WorldWideScience

Sample records for annular core research

  1. Facility modernization Annular Core Research Reactor

    International Nuclear Information System (INIS)

    The Annular Core Research Reactor (ACRR) has undergone numerous modifications since its conception in response to program needs. The original reactor fuel, which was special U-ZrH TRIGA fuel designed primarily for pulsing, has been replaced with a higher pulsing capacity BeO fuel. Other advanced operating modes which use this increased capability, in addition to the pulse and steady state, have been incorporated to tailor power histories and fluences to the experiments. Various experimental facilities have been developed that range from a radiography facility to a 50 cm diameter External Fuel Ring Cavity (FREC) using 180 of the original ZrH fuel elements. Currently a digital reactor console is being produced with GA, which will give enhanced monitoring capabilities of the reactor parameters while leaving the safety-related shutdown functions with analog technology. (author)

  2. Critical heat flux prediction for the annular core research reactor

    International Nuclear Information System (INIS)

    This paper reports on best estimate predictions of Critical Heat Flux Ratio (CHFR) obtained to support the upgrade of the Annular Core Research Reactor (ACRR) at Sandia National Laboratories for 2 to 4 MWt. The CHF productions are based on the University of New Mexico's (UNM)-CHF correlations in conjunction with the Global Conditions Hypothesis (GCH). Results indicate that for the range of inlet water temperature of 293 to 333 K, CHFR predictions range from 3.9 to 2.1, which is more than sufficient to support the proposed ACRR upgrade

  3. Characterization of Novel Calorimeters in the Annular Core Research Reactor *

    Directory of Open Access Journals (Sweden)

    Hehr Brian D.

    2016-01-01

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

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

    Science.gov (United States)

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

    2016-02-01

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

  5. MCNP/MCNPX model of the annular core research reactor.

    Energy Technology Data Exchange (ETDEWEB)

    DePriest, Kendall Russell; Cooper, Philip J.; Parma, Edward J., Jr. (.,; .)

    2006-10-01

    Many experimenters at the Annular Core Research Reactor (ACRR) have a need to predict the neutron/gamma environment prior to testing. In some cases, the neutron/gamma environment is needed to understand the test results after the completion of an experiment. In an effort to satisfy the needs of experimenters, a model of the ACRR was developed for use with the Monte Carlo N-Particle transport codes MCNP [Br03] and MCNPX [Wa02]. The model contains adjustable safety, transient, and control rods, several of the available spectrum-modifying cavity inserts, and placeholders for experiment packages. The ACRR model was constructed such that experiment package models can be easily placed in the reactor after being developed as stand-alone units. An addition to the 'standard' model allows the FREC-II cavity to be included in the calculations. This report presents the MCNP/MCNPX model of the ACRR. Comparisons are made between the model and the reactor for various configurations. Reactivity worth curves for the various reactor configurations are presented. Examples of reactivity worth calculations for a few experiment packages are presented along with the measured reactivity worth from the reactor test of the experiment packages. Finally, calculated neutron/gamma spectra are presented.

  6. A Preliminary Calculation of Annular Core Design for a High-flux Advanced Research Reactor

    International Nuclear Information System (INIS)

    Many of research reactors in operation over the world become old and the number of research reactors is expected to be reduced around 1/3 within a next decade. So it may be necessary to prepare in advance for the future demands of research reactors with a high performance. Therefore, based on the HANARO experiences through design to operation, a concept development of an improved research reactor is under doing. In this paper, 10 MW conceptual annular core is proposed and its basic characteristics were analyzed as a preliminary step

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

    OpenAIRE

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

    2016-01-01

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

  8. Design and fabrication of the instrumented fuel elements for the Annular Core Research Reactor (ACRR)

    International Nuclear Information System (INIS)

    This report describes the design and fabrication techniques for the instrumented fuel elements of the Annular Core Research Reactor (ACRR). The thermocouple assemblies were designed and fabricated at Sandia Laboratories while the instrumented elements were assembled at Los Alamos Scientific Laboratory. In order to satisfy the ACRR's Technical Specifications, the thermocouples are required to measure temperature in excess of 18000C under rapid heating conditions. Because of the potentially high failure rates for thermocouples in such environments, the instrumented fuel elements are designed so that the thermocouples can be replaced easily

  9. Safety analysis for operating the Annular Core Research Reactor with the central cavity liner removed

    International Nuclear Information System (INIS)

    Isotope production in the Annular Core Research Reactor requires highly enriched uranium targets to be irradiated in the high flux central region of the core. In order to accomplish this goal, the central cavity liner has been removed to allow for the eventual placement of targets in that region. This safety evaluation presents the analysis associated with operating the reactor in the steady state mode with the central cavity liner removed and the central region of the core filled with water and aluminum void targets. The reactor operation with enriched, uranium loaded targets will be analyzed in a future analysis document. This analysis describes only the operation of the reactor in the steady state mode; consideration of pulse mode operations with the liner removed is not presented

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

    Directory of Open Access Journals (Sweden)

    Kaiser Krista

    2016-01-01

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

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

    Science.gov (United States)

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

    2016-02-01

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

  12. ACRR [Annular Core Research Reactor] fission product release tests: ST-1 and ST-2

    International Nuclear Information System (INIS)

    Two experiments (ST-1 and ST-2) have been performed in the Annular Core Research Reactor (ACER) at Sandia National Laboratories (SNLA) to obtain time-resolved data on the release of fission products from irradiated fuels under light water reactor (LWR) severe accident conditions. Both experiments were conducted in a highly reducing environment at maximum fuel temperatures of greater than 2400 K. These experiments were designed specifically to investigate the effect of increased total pressure on fission product release; ST-1 was performed at approximately 0.16 MPa and ST-2 was run at 1.9 MPa, whereas other parameters were matched as closely as possible. Release rate data were measured for Cs, I, Ba, Sr, Eu, Te, and U. The release rates were higher than predicted by existing codes for Ba, Sr, Eu, and U. Te release was very low, but Te did not appear to be sequestered by the zircaloy cladding; it was evenly distributed in the fuel. In addition, in posttest analysis a unique fuel morphology (fuel swelling) was observed which may have enhanced fission product release, especially in the high pressure test (ST-2). These data are compared with analytical results from the CORSOR correlation and the VICTORIA computer model. 8 refs., 8 figs., 2 tabs

  13. Neutron Environment Characterization of the Central Cavity in the Annular Core Research Reactor *

    Directory of Open Access Journals (Sweden)

    Parma Edward J.

    2016-01-01

    Full Text Available Characterization of the neutron environment in the central cavity of the Sandia National Laboratories' Annular Core Research Reactor (ACRR is important in order to provide experimenters with the most accurate spectral information and maintain a high degree of fidelity in performing reactor experiments. Characterization includes both modeling and experimental efforts. Building accurate neutronic models of the ACRR and the central cavity “bucket” environments that can be used by experimenters is important in planning and designing experiments, as well as assessing the experimental results and quantifying uncertainties. Neutron fluence characterizations of two bucket environments, LB44 and PLG, are presented. These two environments are used frequently and represent two extremes in the neutron spectrum. The LB44 bucket is designed to remove the thermal component of the neutron spectrum and significantly attenuate the gamma-ray fluence. The PLG bucket is designed to enhance the thermal component of the neutron spectrum and attenuate the gamma-ray fluence. The neutron characterization for each bucket was performed by irradiating 20 different activation foil types, some of which were cadmium covered, resulting in 37 different reactions at the peak axial flux location in each bucket. The dosimetry results were used in the LSL-M2 spectrum adjustment code with a 640-energy group MCNP-generated trial spectrum, self-shielding correction factors, the SNLRML or IRDFF dosimetry cross-section library, trial spectrum uncertainty, and trial covariance matrix, to generate a least-squares adjusted neutron spectrum, spectrum uncertainty, and covariance matrix. Both environment character-izations are well documented and the environments are available for use by experimenters.

  14. Neutron Environment Characterization of the Central Cavity in the Annular Core Research Reactor

    Science.gov (United States)

    Parma, Edward J.; Naranjo, Gerald E.; Lippert, Lance L.; Vehar, David W.

    2016-02-01

    Characterization of the neutron environment in the central cavity of the Sandia National Laboratories' Annular Core Research Reactor (ACRR) is important in order to provide experimenters with the most accurate spectral information and maintain a high degree of fidelity in performing reactor experiments. Characterization includes both modeling and experimental efforts. Building accurate neutronic models of the ACRR and the central cavity "bucket" environments that can be used by experimenters is important in planning and designing experiments, as well as assessing the experimental results and quantifying uncertainties. Neutron fluence characterizations of two bucket environments, LB44 and PLG, are presented. These two environments are used frequently and represent two extremes in the neutron spectrum. The LB44 bucket is designed to remove the thermal component of the neutron spectrum and significantly attenuate the gamma-ray fluence. The PLG bucket is designed to enhance the thermal component of the neutron spectrum and attenuate the gamma-ray fluence. The neutron characterization for each bucket was performed by irradiating 20 different activation foil types, some of which were cadmium covered, resulting in 37 different reactions at the peak axial flux location in each bucket. The dosimetry results were used in the LSL-M2 spectrum adjustment code with a 640-energy group MCNP-generated trial spectrum, self-shielding correction factors, the SNLRML or IRDFF dosimetry cross-section library, trial spectrum uncertainty, and trial covariance matrix, to generate a least-squares adjusted neutron spectrum, spectrum uncertainty, and covariance matrix. Both environment character-izations are well documented and the environments are available for use by experimenters. Work supported by the United States Department of Energy at Sandia National Laboratories. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned

  15. Safety analysis for operating the Annular Core Research Reactor with Cintichem-type targets installed in the central region of the core

    International Nuclear Information System (INIS)

    Production of the molybdenum-99 isotope at the Annular Core Research Reactor requires highly enriched, uranium oxide loaded targets to be irradiated for several days in the high neutron-flux region of the core. This report presents the safety analysis for the irradiation of up to seven Cintichem-type targets in the central region of the core and compares the results to the Annular Core Research Reactor Safety Analysis Report. A 19 target grid configuration is presented that allows one to seven targets to be irradiated, with the remainder of the grid locations filled with aluminum ''void'' targets. Analyses of reactor, neutronic, thermal hydraulics, and heat transfer calculations are presented. Steady-state operation and accident scenarios are analyzed with the conclusion that the reactor can be operated safely with seven targets in the grid, and no additional risk to the public

  16. Safety analysis for operating the Annular Core Research Reactor with Cintichem-type targets installed in the central region of the core

    Energy Technology Data Exchange (ETDEWEB)

    PARMA JR.,EDWARD J.

    2000-01-01

    Production of the molybdenum-99 isotope at the Annular Core Research Reactor requires highly enriched, uranium oxide loaded targets to be irradiated for several days in the high neutron-flux region of the core. This report presents the safety analysis for the irradiation of up to seven Cintichem-type targets in the central region of the core and compares the results to the Annular Core Research Reactor Safety Analysis Report. A 19 target grid configuration is presented that allows one to seven targets to be irradiated, with the remainder of the grid locations filled with aluminum ''void'' targets. Analyses of reactor, neutronic, thermal hydraulics, and heat transfer calculations are presented. Steady-state operation and accident scenarios are analyzed with the conclusion that the reactor can be operated safely with seven targets in the grid, and no additional risk to the public.

  17. Critical heat flux predictions for the Sandia Annular Core Research Reactor

    International Nuclear Information System (INIS)

    This study provides best estimate predictions of the Critical Heat Flux (CHF) and the Critical Heat Flux Ratio (CHFR) to support the proposed upgrade of the Annual Core Research Reactor (ACRR) at Sandia National Laboratories (SNL) from its present value of 2 MWt to 4 MWt. These predictions are based on the University of New Mexico (UNM) - CHF correlation, originally developed for uniformly heated vertical annuli. The UNM-CHF correlation is applicable to low-flow and low-pressure conditions, which are typical of those in the ACRR. The three hypotheses that examined the effect of the nonuniform axial heat flux distribution in the ACRR core are (1) the local conditions hypotheses, (2) the total power hypothesis, and (3) the global conditions hypothesis. These hypotheses, in conjunction with the UNM-CHF correlation, are used to estimate the CHF and CHFR in the ACRR. Because the total power hypothesis predictions of power per rod at CHF are approximately 15%-20% lower than those corresponding to saturation exit conditions, it can be concluded that the total power hypothesis considerably underestimates the CHF for nonuniformly heated geometries. This conclusion is in agreement with previous experimental results. The global conditions hypothesis, which is more conservative and more accurate of the other two, provides the most reliable predictions of CHF/CHFR for the ACRR. The global conditions hypothesis predictions of CHFR varied between 2.1 and 3.9, with the higher value corresponding to the lower water inlet temperature of 20 degrees C

  18. Lessons Learned from Sandia National Laboratories' Operational Readiness Review of the Annular Core Research Reactor (ACRR)

    International Nuclear Information System (INIS)

    The Sandia ACRR (a Hazard Category 2 Nuclear Reactor Facility) was defueled in June 1997 to modify the reactor core and control system to produce medical radioisotopes for the Department of Energy (DOE) Isotope Production Program. The DOE determined that an Operational Readiness Review (ORR) was required to confirm readiness to begin operations within the revised safety basis. This paper addresses the ORR Process, lessons learned from the Sandia and DOE ORRS of the ACRR, and the use of the ORR to confirm authorization basis implementation

  19. Review of Transient Fuel Test Results at Sandia National Laboratories and the Potential for Future Fast Reactor Fuel Transient Testing in the Annular Core Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven A.; Pickard, Paul S.; Parma, Edward J.; Vernon, Milton E.; Kelly, John; Tikare, Veena [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

    2009-06-15

    Reactor driven transient tests of fast reactor fuels may be required to support the development and certification of new fuels for Fast Reactors. The results of the transient fuel tests will likely be needed to support licensing and to provide validation data to support the safety case for a variety of proposed fast fuel types and reactors. In general reactor driven transient tests are used to identify basic phenomenology during reactor transients and to determine the fuel performance limits and margins to failure during design basis accidents such as loss of flow, loss of heat sink, and reactivity insertion accidents. This paper provides a summary description of the previous Sandia Fuel Disruption and Transient Axial Relocation tests that were performed in the Annular Core Research Reactor (ACRR) for the U.S. Nuclear Regulatory Commission almost 25 years ago. These tests consisted of a number of capsule tests and flowing gas tests that used fission heating to disrupt fresh and irradiated MOX fuel. The behavior of the fuel disruption, the generation of aerosols and the melting and relocation of fuel and cladding was recorded on high speed cinematography. This paper will present videos of the fuel disruption that was observed in these tests which reveal stark differences in fuel behavior between fresh and irradiated fuel. Even though these tests were performed over 25 years ago, their results are still relevant to today's reactor designs. These types of transient tests are again being considered by the Advanced Fuel Cycle Initiative to support the Global Nuclear Energy Partnership because of the need to perform tests on metal fuels and transuranic fuels. Because the Annular Core Research Reactor is the only transient test facility available within the US, a brief summary of Sandia's continued capability to perform these tests in the ACRR will also be provided. (authors)

  20. Core-annular flow through a horizontal pipe: Hydrodynamic counterbalancing of buoyancy force on core

    NARCIS (Netherlands)

    Ooms, G.; Vuik, C.; Poesio, P.

    2007-01-01

    A theoretical investigation has been made of core-annular flow: the flow of a high-viscosity liquid core surrounded by a low-viscosity liquid annular layer through a horizontal pipe. Special attention is paid to the question of how the buoyancy force on the core, caused by a density difference betwe

  1. Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

    Science.gov (United States)

    Peterson, Per F.

