weimin

    2015-01-01

    This paper presents a new transverse flux permanent magnet machine. In a normal transverse flux machine, complicated 3-D flux paths often exist. Such 3-D flux paths would require the use of soft magnetic composites material instead of laminations for construction of the machine stator. In the new...... machine topology proposed in this paper, by advantageously utilizing the magnetic flux path provided by an additional rotor, use of laminations that allow 2-D flux paths only will be sufficient to accomplish the required 3-D flux paths. The machine also has a high torque density and is therefore...

  5. Relation of middle molecules levels and oxidative stress to erythropoietin requirements in high-flux versus low-flux hemodialysis

    Directory of Open Access Journals (Sweden)

    Hala S El-Wakil

    2013-01-01

    Full Text Available The objective of this study is to investigate the serum beta-2-microglobulin (B2MG and advanced oxidation protein products (AOPP as middle molecule uremic toxins and protein carbonyl (PCO as oxidative stress marker in uremic patients undergoing high-flux versus low-flux hemodialysis (HD and to correlate their levels to the erythropoietin requirements for those patients. Twenty patients on chronic low-flux HD were recruited in the study. At the start of the study, all patients underwent high-flux HD for eight weeks, followed by low-flux HD for two weeks as a washout period. The patients were then subjected to another eight weeks of low-flux HD. Blood samples were obtained at the beginning and at the end of the high-flux period and the low-flux period. The mean erythropoietin dose for patients using high-flux HD was significantly lower than that for low-flux HD (P = 0.0062. Post-high flux, the B2MG and PCO levels were significantly lower than the pre-high-flux levels (P = 0.026 and 0.0005, respectively, but no significant change was observed in AOPP (P = 0.68. Post-low flux, the B2MG, AOPP and PCO were significantly higher than the pre-low-flux levels (P = 0.0002, 0.021 and <0.0001, respectively. Post-low flux, the B2MG and PCO were significantly higher than the post-high-flux levels (P <0.0001, but no significant difference was observed in AOPP (P = 0.11. High-flux HD results in reduction of some of the middle molecule toxins and PCO levels better than low-flux HD, and is associated with a better response to erythropoietin.

  6. High heat flux transport by microbubble emission boiling

    Science.gov (United States)

    Suzuki, Koichi

    2007-10-01

    In highly subcooled flow boiling, coalescing bubbles on the heating surface collapse to many microbubbles in the beginning of transition boiling and the heat flux increases higher than the ordinary critical heat flux. This phenomenon is called Microbubble Emission Boiling, MEB. It is generated in subcooled flow boiling and the maximum heat flux reaches about 1 kW/cm2(10 MW/m2) at liquid subcooling of 40 K and liquid velocity of 0.5 m/s for a small heating surface of 10 mm×10 mm which is placed at the bottom surface of horizontal rectangular channel. The high pressure in the channel is observed at collapse of the coalescing bubbles and it is closely related the size of coalescing bubbles. Periodic pressure waves are observed in MEB and the heat flux increases linearly in proportion to the pressure frequency. The frequency is considered the frequency of liquid-solid exchange on the heating surface. For the large sized heating surface of 50 mm length×20 mm width, the maximum heat flux obtained is 500 W/cm2 (5 MW/m2) at liquid subcooling of 40 K and liquid velocity of 0.5 m/s. This is considerably higher heat flux than the conventional cooling limit in power electronics. It is difficult to remove the high heat flux by MEB for a longer heating surface than 50 mm by single channel type. A model of advanced cooling device is introduced for power electronics by subcooled flow boiling with impinging jets. Themaxumum cooling heat flux is 500 W/cm2 (5 MW/m2). Microbubble emission boiling is useful for a high heat flux transport technology in future power electronics used in a fuel-cell power plant and a space facility.

  7. The Status of the US High-Temperature Gas Reactors

    Directory of Open Access Journals (Sweden)

    Andrew C. Kadak

    2016-03-01

    Full Text Available In 2005, the US passed the Energy Policy Act of 2005 mandating the construction and operation of a high-temperature gas reactor (HTGR by 2021. This law was passed after a multiyear study by national experts on what future nuclear technologies should be developed. As a result of the Act, the US Congress chose to develop the so-called Next-Generation Nuclear Plant, which was to be an HTGR designed to produce process heat for hydrogen production. Despite high hopes and expectations, the current status is that high temperature reactors have been relegated to completing research programs on advanced fuels, graphite and materials with no plans to build a demonstration plant as required by the US Congress in 2005. There are many reasons behind this diminution of HTGR development, including but not limited to insufficient government funding requirements for research, unrealistically high temperature requirements for the reactor, the delay in the need for a “hydrogen” economy, competition from light water small modular light water reactors, little utility interest in new technologies, very low natural gas prices in the US, and a challenging licensing process in the US for non-water reactors.

  8. Liquid-phase catalytic reactor combined with measurement of hot electron flux and chemiluminescence

    Science.gov (United States)

    Nedrygailov, Ievgen I.; Lee, Changhwan; Moon, Song Yi; Lee, Hyosun; Park, Jeong Young

    2016-11-01

    Understanding the role of electronically nonadiabatic interactions during chemical reactions on metal surfaces in liquid media is of great importance for a variety of applications including catalysis, electrochemistry, and environmental science. Here, we report the design of an experimental apparatus for detection of the highly excited (hot) electrons created as a result of nonadiabatic energy transfer during the catalytic decomposition of hydrogen peroxide on thin-film metal-semiconductor nanodiodes. The apparatus enables the measurement of hot electron flows and related phenomena (e.g., surface chemiluminescence) as well as the corresponding reaction rates at different temperatures. The products of the chemical reaction can be characterized in the gaseous phase by means of gas chromatography. The combined measurement of hot electron flux, catalytic activity, and light emission can lead to a fundamental understanding of the elementary processes occurring during the heterogeneous catalytic reaction.

  9. Generation IV Reactors Integrated Materials Technology Program Plan: Focus on Very High Temperature Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, William R [ORNL; Burchell, Timothy D [ORNL; Katoh, Yutai [ORNL; McGreevy, Timothy E [ORNL; Nanstad, Randy K [ORNL; Ren, Weiju [ORNL; Snead, Lance Lewis [ORNL; Wilson, Dane F [ORNL

    2008-08-01

    Since 2002, the Department of Energy's (DOE's) Generation IV Nuclear Energy Systems (Gen IV) Program has addressed the research and development (R&D) necessary to support next-generation nuclear energy systems. The six most promising systems identified for next-generation nuclear energy are described within this roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor-SCWR and the Very High Temperature Reactor-VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor-GFR, the Lead-cooled Fast Reactor-LFR, and the Sodium-cooled Fast Reactor-SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides and may provide an alternative to accelerator-driven systems. At the inception of DOE's Gen IV program, it was decided to significantly pursue five of the six concepts identified in the Gen IV roadmap to determine which of them was most appropriate to meet the needs of future U.S. nuclear power generation. In particular, evaluation of the highly efficient thermal SCWR and VHTR reactors was initiated primarily for energy production, and evaluation of the three fast reactor concepts, SFR, LFR, and GFR, was begun to assess viability for both energy production and their potential contribution to closing the fuel cycle. Within the Gen IV Program itself, only the VHTR class of reactors was selected for continued development. Hence, this document will address the multiple activities under the Gen IV program that contribute to the development of the VHTR. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of

  10. High-irradiance reactors with unfolded aplanatic optics.

    Science.gov (United States)

    Feuermann, Daniel; Gordon, Jeffrey M

    2008-11-01

    Reconstituting the intense irradiance of short-arc discharge lamps at a remote target, at high radiative efficiency, represents a central challenge in the design of high-temperature chemical reactors, heightened by the need for high numerical aperture at both the target and the source. Separating the optical system from both the source and the reactor allows pragmatic operation, monitoring, and control. We explore near-field unfolded aplanats as feasible solutions and report measurements for a prototype that constitutes a double-ellipsoid mirror. We also propose compound unfolded aplanats that collect lamp emission over all angles (in lieu of light recycling optics) and irradiate the reactor over nearly its full circumference.

  11. Reactor G1: high power experiments; Experiences a forte puissance

    Energy Technology Data Exchange (ETDEWEB)

    Laage, F. de; Teste du Baillet, A.; Veyssiere, A.; Wanner, G. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Retel, H. [Societe Rateau, D.E.A. (France)

    1957-07-01

    The experiments carried out in the starting-up programme of the reactor G1 comprised a series of tests at high power, which allowed the following points to be studied: 1- Effect of poisoning by Xenon (absolute value, evolution). 2- Temperature coefficients of the uranium and graphite for a temperature distribution corresponding to heating by fission. 3- Effect of the pressure (due to the coiling system) on the reactivity. 4- Calibration of the security rods as a function of their position in the pile (1). 5- Temperature distribution of the graphite, the sheathing, the uranium and the air leaving the canals, in a pile running normally at high power. 6- Neutron flux distribution in a pile running normally at high power. 7- Determination of the power by nuclear and thermodynamic methods. These experiments have been carried out under two very different pile conditions. From the 1. to the 15. of August 1956, a series of power increases, followed by periods of stabilisation, were induced in a pile containing uranium only, in 457 canals, amounting to about 34 tons of fuel. A knowledge of the efficiency of the control rods in such a pile has made it possible to measure with good accuracy the principal effects at high temperatures, that is, to deal with points 1, 2, 3, 5. Flux charts giving information on the variations of the material Laplacian and extrapolation lengths in the reflector have been drawn up. Finally the thermodynamic power has been measured under good conditions, in spite of some installation difficulties. On September 16, the pile had its final charge of 100 tons. All the canals were loaded, 1,234 with uranium and 53 (i.e. exactly 4 per cent of the total number) with thorium uniformly distributed in a square lattice of 100 cm side. Since technical difficulties prevented the calibration of the control rods, the measurements were limited to the determination of the thermodynamic power and the temperature distributions (points 5 and 7). This report will

  12. Investigating the use of nanofluids to improve high heat flux cooling systems

    CERN Document Server

    Barrett, T R; Flinders, K; Sergis, A; Hardalupas, Y

    2013-01-01

    The thermal performance of high heat flux components in a fusion reactor could be enhanced significantly by the use of nanofluid coolants, suspensions of a liquid with low concentrations of solid nanoparticles. However, before they are considered viable for fusion, the long-term behaviour of nanofluids must be investigated. This paper reports an experiment which is being prepared to provide data on nanofluid stability, settling and erosion in a HyperVapotron device. Procedures are demonstrated for nanofluid synthesis and quality assessment, and the fluid sample analysis methods are described. The end results from this long-running experiment are expected to allow an initial assessment of the suitability of nanofluids as coolants in a fusion reactor.

  13. High precision photon flux determination for photon tagging experiments

    Science.gov (United States)

    Teymurazyan, A.; Ahmidouch, A.; Ambrozewicz, P.; Asratyan, A.; Baker, K.; Benton, L.; Burkert, V.; Clinton, E.; Cole, P.; Collins, P.; Dale, D.; Danagoulian, S.; Davidenko, G.; Demirchyan, R.; Deur, A.; Dolgolenko, A.; Dzyubenko, G.; Ent, R.; Evdokimov, A.; Feng, J.; Gabrielyan, M.; Gan, L.; Gasparian, A.; Glamazdin, A.; Goryachev, V.; Hardy, K.; He, J.; Ito, M.; Jiang, L.; Kashy, D.; Khandaker, M.; Kolarkar, A.; Konchatnyi, M.; Korchin, A.; Korsch, W.; Kosinov, O.; Kowalski, S.; Kubantsev, M.; Kubarovsky, V.; Larin, I.; Lawrence, D.; Li, X.; Martel, P.; Matveev, V.; McNulty, D.; Mecking, B.; Milbrath, B.; Minehart, R.; Miskimen, R.; Mochalov, V.; Nakagawa, I.; Overby, S.; Pasyuk, E.; Payen, M.; Pedroni, R.; Prok, Y.; Ritchie, B.; Salgado, C.; Shahinyan, A.; Sitnikov, A.; Sober, D.; Stepanyan, S.; Stevens, W.; Underwood, J.; Vasiliev, A.; Vishnyakov, V.; Wood, M.; Zhou, S.

    2014-12-01

    The Jefferson Laboratory PrimEx Collaboration has developed and implemented a method to control the tagged photon flux in photoproduction experiments at the 1% level over the photon energy range from 4.9 to 5.5 GeV. This method has been successfully implemented in a high precision measurement of the neutral pion lifetime. Here, we outline the experimental equipment and the analysis techniques used to accomplish this. These include the use of a total absorption counter for absolute flux calibration, a pair spectrometer for online relative flux monitoring, and a new method for post-bremsstrahlung electron counting.

  14. A High-Speed Baseline Restorer for Neutron Flux Detection in ITER

    Institute of Scientific and Technical Information of China (English)

    曹宏睿; 李世平; 徐修峰; 袁国梁; 杨青巍; 阴泽杰

    2012-01-01

    A neutron flux monitor .(NFM) is a key diagnostic system in the International Ther- monuclear Experimental Reactor (ITER), and may provide readings of a series of important parameters in fusion reaction processes. As a valuable part of the main electronics system of the NFM, the high-speed baseline restorer we designed is an important signal conversion plug-in which can restore the input signal baseline offset to a zero level, while keeping the output pulse signal waveform from the preamplifier basically unchanged.

  15. Comparing changes in plasma and skin autofluorescence in low-flux versus high-flux hemodialysis

    NARCIS (Netherlands)

    Ramsauer, Bernd; Engels, Gerwin; Arsov, Stefan; Hadimeri, Henrik; Sikole, Aleksandar; Graaff, Reindert; Stegmayr, Bernd

    2015-01-01

    Background: Tissue advanced glycation end products (AGE) are increased in hemodialysis (HD) patients, especially those with cardiovascular complications. Skin autofluorescence (skin-AF) can noninvasively estimate the accumulation of AGE in tissue. The aim was to clarify whether HD using a high-flux

  16. Push technology at Argonne National Laboratory.

    Energy Technology Data Exchange (ETDEWEB)

    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).

  17. Visualization Study on High Heat Flux Boiling and Critical Heat Flux

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Satbyoul; Kim, Hyungdae [Kyung Hee University, Yongin (Korea, Republic of)

    2015-10-15

    In this study, an integrated visible and infrared-based experimental method is introduced to simultaneously measure the details of high-resolution liquid-vapor phase and heat transfer distributions on a heated wall. The dynamics and heat transfer at high heat flux boiling and critical heat flux were observed. The experiment was conducted in pool of saturated water under atmospheric pressure. There have been many studies to examine the physical mechanisms of nucleation boiling and critical heat flux over several decades. Several visible and infrared-based optical techniques for time-resolved high resolution measurements for liquid-vapor phase and heater surface temperature during boiling have been introduced to understand the characteristics and mechanisms of them. Liquid-vapor phase, temperature, and heat flux distributions on the heated surface were measured during pool boiling of water using the integrated total reflection and infrared thermometry technique. Qualitative examination of the data for high heat flux boiling and CHF was performed. The main contributions of this work are summarized below. The existence and behavior of dry patches lead the way toward CHF condition. Therefore, the mechanistic modeling of the CHF phenomenon necessarily needs to include the physical parameters related to dynamics of the large dry patch such as life time and size. In addition to the dynamic behavior of the dry patch, the thermal behavior of the hot patch is also important. Even though the dry area was rewetted, the stored thermal energy in the hot patch can be remained if the rewetting time is short and the subsequent dry patch is regenerated quickly.

  18. Evaluation of the thermal neutron flux in the core of IPEN/MB-01 reactor using the code Monte Carlo (MCNP)

    Energy Technology Data Exchange (ETDEWEB)

    Salome, Jean A.D.; Cardoso, Fabiano; Faria, Rochkhudson B.; Pereira, Claubia, E-mail: jadsalome@yahoo.com.br, E-mail: fabinuclear@yahoo.com.br, E-mail: rockdefaria@yahoo.com.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2015-07-01

    The IPEN/MB-01 reactor, located in the city of Sao Paulo - Brazil, reached its first criticality on the year of 1988. The reactor is characterized by a low output power of 100 W only, even because its purpose is to produce knowledge about nuclear power plants on a smaller geometric scale without the requirement of an extremely complex cooling system. The use of devices such as this it is very interesting because it achieves the demands of nuclear engineering about the neutronic parameters needed in the design of large nuclear plants through relatively simple and inexpensive methods. In this paper, the computational mathematical code MCNP5 is used to perform the calculation of the thermal neutron flux in the core of the IPEN/MB-01 reactor. To do this is used an experiment from the LEU-COMP-THERM-077 benchmark that represents the standard rectangular configuration of the IPEN/MB-01 reactor. The thermal neutron flux is calculated at some axial planes of different heights and, after that, axial profiles of the thermal neutron flux are done and compared to experimental results issued previously. The experimental values used as reference refer to a cylindrical configuration of the core of the reactor. Finally, the pertinence and relevance of the results are checked. With this work is expected to produce more knowledge about the dynamics of neutron flux in the core of the IPEN/MB-01 reactor. (author)

  19. Remote field eddy current technique for gap measurement of horizontal flux detector guide tube in pressurized heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Hoon; Jung, Hyun Kyu; Yang, Dong Ju; Cheong, Yong Moo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2004-11-15

    The fuel channels including the pressure tube(PT) and the calandria tube(CT) are important components of the pressurized heavy water reactor(PHWR). A sagging of fuel channel increases by heat and radiation exposure with the increasing operation time. The contact of fuel channel to the Horizontal flux Detector(HFD) guide tube is needed for the power plant safety. In order to solve this safety issue, the electromagnetic technique was applied to measure the status of the guide tube. The Horizontal flux Detector(HFD) guide tube and the Calandria tube(CT) in the Pressurized Heavy Water Reactor(PHWR) are cross-aligned horizontally. The remote field eddy current(RFEC) technology is applied for gap measurement between the HFD guide tube and the CT HFD guide tube can be detected by inserting the RFEC probe into pressure tube(PT) at the crossing point directly. The RFEC signals using the volume integral method(VIM) were simulated for obtaining the optimal inspection parameters. This paper shows that the simulated eddy current signals and the experimental results in variance with the CT/HFD gap.

  20. Control rod drive for high temperature gas cooled reactor

    Institute of Scientific and Technical Information of China (English)

    DengJun-Xian; XuJi-Ming; 等

    1998-01-01

    This control rod drive is developed for HTR-10 high temperature gas cooled test reactor.The stepmotor is prefered to improve positioning of the control rod and the scram behavior.The preliminary test in 1600170 ambient temperature shows that the selected stepmotor and transmission system can meet the main operation function requirements of HTR-10.

  1. High Field Seeking State Atom Laser and Properties of Flux

    Institute of Scientific and Technical Information of China (English)

    XIA Lin; XIONG Wei; YANG Fan; YI Lin; ZHOU Xiao-Ji; CHEN Xu-Zong

    2008-01-01

    We present an experimental study on the continuous atom laser. The experiments show that a high field seeking state atom laser with stable flux can be formed by increasing the strength of outcoupling before large density fluctuations appear. It is easy to obtain a long length or high speed output with this kind of atom laser.

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

    Energy Technology Data Exchange (ETDEWEB)

    Westfall, C.

    2005-09-20

    equipment that cost $100,000 by the 1970s alongside work at 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

  3. A high resolution pneumatic stepping actuator for harsh reactor environments

    Science.gov (United States)

    Tippetts, Thomas B.; Evans, Paul S.; Riffle, George K.

    1993-01-01

    A reactivity control actuator for a high-power density nuclear propulsion reactor must be installed in close proximity to the reactor core. The energy input from radiation to the actuator structure could exceed hundreds of W/cc unless low-cross section, low-absorptivity materials are chosen. Also, for post-test handling and subsequent storage, materials should not be used that are activated into long half-life isotopes. Pneumatic actuators can be constructed from various reactor-compatible materials, but conventional pneumatic piston actuators generally lack the stiffness required for high resolution reactivity control unless electrical position sensors and compensated electronic control systems are used. To overcome these limitations, a pneumatic actuator is under development that positions an output shaft in response to a series of pneumatic pulses, comprising a pneumatic analog of an electrical stepping motor. The pneumatic pulses are generated remotely, beyond the strong radiation environment, and transmitted to the actuator through tubing. The mechanically simple actuator uses a nutating gear harmonic drive to convert motion of small pistons directly to high-resolution angular motion of the output shaft. The digital nature of this actuator is suitable for various reactor control algorithms but is especially compatible with the three bean salad algorithm discussed by Ball et al. (1991).

  4. Flux attenuation at NREL`s High-Flux Solar Furnace

    Energy Technology Data Exchange (ETDEWEB)

    Bingham, C E; Scholl, K L; Lewandowski, A A

    1994-10-01

    The High-Flux Solar Furnace (HFSF) at the National Renewable Energy Laboratory (NREL) has a faceted primary concentrator and a long focal-length-to-diameter ratio (due to its off-axis design). Each primary facet can be aimed individually to produce different flux distributions at the target plane. Two different types of attenuators are used depending on the flux distribution. A sliding-plate attenuator is used primarily when the facets are aimed at the same target point. The alternate attenuator resembles a venetian blind. Both attenuators are located between the concentrator and the focal point. The venetian-blind attenuator is primarily used to control the levels of sunlight failing on a target when the primary concentrators are not focused to a single point. This paper will demonstrate the problem of using the sliding-plate attenuator with a faceted concentrator when the facets are not aimed at the same target point. We will show that although the alternate attenuator necessarily blocks a certain amount of incoming sunlight, even when fully open, it provides a more even attenuation of the flux for alternate aiming strategies.

  5. 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

    Energy Technology Data Exchange (ETDEWEB)

    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).

  6. Generation IV Reactors Integrated Materials Technology Program Plan: Focus on Very High Temperature Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, William R [ORNL; Burchell, Timothy D [ORNL; Katoh, Yutai [ORNL; McGreevy, Timothy E [ORNL; Nanstad, Randy K [ORNL; Ren, Weiju [ORNL; Snead, Lance Lewis [ORNL; Wilson, Dane F [ORNL

    2008-08-01

    Since 2002, the Department of Energy's (DOE's) Generation IV Nuclear Energy Systems (Gen IV) Program has addressed the research and development (R&D) necessary to support next-generation nuclear energy systems. The six most promising systems identified for next-generation nuclear energy are described within this roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor-SCWR and the Very High Temperature Reactor-VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor-GFR, the Lead-cooled Fast Reactor-LFR, and the Sodium-cooled Fast Reactor-SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides and may provide an alternative to accelerator-driven systems. At the inception of DOE's Gen IV program, it was decided to significantly pursue five of the six concepts identified in the Gen IV roadmap to determine which of them was most appropriate to meet the needs of future U.S. nuclear power generation. In particular, evaluation of the highly efficient thermal SCWR and VHTR reactors was initiated primarily for energy production, and evaluation of the three fast reactor concepts, SFR, LFR, and GFR, was begun to assess viability for both energy production and their potential contribution to closing the fuel cycle. Within the Gen IV Program itself, only the VHTR class of reactors was selected for continued development. Hence, this document will address the multiple activities under the Gen IV program that contribute to the development of the VHTR. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of

  7. Development activities of the high heat flux scraper element

    Energy Technology Data Exchange (ETDEWEB)

    Boscary, J., E-mail: jean.boscary@ipp.mpg.de [Max Planck Institute for Plasma Physics, Garching (Germany); Lore, J.; Lumsdaine, A. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Maier, M. [Max Planck Institute for Plasma Physics, Garching (Germany); McGinnis, D. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Peacock, A.; Tretter, J. [Max Planck Institute for Plasma Physics, Garching (Germany)

    2015-10-15

    The function of the high heat flux scraper element is to reduce the heat loads on the element ends of the actively cooled divertor of Wendelstein 7-X. The scraper element is actively water cooled to remove up to 550 kW steady state power load, with localized heat fluxes as high as 20 MW/m{sup 2}. Its surface area, 0.17 m{sup 2}, is contoured to optimally intercept both upstream and downstream particle fluxes. The plasma facing surface is made of 24 individual scraper fingers based on the monoblock technology. Each scraper finger is 247 mm long and 28 mm wide and has 13 monoblocks made of CFC NB31 bonded by hot isostatic pressing onto a CuCrZr cooling tube equipped with a copper twisted tape. Development activities, described here, include the design and fabrication of prototypes to validate the different technologies selected for the scraper element design to prepare a possible production.

  8. Performance characterization of the SERI High-Flux Solar Furnace

    Energy Technology Data Exchange (ETDEWEB)

    Lewandowski, A.; Bingham, C. (Solar Energy Research Inst., Golden, CO (United States)); O' Gallagher, J.; Winston, R.; Sagie, D. (Univ. of Chicago, IL (United States))

    1991-12-01

    This paper describes a unique, new solar furnace at the Solar Energy Research Institute (SERI) that can generate a wide range of flux concentrations to support research in areas including materials processing, high-temperature detoxification and high-flux optics. The furnace is unique in that it uses a flat, tracking heliostat along with a long focal length-to-diameter (f/D) primary concentrator in an off-axis configuration. The experiments are located inside a building completely outside the beam between the heliostat and primary concentrator. The long f/D ratio of the primary concentrator was designed to take advantage of a nonimaging secondary concentrator to significantly increase the flux concentration capabilities of the system. Results are reported for both the single-stage and two-stage configurations. (orig.).

  9. Model biases in high-burnup fast reactor simulations

    Energy Technology Data Exchange (ETDEWEB)

    Touran, N.; Cheatham, J.; Petroski, R. [TerraPower LLC, 11235 S.E. 6th St, Bellevue, WA 98004 (United States)

    2012-07-01

    A new code system called the Advanced Reactor Modeling Interface (ARMI) has been developed that loosely couples multiscale, multiphysics nuclear reactor simulations to provide rapid, user-friendly, high-fidelity full systems analysis. Incorporating neutronic, thermal-hydraulic, safety/transient, fuel performance, core mechanical, and economic analyses, ARMI provides 'one-click' assessments of many multi-disciplined performance metrics and constraints that historically require iterations between many diverse experts. The capabilities of ARMI are implemented in this study to quantify neutronic biases of various modeling approximations typically made in fast reactor analysis at an equilibrium condition, after many repetitive shuffles. Sensitivities at equilibrium that result in very high discharge burnup are considered ( and >20% FIMA), as motivated by the development of the Traveling Wave Reactor. Model approximations discussed include homogenization, neutronic and depletion mesh resolution, thermal-hydraulic coupling, explicit control rod insertion, burnup-dependent cross sections, fission product model, burn chain truncation, and dynamic fuel performance. The sensitivities of these approximations on equilibrium discharge burnup, k{sub eff}, power density, delayed neutron fraction, and coolant temperature coefficient are discussed. (authors)

  10. Process heat cogeneration using a high temperature reactor

    Energy Technology Data Exchange (ETDEWEB)

    Alonso, Gustavo, E-mail: gustavoalonso3@gmail.com [Instituto Nacional de Investigaciones Nucleares, Carretera Mexico-Toluca s/n, Ocoyoacac, Edo. De Mexico 52750 (Mexico); Instituto Politécnico Nacional, Unidad Profesional Adolfo Lopez Mateos, Ed. 9, Lindavista, D.F. 07300 (Mexico); Ramirez, Ramon [Instituto Nacional de Investigaciones Nucleares, Carretera Mexico-Toluca s/n, Ocoyoacac, Edo. De Mexico 52750 (Mexico); Valle, Edmundo del [Instituto Politécnico Nacional, Unidad Profesional Adolfo Lopez Mateos, Ed. 9, Lindavista, D.F. 07300 (Mexico); Castillo, Rogelio [Instituto Nacional de Investigaciones Nucleares, Carretera Mexico-Toluca s/n, Ocoyoacac, Edo. De Mexico 52750 (Mexico)

    2014-12-15

    Highlights: • HTR feasibility for process heat cogeneration is assessed. • A cogeneration coupling for HTR is proposed and process heat cost is evaluated. • A CCGT process heat cogeneration set up is also assessed. • Technical comparison between both sources of cogeneration is performed. • Economical competitiveness of the HTR for process heat cogeneration is analyzed. - Abstract: High temperature nuclear reactors offer the possibility to generate process heat that could be used in the oil industry, particularly in refineries for gasoline production. These technologies are still under development and none of them has shown how this can be possible and what will be the penalty in electricity generation to have this additional product and if the cost of this subproduct will be competitive with other alternatives. The current study assesses the likeliness of generating process heat from Pebble Bed Modular Reactor to be used for a refinery showing different plant balances and alternatives to produce and use that process heat. An actual practical example is presented to demonstrate the cogeneration viability using the fact that the PBMR is a modular small reactor where the cycle configuration to transport the heat of the reactor to the process plant plays an important role in the cycle efficiency and in the plant economics. The results of this study show that the PBMR would be most competitive when capital discount rates are low (5%), carbon prices are high (>30 US$/ton), and competing natural gas prices are at least 8 US$/mmBTU.

  11. Burning high-level TRU waste in fusion fission reactors

    Science.gov (United States)

    Shen, Yaosong

    2016-09-01

    Recently, the concept of actinide burning instead of a once-through fuel cycle for disposing spent nuclear fuel seems to get much more attention. A new method of burning high-level transuranic (TRU) waste combined with Thorium-Uranium (Th-U) fuel in the subcritical reactors driven by external fusion neutron sources is proposed in this paper. The thorium-based TRU fuel burns all of the long-lived actinides via a hard neutron spectrum while outputting power. A one-dimensional model of the reactor concept was built by means of the ONESN_BURN code with new data libraries. The numerical results included actinide radioactivity, biological hazard potential, and much higher burnup rate of high-level transuranic waste. The comparison of the fusion-fission reactor with the thermal reactor shows that the harder neutron spectrum is more efficient than the soft. The Th-U cycle produces less TRU, less radiotoxicity and fewer long-lived actinides. The Th-U cycle provides breeding of 233U with a long operation time (>20 years), hence significantly reducing the reactivity swing while improving safety and burnup.

  12. High Throughput Facility

    Data.gov (United States)

    Federal Laboratory Consortium — Argonne?s high throughput facility provides highly automated and parallel approaches to material and materials chemistry development. The facility allows scientists...

  13. Preliminary Demonstration Reactor Point Design for the Fluoride Salt-Cooled High-Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Greenwood, Michael Scott [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-12-01

    Development of the Fluoride Salt-Cooled High-Temperature Reactor (FHR) Demonstration Reactor (DR) is a necessary intermediate step to enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will utilize known, mature technology to close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include tristructural-isotropic (TRISO) particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell heat exchangers. This report provides an update on the development of the FHR DR. At this writing, the core neutronics and thermal hydraulics have been developed and analyzed. The mechanical design details are still under development and are described to their current level of fidelity. It is anticipated that the FHR DR can be operational within 10 years because of the use of low-risk, near-term technology options.

  14. Report on the joint meeting of the Division of Development and Technology Plasma/Wall Interaction and High Heat Flux Materials and Components Task Groups

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, K.L. (ed.)

    1985-10-01

    This report of the Joint Meeting of the Division of Development and Technology Plasma/Wall Interaction and High Heat Flux Materials and Components Task Groups contains contributing papers in the following areas: Plasma/Materials Interaction Program and Technical Assessment, High Heat Flux Materials and Components Program and Technical Assessment, Pumped Limiters, Ignition Devices, Program Planning Activities, Compact High Power Density Reactor Requirements, Steady State Tokamaks, and Tritium Plasma Experiments. All these areas involve the consideration of High Heat Flux on Materials and the Interaction of the Plasma with the First Wall. Many of the Test Facilities are described as well. (LSP)

  15. Studies of low temperature, low flux radiation embrittlement of nuclear reactor structural materials. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Odette, G.R.; Lucas, G.E.

    1993-06-01

    There are several existing research programs which have components pertinent to the issue of low flux/low temperature embrittlement; in particular, examination of the Shippingport shield tank which has been exposed to low flux and relatively low temperature is being performed by ANL, and evaluation of low temperature embrittlement in A508 and A533B steels in support of the HTGR is currently being performed by ORNL. However, these programs are not specifically directed at the broader issue of low flux/low temperature embrittlement in a range of structural steels. Hence, the authors coordinated their effort with these programs so that their investigations were complementary to existing programs, and they focused on a set of materials which expand the data base developed in these programs. In particular, the authors have investigated embrittlement phenomena in steels that are similar to those used in support structure.

  16. High temperature reactors for cogeneration applications

    Energy Technology Data Exchange (ETDEWEB)

    Verfondern, Karl [Forschungszentrum Juelich (Germany). IEK-6; Allelein, Hans-Josef [Forschungszentrum Juelich (Germany). IEK-6; RWTH Aachen (Germany). Lehrstuhl fuer Reaktorsicherheit und -technik (LRST)

    2016-05-15

    There is a large potential for nuclear energy also in the non-electric heat market. Many industrial sectors have a high demand for process heat and steam at various levels of temperature and pressure to be provided for desalination of seawater, district heating, or chemical processes. The future generation of nuclear plants will be capable to enter the wide field of cogeneration of heat and power (CHP), to reduce waste heat and to increase efficiency. This requires an adjustment to multiple needs of the customers in terms of size and application. All Generation-IV concepts proposed are designed for coolant outlet temperatures above 500 C, which allow applications in the low and medium temperature range. A VHTR would even be able to cover the whole temperature range up to approx. 1 000 C.

  17. Alcohol synthesis in a high-temperature slurry reactor

    Energy Technology Data Exchange (ETDEWEB)

    Roberts, G.W.; Marquez, M.A.; McCutchen, M.S. [North Carolina State Univ., Raleigh, NC (United States)

    1995-12-31

    The overall objective of this contract is to develop improved process and catalyst technology for producing higher alcohols from synthesis gas or its derivatives. Recent research has been focused on developing a slurry reactor that can operate at temperatures up to about 400{degrees}C and on evaluating the so-called {open_quotes}high pressure{close_quotes} methanol synthesis catalyst using this reactor. A laboratory stirred autoclave reactor has been developed that is capable of operating at temperatures up to 400{degrees}C and pressures of at least 170 atm. The overhead system on the reactor is designed so that the temperature of the gas leaving the system can be closely controlled. An external liquid-level detector is installed on the gas/liquid separator and a pump is used to return condensed slurry liquid from the separator to the reactor. In order to ensure that gas/liquid mass transfer does not influence the observed reaction rate, it was necessary to feed the synthesis gas below the level of the agitator. The performance of a commercial {open_quotes}high pressure {close_quotes} methanol synthesis catalyst, the so-called {open_quotes}zinc chromite{close_quotes} catalyst, has been characterized over a range of temperature from 275 to 400{degrees}C, a range of pressure from 70 to 170 atm., a range of H{sub 2}/CO ratios from 0.5 to 2.0 and a range of space velocities from 2500 to 10,000 sL/kg.(catalyst),hr. Towards the lower end of the temperature range, methanol was the only significant product.