    2013-05-14

    A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

  2. Criticality Benchmark Analysis of the HTTR Annular Startup Core Configurations

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2009-11-01

    One of the high priority benchmarking activities for corroborating the Next Generation Nuclear Plant (NGNP) Project and Very High Temperature Reactor (VHTR) Program is evaluation of Japan's existing High Temperature Engineering Test Reactor (HTTR). The HTTR is a 30 MWt engineering test reactor utilizing graphite moderation, helium coolant, and prismatic TRISO fuel. A large amount of critical reactor physics data is available for validation efforts of High Temperature Gas-cooled Reactors (HTGRs). Previous international reactor physics benchmarking activities provided a collation of mixed results that inaccurately predicted actual experimental performance.1 Reevaluations were performed by the Japanese to reduce the discrepancy between actual and computationally-determined critical configurations.2-3 Current efforts at the Idaho National Laboratory (INL) involve development of reactor physics benchmark models in conjunction with the International Reactor Physics Experiment Evaluation Project (IRPhEP) for use with verification and validation methods in the VHTR Program. Annular cores demonstrate inherent safety characteristics that are of interest in developing future HTGRs.

  3. Criticality Benchmark Analysis of the HTTR Annular Startup Core Configurations

    International Nuclear Information System (INIS)

    One of the high priority benchmarking activities for corroborating the Next Generation Nuclear Plant (NGNP) Project and Very High Temperature Reactor (VHTR) Program is evaluation of Japan's existing High Temperature Engineering Test Reactor (HTTR). The HTTR is a 30 MWt engineering test reactor utilizing graphite moderation, helium coolant, and prismatic TRISO fuel. A large amount of critical reactor physics data is available for validation efforts of High Temperature Gas-cooled Reactors (HTGRs). Previous international reactor physics benchmarking activities provided a collation of mixed results that inaccurately predicted actual experimental performance.1 Reevaluations were performed by the Japanese to reduce the discrepancy between actual and computationally-determined critical configurations.2-3 Current efforts at the Idaho National Laboratory (INL) involve development of reactor physics benchmark models in conjunction with the International Reactor Physics Experiment Evaluation Project (IRPhEP) for use with verification and validation methods in the VHTR Program. Annular cores demonstrate inherent safety characteristics that are of interest in developing future HTGRs.

  4. Annular core for modular high temperature gas-cooled reactor (MHTGR)

    International Nuclear Information System (INIS)

    The active core of the 350 MW(t) MHTGR is annular in configuration, shaped to provide a large external surface-to-volume ratio for the transport of heat radially to the reactor vessel in case of a loss of coolant flow. For a given fuel temperature limit, the annular core provides approximately 40 % greater power output over a typical cylindrical configuration. The reactor core is made up of columns of hexagonal blocks, each 793-mm high and 360-mm wide. The active core is 3.5 m in o.d., 1.65 m in i.d., and 7.93 m tall. Fuel elements contain TRISO-coated microspheres of 19.8 % enriched uranium oxycarbide and of fertile thorium oxide. The core is controlled by 30 control rods which enter the inner and outer side reflectors from above. (author)

  5. Experimental study on large diameter drilling in hard rock annular coring

    Institute of Scientific and Technical Information of China (English)

    Yinzhu WU; Guochun YANG; Wenchen WANG

    2008-01-01

    Based on analyzing method of large diameter hard rock drilling at home and abroad, the authors proposed a set of drilling of large diameter hard rock annular coring in low energy consumption, low cost and high efficiency. The prototype of drilling tools was designed and was made. The experimental result of the prototype indicates that this plan and technology are feasible and reach the anticipated object of design. A set of drilling tools has been offered for the constructs of large diameter hard rock coring.

  6. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    International Nuclear Information System (INIS)

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  7. Dynamic Response Control of Three-Layered Annular Plate Due to Various Parametres of Electrorheological Core

    Directory of Open Access Journals (Sweden)

    Pawlus Dorota

    2016-03-01

    Full Text Available The paper presents dynamic responses of annular plate composed of three layers. The middle layer of the plate has electrorheological properties expressed by the Bingham body model. The plate is loaded in the plane of facings with time-dependent forces. The electrorheological effect is observed in the area of supercritical plate behaviour. The influence of both material properties and geometrical dimensions of the core on plate behaviour is examined. The problem is solved analytically and numerically using the orthogonalization method and the finite difference method. Comparison of the results obtained using the finite difference and the finite element methods for a plate in critical state is shown. The numerical calculations are carried out for axisymmetric and asymmetric plate modes. The presented diagrams show the plate reaction to the changes in values of plate parameters and indicate that the supercritical control of plate work is possible.

  8. KNK II third core: design report for the annular fuel elements on the central position to accommodate material test inserts NZ 402 and NZ 403

    International Nuclear Information System (INIS)

    Since August 1984 irradiation experiments with temperature controlled pressure tube probes are being performed in the central position of KNK II. This is part of a long-term experimental program for the development of irradiation resistant reactor materials, which shall also be continued in the third core. The necessary irradiation channel is provided by a special annular fuel element. The present report describes the annular fuel elements for the third core. Aspects of the subassembly design are considered on the basis of the annular element design for the second core and the standard elements of the third core. Two annular elements NZ 402 and NZ 403 (as reserve) are available. It is demonstrated that the expected loadings will allow an unperturbed operation of the annular elements on the central position of the third core

  9. Annular seed-blanket thorium fuel core concepts for heavy water moderated reactors

    International Nuclear Information System (INIS)

    New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen is a 35-element bundle made with a homogeneous mixture of reactor grade Pu and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several annular heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that the various core concepts can achieve a fissile utilization that is up to 30% higher than is currently achieved in a PT-HWR using conventional natural uranium fuel bundles. Up to 67% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 363 kg/year of U-233 is produced. Seed-blanket cores with ∼50% content of low-power blanket bundles may require power de-rating (∼58% to 65%) to avoid exceeding maximum limits for peak channel power, bundle power and linear element ratings. (authors)

  10. Conceptual design study of Pebble Bed Type High Temperature Gas-cooled Reactor with annular core structure

    International Nuclear Information System (INIS)

    This report presents the Conceptual Design Study of Pebble Bed Type High Temperature Gas-cooled Reactor with Annular Core Structure. From this study, it is made clear that the thermal power of the Pebble Bed Type Reactor can be increased to 500MW through introducing the annular core structure without losing the inherent safe characteristics (in the coolant depressurization accident, the fuel temperature does not exceed the temperature where the fuel defect begins.) This thermal power is two times higher than the inherent safe Pebble Bed Type High temperature Gas-cooled Reactor (MHTGR) designed in West Germany. From this result, it is foreseen that the ratio of the plant cost to the reactor power is reduced and the economy of the plant operation is improved. The reactor performances e.g. fuel burnup and fuel temperature are maintained in same level of the MHTGR. (author)

  11. Experimental research on dryout point of flow boiling in narrow annular channels

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    An experimental research on the dryout point of flow boiling in narrow annular channels under low mass flux with 1.55 mm and 1.05 mm annular gap, respectively, is conducted. Distilled water is used as working fluid and the range of pressure is limited within 2.0~4.0 MPa and that of mass flux is 26.0~69.0 kg·m-2·s-1. The relation of critical heat flux (CHF) and critical qualities with mass flux and pressure are revealed. It is found that the critical qualities decrease with the increasing mass flux and increase with the increasing inlet qualities in externally heated annuli.Under the same conditions, critical qualities in the outer tube are always larger than those in the inner tube. The appearance of dryout point in bilaterally heated narrow annuli can be judged according to the ratio of qo/qi.

  12. Laser anemometer measurements in an annular cascade of core turbine vanes and comparison with theory

    Science.gov (United States)

    Goldman, L. J.; Seashultz, R. G.

    1982-01-01

    Laser measurements were made in an annular cascade of stator vanes operating at an exit critical velocity ratio of 0.78. Velocity and flow angles in the blade to blade plane were obtained at every 10 percent of axial chord within the passage and at 1/2 axial chord downstream of the vanes for radial positions near the hub, mean and tip. Results are presented in both plot and tabulated form and are compared with calculations from an inviscid, quasi three dimensional computer program. The experimental measurements generally agreed well with these theoretical calculations, an indication of the usefulness of this analytic approach.

  13. Nonlinear stability of oscillatory core-annular flow: A generalized Kuramoto-Sivashinsky equation with time periodic coefficients

    Science.gov (United States)

    Coward, Adrian V.; Papageorgiou, Demetrios T.; Smyrlis, Yiorgos S.

    1994-01-01

    In this paper the nonlinear stability of two-phase core-annular flow in a pipe is examined when the acting pressure gradient is modulated by time harmonic oscillations and viscosity stratification and interfacial tension is present. An exact solution of the Navier-Stokes equations is used as the background state to develop an asymptotic theory valid for thin annular layers, which leads to a novel nonlinear evolution describing the spatio-temporal evolution of the interface. The evolution equation is an extension of the equation found for constant pressure gradients and generalizes the Kuramoto-Sivashinsky equation with dispersive effects found by Papageorgiou, Maldarelli & Rumschitzki, Phys. Fluids A 2(3), 1990, pp. 340-352, to a similar system with time periodic coefficients. The distinct regimes of slow and moderate flow are considered and the corresponding evolution is derived. Certain solutions are described analytically in the neighborhood of the first bifurcation point by use of multiple scales asymptotics. Extensive numerical experiments, using dynamical systems ideas, are carried out in order to evaluate the effect of the oscillatory pressure gradient on the solutions in the presence of a constant pressure gradient.

  14. Stability of core-annular flow of power-law fluids in the presence of interfacial surfactant

    Institute of Scientific and Technical Information of China (English)

    2010-01-01

    The shear-thinning influence on the core-annular flow stability of two immiscible power-law fluids is considered by making a linear stability analysis.The flow is driven by an axial pressure gradient in a straight pipe with the interface between the two fluids occupied by an insoluble surfactant.Given the basic flow for this core-annular arrangement,the analytical solution is obtained with respect to the power-law fluid model.The linearized equations for the evolution of infinitesimal disturbances are derived and the stability problem is formulated as a generalized matrix eigenvalue problem,which is solved by using the software package Matlab based on the QZ algorithm.The shear-thinning property is found to have marked influence on the power-law fluid core-annular flow stability,which is reflected in various aspects.First,the capillary instability is magnified by the shear-thinning property,which may lead to an essential difference between power-law and Newtonian fluid flows.Especially when the interface is close to the pipe wall,the power-law fluid flow may be unstable while the Newtonian fluid flow is stable.Second,under disturbances to the interface a velocity discontinuity at the interface appears which is destabilizing to the flow.The magnitude of this velocity discontinuity is affected by the power-law index and the flow stability is influenced correspondingly.Besides,the shear-thinning property may induce new stability modes which do not appear in the Newtonian fluid flow.The flow stability shows much dependence on the interface location,the role of which was neglected in most previous studies.The shear-thinning fluid flow is more unstable to long wave disturbances when the interface is close to the pipe wall,while the Newtonian fluid flow is more unstable when the interface is close to the pipe centerline.But this trend is changed by the addition of interfacial surfactant,for which the power-law fluid flow is more stable no matter where the interface is

  15. Physics and behaviour during a ULOF of an innovative heterogeneous annular FBR core

    International Nuclear Information System (INIS)

    The major conclusions: • The reduction of the Na void worth is a way allowing a strong improvement of the dynamic behavior in very severe ULOF transient (10 s halving time), possibly allowing to avoid Na boiling; • 1st order effects: Na density ( 0); • 2nd order effects: - Mass flow gaggling scheme (as a function of the core neutronics); - Other feed-back effects: diagrid, driveline feed-back. → Very strong impact of uncertainties: Thermalhydraulic models & codes, drive-line feed-back modeling; • Methodology for feed-back coefficient calculation (example: in this calculation the Na density effect is linearized from nominal to 100% void, anticonservative in case of no Na boiling); • Core neutronics: nuclear data, models. → Even in case of no Na boiling, the critical events will be: • Fuel cladding and S/A wrapper behavior at very high temperature; • Upper core structures behavior

  16. Annular pancreas

    Science.gov (United States)

    ... page: //medlineplus.gov/ency/article/001142.htm Annular pancreas To use the sharing features on this page, please enable JavaScript. An annular pancreas is a ring of pancreatic tissue that encircles ...

  17. USGS Core Research Center (CRC) Collection of Core

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Core Research Center (CRC) was established in 1974 by the U.S. Geological Survey (USGS) to preserve valuable rock cores for use by scientists and educators from...

  18. The axisymmetric long-wave interfacial stability of core-annular flow of power-law fluid with surfactant

    Science.gov (United States)

    Sun, Xue-Wei; Peng, Jie; Zhu, Ke-Qin

    2012-02-01

    The long wave stability of core-annular flow of power-law fluids with an axial pressure gradient is investigated at low Reynolds number. The interface between the two fluids is populated with an insoluble surfactant. The analytic solution for the growth rate of perturbation is obtained with long wave approximation. We are mainly concerned with the effects of shear-thinning/thickening property and interfacial surfactant on the flow stability. The results show that the influence of shear-thinning/thickening property accounts to the change of the capillary number. For a clean interface, the shear-thinning property enhances the capillary instability when the interface is close to the pipe wall. The converse is true when the interface is close to the pipe centerline. For shear-thickening fluids, the situation is reversed. When the interface is close to the pipe centerline, the capillary instability can be restrained due to the influence of surfactant. A parameter set can be found under which the flow is linearly stable.

  19. Reactivity initiated accident (RIA) type tests and annular core pulse reactor (ACPR) operational experience

    International Nuclear Information System (INIS)

    This paper describes the test conducted to investigate the failure threshold of the fuel when subject to RIA, accomplished in the TRIGA ACPR Nuclear Research Institute, Pitesti. The reactor facility, the capsule used in experiments and the experimental results are presented. The failure threshold was determined at 200 cal/g for an atmospheric gap pressure comparable with similar tests. The failure threshold decreases with increasing gap pressure. The tests proved useful for a better understanding of the fuel behavior in the transient conditions. As it is known RIA is not a common accident for the CANDU reactors, but the fuel failure mechanism can be similar to other type of accidents as LOCA and PCM. The program will be continued, with better instrumentation for the fuel sample and also independent instrumentation to measure pulse characteristics with better statistics. A new project for the experimental fuel elements must be considered to eliminate fuel-endcap interactions. (author)

  20. A research on the mechanisms of transition from annular flow in two-phase pipeline flow

    International Nuclear Information System (INIS)

    Various kinds mechanisms of transitions from two-phase annular flow in tubes were studied and modelled, and the affection factors on the transitions were also discussed. Some mathematical equations and transition criteria for every mechanisms presented were derived, and an unified general criterion for the annular flow transitions in whole range of pipe inclinations was recommended. The boundaries predicted show good agreement with the air-water two-phase experimental data

  1. Effect of moderator density distribution of annular flow on fuel assembly neutronic characteristics in boiling water reactor cores

    International Nuclear Information System (INIS)

    The effect of the moderator density distribution of annular flow on the fuel assembly neutronic characteristics in a boiling water nuclear reactor was investigated using the SRAC95 code system. For the investigation, a model of annular flow for fuel assembly calculation was utilized. The results of the assembly calculation with the model (Method 1) and those of the fuel assembly calculation with the uniform void fraction distribution (Method 2) were compared. It was found that Method 2 underestimates the infinite multiplication factor in the fuel assembly including the gadolinia rod (type 1 assembly). This phenomenon is explained by the fact that the capture rate in the thermal energy region in gadolinia fuel is estimated to be smaller when the liquid film of annular flow at the fuel rod surface is considered. A burnup calculation was performed under the condition of a void fraction of 65% and a volumetric fraction of the liquid film in liquid phase of 1. It is found that Method 2 underestimates the infinite multiplication factor in comparison to Method 1 in the early stage of burnup, and that Method 2 becomes to overestimate the factor after a certain degree of burnup. This is because Method 2 overestimates the depletion rate of the gadolinia. (author)

  2. Seismic research on graphite reactor core

    International Nuclear Information System (INIS)

    Background: Reactors with graphite core structure include production reactor, water-cooled graphite reactor, gas-cooled reactor, high-temperature gas-cooled reactor and so on. Multi-body graphite core structure has nonlinear response under seismic excitation, which is different from the response of general civil structure, metal connection structure or bolted structure. Purpose: In order to provide references for the designing and construction of HTR-PM. This paper reviews the history of reactor seismic research evaluation from certain countries, and summarizes the research methods and research results. Methods: By comparing the methods adopted in different gas-cooled reactor cores, inspiration for our own HTR seismic research was achieved. Results and Conclusions: In this paper, the research ideas of graphite core seismic during the process of designing, constructing and operating HTR-10 are expounded. Also the project progress of HTR-PM and the research on side reflection with the theory of similarity is introduced. (authors)

  3. Research on plasma core reactors

    Science.gov (United States)

    Jarvis, G. A.; Barton, D. M.; Helmick, H. H.; Bernard, W.; White, R. H.

    1976-01-01

    Experiments and theoretical studies are being conducted for NASA on critical assemblies with one-meter diameter by one-meter long low-density cores surrounded by a thick beryllium reflector. These assemblies make extensive use of existing nuclear propulsion reactor components, facilities, and instrumentation. Due to excessive porosity in the reflector, the initial critical mass was 19 kg U(93.2). Addition of a 17 cm thick by 89 cm diameter beryllium flux trap in the cavity reduced the critical mass to 7 kg when all the uranium was in the zone just outside the flux trap. A mockup aluminum UF6 container was placed inside the flux trap and fueled with uranium-graphite elements. Fission distributions and reactivity worths of fuel and structural materials were measured. Finally, an 85,000 cu cm aluminum canister in the central region was fueled with UF6 gas and fission density distributions determined. These results are to be used to guide the design of a prototype plasma core reactor which will test energy removal by optical radiation.

  4. Behavior of an heterogeneous annular FBR core during an unprotected loss of flow accident: Analysis of the primary phase with SAS-SFR

    International Nuclear Information System (INIS)

    In the framework of a substantial improvement on FBR core safety connected to the development of a new Gen IV reactor type, heterogeneous core with innovative features are being carefully analyzed in France since 2009. At EDF R and D, the main goal is to understand whether a strong reduction of the Na-void worth - possibly attempting a negative value - allows a significant improvement of the core behavior during an unprotected loss of flow accident. Also, the physical behavior of such a core is of interest, before and beyond the (possible) onset of Na boiling. Hence, a cutting-edge heterogeneous design, featuring an annular shape, a Na-plena with a B4C plate and a stepwise modulation of fissile core heights, was developed at EDF by means of the SDDS methodology, with a total Na-void worth of -1 $. The behavior of such a core during the primary phase of a severe accident, initiated by an unprotected loss of flow, is analyzed by means of the SAS-SFR code. This study is carried-out at KIT and EDF, in the framework of a scientific collaboration on innovative FBR severe accident analyses. The results show that the reduction of the Na-void worth is very effective, but is not sufficient alone to avoid Na-boiling and, hence, to prevent the core from entering into the primary phase of a severe accident. Nevertheless, the grace time up to boiling onset is greatly enhanced in comparison to a more traditional homogeneous core design, and only an extremely low fraction of the fuel (<0.1%) enters into melting at the end of this phase. A sensitivity analysis shows that, due to the inherent neutronic characteristics of such a core, the gagging scheme plays a major role on the core behavior: indeed, an improved 4-zones gagging scheme, associated with an enhanced control rod drive line expansion feed-back effect, finally prevents the core from entering into sodium boiling. This major conclusion highlights both the progress already accomplished and the need for more detailed future

  5. Dual-core TRIGA research and materials testing reactor

    International Nuclear Information System (INIS)

    General Atomic Company is under contract from the Romanian Institute for Nuclear Technologies to design, fabricate, and install a research reactor in support of the Romanian National Program for Power Reactor Development. The goal was to select a design concept that provided reasonably high neutron fluxes for long term testing of various fuel-cladding-coolant combinations and also provide high performance pulsing for transient testing of fuel specimens. An effective solution was achieved by the selection of a 14 MW steady-state TRIGA reactor for high flux endurance testing, and an Annular Core Pulsing Reactor (ACPR) for high performance pulsing testing, with both reactors mounted in the same reactor tank and operated independently. The fuel bundles for the steady-state reactor consist of 25 uranium-zirconium hydride rods clad in stainless steel arranged in a square 5 x 5 array. The steady-state core is provided with downflow cooling at a rate of approximately 275 gpm/bundle. Bundle flow tests will be performed with both heated and unheated models. The core will be optimized for peak thermal neutron flux and reactivity lifetime within the constraint of a peak fuel meat temperature of 7500C. The operation of the steady-state reactor at a power level of 14 MW will yield peak unperturbed thermal neutron fluxes in the central experiment position of 2.9 x 1014 n/cm2-sec. The corresponding fast neutron flux (less than 1.125 keV) will be 2.6 x 1014 nv. (U.S.)

  6. Granuloma annulare.

    Science.gov (United States)

    Gupta, Diptesh; Hess, Brian; Bachegowda, Lohith

    2010-01-01

    We present a case of a 77-year-old, diabetic male with a 20-year history of a migratory erythematous, asymptomatic, generalized, nonscaly, and nonitchy rash that started over the dorsum of his left hand. On examination, there were multiple annular erythematous plaques, distributed symmetrically and diffusely over his torso and arms, with central clearing and no scales. A punch biopsy of the skin helped us to arrive at the diagnosis of a generalized granuloma annulare (GA). GA is a benign, self-limiting skin condition of unknown etiology that is often asymptomatic. The cause of this condition is unknown, but it has been associated with diabetes mellitus, infections such as HIV, and malignancies such as lymphoma. These lesions typically start as a ring of flesh-colored papules that slowly progress with central clearing. Lack of symptoms, scaling, or associated vesicles helps to differentiate GA from other skin conditions such as tinea corporis, pityriasis rosea, psoriasis, or erythema annulare centrifugum. Treatment is often not needed as the majority of these lesions are self-resolving within 2 years. Treatment may be pursued for cosmetic reasons. Available options include high-dose steroid creams, PUVA, cryotherapy, or drugs such as niacinamide, infliximab, Dapsone, and topical calcineurin inhibitors. PMID:20209383

  7. Behaviour of steel pipe exposed to fouling by heavy oil during core-annular flow; Comportamento de tubo de aco exposto a sujeira de oleo pesado durante escoamento nucleo-anular

    Energy Technology Data Exchange (ETDEWEB)

    Barbosa, Adriana; Bannwart, Antonio C. [Universidade Estadual de Campinas, SP (Brazil). Faculdade de Engenharia Mecanica. Dept. de Engenharia de Petroleo

    2004-07-01

    The use of water-assisted technologies such as core-annular flow to the pipelines of viscous oils has been proposed as an attractive alternative for production and transportation of heavy crudes in both onshore and offshore scenarios. Usually, core-annular flow can be created by injecting a relatively small water flow rate laterally in the pipe, so as to form a thin water annulus surrounding the viscous oil, which is pumped through the center. The reduction in friction losses obtained thanks to lubrication by water is significant, since the pressure drop in a steady state core flow becomes comparable to water flow only. For a complete assessment of core flow technology, however, unwanted effects associated with possible oil adhesion onto the pipe wall should be investigated, since these may cause severe fouling of the wall and pressure drop increase. It has been observed that oil adhesion on metallic surfaces may occur for certain types of crude and oilphilic pipe materials. In this work we present results of pressure drop monitoring during 35 hour-operation of a heavy oil-water core annular flow in a 26.08 mm. i.d. horizontal steel pipe. The oil used is described in terms of its main components and the results of static wet ability tests are also presented for comparison (author)

  8. Shut-down margin study for the next generation VVER-1000 reactor including 13 x 13 hexagonal annular assemblies

    International Nuclear Information System (INIS)

    Highlights: → Shut-Down Margin (SDM) for the next generation annular fuel core of typical VVER-1000, 13 x 13 assemblies are calculated. → The MCNP-5 code is run for many cases with different core burn up at various core temperatures. → There is a substantial drop in SDM in the case of annular fuel for the same power level. → SDM for our proposed VVER-1000 annular pins is calculated for specific average fuel burn up values at the BOC, MOC, and EOC. - Abstract: Shut-Down Margin (SDM) for the next generation annular fuel core of typical VVER-1000, 13 x 13 assemblies are calculated as the main aim of the present research. We have applied the MCNP-5 code for many cases with different values of core burn up at various core temperatures, and therefore their corresponding coolant densities and boric acid concentrations. There is a substantial drop in SDM in the case of annular fuel for the same power level. Specifically, SDM for our proposed VVER-1000 annular pins is calculated when the average fuel burn up values at the BOC, MOC, and EOC are 0.531, 11.5, and 43 MW-days/kg-U, respectively.

  9. Health services research doctoral core competencies

    OpenAIRE

    Holve Erin; Martin Diane P; Forrest Christopher B; Millman Anne

    2009-01-01

    Abstract This manuscript presents an initial description of doctoral level core competencies for health services research (HSR). The competencies were developed by a review of the literature, text analysis of institutional accreditation self-studies submitted to the Council on Education for Public Health, and a consensus conference of HSR educators from US educational institutions. The competencies are described in broad terms which reflect the unique expertise, interests, and preferred learn...

  10. Health services research doctoral core competencies

    Directory of Open Access Journals (Sweden)

    Holve Erin

    2009-06-01

    Full Text Available Abstract This manuscript presents an initial description of doctoral level core competencies for health services research (HSR. The competencies were developed by a review of the literature, text analysis of institutional accreditation self-studies submitted to the Council on Education for Public Health, and a consensus conference of HSR educators from US educational institutions. The competencies are described in broad terms which reflect the unique expertise, interests, and preferred learning methods of academic HSR programs. This initial set of core competencies is published to generate further dialogue within and outside of the US about the most important learning objectives and methods for HSR training and to clarify the unique skills of HSR training program graduates.

  11. USGS Core Research Center (CRC) Collection of Cuttings

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Core Research Center (CRC) was established in 1974 by the U.S. Geological Survey (USGS) to preserve valuable rock cores for use by scientists and educators from...

  12. Analysis and research status of severe core damage accidents

    International Nuclear Information System (INIS)

    The Severe Core Damage Research and Analysis Task Force was established in Nuclear Safety Research Center, Tokai Research Establishment, JAERI, in May, 1982 to make a quantitative analysis on the issues related with the severe core damage accident and also to survey the present status of the research and provide the required research subjects on the severe core damage accident. This report summarizes the results of the works performed by the Task Force during last one and half years. The main subjects investigated are as follows; (1) Discussion on the purposes and necessities of severe core damage accident research, (2) proposal of phenomenological research subjects required in Japan, (3) analysis of severe core damage accidents and identification of risk dominant accident sequences, (4) investigation of significant physical phenomena in severe core damage accidents, and (5) survey of the research status. (author)

  13. Annular Flow Distribution test

    International Nuclear Information System (INIS)

    This report documents the Babcock and Wilcox (B ampersand W) Annular Flow Distribution testing for the Savannah River Laboratory (SRL). The objective of the Annular Flow Distribution Test Program is to characterize the flow distribution between annular coolant channels for the Mark-22 fuel assembly with the bottom fitting insert (BFI) in place. Flow rate measurements for each annular channel were obtained by establishing ''hydraulic similarity'' between an instrumented fuel assembly with the BFI removed and a ''reference'' fuel assembly with the BFI installed. Empirical correlations of annular flow rates were generated for a range of boundary conditions

  14. Future Directions for Research on Core Competencies

    Science.gov (United States)

    Bradshaw, Catherine P.; Guerra, Nancy G.

    2008-01-01

    This concluding commentary highlights common themes that emerged across the chapters in this volume. We identify strengths and limitations of the core competencies framework and discuss the importance of context, culture, and development for understanding the role of the core competencies in preventing risk behavior in adolescence. We also outline…

  15. Approach to development of high flux research reactor with pebble-bed core

    International Nuclear Information System (INIS)

    Full text: The research nuclear reactor of a basin-type IRT with the designed power of 1 MW was put into operation in 'Sosny' settlement not far from Minsk-city in the Republic of Belarus in 1962. In 1971 after its modernization the power was increased up to 4 MW and maximum density of neutron flux in the core was: Thermal 5·1013 neutr./cm2.s Fast (E>0.8 MeV) 2·1013 neutr./cm2.s The reactor has been used for carrying out investigations in the field of solid-state physics, radiation construction materials, radiobiology, gaseous chemically reacting coolants and others. After the Chernobyl NPP accident, in the former USSR the requirements on safety of nuclear reactors have become sufficiently stricter. As to some parameters these requirements became the same as for reactors of nuclear power plants. In this connection the reactor in 'Sosny' settlement did not answer these new requirements by a number of performances such as seismicity of building, efficiency of control and protection system, corrosion in the reactor vessel and others, and it was shutdown in 1987 and its decommissioning was performed during 1988-1999. At the Joint Institute of Power and Nuclear Research -'SOSNY' have been carried out investigations on feasibility of creation of the research reactor with pebble-bed core. The concept of such reactor supposes using the following technical approaches: - Using as fuel the brought sphere micro fuel elements with the diameter of 500-750 mkm to an industrial level; - Organization of reactor operation in the regime with minimum possible fueling with 235U; - Implementation of hydraulic loading - unloading of micro fuel elements with the frequency of one or several days. Physical calculations of the core were carried out with the help of MCU-RFFI program based on the Monte-Carlo method. Two configurations of the pebble-bed core in the high flux reactor have been considered. The first configuration is the core with a neutron trap and an annular fuel layer formed

  16. Core neutronics of a swimming pool research reactor

    International Nuclear Information System (INIS)

    The initial cores of the 5 MW swimming pool research reactor of the Nuclear Research Centre, Tehran have been analyzed using the computer codes METHUSELAH and EQUIPOISE. The effective multiplication factor, critical mass, moderator temperature and void coefficients of the core have been calculated and compared with vendor's values. Calculated values agree reasonably well with the vendor's results. (author)

  17. Numerical investigation on the enhancement capability of annular chimney towards natural convective heat transfer in the interior zone of scaled down FBR core catcher

    International Nuclear Information System (INIS)

    Full text of publication follows: A numerical study has been carried out to determine the influence of annular cylindrical chimney on buoyancy-induced flow in the dished end cavity of scaled down Fast Breeder Reactor. Results are presented for (i) cylindrical chimney configuration and (ii) annular chimney configuration occupying the center of the circular plate. Two dimensional laminar simulations are obtained by solving the fully elliptical governing equations of flow and energy. The fluid is Newtonian and incompressible and satisfies the Boussinesq approximation. Results for the upward facing isothermal circular plate with chimney configurations in confined enclosure are analyzed. The velocity fields and isotherms are studied extensively to assess the impact of both geometries on the flow structure, dynamics and overall heat transfer characteristics in the cavity, towards enhancement of natural convective heat transfer. The predicted results for the cylindrical chimney are compared with known experimental results. The results are of interest to post accident heat removal in fast breeder reactors (FBR). (authors)

  18. Experimental research of reactor core flooding

    International Nuclear Information System (INIS)

    The results are presented of experiments performed with the aim of finding the influence of the method of fixing the thermocouples for measuring the distribution of temperature to the wall of fuel pin simulator. This influence was found for the purpose of emergency core flooding. First experimental results on the effect of nitrogen dissolved in the water on the velocity of the cooling wave are given. These experiments were carried out under the following conditions: initial temperature in pin centre 300 to 600 degC, velocity of water at the inlet into the measuring section 3.5 to 20 cm/s, and atmospheric pressure in the model. (author)

  19. Uncertainty Evaluation for Core Thermal Power in a Research Reactor

    International Nuclear Information System (INIS)

    The Jordan Research and Training Reactor (JRTR) also has three separated and independent channels of the neutron detectors to measure the core power. To calibrate these detectors, the thermal power of Primary Cooling System (PCS) which cools down the heat generated in reactor core is used as calibration reference. The core thermal power can be estimated by the measured values of the mass flow rate, core inlet temperature, and core outlet temperature of coolant in the PCS. In general, the uncertainty of the core thermal power is required to be controlled below a certain value. To meet this requirement, the uncertainty of core thermal power should be evaluated based on the uncertainty of the measured parameters. In this paper, the uncertain evaluation is conducted with variation of the uncertainty of the measured parameters such as mass flow rate, core inlet temperature, core outlet temperature. In addition, the numbers of inlet and outlet temperature are considered to get a higher allowable uncertainty of temperature sensors. The core thermal power uncertainty has been valuated according to measuring parameters such as mass flow rate, temperatures, and number of RTDs. In this parametric study, allowable uncertainties for measuring devices have been obtained to guarantee 5% of the core thermal power uncertainty

  20. Uncertainty Evaluation for Core Thermal Power in a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sunil; Seo, Kyoung-Woo; Kim, Seong-Hoon; Chi, Dae-Young; Park, Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The Jordan Research and Training Reactor (JRTR) also has three separated and independent channels of the neutron detectors to measure the core power. To calibrate these detectors, the thermal power of Primary Cooling System (PCS) which cools down the heat generated in reactor core is used as calibration reference. The core thermal power can be estimated by the measured values of the mass flow rate, core inlet temperature, and core outlet temperature of coolant in the PCS. In general, the uncertainty of the core thermal power is required to be controlled below a certain value. To meet this requirement, the uncertainty of core thermal power should be evaluated based on the uncertainty of the measured parameters. In this paper, the uncertain evaluation is conducted with variation of the uncertainty of the measured parameters such as mass flow rate, core inlet temperature, core outlet temperature. In addition, the numbers of inlet and outlet temperature are considered to get a higher allowable uncertainty of temperature sensors. The core thermal power uncertainty has been valuated according to measuring parameters such as mass flow rate, temperatures, and number of RTDs. In this parametric study, allowable uncertainties for measuring devices have been obtained to guarantee 5% of the core thermal power uncertainty.