  18. High conductivity Be-Cu alloys for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lilley, E.A. [NGK Metals Corp., Reading, PA (United States); Adachi, Takao; Ishibashi, Yoshiki [NGK Insulators, Ltd., Aichi-ken (Japan)

    1995-09-01

    The optimum material has not yet been identified. This will result in heat from plasma to the first wall and divertor. That is, because of cracks and melting by thermal power and shock. Today, it is considered to be some kinds of copper, alloys, however, for using, it must have high conductivity. And it is also needed another property, for example, high strength and so on. We have developed some new beryllium copper alloys with high conductivity, high strength, and high endurance. Therefore, we are introducing these new alloys as suitable materials for the heat sink in fusion reactors.

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

    Science.gov (United States)

    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.

  20. Baseline Concept Description of a Small Modular High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hans Gougar

    2014-05-01

    The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the

  1. Baseline Concept Description of a Small Modular High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gougar, Hans D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-10-01

    The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the

  2. Advanced High-Temperature, High-Pressure Transport Reactor Gasification

    Energy Technology Data Exchange (ETDEWEB)

    Michael L. Swanson

    2005-08-30

    The transport reactor development unit (TRDU) was modified to accommodate oxygen-blown operation in support of a Vision 21-type energy plex that could produce power, chemicals, and fuel. These modifications consisted of changing the loop seal design from a J-leg to an L-valve configuration, thereby increasing the mixing zone length and residence time. In addition, the standpipe, dipleg, and L-valve diameters were increased to reduce slugging caused by bubble formation in the lightly fluidized sections of the solid return legs. A seal pot was added to the bottom of the dipleg so that the level of solids in the standpipe could be operated independently of the dipleg return leg. A separate coal feed nozzle was added that could inject the coal upward into the outlet of the mixing zone, thereby precluding any chance of the fresh coal feed back-mixing into the oxidizing zone of the mixing zone; however, difficulties with this coal feed configuration led to a switch back to the original downward configuration. Instrumentation to measure and control the flow of oxygen and steam to the burner and mix zone ports was added to allow the TRDU to be operated under full oxygen-blown conditions. In total, ten test campaigns have been conducted under enriched-air or full oxygen-blown conditions. During these tests, 1515 hours of coal feed with 660 hours of air-blown gasification and 720 hours of enriched-air or oxygen-blown coal gasification were completed under this particular contract. During these tests, approximately 366 hours of operation with Wyodak, 123 hours with Navajo sub-bituminous coal, 143 hours with Illinois No. 6, 106 hours with SUFCo, 110 hours with Prater Creek, 48 hours with Calumet, and 134 hours with a Pittsburgh No. 8 bituminous coal were completed. In addition, 331 hours of operation on low-rank coals such as North Dakota lignite, Australian brown coal, and a 90:10 wt% mixture of lignite and wood waste were completed. Also included in these test campaigns was

  3. Heat flux limits on the plasma-facing components for a commercial fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wang, X.R.; Tillack, M.S. [Univ. of California, San Diego, La Jolla, CA (United States); Sze, D.K. [Argonne National Lab., IL (United States); Wong, C.P.C. [General Atomics, San Diego, CA (United States)

    1995-12-31

    In this work, the heat flux limits of conventional plasma-facing components (PFC) were examined. The limits are based on maximum allowable temperature and stress levels in the structures. The substrate materials considered were V, SiC composite and HT-9. The use of Cu also was considered. However, low temperature limits, activation and very limited radiation damage life time, make the using of Cu in a commercial power plant unattractive. With selected heat transfer enhancement, the heat flux allowable is about 5.3 MW/m{sup 2} for lithium-cooled V-alloy, 2.7 MW/m{sup 2} for helium-cooled SiC composite, and 2.7 MW/m{sup 2} for helium/water-cooled HT-9. Compared with the maximum heat flux attainable with Cu and cold water (13.4 MW/m{sup 2}), acceptable power plant materials place severe restrictions on heat removal. The thermal conductivity of SiC composite at 1,000 C and after irradiation is a factor of several lowered than the value the authors used. This indicates a need to examine the heat transfer problems associated with PFC, in terms of material development and enhancement in heat transfer. Physics regimes which can provide low peak and average heat flux should be pursued.

  4. Flux-freezing breakdown in high-conductivity magnetohydrodynamic turbulence.

    Science.gov (United States)

    Eyink, Gregory; Vishniac, Ethan; Lalescu, Cristian; Aluie, Hussein; Kanov, Kalin; Bürger, Kai; Burns, Randal; Meneveau, Charles; Szalay, Alexander

    2013-05-23

    The idea of 'frozen-in' magnetic field lines for ideal plasmas is useful to explain diverse astrophysical phenomena, for example the shedding of excess angular momentum from protostars by twisting of field lines frozen into the interstellar medium. Frozen-in field lines, however, preclude the rapid changes in magnetic topology observed at high conductivities, as in solar flares. Microphysical plasma processes are a proposed explanation of the observed high rates, but it is an open question whether such processes can rapidly reconnect astrophysical flux structures much greater in extent than several thousand ion gyroradii. An alternative explanation is that turbulent Richardson advection brings field lines implosively together from distances far apart to separations of the order of gyroradii. Here we report an analysis of a simulation of magnetohydrodynamic turbulence at high conductivity that exhibits Richardson dispersion. This effect of advection in rough velocity fields, which appear non-differentiable in space, leads to line motions that are completely indeterministic or 'spontaneously stochastic', as predicted in analytical studies. The turbulent breakdown of standard flux freezing at scales greater than the ion gyroradius can explain fast reconnection of very large-scale flux structures, both observed (solar flares and coronal mass ejections) and predicted (the inner heliosheath, accretion disks, γ-ray bursts and so on). For laminar plasma flows with smooth velocity fields or for low turbulence intensity, stochastic flux freezing reduces to the usual frozen-in condition.

  5. Advanced High Temperature Reactor Systems and Economic Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Peretz, Fred J [ORNL; Qualls, A L [ORNL

    2011-09-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a large-output [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR's large thermal output enables direct comparison of its performance and requirements with other high output reactor concepts. As high-temperature plants, FHRs can support either high-efficiency electricity generation or industrial process heat production. The AHTR analysis presented in this report is limited to the electricity generation mission. FHRs, in principle, have the potential to be low-cost electricity producers while maintaining full passive safety. However, no FHR has been built, and no FHR design has reached the stage of maturity where realistic economic analysis can be performed. The system design effort described in this report represents early steps along the design path toward being able to predict the cost and performance characteristics of the AHTR as well as toward being able to identify the technology developments necessary to build an FHR power plant. While FHRs represent a distinct reactor class, they inherit desirable attributes from other thermal power plants whose characteristics can be studied to provide general guidance on plant configuration, anticipated performance, and costs. Molten salt reactors provide experience on the materials, procedures, and components necessary to use liquid fluoride salts. Liquid metal reactors provide design experience on using low-pressure liquid coolants, passive decay heat removal, and hot refueling. High temperature gas-cooled reactors provide experience with coated particle fuel and graphite components. Light water reactors (LWRs) show the potentials of transparent, high-heat capacity coolants with low chemical reactivity. Modern coal-fired power plants provide design experience

  6. A new efficient empirical correlation for filtrate flux in slurry bubble column reactor of a gas-to-liquid process

    Energy Technology Data Exchange (ETDEWEB)

    Hemmati, Mohammad Reza [Entekhab Petrochemical Co., Tehran (Iran, Islamic Republic of); Khodagholi, Mohammad Ali [Research Institute of Petroleum Industry, Tehran (Iran, Islamic Republic of)

    2015-12-15

    Gas to Liquid has recently become of great interest. In this technology slurry bubble column reactors are favored for many reasons. Separation of liquid wax from the slurry is still a major problem that may be done by internal or external filtration. A system of sintered metal candle filters are designed and operated to collect experimental data of internal filtration. Data for 4 and 8 micron filter elements with different pressure differences and kinematic viscosity were collected. Data analysis revealed that these data could be correlated as a simple function of time, pressure drop and kinematic viscosity. This new and efficient correlation shows excellent ability to reproduce original data at moderate filtration conditions, but it is less precious in severe conditions. It was understood that main reason for this behavior is different filtrate flux regimes through filter media pores, led to inability of a single correlation to fit both regimes properly.

  7. Heat flux mitigation by impurity seeding in high-field tokamaks

    Science.gov (United States)

    Reinke, M. L.

    2017-03-01

    The ability for tokamaks to exhaust power in the boundary via impurity radiation is explored using empirical scalings and a simple 0D exhaust model, focusing on the scaling with toroidal field and major radius. By combining a scaling for the heat flux width and the L-H threshold power, the parallel heat flux in the SOL is shown to scale strongly with magnetic field, {{q}\\parallel}∼ B\\text{T}2.52 while having little to no scaling with machine size, {{q}\\parallel}∼ {{R}0.16} . Despite the increased heat flux at high field, it is shown that target temperatures relevant to detachment can be reached with finite main-ion dilution for a variety of impurity seeding gases, although non-equilibrium ionization balance is required in most cases. The necessary impurity fractions are estimated to scale like {{f}Z}∼ B\\text{T}0.88{{R}1.33} , a result that is facilitated by an increase in upstream temperature at high {{q}\\parallel} relative to peaks in the impurity cooling-curves. This scaling indicates that for optimizing reactors, minimizing device size while maximizing toroidal field, an approach shown to be consistent with energy confinement scaling, will also maximize the feasibility of reaching detachment at the lowest dilution. Despite this, analysis suggests an increase in the impurity fractions relative to existing devices will be required to exhaust power in a reactor-scale tokamak, with validation of impurity radiation physics required before both simple and detailed models can make reliable predictions of absolute f Z .

  8. Rapid Photocatalytic Degradation of Methylene Blue under High Photon Flux UV Irradiation: Characteristics and Comparison with Routine Low Photon Flux

    OpenAIRE

    2012-01-01

    This study examined the photocatalytic degradation efficiency under high UV photon flux (intensity normalized by photon energy) irradiation; the incident UV photon flux was 1 . 7 1 × 1 0 − 6 − 3 . 1 3 × 1 0 − 6 einstein c m − 2   s − 1 made by a super high-intensity UV apparatus. A comparative study between high photon flux photocatalytic process and routine low photon flux photocatalytic process for methylene blue degradation has been made in aqueous solution. The experimental results showed...

  9. BVOC ecosystem flux measurements at a high latitude wetland site

    Directory of Open Access Journals (Sweden)

    T. Holst

    2010-02-01

    Full Text Available In this study, we present summertime concentrations and fluxes of biogenic volatile organic compounds (BVOCs measured at a sub-arctic wetland in northern Sweden using a disjunct eddy-covariance (DEC technique based on a proton transfer reaction mass spectrometer (PTR-MS. The vegetation at the site was dominated by Sphagnum, Carex and extit{Eriophorum} spp. The measurements reported here cover a period of 50 days (1 August to 19 September 2006, approximately one half of the growing season at the site, and allowed to investigate the effect of day-to-day variation in weather as well as of vegetation senescence on daily BVOC fluxes, and on their temperature and light responses. The sensitivity drift of the DEC system was assessed by comparing H3O+-ion cluster formed with water molecules (H3O+(H2O at m37 with water vapour concentration measurements made using an adjacent humidity sensor, and the applicability of the DEC method was analysed by a comparison of sensible heat fluxes for high frequency and DEC data obtained from the sonic anemometer. These analyses showed no significant PTR-MS sensor drift over a period of several weeks and only a small flux-loss due to high-frequency spectrum omissions. This loss was within the range expected from other studies and the theoretical considerations.

    Standardised (20 °C and 1000 μmol m−2 s−1 PAR summer isoprene emission rates found in this study of 329 μg C m−2 (ground area h−1 were comparable with findings from more southern boreal forests, and fen-like ecosystems. On a diel scale, measured fluxes indicated a stronger temperature dependence than emissions from temperate or (subtropical ecosystems. For the first time, to our knowledge, we report ecosystem methanol fluxes from a sub-arctic ecosystem. Maximum daytime emission fluxes were around 270 μg m−2 h−1

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

    Energy Technology Data Exchange (ETDEWEB)

    Westfall, C.

    2005-09-20

    equipment that cost $100,000 by the 1970s alongside work at 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

  11. Uncertainty of calorimeter measurements at NREL's high flux solar furnace

    Science.gov (United States)

    Bingham, C. E.

    1991-12-01

    The uncertainties of the calorimeter and concentration measurements at the High Flux Solar Furnace (HFSF) at the National Renewable Energy Laboratory (NREL) are discussed. Two calorimeter types have been used to date. One is an array of seven commercially available circular foil calorimeters (gardon or heat flux gages) for primary concentrator peak flux (up to 250 W/sq cm). The second is a cold-water calorimeter designed and built by the University of Chicago to measure the average exit power of the reflective compound parabolic secondary concentrator used at the HFSF (over 3.3 kW across a 1.6/sq cm) exit aperture, corresponding to a flux of about 2 kW/sq cm. This paper discussed the uncertainties of the calorimeter and pyrheliometer measurements and resulting concentration calculations. The measurement uncertainty analysis is performed according to the ASME/ANSI standard PTC 19.1 (1985). Random and bias errors for each portion of the measurement are analyzed. The results show that as either the power or the flux is reduced, the uncertainties increase. Another calorimeter is being designed for a new, refractive secondary which will use a refractive material to produce a higher average flux (5 kW/sq cm) than the reflective secondary. The new calorimeter will use a time derivative of the fluid temperature as a key measurement of the average power out of the secondary. A description of this calorimeter and test procedure is also presented, along with a pre-test estimate of major sources of uncertainty.

  12. Uncertainty of calorimeter measurements at NREL's high flux solar furnace

    Energy Technology Data Exchange (ETDEWEB)

    Bingham, C.E.

    1991-12-01

    The uncertainties of the calorimeter and concentration measurements at the High Flux Solar Furnace (HFSF) at the National Renewable Energy Laboratory (NREL) are discussed. Two calorimeter types have been used to date. One is an array of seven commercially available circular foil calorimeters (gardon or heat flux gages) for primary concentrator peak flux (up to 250 W/cm{sup 2}). The second is a cold-water calorimeter designed and built by the University of Chicago to measure the average exit power of the reflective compound parabolic secondary concentrator used at the HFSF (over 3.3 kW across a 1.6cm{sup {minus}2} exit aperture, corresponding to a flux of about 2 kW/cm{sup 2}). This paper discussed the uncertainties of the calorimeter and pyrheliometer measurements and resulting concentration calculations. The measurement uncertainty analysis is performed according to the ASME/ANSI standard PTC 19.1 (1985). Random and bias errors for each portion of the measurement are analyzed. The results show that as either the power or the flux is reduced, the uncertainties increase. Another calorimeter is being designed for a new, refractive secondary which will use a refractive material to produce a higher average flux (5 kW/cm{sup 2}) than the reflective secondary. The new calorimeter will use a time derivative of the fluid temperature as a key measurement of the average power out of the secondary. A description of this calorimeter and test procedure is also presented, along with a pre-test estimate of major sources of uncertainty. 8 refs., 4 figs., 3 tabs.

  13. A reactor for high-throughput high-pressure nuclear magnetic resonance spectroscopy

    Science.gov (United States)

    Beach, N. J.; Knapp, S. M. M.; Landis, C. R.

    2015-10-01

    The design of a reactor for operando nuclear magnetic resonance (NMR) monitoring of high-pressure gas-liquid reactions is described. The Wisconsin High Pressure NMR Reactor (WiHP-NMRR) design comprises four modules: a sapphire NMR tube with titanium tube holder rated for pressures as high as 1000 psig (68 atm) and temperatures ranging from -90 to 90 °C, a gas circulation system that maintains equilibrium concentrations of dissolved gases during gas-consuming or gas-releasing reactions, a liquid injection apparatus that is capable of adding measured amounts of solutions to the reactor under high pressure conditions, and a rapid wash system that enables the reactor to be cleaned without removal from the NMR instrument. The WiHP-NMRR is compatible with commercial 10 mm NMR probes. Reactions performed in the WiHP-NMRR yield high quality, information-rich, and multinuclear NMR data over the entire reaction time course with rapid experimental turnaround.

  14. Argonne Code Center: benchmark problem book

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

    This report is a supplement to the original report, published in 1968, as revised. The Benchmark Problem Book is intended to serve as a source book of solutions to mathematically well-defined problems for which either analytical or very accurate approximate solutions are known. This supplement contains problems in eight new areas: two-dimensional (R-z) reactor model; multidimensional (Hex-z) HTGR model; PWR thermal hydraulics--flow between two channels with different heat fluxes; multidimensional (x-y-z) LWR model; neutron transport in a cylindrical ''black'' rod; neutron transport in a BWR rod bundle; multidimensional (x-y-z) BWR model; and neutronic depletion benchmark problems. This supplement contains only the additional pages and those requiring modification. (RWR)

  15. CFD simulation on critical heat flux of flow boiling in IVR-ERVC of a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Xiang, E-mail: zhangxiang3@snptc.com.cn [State Nuclear Power Technology Research & Development Center, South Area, Future Science and Technology Park, Chang Ping District, Beijing 102209 (China); Hu, Teng [State Nuclear Power Technology Research & Development Center, South Area, Future Science and Technology Park, Chang Ping District, Beijing 102209 (China); Chen, Deqi, E-mail: chendeqi@cqu.edu.cn [Key Laboratory of Low-grade Energy Utilization Technologies and Systems, Chongqing University, 400044 (China); Zhong, Yunke; Gao, Hong [Key Laboratory of Low-grade Energy Utilization Technologies and Systems, Chongqing University, 400044 (China)

    2016-08-01

    Highlights: • CFD simulation on CHF of boiling two-phase flow in ERVC is proposed. • CFD simulation result of CHF agrees well with that of experimental result. • The characteristics of boiling two-phase flow and boiling crisis are analyzed. - Abstract: The effectiveness of in-vessel retention (IVR) by external reactor vessel cooling (ERVC) strongly depends on the critical heat flux (CHF). As long as the local CHF does not exceed the local heat flux, the lower head of the pressure vessel can be cooled sufficiently to prevent from failure. In this paper, a CFD simulation is carried out to investigate the CHF of ERVC. This simulation is performed by a CFD code fluent couple with a boiling model by UDF (User-Defined Function). The experimental CHF of ERVC obtained by State Nuclear Power Technology Research and Development Center (SNPTRD) is used to validate this CFD simulation, and it is found that the simulation result agrees well with the experimental result. Based on the CFD simulation, detailed analysis focusing on the pressure distribution, velocity distribution, void fraction distribution, heating wall temperature distribution are proposed in this paper.

  16. Flow Induced Vibration Program at Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    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.

  17. High-Density Plasma Reactors: Simulations for Design

    Science.gov (United States)

    Hash, David B.; Meyyappan, Meyya; Arnold, James O. (Technical Monitor)

    1998-01-01

    The development of improved and more efficient plasma reactors is a costly process for the semiconductor industry. Until five years ago, the Industry made most of its advancements through a trial and error approach. More recently, the role of computational modeling in the design process has increased. Both conventional computational fluid dynamics (CFD) techniques like Navier-Stokes solvers as well as particle simulation methods are used to model plasma reactor flowfields. However, since high-density plasma reactors generally operate at low gas pressures on the order of 1 to 10 mTorr, a particle simulation may be necessary because of the failure of CFD techniques to model rarefaction effects. The direct simulation Monte Carlo method is the most widely accepted and employed particle simulation tool and has previously been used to investigate plasma reactor flowfields. A plasma DSMC code is currently under development at NASA Ames Research Center with its foundation as the object-oriented parallel Cornell DSMC code, MONACO. The present investigation is a follow up of a neutral flow investigation of the effects of process parameters as well as reactor design on etch rate and etch rate uniformity. The previous work concentrated on silicon etch of a chlorine flow in a configuration typical of electron cyclotron resonance (ECR) or helical resonator type reactors. The effects of the plasma on the dissociation chemistry were modeled by making assumptions about the electron temperature and number density. The electrons or ions themselves were not simulated.The present work extends these results by simulating the charged species.The electromagnetic fields are calculated such that power deposition is modeled self-consistently. Electron impact reactions are modeled along with mechanisms for charge exchange. An bipolar diffusion assumption is made whereby electrons remain tied to the ions. However, the velocities of tile electrons are allowed to be modified during collisions

  18. Overview of results of the first phase of validation activities for the IFMIF High Flux Test Module

    Energy Technology Data Exchange (ETDEWEB)

    Arbeiter, Frederik, E-mail: frederik.arbeiter@kit.edu [Karlsruhe Institute of Technology, Karlsruhe (Germany); Chen Yuming; Dolensky, Bernhard; Freund, Jana; Heupel, Tobias; Klein, Christine; Scheel, Nicola; Schlindwein, Georg [Karlsruhe Institute of Technology, Karlsruhe (Germany)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer Validation of computational fluid dynamics (CFD) modeling approach for application in the IFMIF High Flux Test Module. Black-Right-Pointing-Pointer Fabrication of prototypes of the irradiation capsules of the IFMIF High Flux Test Module. - Abstract: The international fusion materials irradiation facility (IFMIF) is projected to create an experimentally validated database of material properties relevant for fusion reactor designs. The IFMIF High Flux Test Module is the dedicated experiment to irradiate alloys in the temperature range 250-550 Degree-Sign C and up to 50 displacements per atom per irradiation cycle. The High Flux Test Module is developed to maximize the specimen payload in the restricted irradiation volume, and to minimize the temperature spread within each specimen bundle. Low pressure helium mini-channel cooling is used to offer a high integration density. Due to the demanding thermo-hydraulic and mechanical conditions, the engineering design process (involving numerical neutronic, thermo-hydraulic and mechanical analyses) is supported by extensive experimental validation activities. This paper reports on the prototype manufacturing, thermo-hydraulic modeling experiments and component tests, as well as on mechanical testing. For the testing of the 1:1 prototype of the High Flux Test Module, a dedicated test facility, the Helium Loop Karlsruhe-Low Pressure (HELOKA-LP) has been taken into service.

  19. High Tc Superconductor Theoretical Models and Electromagnetic Flux Characteristics

    Institute of Scientific and Technical Information of China (English)

    JIN Jian-xun

    2006-01-01

    High Tc Superconductors (HTS) have special electromagnetic characteristics and phenomena. Effort has been made in order to theoretically understand the applied HTS superconductivity and HTS behaviors for practical applications, various theoretical models related to the HTS electromagnetic properties have been developed. The theoretical models and analytic methods are summarized with regard to understanding the HTS magnetic flux characteristic which is one of the most critical issues related to HTS applications such as for HTS magnetic levitation application.

  20. High flux inductors for the rapid heating of steel products

    Energy Technology Data Exchange (ETDEWEB)

    Pierret, R.; Griffay, G.; Galbrun, F. [Institut de Recherches de la Siderurgie Francaise (IRSID), 78 - Saint-Germain-en-Laye (France); Hellegouaec`h, J.; Prost, G.

    1995-03-01

    To reduce investment and operating costs of electroheating processes of long products by induction, we developed a new multilayed inductor with high flux density which represents a real technological step in regard of conventional technics: 4 MV/m{sup 2} instead of 1MW/m{sup 2}, efficiency of 85% instead of 55%, compacity and low costs of maintenance. The new technology can also be used with success in flat products plants. (authors). 10 figs., 1 tab.

  1. Determination of the neutron flux in the reactor zones with the strong neutron absorption and leakage

    OpenAIRE

    Ljubenov Vladan; Milošević Miodrag 1

    2004-01-01

    The procedures for the numerical and experimental determination of the neutron flux in the zones with the strong neutron absorption and leakage are described in this paper. Numerical procedure is based on the application of the SCALE-4.4a code system where the Dancoff factors are determined by the VEGA2DAN code. Two main parts of the experimental methodology are measurement of the activity of irradiated foils and determination of the averaged neutron absorption cross-section in the foils by t...

  2. The Status of the US High-Temperature Gas Reactors

    OpenAIRE

    2016-01-01

    In 2005, the US passed the Energy Policy Act of 2005 mandating the construction and operation of a high-temperature gas reactor (HTGR) by 2021. This law was passed after a multiyear study by national experts on what future nuclear technologies should be developed. As a result of the Act, the US Congress chose to develop the so-called Next-Generation Nuclear Plant, which was to be an HTGR designed to produce process heat for hydrogen production. Despite high hopes and expectations, the current...

  3. Design and construction of an automatic measurement electronic system and graphical neutron flux for the subcritical reactor; Diseno y construccion de un sistema electronico automatico de medicion y graficado del flujo neutronico para el reactor subcritico

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez M, J.L.; Balderas, E.G.; Rivero G, T. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1997-07-01

    The National Institute of Nuclear Research (ININ) has in its installations with a nuclear subcritical reactor which was designed and constructed with the main purpose to be used in the nuclear sciences education in the Physics areas and Reactors engineering. Within the nuclear experiments that can be realized in this reactor are very interesting those about determinations of neutron and gamma fluxes spectra, since starting from these some interesting nuclear parameters can be obtained. In order to carry out this type of experiments different radioactive sources are used which exceed the permissible doses by far to human beings. Therefore it is necessary the remote handling as of the source as of detectors used in different experiments. In this work it is presented the design of an electronic system which allows the different positions inside of the tank of subcritical reactor at ININ over the radial and axial axes in manual or automatic ways. (Author)

  4. Brookhaven leak reactor to close

    CERN Multimedia

    MacIlwain, C

    1999-01-01

    The DOE has announced that the High Flux Beam Reactor at Brookhaven is to close for good. Though the news was not unexpected researchers were angry the decision had been taken before the review to assess the impact of reopening the reactor had been concluded (1 page).

  5. Thermal hydraulic test for reactor safety system - Critical heat flux experiment and development of prediction models

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Soon Heung; Baek, Won Pil; Yang, Soo Hyung; No, Chang Hyun [Korea Advanced Institute of Science and Technology, Taejon (Korea)

    2000-04-01

    To acquire CHF data through the experiments and develop prediction models, research was conducted. Final objectives of research are as follows: 1) Production of tube CHF data for low and middle pressure and mass flux and Flow Boiling Visualization. 2) Modification and suggestion of tube CHF prediction models. 3) Development of fuel bundle CHF prediction methodology base on tube CHF prediction models. The major results of research are as follows: 1) Production of the CHF data for low and middle pressure and mass flux. - Acquisition of CHF data (764) for low and middle pressure and flow conditions - Analysis of CHF trends based on the CHF data - Assessment of existing CHF prediction methods with the CHF data 2) Modification and suggestion of tube CHF prediction models. - Development of a unified CHF model applicable for a wide parametric range - Development of a threshold length correlation - Improvement of CHF look-up table using the threshold length correlation 3) Development of fuel bundle CHF prediction methodology base on tube CHF prediction models. - Development of bundle CHF prediction methodology using correction factor. 11 refs., 134 figs., 25 tabs. (Author)

  6. Filtration behavior of casein glycomacropeptide (CGMP) in an enzymatic membrane reactor: fouling control by membrane selection and threshold flux operation

    DEFF Research Database (Denmark)

    Luo, Jianquan; Morthensen, Sofie Thage; Meyer, Anne S.

    2014-01-01

    . In this study, the filtration performance and fouling behavior during ultrafiltration (UF) of CGMP for the enzymatic production of 3′-sialyllactose were investigated. A 5kDa regenerated cellulose membrane with high anti-fouling performance, could retain CGMP well, permeate 3′-sialyllactose, and was found...... concentration on the threshold flux were studied based on the resistance-in-series model. Higher hydrophilicity of the membrane, elevated pH and agitation, and lower CGMP concentration were found to increase the threshold flux and decrease membrane fouling....

  7. Calculation of neutron and gamma fluxes in support to the interpretation of measuring devices irradiated in the core periphery of the OSIRIS Material Testing Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Malouch, Fadhel [Alternative Energies and Atomic Energy Commission - CEA, Saclay Center, DEN/DANS/DM2S/SERMA, F-91191 Gif-sur-Yvette Cedex (France)

    2015-07-01

    Technological irradiations carried out in material testing reactors (MTRs) are used to study the behavior of materials under irradiation conditions required by different types of nuclear power plants (NPPs). For MTRs, specific instrumentation is required for the experiment monitoring and for the characterization of irradiation conditions, in particular the flux of neutrons and photons. To measure neutron and photon flux in experimental locations, different sensors can be used, such as SPNDs (self-powered neutron detectors), SPGDs (self-powered gamma detectors) and ionization chambers. These sensors involve interactions producing ultimately a measurable electric current. Various sensors have been recently tested in the core periphery of the OSIRIS reactor (located at the CEA-Saclay center) in order to qualify their responses to the neutron and the photon flux. One of the key input data for this qualification is to have a relevant evaluation of neutron and gamma fluxes at the irradiation location. The objective of this work is to evaluate the neutron and the gamma flux in the core periphery of the OSIRIS reactor. With this intention, specific neutron-photonic three-dimensional calculations have been performed and are mainly based on the TRIPOLI-4{sup R} three-dimensional continuous-energy Monte Carlo code, developed by CEA (Saclay Center) and extensively validated against reactor dosimetry benchmarks. In the case of the OSIRIS reactor, TRIPOLI-4{sup R} code has been validated against experimental results based on neutron flux and nuclear heating measurements performed in ex-core and in-core experiments. In this work, simultaneous contribution of neutrons and gamma photons in the core periphery is considered using neutron-photon coupled transport calculations. Contributions of prompt and decay photons have been taken into account for the gamma flux calculation. Specific depletion codes are used upstream to provide the decay-gamma sources required by TRIPOLI-4

  8. The final power calibration of the IPEN/MB-01 nuclear reactor for various configurations obtained from the measurements of the absolute average neutron flux

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alexandre Fonseca Povoa da, E-mail: alexandre.povoa@mar.mil.br [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Bitelli, Ulysses d' Utra; Mura, Luiz Ernesto Credidio; Lima, Ana Cecilia de Souza; Betti, Flavio; Santos, Diogo Feliciano dos, E-mail: ubitelli@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    The use of neutron activation foils is a widely spread technique applied to obtain nuclear parameters then comparing the results with those calculated using specific methodologies and available nuclear data. By irradiation of activation foils and subsequent measurement of its induced activity, it is possible to determine the neutron flux at the position of irradiation. The power level during operation of the reactor is a parameter which is directly proportional to the average neutron flux throughout the core. The objective of this work is to gather data from irradiation of gold foils symmetrically placed along a cylindrically configured core which presents only a small excess reactivity in order to derive the power generated throughout the spatial thermal and epithermal neutron flux distribution over the core of the IPEN/MB-01 Nuclear Reactor, eventually lending to a proper calibration of its nuclear channels. The foils are fixed in a Lucite plate then irradiated with and without cadmium sheaths so as to obtain the absolute thermal and epithermal neutron flux. The correlation between the average power neutron flux resulting from the gold foils irradiation, and the average power digitally indicated by the nuclear channel number 6, allows for the calibration of the nuclear channels of the reactor. The reactor power level obtained by thermal neutron flux mapping was (74.65 ± 2.45) watts to a mean counting per seconds of 37881 cps to nuclear channel number 10 a pulse detector, and 0.719.10{sup -5} ampere to nuclear linear channel number 6 (a non-compensated ionization chamber). (author)

  9. High Energy Atmospheric Neutrino Fluxes From a Realistic Primary Spectrum

    Science.gov (United States)

    Campos Penha, Felipe; Dembinski, Hans; Gaisser, Thomas K.; Tilav, Serap

    2016-03-01

    Atmospheric neutrino fluxes depend on the energy spectrum of primary nucleons entering the top of the atmosphere. Before the advent of AMANDA and the IceCube Neutrino Observatory, measurements of the neutrino fluxes were generally below ~ 1TeV , a regime in which a simple energy power law sufficed to describe the primary spectrum. Now, IceCube's muon neutrino data extends beyond 1PeV , including a combination of neutrinos from astrophysical sources with background from atmospheric neutrinos. At such high energies, the steepening at the knee of the primary spectrum must be accounted for. Here, we describe a semi-analytical approach for calculating the atmospheric differential neutrino fluxes at high energies. The input is a realistic primary spectrum consisting of 4 populations with distinct energy cutoffs, each with up to 7 representative nuclei, where the parameters were extracted from a global fit. We show the effect of each component on the atmospheric neutrino spectra, above 10TeV . The resulting features follow directly from recent air shower measurements included in the fit. Felipe Campos Penha gratefully acknowledges financial support from CAPES (Processo BEX 5348/14-5), CNPq (Processo 142180/2012-2), and the Bartol Research Institute.