  1. Core management and full core conversion status of the Dalat nuclear research reactor

    International Nuclear Information System (INIS)

    The Dalat Nuclear Research Reactor (DNRR) is a pool type research reactor which was reconstructed in 1982 from the old 250 kW TRIGA-MARK II reactor. The reconstructed reactor reached its initial criticality in November 1983 and attained it nominal power of 500 kW in February 1984. The DNRR uses Russian fuel assemblies (FAs) type VVR-M2. The first fuel reloading was executed in April 1994 after more than 10 years of operation with 89 HEU FAs. The 11 new HEU FAs were added in the core periphery, at previous beryllium element locations. After reloading the working configuration of reactor core consisted of 100 HEU FAs. Second reloading for Dalat Nuclear Research Reactor was realized in March 2002. The 4 new HEU FAs were added in the core periphery, at previous beryllium element locations. After reloading the working configuration of reactor core consisted of 104 HEU FAs. The third fuel reloading by shuffling of HEU FAs was executed in June 2004. The shuffling of 16 HEU FAs with highest burn up in the centre of the core and 16 HEU FAs with low burn up in the core periphery was done. The working configuration of reactor core kept unchanged of 104 HEU FAs. The fourth fuel reloading was executed in November 2006. The 2 new HEU FAs were loaded in the core periphery, at previous locations of wet irradiation channel and dry irradiation channel. After reloading the working configuration of reactor core consisted of 106 HEU FAs. Contracts for reactor core conversion between USA, Russia, Vietnam and the International Atomic Energy Agency for Nuclear fuel manufacture and supply for DNRR and Return of Russian-origin non-irradiated highly enriched uranium fuel to the Russian Federation have been realized in 2007. The 35 fresh HEU FAs were sent back to Russian Federation. The 36 new LEU FAs from Russian Federation have been received. According to the results of design and safety analyses performed by the joint study between RERTR Program at Argonne National Laboratory and Vietnam

  2. 生物质固体成型燃料环模成型技术研究进展%The research progress in biomass annular mould forming for fuel technology

    Institute of Scientific and Technical Information of China (English)

    欧阳双平; 侯书林; 赵立欣; 田宜水; 孟海波

    2011-01-01

    综合分析了国内外生物质固体成型燃料环模成型技术、成型设备及产业发展现状.比较了生物质环模颗粒成型机和生物质环模压块成型机的性能和产品,指出了生物质固体成型燃料环模技术及设备存在着基础理论薄弱、原料适应差、易损件寿命短等问题;提出了我国生物质固体成型燃料环模成型技术的发展方向.%The research status of annular mould forming for fuel technology, forming equipment and the industrial development has been comprehensively analyzed, the performance and the product of annular mould biomass pellet machine and annular mould biomass briquette machine have been compared, the problems of biomass annular mould forming for fuel technology and equipment, such as weak theory foundation, poor feedstock suitability, short life span of wearing part, were pointed out, the development orientation of biomass annular mould forming for fuel technology in China was proposed.

  3. Annular pancreas (image)

    Science.gov (United States)

    Annular pancreas is an abnormal ring or collar of pancreatic tissue that encircles the duodenum (the part of the ... intestine that connects to stomach). This portion of pancreas can constrict the duodenum and block or impair ...

  4. Core conversion effects on the safety analysis of research reactors

    International Nuclear Information System (INIS)

    The safety related parameters of the 5 MW Democritus research reactor that will be affected by the scheduled core conversion to use LEU instead of HEU are considered. The analysis of the safety related items involved in such a core conversion, mainly the consequences due to MCA, DBA, etc., is of a general nature and can, therefore, be applied to other similar pool type reactors as well. (T.A.)

  5. Vibration model of a pressurized water reactor which takes into account the fluid influence in the annular gap between core barrel and pressure vessel

    International Nuclear Information System (INIS)

    A theoretical vibration model of a pressurized water reactor is established and studied which takes into account the fluid-structure interaction of the coupled three-dimensional system reactor pressure vessel-core barrel (reactor cavity). Vibration differential equations are derived only for the two-dimensional movement; the eigenfrequencies and amplitude ratios of the undamped system as well as a dimensionless damping factor of cavity vibrations are calculated with the data of the WWER-440, and discussed. (orig.)

  6. Biospecimen Core Resource - Office of Cancer Clinical Proteomics Research

    Science.gov (United States)

    The purpose of this notice is to notify the community that the National Cancer Institute's (NCI’s) Office of Cancer Clinical Proteomics Research (OCCPR) is seeking sources to establish a Biospecimen Core Resource (BCR), capable of receiving, qualifying, processing, and distributing annotated biospecimens.

  7. Generalized granuloma annulare

    Directory of Open Access Journals (Sweden)

    Khatri M

    1995-01-01

    Full Text Available A 35-years-old female patient had generalized pruritic papular lesions, distributed like dermatitis herpetiformis for last 4 years. Histopathologic changes were typical of granuloma annulare with negative results of direct immunofluorescence. The patient did not have association of diabetes mellitus or any other systemic disease. She failed to respond to dapsone therapy and 13-cis-retinoic acid.

  8. Oscillating annular liquid membranes

    International Nuclear Information System (INIS)

    The response of annular liquid membranes (e.g. used as protection systems in laser fusion reactors) to sinusoidal mass flow rate fluctuations at the nozzle exit is analyzed as a function of the amplitude and frequency of the axial velocity fluctuations at the nozzle exit and thermodynamic compression of the gas enclosed by the membrane. The pressure of the gases enclosed by the annular membrane and the axial distance at which the annular membrane merges on the symmetry axis are periodic functions of time which have the same period as that of the mass flow rate fluctuations at the nozzle exit. They are also nearly sinusoidal functions of time for small amplitudes of the mass flow rate fluctuations at the nozzle exit, and exhibit delay and lag times with respect to the sinusoidal axial velocity fluctuations at the nozzle exit. The delay and the lag times are functions of the amplitude and frequency of the mass flow rate fluctuations at the nozzle exit and the polytropic exponent. The amplitudes of both the pressure of the gases enclosed by the annular liquid membrane and the convergence length increase and decrease, resp., as the amplitude and frequency of the mass flow rate fluctuations at the nozzle exit, resp., are increased. They also increase as the polytropic exponent is increased. (orig.)

  9. Annular Planar Monopole Antennas

    OpenAIRE

    Chen, Z. N.; Ammann, Max; Chia, W.Y. W.; See, T.S. P.

    2002-01-01

    A type of annular planar monopole antenna is presented. The impedance and radiation characteristics of the monopole with different holes and feed gaps are experimentally examined. The measured results demonstrate that the proposed antenna is capable of providing significantly broad impedance bandwidth with acceptable radiation performance.

  10. Physics research for molten salt reactor with different core boundaries

    International Nuclear Information System (INIS)

    Background: Unlike the traditional solid fuel reactor with fixed boundary conditions, the inlet and outlet pipe and the core of molten salt reactor fuel is connected so that the flowing liquid fuel can travel freely between the pipe and core. Purpose: This article has made systematic study of the influence of different molten salt fuel regions on reactor physics, including the top and bottom fuel of the core vessel and the pipe fuel. The core physics was researched under different boundary conditions, and the region of the effective core was indicated subsequently. Methods: MSRE was taken as the reference reactor and the calculation was completed based on the Monte Carlo Code MCNP. Results: Results show that the fuel in the top and bottom of vessel impacts on keff and energy spectrum obviously. The influence of outlet pipe on keff was negligible when pipe radius less than 25 cm, and the perturbation of outlet pipe on the keff could be neglected when its length more than 20 cm. Conclusions: Results provide rational theory for the design of MSR and the development of computation code. (authors)

  11. Core Management of the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    The Dalat nuclear research reactor (DNRR) is a pool-type research reactor which was reconstructed from the 250 kW TRIGA-MARK II reactor. The reconstructed reactor reached its initial criticality in November 1983 and attained its nominal power of 500 kW in February 1984. The DNRR uses Russian fuel assemblies, type WWR-SM. The first fuel reloading was executed in April 1994 after more than ten years of operation with 89 fuel assemblies. Research on core management of DNRR with the purpose of maintaining safe operation and effective utilization of reserve fuel assem- blies is being carried out at the Nuclear Research Institute. Calculations of fuel burn-up for the Dalat nuclear research reactor are carried out based on the cell calculation program WIMS and two diffusion calculation programs HEXAGA and HEXNOD. Experimental measurement of fuel burn-up for the Dalat nuclear research reactor was realized by a measurement method of long-life isotopes from fission products. Optimum second fuel reloading and future refuelling for DNRR have been gained. A second fuel reloading for the Dalat nuclear research reactor was realized in March 2002. After reloading the working configuration of the reactor, the core consisted of 104 fuel assemblies. Research results for future refuelling for DNRR show that with 36 reserve fuel assemblies, the reactor will be operated for at least 17 851 h at nominal power since the second fuel reloading. (author)

  12. Interfacial friction in low flowrate vertical annular flow

    International Nuclear Information System (INIS)

    During boil-off and reflood transients in nuclear reactors, the core liquid inventory and inlet flowrate are largely determined by the interfacial friction in the reactor core. For these transients, annular flow occurs at relatively modest liquid flowrates and at the low heat fluxes typical of decay heat conditions. The resulting low vapor Reynolds numbers, are out of the data range used to develop the generally accepted interfacial friction relations for annular flow. In addition, most existing annular flow data comes from air/liquid adiabatic experiments with fully developed flows. By contrast, in a reactor core, the flow is continuously developing along the heated length as the vapor flowrate increases and the flow regimes evolve from bubbly to annular flow. Indeed, the entire annular flow regime may exist only over tens of L/D's. Despite these limitations, many of the advanced reactor safety analysis codes employ the Wallis model for interfacial friction in annular flow. Our analyses of the conditions existing at the end-of-reflood in the PERICLES tests have indicated that the Wallis model seriously underestimates the interfacial shear for low vapor velocity cocurrent upflow. To extend the annular flow data base to diabatic low flowrate conditions, the DADINE tests were re-analyzed. In these tests, both pressure drop and local cross-section averaged void fractions were measured. Thus, both the wall and interfacial shear can be deduced. Based on the results of this analysis, a new correlation is proposed for interfacial friction in annular flow. (authors). 5 figs., 12 refs

  13. An Annular Gap Acceleration Model for γ-ray Emission of Pulsars

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    If the binding energy of the pulsar's surface is not so high (the case of a neutron star), both negative and positive charges will flow out freely from the surface of the star. An annular free flow model for γ-ray emission of pulsars is suggested. It is emphasized that:(1) Two kinds of acceleration regions (annular and core) need to be taken into account. The annular acceleration region is defined by the magnetic field lines that cross the null charge surface within the light cylinder. (2) If the potential drop in the annular region of a pulsar is high enough (normally the case for young pulsars), charges in both the annular and the core regions could be accelerated and produce primary gamma-rays. Secondary pairs are generated in both regions and stream outwards to power the broadband radiations. (3) The potential drop grows more rapidly in the annular region than in the core region. The annular acceleration process is a key process for producing the observed wide emission beams. (4)The advantages of both the polar cap and outer gap models are retained in this model. The geometric properties of the γ-ray emission from the annular flow are analogous to that presented in a previous work by Qiao et al., which match the observations well. (5) Since charges with different signs leave the pulsar through the annular and the core regions respectively, the current closure problem can be partially solved.

  14. Hydrodynamic stability of inverted annular flow in an adiabatic simulation

    International Nuclear Information System (INIS)

    In experiments, inverted annular flow was simulated adiabatically with turbulent water jets, issuing downward from long aspect nozzles, enclosed in gas annuli. Velocities, diameters, and gas species were varied, and core jet length, shape, break-up mode, and dispersed-core droplet sizes were recorded at approximately 750 data points. Inverted annular flow was observed to develop into inverted slug flow at low relative velocities, and into dispersed droplet flow at high relative velocities. For both of the above transitions from inverted annular flow, a correlation for core jet length was developed by extending work done on free liquid jets to include this new, coaxial, jet disintegration phenomenon. The result, showing length dependence upon diameter, jet Reynolds number, jet Weber number, void fraction, and gas Weber number, correlates the data well, especially at moderate-to-large relative velocities

  15. CREATIVE CORE WITHIN THE SCIENTIFIC RESEARCH OF EDUCATIONAL PROBLEMS

    Directory of Open Access Journals (Sweden)

    Vladimir I. Zagvyazinsky

    2014-01-01

    Full Text Available Abstract. The authors continue to explore the issue of practical methodology of research studies (Education and Science No 8, 2014. The aim of the study is to reveal theoretical and practical issues of empirical-research activity at the stages of conceptual problem understanding of pedagogical research, ideas promotion, development of the conception of their practical implementation and formation of scientifically grounded constructive hypotheses. Methods. The applied methods include theoretical analysis, synthesis, abstraction, idealization, generalization, specification, comparison, classification, extrapolation, modelling and hermeneutic interpretation. Results. Scientifically grounded description of the process of creative research of educational problems is presented in the paper. The content of such notions as «concept», «creative core», «idea», «plot», and «research hypothesis» is revealed. The creative core of scientific research is considered in close relation to the key aspects of the scientific inquiry: problematics definition, research topic and subject selection, clarification of initial facts and theoretical statements, definition of the conceptual framework, determination of essential novelty. The authors explore the mechanisms of the scientific research through the analysis of the educational situation and initial facts to the formation of a conceptual framework of research and the further working out of its instrumental component. The creative core of research is characterised as its most difficult element including the inception of idea transformation, its realisation into some conception and expansion into productive hypothesis. The authors describe the types of hypotheses, such as working and scientific (ready, functional and explanatory. The logical and gnoseological structure of a scientific hypothesis is shown in the study. The methodological requirements to the formulation of the scientific hypotheses are

  16. Diffractive analysis of annular resonators.

    Science.gov (United States)

    Morin, M; Bélanger, P A

    1992-04-20

    The modal properties of annular resonators are investigated by using an approximate version of the Kirchhoff-Fresnel integral. It is shown that the radial diffraction of a thin annular beam with a large inside radius is similar to that of a cylindrical field distribution. This permits the formal demonstration of the equivalence that exists between large Fresnel number annular resonators and infinite strip resonators. The model explains the properties of annular resonators that have been observed either experimentally or numerically by others, such as the lack of azimuthal discrimination. PMID:20720842

  17. Reactor core materials research and integrated material database establishment

    International Nuclear Information System (INIS)

    Mainly two research areas were covered in this project. One is to establish the integrated database of nuclear materials, and the other is to study the behavior of reactor core materials, which are usually under the most severe condition in the operating plants. During the stage I of the project (for three years since 1999) in- and out of reactor properties of stainless steel, the major structural material for the core structures of PWR (Pressurized Water Reactor), were evaluated and specification of nuclear grade material was established. And the damaged core components from domestic power plants, e.g. orifice of CVCS, support pin of CRGT, etc. were investigated and the causes were revealed. To acquire more resistant materials to the nuclear environments, development of the alternative alloys was also conducted. For the integrated DB establishment, a task force team was set up including director of nuclear materials technology team, and projector leaders and relevant members from each project. The DB is now opened in public through the Internet

  18. Core conversion of the Portuguese research reactor to LEU fuel

    International Nuclear Information System (INIS)

    Core conversion of the Portuguese Research Reactor (RPI) to LEU fuel is being performed within IAEA's Technical Cooperation project POR/4/016, with financial support from the US and Portugal. CERCA was selected as manufacturer of the LEU assemblies by the IAEA after an international call for bids. CERCA provided a comprehensive package to the RPI which included the mechanical verification of the design of the assemblies, their manufacture and arrangements for a joint inspection of the finished assemblies. The LEU fuel assemblies were manufactured within 8 months upon final approval of the design. The safety analyses for the core conversion to LEU fuel were made with the assistance of the RERTR program and were submitted for review by the IAEA and by Portuguese authorities in January 2007. Revised documents were submitted in June 2007 addressing the issues raised during review. Regulatory approval was received in early August and core conversion was done in early September. All measured safety parameters are within the defined acceptance limits. Operation at full power is expected by the end of October. (author)

  19. Elastic instability in stratified core annular flow

    CERN Document Server

    Bonhomme, Oriane; Leng, Jacques; Colin, Annie

    2010-01-01

    We study experimentally the interfacial instability between a layer of dilute polymer solution and water flowing in a thin capillary. The use of microfluidic devices allows us to observe and quantify in great detail the features of the flow. At low velocities, the flow takes the form of a straight jet, while at high velocities, steady or advected wavy jets are produced. We demonstrate that the transition between these flow regimes is purely elastic -- it is caused by viscoelasticity of the polymer solution only. The linear stability analysis of the flow in the short-wave approximation captures quantitatively the flow diagram. Surprisingly, unstable flows are observed for strong velocities, whereas convected flows are observed for low velocities. We demonstrate that this instability can be used to measure rheological properties of dilute polymer solutions that are difficult to assess otherwise.