  10. Characterization of the neutron flux in the Hohlraum of the thermal column of the TRIGA Mark III reactor of the ININ; Caracterizacion del flujo neutronico en el Hohlraum de la columna termica del reactor TRIGA Mark III del ININ

    Energy Technology Data Exchange (ETDEWEB)

    Delfin L, A.; Palacios, J.C.; Alonso, G. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: adl@nuclear.inin.mx

    2006-07-01

    Knowing the magnitude of the neutron flux in the reactor irradiation facilities, is so much importance for the operation of the same one, like for the investigation developing. Particularly, knowing with certain precision the spectrum and the neutron flux in the different positions of irradiation of a reactor, it is essential for the evaluation of the results obtained for a certain irradiation experiment. The TRIGA Mark III reactor account with irradiation facilities designed to carry out experimentation, where the reactor is used like an intense neutron source and gamma radiation, what allows to make irradiations of samples or equipment in radiation fields with components and diverse levels in the different facilities, one of these irradiation facilities is the Thermal Column where the Hohlraum is. In this work it was carried out a characterization of the neutron flux inside the 'Hohlraum' of the irradiation facility Thermal Column of the TRIGA Mark III reactor of the Nuclear Center of Mexico to 1 MW of power. It was determined the sub cadmic neutron flux and the epi cadmic by means of the neutron activation technique of thin sheets of gold. The maps of the distribution of the neutron flux for both energy groups in three different positions inside the 'Hohlraum' are presented, these maps were obtained by means of the irradiation of undressed thin activation sheets of gold and covered with cadmium in arrangements of 10 x 12, located parallel to 11.5 cm, 40.5 cm and 70.5 cm to the internal wall of graphite of the installation in inverse address to the position of the reactor core. Starting from the obtained values of neutron flux it was found that, for the same position of the surface of irradiation of the experimental arrangement, the relative differences among the values of neutron flux can be of 80%, and that the differences among different positions of the irradiation surfaces can vary until in a one order of magnitude. (Author)

  11. Protective coatings for very high temperature reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Cabet, C.; Guerre, C. [Service de la Corrosion et du Comportement des Materiaux dans leur Environnement, DEN/DANS/DPC, CEA Saclay, 91191 Gif sur Yvette (France); Thieblemont, F. [Optoelectronics Materials Laboratory, Materials and Interfaces Department, Weizmann Institute of Science, Rehovot (Israel)

    2008-07-15

    The future very high temperature reactors (VHTR) are nuclear systems that shall operate at a maximum temperature of about 950 C. Primary circuit materials thus require good creep and corrosion resistance on very long time. Use of high-strength alloys with protective coatings could significantly improve the service life of high temperature reactor components. However, coating systems are mainly designed for shorter term purposes, often under extremely aggressive atmospheres, that cannot be extrapolated to the VHTR environment. We present our first investigations on the environmental resistance of Alloy 800H coated with two different protective systems under VHTR representative conditions: NiAl(Pt)/EBPVD ZrO{sub 2}(Y) and NiCrAl(Y)/CVD ZrO{sub 2}(Y). Isothermal exposures were carried out up to 1000 h at 950 C in impure helium. This specific atmosphere was shown to induce formation of a surface oxide scale together with carburisation of the bare Alloy 800H. After high temperature exposure to impure helium, the microstructure of the coated specimens has changed due to both thermal ageing and corrosion. Performances of the two coating systems are compared regarding the VHTR application. (Abstract Copyright [2008], Wiley Periodicals, Inc.)

  12. Reactor

    Science.gov (United States)

    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.

  13. Detailed studies of Minor Actinide transmutation-incineration in high-intensity neutron fluxes

    Energy Technology Data Exchange (ETDEWEB)

    Bringer, O. [CEA/Saclay/DSM/DAPNIA, Gif-sur-Yvette (France); Al Mahamid, I. [Lawrence Berkeley National Laboratory, E.H. and S. Div., CA (United States); Blandin, C. [CEA/Cadarache/DEN/DER/SPEX, Saint-Paul-lez-Durances (France); Chabod, S. [CEA/Saclay/DSM/DAPNIA, Gif-sur-Yvette (France); Chartier, F. [CEA/Cadarache/DEN/DPC/SECR, Gif-sur-Yvette (France); Dupont, E.; Fioni, G. [CEA/Saclay/DSM/DAPNIA, Gif-sur-Yvette (France); Isnard, H. [CEA/Cadarache/DEN/DPC/SECR, Gif-sur-Yvette (France); Letourneau, A.; Marie, F. [CEA/Saclay/DSM/DAPNIA, Gif-sur-Yvette (France); Mutti, P. [Institut Laue-Langevin, Grenoble (France); Oriol, L. [CEA/Cadarache/DEN/DER/SPEX, Saint-Paul-lez-Durances (France); Panebianco, S.; Veyssiere, C. [CEA/Saclay/DSM/DAPNIA, Gif-sur-Yvette (France)

    2006-07-01

    The Mini-INCA project is dedicated to the measurement of incineration-transmutation chains and potentials of minor actinides in high-intensity thermal neutron fluxes. In this context, new types of detectors and methods of analysis have been developed. The {sup 241}Am and {sup 232}Th transmutation-incineration chains have been studied and several capture and fission cross sections measured very precisely, showing some discrepancies with existing data or evaluated data. An impact study was made on different based-like GEN-IV reactors. It underlines the necessity to proceed to precise measurements for a large number of minor-actinides that contribute to these future incineration scenarios. (authors)

  14. Multi-scale Control and Enhancement of Reactor Boiling Heat Flux by Reagents and Nanoparticles

    Energy Technology Data Exchange (ETDEWEB)

    Manglik, R M; Athavale, A; Kalaikadal, D S; Deodhar, A; Verma, U

    2011-09-02

    The phenomenological characterization of the use of non-invasive and passive techniques to enhance the boiling heat transfer in water has been carried out in this extended study. It provides fundamental enhanced heat transfer data for nucleate boiling and discusses the associated physics with the aim of addressing future and next-generation reactor thermal-hydraulic management. It essentially addresses the hypothesis that in phase-change processes during boiling, the primary mechanisms can be related to the liquid-vapor interfacial tension and surface wetting at the solidliquid interface. These interfacial characteristics can be significantly altered and decoupled by introducing small quantities of additives in water, such as surface-active polymers, surfactants, and nanoparticles. The changes are fundamentally caused at a molecular-scale by the relative bulk molecular dynamics and adsorption-desorption of the additive at the liquid-vapor interface, and its physisorption and electrokinetics at the liquid-solid interface. At the micro-scale, the transient transport mechanisms at the solid-liquid-vapor interface during nucleation and bubblegrowth can be attributed to thin-film spreading, surface-micro-cavity activation, and micro-layer evaporation. Furthermore at the macro-scale, the heat transport is in turn governed by the bubble growth and distribution, macro-layer heat transfer, bubble dynamics (bubble coalescence, collapse, break-up, and translation), and liquid rheology. Some of these behaviors and processes are measured and characterized in this study, the outcomes of which advance the concomitant fundamental physics, as well as provide insights for developing control strategies for the molecular-scale manipulation of interfacial tension and surface wetting in boiling by means of polymeric reagents, surfactants, and other soluble surface-active additives.

  15. Aging study of boiling water reactor high pressure injection systems

    Energy Technology Data Exchange (ETDEWEB)

    Conley, D.A.; Edson, J.L.; Fineman, C.F. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1995-03-01

    The purpose of high pressure injection systems is to maintain an adequate coolant level in reactor pressure vessels, so that the fuel cladding temperature does not exceed 1,200{degrees}C (2,200{degrees}F), and to permit plant shutdown during a variety of design basis loss-of-coolant accidents. This report presents the results of a study on aging performed for high pressure injection systems of boiling water reactor plants in the United States. The purpose of the study was to identify and evaluate the effects of aging and the effectiveness of testing and maintenance in detecting and mitigating aging degradation. Guidelines from the United States Nuclear Regulatory Commission`s Nuclear Plant Aging Research Program were used in performing the aging study. Review and analysis of the failures reported in databases such as Nuclear Power Experience, Licensee Event Reports, and the Nuclear Plant Reliability Data System, along with plant-specific maintenance records databases, are included in this report to provide the information required to identify aging stressors, failure modes, and failure causes. Several probabilistic risk assessments were reviewed to identify risk-significant components in high pressure injection systems. Testing, maintenance, specific safety issues, and codes and standards are also discussed.

  16. High temperature ceramic membrane reactors for coal liquid upgrading

    Energy Technology Data Exchange (ETDEWEB)

    Tsotsis, T.T. (University of Southern California, Los Angeles, CA (United States). Dept. of Chemical Engineering); Liu, P.K.T. (Aluminum Co. of America, Pittsburgh, PA (United States)); Webster, I.A. (Unocal Corp., Los Angeles, CA (United States))

    1992-01-01

    Membrane reactors are today finding extensive applications for gas and vapor phase catalytic reactions (see discussion in the introduction and recent reviews by Armor [92], Hsieh [93] and Tsotsis et al. [941]). There have not been any published reports, however, of their use in high pressure and temperature liquid-phase applications. The idea to apply membrane reactor technology to coal liquid upgrading has resulted from a series of experimental investigations by our group of petroleum and coal asphaltene transport through model membranes. Coal liquids contain polycyclic aromatic compounds, which not only present potential difficulties in upgrading, storage and coprocessing, but are also bioactive. Direct coal liquefaction is perceived today as a two-stage process, which involves a first stage of thermal (or catalytic) dissolution of coal, followed by a second stage, in which the resulting products of the first stage are catalytically upgraded. Even in the presence of hydrogen, the oil products of the second stage are thought to equilibrate with the heavier (asphaltenic and preasphaltenic) components found in the feedstream. The possibility exists for this smaller molecular fraction to recondense with the unreacted heavy components and form even heavier undesirable components like char and coke. One way to diminish these regressive reactions is to selectively remove these smaller molecular weight fractions once they are formed and prior to recondensation. This can, at least in principle, be accomplished through the use of high temperature membrane reactors, using ceramic membranes which are permselective for the desired products of the coal liquid upgrading process. An additional incentive to do so is in order to eliminate the further hydrogenation and hydrocracking of liquid products to undesirable light gases.

  17. Determination of nitrogen in wheat flour through Activation analysis using Fast neutron flux of a Thermal nuclear reactor; Determinacion de nitrogeno en harina de trigo mediante analisis por activacion empleando el flujo de neutrones rapidos de un reactor nuclear termico

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez G, T

    1976-07-01

    In this work is done a technical study for determining Nitrogen (protein) and other elements in wheat flour Activation analysis, with Fast neutrons from a Thermal nuclear reactor. Initially it is given an introduction about the basic principles of the methods of analysis. Equipment used in Activation analysis and a brief description of the neutron source (Thermal nuclear reactor). The realized experiments for determining the flux form in the irradiation site, the half life of N-13 and the interferences due to the sample composition are included too. Finally, the obtained results by Activation and the Kjeldahl method are tabulated. (Author)

  18. NEET Enhanced Micro Pocket Fission Detector for High Temperature Reactors - FY15 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Unruh, Troy [Idaho National Lab. (INL), Idaho Falls, ID (United States); McGregor, Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ugorowski, Phil [Idaho National Lab. (INL), Idaho Falls, ID (United States); Reichenberger, Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ito, Takashi [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    A new project, that is a collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Atomic Energy Agency, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), has been initiated by the Nuclear Energy Enabling Technologies (NEET) Advanced Sensors and Instrumentation (ASI) program for developing and testing High Temperature Micro-Pocket Fission Detectors (HT MPFD), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package for temperatures up to 800 °C. The MPFD technology utilizes a small, multi-purpose, robust, in-core parallel plate fission chamber and thermocouple. As discussed within this report, the small size, variable sensitivity, and increased accuracy of the MPFD technology represent a revolutionary improvement over current methods used to support irradiations in US Material Test Reactors (MTRs). Previous research conducted through NEET ASI1-3 has shown that the MPFD technology could be made robust and was successfully tested in a reactor core. This new project will further the MPFD technology for higher temperature regimes and other reactor applications by developing a HT MPFD suitable for temperatures up to 800 °C. This report summarizes the research progress for year one of this three year project. Highlights from research accomplishments include: A joint collaboration was initiated between INL, KSU, and CEA. Note that CEA is participating at their own expense because of interest in this unique new sensor. An updated HT MPFD design was developed. New high temperature-compatible materials for HT MPFD construction were procured. Construction methods to support the new design were evaluated at INL. Laboratory evaluations of HT MPFD were initiated. Electrical contact and fissile material plating has been performed at KSU. Updated detector electronics are undergoing evaluations at KSU. A

  19. Reactor Neutrinos

    OpenAIRE

    Soo-Bong Kim; Thierry Lasserre; Yifang Wang

    2013-01-01

    We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very ...

  20. High Temperature Gas-Cooled Test Reactor Options Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    Preliminary scoping calculations are being performed for a 100 MWt gas-cooled test reactor. The initial design uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to identify some reactor design features to investigate further. Current status of the effort is described.

  1. Laboratory Measurement of Enthalpy Flux in High Winds

    Science.gov (United States)

    Jeong, D.; Haus, B. K.; Donelan, M. A.; Zhang, J.

    2006-12-01

    The intensity of tropical cyclones is sensitive to the rates at which enthalpy and momentum are transferred between sea and air in the high-wind core of the storm. Present models of the wind dependence of these transfer rates, does not allow for storms of greater than marginal hurricane intensity. Recent studies have shown that there is a saturation of the bulk drag coefficient in high winds, however more information on the enthalpy flux is required. In particular the role that sea spray plays in enhancing the enthalpy transfer at very high wind speeds is not known. The coefficients for sensible and latent heat transfer (Stanton and Dalton numbers) were measured in the 15-m wind-wave facility at the University of Miami's Air-Sea Interaction Saltwater Tank (ASIST). The wind speed (referred to 10m) was explored over a range of 0 to 45 m/s, covering a full range of aerodynamic conditions from smooth to fully rough. Experiments were designed with water temperatures set between 2 and 5° C above/below the air temperature, with precision thermistors (± 0.002° C) to monitor temperature and Li-Cor infra-red absorption devices to monitor specific humidity changes at upstream and downstream ends of the wave tank during the experiment. The calorimetric use of a wind-wave tank gave precise flux estimates, and experiments were repeated at different Bowen ratios to allow the separation of the heat and moisture parts of the transfer. The effect of spray on the moisture flux was reflected in the drop in temperature along the air path from upstream to downstream and this made it possible to estimate the total spray evaporated in the air column.

  2. Operational characteristics of the high flux plasma generator Magnum-PSI

    Energy Technology Data Exchange (ETDEWEB)

    Eck, H.J.N. van, E-mail: h.j.n.vaneck@differ.nl [FOM Institute DIFFER, Dutch Institute For Fundamental Energy Research, Association EURATOM-FOM, Trilateral Euregio Cluster, P.O. Box 1207, 3430 BE Nieuwegein (Netherlands); Abrams, T. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Berg, M.A. van den; Brons, S.; Eden, G.G. van [FOM Institute DIFFER, Dutch Institute For Fundamental Energy Research, Association EURATOM-FOM, Trilateral Euregio Cluster, P.O. Box 1207, 3430 BE Nieuwegein (Netherlands); Jaworski, M.A.; Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Meiden, H.J. van der; Morgan, T.W.; Pol, M.J. van de; Scholten, J.; Smeets, P.H.M.; De Temmerman, G.; Vries, P.C. de; Zeijlmans van Emmichoven, P.A. [FOM Institute DIFFER, Dutch Institute For Fundamental Energy Research, Association EURATOM-FOM, Trilateral Euregio Cluster, P.O. Box 1207, 3430 BE Nieuwegein (Netherlands)

    2014-10-15

    Highlights: •We have described the design and capabilities of the plasma experiment Magnum-PSI. •The plasma conditions are well suited for PSI studies in support of ITER. •Quasi steady state heat fluxes over 10 MW m{sup −2} have been achieved. •Transient heat and particle loads can be generated to simulate ELM instabilities. •Lithium coating can be applied to the surfaces of samples under vacuum. -- Abstract: In Magnum-PSI (MAgnetized plasma Generator and NUMerical modeling for Plasma Surface Interactions), the high density, low temperature plasma of a wall stabilized dc cascaded arc is confined to a magnetized plasma beam by a quasi-steady state axial magnetic field up to 1.3 T. It aims at conditions that enable fundamental studies of plasma–surface interactions in the regime relevant for fusion reactors such as ITER: 10{sup 23}–10{sup 25} m{sup −2} s{sup −1} hydrogen plasma flux densities at 1–5 eV. To study the effects of transient heat loads on a plasma-facing surface, a high power pulsed magnetized arc discharge has been developed. Additionally, the target surface can be transiently heated with a pulsed laser system during plasma exposure. In this contribution, the current status, capabilities and performance of Magnum-PSI are presented.

  3. Three new nondestructive evaluation tools based on high flux neutron sources

    Energy Technology Data Exchange (ETDEWEB)

    Hubbard, C.R.; Raine, D.; Peascoe, R.; Wright, M. [and others

    1997-03-01

    Nondestructive evaluation methods and systems based on specific attributes of neutron interactions with materials are being developed. The special attributes of neutrons are low attenuation in most engineering materials, strong interaction with low Z elements, and epithermal neutron absorption resonances. The three methods under development at ORNL include neutron based tomography and radiography; through thickness, nondestructive texture mapping; and internal, noninvasive temperature measurement. All three techniques require high flux sources such as the High Flux Isotope Reactor, a steady state source, or the Oak Ridge Electron Linear Accelerator, a pulsed neutron source. Neutrons are quite penetrating in most engineering materials and thus can be useful to detect internal flaws and features. Hydrogen atoms, such as in a hydrocarbon fuel, lubricant, or a metal hydride, are relatively opaque to neutron transmission and thus neutron based tomography/radiography is ideal to image their presence. Texture, the nonrandom orientation of crystalline grains within materials, can be mapped nondestructively using neutron diffraction methods. Epithermal neutron resonance absorption is being studied as a noncontacting temperature sensor. This paper highlights the underlying physics of the methods, progress in development, and the potential benefits for science and industry of the three facilities.

  4. Enzymatically active high-flux selectively gas-permeable membranes

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, Ying-Bing; Cecchi, Joseph L.; Rempe, Susan; FU, Yaqin; Brinker, C. Jeffrey

    2016-01-26

    An ultra-thin, catalyzed liquid transport medium-based membrane structure fabricated with a porous supporting substrate may be used for separating an object species such as a carbon dioxide object species. Carbon dioxide flux through this membrane structures may be several orders of magnitude higher than traditional polymer membranes with a high selectivity to carbon dioxide. Other gases such as molecular oxygen, molecular hydrogen, and other species including non-gaseous species, for example ionic materials, may be separated using variations to the membrane discussed.

  5. High-flux solar photon processes: Opportunities for applications

    Energy Technology Data Exchange (ETDEWEB)

    Steinfeld, J I; Coy, S L; Herzog, H; Shorter, J A; Schlamp, M; Tester, J W; Peters, W A [Massachusetts Inst. of Tech., Cambridge, MA (United States)

    1992-06-01

    The overall goal of this study was to identify new high-flux solar photon (HFSP) processes that show promise of being feasible and in the national interest. Electric power generation and hazardous waste destruction were excluded from this study at sponsor request. Our overall conclusion is that there is promise for new applications of concentrated solar photons, especially in certain aspects of materials processing and premium materials synthesis. Evaluation of the full potential of these and other possible applications, including opportunities for commercialization, requires further research and testing. 100 refs.

  6. BVOC ecosystem flux measurements at a high latitude wetland site

    Directory of Open Access Journals (Sweden)

    T. Holst

    2008-12-01

    Full Text Available In this study, we present summertime concentrations and fluxes of biogenic volatile organic compounds (BVOCs measured at a sub-arctic wetland in northern Sweden using a disjunct eddy-covariance (DEC technique based on a proton transfer reaction mass spectrometer (PTR-MS. The vegetation at the site was dominated by Sphagnum, Carex and Eriophorum spp. The performance of the DEC system was assessed by comparing H3O+-ion cluster formed with water molecules (H3O+(H2O at m37 with water vapour concentration measurements made using an adjacent humidity sensor, and from a comparison of sensible heat fluxes for high frequency and DEC data obtained from the sonic anemometer. These analyses showed no significant PTR-MS sensor drift over a period of several weeks and only a small flux-loss due to high-frequency spectrum omissions. This loss was within the range expected from other studies and the theoretical considerations.

    Standardised (20°C and 1000 μmol m−2 s−1 PAR summer isoprene emission rates of 323 μg C m−2 (ground area h−1 were comparable with findings from more southern boreal forests, and fen-like ecosystems. On a diel scale, measured fluxes indicated a stronger temperature dependence when compared with emissions from temperate or (subtropical ecosystems. For the first time, to our knowledge, we report ecosystem methanol fluxes from a sub-arctic ecosystem. Maximum daytime emission fluxes were around 270 μg m−2 h−1 (ca. 100 μg C m−2 h-1 and measurements indicated some nocturnal deposition.

    The measurements reported here covered a period of 50 days (1 August to 19 September 2006, approximately one half of the growing season at the site, and allowed to investigate the effect of vegetation senescence on daily BVOC fluxes and on their temperature and light

  7. High-flux solar furnace processing of silicon solar cells

    Energy Technology Data Exchange (ETDEWEB)

    Tsuo, Y.S.; Pitts, J.R.; Landry, M.D.; Menna, P.; Bingham, C.E.; Lewandowski, A.; Ciszek, T.F. [National Renewable Energy Laboratory, Golden, CO (United States)

    1996-06-10

    We used a 10-kW, high-flux solar furnace (HFSF) to diffuse the front-surface n{sup +}-p junction and the back-surface p-p{sup +} junction of single-crystal silicon solar cells in one processing step. We found that all of these HFSF-processed cells have better conversion efficiencies than control cells of identical structures fabricated by conventional furnace diffusion methods. We also used the HFSF to crystallize a-Si:H thin films on glass, to texture crystalline silicon surfaces, to deposit gold contacts on silicon wafers, and to getter impurities from metallurgical grade silicon. HFSF processing offers several advantages over conventional furnace processing: (1) it provides a cold-wall process, which reduces contamination; (2) temperature versus time profiles can be precisely controlled; (3) wavelength, intensity, and spatial distribution of the incident solar flux can be controlled and changed rapidly; (4) a number of high-temperature processing steps can be performed simultaneously; and (5) combined quantum and thermal effects may benefit overall cell performance. We conclude that HFSF processing of silicon solar cells has the potential to improve cell efficiency, reduce cell fabrication costs, and also be an environmentally friendly manufacturing method. We have also demonstrated that the HFSF can be used to achieve solid-phase crystallization of a-Si:H at very high speed

  8. Blistering on tungsten surface exposed to high flux deuterium plasma

    Science.gov (United States)

    Xu, H. Y.; Liu, W.; Luo, G. N.; Yuan, Y.; Jia, Y. Z.; Fu, B. Q.; De Temmerman, G.

    2016-04-01

    The blistering behaviour of tungsten surfaces exposed to very high fluxes (1-2 × 1024/m2/s) of low energy (38 eV) deuterium plasmas was investigated as a function of ion fluence (0.2-7 × 1026 D/m2) and surface temperature (423-873 K). Blisters were observed under all conditions, especially up to temperatures of 873 K. The blister parameters are evaluated with blister size, blister density and surface coverage. The blister size always peaked at less than 0.5 μm and no blister larger than 10 μm is observed even at high fluence. The blister densities are found in high magnitude of 106 blisters/m2, with the surface coverages lower than 2%. The formation of cracks in the sub-surface region was observed by cross-section imaging. Changes in blister size and shape with fluence and temperature suggest processes of predominantly nucleation and subsequent growth of blisters. The smaller blister size is considered to be caused by a combination of flux-related effects such as enhanced defect formation in the near surface region, reduced deuterium diffusivity and relatively short exposure times.

  9. Effects of Low Energy and High Flux Helium/Hydrogen Plasma Irradiation on Tungsten as Plasma Facing Material

    Institute of Scientific and Technical Information of China (English)

    Ye Minyou

    2005-01-01

    The High-Z material tungsten (W) has been considered as a plasma facing material in the divertor region of ITER (International Thermonuclear Experimental Reactor). In ITER, the divertor is expected to operate under high particle fluxes (> 1023 m-2s-1) from the plasma as well as from intrinsic impurities with a very low energy (< 200 eV). During the past dacade, the effects of plasma irradiation on tungsten have been studied extensively as functions of the ion energy,fluence and surface temperature in the burning plasma conditions. In this paper, recent results concerning blister and bubble formations on the tungsten surface under low energy (< 100 eV) and high flux (> 1021 m-2s-1) He/H plasma irradiation are reviewed to gain a better understanding of the performance of tungsten as a plasma facing material under the burning plasma conditions.

  10. Neutron flux calculation and fluence in the encircling of the core and vessel of a reactor BWR; Calculo de flujo neutronico y fluencia en la envolvente del nucleo y la vasija de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Martinez C, E.

    2011-07-01

    One of the main objectives related to the safety of any nuclear power plant, including the nuclear power plant of Laguna Verde is to ensure the structural integrity of reactor pressure vessel. To identify and quantify the damage caused by neutron irradiation in the vessel of any nuclear reactor, it is necessary to know both the neutron flux and the neutron fluence that the vessel has been receiving during its operation lifetime, and that the damage observed by mechanical testing are products of microstructural effects induced by neutron irradiation; therefore, it is important the study and prediction of the neutron flux in order to have a better understanding of the damage that these materials are receiving. The calculation here described uses the DORT code, which solves the neutron transport equation in discrete ordinates in two dimensions (x-y, r-{theta} and r-z), according to a regulatory guide, it should make an approximation of the neutron flux in three dimensions by the so called synthesis method. It is called in that way because it achieves a representation of 3 Dimensional neutron flux combining or summarizing the fluxes calculated by DORT r-{theta}, r-z and r. This work presents the application of synthesis method, according to Regulatory Guide 1190, to determine the 3 Dimensional fluxes in internal BWR reactor using three different spatial meshes. The results of the neutron flux and fluence, using three different meshes in the directions r, {theta} and z were compared with results reported in the literature obtaining a difference not larger than 9.61%, neutron flux reached its maximum, 1.58 E + 12 n/cm{sup 2}s, at a height H 4 (239.07 cm) and angle 32.236{sup o} in the core shroud and 4.00 E + 09 n/cm{sup 2}s at a height H 4 and angle 35.27{sup o} in the inner wall of the reactor vessel, positions that are consistent to within {+-}10% over the ones reported in the literature. (Author)

  11. High-flux solar furnace processing of silicon solar cells

    Energy Technology Data Exchange (ETDEWEB)

    Tsuo, Y.S.; Pitts, J.R.; Landry, M.D.; Bingham, C.E.; Lewandowski, A.; Ciszek, T.F. [National Renewable Energy Lab., Golden, CO (United States)

    1994-12-31

    The authors used a 10-kW high-flux solar furnace (HFSF) to diffuse the front-surface n{sup +}-p junction and the back-surface p-p{sup +} junction of single-crystal silicon solar cells in one processing step. They found that all of the HFSF-processed cells have better conversion efficiencies than control cells of identical structures fabricated by conventional furnace diffusion methods. HFSF processing offers several advantages that may contribute to improved solar cell efficiency: (1) it provides a cold-wall process, which reduces contamination; (2) temperature versus time profiles can be precisely controlled; (3) wavelength, intensity, and spatial distribution of the incident solar flux can be controlled and changed rapidly; (4) a number of high-temperature processing steps can be performed simultaneously; and (5) combined quantum and thermal effects may benefit overall cell performance. The HFSF has also been successfully used to texture the surface of silicon wafers and to crystallize a-Si:H thin films on glass.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  13. R D activities at Argonne National Laboratory for the application of base seismic isolation in nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    Seidensticker, R.W.

    1991-01-01

    Argonne National Laboratory (ANL) has been deeply involved in the development of seismic isolation for use in nuclear facilities for the past decade. Initial focus of these efforts has been on the use of seismic isolation for advanced liquid metal reactors (LMR). Subsequent efforts in seismic isolation at ANL included a lead role in an accelerated development program for possible use of seismic isolation for the DOE's New Production reactors (NPR). Under funding provided by the National Science Foundation (NSF) Argonne is currently working with Shimizu in a joint United States-Japanese program on response of seismically-isolated buildings to actual earthquakes. The results of recent work in the seismic isolation program elements are described in this paper. The current Status of these programs is presented along with an assessment of work still needed to bring the benefits of this emerging technology to full potential in nuclear reactors and other nuclear facilities. 38 refs., 3 figs.

  14. High Temperature Fission Chamber for He- and FLiBe-cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bell, Zane W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Giuliano, Dominic R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holcomb, David Eugene [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lance, Michael J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Miller, Roger G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Warmack, Robert J. Bruce [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Mark J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    We have evaluated candidate technologies for in-core fission chambers for high-temperature reactors to monitor power level via measurements of neutron flux from start-up through full power at up to 800°C. This research is important because there are no commercially available instruments capable of operating above 550 °C. Component materials and processes were investigated for fission chambers suitable for operation at 800 °C in reactors cooled by molten fluoride salt (FLiBe) or flowing He, with an emphasis placed on sensitivity (≥ 1 cps/nv), service lifetime (2 years at full power), and resistance to direct immersion in FLiBe. The latter gives the instrument the ability to survive accidents involving breach of a thimble. The device is envisioned to be a two-gap, three-electrode instrument constructed from concentric nickel-plated alumina cylinders and using a noble gas–nitrogen fill-gas. We report the results of measurements and calculations of the response of fill gasses, impurity migration in nickel alloy, brazing of the alumina insulator, and thermodynamic calculations.

  15. Dialyzer Reuse and Outcomes of High Flux Dialysis.

    Directory of Open Access Journals (Sweden)

    Christos Argyropoulos

    Full Text Available The bulk of randomized trial evidence for the expanding use of High Flux (HF hemodialysis worldwide comes from two randomized controlled trials, one of which (HEMODIALYSIS, HEMO allowed, while the other (Membrane Outcomes Permeability, MPO excluded, the reuse of membranes. It is not known whether dialyzer reuse has a differential impact on outcomes with HF vs low flyx (LF dialyzers.Proportional Hazards Models and Joint Models for longitudinal measures and survival outcomes were used in HEMO to analyze the relationship between β2-microglobulin (β2M concentration, flux, and reuse. Meta-analysis and regression techniques were used to synthesize the evidence for HF dialysis from HEMO and MPO.In HEMO, minimally reused (< 6 times HF dialyzers were associated with a hazard ratio (HR of 0.67 (95% confidence interval, 95%CI: 0.48-0.92, p = 0.015, 0.64 (95%CI: 0.44 - 0.95, p = 0.03, 0.61 (95%CI: 0.41 - 0.90, p = 0.012, 0.53 (95%CI: 0.28 - 1.02, p = 0.057 relative to minimally reused LF ones for all cause, cardiovascular, cardiac and infectious mortality respectively. These relationships reversed for extensively reused membranes (p for interaction between reuse and flux < 0.001, p = 0.005 for death from all cause and cardiovascular causes, while similar trends were noted for cardiac and infectious mortality (p of interaction between reuse and flux of 0.10 and 0.08 respectively. Reduction of β2M explained only 1/3 of the effect of minimally reused HF dialyzers on all cause mortality, while non-β2M related factors explained the apparent attenuation of the benefit with more extensively reused dialyzers. Meta-regression of HEMO and MPO estimated an adjusted HR of 0.63 (95% CI: 0.51-0.78 for non-reused HF dialyzers compared with non-reused LF membranes.This secondary analysis and synthesis of two large hemodialysis trials supports the widespread use of HF dialyzers in clinical hemodialysis over the last decade. A mechanistic understanding of the effects of

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

    Science.gov (United States)

    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.

  17. A Catalytically Active Membrane Reactor for Fast, Highly Exothermic, Heterogeneous Gas Reactions. A Pilot Plant Study

    NARCIS (Netherlands)

    Veldsink, Jan W.; Versteeg, Geert F.; Swaaij, Wim P.M. van

    1995-01-01

    Membrane reactors have been frequently studied because of their ability to combine chemical activity and separation properties into one device. Due to their thermal stability and mechanical strength, ceramic membranes are preferred over polymeric ones, but small transmembrane fluxes obstruct a wides

  18. Engineered Coatings for Ni Alloys in High Temperature Reactors

    Science.gov (United States)

    Clark, Elizabeth A.; Yang, James Y.; Kumar, Deepak; Was, Gary S.; Levi, Carlos G.

    2013-02-01

    Alloy 617 is a candidate material for the intermediate heat exchanger of the He-cooled very high temperature reactor. At target temperatures ≥1223 K (950 °C), low level impurities in the gas stream may cause carburization, decarburization, and/or oxidation of 617 with deleterious effects on its mechanical properties. The chromia scale formed naturally by 617 does not provide adequate protection in the expected environment. Alpha alumina offers a greater potential as an effective diffusion barrier with superior stability, but it requires modification of the alloy surface. This work explores two approaches to surface modification based on aluminizing, either alone or in combination with FeCrAlY cladding, followed by pre-oxidation to form alpha alumina. Both approaches yield coatings with promising diffusional stability on alloy 617. Initial corrosion studies in impure He environments reveal that the alpha alumina is stable and protects the underlying substrate in both carburizing and decarburizing environments.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  20. Design and Calibration of a Novel High Temperature Heat Flux Sensor

    OpenAIRE

    Raphael-Mabel, Sujay Anand

    2005-01-01

    Heat flux gages are important in applications where measurement of the transfer of energy is more important than measurement of the temperature itself. There is a need for a heat flux sensor that can perform reliably for long periods of time in high temperature and high heat flux environment. The primary objective is to design and build a heat flux sensor that is capable of operating for extended periods of time in a high heat flux and high temperature environment. A High Temperature Heat F...

  1. High-energy tritium beams as current drivers in tokamak reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mikkelsen, D.R.; Grisham, L.R.

    1983-04-01

    The effect on neutral-beam design and reactor performance of using high-energy (approx. 3-10 MeV) tritium neutral beams to drive steady-state tokamak reactors is considered. The lower current of such beams leads to several advantages over lower-energy neutral beams. The major disadvantage is the reduction of the reactor output caused by the lower current-drive efficiency of the high-energy beams.

  2. Deterministic Modeling of the High Temperature Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ortensi, J.; Cogliati, J. J.; Pope, M. A.; Ferrer, R. M.; Ougouag, A. M.

    2010-06-01

    Idaho National Laboratory (INL) is tasked with the development of reactor physics analysis capability of the Next Generation Nuclear Power (NGNP) project. In order to examine INL’s current prismatic reactor deterministic analysis tools, the project is conducting a benchmark exercise based on modeling the High Temperature Test Reactor (HTTR). This exercise entails the development of a model for the initial criticality, a 19 column thin annular core, and the fully loaded core critical condition with 30 columns. Special emphasis is devoted to the annular core modeling, which shares more characteristics with the NGNP base design. The DRAGON code is used in this study because it offers significant ease and versatility in modeling prismatic designs. Despite some geometric limitations, the code performs quite well compared to other lattice physics codes. DRAGON can generate transport solutions via collision probability (CP), method of characteristics (MOC), and discrete ordinates (Sn). A fine group cross section library based on the SHEM 281 energy structure is used in the DRAGON calculations. HEXPEDITE is the hexagonal z full core solver used in this study and is based on the Green’s Function solution of the transverse integrated equations. In addition, two Monte Carlo (MC) based codes, MCNP5 and PSG2/SERPENT, provide benchmarking capability for the DRAGON and the nodal diffusion solver codes. The results from this study show a consistent bias of 2–3% for the core multiplication factor. This systematic error has also been observed in other HTTR benchmark efforts and is well documented in the literature. The ENDF/B VII graphite and U235 cross sections appear to be the main source of the error. The isothermal temperature coefficients calculated with the fully loaded core configuration agree well with other benchmark participants but are 40% higher than the experimental values. This discrepancy with the measurement stems from the fact that during the experiments the

  3. High Flux Commercial Illumination Solution with Intelligent Controls

    Energy Technology Data Exchange (ETDEWEB)

    Camil Ghiu

    2012-04-30

    This report summarizes the work performed at OSRAM SYLVANIA under US Department of Energy contract DE-EE0003241 for developing a high efficiency LED-based luminaire. A novel light engine module (two versions: standard and super), power supply and luminaire mechanical parts were designed and tested. At steady-state, the luminaire luminous flux is 3156 lumens (lm), luminous efficacy 97.4 LPW and CRI (Ra) 88 at a correlated color temperature (CCT) of 3507K. When the luminaire is fitted with the super version of the light engine the efficacy reaches 130 LPW. In addition, the luminaire is provided with an intelligent control network capable of additional energy savings. The technology developed during the course of this project has been incorporated into a family of products. Recently, the first product in the family has been launched.

  4. MCO Membranes: Enhanced Selectivity in High-Flux Class

    Science.gov (United States)

    Boschetti-de-Fierro, Adriana; Voigt, Manuel; Storr, Markus; Krause, Bernd

    2015-12-01

    Novel MCO high-flux membranes for hemodialysis have been developed with optimized permeability, allowing for filtration close to that of the natural kidney. A comprehensive in vitro characterization of the membrane properties by dextran filtration is presented. The sieving profile of pristine membranes, as well as that of membranes exposed to blood for 40 minutes, are described. The effective pore size (Stokes-Einstein radius) was estimated from filtration experiments before and after blood exposure, and results were compared to hydrodynamic radii of middle and large uremic toxins and essential proteins. The results indicate that the tailored pore sizes of the MCO membranes promote removal of large toxins while ensuring the retention of albumin.

  5. High-quality Critical Heat Flux in Horizontally Coiled Tubes

    Institute of Scientific and Technical Information of China (English)

    1995-01-01

    An investigation on the high-quality dryout in two electrically heated coiled tubes with horizontally helix axes is reported.The temperature profiles both along the tube and around the circumference are measured.and it is found that the temperature profiles around the circumference are not identical for the corss-sections at different parts of the coil.The “local condition hypothesis” seems applicable under present conditions,and the critical heat flux qcr decreases with increasing critical quality xcr.The CHF increases as mass velocity and ratio of tube diameter to coil diameter(d/D) increases,and it seems not to be affected hby the system pressure.The CHF is larger with coils than that with straight tubes,and the difference increases with increasing mass velocity and d/D.

  6. Calculation of intermediate neutron flux in the radial reflectors of graphite reactors, comparison with experiments; Calcul du flux de neutrons intermediaires dans les reflecteurs lateraux des piles a graphite. Comparaison avec l'experience

    Energy Technology Data Exchange (ETDEWEB)

    Brisbois, J.; Vergnaud, T.; Oceraies, Y

    1967-12-01

    In a graphite pile, EDF or Inca type reactor, it is necessary to know the value of the intermediate neutron flux at the output of the lateral reflector in order to determine more precisely the neutron flux at the level of ionisation chambers. A sub critical pile of graphite and natural uranium was built, allowing to reconstitute the geometry of the radiation sources and the disposition of inferior and lateral protections of these piles. This pile is supplied with thermal neutrons coming from the Nereide light water type reactor. Some measurements of intermediate neutron flux have been made in this pile in order to establish a formalism for neutron flux calculation in slowing down in a whole core-lateral reflector, from the distribution of the thermal neutrons flux in the core. The flux calculation is done by age theory in three dimensions, in two homogenous media, separated by an axially semi infinite and laterally finite plane. One of these media includes a distribution of source. The constants are modified in order to take into account the presence of empty channels in the stacking. These calculations are done by the Malaga code. The checking of the formalism has been made in a greater complex geometry of these reactors that introduces an uncertainty factor in the comparison of results. We can however tell that we estimate correctly the variation of the intermediate neutrons flux in the core as well as its descending in a holed lateral reflector. The ratio between the calculation and the experiment is inferior to 2 or 3. Most of the time to a factor 2. [French] Dans une pile a graphite, du type EdF ou Inca, il est necessaire de connaitre la valeur du flux de neutrons intermediaires a la sortie du reflecteur lateral, afin de determiner avec plus de precision le flux de neutrons au niveau des chambres d'ionisation. Il a ete construit un empilement sous-critique, graphite uranium naturel, qui permet de reconstituer la geometrie des sources de rayonnement et la

  7. Neutron analysis of the fuel of high temperature nuclear reactors; Analisis neutronico del combustible de reactores nucleares de alta temperatura

    Energy Technology Data Exchange (ETDEWEB)

    Bastida O, G. E.; Francois L, J. L., E-mail: gbo729@yahoo.com.mx [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

    2014-10-15

    In this work a neutron analysis of the fuel of some high temperature nuclear reactors is presented, studying its main features, besides some alternatives of compound fuel by uranium and plutonium, and of coolant: sodium and helium. For this study was necessary the use of a code able to carry out a reliable calculation of the main parameters of the fuel. The use of the Monte Carlo method was convenient to simulate the neutrons transport in the reactor core, which is the base of the Serpent code, with which the calculations will be made for the analysis. (Author)

  8. High-temperature membrane reactors: potential and problems

    NARCIS (Netherlands)

    Saracco, G.; Neomagus, H.W.J.P.; Versteeg, G.F.; Swaaij, van W.P.M.

    1999-01-01

    The most recent literature in the field of membrane reactors is reviewed, four years after an analogous effort of ours (Saracco et al., 1994), describing shortly the potentials of these reactors, which now seem to be well established, and focusing mostly on problems towards practical exploitation. S

  9. High-temperature membrane reactors : potential and problems

    NARCIS (Netherlands)

    Saracco, G.; Neomagus, H.W.J.P.; Versteeg, G.F.; Swaaij, W.P.M. van

    1999-01-01

    The most recent literature in the field of membrane reactors is reviewed, four years after an analogous effort of ours, describing shortly the potentials of these reactors, which now seem to be well established, and focusing mostly on problems towards practical exploitation. Since then, progress has

  10. Development of a computer code for neutronic calculations of a hexagonal lattice of nuclear reactor using the flux expansion nodal method

    Directory of Open Access Journals (Sweden)

    Mohammadnia Meysam

    2013-01-01

    Full Text Available The flux expansion nodal method is a suitable method for considering nodalization effects in node corners. In this paper we used this method to solve the intra-nodal flux analytically. Then, a computer code, named MA.CODE, was developed using the C# programming language. The code is capable of reactor core calculations for hexagonal geometries in two energy groups and three dimensions. The MA.CODE imports two group constants from the WIMS code and calculates the effective multiplication factor, thermal and fast neutron flux in three dimensions, power density, reactivity, and the power peaking factor of each fuel assembly. Some of the code's merits are low calculation time and a user friendly interface. MA.CODE results showed good agreement with IAEA benchmarks, i. e. AER-FCM-101 and AER-FCM-001.

  11. Calculations of the thermal and fast neutron fluxes in the Syrian miniature neutron source reactor using the MCNP-4C code.

    Science.gov (United States)

    Khattab, K; Sulieman, I

    2009-04-01

    The MCNP-4C code, based on the probabilistic approach, was used to model the 3D configuration of the core of the Syrian miniature neutron source reactor (MNSR). The continuous energy neutron cross sections from the ENDF/B-VI library were used to calculate the thermal and fast neutron fluxes in the inner and outer irradiation sites of MNSR. The thermal fluxes in the MNSR inner irradiation sites were also measured experimentally by the multiple foil activation method ((197)Au (n, gamma) (198)Au and (59)Co (n, gamma) (60)Co). The foils were irradiated simultaneously in each of the five MNSR inner irradiation sites to measure the thermal neutron flux and the epithermal index in each site. The calculated and measured results agree well.

  12. High-flux neutron source based on a liquid-lithium target

    Science.gov (United States)

    Halfon, S.; Feinberg, G.; Paul, M.; Arenshtam, A.; Berkovits, D.; Kijel, D.; Nagler, A.; Eliyahu, I.; Silverman, I.

    2013-04-01

    A prototype compact Liquid Lithium Target (LiLiT), able to constitute an accelerator-based intense neutron source, was built. The neutron source is intended for nuclear astrophysical research, boron neutron capture therapy (BNCT) in hospitals and material studies for fusion reactors. The LiLiT setup is presently being commissioned at Soreq Nuclear research Center (SNRC). The lithium target will produce neutrons through the 7Li(p,n)7Be reaction and it will overcome the major problem of removing the thermal power generated by a high-intensity proton beam, necessary for intense neutron flux for the above applications. The liquid-lithium loop of LiLiT is designed to generate a stable lithium jet at high velocity on a concave supporting wall with free surface toward the incident proton beam (up to 10 kW). During off-line tests, liquid lithium was flown through the loop and generated a stable jet at velocity higher than 5 m/s on the concave supporting wall. The target is now under extensive test program using a high-power electron-gun. Up to 2 kW electron beam was applied on the lithium flow at velocity of 4 m/s without any flow instabilities or excessive evaporation. High-intensity proton beam irradiation will take place at SARAF (Soreq Applied Research Accelerator Facility) superconducting linear accelerator currently in commissioning at SNRC.

  13. High Heat Flux Burnout in Subcooled Flow Boiling

    Institute of Scientific and Technical Information of China (English)

    G.P.Celata; M.Cumo; 等

    1995-01-01

    The paper reports the results of an experimental research carried out at the Heat transfer divison of the Energy Department,C.R.Casaccia,on the thermal hydraulic characterization of subcooled flow boiling CHF under typical conditions of thermonuclear fusion reactors.I.e.high liquid velocity and subcooling.The experiment was carried out exploring the following parameters:channel diameter(from 2.5to 8.0 mm),heated length(10 and 15cm) ,liquid velocity (from 2 to 40m/s),exit pressure(from atmospheric to 5.0 MPa),inlet temperature(from 30 to 80℃),channel orientation (vertical and horizontal),A maximum CHF value of 60.6MW/m2 has been obtained under the following conditions:Tin-30°,p=2.5MPa,u=40m/s,D=2.5mm(smooth channel) Turbulence promoters(helically coiled wires)have been employed to further enhance the CHF attainable with subcooled flow boiling.Helically coiled wires allow an increase of 50% of the maximum CHF obtained with smooth channels.

  14. High temperature gas-cooled reactor: gas turbine application study

    Energy Technology Data Exchange (ETDEWEB)

    1980-12-01

    The high-temperature capability of the High-Temperature Gas-Cooled Reactor (HTGR) is a distinguishing characteristic which has long been recognized as significant both within the US and within foreign nuclear energy programs. This high-temperature capability of the HTGR concept leads to increased efficiency in conventional applications and, in addition, makes possible a number of unique applications in both electrical generation and industrial process heat. In particular, coupling the HTGR nuclear heat source to the Brayton (gas turbine) Cycle offers significant potential benefits to operating utilities. This HTGR-GT Application Study documents the effort to evaluate the appropriateness of the HTGR-GT as an HTGR Lead Project. The scope of this effort included evaluation of the HTGR-GT technology, evaluation of potential HTGR-GT markets, assessment of the economics of commercial HTGR-GT plants, and evaluation of the program and expenditures necessary to establish HTGR-GT technology through the completion of the Lead Project.

  15. Spent fuel treatment and mineral waste form development at Argonne National Laboratory-West

    Energy Technology Data Exchange (ETDEWEB)

    Goff, K.M.; Benedict, R.W.; Bateman, K. [Argonne National Lab., Idaho Falls, ID (United States); Lewis, M.A.; Pereira, C. [Argonne National Lab., IL (United States); Musick, C.A. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1996-07-01

    At Argonne National Laboratory-West (ANL-West) there are several thousand kilograms of metallic spent nuclear fuel containing bond sodium. This fuel will be treated in the Fuel Conditioning Facility (FCF) at ANL-West to produce stable waste forms for storage and disposal. Both mineral and metal high-level waste forms will be produced. The mineral waste form will contain the active metal fission products and the transuranics. Cold small-scale waste form testing has been on-going at Argonne in Illinois. Large-scale testing is commencing at ANL-West.

  16. Status of the design concepts for a high fluence fast pulse reactor (HFFPR)

    Energy Technology Data Exchange (ETDEWEB)

    Philbin, J.S.; Nelson, W.E.; Rosenstroch, B.

    1978-10-01

    The report describes progress that has been made on the design of a High Fluence Fast Pulse Reactor (HFFPR) through the end of calendar year 1977. The purpose of this study is to present design concepts for a test reactor capable of accommodating large scale reactor safety tests. These concepts for reactor safety tests are adaptations of reactor concepts developed earlier for DOE/OMA for the conduct of weapon effects tests. The preferred driver core uses fuel similar to that developed for Sandia's ACPR upgrade. It is a BeO/UO/sub 2/ fuel that is gas cooled and has a high volumetric heat capacity. The present version of the design can drive large (217) pin bundles of prototypically enriched mixed oxide fuel well beyond the fuel's boiling point. Applicability to specific reactor safety accident scenarios and subsequent design improvements will be presented in future reports on this subject.

  17. Argonne Bubble Experiment Thermal Model Development II

    Energy Technology Data Exchange (ETDEWEB)

    Buechler, Cynthia Eileen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    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.

  18. A new high-flux solar furnace for high-temperature thermochemical research

    Energy Technology Data Exchange (ETDEWEB)

    Haueter, P.; Seitz, T.; Steinfeld, A. [Paul Scherrer Inst., Villigen (Switzerland). Solar Process Technology Group

    1999-02-01

    A new high-flux solar furnace, capable of delivering up to 40 kW at peak concentration ratios exceeding 5000, is operational at PSI. Its optical design characteristics, main engineering features, and operating performance are described. This solar concentrating facility will be used principally for investigating the thermochemical processing of solar fuels at temperatures as high as 2500 K.

  19. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-01-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  20. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  1. Tritium Formation and Mitigation in High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Carl Stoots

    2012-08-01

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. In order to prevent the tritium contamination of proposed reactor buildings and surrounding sites, this paper examines the root causes and potential solutions for the production of this radionuclide, including materials selection and inert gas sparging. A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750°C. Results of the diffusion model are presented for one steadystate value of tritium production in the reactor.

  2. High photon flux table-top coherent extreme ultraviolet source

    CERN Document Server

    Hädrich, Steffen; Rothhardt, Jan; Krebs, Manuel; Hoffmann, Armin; Pronin, Oleg; Pervak, Vladimir; Limpert, Jens; Tünnermann, Andreas

    2014-01-01

    High harmonic generation (HHG) enables extreme ultraviolet radiation with table-top setups. Its exceptional properties, such as coherence and (sub)-femtosecond pulse durations, have led to a diversity of applications. Some of these require a high photon flux and megahertz repetition rates, e.g. to avoid space charge effects in photoelectron spectroscopy. To date this has only been achieved with enhancement cavities. Here, we establish a novel route towards powerful HHG sources. By achieving phase-matched HHG of a megahertz fibre laser we generate a broad plateau (25 eV - 40 eV) of strong harmonics, each containing more than $10^{12}$ photons/s, which constitutes an increase by more than one order of magnitude in that wavelength range. The strongest harmonic (H25, 30 eV) has an average power of 143 $\\mu$W ($3\\cdot10^{13}$ photons/s). This concept will greatly advance and facilitate applications in photoelectron or coincidence spectroscopy, coherent diffractive imaging or (multidimensional) surface science.

  3. A Compact, High-Flux Cold Atom Beam Source

    Science.gov (United States)

    Kellogg, James R.; Kohel, James M.; Thompson, Robert J.; Aveline, David C.; Yu, Nan; Schlippert, Dennis

    2012-01-01

    The performance of cold atom experiments relying on three-dimensional magneto-optical trap techniques can be greatly enhanced by employing a highflux cold atom beam to obtain high atom loading rates while maintaining low background pressures in the UHV MOT (ultra-high vacuum magneto-optical trap) regions. Several techniques exist for generating slow beams of cold atoms. However, one of the technically simplest approaches is a two-dimensional (2D) MOT. Such an atom source typically employs at least two orthogonal trapping beams, plus an additional longitudinal "push" beam to yield maximum atomic flux. A 2D atom source was created with angled trapping collimators that not only traps atoms in two orthogonal directions, but also provides a longitudinal pushing component that eliminates the need for an additional push beam. This development reduces the overall package size, which in turn, makes the 2D trap simpler, and requires less total optical power. The atom source is more compact than a previously published effort, and has greater than an order of magnitude improved loading performance.

  4. Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor

    Energy Technology Data Exchange (ETDEWEB)

    B. Boer; A. M. Ougouag

    2010-09-01

    The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a ”standard,” UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge

  5. Continuous-flow high pressure hydrogenation reactor for optimization and high-throughput synthesis.

    Science.gov (United States)

    Jones, Richard V; Godorhazy, Lajos; Varga, Norbert; Szalay, Daniel; Urge, Laszlo; Darvas, Ferenc

    2006-01-01

    This paper reports on a novel continuous-flow hydrogenation reactor and its integration with a liquid handler to generate a fully automated high-throughput hydrogenation system for library synthesis. The reactor, named the H-Cube, combines endogenous hydrogen generation from the electrolysis of water with a continuous flow-through system. The system makes significant advances over current batch hydrogenation reactors in terms of safety, reaction validation efficiency, and rates of reaction. The hydrogenation process is described along with a detailed description of the device's main parts. The reduction of a series of functional groups, varying in difficulty up to 70 degrees C and 70 bar are also described. The paper concludes with the integration of the device into an automated liquid handler followed by the reduction of a nitro compound in a high throughput manner. The system is fully automated and can conduct 5 reactions in the time it takes to perform and workup one reaction manually on a standard batch reactor.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. High-throughput DNA droplet assays using picoliter reactor volumes.

    Science.gov (United States)

    Srisa-Art, Monpichar; deMello, Andrew J; Edel, Joshua B

    2007-09-01

    The online characterization and detection of individual droplets at high speeds, low analyte concentrations, and perfect detection efficiencies is a significant challenge underpinning the application of microfluidic droplet reactors to high-throughput chemistry and biology. Herein, we describe the integration of confocal fluorescence spectroscopy as a high-efficiency detection method for droplet-based microfluidics. Issues such as surface contamination, rapid mixing, and rapid detection, as well as low detections limits have been addressed with the approach described when compared to conventional laminar flow-based fluidics. Using such a system, droplet size, droplet shape, droplet formation frequencies, and droplet compositions can be measured accurately and precisely at kilohertz frequencies. Taking advantage of this approach, we demonstrate a high-throughput biological assay based on fluorescence resonance energy transfer (FRET). By attaching a FRET donor (Alexa Fluor 488) to streptavidin and labeling a FRET acceptor (Alexa Fluor 647) on one DNA strand and biotin on the complementary strand, donor and acceptor molecules are brought in proximity due to streptavidin-biotin binding, resulting in FRET. Fluorescence bursts of the donor and acceptor from each droplet can be monitored simultaneously using separate avalanche photodiode detectors operating in single photon counting mode. Binding assays were investigated and compared between fixed streptavidin and DNA concentrations. Binding curves fit perfectly to Hill-Waud models, and the binding ratio between streptavidin and biotin was evaluated and found to be in agreement with the biotin binding sites on streptavidin. FRET efficiency for this FRET pair was also investigated from the binding results. Efficiency results show that this detection system can precisely measure FRET even at low FRET efficiencies.

  8. Flux-measuring approach of high temperature metal liquid based on BP neural networks

    Institute of Scientific and Technical Information of China (English)

    胡燕瑜; 桂卫华; 李勇刚

    2003-01-01

    A soft-measuring approach is presented to measure the flux of liquid zinc with high temperature andcausticity. By constructing mathematical model based on neural networks, weighing the mass of liquid zinc, the fluxof liquid zinc is acquired indirectly, the measuring on line and flux control are realized. Simulation results and indus-trial practice demonstrate that the relative error between the estimated flux value and practical measured flux value islower than 1.5%, meeting the need of industrial process.

  9. Adaptation of a High Frequency Ultrasonic Transducer to the Measurement of Water Temperature in a Nuclear Reactor

    Science.gov (United States)

    Zaz, G.; Calzavara, Y.; Le Clézio, E.; Despaux, G.

    Most high flux reactors possess for research purposes fuel elements composed of plates. Their relative distance is a crucial parameter, particularly concerning the irradiation history. For the High Flux Reactor (RHF) of the Institute Laue-Langevin (ILL), the measurement of this distance with a microscopic resolution becomes extremely challenging. To address this issue, a specific ultrasonic transducer, presented in a first paper, has been designed and manufactured to be inserted into the 1.8 mm width channel existing between curved fuel plates. It was set on a blade yielding a total device thickness of 1 mm. To achieve the expected resolution, the system is excited with frequencies up to 70 MHz and integrated into a set of high frequency acquisition instruments. Thanks to a specific signal processing, this device allows the distance measurement through the evaluation of the ultrasonic wave time of fight. One of the crucial points is then the evaluation of the local water temperature inside the water channel. To obtain a precise estimation of this parameter, the ultrasonic sensor is used as a thermometer thanks to the analysis of the spectral components of the acoustic signal propagating inside the sensor multilayered structure. The feasibility of distance measurement was proved during the December 2013 experiment in the RHF fuel element of the ILL. Some of the results will be presented as well as some experimental constraints identified to improve the accuracy of the measurement in future works.

  10. Design, Test and Demonstration of Saturable Reactor High-Temperature Superconductor Fault Current Limiters

    Energy Technology Data Exchange (ETDEWEB)

    Darmann, Frank [Zenergy Power, Inc., Burlingame, CA (United States); Lombaerde, Robert [Zenergy Power, Inc., Burlingame, CA (United States); Moriconi, Franco [Zenergy Power, Inc., Burlingame, CA (United States); Nelson, Albert [Zenergy Power, Inc., Burlingame, CA (United States)

    2012-03-01

    Zenergy Power has successfully designed, built, tested, and installed in the US electrical grid a saturable reactor Fault Current Limiter. Beginning in 2007, first as SC Power Systems and from 2008 as Zenergy Power, Inc., ZP used DOE matching grant and ARRA funds to help refine the design of the saturated reactor fault current limiter. ZP ultimately perfected the design of the saturated reactor FCL to the point that ZP could reliably design a suitable FCL for most utility applications. Beginning with a very basic FCL design using 1G HTS for a coil housed in a LN2 cryostat for the DC bias magnet, the technology progressed to a commercial system that was offered for sale internationally. Substantial progress was made in two areas. First, the cryogenics cooling system progressed from a sub-cooled liquid nitrogen container housing the HTS coils to cryostats utilizing dry conduction cooling and reaching temperatures down to less than 20 degrees K. Large, round cryostats with warm bore diameters of 1.7 meters enabled the design of large tanks to hold the AC components. Second, the design of the AC part of the FCL was refined from a six legged spider design to a more compact and lighter design with better fault current limiting capability. Further refinement of the flux path and core shape led to an efficient saturated reactor design requiring less Ampere-turns to saturate the core. In conclusion, the development of the saturable reactor FCL led to a more efficient design not requiring HTS magnets and their associated peripheral equipment, which yielded a more economical product in line with the electric utility industry expectations. The original goal for the DOE funding of the ZP project Design, Test and Demonstration of Saturable Reactor High-Temperature Superconductor Fault Current Limiters was to stimulate the HTS wire industry with, first 1G, then 2G, HTS wire applications. Over the approximately 5 years of ZP's product development program, the amount of HTS

  11. Development of high-fidelity multiphysics system for light water reactor analysis

    Science.gov (United States)

    Magedanz, Jeffrey W.

    There has been a tendency in recent years toward greater heterogeneity in reactor cores, due to the use of mixed-oxide (MOX) fuel, burnable absorbers, and longer cycles with consequently higher fuel burnup. The resulting asymmetry of the neutron flux and energy spectrum between regions with different compositions causes a need to account for the directional dependence of the neutron flux, instead of the traditional diffusion approximation. Furthermore, the presence of both MOX and high-burnup fuel in the core increases the complexity of the heat conduction. The heat transfer properties of the fuel pellet change with irradiation, and the thermal and mechanical expansion of the pellet and cladding strongly affect the size of the gap between them, and its consequent thermal resistance. These operational tendencies require higher fidelity multi-physics modeling capabilities, and this need is addressed by the developments performed within this PhD research. The dissertation describes the development of a High-Fidelity Multi-Physics System for Light Water Reactor Analysis. It consists of three coupled codes -- CTF for Thermal Hydraulics, TORT-TD for Neutron Kinetics, and FRAPTRAN for Fuel Performance. It is meant to address these modeling challenges in three ways: (1) by resolving the state of the system at the level of each fuel pin, rather than homogenizing entire fuel assemblies, (2) by using the multi-group Discrete Ordinates method to account for the directional dependence of the neutron flux, and (3) by using a fuel-performance code, rather than a Thermal Hydraulics code's simplified fuel model, to account for the material behavior of the fuel and its feedback to the hydraulic and neutronic behavior of the system. While the first two are improvements, the third, the use of a fuel-performance code for feedback, constitutes an innovation in this PhD project. Also important to this work is the manner in which such coupling is written. While coupling involves combining

  12. A preliminary assessment of the effects of heat flux distribution and penetration on the creep rupture of a reactor vessel lower head

    Energy Technology Data Exchange (ETDEWEB)

    Chu, T.Y.; Bentz, J.; Simpson, R. [Sandia National Labs., Albuquerque, NM (United States); Witt, R. [Univ. of Wisconsin, Madison, WI (United States)

    1997-02-01

    The objective of the Lower Head Failure (LHF) Experiment Program is to experimentally investigate and characterize the failure of the reactor vessel lower head due to thermal and pressure loads under severe accident conditions. The experiment is performed using 1/5-scale models of a typical PWR pressure vessel. Experiments are performed for various internal pressure and imposed heat flux distributions with and without instrumentation guide tube penetrations. The experimental program is complemented by a modest modeling program based on the application of vessel creep rupture codes developed in the TMI Vessel Investigation Project. The first three experiments under the LHF program investigated the creep rupture of simulated reactor pressure vessels without penetrations. The heat flux distributions for the three experiments are uniform (LHF-1), center-peaked (LHF-2), and side-peaked (LHF-3), respectively. For all the experiments, appreciable vessel deformation was observed to initiate at vessel wall temperatures above 900K and the vessel typically failed at approximately 1000K. The size of failure was always observed to be smaller than the heated region. For experiments with non-uniform heat flux distributions, failure typically occurs in the region of peak temperature. A brief discussion of the effect of penetration is also presented.

  13. Two-dimensional DORT discrete ordinates X-Y geometry neutron flux calculations for the Halden Heavy Boiling Water Reactor core configurations

    Energy Technology Data Exchange (ETDEWEB)

    Slater, C.O.

    1990-07-01

    Results are reported for two-dimensional discrete ordinates, X-Y geometry calculations performed for seven Halden Heavy Boiling Water Reactor core configurations. The calculations were performed in support of an effort to reassess the neutron fluence received by the reactor vessel. Nickel foil measurement data indicated considerable underprediction of fluences by the previously used multigroup removal- diffusion method. Therefore, calculations by a more accurate method were deemed appropriate. For each core configuration, data are presented for (1) integral fluxes in the core and near the vessel wall, (2) neutron spectra at selected locations, (3) isoflux contours superimposed on the geometry models, (4) plots of the geometry models, and (5) input for the calculations. The initial calculations were performed with several mesh sizes. Comparisons of the results from these calculations indicated that the uncertainty in the calculated fluxes should be less than 10%. However, three-dimensional effects (such as axial asymmetry in the fuel loading) could contribute to much greater uncertainty in the calculated neutron fluxes. 7 refs., 22 figs., 11 tabs.

  14. Anisotropic flux pinning in high T{sub c} superconductors

    Energy Technology Data Exchange (ETDEWEB)

    Kolesnik, S. [Institute of Physics, Polish Academy of Sciences, Al. Lotnikow 32/46, PL-02668, Warszawa (Poland); Igalson, J. [Institute of Physics, Polish Academy of Sciences, Al. Lotnikow 32/46, PL-02668, Warszawa (Poland); Skoskiewicz, T. [Institute of Physics, Polish Academy of Sciences, Al. Lotnikow 32/46, PL-02668, Warszawa (Poland); Szymczak, R. [Institute of Physics, Polish Academy of Sciences, Al. Lotnikow 32/46, PL-02668, Warszawa (Poland); Baran, M. [Institute of Physics, Polish Academy of Sciences, Al. Lotnikow 32/46, PL-02668, Warszawa (Poland); Pytel, K. [Institute of Atomic Energy, Swierk (Poland); Pytel, B. [Institute of Atomic Energy, Swierk (Poland)

    1995-02-09

    In this paper we present a comparison of the results of FC magnetization measurements on several Pb-Sr-(Y,Ca)-Cu-O crystals representing various levels of flux pinning. The pinning centers in our crystals have been set up during the crystal growth process or introduced by neutron irradiation. Some possible explanations of the observed effects, including surface barrier, flux-center distribution and sample-shape effects, are discussed. ((orig.)).

  15. A test study on treatment of high-strength polyester wastewater with anaerobic reactor

    Institute of Scientific and Technical Information of China (English)

    韩洪军; 陈秀荣; 徐春艳

    2002-01-01

    The treatment of polyester wastewater using Up-flow activated sludge bed anaerobic filer ( UASB-AF), demonstrated that UASB-AF reactors has a high efficiency, its volume loading is 10 ~ 12 kgCOD/( m3 @d) ,HRT is 22 ~24 h, and the removal of COD is about 80%. The reactor has advantage of fast starting andenduring pulse loading.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

    Energy Technology Data Exchange (ETDEWEB)

    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 1979

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-03-01

    The results of the environmental monitoring program at Argonne National Laboratory for 1979 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, Argonne effluent water, soil, grass, bottom sediment, and foodstuffs; for a variety of chemical constituents in air, surface water, and Argonne effluent water; and of the environemetal penetrating radiation dose. Sample collections and measurements were made at the site boundary and off the Argonne site for comparison purposes. Some on-site measuremenets 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 and are compared with applicable environmental quality standards. The potential radiation dose to off-site population groups is also estimated.

  19. Environmental monitoring at Argonne National Laboratory. Annual report for 1983

    Energy Technology Data Exchange (ETDEWEB)

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

    1984-03-01

    The results of the environmental monitoring program at Argonne National Laboratory for 1983 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 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 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. 19 references, 8 figures, 49 tables.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  1. Surface morphology and deuterium retention in tungsten exposed to high flux D plasma at high temperatures

    NARCIS (Netherlands)

    Jia, Y. Z.; De Temmerman, G.; Luo, G. N.; Xu, H.Y.; Li, C.; Fu, B. Q.; Liu, W.

    2015-01-01

    Surface modifications and deuterium retention induced in tungsten by high fluxes (1024 m−2 s−1) low energy (38 eV) deuterium ions were studied as a function of surface temperature. Blister formation was studied by scanning electron microscopy and electron backscatter diffraction, while deuterium ret

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. Measurement of the thermal and fast neutron flux in a research reactor with a Li and Th loaded optical fibre detector

    CERN Document Server

    Yamane, Y; Misawa, T; Karlsson, J K H; Pázsit, I

    1999-01-01

    The spatial dependence of thermal and fast neutron flux was measured axially in the core of a 1 MW research reactor. The measurements were made by a thin optical fibre detector with a neutron sensitive ZnS(Ag) scintillation tip. For thermal neutrons sup 6 Li was used, whereas for fast neutrons sup 2 sup 3 sup 2 Th was used as neutron converter. The spatial dependence was measured by moving the fibre axially with a uniform speed. The measurement takes a few minutes, compared to up to 10 h with the conventional wire activation method. Comparison with traditional measurements shows a good agreement. (author)

  4. Autonomous, high-resolution observations of particle flux in the oligotrophic ocean

    Directory of Open Access Journals (Sweden)

    M. L. Estapa

    2013-08-01

    Full Text Available Observational gaps limit our understanding of particle flux attenuation through the upper mesopelagic because available measurements (sediment traps and radiochemical tracers have limited temporal resolution, are labor-intensive, and require ship support. Here, we conceptually evaluate an autonomous, optical proxy-based method for high-resolution observations of particle flux. We present four continuous records of particle flux collected with autonomous profiling floats in the western Sargasso Sea and the subtropical North Pacific, as well as one shorter record of depth-resolved particle flux near the Bermuda Atlantic Time-series Study (BATS and Oceanic Flux Program (OFP sites. These observations illustrate strong variability in particle flux over very short (~1-day timescales, but at longer timescales they reflect patterns of variability previously recorded during sediment trap time series. While particle flux attenuation at BATS/OFP agreed with the canonical power-law model when observations were averaged over a month, flux attenuation was highly variable on timescales of 1–3 days. Particle fluxes at different depths were decoupled from one another and from particle concentrations and chlorophyll fluorescence in the immediately overlying surface water, consistent with horizontal advection of settling particles. We finally present an approach for calibrating this optical proxy in units of carbon flux, discuss in detail the related, inherent physical and optical assumptions, and look forward toward the requirements for the quantitative application of this method in highly time-resolved studies of particle export and flux attenuation.

  5. Autonomous, high-resolution observations of particle flux in the oligotrophic ocean

    Directory of Open Access Journals (Sweden)

    M. L. Estapa

    2013-01-01

    Full Text Available Observational gaps limit our understanding of particle flux attenuation through the upper mesopelagic because available measurements (sediment traps and radiochemical tracers have limited temporal resolution, are labor-intensive, and require ship support. Here, we conceptually evaluate an autonomous, optical proxy-based method for high-resolution observations of particle flux. We present four continuous records of particle flux collected with autonomous, profiling floats in the western Sargasso Sea and the subtropical North Pacific, as well as one shorter record of depth-resolved particle flux near the Bermuda Atlantic Timeseries Study (BATS and Oceanic Flux Program (OFP sites. These observations illustrate strong variability in particle flux over very short (~1 day timescales, but at longer timescales they reflect patterns of variability previously recorded during sediment trap timeseries. While particle flux attenuation at BATS/OFP agreed with the canonical power-law model when observations were averaged over a month, flux attenuation was highly variable on timescales of 1–3 days. Particle fluxes at different depths were decoupled from one another and from particle concentrations and chlorophyll fluorescence in the immediately-overlying surface water, consistent with horizontal advection of settling particles. We finally present an approach for calibrating this optical proxy in units of carbon flux, discuss in detail the related, inherent physical and optical assumptions, and look forward toward the requirements for the quantitative application of this method in highly time-resolved studies of particle export and flux attenuation.