  20. Azimuthally forced flames in an annular combustor

    Science.gov (United States)

    Worth, Nicholas; Dawson, James; Mastorakos, Epaminondas

    2015-11-01

    Thermoacoustic instabilities are more likely to occur in lean burn combustion systems, making their adoption both difficult and costly. At present, our knowledge of such phenomena is insufficient to produce an inherently stable combustor by design, and therefore an improved understanding of these instabilities has become the focus of a significant research effort. Recent experimental and numerical studies have demonstrated that the symmetry of annular chambers permit a range of self-excited azimuthal modes to be generated in annular geometry, which can make the study of isolated modes difficult. While acoustic forcing is common in single flame experiments, no equivalent for forced azimuthal modes in an annular chamber have been demonstrated. The present investigation focuses on the novel application of acoustic forcing to a laboratory scale annular combustor, in order to generate azimuthal standing wave modes at a prescribed frequency and amplitude. The results focus on the ability of the method to isolate the mode of oscillation using experimental pressure and high speed OH* measurements. The successful excitation of azimuthal modes demonstrated represents an important step towards improving our fundamental understanding of this phenomena in practically relevant geometry.

  1. Research on the annular lithium jet concept for future laser-fusion reactors. Final report No. 3, Sept 1977--Dec 1978

    International Nuclear Information System (INIS)

    Experiments have been performed to determine the length for convergence or closure of a vertical, hollow annular water jet due to the action of surface tension forces. The data agree well with theoretical predictions up to a velocity of about 3 m/s. At higher velocities, the convergence lengths are less than predicted and this is attributed to the jet acting as an ejector pump and thereby reducing the air pressure inside the annulus to slightly sub-atmospheric values. The stability of such a jet is also discussed in the light of the fact that no hydrodynamic instabilities have been observed to date. Finally the results of a series of experiments on the flow spreading or splitting due to the presence of wedge-shaped obstacles in the path of the annular jet flow are described

  2. Detonation Initiation by Annular Jets and Shock Waves

    OpenAIRE

    Shepherd, Joseph E.

    2005-01-01

    The objective of this research is to experimentally determine the feasibility of initiating detonation in fuel-air mixtures using only the energy in hot, compressed air. The existing 6-inch shock tube at Caltech was used to create hot, high-pressure air behind a reflected shock wave. The hot air created an imploding annular shock wave when it jetted through an annular orifice into a 76-mm-diameter, 1-m-long tube attached to the end of the shock tube. A special test section with an annular ...

  3. 基于环网的短路电流计算方法研究%Research of short circuit current calculating method based on annular grid

    Institute of Scientific and Technical Information of China (English)

    陶建鑫; 朱本坤

    2013-01-01

    环形供电网络在舰船上被广泛应用,为了准确计算环形网络的短路电流,本文采用发电机的等效处理来计算环形网络的短路电流,通过发电机与阻抗串联和与其他发电机进行并联从而将网络简化,并使用MATLAB中的Simpowersystems工具箱对实例进行仿真验证,结果表明该方法有较高的准确性和精度,为复杂结构的电力系统的短路电流计算奠定了基础.%Annular grid has been widely used on ships, In order to accurately calculate the short circuit current of the annular, In this paper, using the equivalent generators to calculate the short circuit current on annular grid, the equivalent generators via generator with impedance in series and then parallel with other generators to simplify network, and use Simpowersystems toolbox in MATLAB to simulation instance, the results show that this method has higher accuracy and precision, Laid the foundation for the complex structure of the power system short-circuit current calculation.

  4. Assessment of Inner Channel Blockage on the Annular Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Shin, C. H.; In, W. K.; Oh, D. S.; Chun, T. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    A dual-cooled annular fuel for a pressurized water reactor (PWR) has been introduced for a significant amount of reactor power uprate. The Korea Atomic Energy Research Institute (KAERI) has been performing a research to develop a dual-cooled annular fuel for the power uprate of 20% in an optimized PWR in Korea, OPR1000. An inner channel blockage is principal one of technical issues of the annular fuel rod. The inner channel in an annular fuel is isolated from the neighbor channels unlike the outer channels. The inner channel will be faced with a DNB accident by the partial blockage. In this paper, the largest fractional channel blockage was assessed by subchannel analysis code MATRA-AF and an end plug design to complement inlet blockage of inner channel was estimated by CFD code, CFD-ACE

  5. Interfacial friction in cocurrent upward annular flow

    Science.gov (United States)

    Hossfeld, L. M.; Bharathan, D.; Wallis, G. B.; Richter, H. J.

    1982-03-01

    Cocurrent upward annular flow is investigated, with an emphasis on correlating and predicting pressure drop. Attention is given to the characteristics of the liquid flow in the film, and the interaction of the core with the film. Alternate approaches are discussed for correlating suitably defined interfacial friction factors. Both approaches are dependent on knowledge of the entrainment in order to make predictions. Dimensional analysis is used to define characteristic parameters of the flow and an effort is made to determine, to the extent possible, the influences of these parameters on the interfacial friction factor.

  6. Axisymmetric annular curtain stability

    International Nuclear Information System (INIS)

    A temporal stability analysis was carried out to investigate the stability of an axially moving viscous annular liquid jet subject to axisymmetric disturbances in surrounding co-flowing viscous gas media. We investigated in this study the effects of inertia, surface tension, the gas-to-liquid density ratio, the inner-to-outer radius ratio and the gas-to-liquid viscosity ratio on the stability of the jet. With an increase in inertia, the growth rate of the unstable disturbances is found to increase. The dominant (or most unstable) wavenumber decreases with increasing Reynolds number for larger values of the gas-to-liquid viscosity ratio. However, an opposite tendency for the most unstable wavenumber is predicted for small viscosity ratio in the same inertia range. The surrounding gas density, in the presence of viscosity, always reduces the growth rate, hence stabilizing the flow. There exists a critical value of the density ratio above which the flow becomes stable for very small viscosity ratio, whereas for large viscosity ratio, no stable flow appears in the same range of the density ratio. The curvature has a significant destabilizing effect on the thin annular jet, whereas for a relatively thick jet, the maximum growth rate decreases as the inner radius increases, irrespective of the surrounding gas viscosity. The degree of instability increases with Weber number for a relatively large viscosity ratio. In contrast, for small viscosity ratio, the growth rate exhibits a dramatic dependence on the surface tension. There is a small Weber number range, which depends on the viscosity ratio, where the flow is stable. The viscosity ratio always stabilizes the flow. However, the dominant wavenumber increases with increasing viscosity ratio. The range of unstable wavenumbers is affected only by the curvature effect. (paper)

  7. Stability of cantilevered coaxial shells with internal and annular flow

    International Nuclear Information System (INIS)

    This paper is a theoretical study of the stability of cantilevered coaxial cylindrical shells conveying incompressible fluid in the annular space in- between and within the inner shell. The viscous effects of the mean flow are taken into account, but the perturbations of the equilibrium state on the basis of which stability is assessed is carried out by means of potential flow theory, thus neglecting unsteady viscous effects which are known to become important for narrow annular flows. Shell displacements are described by Flugge's equations of motion. Solution of the coupled fluid-structure equations is carried out by means of the Fourier Transform Method. The main finding of this research is that stability is lost by flutter for internal flow, according to both the inviscid and viscous variants of the theory; for annular flow, however, whereas inviscid theory predicts loss of stability by flutter, viscous theory (with dissipative effects included) predicts that the shell loses stability by divergence and then, at appreciably higher flow, by flutter. Reduction of the annular gap generally destabilizes the system; while increased steady viscous effects slightly stabilize the system for internal flow, they strongly destabilize it for annular flow. Increasing the length of the shell destabilizes the system for both internal and annular flows. The presence of internal flow in addition to annular flow tends to stabilize the system vis-a-vis the case of annular flow, but only at low flow velocities, having the opposite effect at higher flows; the same effects arise when the main flow is internal and an annular flow added to the system

  8. Hydraulic lift-off issues for application of high performance annular fuels in pressurized water reactors

    International Nuclear Information System (INIS)

    Highlights: • Pin and assembly lift-off forces are compared between solid and annular fuel. • Annular fuel experiences much stronger uplift forces. • Much stronger hold-down forces are required by annular fuel assembly. • Engineering modifications for hold-down mechanisms are required by annular fuel. - Abstract: In the PWR core, the fuel assembly is firmly seated on the lower core plate during operation. However, if the hydraulic force exerted on the fuel assembly by coolant flow is too large and the fuel assembly is lifted-off from the lower core plate, the excessive vibration will cause fuel failure. Therefore, the hydraulic lift-off issue needs to be addressed when the advanced fuel assembly is developed. It has been shown that the advanced annular fuel design with internal cooling allows power uprating up to 50% while the peak temperature of the fuel can be reduced and the MDNBR can be maintained. However, if the coolant condition in the core is kept unchanged, increasing the core power by 50% requires the core flow rate also increase proportionally, which will give rise to the hydraulic lift-off, an important issue to be addressed. In this paper, taking the 17 × 17 solid fuel design as the reference, the hydraulic lift-off issue is investigated for proposed 12 × 12 and 13 × 13 annular fuel designs. Both the steady-state and start-up operating conditions are evaluated. It is found that the hydraulic lift-off indeed is an issue for annular fuel design which requires careful analysis. By comparison, the lift-off forces and hold-down forces required for the externally and internally cooled annular fuels (13 × 13 and 12 × 12 arrays) are several times larger than that of the referenced solid fuel (17 × 17 array). Therefore, the hold-down mechanism for annular fuel needs to be carefully designed

  9. Annular Hybrid Rocket Motor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Engineers at SpaceDev have conducted a preliminary design and analysis of a proprietary annular design concept for a hybrid motor. A U.S. Patent application has...

  10. Manufacture of annular cermet articles

    Science.gov (United States)

    Forsberg, Charles W.; Sikka, Vinod K.

    2004-11-02

    A method to produce annular-shaped, metal-clad cermet components directly produces the form and avoids multiple fabrication steps such as rolling and welding. The method includes the steps of: providing an annular hollow form with inner and outer side walls; filling the form with a particulate mixture of ceramic and metal; closing, evacuating, and hermetically sealing the form; heating the form to an appropriate temperature; and applying force to consolidate the particulate mixture into solid cermet.

  11. Analytical solution of neutron transport equation in an annular reactor with a rotating pulsed source

    International Nuclear Information System (INIS)

    In this study, an analytical solution of the neutron transport equation in an annular reactor is presented with a short and rotating neutron source of the type S(x) δ (x- Vt), where V is the speed of annular pulsed reactor. The study is an extension of a previous study by Williams [12] carried out with a pulsed source of the type S(x) δ (t). In the new concept of annular pulsed reactor designed to produce continuous high flux, the core consists of a subcritical annular geometry pulsed by a rotating modulator, producing local super prompt critical condition, thereby giving origin to a rotating neutron pulse. An analytical solution is obtained by opening up of the annular geometry and applying one energy group transport theory in one dimension using applied mathematical techniques of Laplace transform and Complex Variables. The general solution for the flux consists of a fundamental mode, a finite number of harmonics and a transient integral. A condition which limits the number of harmonics depending upon the circumference of the annular geometry has been obtained. Inverse Laplace transform technique is used to analyse instability condition in annular reactor core. A regenerator parameter in conjunction with perimeter of the ring and nuclear properties is used to obtain stable and unstable harmonics and to verify if these exist. It is found that the solution does not present instability in the conditions stated in the new concept of annular pulsed reactor. (author)

  12. The influence of annular seal clearance to the critical speed of the multistage pump

    International Nuclear Information System (INIS)

    In the multistage pump of high head, pressure difference in two ends of annular seal clearance and rotor eccentric would produce the sealing fluid force, the effect of which can be expressed by a damping and stiffness coefficient. It has a great influence on the critical speed of the rotor system. In order to research the influence of the annular seal to the rotor system, this paper used CFD method to conduct the numerical simulation for the flow field of annular seal clearance. The radial and tangential forces were obtained to calculate the annular dynamic coefficients. Also dynamic coefficient were obtained by Matlab. The rotor system was modeled using ANSYS finite software and the critical speed with and without annular seal clearance were calculated. The result shows: annular seal's fluid field is under the comprehensive effect of pressure difference and rotor entrainment. Due to the huge pressure difference in front annular seal, fluid flows under pressure difference; the low pressure difference results in the more obvious effect on the clearance field in back annular seal. The first order critical speed increases greatly with the annular seal clearance; while the average growth rate of the second order critical speed is only 3.2%; the third and fourth critical speed decreases little. Based on the above result, the annular seal has great influence to the first order speed, while has little influence on the rest

  13. Adiabatic Steam-Water Annular Flow in an Annular Geometry

    DEFF Research Database (Denmark)

    Andersen, P. S.; Würtz, J.

    1981-01-01

    Experimental results for fully developed steam-water annular flow in annular geometries are presented. Rod and tube film flow rates and axial pressure gradients were measured for mass fluxes between 500 and 2000 kg/m2s, steam qualities between 20 and 60 per cent and pressures ranging from 3 to 9...... MPa. It was found that the measured tube film flow rate per unit tube perimeter is always many times greater than the corresponding rod film flow rate. Possible explanations for this asymmetry are discussed....

  14. CORE

    DEFF Research Database (Denmark)

    Krigslund, Jeppe; Hansen, Jonas; Hundebøll, Martin;

    2013-01-01

    different flows. Instead of maintaining these approaches separate, we propose a protocol (CORE) that brings together these coding mechanisms. Our protocol uses random linear network coding (RLNC) for intra- session coding but allows nodes in the network to setup inter- session coding regions where flows...... increase the benefits of XORing by exploiting the underlying RLNC structure of individual flows. This goes beyond providing additional reliability to each individual session and beyond exploiting coding opportunistically. Our numerical results show that CORE outperforms both forwarding and COPE......-like schemes in general. More importantly, we show gains of up to 4 fold over COPE-like schemes in terms of transmissions per packet in one of the investigated topologies....

  15. Development of a core follow calculational system for research reactors

    International Nuclear Information System (INIS)

    Over the last few years a comprehensive Pressurized Water Reactor and Materials Testing Reactor core analysis code system based on modern reactor physics methods has been under development by the Atomic Energy Corporation of South Africa. This system, known as OSCAR-3, will incorporate a customized graphical user interface and data management system to ensure user-friendliness and good quality control. The system has now reached the stage of development where it can be used for practical MTR core analyses. This paper describes the current capabilities of the components of the OSCAR-3 package, their integration within the package, and outlines future developments. 10 refs., 1 tab., 1 fig

  16. Neutronic Design of the First Core of the Replacement Research Reactor

    International Nuclear Information System (INIS)

    The paper describes the general neutronic characteristics of the first core of the replacement research reactor (RRR) for the Australian Nuclear Science and Technology Organisation (ANSTO). A compact core with 16 FA has been designed to fulfil all the very demanding neutronic requirements of the RRR facility. The contractual performance parameters must be verified for the equilibrium core; a very important design effort was carried out in the initial fresh core to have a similar performance. The description covers different aspects of the neutronic design: a detailed nuclear design of U3Si2 first core, the design calculation tools, together with a comparison of the first core performance against the core design criteria and the equilibrium core performance. (author)

  17. A Research Agenda for the Common Core State Standards: What Information Do Policymakers Need?

    Science.gov (United States)

    Rentner, Diane Stark; Ferguson, Maria

    2014-01-01

    This report looks specifically at the information and data needs of policymakers related to the Common Core State Standards (CCSS) and the types of research that could provide this information. The ideas in this report were informed by a series of meetings and discussions about a possible research agenda for the Common Core, sponsored by the…

  18. Core Journals of the Rapidly Changing Research Front of "Superconductivity."

    Science.gov (United States)

    Brooks, Terrence A.

    1989-01-01

    Analyzes journals contributing to the literature on superconductivity from 1973 to 1987. Reveals that the literature has tripled in absolute size and has grown in intensity. Identifies core journals and reveals a communication pattern dominated by two journals, "Physical Review Letters" and "Physical Review B." Critiques problems of bibliometric…

  19. Visualization of the annular synthetic jet

    Czech Academy of Sciences Publication Activity Database

    Broučková, Zuzana; Trávníček, Zdeněk; Šafařík, Pavel

    Praha: Ústav termomechaniky AV ČR, v. v. i, 2012 - (Šimurda, D.; Kozel, K.), s. 13-16 ISBN 978-80-87012-40-6. [Topical Problems of Fluid Mechanics 2012. Praha (CZ), 15.02.2012-17.02.2012] R&D Projects: GA AV ČR(CZ) IAA200760801; GA ČR(CZ) GCP101/11/J019 Institutional research plan: CEZ:AV0Z20760514 Keywords : synthetic jet * annular jet * visualization Subject RIV: BK - Fluid Dynamics

  20. Phase flow rate measurements of annular flows

    OpenAIRE

    Al-Yarubi, Qahtan

    2010-01-01

    In the international oil and gas industry multiphase annular flow in pipelines and wells is extremely important, but not well understood. This thesis reports the development of an efficient and cheap method for measuring the phase flow rates in two phase annular and annular mist flow, in which the liquid phase is electrically conducting, using ultrasonic and conductance techniques. The method measures changes in the conductance of the liquid film formed during annular flow and uses these to c...

  1. Infusing Qualitative Research Experiences into Core Counseling Curriculum Courses

    Science.gov (United States)

    Letourneau, Jade L. H.

    2015-01-01

    Many calls to action for promoting research with counselors-in-training and producing research-practitioners have been published over the past few decades (Balkin 2013; Granello and Granello 1998; Heppner and Anderson 1985), yet the research-practice gap remains. This article explores how qualitative research may help bridge that gap and offers…

  2. Hydrodynamics of annular-dispersed flow

    International Nuclear Information System (INIS)

    The interfacial drag, droplet entrainment, and droplet size distributions are important for detailed mechanistic modeling of annular dispersed two-phase flow. In view of this, recently developed correlations for these parameters are presented and discussed in this paper. The drag correlations for multiple fluid particle systems have been developed from a similarity hypothesis based on the mixture viscosity model. The results show that the drag coefficient depends on the particle Reynolds number and droplet concentration. The onset on droplet entrainment significantly alters the mechanisms of mass, momentum, and energy transfer between the film and gas core flow as well as the transfer between the two-phase mixture and the wall. By assuming the roll wave entrainment mechanism, the correlations for the amount of entrained droplet as well as for the droplet size distribution have been obtained from a simple model in collaboration with a large number of data

  3. Analysis of fuel management pattern of research reactor core of the MTR type design

    International Nuclear Information System (INIS)

    Research reactor core design needs neutronics parameter calculation use computer codes. Research reactor MTR type is very interested because can be used as research and also a radioisotope production. The research reactor in Indonesia right now is already 25 years old. Therefore, it is needed to design a new research reactor as a compact core. Recent research reactor core is not enough to meet criteria acceptance in the UCD which already determined namely thermal neutron flux in the core is 1.0x1015 n/cm2s. so that it is necessary to be redesign the alternative core design. The new research reactor design is a MTR type with 5x5 configuration core, uses U9Mo-Al fuel, 70 cm of high and uses two certainly fuel management pattern. The aim of this research is to achieve neutron flux in the core to meet the criteria acceptance in the UCD. Calculation is done by using WIMSD-B, Batan-FUEL and Batan-3DIFF codes. The neutronic parameters to be achieved by this calculation are the power level of 50 MW thermal and core cycle of 20 days. The neutronics parameter calculation is done for new U-9Mo-Al fuel with variation of densities.The result of calculation showed that the fresh core with 5x5 configuration, 360 gram, 390 gram and 450 gram of fuel loadings have meet safety margin and acceptance criteria in the UCD at the thermal neutron flux is more then 1.0 x 1015 n/cm2s. But for equilibrium core is only the 450 gram of loading meet the acceptance criteria. (author)

  4. Subcutaneous granuloma annulare: radiologic appearance

    International Nuclear Information System (INIS)

    Objective. Granuloma annulare is an uncommon benign inflammatory dermatosis characterized by the formation of dermal papules with a tendency to form rings. There are several clinically distinct forms. The subcutaneous form is the most frequently encountered by radiologists, with the lesion presenting as a superficial mass. There are only a few scattered reports of the imaging appearance of this entity in the literature. We report the radiologic appearance of five cases of subcutaneous granuloma annulare. Design and patients. The radiologic images of five patients (three male, two female) with subcutaneous granuloma annulare were retrospectively studied. Mean patient age was 6.4 years (range, 2-13 years). The lesions occurred in the lower leg (two), foot, forearm, and hand. MR images were available for all lesions, gadolinium-enhanced imaging in three cases, radiographs in four, and bone scintigraphy in one. Results. Radiographs showed unmineralized nodular masses localized to the subcutaneous adipose tissue. The size range, in greatest dimension on imaging studies, was 1-4 cm. MR images show a mass with relatively decreased signal intensity on all pulse sequences, with variable but generally relatively well defined margins. There was extensive diffuse enhancement following gadolinium administration. Conclusion. The radiologic appearance of subcutaneous granuloma annulare is characteristic, typically demonstrating a nodular soft-tissue mass involving the subcutaneous adipose tissue. MR images show a mass with relatively decreased signal intensity on all pulse sequences and variable but generally well defined margins. There is extensive diffuse enhancement following gadolinium administration. Radiographs show a soft-tissue mass or soft-tissue swelling without evidence of bone involvement or mineralization. This radiologic appearance in a young individual is highly suggestive of subcutaneous granuloma annulare. (orig.)

  5. Development of conceptual nuclear design of 10MWt research reactor core

    International Nuclear Information System (INIS)

    KAERI has been devoted to develop export-oriented research reactors for a growing world-wide demand of new research reactor construction. Their ambition is that design of Korean research reactor must be competitive in commercial and technological based on the experience of the HANARO core design concept with thermal power of 30MW. They are developing a new research reactor named Advanced HANARO research Reactor (AHR) with thermal power of 20 MW. KAERI has export records of nuclear technology. In 1954-1967 two series of pool type research reactors based on the Russian design, VVR type and IRT type, have been constructed and commissioned in some countries as well as Russia. Nowadays Russian design is introducing again for export to developing countries such as Union of Myanmar. Therefore the objective of this research is that to build and innovative 10 MW research reactor core design based on the concept of HANARO core design to be competitive with Russian research reactor core design. system tool of HELIOS was used at the first stage in both cases which are research reactor using tubular type fuel assemblies and that reactor using pin type fuel assemblies. The reference core design of first kind of research reactor includes one in-core irradiation site at the core center. The neutron flux evaluations for core as well as reflector region were done through logical consistency of neutron flux distributions for individual assemblies. In order to find the optimum design, the parametric studies were carried out for assembly pitch, active fuel length, number of fuel ring in each assembly and so on. Design result shows the feasibility to have high neutron flux at in-core irradiation site. The second kind of research reactor is used the same kind of assemblies as HANARO and hence there is no optimization about basic design parameters. That core has only difference composition of assemblies and smaller specific power than HANARO. Since it is a reference core at first stage

  6. Value and challenges of research on health professions' core subjects in education.

    Science.gov (United States)

    Hooper, Barbara; Krishnagiri, Sheama; Pollie Price, M; Bilics, Andrea R; Taff, Steven D; Mitcham, Maralynne D

    2014-01-01

    Professions are organized around central concerns, or core subjects. Knowledge of a field's core subject is indispensable to effective practice, reasoning, and professional identity. In health professions education, however, core subjects are often obscured by the plethora of topics and skills that must be taught, rendering them largely implicit in the learning process. Core subjects and how they are addressed in curricula thus remain under-researched in health professions education. The scarcity of research can be attributed to the need for (1) explicating core subjects as the basis for learning, (2) language that describes professional education as connecting all learning to a field's core, and (3) research methods that go beyond early phases of research development, including a conceptual framework for understanding and studying core subjects. This paper presents strategies addressing each of these challenges that were developed through a pilot and a subsequent large national study of occupational therapy education. These strategies provide a foundation for dialogue and future research on the nature and function of core subjects in health professions education. PMID:25433182

  7. Annular-Efficient Triangulations of 3-manifolds

    CERN Document Server

    Jaco, William

    2011-01-01

    A triangulation of a compact 3-manifold is annular-efficient if it is 0-efficient and the only normal, incompressible annuli are thin edge-linking. If a compact 3-manifold has an annular-efficient triangulation, then it is irreducible, boundary-irreducible, and an-annular. Conversely, it is shown that for a compact, irreducible, boundary-irreducible, and an-annular 3-manifold, any triangulation can be modified to an annular-efficient triangulation. It follows that for a manifold satisfying this hypothesis, there are only a finite number of boundary slopes for incompressible and boundary-incompressible surfaces of a bounded Euler characteristic.

  8. McCARD for neutronics design and analysis of research reactor cores

    International Nuclear Information System (INIS)

    McCARD is a Monte Carlo (MC) neutron-photon transport simulation code developed exclusively for the neutronics design and analysis of nuclear reactor cores. McCARD is equipped with the hierarchical modeling and scripting functions, the CAD-based geometry processing module, the adjoint-weighted kinetics parameter and source multiplication factor estimation modules as well as the burnup analysis capability for the neutronics design and analysis of both research and power reactor cores. This paper highlights applicability of McCARD for the research reactor core neutronics analysis, as demonstrated for Kyoto University Critical Assembly, HANARO research reactor, and YALINA subcritical facility. (authors)

  9. Improvement of image processing algorithms for annular flow

    International Nuclear Information System (INIS)

    Annular flow occurs in a wide range of industrial heat-transfer equipment, including the top of a BWR core, in the steam generator of a PWR, and in postulated accident scenarios including critical heat flux (CHF) by dryout. The modeling of annular flow often requires information regarding the average thickness of liquid film at the periphery of the flow channel as a measurement of film roughness (film roughness concept). More recently, two-region modeling efforts require wave intermittency as a measurement of disturbance wave (as opposed to base film thickness) contribution to gas-to-liquid momentum transfer and pressure loss. The present work focuses on the characterization of film behaviors in annular flow using quantitative visualization. The data reduction codes for planar laser-induced flourescence (PLIF) imaging and back-lit quartz tube imaging have been further developed to improve measurement accuracy. Film thickness distribution (base film and wave), disturbance wave length, and wave intermittency estimates have been updated and applied to a recent two-region annular flow model. Outputs of average film thickness, pressure gradient, and average wave velocity have been modeled with mean absolute errors of 8.70%, 17.42%, and 19.14%, respectively. (author)

  10. Thermal hydraulic performance assessment of dual-cooled annular nuclear fuel for OPR-1000

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Chang-Hwan, E-mail: shinch@kaeri.re.kr [LWR Fuel Development Division, Korea Atomic Energy Research Institute, 989-111 Daedeok-Daero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Chun, Tae-Hyun, E-mail: thchun@kaeri.re.kr [LWR Fuel Development Division, Korea Atomic Energy Research Institute, 989-111 Daedeok-Daero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Oh, Dong-Seok, E-mail: dsoh1@kaeri.re.kr [LWR Fuel Development Division, Korea Atomic Energy Research Institute, 989-111 Daedeok-Daero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); In, Wang-Kee, E-mail: wkin@kaeri.re.kr [LWR Fuel Development Division, Korea Atomic Energy Research Institute, 989-111 Daedeok-Daero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of)

    2012-02-15

    Highlights: Black-Right-Pointing-Pointer A thermal hydraulic performance of a 12 Multiplication-Sign 12 annular fuel array is evaluated. Black-Right-Pointing-Pointer The subchannel analysis code for the dual-cooled annular fuel, MATRA-AF is validated. Black-Right-Pointing-Pointer We evaluate the sensitivity for geometry tolerances and operating parameter. Black-Right-Pointing-Pointer We decide the essential design parameters to uprate the power generation by dual-cooled annular fuel. Black-Right-Pointing-Pointer A thermal margin amount accommodating a 20% power-uprate seems viable. - Abstract: An internally and externally cooled annular fuel was proposed for an advance PWR, which can endure substantial power uprating. KAERI is pursuing the development for a reloading of power uprated annular fuel for the operating PWR reactors of OPR-1000. In this paper, the characteristics and verification of the MATRA-AF are described. The thermal hydraulic performance of a 12 Multiplication-Sign 12 annular fuel is calculated for the major design parameters and its performance is compared against the reference 16 Multiplication-Sign 16 cylindrical fuel assembly. In particular, the enhancements of the thermal hydraulic performance of dual-cooled annular fuel are estimated for the 100% normal power reactor core. The purpose of this study is to estimate a normal power for OPR-1000 with dual-cooled annular fuel, and ultimately to assess the feasibility of 120% core power. The parametric study was carried out for the fuel rod dimension, gap conductance, thermal diffusion coefficients, and pressure loss of the spacer grids. As a result of the analysis on the nominal power, annular fuel showed a sufficient margin available on DNB and fuel pellet temperature relative to cylindrical fuel. The margin amount seems accommodating a 20% power-uprate seems viable.

  11. A Quantitative Assessment of the Research Chefs Association Core Competencies for the Practicing Culinologist

    Science.gov (United States)

    Bissett, Rachel L.; Cheng, Michael S. H.; Brannan, Robert G.

    2010-01-01

    Professional organizations have linked core competency to professional success and competitive strategy. The Research Chefs Assn. (RCA) recently released 43 core competencies for practicing culinologists. Culinology[R] is a profession that links skills of culinary arts and food science and technology in the development of food products. An online…

  12. Managing of Core Competemces: A Case study of SP Research Technical Institute

    OpenAIRE

    Zehra, Khizran; Wong, Stephanie

    2009-01-01

      Master's Thesis in Innovation and Business Creation ______________________________________________________ Title: Managing of Core Competences: A case study of SP Technical Research Institute Authors: Khizran Zehra & Stephanie Wong Supervisor: Mike Danilovic Date: May 2009 Key Words: Core Competences, Technological Changes, Sustainable Development, Sustained Competitive Advantage and Organizational Learning. _______________________________________________________________ A...