  6. Modeling and simulation of high-pressure industrial autoclave polyethylene reactor

    Directory of Open Access Journals (Sweden)

    2008-01-01

    Full Text Available High-pressure technology for polyethylene production has been widely used by industries around the world. A good model for the reactor fluid dynamics is essential to set the operating conditions of an autoclave reactor. The high-pressure autoclave reactor model developed in this work was based on a non-isothermal dynamic model, where PID control equations are used to maintain the operation at the unstable steady state. The kinetic mechanism to describe the polymerization rate and molecular weight averages are presented. The model is capable of computing temperature, concentration gradients and polymer characteristics. The model was validated for an existing industrial reactor and data for production of homopolymer polyethylene and has represented well the behavior of the autoclave reactor used in ethylene homopolymerization.

  7. ANALYSIS OF A HIGH TEMPERATURE GAS-COOLED REACTOR POWERED HIGH TEMPERATURE ELECTROLYSIS HYDROGEN PLANT

    Energy Technology Data Exchange (ETDEWEB)

    M. G. McKellar; E. A. Harvego; A. M. Gandrik

    2010-11-01

    An updated reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production has been developed. The HTE plant is powered by a high-temperature gas-cooled reactor (HTGR) whose configuration and operating conditions are based on the latest design parameters planned for the Next Generation Nuclear Plant (NGNP). The current HTGR reference design specifies a reactor power of 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 322°C and 750°C, respectively. The reactor heat is used to produce heat and electric power to the HTE plant. A Rankine steam cycle with a power conversion efficiency of 44.4% was used to provide the electric power. The electrolysis unit used to produce hydrogen includes 1.1 million cells with a per-cell active area of 225 cm2. The reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes a steam-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The overall system thermal-to-hydrogen production efficiency (based on the higher heating value of the produced hydrogen) is 42.8% at a hydrogen production rate of 1.85 kg/s (66 million SCFD) and an oxygen production rate of 14.6 kg/s (33 million SCFD). An economic analysis of this plant was performed with realistic financial and cost estimating The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.03/kg of hydrogen was calculated assuming an internal rate of return of 10% and a debt to equity ratio of 80%/20% for a reactor cost of $2000/kWt and $2.41/kg of hydrogen for a reactor cost of $1400/kWt.

  8. Tritium Formation and Mitigation in High-Temperature Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Carl Stoots; Hans A. Schmutz

    2013-03-01

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. To prevent the tritium contamination of proposed reactor buildings and surrounding sites, this study examines the root causes and potential mitigation strategies for permeation of tritium (such as: materials selection, inert gas sparging, etc...). A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750 degrees C. Results of the diffusion model are presented for a steady production of tritium

  9. Tritium Formation and Mitigation in High-Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Carl Stoots

    2012-10-01

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. To prevent the tritium contamination of proposed reactor buildings and surrounding sites, this study examines the root causes and potential mitigation strategies for permeation of tritium (such as: materials selection, inert gas sparging, etc...). A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750 degrees C. Results of the diffusion model are presented for a steady production of tritium

  10. Fuel element concept for long life high power nuclear reactors

    Science.gov (United States)

    Mcdonald, G. E.; Rom, F. E.

    1969-01-01

    Nuclear reactor fuel elements have burnups that are an order of magnitude higher than can currently be achieved by conventional design practice. Elements have greater time integrated power producing capacity per unit volume. Element design concept capitalizes on known design principles and observed behavior of nuclear fuel.

  11. An Evaluation Report on the High Temperature Design of the KALIMER-600 Reactor Structures

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chang Gyu; Lee, Jae Han

    2007-03-15

    This report is on the validity evaluation of high temperature structural design for the reactor structures and piping of the pool-type Liquid Metal Reactor, KALIMER-600 subjected to the high temperature thermal load condition. The structural concept of the Upper Internal Structure located above the core is analyzed and the adequate UIS conceptual design for KALIMER-600 is proposed. Also, the high temperature structural integrity of the thermal liner which is to protect the UIS bottom plate from the high frequency thermal fatigue damage was evaluated by the thermal stripping analysis. The high temperature structural design of the reactor internal structure by considering the reactor startup-shutdown cycle was carried out and the structural integrity of it for a normal operating condition as well as the transient condition of the primary pump trip accident was confirmed. Additionally the structure design of the reactor internal structural was changed to prevent the non-uniform deformation of the primary pump which is induced by the thermal expansion difference between the reactor head and the baffle plate. The arrangement of the IHTS piping system which is a part of the reactor system is carried out and the structural integrity and the accumulated deformation by considering the reactor startup-shutdown cycle of a normal operating condition were evaluated. The structural integrity and the accumulated deformation of the PDRC hot leg piping by considering the PDRC operating condition were evaluated. The validity of KALIMER-600 high temperature structural design is confirmed through this study, and it is clearly found that the methodology research to evaluate the structural integrity considering the reactor life time of 60 years ensured is necessary.

  12. Surface morphology and deuterium retention in tungsten exposed to high flux D plasma at high temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Jia, Y.Z. [Laboratory of Advanced Materials, School of Materials Science and Engineering, Tsinghua University, Beijing 100084 (China); De Temmerman, G. [FOM Institute DIFFER-Dutch Institute for Fundamental Energy Research, Edisonbaan 14, 3439 MN Nieuwegein (Netherlands); ITER Organization, Route de Vinon-sur-Verdon-CS90046, 13067 St Paul Lez Durance Cedex (France); Luo, G.-N. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Xu, H.Y. [Science and Technology on Surface Physics and Chemistry Laboratory, Mianyang, Sichuan 621907 (China); Li, C.; Fu, B.Q. [Laboratory of Advanced Materials, School of Materials Science and Engineering, Tsinghua University, Beijing 100084 (China); Liu, W., E-mail: liuw@mail.tsinghua.edu.cn [Laboratory of Advanced Materials, School of Materials Science and Engineering, Tsinghua University, Beijing 100084 (China)

    2015-02-15

    Surface modifications and deuterium retention induced in tungsten by high fluxes (10{sup 24} m{sup −2} s{sup −1}) low energy (38 eV) deuterium ions were studied as a function of surface temperature. Blister formation was studied by scanning electron microscopy and electron backscatter diffraction, while deuterium retention was measured by thermal desorption spectroscopy. Blisters are observed on the surface exposed at different temperatures, ranging from 493 K to 1273 K. The blister density and D retention decrease with the increasing exposure temperature. The formation of blisters at high temperatures is attributed to the high flux of D plasma. At 943 K, with the increasing fluence, there is trend to the saturation of D retention and blister density. The defects caused by plasma exposure have an important effect on the D trapping and blistering behavior. The formation of blisters has a strong relationship with slipping system of tungsten.

  13. Study of neutronic flux in IPR-R1 reactor with MCNPX; Estudo do fluxo neutronico no reator IPR-R1 com o MCNPX

    Energy Technology Data Exchange (ETDEWEB)

    Melo, J.A.S.; Castrillo, L.S., E-mail: julio.angelo@poli.br, E-mail: lazara@poli.br [Universidade de Pernambuco (UPE), Recife, PE (Brazil). Escola Politecnica; Oliveira, R.M.B.M., E-mail: romero.matias@educacao.pe.gov.br [Secretaria Executiva de Educacao do Estado de Pernambuco (SEE), Recife, PE (Brazil)

    2016-11-01

    MCNPX computer code, one of the latest versions of code MCNP transport were used to study the flux distribution and its neutronic fluence as a function of energy in two research reactor irradiation IPR-R1. The model developed was validated with research conducted by Dalle (2005). Initially, in the simulation is considered fresh fuel whose core configuration contained three neutron rods control, being two of them 100% ejected while the other inserted 3,1 x 10{sup -1} m deep, as adopted in the literature situation. The neutron source used was the critical type, through KSRC card. The results of the neutron flow and neutronic fluence were obtained in the central tube and the turntable on a range of energy spectrum that ranged from 1.0 x 10{sup -9} MeV to 10 MeV, showing good correlations with the model used in validation. Finally, a hypothetical situation wherein the three reactor control rods are ejected simultaneously was simulated. The simulation results showed an increase in the neutron flux of 7% in the central tube and 5% on the turntable.

  14. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. High-throughput reactor system with individual temperature control for the investigation of monolith catalysts.

    Science.gov (United States)

    Dellamorte, Joseph C; Vijay, Rohit; Snively, Christopher M; Barteau, Mark A; Lauterbach, Jochen

    2007-07-01

    A high-throughput parallel reactor system has been designed and constructed to improve the reliability of results from large diameter catalysts such as monoliths. The system, which is expandable, consists of eight quartz reactors, 23.5 mm in diameter. The eight reactors were designed with separate K type thermocouples and radiant heaters, allowing for the independent measurement and control of each reactor temperature. This design gives steady state temperature distributions over the eight reactors within 0.5 degrees C of a common setpoint from 50 to 700 degrees C. Analysis of the effluent from these reactors is performed using rapid-scan Fourier transform infrared (FTIR) spectroscopic imaging. The integration of this technique to the reactor system allows a chemically specific, truly parallel analysis of the reactor effluents with a time resolution of approximately 8 s. The capabilities of this system were demonstrated via investigation of catalyst preparation conditions on the direct epoxidation of ethylene, i.e., on the ethylene conversion and the ethylene oxide selectivity. The ethylene, ethylene oxide, and carbon dioxide concentrations were calibrated based on spectra from FTIR imaging using univariate and multivariate chemometric techniques. The results from this analysis showed that the calcination conditions significantly affect the ethylene conversion, with a threefold increase in the conversion when the catalyst was calcined for 3 h versus 12 h at 400 degrees C.

  16. Markets for reactor-produced non-fission radioisotopes

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, R.G.

    1995-01-01

    Current market segments for reactor produced radioisotopes are developed and reported from a review of current literature. Specific radioisotopes studied in is report are the primarily selected from those with major medical or industrial markets, or those expected to have strongly emerging markets. Relative market sizes are indicated. Special emphasis is given to those radioisotopes that are best matched to production in high flux reactors such as the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory or the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory. A general bibliography of medical and industrial radioisotope applications, trends, and historical notes is included.

  17. High conversion pressurized water reactor with boiling channels

    Energy Technology Data Exchange (ETDEWEB)

    Margulis, M., E-mail: maratm@post.bgu.ac.il [The Unit of Nuclear Engineering, Ben Gurion University of the Negev, POB 653, Beer Sheva 84105 (Israel); Shwageraus, E., E-mail: es607@cam.ac.uk [Department of Engineering, University of Cambridge, CB2 1PZ Cambridge (United Kingdom)

    2015-10-15

    Highlights: • Conceptual design of partially boiling PWR core was proposed and studied. • Self-sustainable Th–{sup 233}U fuel cycle was utilized in this study. • Seed-blanket fuel assembly lattice optimization was performed. • A coupled Monte Carlo, fuel depletion and thermal-hydraulics studies were carried out. • Thermal–hydraulic analysis assured that the design matches imposed safety constraints. - Abstract: Parametric studies have been performed on a seed-blanket Th–{sup 233}U fuel configuration in a pressurized water reactor (PWR) with boiling channels to achieve high conversion ratio. Previous studies on seed-blanket concepts suggested substantial reduction in the core power density is needed in order to operate under nominal PWR system conditions. Boiling flow regime in the seed region allows more heat to be removed for a given coolant mass flow rate, which in turn, may potentially allow increasing the power density of the core. In addition, reduced moderation improves the breeding performance. A two-dimensional design optimization study was carried out with BOXER and SERPENT codes in order to determine the most attractive fuel assembly configuration that would ensure breeding. Effects of various parameters, such as void fraction, blanket fuel form, number of seed pins and their dimensions, on the conversion ratio were examined. The obtained results, for which the power density was set to be 104 W/cm{sup 3}, created a map of potentially feasible designs. It was found that several options have the potential to achieve end of life fissile inventory ratio above unity, which implies potential feasibility of a self-sustainable Thorium fuel cycle in PWRs without significant reduction in the core power density. Finally, a preliminary three-dimensional coupled neutronic and thermal–hydraulic analysis for a single seed-blanket fuel assembly was performed. The results indicate that axial void distribution changes drastically with burnup. Therefore

  18. Intensification of highly exothermic fast reaction by multi-injection microstructured reactor

    OpenAIRE

    Haber, Julien; Jiang, Bo; Maeder, Thomas; Borhani, Navid; Thome, John; Renken, Albert; Kiwi-Minsker, Lioubov

    2014-01-01

    Microstructured reactors (MSR) with characteristic dimensions below 100 μm are warranted to maintain close to isothermal conditions when carrying out quasi-instantaneous highly exothermic reactions. Unfortunately, such small dimensions increase the risk of clogging, create high pressure drop and are costly to number-up. The multi-injection (MI) MSR, where one of the reactants is added stepwise along the reactor length, allows working with larger dimensions (diameter >500 μm) while maintaining...

  19. Time Exceedances for High Intensity Solar Proton Fluxes

    Science.gov (United States)

    Xapsos, Michael A.; Stauffer, Craig A.; Jordan, Thomas M.; Adam, James H., Jr.; Dietrich, William F.

    2011-01-01

    A model is presented for times during a space mission that specified solar proton flux levels are exceeded. This includes both total time and continuous time periods during missions. Results for the solar maximum and solar minimum phases of the solar cycle are presented and compared for a broad range of proton energies and shielding levels. This type of approach is more amenable to reliability analysis for spacecraft systems and instrumentation than standard statistical models.

  20. High Flux, Fouling Resistant Membranes for RO Pretreatment

    Science.gov (United States)

    2012-05-31

    improved flux and fouling resistance . This in turn will lead to more compact and efficient filtration systems that reduce maintenance and downtime...size, and resistance to protein fouling by BSA filtration . Hollow Fiber Development and Testing Hollow fibers were produced by a generally...layer during dead- end filtration . The shear forces in cross-flow operation can impact the effective particle size in the feed and compact the cake

  1. Instrument for thermal radiation flux measurement in high temperature gas flow (Cuernavaca instrument)

    Energy Technology Data Exchange (ETDEWEB)

    Afgan, N.H. [Universidade Tecnica, Lisbon (Portugal); Leontiev, A.I. [Moscow State Technical University (Russian Federation)

    1995-05-01

    A new instrument for hemispherical radiation heat flux measurement is proposed. It is based on the theory of blow of the boundary layer, taking into account that at the critical mass flow rate through the porous surface the thermal boundary layer is blown off and only radiation flux from high temperature gases reaches the porous surface. With the measurement of blow of gas flow and the temperature of the porous material, the respective heat flux is obtained. (author)

  2. Effect of high-flux astronomical sources on the constellation-X microcalorimeter spectrometer

    Energy Technology Data Exchange (ETDEWEB)

    Figueroa-Feliciano, E. E-mail: enectali.figueroa@gsfc.nasa.gov; Bandler, S.; Boyce, K.; Chervenak, J.; Finkbeiner, F.; Kelley, R.; Lindeman, M.A.; Porter, S.F.; Saab, T.; Stahle, C.K

    2004-03-11

    We study the effects of high flux on transition-edge sensors by using a non-linear microcalorimeter model with parameters suitable for the Constellation-X mission. A photon event list was created that simulated the flux and spectrum of the Crab nebula incident on the detector when folded through the Constellation-X optics and filters. Simulations show that even at Crab levels of flux, the Con-X microcalorimeter does not 'latch' into the normal state.

  3. Modelling deuterium release from tungsten after high flux high temperature deuterium plasma exposure

    Science.gov (United States)

    Grigorev, Petr; Matveev, Dmitry; Bakaeva, Anastasiia; Terentyev, Dmitry; Zhurkin, Evgeny E.; Van Oost, Guido; Noterdaeme, Jean-Marie

    2016-12-01

    Tungsten is a primary candidate for plasma facing materials for future fusion devices. An important safety concern in the design of plasma facing components is the retention of hydrogen isotopes. Available experimental data is vast and scattered, and a consistent physical model of retention of hydrogen isotopes in tungsten is still missing. In this work we propose a model of non-equilibrium hydrogen isotopes trapping under fusion relevant plasma exposure conditions. The model is coupled to a diffusion-trapping simulation tool and is used to interpret recent experiments involving high plasma flux exposures. From the computational analysis performed, it is concluded that high flux high temperature exposures (T = 1000 K, flux = 1024 D/m2/s and fluence of 1026 D/m2) result in generation of sub-surface damage and bulk diffusion, so that the retention is driven by both sub-surface plasma-induced defects (bubbles) and trapping at natural defects. On the basis of the non-equilibrium trapping model we have estimated the amount of H stored in the sub-surface region to be ∼10-5 at-1, while the bulk retention is about 4 × 10-7 at-1, calculated by assuming the sub-surface layer thickness of about 10 μm and adjusting the trap concentration to comply with the experimental results for the integral retention.

  4. Modular High Temperature Gas-Cooled Reactor Safety Basis and Approach

    Energy Technology Data Exchange (ETDEWEB)

    David Petti; Jim Kinsey; Dave Alberstein

    2014-01-01

    Various international efforts are underway to assess the safety of advanced nuclear reactor designs. For example, the International Atomic Energy Agency has recently held its first Consultancy Meeting on a new cooperative research program on high temperature gas-cooled reactor (HTGR) safety. Furthermore, the Generation IV International Forum Reactor Safety Working Group has recently developed a methodology, called the Integrated Safety Assessment Methodology, for use in Generation IV advanced reactor technology development, design, and design review. A risk and safety assessment white paper is under development with respect to the Very High Temperature Reactor to pilot the Integrated Safety Assessment Methodology and to demonstrate its validity and feasibility. To support such efforts, this information paper on the modular HTGR safety basis and approach has been prepared. The paper provides a summary level introduction to HTGR history, public safety objectives, inherent and passive safety features, radionuclide release barriers, functional safety approach, and risk-informed safety approach. The information in this paper is intended to further the understanding of the modular HTGR safety approach. The paper gives those involved in the assessment of advanced reactor designs an opportunity to assess an advanced design that has already received extensive review by regulatory authorities and to judge the utility of recently proposed new methods for advanced reactor safety assessment such as the Integrated Safety Assessment Methodology.

  5. Magnetic fusion energy plasma interactive and high heat flux components. Volume II. Technical assessment of the critical issues and problem areas in high heat flux materials and component development

    Energy Technology Data Exchange (ETDEWEB)

    Abdou, M.A.; Boyd, R.D.; Easor, J.R.; Gauster, W.B.; Gordon, J.D.; Mattas, R.F.; Morgan, G.D.; Ulrickson, M.A,; Watson, R.D.; Wolfer, W.G,

    1984-06-01

    A technical assessment of the critical issues and problem areas for high heat flux materials and components (HHFMC) in magnetic fusion devices shows these problems to be of critical importance for the successful operation of near-term fusion experiments and for the feasibility and attractiveness of long-term fusion reactors. A number of subgroups were formed to assess the critical HHFMC issues along the following major lines: (1) source conditions, (2) systems integration, (3) materials and processes, (4) thermal hydraulics, (5) thermomechanical response, (6) electromagnetic response, (7) instrumentation and control, and (8) test facilities. The details of the technical assessment are presented in eight chapters. The primary technical issues and needs for each area are highlighted.

  6. Joining and fabrication techniques for high temperature structures including the first wall in fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ho Jin; Lee, B. S.; Kim, K. B

    2003-09-01

    The materials for PFC's (Plasma Facing Components) in a fusion reactor are severely irradiated with fusion products in facing the high temperature plasma during the operation. The refractory materials can be maintained their excellent properties in severe operating condition by lowering surface temperature by bonding them to the high thermal conducting materials of heat sink. Hence, the joining and bonding techniques between dissimilar materials is considered to be important in case of the fusion reactor or nuclear reactor which is operated at high temperature. The first wall in the fusion reactor is heated to approximately 1000 .deg. C and irradiated severely by the plasma. In ITER, beryllium is expected as the primary armour candidate for the PFC's; other candidates including W, Mo, SiC, B4C, C/C and Si{sub 3}N{sub 4}. Since the heat affected zones in the PFC's processed by conventional welding are reported to have embrittlement and degradation in the sever operation condition, both brazing and diffusion bonding are being considered as prime candidates for the joining technique. In this report, both the materials including ceramics and the fabrication techniques including joining technique between dissimilar materials for PFC's are described. The described joining technique between the refractory materials and the dissimilar materials may be applicable for the fusion reactor and Generation-4 future nuclear reactor which are operated at high temperature and high irradiation.

  7. Application of nondestructive methods for qualification of high density fuels in the IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Jose E.R.; Silva, Antonio T.; Domingos, Douglas B.; Terremoto, Luis A.A., E-mail: jersilva@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    IPEN/CNEN-SP manufactures fuels to be used in its research reactor - the IEA-R1. To qualify those fuels, it is necessary to check if they have a good performance under irradiation. As Brazil still does not have nuclear research reactors with high neutron fluxes, or suitable hot cells for carrying out post-irradiation examination of nuclear fuels, IPEN/CNEN-SP has conducted a fuel qualification program based on the use of uranium compounds (U{sub 3}O{sub 8} and U{sub 3}Si{sub 2} dispersed in Al matrix) internationally tested and qualified to be used in research reactors, and has attained experience in the technological development stages for the manufacturing of fuel plates, irradiation and non-destructive post-irradiation testing. Fuel elements containing low volume fractions of fuel in the dispersion were manufactured and irradiated successfully directly in the core of the IEA-R1. However, there are plans at IPEN/CNEN-SP to increase the uranium density of the fuels. Ten fuel miniplates (five containing U{sub 3}O{sub 8}-Al and five containing U{sub 3}Si{sub 2}-Al), with densities of 3.2 gU/cm{sup 3} and 4.8 gU/cm{sup 3} respectively, are being irradiated inside an irradiation device placed in a peripheral position of the IEA-R1 core. Non-destructive methods will be used to evaluate irradiation performance of the fuel miniplates after successive cycles of irradiation, by means: monitoring the reactor parameters during operation; periodic underwater visual inspection of fuel miniplates, eventual sipping test for fuel miniplates suspected of leakage and underwater measuring of the miniplate thickness for assessment of the fuel miniplate swelling. (author)

  8. Argonne National Laboratory research offers clues to Alzheimer's plaques

    CERN Multimedia

    2003-01-01

    Researchers from Argonne National Laboratory and the University of Chicago have developed methods to directly observe the structure and growth of microscopic filaments that form the characteristic plaques found in the brains of those with Alzheimer's Disease (1 page).

  9. Thermal barrier coatings (TBC's) for high heat flux thrust chambers

    Science.gov (United States)

    Bradley, Christopher M.

    -section components has become critical, but at the same time the service conditions have put our best alloy systems to their limits. As a result, implementation of cooling holes and thermal barrier coatings are new advances in hot-section technologies now looked at for modifications to reach higher temperature applications. Current thermal barrier coatings used in today's turbine applications is known as 8%yttria-stabilized zirconia (YSZ) and there are no coatings for current thrust chambers. Current research is looking at the applicability of 8%yttria-stabilized hafnia (YSH) for turbine applications and the implementation of 8%YSZ onto thrust chambers. This study intends to determine if the use of thermal barrier coatings are applicable for high heat flux thrust chambers using industrial YSZ will be advantageous for improvements in efficiency, thrust and longer service life by allowing the thrust chambers to be used more than once.

  10. Single-pass high harmonic generation at high repetition rate and photon flux

    Science.gov (United States)

    Hädrich, Steffen; Rothhardt, Jan; Krebs, Manuel; Demmler, Stefan; Klenke, Arno; Tünnermann, Andreas; Limpert, Jens

    2016-09-01

    Sources of short wavelength radiation with femtosecond to attosecond pulse durations, such as synchrotrons or free electron lasers, have already made possible numerous, and will facilitate more, seminal studies aimed at understanding atomic and molecular processes on fundamental length and time scales. Table-top sources of coherent extreme ultraviolet to soft x-ray radiation enabled by high harmonic generation (HHG) of ultrashort pulse lasers have also gained significant attention in the last few years due to their enormous potential for addressing a plethora of applications, therefore constituting a complementary source to large-scale facilities (synchrotrons and free electron lasers). Ti:sapphire based laser systems have been the workhorses for HHG for decades, but are limited in repetition rate and average power. On the other hand, it has been widely recognized that fostering applications in fields such as photoelectron spectroscopy and microscopy, coincidence detection, coherent diffractive imaging and frequency metrology requires a high repetition rate and high photon flux HHG sources. In this article we will review recent developments in realizing the demanding requirement of producing a high photon flux and repetition rate at the same time. Particular emphasis will be put on suitable ultrashort pulse and high average power lasers, which directly drive harmonic generation without the need for external enhancement cavities. To this end we describe two complementary schemes that have been successfully employed for high power fiber lasers, i.e. optical parametric chirped pulse amplifiers and nonlinear pulse compression. Moreover, the issue of phase-matching in tight focusing geometries will be discussed and connected to recent experiments. We will highlight the latest results in fiber laser driven high harmonic generation that currently produce the highest photon flux of all existing sources. In addition, we demonstrate the first promising applications and

  11. Extension of the reactor dynamics code MGT-3D for pebblebed and blocktype high-temperature-reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shi, Dunfu

    2015-07-01

    The High Temperature Gas cooled Reactor (HTGR) is an improved, gas cooled nuclear reactor. It was chosen as one of the candidates of generation IV nuclear plants [1]. The reactor can be shut down automatically because of the negative reactivity feedback due to the temperature's increasing in designed accidents. It is graphite moderated and Helium cooled. The residual heat can be transferred out of the reactor core by inactive ways as conduction, convection, and thermal radiation during the accident. In such a way, a fuel temperature does not go beyond a limit at which major fission product release begins. In this thesis, the coupled neutronics and fluid mechanics code MGT-3D used for the steady state and time-dependent simulation of HTGRs, is enhanced and validated [2]. The fluid mechanics part is validated by SANA experiments in steady state cases as well as transient cases. The fuel temperature calculation is optimized by solving the heat conduction equation of the coated particles. It is applied in the steady state and transient simulation of PBMR, and the results are compared to the simulation with the old overheating model. New approaches to calculate the temperature profile of the fuel element of block-type HTGRs, and the calculation of the homogeneous conductivity of composite materials are introduced. With these new developments, MGT-3D is able to simulate block-type HTGRs as well. This extended MGT-3D is used to simulate a cuboid ceramic block heating experiment in the NACOK-II facility. The extended MGT-3D is also applied to LOFC and DLOFC simulation of GT-MHR. It is a fluid mechanics calculation with a given heat source. This calculation result of MGT-3D is verified with the calculation results of other codes. The design of the Japanese HTTR is introduced. The deterministic simulation of the LOFC experiment of HTTR is conducted with the Monte-Carlo code Serpent and MGT-3D, which is the LOFC Project organized by OECD/NEA [3]. With Serpent the burnup

  12. A Review of Mold Flux Development for the Casting of High-Al Steels

    Science.gov (United States)

    Wang, Wanlin; Lu, Boxun; Xiao, Dan

    2016-02-01

    Mold flux plays key roles during the continuous casting process of molten steel, which accounts for the quality of final slabs. With the development of advanced high strength steels (AHSS), certain amounts of Al have been added into steels that would introduce severe slag/metal interaction problems during process of continuous casting. The reaction is between Al and SiO2 that is the major component in the mold flux system. Intensive efforts have been conducted to optimize the mold flux and a CaO-Al2O3-based mold flux system has been proposed, which shows the potential to be applied for the casting process of AHSS. The latest developments for this new mold flux system were summarized with the aim to offer technical guidance for the design of new generation mold flux system for the casting of AHSS.

  13. Translational symmetry of high order tokamak flux surface shaping in gyrokinetics

    CERN Document Server

    Ball, Justin; Barnes, Michael

    2015-01-01

    A particular translational symmetry of the local nonlinear $\\delta f$ gyrokinetic model is demonstrated analytically and verified numerically. This symmetry shows that poloidally translating all the flux surface shaping effects with large poloidal mode number by a single tilt angle has an exponentially small effect on the transport properties of a tokamak. This is shown using a generalization of the Miller local equilibrium model to specify an arbitrary flux surface geometry. With this geometry specification we find that, when performing an expansion in large flux surface shaping mode number, the governing equations of gyrokinetics are symmetric in the poloidal translation of the high order shaping effects. This allows us to take the fluxes from a single configuration and calculate the fluxes in any configuration that can be produced by translating the large mode number shaping effects. This creates a distinction between tokamaks with mirror symmetric flux surfaces and tokamaks without mirror symmetry, which ...

  14. An easily regenerable enzyme reactor prepared from polymerized high internal phase emulsions.

    Science.gov (United States)

    Ruan, Guihua; Wu, Zhenwei; Huang, Yipeng; Wei, Meiping; Su, Rihui; Du, Fuyou

    2016-04-22

    A large-scale high-efficient enzyme reactor based on polymerized high internal phase emulsion monolith (polyHIPE) was prepared. First, a porous cross-linked polyHIPE monolith was prepared by in-situ thermal polymerization of a high internal phase emulsion containing styrene, divinylbenzene and polyglutaraldehyde. The enzyme of TPCK-Trypsin was then immobilized on the monolithic polyHIPE. The performance of the resultant enzyme reactor was assessed according to the conversion ability of Nα-benzoyl-l-arginine ethyl ester to Nα-benzoyl-l-arginine, and the protein digestibility of bovine serum albumin (BSA) and cytochrome (Cyt-C). The results showed that the prepared enzyme reactor exhibited high enzyme immobilization efficiency and fast and easy-control protein digestibility. BSA and Cyt-C could be digested in 10 min with sequence coverage of 59% and 78%, respectively. The peptides and residual protein could be easily rinsed out from reactor and the reactor could be regenerated easily with 4 M HCl without any structure destruction. Properties of multiple interconnected chambers with good permeability, fast digestion facility and easily reproducibility indicated that the polyHIPE enzyme reactor was a good selector potentially applied in proteomics and catalysis areas.

  15. High Throughput Atomic Layer Deposition Processes: High Pressure Operations, New Reactor Designs, and Novel Metal Processing

    Science.gov (United States)

    Mousa, MoatazBellah Mahmoud

    Atomic Layer Deposition (ALD) is a vapor phase nano-coating process that deposits very uniform and conformal thin film materials with sub-angstrom level thickness control on various substrates. These unique properties made ALD a platform technology for numerous products and applications. However, most of these applications are limited to the lab scale due to the low process throughput relative to the other deposition techniques, which hinders its industrial adoption. In addition to the low throughput, the process development for certain applications usually faces other obstacles, such as: a required new processing mode (e.g., batch vs continuous) or process conditions (e.g., low temperature), absence of an appropriate reactor design for a specific substrate and sometimes the lack of a suitable chemistry. This dissertation studies different aspects of ALD process development for prospect applications in the semiconductor, textiles, and battery industries, as well as novel organic-inorganic hybrid materials. The investigation of a high pressure, low temperature ALD process for metal oxides deposition using multiple process chemistry revealed the vital importance of the gas velocity over the substrate to achieve fast depositions at these challenging processing conditions. Also in this work, two unique high throughput ALD reactor designs are reported. The first is a continuous roll-to-roll ALD reactor for ultra-fast coatings on porous, flexible substrates with very high surface area. While the second reactor is an ALD delivery head that allows for in loco ALD coatings that can be executed under ambient conditions (even outdoors) on large surfaces while still maintaining very high deposition rates. As a proof of concept, part of a parked automobile window was coated using the ALD delivery head. Another process development shown herein is the improvement achieved in the selective synthesis of organic-inorganic materials using an ALD based process called sequential vapor

  16. Progress on the realization of a new GEM based neutron diagnostic concept for high flux neutron beams

    Energy Technology Data Exchange (ETDEWEB)

    Croci, G.; Tardocchi, M. [Istituto di Fisica del Plasma, Associazione EURATOM-ENEA-CNR, Milano, Italy and INFN, Sez. di Milano-Bicocca, Milano (Italy); Rebai, M.; Cippo, E. Perelli; Gorini, G. [Dipartimento di Fisica, Università degli Studi di Milano-Bicocca, Milano, Italy and INFN, Sez. di Milano-Bicocca, Milano (Italy); Cazzaniga, C. [Dipartimento di Fisica, Università degli Studi di Milano-Bicocca, Milano (Italy); Palma, M. Dalla; Pasqualotto, R.; Tollin, M. [Consorzio RFX - Associazione Euratom-Enea sulla Fusione, Padova (Italy); Grosso, G.; Muraro, A. [Istituto di Fisica del Plasma, Associazione EURATOM-ENEA-CNR, Milano (Italy); Murtas, F.; Claps, G. [INFN, Laboratori Nazionali di Frascati, Frascati (Roma) (Italy); Cavenago, M. [INFN, Laboratori Nazionali di Legnaro, Legnaro (Padova) (Italy)

    2014-08-21

    Fusion reactors will need high flux neutron detectors to diagnose the deuterium-deuterium and deuterium-tritium. A candidate detection technique is the Gas Electron Multiplier (GEM). New GEM based detectors are being developed for application to a neutral deuterium beam test facility. The proposed detection system is called Close-contact Neutron Emission Surface Mapping (CNESM). The diagnostic aims at providing the map of the neutron emission due to interaction of the deuterium beam with the deuterons implanted in the beam dump surface. This is done by placing a detector in close contact, right behind the dump. CNESM uses nGEM detectors, i.e. GEM detectors equipped with a cathode that also serves as neutron-proton converter foil. After the realization and test of several small area prototypes, a full size prototype has been realized and tested with laboratory sources. Test on neutron beams are foreseen for the next months.

  17. Progress on the realization of a new GEM based neutron diagnostic concept for high flux neutron beams

    Science.gov (United States)

    Croci, G.; Rebai, M.; Cazzaniga, C.; Palma, M. Dalla; Grosso, G.; Muraro, A.; Murtas, F.; Claps, G.; Pasqualotto, R.; Cippo, E. Perelli; Tardocchi, M.; Tollin, M.; Cavenago, M.; Gorini, G.