  13. Comparative study of research reactor core utilizing LEU and mixed (LEU and HEU) fuels

    International Nuclear Information System (INIS)

    Two cores of a swimming pool type research reactor, PARR-1, comprising of i) Low Enriched Uranium (LEU) fuel only , ii) LEU fuel mixed with High Enriched Uranium (HEU) fuel, have been analyzed. This study aims to utilize the partially burnt HEU spent fuel elements from the spent fuel rack, with burnup much less than their designed discharge burnup limit, discharged from the reactor core at the time of dismantling the HEU core during the implementation of world wide core conversion project from HEU to LEU in mid 1980's. For this, some reactor physics characteristic parameters , important from reactor operation, control and safety point of view, have been calculated and compared for the above mentioned two cores. These results included, core criticality, excess reactivity, shutdown margin, integrated control rods' worth, flux/power distribution, power peaking factors and the reactivity feed back coefficients for both these cores. Reactor lattice and 3- dimensional core analysis codes, WIMS-D/4 and CITATION were employed for the calculations. For the mix-fueled core, excess reactivity is found to be on higher side, 617 pcm, and accordingly decrease in its shutdown margin is predicted as compared with the values for LEU core. This is due to the effectiveness of less burnt HEU fuel elements in the mix-fueled core. However, other parameters do not show any significant difference for both these cores, due to the location of less burnt HEU fuel element at the core periphery. These results provide the basis for the operation of the research reactor utilizing mixed fuel without affecting its performance from safety and utilization point of view. (author)

  14. A survey of core research in information systems

    CERN Document Server

    Sidorova, Anna; Torres, Russell; Johnson, Vess

    2013-01-01

    The Information Systems (IS) discipline was founded on the intersection of computer science and organizational sciences, and produced a rich body of research on topics ranging from database design and the strategic role of IT to website design and online consumer behavior. In this book, the authors provide an introduction to the discipline, its development, and the structure of IS research, at a level that is appropriate for emerging and current IS scholars. Guided by a bibliometric study of all research articles published in eight premier IS research journals over a 20-year period, the author

  15. Safety And Transient Analyses For Full Core Conversion Of The Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Preparing for full core conversion of Dalat Nuclear Research Reactor (DNRR), safety and transient analyses were carried out to confirm about ability to operate safely of proposed Low Enriched Uranium (LEU) working core. The initial LEU core consisting 92 LEU fuel assemblies and 12 Beryllium rods was analyzed under initiating events of uncontrolled withdrawal of a control rod, cooling pump failure, earthquake and fuel cladding fail. Working LEU core response were evaluated under these initial events based on RELAP/Mod3.2 computer code and other supported codes like ORIGEN, MCNP and MACCS2. Obtained results showed that safety of the reactor is maintained for all transients/accidents analyzed. (author)

  16. Archive of Geosample Information from the British Ocean Sediment Core Research Facility (BOSCORF)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The British Ocean Sediment Core Research Facility (BOSCORF), National Oceanography Centre, is a contributor to the Index to Marine and Lacustrine Geological Samples...

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

    International Nuclear Information System (INIS)

    Highlights: ► Kinetic parameters of Tehran research reactor mixed-core have been calculated. ► Burn-up effect on TRR kinetics parameters has been studied. ► Replacement of LEU-CFE with HEU-CFE in the TRR core has been investigated. ► Results of each mixed core were compared to the reference core. ► Calculation of kinetic parameters are necessary for reactivity and power excursion transient analysis. - Abstract: In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTRPC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change

  18. core calculations for ETRR-1 research reactor upgrading

    International Nuclear Information System (INIS)

    nuclear research reactors play an important role in supporting the nuclear energy program for most countries. research reactors are categorized according to the type of fuel, fuel enrichment, type of moderator and reflector, the power of the reactor and its application. most reactors initially operated at low power then an era began to up-rate the power by changing the fuel type, improving the thermal-hydraulic system performance and modifying the control system to comply with the new trends in research reactors and its applications. in this thesis, we carried out static calculation for the egyptian first research reactor ETRR-1 to evaluate its power upgrade possibility. firstly, we carried out cell calculation using WIMSD/4 code to study the variation of the infinite multiplication factor with the variation of fuel enrichment, lattice pitch and adding heavy water by increasing percentage to the ordinary water coolant

  19. Core elements of programmatic research in nursing: a case study.

    Science.gov (United States)

    Koch, Tina; Rolfe, Gary; Kralik, Debbie

    2005-01-01

    In this invited paper Tina Koch and Debbie Kralik present the establishment of a research program outside the precincts of a university and we ask Gary Rolfe to provide a commentary from the perspective of an academic. We argue that a dedicated research unit, with a clearly articulated philosophy and in response to research questions from clients, community and practitioners, provides the focus to drive the program. Although we have infrastructure from the RDNS Foundation, obtaining external funding to support our program is a central activity. Discernable outcomes of our collaborative inquiries are described as participants with whom we research narrate aspects of their experience, leading to enhancement of self agency and quality of life. We illustrate the reform potential as groups of research participants develop sustainable people networks. Most importantly, theoretical development is ongoing describing transition (ways in which people are able to take a chronic illness into their lives and move on) and better understanding on ways in which health care professionals can facilitate transition. Evidence based news letters are written in collaboration with practitioners, however we ponder about ways to further our research findings in practice. Gary Rolfe speculates about intermural or extramural research programs. He frames his response using Brand's criteria to research program decisions. In order of importance he asks: (1) will it be fun? (2) will we learn anything from it? (3) will it make the world a better place? (4) will it earn enough money to pay for the first three? Gary argues that one of the luxuries of working within the university sector as an academic is that he can occasionally ignore question four. PMID:16619898

  20. Advances in Research on Modern Agricultural Development in Grain Production Core Area of China

    Institute of Scientific and Technical Information of China (English)

    Yan; LIU

    2015-01-01

    Grain production core area is key region of modern agricultural development in China. Through summarizing related literature about grain production area and modern agricultural development researches both at home and abroad,it obtained characteristics and existing problems in the modern agricultural development of the grain production core area. It is found that there are many research perspectives in modern agricultural development of the grain production core area. On the basis of analyzing the grain production core area and connotation,mode and evaluation of the modern agricultural development,it is concluded that further study should be carried out for adopting which development mode and how to make evaluation,so as to provide theoretical guidance for balanced development of modern agriculture in grain production core area of different regions.

  1. Core elements of developmental epidemiologically based prevention research.

    Science.gov (United States)

    Kellam, S G; Koretz, D; Mościcki, E K

    1999-08-01

    In the early 1990's, important progress was documented in prevention research on mental and behavioral disorders, with recommendations for a prevention research agenda. One of the earliest implementation efforts was the workshop, "A Scientific Structure for the Emerging Field of Prevention Research," sponsored by the National Institute of Mental Health and The Johns Hopkins University Prevention Research Center, and held in Baltimore, Maryland, in December of 1994. The purpose of the workshop was to merge three perspectives from the traditionally disparate areas of epidemiology, life course development, and intervention trials technology into an integrated, interdisciplinary effort that would define a scientific structure enabling rapid advancement in prevention science. As a consequence of that workshop, the papers were written that are contained in this and the next special issue on prevention of the American Journal of Community Psychology. This first paper is a description of the salient features of developmental epidemiologically-based prevention research. Beyond the above three perspectives, we discuss the role of developmental and intervention theories; measurement of implementation, mediators, and moderators, including multi-stage sampling and measurement; the central role of multilevel growth modeling; concepts of attributable risk and prevented fraction; proximal/distal modeling and effect sizes; and partnerships between researchers and communities. PMID:10573831

  2. Demonstration of core neutronic calculation for research and training reactors via SCALE4.4

    International Nuclear Information System (INIS)

    In this work, full core modelling is performed to improve neutronic analyses capability for nuclear research reactors using SCALE4.4 system. KENOV.a module of SCALE4.4 system is utilized for full core neutronic analysis. The ORIGEN-S module is also coupled with the KENOV.a module to perform burnup dependent core analyses. Results of control rod worths for 1st cycle of Cekmece TR-2 research reactor are presented. In particular, coupling of KENOV.a and ORIGEN-S modules of SCALE4.4 is discussed. The preliminary results of 2-D burnup dependent neutronic calculations are also given. (author)

  3. Researches on a reactor core in heavy ion inertial fusion

    CERN Document Server

    Kondo, S; Iinuma, T; Kubo, K; Kato, H; Kawata, S; Ogoyski, A I

    2016-01-01

    In this paper a study on a fusion reactor core is presented in heavy ion inertial fusion (HIF), including the heavy ion beam (HIB) transport in a fusion reactor, a HIB interaction with a background gas, reactor cavity gas dynamics, the reactor gas backflow to the beam lines, and a HIB fusion reactor design. The HIB has remarkable preferable features to release the fusion energy in inertial fusion: in particle accelerators HIBs are generated with a high driver efficiency of ~30-40%, and the HIB ions deposit their energy inside of materials. Therefore, a requirement for the fusion target energy gain is relatively low, that would be ~50 to operate a HIF fusion reactor with a standard energy output of 1GW of electricity. In a fusion reactor the HIB charge neutralization is needed for a ballistic HIB transport. Multiple mechanical shutters would be installed at each HIB port at the reactor wall to stop the blast waves and the chamber gas backflow, so that the accelerator final elements would be protected from the ...

  4. Neutronics conceptual design of the innovative research reactor core using uranium molybdenum fuel

    International Nuclear Information System (INIS)

    The multipurpose of research reactor utilization make many countries build the new research reactor. Trend of this reactor for this moment is multipurpose reactor type with a compact core to get high neutron flux at the low or medium level of power. The research newest. Reactor in Indonesia right now is already 25 year old. Therefore, it is needed to design a new research reactor, called innovative research reactor (IRR) and then as an alternative to replace the old research reactor. The aim of this research is to get the optimal configuration of equilibrium core with the acceptance criteria are minimum thermal neutron flux is 2.5E14 n/cm2 s at the power level of 20 MW (minimum), length of cycle of more than 40 days, and the most efficient of using fuel in the core. Neutronics design has been performed for new fuel of U-9Mo-AI with various fuel density and reflector. Design calculation has been performed using WIMSD-5B and BATAN-FUEL computer codes. The calculation result of the conceptual design shows four core configurations namely 5x5, 5x7, 6x5 and 6x6. The optimalization result for equilibrium core of innovative research reactor is the 5x5 configuration with 450 gU fuel loading, berilium reflector, maximum thermal neutron flux at reflector is 3.33E14 n/cm2 sand length of cycle is 57 days is the most optimal of IRR. (author)

  5. Droplet sizes, dynamics and deposition in vertical annular flow

    International Nuclear Information System (INIS)

    The role of droplets in vertical upwards annular flow is investigated, focusing on the droplet size distributions, dynamics, and deposition phenomena. An experimental program was performed based on a new laser optical technique developed in these laboratories and implemented here for annular flow. This permitted the simultaneous measurement of droplet size, axial and radial velocity. The dependence of droplet size distributions on flow conditions is analyzed. The Upper-Log Normal function proves to be a good model for the size distribution. The mechanism controlling the maximum stable drop size was found to result from the interaction of the pressure fluctuations of the turbulent flow of the gas core with the droplet. The average axial droplet velocity showed a weak dependence on gas rates. This can be explained once the droplet size distribution and droplet size-velocity relationship are analyzed simultaneously. The surprising result from the droplet conditional analysis is that larger droplet travel faster than smaller ones. This dependence cannot be explained if the drag curves used do not take into account the high levels of turbulence present in the gas core in annular flow. If these are considered, then interesting new situations of multiplicity and stability of droplet terminal velocities are encountered. Also, the observed size-velocity relationship can be explained. A droplet deposition is formulated based on the particle inertia control. This permitted the calculation of rates of drop deposition directly from the droplet size and velocities data

  6. CFD Simulation of Annular Centrifugal Extractors

    OpenAIRE

    Vedantam, S.; Wardle, K. E.; Tamhane, T. V.; Ranade, V. V.; Joshi, J. B.

    2012-01-01

    Annular centrifugal extractors (ACE), also called annular centrifugal contactors offer several advantages over the other conventional process equipment such as low hold-up, high process throughput, low residence time, low solvent inventory and high turn down ratio. The equipment provides a very high value of mass transfer coefficient and interfacial area in the annular zone because of the high level of power consumption per unit volume and separation inside the rotor due to the high g of cent...

  7. Granuloma annulare in herpes zoster scars.

    Science.gov (United States)

    Ohata, C; Shirabe, H; Takagi, K; Kawatsu, T

    2000-03-01

    A 54-year-old Japanese female developed granuloma annulare twice in herpes zoster scars. Soon after the second event, she developed ulcerative colitis, which was well controlled by sulfonamides and corticosteroid suppository. She had no history of diabetes mellitus. There was no recurrence of granuloma annulare by June of 1999. Granuloma annulare might have contributed to the complications of ulcerative colitis, although this had not been noticed before. PMID:10774142

  8. Reference equilibrium core with central flux irradiation facility for Pakistan research reactor-1

    International Nuclear Information System (INIS)

    In order to assess various core parameters a reference equilibrium core with Low Enriched Uranium (LEU) fuel for Pakistan Research Reactor (PARR-1) was assembled. Due to increased volume of reference core, the average neutron flux reduced as compared to the first higher power operation. To get a higher neutron flux an irradiation facility was created in centre of the reference equilibrium core where the advantage of the neutron flux peaking was taken. Various low power experiments were performed in order to evaluate control rods worth and neutron flux mapping inside the core. The neutron flux inside the central irradiation facility almost doubled. With this arrangement reactor operation time was cut down from 72 hours to 48 hours for the production of the required specific radioactivity. (author)

  9. Systematic comparison of the use of annular and Zernike circle polynomials for annular wavefronts

    NARCIS (Netherlands)

    Mahajan, V.N.; Aftab, M.

    2010-01-01

    The theory of wavefront analysis of a noncircular wavefront is given and applied for a systematic comparison of the use of annular and Zernike circle polynomials for the analysis of an annular wavefront. It is shown that, unlike the annular coefficients, the circle coefficients generally change as t

  10. Research progress and recommendations on reactor pressure vessel integrity under hypothetical core melt down accident

    International Nuclear Information System (INIS)

    Background: It is very important to ensure the integrity of the reactor pressure vessel under core melt down accident. The high-temperature creep failure is the main failure mode of the reactor pressure vessel under core melt down accident. Purpose: This paper is to present an overview of research status and progress on high-temperature creep behavior of reactor pressure vessel considering the hypothetical core melt down scenario. Methods: Emphasis is placed on accomplished achievements in creep tests, scale model experiments and numerical simulation, and the domestic newly research productions on high-temperature creep behavior of reactor pressure vessel structure integrity. Conclusions: This paper also discusses the limitations of existing researches and indicates future research directions, such as multi-axis tensile tests, analysis of three-dimensional coupling temperature field, scaled model tests, and so on. (authors)

  11. ORPHEE research reactor: 3D core depletion calculation using Monte-Carlo code TRIPOLI-4®

    Science.gov (United States)

    Damian, F.; Brun, E.

    2014-06-01

    ORPHEE is a research reactor located at CEA Saclay. It aims at producing neutron beams for experiments. This is a pool-type reactor (heavy water), and the core is cooled by light water. Its thermal power is 14 MW. ORPHEE core is 90 cm height and has a cross section of 27x27 cm2. It is loaded with eight fuel assemblies characterized by a various number of fuel plates. The fuel plate is composed of aluminium and High Enriched Uranium (HEU). It is a once through core with a fuel cycle length of approximately 100 Equivalent Full Power Days (EFPD) and with a maximum burnup of 40%. Various analyses under progress at CEA concern the determination of the core neutronic parameters during irradiation. Taking into consideration the geometrical complexity of the core and the quasi absence of thermal feedback for nominal operation, the 3D core depletion calculations are performed using the Monte-Carlo code TRIPOLI-4® [1,2,3]. A preliminary validation of the depletion calculation was performed on a 2D core configuration by comparison with the deterministic transport code APOLLO2 [4]. The analysis showed the reliability of TRIPOLI-4® to calculate a complex core configuration using a large number of depleting regions with a high level of confidence.