    2014-08-01

    Fusion reactors will need high flux neutron detectors to diagnose the deuterium-deuterium and deuterium-tritium. A candidate detection technique is the Gas Electron Multiplier (GEM). New GEM based detectors are being developed for application to a neutral deuterium beam test facility. The proposed detection system is called Close-contact Neutron Emission Surface Mapping (CNESM). The diagnostic aims at providing the map of the neutron emission due to interaction of the deuterium beam with the deuterons implanted in the beam dump surface. This is done by placing a detector in close contact, right behind the dump. CNESM uses nGEM detectors, i.e. GEM detectors equipped with a cathode that also serves as neutron-proton converter foil. After the realization and test of several small area prototypes, a full size prototype has been realized and tested with laboratory sources. Test on neutron beams are foreseen for the next months.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  19. Development of a version of the reactor dynamics code DYN3D applicable for High Temperature Reactors; Entwicklung einer Version des Reaktordynamikcodes DYN3D fuer Hochtemperaturreaktoren. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Rohde, Ulrich; Apanasevich, Pavel; Baier, Silvio; Duerigen, Susan; Fridman, Emil; Grahn, Alexander; Kliem, Soeren; Merk, Bruno

    2012-07-15

    Based on the reactor dynamics code DYN3D for the simulation of transient processes in Light Water Reactors, a code version DYN3D-HTR for application to graphitemoderated, gas-cooled block-type high temperature reactors has been developed. This development comprises: - the methodical improvement of the 3D steady-state neutron flux calculation for the hexagonal geometry of the HTR fuel element blocks - the development of methods for the generation of homogenised cross section data taking into account the double heterogeneity of the fuel element block structure - the implementation of a 3D model for heat conduction and heat transport in the graphite matrix. The nodal method for neutron flux calculation based on SP3 transport approximation was extended to hexagonal fuel element geometry, where the hexagons are subdivided into triangles, thus the method had finally to be derived for triangular geometry. In triangular geometry, a subsequent subdivision of the hexagonal elements can be considered, and therefore, the effect of systematic mesh refinement can be studied. The algorithm was verified by comparison with Monte Carlo reference solutions, on the node-wise level, as well as also on the pin-wise level. New procedures were developed for the homogenization of the double-heterogeneous fuel element structures. One the one hand, the so-called Reactivity equivalent Physical Transformation (RPT), the two-step homogenization method based on 2D deterministic lattice calculations, was extended to cells with different temperatures of the materials. On the other hand, the progress in development of Monte Carlo methods for spectral calculations, in particular the development of the code SERPENT, opened a new, fully consistent 3D approach, where all details of the structures on fuel particle, fuel compact and fuel block level can be taken into account within one step. Moreover, a 3D heat conduction and heat transport model was integrated into DYN3D to be able to simulate radial

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

    Energy Technology Data Exchange (ETDEWEB)

    Davis, T. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Gomez, J. L. [Argonne National Lab. (ANL), Argonne, IL (United States); Moos, L. P. [Argonne National Lab. (ANL), Argonne, IL (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  5. Transmutation of minor actinides discharged from LMFBR spent fuel in a high power density fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Uebeyli, Mustafa E-mail: mubeyli@gazi.edu.tr

    2004-12-01

    Significant amounts of nuclear wastes consisting of plutonium, minor actinides and long lived fission products are produced during the operation of commercial nuclear power plants. Therefore, the destruction of these wastes is very important with respect to public health, environment and also the future of nuclear energy. In this study, transmutation of minor actinides (MAs) discharged from LMFBR spent fuel in a high power density fusion reactor has been investigated under a neutron wall load of 10 MW/m{sup 2} for an operation period of 10 years. Also, the effect of MA percentage on the transmutation has been examined. The fuel zone, containing MAs as spheres cladded with W-5Re, has been located behind the first wall to utilize the high neutron flux for transmutation effectively. Helium at 40 atm has been used as an energy carrier. At the end of the operation period, the total burning and transmutation are greater than the total buildups in all investigated cases, and very high burnups (420-470 GWd/tHM) are reached, depending on the MA content. The total transmutation rate values are 906 and 979 kg/GW{sub th} year at startup and decrease to 140 and 178 kg/GW{sub th} year at the end of the operation for fuel with 10% and 20% MA, respectively. Over an operation period of 10 years, the effective half lives decrease from 2.38, 2.21 and 3.08 years to 1.95, 1.80 and 2.59 years for {sup 237}Np, {sup 241}Am and {sup 243}Am, respectively. Total atomic densities decrease exponentially during the operation period. The reductions in the total atomic densities with respect to the initial ones are 79%, 81%, 82%, 83%, 85% and 86% for 10%, 12%, 14%, 16%, 18% and 20% MAs, respectively.

  6. Design of and initial results from a highly instrumented reactor for atmospheric chemistry (HIRAC

    Directory of Open Access Journals (Sweden)

    D. R. Glowacki

    2007-07-01

    Full Text Available The design of a Highly Instrumented Reactor for Atmospheric Chemistry (HIRAC is described and initial results obtained from HIRAC are presented. The ability of HIRAC to perform in-situ laser-induced fluorescence detection of OH and HO2 radicals with the Fluorescence Assay by Gas Expansion (FAGE technique establishes it as internationally unique for a chamber of its size and pressure/temperature variable capabilities. In addition to the FAGE technique, HIRAC features a suite of analytical instrumentation, including: a multipass FTIR system; a conventional gas chromatography (GC instrument and a GC instrument for formaldehyde detection; and NO/NO2, CO, O3, and H2O vapour analysers. Ray tracing simulations and measurements of the blacklamp flux have been utilized to develop a detailed model of the radiation field within HIRAC. Comparisons between the analysers and the FTIR coupled to HIRAC have been performed, and HIRAC has also been used to investigate pressure dependent kinetics of the chlorine atom reaction with ethene and the reaction of O3 and t-2-butene. The results obtained are in good agreement with literature recommendations and Master Chemical Mechanism predictions. HIRAC thereby offers a highly instrumented platform with the potential for: (1 high precision kinetics investigations over a range of atmospheric conditions; (2 detailed mechanism development, significantly enhanced according to its capability for measuring radicals; and (3 field instrument intercomparison, calibration, development, and investigations of instrument response under a range of atmospheric conditions.

  7. Transient modeling of the thermohydraulic behavior of high temperature heat pipes for space reactor applications

    Science.gov (United States)

    Hall, Michael L.; Doster, Joseph M.

    1986-01-01

    Many proposed space reactor designs employ heat pipes as a means of conveying heat. Previous researchers have been concerned with steady state operation, but the transient operation is of interest in space reactor applications due to the necessity of remote startup and shutdown. A model is being developed to study the dynamic behavior of high temperature heat pipes during startup, shutdown and normal operation under space environments. Model development and preliminary results for a hypothetical design of the system are presented.

  8. A high-flux BEC source for mobile atom interferometers

    CERN Document Server

    Rudolph, Jan; Grzeschik, Christoph; Sternke, Tammo; Grote, Alexander; Popp, Manuel; Becker, Dennis; Müntinga, Hauke; Ahlers, Holger; Peters, Achim; Lämmerzahl, Claus; Sengstock, Klaus; Gaaloul, Naceur; Ertmer, Wolfgang; Rasel, Ernst M

    2015-01-01

    Quantum sensors based on coherent matter-waves are precise measurement devices whose ultimate accuracy is achieved with Bose-Einstein condensates (BEC) in extended free fall. This is ideally realized in microgravity environments such as drop towers, ballistic rockets and space platforms. However, the transition from lab-based BEC machines to robust and mobile sources with comparable performance is a technological challenge. Here we report on the realization of a miniaturized setup, generating a flux of $4 \\times 10^5$ quantum degenerate $^{87}$Rb atoms every 1.6 s. Ensembles of $1 \\times 10^5$ atoms can be produced at a 1 Hz rate. This is achieved by loading a cold atomic beam directly into a multi-layer atom chip that is designed for efficient transfer from laser-cooled to magnetically trapped clouds. The attained flux of degenerate atoms is on par with current lab-based experiments while offering significantly higher repetition rates. The compact and robust design allows for mobile operation in a variety of...

  9. Local Neutron Flux Distribution Measurements by Wire-Dosimetry in the AMMON Experimental Program in the EOLE Reactor

    Directory of Open Access Journals (Sweden)

    Gruel A.

    2016-01-01

    Full Text Available Dosimetry measurements were carried out during the AMMON experimental program, in the EOLE facility. Al-0.1 wt% Au wires were positioned along curved fuel plates of JHR-type assemblies to investigate the azimuthal and axial gold capture rate profiles, directly linked to the thermal and epithermal flux. After irradiation, wires were cut into small segments (a few mm, and the gold capture rate of each part was measured by gamma spectrometry on the MADERE platform. This paper presents results in the “hafnium” configuration, and more specifically the azimuthal flux profile characterization. The final uncertainty on each measured wire lies below 1% (at 2 standard deviations. Experimental profiles are in a good agreement against Monte Carlo calculations, and the 4% capture rate increase at the plate edge is well observed. The flux dissymmetry due to assembly position in the core is also measured, and shows a 10% discrepancy between the two edges of the plate.

  10. Fuel-Cycle and Nuclear Material Disposition Issues Associated with High-Temperature Gas Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shropshire, D.E.; Herring, J.S.

    2004-10-03

    The objective of this paper is to facilitate a better understanding of the fuel-cycle and nuclear material disposition issues associated with high-temperature gas reactors (HTGRs). This paper reviews the nuclear fuel cycles supporting early and present day gas reactors, and identifies challenges for the advanced fuel cycles and waste management systems supporting the next generation of HTGRs, including the Very High Temperature Reactor, which is under development in the Generation IV Program. The earliest gas-cooled reactors were the carbon dioxide (CO2)-cooled reactors. Historical experience is available from over 1,000 reactor-years of operation from 52 electricity-generating, CO2-cooled reactor plants that were placed in operation worldwide. Following the CO2 reactor development, seven HTGR plants were built and operated. The HTGR came about from the combination of helium coolant and graphite moderator. Helium was used instead of air or CO2 as the coolant. The helium gas has a significant technical base due to the experience gained in the United States from the 40-MWe Peach Bottom and 330-MWe Fort St. Vrain reactors designed by General Atomics. Germany also built and operated the 15-MWe Arbeitsgemeinschaft Versuchsreaktor (AVR) and the 300-MWe Thorium High-Temperature Reactor (THTR) power plants. The AVR, THTR, Peach Bottom and Fort St. Vrain all used fuel containing thorium in various forms (i.e., carbides, oxides, thorium particles) and mixtures with highly enriched uranium. The operational experience gained from these early gas reactors can be applied to the next generation of nuclear power systems. HTGR systems are being developed in South Africa, China, Japan, the United States, and Russia. Elements of the HTGR system evaluated included fuel demands on uranium ore mining and milling, conversion, enrichment services, and fuel fabrication; fuel management in-core; spent fuel characteristics affecting fuel recycling and refabrication, fuel handling, interim

  11. The Advanced High-Temperature Reactor (AHTR) for Producing Hydrogen to Manufacture Liquid Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Peterson, P.F.; Ott, L.

    2004-10-06

    Conventional world oil production is expected to peak within a decade. Shortfalls in production of liquid fuels (gasoline, diesel, and jet fuel) from conventional oil sources are expected to be offset by increased production of fuels from heavy oils and tar sands that are primarily located in the Western Hemisphere (Canada, Venezuela, the United States, and Mexico). Simultaneously, there is a renewed interest in liquid fuels from biomass, such as alcohol; but, biomass production requires fertilizer. Massive quantities of hydrogen (H2) are required (1) to convert heavy oils and tar sands to liquid fuels and (2) to produce fertilizer for production of biomass that can be converted to liquid fuels. If these liquid fuels are to be used while simultaneously minimizing greenhouse emissions, nonfossil methods for the production of H2 are required. Nuclear energy can be used to produce H2. The most efficient methods to produce H2 from nuclear energy involve thermochemical cycles in which high-temperature heat (700 to 850 C) and water are converted to H2 and oxygen. The peak nuclear reactor fuel and coolant temperatures must be significantly higher than the chemical process temperatures to transport heat from the reactor core to an intermediate heat transfer loop and from the intermediate heat transfer loop to the chemical plant. The reactor temperatures required for H2 production are at the limits of practical engineering materials. A new high-temperature reactor concept is being developed for H2 and electricity production: the Advanced High-Temperature Reactor (AHTR). The fuel is a graphite-matrix, coated-particle fuel, the same type that is used in modular high-temperature gas-cooled reactors (MHTGRs). The coolant is a clean molten fluoride salt with a boiling point near 1400 C. The use of a liquid coolant, rather than helium, reduces peak reactor fuel and coolant temperatures 100 to 200 C relative to those of a MHTGR. Liquids are better heat transfer fluids than gases

  12. Investigating the Spectral Anomaly with Different Reactor Antineutrino Experiments

    CERN Document Server

    Buck, Christian; Haser, Julia; Lindner, Manfred

    2015-01-01

    The spectral shape of reactor antineutrinos measured in recent experiments shows anomalies in the neutrino flux predictions. New precision measurements of the reactor neutrino spectra as well as more complete input in nuclear data bases are needed to resolve the observed discrepancies between flux models and experimental results. We combine experiments at reactors which are highly enriched in ${}^{235}$U with commercial reactors with typically lower enrichment to gain new insights into the origin of the anomalous neutrino spectrum. As an example, we discuss the option of a direct comparison of the measured shape in the currently running Double Chooz near detector and the upcoming Stereo experiment.

  13. Measurement of the Ratio of High Energy Neutron in the Pulse Nuclear Reactor

    Institute of Scientific and Technical Information of China (English)

    MAO; Guo-shu; DING; You-qian; YANG; Lei; MA; Peng; YU; Zhen-hua

    2012-01-01

    <正>In the production of radioisotopes and neutron activation analysis, the fast neutron densities are very important to estimate the yields of the radioisotopes. In order to determine the fast neutron flux ratio, different foils are used to measure the thermal neutron flux and the fast neutron flux. In this paper 238U was used as only a monitor to measure the ratio of high energy neutron (>6 MeV). By measuring the

  14. Thermal analysis of heat and power plant with high temperature reactor and intermediate steam cycle

    Directory of Open Access Journals (Sweden)

    Fic Adam

    2015-03-01

    Full Text Available Thermal analysis of a heat and power plant with a high temperature gas cooled nuclear reactor is presented. The main aim of the considered system is to supply a technological process with the heat at suitably high temperature level. The considered unit is also used to produce electricity. The high temperature helium cooled nuclear reactor is the primary heat source in the system, which consists of: the reactor cooling cycle, the steam cycle and the gas heat pump cycle. Helium used as a carrier in the first cycle (classic Brayton cycle, which includes the reactor, delivers heat in a steam generator to produce superheated steam with required parameters of the intermediate cycle. The intermediate cycle is provided to transport energy from the reactor installation to the process installation requiring a high temperature heat. The distance between reactor and the process installation is assumed short and negligable, or alternatively equal to 1 km in the analysis. The system is also equipped with a high temperature argon heat pump to obtain the temperature level of a heat carrier required by a high temperature process. Thus, the steam of the intermediate cycle supplies a lower heat exchanger of the heat pump, a process heat exchanger at the medium temperature level and a classical steam turbine system (Rankine cycle. The main purpose of the research was to evaluate the effectiveness of the system considered and to assess whether such a three cycle cogeneration system is reasonable. Multivariant calculations have been carried out employing the developed mathematical model. The results have been presented in a form of the energy efficiency and exergy efficiency of the system as a function of the temperature drop in the high temperature process heat exchanger and the reactor pressure.

  15. High precision, medium flux rate CZT spectroscopy for coherent scatter imaging

    Science.gov (United States)

    Greenberg, Joel A.; Hassan, Mehadi; Brady, David J.; Iniewski, Kris

    2016-05-01

    CZT detectors are primary candidates for many next-generation X-ray imaging systems. These detectors are typically operated in either a high precision, low flux spectroscopy mode or a low precision, high flux photon counting mode. We demonstrate a new detector configuration that enables operation in a high precision, medium flux spectroscopy mode, which opens the potential for a variety of new applications in medical imaging, non-destructive testing and baggage scanning. In particular, we describe the requirements of a coded aperture coherent scattering X-ray system that can perform fast imaging with accurate material discrimination.

  16. Compact high-flux source of cold sodium atoms

    Science.gov (United States)

    Lamporesi, G.; Donadello, S.; Serafini, S.; Ferrari, G.

    2013-06-01

    We present a compact source of cold sodium atoms suitable for the production of quantum degenerate gases and versatile for a multi-species experiment. The magnetic field produced by permanent magnets allows to simultaneously realize a Zeeman slower and a two-dimensional magneto-optical trap (MOT) within an order of magnitude smaller length than standard sodium sources. We achieve an atomic flux exceeding 4 × 109 atoms/s loaded in a MOT, with a most probable longitudinal velocity of 20 m/s, and a brightness larger than 2.5 × 1012 atoms/s/sr. This atomic source allows us to produce pure Bose-Einstein condensates with more than 107 atoms and a background pressure limited lifetime of 5 min.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  18. High-temperature nuclear reactor power plant cycle for hydrogen and electricity production – numerical analysis

    Directory of Open Access Journals (Sweden)

    Dudek Michał

    2016-01-01

    Full Text Available High temperature gas-cooled nuclear reactor (called HTR or HTGR for both electricity generation and hydrogen production is analysed. The HTR reactor because of the relatively high temperature of coolant could be combined with a steam or gas turbine, as well as with the system for heat delivery for high-temperature hydrogen production. However, the current development of HTR’s allows us to consider achievable working temperature up to 750°C. Due to this fact, industrial-scale hydrogen production using copper-chlorine (Cu-Cl thermochemical cycle is considered and compared with high-temperature electrolysis. Presented calculations show and confirm the potential of HTR’s as a future solution for hydrogen production without CO2 emission. Furthermore, integration of a hightemperature nuclear reactor with a combined cycle for electricity and hydrogen production may reach very high efficiency and could possibly lead to a significant decrease of hydrogen production costs.

  19. Design description and validation results for the IFMIF High Flux Test Module as outcome of the EVEDA phase

    Directory of Open Access Journals (Sweden)

    F. Arbeiter

    2016-12-01

    Full Text Available During the Engineering Validation and Engineering Design Activities (EVEDA phase (2007-2014 of the International Fusion Materials Irradiation Facility (IFMIF, an advanced engineering design of the High Flux Test Module (HFTM has been developed with the objective to facilitate the controlled irradiation of steel samples in the high flux area directly behind the IFMIF neutron source. The development process addressed included manufacturing techniques, CAD, neutronic, thermal-hydraulic and mechanical analyses complemented by a series of validation activities. Validation included manufacturing of 1:1 parts and mockups, test of prototypes in the FLEX and HELOKA-LP helium loops of KIT for verification of the thermal and mechanical properties, and irradiation of specimen filled capsule prototypes in the BR2 test reactor. The prototyping activities were backed by several R&D studies addressing focused issues like handling of liquid NaK (as filling medium and insertion of Small Specimen Test Technique (SSTT specimens into the irradiation capsules. This paper provides an up-todate design description of the HFTM irradiation device, and reports on the achieved performance criteria related to the requirements. Results of the validation activities are accounted for and the most important issues for further development are identified.

  20. Analysis of two-phase flow instability in helical tube steam generator in high temperature gas cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Yu; Lv, Xuefeng; Wang, Shengfei; Niu, Fenglei; Tian, Li [North China Electric Power Univ., Beijing (Switzerland)

    2012-03-15

    The steam generator composed of multi-helical tubes is used in high temperature gas cooled reactors and two-phase flow instability should be avoided in design. And density-wave oscillation which is mainly due to flow, density and the relationship between the pressure drop delays and feedback effects is one of the two-phase flow instability phenomena easily to occur. Here drift-flux model is used to simulate the performance of the fluid in the secondary side and frequency domain and time domain methods are used to evaluate whether the density-wave oscillation will happen or not. Several operating conditions with nominal power from 15% to 30% are calculated in this paper. The results of the two methods are in accordance, flow instability will occur when power is less than 20% nominal power, which is also according with the result of the experiments well.

  1. On0Line Fuel Failure Monitor for Fuel Testing and Monitoring of Gas Cooled Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ayman I. Hawari; Mohamed A. Bourham

    2010-04-22

    IVery High Temperature Reactors (VHTR) utilize the TRISO microsphere as the fundamental fuel unit in the core. The TRISO microsphere (~ 1- mm diameter) is composed of a UO2 kernel surrounded by a porous pyrolytic graphite buffer, an inner pyrolytic graphite layer, a silicon carbide (SiC) coating, and an outer pyrolytic graphite layer. The U-235 enrichment of the fuel is expected to range from 4% – 10% (higher enrichments are also being considered). The layer/coating system that surrounds the UO2 kernel acts as the containment and main barrier against the environmental release of radioactivity. To understand better the behavior of this fuel under in-core conditions (e.g., high temperature, intense fast neutron flux, etc.), the US Department of Energy (DOE) is launching a fuel testing program that will take place at the Advanced Test Reactor (ATR) located at Idaho National Laboratory (INL). During this project North Carolina State University (NCSU) researchers will collaborate with INL staff for establishing an optimized system for fuel monitoring for the ATR tests. In addition, it is expected that the developed system and methods will be of general use for fuel failure monitoring in gas cooled VHTRs.

  2. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

    2010-09-01

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450

  3. Nanostructures and pinholes on W surfaces exposed to high flux D plasma at high temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Jia, Y.Z., E-mail: jaja880816@aliyun.com [Laboratory of Advanced Materials, School of Materials Science and Engineering, Tsinghua University, Beijing 100084 (China); Liu, W., E-mail: liuw@mail.tsinghua.edu.cn [Laboratory of Advanced Materials, School of Materials Science and Engineering, Tsinghua University, Beijing 100084 (China); Xu, B. [Laboratory of Advanced Materials, School of Materials Science and Engineering, Tsinghua University, Beijing 100084 (China); Luo, G.-N. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Li, C.; Fu, B.Q. [Laboratory of Advanced Materials, School of Materials Science and Engineering, Tsinghua University, Beijing 100084 (China); De Temmerman, G. [FOM Institute DIFFER-Dutch Institute for Fundamental Energy Research, Edisonbaan 14, 3439 MN Nieuwegein (Netherlands); ITER Organization, Route de Vinon-Sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France)

    2015-08-15

    Nanostructures and pinholes formed on tungsten surface exposed to high fluxes (10{sup 24} m{sup −2} s{sup −1}) deuterium ions at 943 K and 1073 K were studied by scanning electron microscopy and electron backscatter diffraction. Nanostructure formation is observed at 943 K and 1073 K, and exhibits a strong dependence on the surface orientation. With increasing fluence, pinholes appear on the surface and are mainly observed on grains with surface normal near [1 1 1]. The pinholes are speculated to be caused by the rupture of bubbles formed near the surface. The formation of pinholes has no obvious relationship with the surface nanostructures.

  4. High temperature indentation tests on fusion reactor candidate materials

    Energy Technology Data Exchange (ETDEWEB)

    Montanari, R. [Dipartimento di Ingegneria Meccanica, Universita di Roma-Tor Vergata, Via del Politecnico 1, I-00133 Rome (Italy)]. E-mail: roberto.montanari@uniroma2.it; Filacchioni, G. [ENEA CR Casaccia, Via Anguillarese 301, I-00060 S.M. di Galeria, Rome (Italy); Iacovone, B. [Dipartimento di Ingegneria Meccanica, Universita di Roma-Tor Vergata, Via del Politecnico 1, I-00133 Rome (Italy); Plini, P. [Dipartimento di Ingegneria Meccanica, Universita di Roma-Tor Vergata, Via del Politecnico 1, I-00133 Rome (Italy); Riccardi, B. [Associazione EURATOM-ENEA sulla Fusione, P.O. Box 65, I-00044 Frascati, Rome (Italy)

    2007-08-01

    Flat-top cylinder indenter for mechanical characterization (FIMEC) is an indentation technique employing cylindrical punches with diameters ranging from 0.5 to 2 mm. The test gives pressure-penetration curves from which the yield stress can be determined. The FIMEC apparatus was developed to test materials in the temperature range from -180 to +200 {sup o}C. Recently, the heating system of FIMEC apparatus has been modified to operate up to 500 {sup o}C. So, in addition to providing yield stress over a more extended temperature range, it is possible to perform stress-relaxation tests at temperatures of great interest for several nuclear fusion reactor (NFR) alloys. Data on MANET-II, F82H mod., Eurofer-97, EM-10, AISI 316 L, Ti6Al4V and CuCrZr are presented and compared with those obtained by mechanical tests with standard methods.

  5. Diagnostic studies on lithium-ion cells at Argonne National Laboratory: an overview

    Science.gov (United States)

    Abraham, Daniel P.

    2010-04-01

    High-power and high-energy lithium-ion cells are being studied at Argonne National Laboratory (Argonne) as part of the U.S. Department of Energy's FreedomCar and Vehicle Technologies (FCVT) program. Cells ranging in capacity from 1 mAh to 1Ah, and containing a variety of electrodes and electrolytes, are examined to determine suitable material combinations that will meet and exceed the FCVT performance, cost, and safety targets. In this article, accelerated aging of 18650-type cells, and characterization of components harvested from these cells, is described. Several techniques that include electrochemical measurements, analytical electron microscopy, and x-ray spectroscopy were used to study the various cell components. Data from these studies were used to identify the most likely contributors to property degradation and determine mechanisms responsible for cell capacity fade and impedance rise.

  6. Methodology of Supervision by Analysis of Thermal Flux for Thermal Conduction of a Batch Chemical Reactor Equipped with a Monofluid Heating/Cooling System

    Directory of Open Access Journals (Sweden)

    Ghania Henini

    2012-01-01

    Full Text Available We present the thermal behavior of a batch reactor to jacket equipped with a monofluid heating/cooling system. Heating and cooling are provided respectively by an electrical resistance and two plate heat exchangers. The control of the temperature of the reaction is based on the supervision system. This strategy of management of the thermal devices is based on the usage of the thermal flux as manipulated variable. The modulation of the monofluid temperature by acting on the heating power or on the opening degrees of an air-to-open valve that delivers the monofluid to heat exchanger. The study shows that the application of this method for the conduct of the pilot reactor gives good results in simulation and that taking into account the dynamics of the various apparatuses greatly improves ride quality of conduct. In addition thermal control of an exothermic reaction (mononitration shows that the consideration of heat generated in the model representation improve the results by elimination any overshooting of the set-point temperature.

  7. High efficiency removal of 2-chlorophenol from drinking water by a hydrogen-based polyvinyl chloride membrane biofilm reactor.

    Science.gov (United States)

    Xia, Siqing; Zhang, Zhiqiang; Zhong, Fohua; Zhang, Jiao

    2011-02-28

    A continuously stirred hydrogen-based membrane biofilm reactor (MBfR) with polyvinyl chloride (PVC) hollow fiber membrane was investigated for removing 2-chlorophenol (2-CP) from contaminated drinking water. The bioreactor startup was achieved by acclimating the microorganisms from a denitrifying and sulfate-reducing MBfR to the drinking water contaminated by 2-CP. The effects of some major factors, including 2-CP loading, H(2) pressure, nitrate loading, and sulfate loading, on the removal of 2-CP by the MBfR were systematically investigated. Although the effluent 2-CP concentration increased with its increasing influent loading, the removing efficiency of 2-CP by the MBfR could be up to 94.7% under a high influent loading (25.71 mg/L d). The removing efficiency of 2-CP by the MBfR could be improved by higher H(2) pressure, and lower influent nitrate concentration and sulfate concentration. A high H(2) pressure can assure enough available H(2) as the electron donor for 2-CP degradation. The competition in the electron donor made nitrate and sulfate inhibit the degradation of 2-CP in the MBfR. The electron flux analyses indicated that the degradation of 2-CP only accounted for a small part of electron flux, and the autohydrogenotrophic bacteria in the MBfR were highly efficient for the 2-CP removal.

  8. Properties of High Basicity Mold Fluxes for Peritectic Steel Slab Casting

    Institute of Scientific and Technical Information of China (English)

    LONG Xiao; HE Sheng-ping; XU Jian-fei; HUO Xu-ling; WANG Qian

    2012-01-01

    In high speed continuous casting of peritectic steel slabs, mold fluxes with high basicity are required for less surface defect product. However, the basicity of remaining liquid slag film tends to decrease in casting process because of the crystallization of 3CaO ·2SiO2 · CaF2. Thus, a way is put forward to improve mold fluxesr properties by raising the original basicity. In order to confirm the possibility of this method, the effect of rising original basicity on the properties of mold fluxes is discussed. Properties of high fluorine based mold fluxes with different basicities and contents of CaF2 , Na2 O, and MgO were measured, respectively. Then, properties of higher basicity mold fluxes were discussed and compared with traditional ones. The results show that increasing the basicity index can improve the melting and flow property of mold fluxes. With the increasing basicity, crystallization rate of mold fluxes increases obviously and crystallization temperature tends to decrease when the basicity exceeds 1.35. The method presen- ted before is proved as a potential way to resolve the contradiction between horizontal heat transfer controlling and solidified shell lubricating for peritectic steel slab casting. But further study on improving the flow property of liquid slag is needed. This work can be used to guide mold fluxes design for high speed continuous casting of peritectic steel slabs.

  9. Analysis on Zero Power Experiment of High Flux Engineering Test Reactor with Three-Dimensional Continuous Energy Monte Carlo Code%三维堆芯连续能量蒙特卡罗程序用于HFETR零功率物理实验计算分析研究

    Institute of Scientific and Technical Information of China (English)

    彭钢

    2012-01-01

    采用三维堆芯连续能量蒙特卡罗程序( MCNP)对高通量工程试验堆(HFETR)零功率物理实验进行计算分析.从计算结果可以看出,在零功率反应堆上,径向铍反射层应当考虑金属铍中的杂质和密度修正,同时需要考虑控制棒过渡段的10B含量修正;而HFETR上的铍块则可以认为是纯金属铍,控制棒过渡段中10B含量也需另作考虑.从计算结果来看,多数参数(堆芯中子有效增殖系数keff、中子通量密度相对分布、γ剂量率相对分布以及多数部件反应性价值)均较好满足误差要求,而个别小反应性部件计算误差较大可能与MCNP程序模型有关.%Three-dimensional, continuous energy Monte Carlo code (MCNP) is adopted to carry out the analysis of zero power experiment of HFETR. From the results, the impurity, density of Beryllium block and the B concentration in control rod transition part should be carefully determined in the analysis of zero power experiment. While in the experiment of HFETR, the Beryllium block is considered as pure metal and the 10B concentration in control rod transition part is different from that of zero power experiment. From the calculation results, these parameters (effective neutron multiplication factor Keff relative distribution of neutron flux density, y dose rate distribution and component reactivity) are quite fit with the experiment. The difference of small reactivity between calculation and experiment is quite large, and may be related to the deficiency of MCNP model.

  10. 13C labeling analysis of sugars by high resolution-mass spectrometry for metabolic flux analysis.

    Science.gov (United States)

    Acket, Sébastien; Degournay, Anthony; Merlier, Franck; Thomasset, Brigitte

    2017-02-14

    Metabolic flux analysis is particularly complex in plant cells because of highly compartmented metabolism. Analysis of free sugars is interesting because it provides data to define fluxes around hexose, pentose, and triose phosphate pools in different compartment. In this work, we present a method to analyze the isotopomer distribution of free sugars labeled with carbon 13 using a liquid chromatography-high resolution mass spectrometry, without derivatized procedure, adapted for Metabolic flux analysis. Our results showed a good sensitivity, reproducibility and better accuracy to determine isotopic enrichments of free sugars compared to our previous methods [5, 6].

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

    Energy Technology Data Exchange (ETDEWEB)

    Thayer, K.J. [ed.

    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.

  12. Plasma resonance and flux dynamics in layered high-Tc superconductors

    DEFF Research Database (Denmark)

    Pedersen, Niels Falsig; Sakai, S.

    2000-01-01

    Flux dynamics of layered high Tc superconductors are considered with special emphasis on the small oscillation modes. In particular we find the dispersion relation for the plasma modes and discuss the spectra to be observed in microwave experiments.......Flux dynamics of layered high Tc superconductors are considered with special emphasis on the small oscillation modes. In particular we find the dispersion relation for the plasma modes and discuss the spectra to be observed in microwave experiments....

  13. Analysis of High Temperature Reactor Control Rod Worth for the Initial and Full Core

    Science.gov (United States)

    Oktajianto, Hammam; Setiawati, Evi; Anam, Khoirul; Sugito, Heri

    2017-01-01

    Control rod is one important component in a nuclear reactor. In nuclear reactor operations the control rod functions to shut down the reactor. This research analyses ten control rods worth of HTR (High Temperature Reactor) at initial and full core. The HTR in this research adopts HTR-10 China and HTR- of pebble bed. Core calculations are performed by using MCNPX code after modelling the entire parts of core in condition of ten control rods fully withdrawn, all control rods in with 20 cm ranges of depth and the use of one control rod. Pebble bed and moderator balls are distributed in the core zone using a Body Centred Cubic (BCC) lattice by ratio of 57:43. The research results are obtained that the use of one control rod will decrease the reactor criticality of 2.04±0.12 %Δk/k at initial core and 1.57±0.10 %Δk/k at full core. The deeper control rods are in, the lesser criticality of reactor is with reactivity of ten control rods of 16.41±0.11 %Δk/k at initial core and 15.43±0.11 %Δk/k at full core. The results show that the use of ten control rods at full core will keep achieving subcritical condition even though the reactivity is smaller than reactivity at initial core.

  14. An Analysis of Testing Requirements for Fluoride Salt Cooled High Temperature Reactor Components

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Cetiner, Sacit M [ORNL; Flanagan, George F [ORNL; Peretz, Fred J [ORNL; Yoder Jr, Graydon L [ORNL

    2009-11-01

    This report provides guidance on the component testing necessary during the next phase of fluoride salt-cooled high temperature reactor (FHR) development. In particular, the report identifies and describes the reactor component performance and reliability requirements, provides an overview of what information is necessary to provide assurance that components will adequately achieve the requirements, and then provides guidance on how the required performance information can efficiently be obtained. The report includes a system description of a representative test scale FHR reactor. The reactor parameters presented in this report should only be considered as placeholder values until an FHR test scale reactor design is completed. The report focus is bounded at the interface between and the reactor primary coolant salt and the fuel and the gas supply and return to the Brayton cycle power conversion system. The analysis is limited to component level testing and does not address system level testing issues. Further, the report is oriented as a bottom-up testing requirements analysis as opposed to a having a top-down facility description focus.

  15. 超临界水冷堆堆芯简化模型流量分配研究%Code Research on Mass Flux Assignment of Spuercritical Water-Cooled Reactor

    Institute of Scientific and Technical Information of China (English)

    李臻洋; 周涛; 孙灿辉

    2011-01-01

    Taking the fuel assembly of thermal spectrum supercritical water-cooled reactor (SCWR) as the research object, and on the condition of average orifice size, the reactor core power distribution is simulated, and the thermal hydraulic calculation model is established and the corresponding program is developed. The coolant mass flux distribution and related parameters distributions in the parallel channels is calculated. The results show that the axial density distribution and distribution core power of each fuel assembly group channels is very inhomogeneous, causing a large mass flux difference, which could be resolved through increasing the orifice size of high power fuel assembly groups.%选取超临界水冷堆(SCWR)燃料组件作为研究对象,在平均孔口尺寸条件下,对堆芯功率分布进行模拟,建立了热工水力计算模型并进行了程序的开发,计算出了各个并联通道内的冷却剂流量以及相关参数分布.结果表明,平均孔口尺寸条件得到的各组群燃料通道轴向密度分布、堆芯功率分布存在较大的不均匀性,致使流量分配存在较大的差异;通过增大高功率组群的孔板尺寸即可得到较为合理的热工水力参数分布.

  16. Procedure of calculation of the spatial distribution of temperatures and heat fluxes in the steam generator of a nuclear power installation with an RBEC fast-neutron reactor

    Science.gov (United States)

    Frolov, A. A.; Sedov, A. A.

    2016-08-01

    A method for combined 3D/1D-modeling of thermohydraulics of a once-through steam generator (SG) based on the joint analysis of three-dimensional thermo- and hydrodynamics of a single-phase heating coolant in the intertube space and one-dimensional thermohydraulics of steam-generating channels (tubes) with the use of well-known friction and heat-transfer correlations under various boiling conditions is discussed. This method allows one to determine the spatial distribution of temperatures and heat fluxes of heat-exchange surfaces of SGs with a single-phase heating coolant in the intertube space and with steam generation within tubes. The method was applied in the analytical investigation of typical operation of a once-through SG of a nuclear power installation with an RBEC fast-neutron heavy-metal reactor that is being designed by Kurchatov Institute in collaboration with OKB GIDROPRESS and Leipunsky Institute of Physics and Power Engineering. Flow pattern and temperature fields were obtained for the heavy-metal heating coolant in the intertube space. Nonuniformities of heating of the steam-water coolant in different heat-exchange tubes and nonuniformities in the distribution of heat fluxes at SG heat-exchange surfaces were revealed.

  17. Advanced high-pressure bench-scale reactor for testing with hot corrosive gases

    Energy Technology Data Exchange (ETDEWEB)

    Abbasian, J.; Bachta, R.P.; Wangerow, J.R. (Inst. of Gas Technology, Chicago, IL (United States)); Mojtahedi, W.; Salo, K. (Enviropower Inc., Espoo (Finland))

    1994-01-01

    A bench-scale, high-pressure/high-temperature fluidized-bed reactor (HPTR) system is described that is capable of operating at a maximum temperature and pressure of 1,000 C and 30 bar in a corrosive atmosphere. The design of the unit is based on a double-shell balanced-pressure system. All the hot parts of the reactor that are wetted by the corrosive (and/or reactive) gases and the entire sampling line are constructed of inert material to prevent corrosion and loss of the reactant gases. The unit has been used for over 200 high-pressure hot coal gas desulfurization tests at 20 bars and up to 750 C without any experimental problem and with excellent sulfur balance, indicating that this reactor system is ideal for testing with reactive and corrosive gases at elevated pressures and temperatures.

  18. Nuclear Rocket Engine Reactor

    CERN Document Server

    Lanin, Anatoly

    2013-01-01

    The development of a nuclear rocket engine reactor (NRER ) is presented in this book. The working capacity of an active zone NRER under mechanical and thermal load, intensive neutron fluxes, high energy generation (up to 30 MBT/l) in a working medium (hydrogen) at temperatures up to 3100 K is displayed. Design principles and bearing capacity of reactors area discussed on the basis of simulation experiments and test data of a prototype reactor. Property data of dense constructional, porous thermal insulating and fuel materials like carbide and uranium carbide compounds in the temperatures interval 300 - 3000 K are presented. Technological aspects of strength and thermal strength resistance of materials are considered. The design procedure of possible emergency processes in the NRER is developed and risks for their origination are evaluated. Prospects of the NRER development for pilotless space devices and piloted interplanetary ships are viewed.

  19. Tiger team assessment of the Argonne Illinois site

    Energy Technology Data Exchange (ETDEWEB)

    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. First in-core simultaneous measurements of nuclear heating and thermal neutron flux obtained with the innovative mobile calorimeter CALMOS inside the OSIRIS reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lepeltier, Valerie; Bubendorff, Jacques; Carcreff, Hubert [Nuclear studies and reactor irradiation Service, CEA Saclay 91191 Gif sur Yvette (France); Salmon, Laurent [Thermalhydraulics and Fluid Mechanics Section, CEA Saclay 91191 Gif sur Yvette, (France)

    2015-07-01

    Nuclear heating inside a MTR reactor has to be known in order to design and to run irradiation experiments which have to fulfill target temperature constraints. This measurement is usually carried out by calorimetry. The innovative calorimetric system, CALMOS, has been studied and built in 2011 for the 70 MWth OSIRIS reactor operated by CEA. Thanks to a new type of calorimetric probe, associated to a specific displacement system, it provides measurements along the fissile height and above the core. This development required preliminary modelling and irradiation of mock-ups of the calorimetric probe in the ex-core area, where nuclear heating rate does not exceed 2 W.g{sup -1}. The calorimeter working modes, the different measurement procedures allowed with such a new probe, the main modeling and experimental results and expected advantages of this new technique have been already presented. However, these first in-core measurements were not performed beyond 6 W.g{sup -1}, due to an inside temperature limitation imposed by a safety authority requirement. In this paper, we present the first in-core simultaneous measurements of nuclear heating and conventional thermal neutron flux obtained by the CALMOS device at the 70 MW nominal reactor power. For the first time, this experimental system was operated in nominal in-core conditions, with nominal neutron flux up to 2.7 10{sup 14} n.cm{sup -2}.s{sup -1} and nuclear heating up to 12 W.g{sup -1}. A comprehensive measurement campaign carried out from 2013 to 2015 inside all accessible irradiation locations of the core, allowed to qualify definitively this new device, not only in terms of measurement ability but also in terms of reliability. After a brief reminder of the calorimetric cell configuration and displacement system specificities, first nuclear heating distributions at nominal power are presented and discussed. In order to reinforce the heating evaluation, a systematic comparison is made between results obtained by

  1. New design targets and new automated technology for the production of radionuclides with high specificity radioactivity in nuclear research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gerasimov, A.S.; Kiselev, G.V. [State Russian Center of the Russian Federation, Moscow (Russian Federation). Institute of Theoretical and Experimental Physics

    1997-10-01

    Current demands of industry require the application of radionuclides with high specific radioactivity under low consumption of neutrons. To provide this aim staff of ITEP Reactor Department investigated the different type AEs of start targets for the production of the main radionuclides; Co-60, Ir-192 and others. In first turn the targets of Co and Ir without the block-effect of neutron flux (with low absorption of neutrons) were investigated. The following principal results were received for example for Ir-192: block effect is equal 0.086 for diameter of Ir target mm and is equal 0.615 for diameter Ir target 0.5mm. It means average neutron flux for Ir target diameter 0.5mm and therefore the production of Ir-192 will be at 10 times more than for diameter 6.0mm. To provide the automated technology of the manufacture of radioactive sources with radionuclides with high specific radioactivity it was proposed that the compound targets for the irradiation of ones and for the management with the irradiated targets. Different types of compound targets were analyzed. (authors)

  2. Numerical Study of High Heat Flux Performances of Flat-Tile Divertor Mock-ups with Hypervapotron Cooling Concept

    Science.gov (United States)

    Chen, Lei; Liu, Xiang; Lian, Youyun; Cai, Laizhong

    2015-09-01

    The hypervapotron (HV), as an enhanced heat transfer technique, will be used for ITER divertor components in the dome region as well as the enhanced heat flux first wall panels. W-Cu brazing technology has been developed at SWIP (Southwestern Institute of Physics), and one W/CuCrZr/316LN component of 450 mm×52 mm×166 mm with HV cooling channels will be fabricated for high heat flux (HHF) tests. Before that a relevant analysis was carried out to optimize the structure of divertor component elements. ANSYS-CFX was used in CFD analysis and ABAQUS was adopted for thermal-mechanical calculations. Commercial code FE-SAFE was adopted to compute the fatigue life of the component. The tile size, thickness of tungsten tiles and the slit width among tungsten tiles were optimized and its HHF performances under International Thermonuclear Experimental Reactor (ITER) loading conditions were simulated. One brand new tokamak HL-2M with advanced divertor configuration is under construction in SWIP, where ITER-like flat-tile divertor components are adopted. This optimized design is expected to supply valuable data for HL-2M tokamak. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2011GB110001 and 2011GB110004)

  3. High rate nitrogen removal by ANAMMOX internal circulation reactor (IC) for old landfill leachate treatment.

    Science.gov (United States)

    Phan, The Nhat; Van Truong, Thi Thanh; Ha, Nhu Biec; Nguyen, Phuoc Dan; Bui, Xuan Thanh; Dang, Bao Trong; Doan, Van Tuan; Park, Joonhong; Guo, Wenshan; Ngo, Huu Hao

    2017-06-01

    This study aimed to evaluate the performance of a high rate nitrogen removal lab-scale ANAMMOX reactor, namely Internal Circulation (IC) reactor, for old landfill leachate treatment. The reactor was operated with pre-treated leachate from a pilot Partial Nitritation Reactor (PNR) using a high nitrogen loading rate ranging from 2 to 10kgNm(-3)d(-1). High rate removal of nitrogen (9.52±1.11kgNm(-3)d(-1)) was observed at an influent nitrogen concentration of 1500mgNL(-1). The specific ANAMMOX activity was found to be 0.598±0.026gN2-NgVSS(-1)d(-1). Analysis of ANAMMOX granules suggested that 0.5-1.0mm size granular sludge was the dominant group. The results of DNA analysis revealed that Candidatus Kueneniastuttgartiensis was the dominant species (37.45%) in the IC reactor, whereas other species like uncultured Bacteroidetes bacterium only constituted 5.37% in the system, but they were still responsible for removing recalcitrant organic matter.

  4. Nitrate removal from high strength nitrate-bearing wastes in granular sludge sequencing batch reactors.

    Science.gov (United States)

    Krishna Mohan, Tulasi Venkata; Renu, Kadali; Nancharaiah, Yarlagadda Venkata; Satya Sai, Pedapati Murali; Venugopalan, Vayalam Purath

    2016-02-01

    A 6-L sequencing batch reactor (SBR) was operated for development of granular sludge capable of denitrification of high strength nitrates. Complete and stable denitrification of up to 5420 mg L(-1) nitrate-N (2710 mg L(-1) nitrate-N in reactor) was achieved by feeding simulated nitrate waste at a C/N ratio of 3. Compact and dense denitrifying granular sludge with relatively stable microbial community was developed during reactor operation. Accumulation of large amounts of nitrite due to incomplete denitrification occurred when the SBR was fed with 5420 mg L(-1) NO3-N at a C/N ratio of 2. Complete denitrification could not be achieved at this C/N ratio, even after one week of reactor operation as the nitrite levels continued to accumulate. In order to improve denitrification performance, the reactor was fed with nitrate concentrations of 1354 mg L(-1), while keeping C/N ratio at 2. Subsequently, nitrate concentration in the feed was increased in a step-wise manner to establish complete denitrification of 5420 mg L(-1) NO3-N at a C/N ratio of 2. The results show that substrate concentration plays an important role in denitrification of high strength nitrate by influencing nitrite accumulation. Complete denitrification of high strength nitrates can be achieved at lower substrate concentrations, by an appropriate acclimatization strategy.

  5. Catalytic liquid marbles: Ag nanowire-based miniature reactors for highly efficient degradation of methylene blue.

    Science.gov (United States)

    Miao, Yue-E; Lee, Hiang Kwee; Chew, Wee Shern; Phang, In Yee; Liu, Tianxi; Ling, Xing Yi

    2014-06-04

    Ag nanowire-based catalytic liquid marbles are fabricated as miniature reactors, which demonstrate highly efficient, support-free and rate-controllable heterogeneous degradation of methylene blue, with catalytic efficiency close to 100%. Our miniature catalytic liquid marbles are essential for reactions involving highly toxic/hazardous or costly reactants, where small volume preliminary reactions are preferred.

  6. Control of Crystal Morphology for Mold Flux During High-Aluminum AHSS Continuous Casting Process

    Science.gov (United States)

    GUO, Jing; SEO, Myung-Duk; SHI, Cheng-Bin; CHO, Jung-Wook; KIM, Seon-Hyo

    2016-08-01

    In the present manuscript, the efforts to control the crystal morphology are carried out aiming at improving the lubrication of lime-alumina-based mold flux for casting advanced high-strength steel with high aluminum. Jackson α factors for crystals of melt crystallization in multi-component mold fluxes are established and reasonably evaluated by applying thermodynamic databases to understand the crystal morphology control both in lime-alumina-based and lime-silica-based mold fluxes. The results show that Jackson α factor and supercooling are the most critical factors to determine the crystal morphology in a mold flux. Crystals precipitating in mold fluxes appear with different morphologies due to their different Jackson α factors and are likely to be more faceted with higher Jackson α factor. In addition, there is a critical supercooling degree for crystal morphology dendritic transition. When the supercooling over the critical value, the crystals transform from faceted shape to dendritic ones in morphology as the kinetic roughening occurs. Typically, the critical supercooling degrees for cuspidine dendritic transition in the lime-silica-based mold fluxes are evaluated to be between 0.05 and 0.06. Finally, addition of a small amount of Li2O in the mold flux can increase the Jackson α factor and decrease the supercooling for cuspidine precipitation; thus, it is favorable to enhance a faceted cuspidine crystal.

  7. Fractal Pattern Growth in Ti-Implanted Steel with High Ion Flux

    Institute of Scientific and Technical Information of China (English)

    张通和; 吴瑜光; 刘安东

    2002-01-01

    We report on the formation of metal nanometre phase and fractal patterns in steel using metal vapour vacuum arc source ion implantation with high ion flux. The dense nanometre phases are cylindrical and well dispersed in the Ti-implanted layer with an ion flux up to 50μA/cm2. The collision fractal pattern is formed in Ti-implanted steel with an ion flux of 25μA/cm2 and the disconnected fractal pattern is observed with an ion flux of 50μA/cm2.The average density ofnanometre phases decreases from 1.2 × 1011/cm2 to 6.5 × 1010/cm2 as the ion flux increases from 25 μA/cm2 to 50 μA/cm2. Fractal pattern growth is in remarkable agreement with Sander's diffusion-limited aggregation model. The alloy clusters have diffused and aggregated in chains forming branches to grow a beautiful tree during Ti implantation with an ion flux ranging from 75μA/cm2 to 85μA/cm2. We discuss the model of fractal pattern growth during ion implantation with high ion flux.

  8. Homopolar Gun for Pulsed Spheromak Fusion Reactors II

    Energy Technology Data Exchange (ETDEWEB)

    Fowler, T

    2004-06-14

    A homopolar gun is discussed that could produce the high currents required for pulsed spheromak fusion reactors even with unit current amplification and open field lines during injection, possible because close coupling between the gun and flux conserver reduces gun losses to acceptable levels. Example parameters are given for a gun compatible with low cost pulsed reactors and for experiments to develop the concept.

  9. Gas-Liquid Mass Transfer in a Slurry Bubble Column Reactor under High Temperature and

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The gas-liquid mass transfer of H2 and CO in a high temperature and high-pressure three-phase slurry bubble column reactor is studied. The gas-liquid volumetric mass transfer coefficients κLα are obtained by measuring the dissolution rate of H2 and CO. The influences of the main operation conditions, such as temperature, pressure, superficial gas velocity and solid concentration, are studied systematically. Two empirical correlations are proposed to predict κLα values for H2 and CO in liquid paraffin/solid particles slurry bubble column reactors.

  10. Advanced High-Temperature Reactor Dynamic System Model Development: April 2012 Status

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A L; Cetiner, M S; Wilson, Jr, T L

    2012-04-30

    The Advanced High-Temperature Reactor (AHTR) is a large-output fluoride-salt-cooled high-temperature reactor (FHR). An early-phase preconceptual design of a 1500 MW(e) power plant was developed in 2011 [Refs. 1 and 2]. An updated version of this plant is shown as Fig. 1. FHRs feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR is designed to be a “walk away” reactor that requires no action to prevent large off-site releases following even severe reactor accidents. This report describes the development of dynamic system models used to further the AHTR design toward that goal. These models predict system response during warmup, startup, normal operation, and limited off-normal operating conditions. Severe accidents that include a loss-of-fluid inventory are not currently modeled. The scope of the models is limited to the plant power system, including the reactor, the primary and intermediate heat transport systems, the power conversion system, and safety-related or auxiliary heat removal systems. The primary coolant system, the intermediate heat transport system and the reactor building structure surrounding them are shown in Fig. 2. These systems are modeled in the most detail because the passive interaction of the primary system with the surrounding structure and heat removal systems, and ultimately the environment, protects the reactor fuel and the vessel from damage during severe reactor transients. The reactor silo also plays an important role during system warmup. The dynamic system modeling tools predict system performance and response. The goal is to accurately predict temperatures and pressures within the primary, intermediate, and power conversion systems and to study the impacts of design changes on those responses. The models are design tools and are not intended to be used in reactor qualification. The important details to capture in the primary

  11. A fast and flexible reactor physics model for simulating neutron spectra and depletion in fast reactors

    Science.gov (United States)

    Recktenwald, Geoff; Deinert, Mark

    2010-03-01

    Determining the time dependent concentration of isotopes within a nuclear reactor core is central to the analysis of nuclear fuel cycles. We present a fast, flexible tool for determining the time dependent neutron spectrum within fast reactors. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to simulate the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. While originally developed for LWR simulations, the model is shown to produce fast reactor spectra that show high degree of fidelity to available fast reactor benchmarks.

  12. Study on radiation damage to high energy accelerator components by irradiation in a nuclear reactor

    CERN Document Server

    Schönbacher, Helmut; Casta, J; Van de Voorde, M H

    1975-01-01

    The structural and other components used in high energy accelerators are continuously exposed to a wide spectrum of high energy particles and electromagnetic radiation. The resulting radiation damage may severely influence the functional capability of accelerator facilities. In order to arrive at an estimate of the service life of various materials in the radiation field, simulating experiments have to be carried out in a nuclear reactor. A large number of organic and inorganic materials, electronic components, metals, etc., intended specifically for use in 400 GeV proton synchrotron of CERN near Geneva, were irradiated in the ASTRA reactor in Seibersdorf near Vienna. The paper reports on the irradiation facilities available in this reactor for this purpose, on the dosimetry methods used, on the most important materials irradiated and on the results obtained in these experiments. (14 refs).

  13. THERMAL NEUTRON FLUX MAPPING ON A TARGET CAPSULE AT RABBIT FACILITY OF RSG-GAS REACTOR FOR USE IN k0-INAA

    Directory of Open Access Journals (Sweden)

    Sutisna Sutisna

    2015-03-01

    Full Text Available Instrumental neutron activation analysis based on the k0 method (k0-INAA requires the availability of the accurate reactor parameter data, in particular a thermal neutron flux that interact with a targets inside the target capsule. This research aims to determine and map the thermal neutron flux inside the capsule and irradiation channels used for the elemental quantification using the k0-AANI. Mapping of the thermal neutron flux (фth on two type of irradiation capsule have been done for RS01 and RS02 facilities of RSG-GAS reactor. Thermal neutron flux determined using Al-0,1%Au alloy through 197Au(n,g 198Au nuclear reaction, while the flux mapping done using statistics R. Thermal neutron flux are calculated using k0-IAEA software provided by IAEA. The results showed the average thermal neutron flux is (5.6±0.3×10+13 n.cm-2.s-1; (5.6±0.4×10+13 n.cm-2.s-1; (5.2±0.4×10+13 n.cm-2.s-1 and (5.3±0.4×10+13 n.cm-2.s-1 for Polyethylene capsule of 1st , 2nd, 3rd and 4th layer respectively. In the case of Aluminum capsule, the thermal neutron flux was lower compared to that on Polyethylene capsule. There were (3.0±0.2×10+13 n.cm-2.s-1; (2.8±0.1×10+13 n.cm-2.s-1; (3.2±0.3×10+13 n.cm-2.s-1 for 1st, 2nd and 3rd layers respectively. For each layer in the capsule, the thermal neutron flux is not uniform and it was no degradation flux in the axial direction, both for polyethylene and aluminum capsules. Contour map of eight layer on polyethylene capsule and six layers on aluminum capsule for RS01 and RS02 irradiation channels had a similar pattern with a small diversity for all type of the irradiation capsule. Keywords: thermal neutron, flux, capsule, NAA   Analisis aktivasi neutron instrumental berbasis metode k0 (k0-AANI memerlukan ketersediaan data parameter reaktor yang akurat, khususnya data fluks neutron termal yang berinteraksi dengan inti sasaran di dalam kapsul target. Penelitian ini bertujuan menentukan dan memetakan fluks neutron termal

  14. High temperature thermocouple and heat flux gauge using a unique thin film-hardware hot juncture

    Science.gov (United States)

    Liebert, C. H.; Holanda, R.; Hippensteele, S. A.; Andracchio, C. A.

    1984-01-01

    A special thin film-hardware material thermocouple (TC) and heat flux gauge concept for a reasonably high temperature and high flux flat plate heat transfer experiment was fabricated and tested to gauge temperatures of 911 K. This concept was developed for minimal disturbance of boundary layer temperature and flow over the plates and minimal disturbance of heat flux through the plates. Comparison of special heat flux gauge Stanton number output at steady-state conditions with benchmark literature data was good and agreement was within a calculated uncertainty of the measurement system. Also, good agreement of special TC and standard TC outputs was obtained and the results are encouraging. Oxidation of thin film thermoelements was a primary failure mode after about 5 of operation.

  15. Design of high temperature irradiation materials inspection cells. (Spent fuel inspection cells) in the High Temperature Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ino, Hiroichi; Ueta, Shouhei; Suzuki, Hiroshi; Sawa, Kazuhiro [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Tobita, Tsutomu [Nuclear Engineering Company, Ltd., Tokai, Ibaraki (Japan)

    2002-01-01

    This report summarizes design requirements and design results for shields, ventilation system and fuel handling devices for the high temperature irradiation materials inspection cells (spent fuel inspection cells). These cells are small cells to carry out few post-irradiation examinations of spent fuels, specimen, etc., which are irradiated in the High Temperature Engineering Test Reactor, since the cells should be built in limited space in the HTTR reactor building, the cells are designed considering relationship between the cells and the reactor building to utilize the limited space effectively. The cells consist of three partitioned hot cells with wall for neutron and gamma-ray shields, ventilation system including filtering units and fuel handling devices. The post-irradiation examinations of the fuels and materials are planed by using the cells and the Hot Laboratory of the Japan Materials Testing Reactor to establish the technology basis on high temperature gas-cooled reactors (HTGRs). In future, irradiation tests and post-irradiation examinations will be carried out with the cells to upgrade present HTGR technologies and to make the innovative basic research on high-temperature engineering. (author)

  16. Reflectance-Based Estimation of Soil Heat Fluxes in the Texas High Plains

    Science.gov (United States)

    Gowda, P. H.; Colaizzi, P. D.; O'Shaughnessy, S.; Ha, W.; Howell, T. A.

    2010-12-01

    Soil heat flux (G) is one of the terms required for estimating evapotranspiration rates using an energy balance. Numerous reflectance-based models are available in the literature for estimating G fluxes. However, these models have shown wide variation in their performance. Therefore, operational ET remote sensing programs may require locally developed/calibrated models for accurately estimating G. The objective of this study was to develop and evaluate reflectance-based empirical G models for the semi-arid Texas High Plains. Soil heat flux was measured at 0.15 hz interval and averaged every 15 minutes at five different locations within a 4.7 ha lysimeter field with Pullman clay loam soil during the 2010 summer growing season. The field was planted to soybean and managed under dryland conditions. In each location, G was measured at 8 cm depth with two Campbell Scientific HFT3 soil heat flux plates. Soil temperature was measured at 2 and 6 cm above the soil heat flux plates. Soil moisture was measured in the 2-8 cm layer using Acclima SDI-12 sensors. Hourly G fluxes at the surface were calculated by adding the measured G fluxes at 8 cm to the energy stored above the heat flux plates. A multispectral radiometer (MSR, CROPSCAN, Inc.) and hand-held thermometer (EVEREST Interscience Inc.) measured surface reflectance in red and near infrared bandwidths and surface temperature (ST), respectively, daily at 11:30 AM CST to be consistent with the Landsat 5 overpass time. Fraction crop cover (FC) was measured by digital photographs taken twice a week. A set of G models was developed for estimating hourly fluxes based on measured reflectance, net radiation, ST, NDVI, and FC,. Resulting models were compared for performance with existing models available in the literature. In this presentation, we will discuss our G models for the Texas High Plains and the statistical results.

  17. Superconducting flux pump for high-temperature superconductor insert coils of NMR magnets

    Science.gov (United States)

    Jeong, S.; Lee, H.; Iwasa, Y.

    2002-05-01

    This paper describes a prototype flux pump recently operated at the MIT Francis Bitter Magnet Laboratory. The results of the prototype flux pump will be used in the development of a full-scale flux pump that will be coupled to a high-temperature superconductor (HTS) insert coil of a high-field NMR magnet. Such an HTS insert is unlikely to operate in persistent mode because of the conductor's low index (n). The flux pump can compensate for field decay in the HTS insert coil and make the insert operate effectively in persistent mode. The flux pump, comprised essentially of a transformer and two switches, all made of superconductor, transfers into the insert coil a fraction of a magnetic energy that is first introduced in the secondary circuit of the transformer by a current supplied to the primary circuit. A prototype flux pump has been designed, fabricated, and operated to demonstrate that a flux pump can indeed supply a small metered current into a load superconducting magnet. A current increment in the range of microamperes has been measured in the magnet after each pumping action. The superconducting prototype flux pump is made of Nb3Sn tape. The pump is placed in a gaseous environment above the liquid helium level to keep its heat dissipation from directly discharged in the liquid; the effluent helium vapor maintains the thermal stability of the flux pump. [This paper is also published in Advances in Cryogenic Engineering Volume 47A, AIP Conference Proceedings Volume 613, pp. 441-448.

  18. Autophagic flux is highly active in early mitosis and differentially regulated throughout the cell cycle

    OpenAIRE

    Li, Zhiyuan; Ji, Xinmiao; Wang, Dongmei; Liu, Juanjuan; Zhang, Xin

    2016-01-01

    Mitosis is a fast process that involves dramatic cellular remodeling and has a high energy demand. Whether autophagy is active or inactive during the early stages of mitosis in a naturally dividing cell is still debated. Here we aimed to use multiple assays to resolve this apparent discrepancy. Although the LC3 puncta number was reduced in mitosis, the four different cell lines we tested all have active autophagic flux in both interphase and mitosis. In addition, the autophagic flux was highl...

  19. Behavior of TPC`s in a high particle flux environment

    Energy Technology Data Exchange (ETDEWEB)

    Etkin, A.; Eisemann, S.E.; Foley, K.J.; Hackenburg, R.W.; Longacre, R.S.; Love, W.A.; Morris, T.W.; Platner, E.D.; Saulys, A.C. [Brookhaven National Lab., Upton, NY (United States); Lindenbaum, S.J. [Brookhaven National Lab., Upton, NY (United States)]|[City Coll., New York, NY (United States); Chan, C.S.; Kramer, M.A.; Zhao, K.H.; Zhu, Y. [City Coll., New York, NY (United States); Hallman, T.J.; Madansky, L. [Johns Hopkins Univ., Baltimore, MD (United States); Ahmad, S.; Bonner, B.E.; Buchanan, J.A.; Chiou, C.N.; Clement, J.M.; Mutchler, G.S.; Roberts, J.B. [Rice Univ., Houston, TX (United States). Bonner Nuclear Labs.

    1991-12-13

    TPC`s (Time Projection Chamber) used in E-810 at the AGS (Alternating Gradient Synchrotron) were exposed to fluxes equivalent to more than 10{sup 7} minimum ionizing particles per second to find if such high fluxes cause gain changes or distortions of the electric field. Initial results of these and other tests are presented and the consequences for the RHIC (Relativistic Heavy Ion Collider) TPC-based experiments are discussed.

  20. Behavior of TPC`s in a high particle flux environment

    Energy Technology Data Exchange (ETDEWEB)

    Etkin, A.; Eiseman, S.E.; Foley, K.J.; Hackenburg, R.W.; Longacre, R.S.; Love, W.A.; Morris, T.W.; Platner, E.D.; Saulys, A.C. [Brookhaven National Lab., Upton, NY (United States); Lindenbaum, S.J. [Brookhaven National Lab., Upton, NY (United States)]|[City Coll., New York, NY (United States); Chan, C.S.; Kramer, M.A.; Zhao, K.H.; Zhu, Y. [City Coll., New York, NY (United States); Hallman, T.J.; Madansky, L. [Johns Hopkins Univ., Baltimore, MD (United States); Ahmad, S.; Bonner, B.E.; Buchanan, J.A.; Chiou, C.N.; Clement, J.M.; Mutchler, G.S.; Roberts, J.B. [Rice Univ., Houston, TX (United States)

    1992-07-08

    TPC`s (Time Projection Chamber) used in E-810 at the AGS (Alternating Gradient Synchrotron) were exposed to fluxes equivalent to more than 10{sup 7} minimum ionizing particles per second to find if such high fluxes cause gain changes or distortions of the electric field. Initial results of these and other tests are presented and the consequences for the RHIC (Relativistic Heavy Ion collider) TPC-based experiments are discussed.

  1. Behavior of TPC`s in a high particle flux environment

    Energy Technology Data Exchange (ETDEWEB)

    Etkin, A.; Eiseman, S.E.; Foley, K.J.; Hackenburg, R.W.; Longacre, R.S.; Love, W.A.; Morris, T.W.; Platner, E.D.; Saulys, A.C.; Lindenbaum, S.J. [Brookhaven National Lab., Upton, NY (United States); Chan, C.S.; Kramer, M.A.; Zhao, K.H.; Zhu, Y. [City College of New York, New York (United States); Hallman, T.J.; Madansky, L. [Johns Hopkins Univ., Baltimore, MD (United States); Ahmad, S.; Bonner, B.E.; Buchanan, J.A.; Chiou, C.N.; Clement, J.M.; Mutchler, G.S.; Roberts, J.B. [Bonner Nuclear Lab., Houston, TX (United States)

    1991-12-31

    TPC`s (Time Projection Chamber) used in E-810 at the TAGS (Alternating Gradient Synchrotron) were exposed to fluxes equivalent to more than 10 minimum ionizing particles per second to find if such high fluxes cause gain changes or distortions of the electric field. Initial results of these and other tests are presented and the consequences for the RHIC (Relativistic Heavy Ion Collider) TPC-based experiments are discussed.

  2. Behavior of TPC`s in a high particle flux environment

    Energy Technology Data Exchange (ETDEWEB)

    Etkin, A.; Eiseman, S.E.; Foley, K.J.; Hackenburg, R.W.; Longacre, R.S.; Love, W.A.; Morris, T.W.; Platner, E.D.; Saulys, A.C. [Brookhaven National Lab., Upton, NY (United States); Lindenbaum, S.J. [Brookhaven National Lab., Upton, NY (United States)]|[City Coll., New York, NY (United States); Chan, C.S.; Kramer, M.A.; Zhao, K.H.; Zhu, Y. [City Coll., New York, NY (United States); Hallman, T.J.; Madansky, L. [Johns Hopkins Univ., Baltimore, MD (United States); Ahmad, S.; Bonner, B.E.; Buchanan, J.A.; Chiou, C.N.; Clement, J.M.; Mutchler, G.S.; Roberts, J.B. [Rice Univ., Houston, TX (United States). Bonner Nuclear Labs.

    1991-12-31

    TPC`s (Time Projection Chamber) used in E-810 at the AGS (Alternating Gradient Synchrotron) were exposed to fluxes equivalent to more than 10{sup 7} minimum ionizing particles per second to find if such high fluxes cause gain changes or distortions of the electric field. Initial results of these and other tests are presented and the consequences for the RHIC (Relativistic Heavy Ion Collider) TPC-based experiments are discussed.

  3. Evaluation of strategies for end storage of high-level reactor fuel; Vurdering av strategier for sluttlagring av hoeyaktivt reaktorbrensel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    This report evaluates a national strategy for end-storage of used high-level reactor fuel from the research reactors at Kjeller and in Halden. This strategy presupposes that all the important phases in handling the high-level material, including temporary storage and deposition, are covered. The quantity of spent fuel from Norwegian reactors is quite small. In addition to the technological issues, ethical, environmental, safety and economical requirements are emphasized.

  4. Temperature-dependent surface modification of Ta due to high-flux, low-energy He+ ion irradiation

    Science.gov (United States)

    Novakowski, T. J.; Tripathi, J. K.; Hassanein, A.

    2015-12-01

    This work examines the response of Tantalum (Ta) as a potential candidate for plasma-facing components (PFCs) in future nuclear fusion reactors. Tantalum samples were exposed to high-flux, low-energy He+ ion irradiation at different temperatures in the range of 823-1223 K. The samples were irradiated at normal incidence with 100 eV He+ ions at constant flux of 1.2 × 1021 ions m-2 s-1 to a total fluence of 4.3 × 1024 ions m-2. An additional Ta sample was also irradiated at 1023 K using a higher ion fluence of 1.7 × 1025 ions m-2 (at the same flux of 1.2 × 1021 ions m-2 s-1), to confirm the possibility of fuzz formation at higher fluence. This higher fluence was chosen to roughly correspond to the lower fluence threshold of fuzz formation in Tungsten (W). Surface morphology was characterized with a combination of field-emission scanning electron microscopy (FE-SEM) and atomic force microscopy (AFM). These results demonstrate that the main mode of surface damage is pinholes with an average size of ∼70 nm2 for all temperatures. However, significantly larger pinholes are observed at elevated temperatures (1123 and 1223 K) resulting from the agglomeration of smaller pinholes. Ex situ X-ray photoelectron spectroscopy (XPS) provides information about the oxidation characteristics of irradiated surfaces, showing minimal exfoliation of the irradiated Ta surface. Additionally, optical reflectivity measurements are performed to further characterize radiation damage on Ta samples, showing gradual reductions in the optical reflectivity as a function of temperature.

  5. Nanoengineering of Flux Pinning Sites in High-Tc Superconductors

    Institute of Scientific and Technical Information of China (English)

    2003-01-01

    Volume pinning forces were determined for a variety of bulk high-Tcsuperconductors of the 123-type from magnetization measurements. By means of scaling of the pinning forces, the acting pinning mechanisms in various temperature ranges were identified. The Nd-based superconductors and some YBCO crystalsexhibited a dominating pinning of the δTc-type (i.e., small, superconducting pinning sites). In contrast to this, the addition of insulating 211 particles provided pinning of the δl-type; providing effective pinning in the entire temperature range acting as a "background" pinning mechanism for the peak effect. Due to the small coherence lengths of the high-Tc compounds, effective pinning sites are defects or particles of nanometer size relative to ξ3. Integral magnetic measurements of the magnetization as a function of temperature in large applied magnetic fields (up to 7 T) revealed that practically all high-Tc compounds were spatially inhomogeneous, which could be caused byoxygen deficiency (YBCO), solid solutions of Nd/Ba (NdBCO and other light rare earth compounds), intergrowths (Bi-based superconductors), and doping by pair-breaking dopants like Zn, Pr. This implies that the superconducting sample consists of stronger and weaker superconducting areas, coupled together. In large appliedfields, this coupling gets broken and the magnetization versus temperature curves revealed more than one superconducting transition. In contrast, irradiation experiments by neutrons, protons, and heavy-ions enabled the artificial introduction of very effective pinning sites into the high-Tc superconductors, thus creating a large variety of different observations using magnetic data. From all these observations, we construct a pinning diagram for bulk high-Tc superconductors explaining many features observed in high-Tc samples.

  6. Argonne's contribution to regional development : successful examples.

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y. I.

    2000-11-14

    Argonne National Laboratory's mission is basic research and technology development to meet national goals in scientific leadership, energy technology, and environmental quality. In addition to its core missions as a national research and development center, Argonne has exerted a positive impact on its regional economic development, has carried out outstanding educational programs not only for college/graduate students but also for pre-college students and teachers, and has fostered partnerships with universities for research collaboration and with industry for shaping the new technological frontiers.

  7. Performance model of the Argonne Voyager multimedia server

    Energy Technology Data Exchange (ETDEWEB)

    Disz, T.; Olson, R.; Stevens, R. [Argonne National Lab., IL (United States). Mathematics and Computer Science Div.

    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.

  8. CATARACT: Computer code for improving power calculations at NREL's high-flux solar furnace

    Science.gov (United States)

    Scholl, K.; Bingham, C.; Lewandowski, A.

    1994-01-01

    The High-Flux Solar Furnace (HFSF), operated by the National Renewable Energy Laboratory, uses a camera-based, flux-mapping system to analyze the distribution and to determine total power at the focal point. The flux-mapping system consists of a diffusively reflecting plate with seven circular foil calorimeters, a charge-coupled device (CCD) camera, an IBM-compatible personal computer with a frame-grabber board, and commercial image analysis software. The calorimeters provide flux readings that are used to scale the image captured from the plate by the camera. The image analysis software can estimate total power incident on the plate by integrating under the 3-dimensional image. Because of the physical layout of the HFSF, the camera is positioned at a 20 angle to the flux mapping plate normal. The foreshortening of the captured images that results represents a systematic error in the power calculations because the software incorrectly assumes the image is parallel to the camera's array. We have written a FORTRAN computer program called CATARACT (camera/target angle correction) that we use to transform the original flux-mapper image to a plane that is normal to the camera's optical axis. A description of the code and the results of experiments performed to verify it are presented. Also presented are comparisons of the total power available from the HFSF as determined from the flux mapping system and theoretical considerations.

  9. Flux flow in high-Tc Josephson junctions

    DEFF Research Database (Denmark)

    Filatrella, G.; Pedersen, Niels Falsig

    1993-01-01

    The possibility of achieving fluxon nucleation in nonhysteretic high-T(c) Josephson junctions due to the presence of inhomogeneities is investigated numerically. For a large range of parameters the I- V characteristics in presence of such discontinuities show a strong similarity with those obtain...

  10. Magnetic flux annihilation waves in inhomogeneous high-temperature superconductors

    NARCIS (Netherlands)

    Rudnev, IA; Khodot, AE; Eremin, AV; Mikhailov, BP

    2004-01-01

    The process of magnetic field penetration into polycrystalline high-T-c superconductors of the YBa2Cu3O7 - x and Bi2Sr2Ca2Cu3O10 - x systems has been studied using traditional magnetooptical methods and scanning Hall probe microscopy. It is established that remagnetization of a sample is accompanied

  11. Study of hydrogen generation plant coupled to high temperature gas cooled reactor

    Science.gov (United States)

    Brown, Nicholas Robert

    Hydrogen generation using a high temperature nuclear reactor as a thermal driving vector is a promising future option for energy carrier production. In this scheme, the heat from the nuclear reactor drives an endothermic water-splitting plant, via coupling, through an intermediate heat exchanger. While both high temperature nuclear reactors and hydrogen generation plants have high individual degrees of development, study of the coupled plant is lacking. Particularly absent are considerations of the transient behavior of the coupled plant, as well as studies of the safety of the overall plant. The aim of this document is to contribute knowledge to the effort of nuclear hydrogen generation. In particular, this study regards identification of safety issues in the coupled plant and the transient modeling of some leading candidates for implementation in the Nuclear Hydrogen Initiative (NHI). The Sulfur Iodine (SI) and Hybrid Sulfur (HyS) cycles are considered as candidate hydrogen generation schemes. Several thermodynamically derived chemical reaction chamber models are coupled to a well-known reference design of a high temperature nuclear reactor. These chemical reaction chamber models have several dimensions of validation, including detailed steady state flowsheets, integrated loop test data, and bench scale chemical kinetics. Eight unique case studies are performed based on a thorough literature review of possible events. The case studies are: (1) feed flow failure from one section of the chemical plant to another, (2) product flow failure (recycle) within the chemical plant, (3) rupture or explosion within the chemical plant, (4) nuclear reactor helium inlet overcooling due to a process holding tank failure, (5) helium inlet overcooling as an anticipated transient without SCRAM, (6) total failure of the chemical plant, (7) parametric study of the temperature in an individual reaction chamber, and (8) control rod insertion in the nuclear reactor. Various parametric

  12. Investigation of Isfahan miniature neutron source reactor (MNSR) for boron neutron capture therapy by MCNP simulation

    OpenAIRE

    S. Z. Kalantari; H Tavakoli; Nami, M.

    2015-01-01

    One of the important neutron sources for Boron Neutron Capture Therapy (BNCT) is a nuclear reactor. It needs a high flux of epithermal neutrons. The optimum conditions of the neutron spectra for BNCT are provided by the International Atomic Energy Agency (IAEA). In this paper, Miniature Neutron Source Reactor (MNSR) as a neutron source for BNCT was investigated. For this purpose, we designed a Beam Shaping Assembly (BSA) for the reactor and the neutron transport from the core of the reactor t...

  13. Growth of a dry spot under a vapor bubble at high heat flux and high pressure

    CERN Document Server

    Nikolayev, Vadim; Lagier, G -L; Hegseth, J

    2016-01-01

    We report a 2D modeling of the thermal diffusion-controlled growth of a vapor bubble attached to a heating surface during saturated boiling. The heat conduction problem is solved in a liquid that surrounds a bubble with a free boundary and in a semi-infinite solid heater by the boundary element method. At high system pressure the bubble is assumed to grow slowly, its shape being defined by the surface tension and the vapor recoil force, a force coming from the liquid evaporating into the bubble. It is shown that at some typical time the dry spot under the bubble begins to grow rapidly under the action of the vapor recoil. Such a bubble can eventually spread into a vapor film that can separate the liquid from the heater thus triggering the boiling crisis (critical heat flux).

  14. Validation of SCALE for High Temperature Gas-Cooled Reactors Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Ilas, Dan [ORNL; Kelly, Ryan P [ORNL; Sunny, Eva E [ORNL

    2012-08-01

    This report documents verification and validation studies carried out to assess the performance of the SCALE code system methods and nuclear data for modeling and analysis of High Temperature Gas-Cooled Reactor (HTGR) configurations. Validation data were available from the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhE Handbook), prepared by the International Reactor Physics Experiment Evaluation Project, for two different HTGR designs: prismatic and pebble bed. SCALE models have been developed for HTTR, a prismatic fuel design reactor operated in Japan and HTR-10, a pebble bed reactor operated in China. The models were based on benchmark specifications included in the 2009, 2010, and 2011 releases of the IRPhE Handbook. SCALE models for the HTR-PROTEUS pebble bed configuration at the PROTEUS critical facility in Switzerland have also been developed, based on benchmark specifications included in a 2009 IRPhE draft benchmark. The development of the SCALE models has involved a series of investigations to identify particular issues associated with modeling the physics of HTGRs and to understand and quantify the effect of particular modeling assumptions on calculation-to-experiment comparisons.

  15. Reactor Testing and Qualification: Prioritized High-level Criticality Testing Needs

    Energy Technology Data Exchange (ETDEWEB)

    S. Bragg-Sitton; J. Bess; J. Werner; G. Harms; S. Bailey

    2011-09-01

    Researchers at the Idaho National Laboratory (INL) were tasked with reviewing possible criticality testing needs to support development of the fission surface power system reactor design. Reactor physics testing can provide significant information to aid in development of technologies associated with small, fast spectrum reactors that could be applied for non-terrestrial power systems, leading to eventual system qualification. Several studies have been conducted in recent years to assess the data and analyses required to design and build a space fission power system with high confidence that the system will perform as designed [Marcille, 2004a, 2004b; Weaver, 2007; Parry et al., 2008]. This report will provide a summary of previous critical tests and physics measurements that are potentially applicable to the current reactor design (both those that have been benchmarked and those not yet benchmarked), summarize recent studies of potential nuclear testing needs for space reactor development and their applicability to the current baseline fission surface power (FSP) system design, and provide an overview of a suite of tests (separate effects, sub-critical or critical) that could fill in the information database to improve the accuracy of physics modeling efforts as the FSP design is refined. Some recommendations for tasks that could be completed in the near term are also included. Specific recommendations on critical test configurations will be reserved until after the sensitivity analyses being conducted by Los Alamos National Laboratory (LANL) are completed (due August 2011).

  16. Development of a fixed-bed anammox reactor with high treatment potential.

    Science.gov (United States)

    Okamoto, Hiroyuki; Kawamura, Kimito; Nishiyama, Takashi; Fujii, Takao; Furukawa, Kenji

    2013-02-01

    A plug-flow type anaerobic ammonium oxidation (anammox) reactor was developed using malt ceramics (MC) produced from carbonized spent grains as the biomass carriers for anammox sludge. Partial nitrified effluent of the filtrate from the sludge dehydrator of a brewery company was used as influent to a 20 L anammox reactor using MC. An average volumetric nitrogen removal rate (VNR) of 8.78 kg-N/m(3)/day was maintained stably for 76 days with 1 h of HRT. In a larger anammox reactor (400 L), an average VNR of 4.84 kg-N/m(3)/day could be maintained for 86 days during the treatment of low strength synthetic inorganic wastewater. As a result of bacterial community analysis for the 20 L anammox reactor, Asahi BRW1, probably originating from the wastewater collected at Asahi Breweries, was detected as the dominant anammox bacterium. These anammox reactors were characterized by a high NH(4)-N removal capacity for low strength wastewater with a short hydraulic retention time.

  17. Gas-Liquid Mass Transfer in a Slurry Bubble Column Reactor under High Temperature andHigh Pressure

    Institute of Scientific and Technical Information of China (English)

    杨卫国; 王金福; 金涌

    2001-01-01

    The gas-liquid mass transfer of H2 and CO in a high temperature and high-pressure three-phase slurry bubble column reactor is studied. The gas-liquid volumetric mass transfer coefficients kLa are obtained by measuring the dissolution rate of H2 and CO. The influences of the main operation conditions, such as temperature, pressure,superficial gas velocity and solid concentration, are studied systematically. Two empirical correlations are proposed to predict kLa values for H2 and CO in liquid paraffln/solid particles slurry bubble column reactors.

  18. Fluoride Salt-Cooled High-Temperature Demonstration Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wysocki, Aaron J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-02-01

    The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include TRISO particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Several preconceptual and conceptual design efforts that have been conducted on FHR concepts bear a significant influence on the FHR DR design. Specific designs include the Oak Ridge National Laboratory (ORNL) advanced high-temperature reactor (AHTR) with 3400/1500 MWt/megawatts of electric output (MWe), as well as a 125 MWt small modular AHTR (SmAHTR) from ORNL. Other important examples are the Mk1 pebble bed FHR (PB-FHR) concept from the University of California, Berkeley (UCB), and an FHR test reactor design developed at the Massachusetts Institute of Technology (MIT). The MIT FHR test reactor is based on a prismatic fuel platform and is directly relevant to the present FHR DR design effort. These FHR concepts are based on reasonable assumptions for credible commercial prototypes. The FHR DR concept also directly benefits from the operating experience of the Molten Salt Reactor Experiment (MSRE), as well as the detailed design efforts for a large molten salt reactor concept and its breeder variant, the Molten Salt Breeder Reactor. The FHR DR technology is most representative of the 3400 MWt AHTR

  19. Evaluation of proposed German safety criteria for high-temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Barsell, A.W.

    1980-05-01

    This work reviews proposed safety criteria prepared by the German Bundesministerium des Innern (BMI) for future licensing of gas-cooled high-temperature reactor (HTR) concepts in the Federal Republic of Germany. Comparison is made with US General Design Criteria (GDCs) in 10CFR50 Appendix A and with German light water reactor (LWR) criteria. Implications for the HTR design relative to the US design and safety approach are indicated. Both inherent characteristics and design features of the steam cycle, gas turbine, and process heat concepts are taken into account as well as generic design options such as a pebble bed or prismatic core.

  20. Development and Testing of a High Capacity Plasma Chemical Reactor in the Ukraine

    Energy Technology Data Exchange (ETDEWEB)

    Reilly, Raymond W.

    2012-07-30

    This project, Development and Testing of a High Capacity Plasma Chemical Reactor in the Ukraine was established at the Kharkiv Institute of Physics and Technology (KIPT). The associated CRADA was established with Campbell Applied Physics (CAP) located in El Dorado Hills, California. This project extends an earlier project involving both CAP and KIPT conducted under a separate CRADA. The initial project developed the basic Plasma Chemical Reactor (PCR) for generation of ozone gas. This project built upon the technology developed in the first project, greatly enhancing the output of the PCR while also improving reliability and system control.

  1. Neutrino mass hierarchy and precision physics with medium-baseline reactors: impact of energy-scale and flux-shape uncertainties

    CERN Document Server

    Capozzi, F; Marrone, A

    2015-01-01

    Nuclear reactors provide intense sources of electron antineutrinos, characterized by few-MeV energy E and unoscillated spectral shape Phi(E). High-statistics observations of reactor neutrino oscillations over medium-baseline distances L ~ O(50) km would provide unprecedented opportunities to probe both the long-wavelength mass-mixing parameters (delta m^2 and theta_12) and the short-wavelength ones (Delta m^2 and theta_13), together with the subtle interference effects associated to the neutrino mass hierarchy (either normal or inverted). In a given experimental setting - here taken as in the JUNO project for definiteness - the achievable hierarchy sensitivity and parameter accuracy depend not only on the accumulated statistics but also on systematic uncertainties, which include (but are not limited to) the mass-mixing priors and the normalizations of signals and backgrounds. We examine, in addition, the effect of introducing smooth deformations of the detector energy scale, E -> E'(E), and of the reactor flu...

  2. Reactor Neutrinos

    Directory of Open Access Journals (Sweden)

    Soo-Bong Kim

    2013-01-01

    Full Text Available We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very recently the most precise determination of the neutrino mixing angle θ13. This paper provides an overview of the upcoming experiments and of the projects under development, including the determination of the neutrino mass hierarchy and the possible use of neutrinos for society, for nonproliferation of nuclear materials, and geophysics.

  3. Fluoride Salt-Cooled High-Temperature Reactor Technology Development and Demonstration Roadmap

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Flanagan, George F [ORNL; Mays, Gary T [ORNL; Pointer, William David [ORNL; Robb, Kevin R [ORNL; Yoder Jr, Graydon L [ORNL

    2013-11-01

    Fluoride salt-cooled High-temperature Reactors (FHRs) are an emerging reactor class with potentially advantageous performance characteristics, and fully passive safety. This roadmap describes the principal remaining FHR technology challenges and the development path needed to address the challenges. This roadmap also provides an integrated overview of the current status of the broad set of technologies necessary to design, evaluate, license, construct, operate, and maintain FHRs. First-generation FHRs will not require any technology breakthroughs, but do require significant concept development, system integration, and technology maturation. FHRs are currently entering early phase engineering development. As such, this roadmap is not as technically detailed or specific as would be the case for a more mature reactor class. The higher cost of fuel and coolant, the lack of an approved licensing framework, the lack of qualified, salt-compatible structural materials, and the potential for tritium release into the environment are the most obvious issues that remain to be resolved.

  4. High Temperature Fusion Reactor Cooling Using Brayton Cycle Based Partial Energy Conversion

    Science.gov (United States)

    Juhasz, Albert J.; Sawicki, Jerzy T.

    2003-01-01

    For some future space power systems using high temperature nuclear heat sources most of the output energy will be used in other than electrical form, and only a fraction of the total thermal energy generated will need to be converted to electrical work. The paper describes the conceptual design of such a partial energy conversion system, consisting of a high temperature fusion reactor operating in series with a high temperature radiator and in parallel with dual closed cycle gas turbine (CCGT) power systems, also referred to as closed Brayton cycle (CBC) systems, which are supplied with a fraction of the reactor thermal energy for conversion to electric power. Most of the fusion reactor's output is in the form of charged plasma which is expanded through a magnetic nozzle of the interplanetary propulsion system. Reactor heat energy is ducted to the high temperature series radiator utilizing the electric power generated to drive a helium gas circulation fan. In addition to discussing the thermodynamic aspects of the system design the authors include a brief overview of the gas turbine and fan rotor-dynamics and proposed bearing support technology along with performance characteristics of the three phase AC electric power generator and fan drive motor.

  5. Hydrogen/Oxygen Reactions at High Pressures and Intermediate Temperatures: Flow Reactor Experiments and Kinetic Modeling

    DEFF Research Database (Denmark)

    Hashemi, Hamid; Christensen, Jakob Munkholt; Glarborg, Peter

    A series of experimental and numerical investigations into hydrogen oxidation at high pressures and intermediate temperatures has been conducted. The experiments were carried out in a high pressure laminar flow reactor at 50 bar pressure and a temperature range of 600–900 K. The equivalence ratio......, ignition occurs at the temperature of 775–800 K. In general, the present model provides a good agreement with the measurements in the flow reactor and with recent data on laminar burning velocity and ignition delay time.......A series of experimental and numerical investigations into hydrogen oxidation at high pressures and intermediate temperatures has been conducted. The experiments were carried out in a high pressure laminar flow reactor at 50 bar pressure and a temperature range of 600–900 K. The equivalence ratio......, the mechanism is used to simulate published data on ignition delay time and laminar burning velocity of hydrogen. The flow reactor results show that at reducing, stoichiometric, and oxidizing conditions, conversion starts at temperatures of 750–775 K, 800–825 K, and 800–825 K, respectively. In oxygen atmosphere...

  6. Analytical Study of High Concentration PCB Paint at the Heavy Water Components Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lowry, N.J.

    1998-10-21

    This report provides results of an analytical study of high concentration PCB paint in a shutdown nuclear test reactor located at the US Department of Energy's Savannah River Site (SRS). The study was designed to obtain data relevant for an evaluation of potential hazards associated with the use of and exposure to such paints.

  7. Sustainability and Efficiency Improvements of Gas-Cooled High Temperature Reactors

    NARCIS (Netherlands)

    Marmier, A.

    2012-01-01

    The work presented in this thesis covers three fundamental aspects of High Temperature Reactor (HTR) performance, namely fuel testing under irradiation for maximized safety and sustainability, fuel architecture for improved economy and sustainability, and a novel Balance of Plant concept to enable f

  8. Practical superconductor development for electrical applications - Argonne National Laboratory quarterly report for the period ending September 30, 2002.

    Energy Technology Data Exchange (ETDEWEB)

    Dorris, S. E.

    2002-12-02

    This is a multiyear experimental research program that focuses on improving relevant material properties of high-T{sub c} superconductors (HTSs) and developing fabrication methods that can be transferred to industry for production of commercial conductors. The development of teaming relationships through agreements with industrial partners is a key element of the Argonne (ANL) program.

  9. Practical superconductor development for electrical power applications - Argonne National Laboratory - quarterly report for the period ending June 30, 2001.

    Energy Technology Data Exchange (ETDEWEB)

    Dorris, S. E.

    2001-08-21

    This is a multiyear experimental research program focused on improving relevant material properties of high-T{sub c} superconductors (HTSs) and on development of fabrication methods that can be transferred to industry for production of commercial conductors. The development of teaming relationships through agreements with industrial partners is a key element of the Argonne (ANL) program.

  10. Mathematical analysis for internal filtration of convection-enhanced high-flux hemodialyzer.

    Science.gov (United States)

    Lee, Jung Chan; Lee, Kyungsoo; Kim, Hee Chan

    2012-10-01

    Structural modifications using a conventional hemodialyzer improved the internal filtration and clearance of middle molecular weight wastes by enhanced convection effect. In this study, we employed a mathematical model describing the internal filtration rate as well as the hemodynamic and hematologic parameters in highflux dialyzer to interpret the previous reported experimental results. Conventional high-flux hemodialysis and convection-enhanced high-flux hemodialysis were configured in the mathematical forms and integrated into the iterative numerical method to predict the internal filtration phenomena inside the dialyzers during dialysis. The distributions of blood pressure, dialysate pressure, oncotic pressure, blood flow rates, dialysate flow rates, local ultrafiltration, hematocrit, protein concentration and blood viscosity along the axial length of dialyzer were calculated in order to estimate the internal filtration volume. The results show that the filtration volumes by internal filtration is two times higher in a convection-enhanced high-flux hemodialyzer than in a conventional high-flux hemodialzer and explains the experimental result of improved clearance of middle molecular size waste in convection-enhanced high-flux hemodialyzer.

  11. Methods to assess high-resolution subsurface gas concentrations and gas fluxes in wetland ecosystems

    DEFF Research Database (Denmark)

    Elberling, Bo; Kühl, Michael; Glud, Ronnie N.

    2013-01-01

    The need for measurements of soil gas concentrations and surface fluxes of greenhouse gases at high temporal and spatial resolution in wetland ecosystem has lead to the introduction of several new analytical techniques and methods. In addition to the automated flux chamber methodology for high......-resolution estimates of greenhouse gas fluxes across the soil-atmosphere interface, these high-resolution methods include microsensors for quantification of spatiotemporal concentration dynamics in O2 and N2O at micrometer scales, fiber-optic optodes for long-term continuous point measurements of O2 concentrations...... and peat soils are highly heterogeneous, containing a mosaic of dynamic macropore systems created by both macrofauna and flora leading to distinct spatial and temporal variations in gas concentration on a scale of millimeters and minutes. Applications of these new methodologies allow measurements...

  12. A new frontier in CO2 flux measurements using a highly portable DIAL laser system

    Science.gov (United States)

    Queiβer, Manuel; Granieri, Domenico; Burton, Mike

    2016-01-01

    Volcanic CO2 emissions play a key role in the geological carbon cycle, and monitoring of volcanic CO2 fluxes helps to forecast eruptions. The quantification of CO2 fluxes is challenging due to rapid dilution of magmatic CO2 in CO2-rich ambient air and the diffuse nature of many emissions, leading to large uncertainties in the global magmatic CO2 flux inventory. Here, we report measurements using a new DIAL laser remote sensing system for volcanic CO2 (CO2DIAL). Two sites in the volcanic zone of Campi Flegrei (Italy) were scanned, yielding CO2 path-amount profiles used to compute fluxes. Our results reveal a relatively high CO2 flux from Campi Flegrei, consistent with an increasing trend. Unlike previous methods, the CO2DIAL is able to measure integrated CO2 path-amounts at distances up to 2000 m using virtually any solid surface as a reflector, whilst also being highly portable. This opens a new frontier in quantification of geological and anthropogenic CO2 fluxes. PMID:27652775

  13. Status of Preconceptual Design of the Advanced High-Temperature Reactor (AHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Ingersoll, D.T.

    2004-07-29

    A new reactor plant concept is presented that combines the benefits of ceramic-coated, high-temperature particle fuel with those of clean, high-temperature, low-pressure molten salt coolant. The Advanced High-Temperature Reactor (AHTR) concept is a collaboration of Oak Ridge National Laboratory, Sandia National Laboratories, and the University of California at Berkeley. The purpose of the concept is to provide an advanced design capable of satisfying the top-level functional requirements of the U.S. Department of Energy Next Generation Nuclear Plant (NGNP), while also providing a technology base that is sufficiently robust to allow future development paths to higher temperatures and larger outputs with highly competitive economics. This report summarizes the status of the AHTR preconceptual design. It captures the results from an intense effort over a period of 3 months to (1) screen and examine potential feasibility concerns with the concept; (2) refine the conceptual design of major systems; and (3) identify research, development, and technology requirements to fully mature the AHTR design. Several analyses were performed and are presented to quantify the AHTR performance expectations and to assist in the selection of several design parameters. The AHTR, like other NGNP reactor concepts, uses coated particle fuel in a graphite matrix. But unlike the other NGNP concepts, the AHTR uses molten salt rather than helium as the primary system coolant. The considerable previous experience with molten salts in nuclear environments is discussed, and the status of high-temperature materials is reviewed. The large thermal inertia of the system, the excellent heat transfer and fission product retention characteristics of molten salt, and the low-pressure operation of the primary system provide significant safety attributes for the AHTR. Compared with helium coolant, a molten salt cooled reactor will have significantly lower fuel temperatures (150-200-C lower) for the

  14. Material Control and Accounting Design Considerations for High-Temperature Gas Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Trond Bjornard; John Hockert

    2011-08-01

    The subject of this report is domestic safeguards and security by design (2SBD) for high-temperature gas reactors, focusing on material control and accountability (MC&A). The motivation for the report is to provide 2SBD support to the Next Generation Nuclear Plant (NGNP) project, which was launched by Congress in 2005. This introductory section will provide some background on the NGNP project and an overview of the 2SBD concept. The remaining chapters focus specifically on design aspects of the candidate high-temperature gas reactors (HTGRs) relevant to MC&A, Nuclear Regulatory Commission (NRC) requirements, and proposed MC&A approaches for the two major HTGR reactor types: pebble bed and prismatic. Of the prismatic type, two candidates are under consideration: (1) GA's GT-MHR (Gas Turbine-Modular Helium Reactor), and (2) the Modular High-Temperature Reactor (M-HTR), a derivative of Areva's Antares reactor. The future of the pebble-bed modular reactor (PBMR) for NGNP is uncertain, as the PBMR consortium partners (Westinghouse, PBMR [Pty] and The Shaw Group) were unable to agree on the path forward for NGNP during 2010. However, during the technology assessment of the conceptual design phase (Phase 1) of the NGNP project, AREVA provided design information and technology assessment of their pebble bed fueled plant design called the HTR-Module concept. AREVA does not intend to pursue this design for NGNP, preferring instead a modular reactor based on the prismatic Antares concept. Since MC&A relevant design information is available for both pebble concepts, the pebble-bed HTGRs considered in this report are: (1) Westinghouse PBMR; and (2) AREVA HTR-Module. The DOE Office of Nuclear Energy (DOE-NE) sponsors the Fuel Cycle Research and Development program (FCR&D), which contains an element specifically focused on the domestic (or state) aspects of SBD. This Material Protection, Control and Accountancy Technology (MPACT) program supports the present work

  15. Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Per [Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering; Greenspan, Ehud [Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering

    2015-02-09

    This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designs are used, the power density of salt- cooled reactors is limited to 10 MW/m3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m3. This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X

  16. Meeting notes of the High Flux Isotope Reactor (HFIR) futures group

    Energy Technology Data Exchange (ETDEWEB)

    Houser, M.M. [comp.

    1995-08-01

    This report is a compilation of the notes from the ten meetings. The group charter is: (1) to identify and characterize the range of possibilities and necessities for keeping the HFIR operating for at least the next 15 years; (2) to identify and characterize the range of possibilities for enhancing the scientific and technical utility of the HFIR; (3) to evaluate the benefits or impacts of these possibilities on the various scientific fields that use the HFIR or its products; (4) to evaluate the benefits or impacts on the operation and maintenance of the HFIR facility and the regulatory requirements; (5) to estimate the costs, including operating costs, and the schedules, including downtime, for these various possibilities; and one possible impact of proposed changes may be to stimulate increased pressure for a reduced enrichment fuel for HFIR.

  17. Aging Assessment of an Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) Service Cable.

    Energy Technology Data Exchange (ETDEWEB)

    Bernstein, Robert; Celina, Mathias Christopher; Redline, Erica Marie; White, II, Gregory Von

    2014-07-01

    Nuclear energy is one industry where aging of safety-related materials and components is of great concern. Many U.S. nuclear power plants are approaching, or have already exceeded, 40 years of age. Analysis comparing the cost of new plant construction versus long-term operation under extended plant licensing through 60 years strongly favors the latter option. To ensure the safe, reliable, and cost-effective long-term operation of nuclear power plants, many systems, structures, and components must be evaluated. Furthermore, as new analytical techniques and testing approaches are developed, it is imperative that we also validate, and if necessary, improve upon the previously employed Institute of Electrical and Electronic Engineers (IEEE) qualification standards originally written in 1974. Fortunately, this daunting task has global support, particularly in light of the new social and political climate surrounding nuclear energy in a post-Fukushima era.

  18. A Level 1+ Probabilistic Safety Assessment of the High Flux Australian Reactor. Vol 3: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-01-01

    The third volume of the Probabilistic Safety Assessment contains supporting information for the PSA as follows: Appendix C (continued) with details of the system analysis and reports for the system/top event models; Appendix D with results of the specific engineering analyses of internal initiating events; Appendix E, containing supporting data for the human performance assessment,; Appendix F with details of the estimation of the frequency of leaks at HIFAR and Appendix G, containing event sequence model and quantification results

  19. Experimental facility for development of high-temperature reactor technology: instrumentation needs and challenges

    OpenAIRE

    Sabharwall Piyush; O’Brien James E.; Yoon SuJong; Sun Xiaodong

    2015-01-01

    A high-temperature, multi-fluid, multi-loop test facility is under development at the Idaho National Laboratory for support of thermal hydraulic materials, and system integration research for high-temperature reactors. The experimental facility includes a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed includ...

  20. Application of Gamma code coupled with turbomachinery models for high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh

    2008-02-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-ofcoolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of a toxic gas, CO, and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. GAMMA code is being developed to implement turbomachinery models in the power conversion unit (PCU) and ultimately models associated with the hydrogen plant. Some preliminary results will be described in this paper.

  1. Development of GAMMA Code and Evaluation for a Very High Temperature gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Chang H; Lim, H.S.; Kim, E.S.; NO, H.C.

    2007-06-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gasses (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. This paper will also include what improvements will be made in the Gamma code for the VHTR.

  2. High field side launch of RF waves: A new approach to reactor actuators

    Science.gov (United States)

    Wallace, G. M.; Baek, S. G.; Bonoli, P. T.; Faust, I. C.; LaBombard, B. L.; Lin, Y.; Mumgaard, R. T.; Parker, R. R.; Shiraiwa, S.; Vieira, R.; Whyte, D. G.; Wukitch, S. J.

    2015-12-01

    Launching radio frequency (RF) waves from the high field side (HFS) of a tokamak offers significant advantages over low field side (LFS) launch with respect to both wave physics and plasma material interactions (PMI). For lower hybrid (LH) waves, the higher magnetic field opens the window between wave accessibility (n∥≡c k∥/ω >√{1 -ωpi 2/ω2+ωpe 2/ωce 2 }+ωp e/|ωc e| ) and the condition for strong electron Landau damping (n∥˜√{30 /Te } with Te in keV), allowing LH waves from the HFS to penetrate into the core of a burning plasma, while waves launched from the LFS are restricted to the periphery of the plasma. The lower n∥ of waves absorbed at higher Te yields a higher current drive efficiency as well. In the ion cyclotron range of frequencies (ICRF), HFS launch allows for direct access to the mode conversion layer where mode converted waves absorb strongly on thermal electrons and ions, thus avoiding the generation of energetic minority ion tails. The absence of turbulent heat and particle fluxes on the HFS, particularly in double null configuration, makes it the ideal location to minimize PMI damage to the antenna structure. The quiescent SOL also eliminates the need to couple LH waves across a long distance to the separatrix, as the antenna can be located close to plasma without risking damage to the structure. Improved impurity screening on the HFS will help eliminate the long-standing issues of high Z impurity accumulation with ICRF. Looking toward a fusion reactor, the HFS is the only possible location for a plasma-facing RF antenna that will survive long-term. By integrating the antenna into the blanket module it is possible to improve the tritium breeding ratio compared with an antenna occupying an equatorial port plug. Blanket modules will require remote handling of numerous cooling pipes and electrical connections, and the addition of transmission lines will not substantially increase the level of complexity. The obvious engineering

  3. Concept of an inherently-safe high temperature gas-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ohashi, Hirofumi; Sato, Hiroyuki; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro [Nuclear Hydrogen and Heat Application Research Center, Japan Atomic Energy Agency, Oarai-machi, Ibaraki-ken, 311-1394 (Japan)

    2012-06-06

    As the challenge to ensure no harmful release of radioactive materials at th