  12. An Evaluation of the Annular Fuel and Bottle-Shaped Fuel Concepts for Sodium Fast Reactors

    OpenAIRE

    Memmott, Matthew; Buongiorno, Jacopo; Hejzlar, Pavel

    2010-01-01

    Two innovative fuel concepts, the internally and externally cooled annular fuel and the bottle-shaped fuel, were investigated with the goal of increasing the power density and reduce the pressure drop in the sodium-cooled fast reactor, respectively. The concepts were explored for both high- and low-conversion core configurations, and metal and oxide fuels. The annular fuel concept is best suited for low-conversion metal-fuelled cores, where it can enable a power uprate of ~20%; the magnitude ...

  13. Public values: core or confusion? Introduction to the centrality and puzzlement of public values research

    NARCIS (Netherlands)

    T. Beck Jørgensen; M.R. Rutgers

    2014-01-01

    This article provides the introduction to a symposium on contemporary public values research. It is argued that the contribution to this symposium represent a Public Values Perspective, distinct from other specific lines of research that also use public value as a core concept. Public administration

  14. Development of a Core Management Tool for the MYRRHA Irradiation Research Facility

    OpenAIRE

    Jaluvka, David

    2015-01-01

    This dissertation develops a core management tool called RELOAD-M capable of optimizing reactor-core fuel loadings for MYRRHA, the future fast-spectrum research facility currently under development at SCK-CEN, Belgium. Such a tool is needed for designing highly efficient loading patterns that reflect various performance objectives of the multipurpose machine. RELOAD-M can solve the single-cycle loading pattern optimization problem, using different metaheuristic optimization methods and reacto...

  15. Confocal Annular Josephson Tunnel Junctions

    Science.gov (United States)

    Monaco, Roberto

    2016-04-01

    The physics of Josephson tunnel junctions drastically depends on their geometrical configurations and here we show that also tiny geometrical details play a determinant role. More specifically, we develop the theory of short and long annular Josephson tunnel junctions delimited by two confocal ellipses. The behavior of a circular annular Josephson tunnel junction is then seen to be simply a special case of the above result. For junctions having a normalized perimeter less than one, the threshold curves in the presence of an in-plane magnetic field of arbitrary orientations are derived and computed even in the case with trapped Josephson vortices. For longer junctions, a numerical analysis is carried out after the derivation of the appropriate motion equation for the Josephson phase. We found that the system is modeled by a modified and perturbed sine-Gordon equation with a space-dependent effective Josephson penetration length inversely proportional to the local junction width. Both the fluxon statics and dynamics are deeply affected by the non-uniform annulus width. Static zero-field multiple-fluxon solutions exist even in the presence of a large bias current. The tangential velocity of a traveling fluxon is not determined by the balance between the driving and drag forces due to the dissipative losses. Furthermore, the fluxon motion is characterized by a strong radial inward acceleration which causes electromagnetic radiation concentrated at the ellipse equatorial points.

  16. Confocal Annular Josephson Tunnel Junctions

    Science.gov (United States)

    Monaco, Roberto

    2016-09-01

    The physics of Josephson tunnel junctions drastically depends on their geometrical configurations and here we show that also tiny geometrical details play a determinant role. More specifically, we develop the theory of short and long annular Josephson tunnel junctions delimited by two confocal ellipses. The behavior of a circular annular Josephson tunnel junction is then seen to be simply a special case of the above result. For junctions having a normalized perimeter less than one, the threshold curves in the presence of an in-plane magnetic field of arbitrary orientations are derived and computed even in the case with trapped Josephson vortices. For longer junctions, a numerical analysis is carried out after the derivation of the appropriate motion equation for the Josephson phase. We found that the system is modeled by a modified and perturbed sine-Gordon equation with a space-dependent effective Josephson penetration length inversely proportional to the local junction width. Both the fluxon statics and dynamics are deeply affected by the non-uniform annulus width. Static zero-field multiple-fluxon solutions exist even in the presence of a large bias current. The tangential velocity of a traveling fluxon is not determined by the balance between the driving and drag forces due to the dissipative losses. Furthermore, the fluxon motion is characterized by a strong radial inward acceleration which causes electromagnetic radiation concentrated at the ellipse equatorial points.

  17. Swiss Federal Energy Research Commission - Annual report 2008; Eidgenoessische Energieforschungskommission CORE. Jahresbericht 2008

    Energy Technology Data Exchange (ETDEWEB)

    Maus, K.

    2009-07-01

    This annual report presents a review of the activities carried out by the Swiss Federal Energy Research Commission CORE in the year 2008. Main points of interest were the definition of a new CORE vision, a review of all research programmes, co-operation and co-ordination with public and private institutes, active consultancy, recommendations for further education and training, improved international information exchange and good communication with business, politics and the general public. The definition of a concept for Swiss energy research for the period 2012 to 2016 is mentioned. The annual report also reports on an internal visit made to various laboratories of the Swiss Federal Institute of Technology in Lausanne and the Energy Center in Zurich. The focussing of CORE activities on particular themes is discussed

  18. Innovative research reactor core designed. Estimation and analysis of gamma heating distribution

    International Nuclear Information System (INIS)

    The Gamma heating value is an important factor needed for safety analysis of each experiments that will be realized on research reactor core. Gamma heat is internal heat source occurs in each irradiation facilities or any material irradiated in reactor core. This value should be determined correctly because of the safety related problems. The gamma heating value is in general depend on. reactor core characteristics, different one and other, and then each new reactor design should be completed by gamma heating data. The Innovative Research Reactor is one of the new reactor design that should be completed with any safety data, including the gamma heating value. For this reasons, calculation and analysis of gamma heating in the hole of reactor core and irradiation facilities in reflector had been done by using of modified and validated Gamset computer code. The result shown that gamma heating value of 11.75 W/g is the highest value at the center of reactor core, higher than gamma heating value of RSG-GAS. However, placement of all irradiation facilities in reflector show that safety characteristics for irradiation facilities of innovative research reactor more better than RSG-GAS reactor. Regarding the results obtained, and based on placement of irradiation facilities in reflector, can be concluded that innovative research reactor more safe for any irradiation used. (author)

  19. Experimental investigation on Heat Transfer Performance of Annular Flow Path Heat Pipe

    Energy Technology Data Exchange (ETDEWEB)

    Kim, In Guk; Kim, Kyung Mo; Jeong, Yeong Shin; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

    2015-05-15

    Mochizuki et al. was suggested the passive cooling system to spent nuclear fuel pool. Detail analysis of various heat pipe design cases was studied to determine the heat pipes cooling performance. Wang et al. suggested the concept PRHRS of MSR using sodium heat pipes, and the transient performance of high temperature sodium heat pipe was numerically simulated in the case of MSR accident. The meltdown at the Fukushima Daiichi nuclear power plants alarmed to the dangers of station blackout (SBO) accident. After the SBO accident, passive decay heat removal systems have been investigated to prevent the severe accidents. Mochizuki et al. suggested the heat pipes cooling system using loop heat pipes for decay heat removal cooling and analysis of heat pipe thermal resistance for boiling water reactor (BWR). The decay heat removal systems for pressurized water reactor (PWR) were suggested using natural convection mechanisms and modification of PWR design. Our group suggested the concept of a hybrid heat pipe with control rod as Passive IN-core Cooling System (PINCs) for decay heat removal for advanced nuclear power plant. Hybrid heat pipe is the combination of the heat pipe and control rod. In the present research, the main objective is to investigate the effect of the inner structure to the heat transfer performance of heat pipe containing neutron absorber material, B{sub 4}C. The main objective is to investigate the effect of the inner structure in heat pipe to the heat transfer performance with annular flow path. ABS pellet was used instead of B{sub 4}C pellet as cylindrical structures. The thermal performances of each heat pipes were measured experimentally. Among them, concentric heat pipe showed the best performance compared with others. 1. Annular evaporation section heat pipe and annular flow path heat pipe showed heat transfer degradation. 2. AHP also had annular vapor space and contact cooling surface per unit volume of vapor was increased. Heat transfer

  20. Soliton bunching in annular Josephson junctions

    DEFF Research Database (Denmark)

    Vernik, I.V; Lazarides, Nickos; Sørensen, Mads Peter;

    1996-01-01

    By studying soliton (fluxon) motion in long annular Josephson junctions it is possible to avoid the influence of the boundaries and soliton-soliton collisions present in linear junctions. A new experimental design consisting of a niobium coil placed on top of an annular junction has been used to...

  1. Bistability and hysteresis of annular impinging jets

    Science.gov (United States)

    Tisovsky, Tomas

    2016-06-01

    In present study, the bistability and hysteresis of annular impinging jets is investigated. Annular impinging jets are simulated using open source CFD code - OpenFOAM. Both flow field patterns of interest are obtained and hysteresis is found by means of dynamic mesh simulation. Effect of nozzle exit velocity on resulting hysteresis loop is also illustrated.

  2. ACT-CCREC Core Research Program: Study Questions and Design. ACT Working Paper Series. WP-2015-01

    Science.gov (United States)

    Cruce, Ty M.

    2015-01-01

    This report provides a non-technical overview of the guiding research questions and research design for the ACT-led core research program conducted on behalf of the GEAR UP College and Career Readiness Evaluation Consortium (CCREC). The core research program is a longitudinal study of the effectiveness of 14 GEAR UP state grants on the academic…

  3. Steady-state thermal hydraulic analysis of the equilibrium core of Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Steady-state thermal hydraulic analysis of Pakistan Research Reactor-1 (PARR-1) has been carried out. RELAP5/Mod 3.4 (a best-estimate system code) was employed. PARR-1 is a swimming pool type research reactor using MTR (Material Testing Reactor) type fuel. It uses low enriched uranium (<20%) fuel with light water flowing from top to bottom under gravity. Standard correlations were employed to compute various parameters, which include: coolant velocity distribution in the core; critical velocity; pressure drop; saturation temperature; temperature distribution in the core, OFI (onset of flow instability) and DNB (departure from nucleate boiling)

  4. Steady-state thermal hydraulic analysis of the equilibrium core of Pakistan research reactor-1

    Energy Technology Data Exchange (ETDEWEB)

    Bokhari, I.H. [SDTP User Group, Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore 45650, Islamabad (Pakistan)], E-mail: ihbokhari@yahoo.co.uk; Mahmood, T.; Chaudri, K.S. [SDTP User Group, Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore 45650, Islamabad (Pakistan)

    2007-10-15

    Steady-state thermal hydraulic analysis of Pakistan Research Reactor-1 (PARR-1) has been carried out. RELAP5/Mod 3.4 (a best-estimate system code) was employed. PARR-1 is a swimming pool type research reactor using MTR (Material Testing Reactor) type fuel. It uses low enriched uranium (<20%) fuel with light water flowing from top to bottom under gravity. Standard correlations were employed to compute various parameters, which include: coolant velocity distribution in the core; critical velocity; pressure drop; saturation temperature; temperature distribution in the core, OFI (onset of flow instability) and DNB (departure from nucleate boiling)

  5. Theoretical research on construction quality real-time monitoring and system integration of core rockfill dam

    Institute of Scientific and Technical Information of China (English)

    2009-01-01

    With the enlargement of core rockfill dam construction scale and the improvement of construction mechanization level, the traditional manual construction quality control method is now difficult to meet the quality and safety demands of modern dam construction, so automatic and real-time dam construction quality monitoring with high-techs is urgently needed. The paper makes theoretical research on construction quality real-time monitoring and system integration of core rockfill dam, proposes implementation method and integrated solution of construction quality real-time monitoring of core rockfill dam construction process, realizes refining, all-whether, entire-process and real-time control and analysis on key links of dam construction, and introduces the application of the construction quality real-time monitoring and system integration technology to a practical core rockfill dam project.

  6. Status of the compact core design for the Munich research reactor

    International Nuclear Information System (INIS)

    A novel 'compact core' has been proposed for our project of substantially modernizing the research reactor FRM at Munich. This core has about 20 cm diameter and 70 cm height, is cooled by H2O and surrounded by a large D2O moderator tank. It makes essential use of the new U3Si/Al dispersion fuel with very high Uranium density now available. We present the results of new, two-dimensional neutronic calculations and give an estimate of the probable burnup and reactivity behaviour of the compact core. We expect that this core can be effectively operated with an unperturbed multiplication factor of about 1.22, and that a maximum thermal neutron flux of 7 to 8·1014cm-,2s-1 can be achieved in the D2O tank at 20 MW reactor power. (author)

  7. Measurement of Quasi-periodic Oscillating Flow Motion in Simulated Dual-cooled Annular Fuel Bundle

    International Nuclear Information System (INIS)

    In order to increase a significant amount of reactor power in OPR1000, KAERI (Korea Atomic Energy Research Institute) has been developing a dual-cooled annular fuel. The dual-cooled annular fuel is simultaneously cooled by the water flow through the inner and the outer channels. KAERI proposed the 12x12 dual-cooled annular fuel array which was designed to be structurally compatible with the 16x16 cylindrical solid fuel array by maintaining the same array size and the guide tubes in the same locations, as shown in Fig. 1. In such a case, due to larger outer diameter of dual-cooled annular fuel than conventional solid fuel, a P/D (Pitch-to-Diameter ratio) of dual cooled annular fuel assembly becomes smaller than that of cylindrical solid fuel. A change in P/D of fuel bundle can cause a difference in the flow mixing phenomena between the dual-cooled annular and conventional cylindrical solid fuel assemblies. In this study, the rod bundle flow motion appearing in a small P/D case is investigated preliminarily using PIV (Particle Image Velocimetry) for dual-cooled annular fuel application

  8. A self-standing two-fluid CFD model for vertical upward two-phase annular flow

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Y., E-mail: yang_liu@mail.dlut.edu.c [Key Laboratory of Ocean Energy Utilization and Energy Conservation of Ministry of Education, Dalian University of Technology, Dalian 116024, Liaoning Province (China); Li, W.Z.; Quan, S.L. [Key Laboratory of Ocean Energy Utilization and Energy Conservation of Ministry of Education, Dalian University of Technology, Dalian 116024, Liaoning Province (China)

    2011-05-15

    Research highlights: A mathematic model for two-phase annular flow is established in this paper. Pressure loss and wall shear stress increase with inlet gas and liquid flow velocities. Droplet mass fraction distribution exhibits a concave profile radially. - Abstract: In this paper, a new two-fluid CFD (computational fluid dynamics) model is proposed to simulate the vertical upward two-phase annular flow. This model solves the basic mass and momentum equations for the gas core region flow and the liquid film flow, where the basic governing equations are accounted for by the commercial CFD package Fluent6.3.26. The liquid droplet flow and the interfacial inter-phase effects are accounted for by the programmable interface of Fluent, UDF (user defined function). Unlike previous models, the present model includes the effect of liquid roll waves directly determined from the CFD code. It is able to provide more detailed and, the most important, self-standing information for both the gas core flow and the film flow as well as the inner tube wall situations.

  9. Reactor core conversion studies of Ghana: Research Reactor-1 and proposal for addition of safety rod

    International Nuclear Information System (INIS)

    The inclusion of an additional safety rod in conjunction with a core conversion study of Ghana Research Reactor-1 (GHARR-1) was carried out using neutronics, thermal hydraulics and burnup codes. The study is based on a recommendation by Integrated Safety Assessment for Research Reactors (INSARP) mission to incorporate a safety rod to the reactor safety system as well as the need to replace the reactor fuel with LEU. Conversion from one fuel type to another requires a complete re-evaluation of the safety analysis. Changes to the reactivity worth, shutdown margin, power density and material properties must be taken into account, and appropriate modifications made. Neutronics analysis including burnup was studied followed by thermal hydraulics analyses which comprise steady state and transients. Four computer codes were used for the analysis; MCNP, REBUS, PLTEP and PARET. The neutronics analysis revealed that the LEU core must be operated at 34 Kw in order to attain the flux of 1.0E12 n/cm2.s as the nominal flux of the HEU core. The auxiliary safety rod placed at a modified irradiation site gives a better worth than the cadmium capsules. For core excess reactivity of 4 mk, 348 fuel pins would be appropriate for the GHARR-1 LEU core. Results indicate that flux level of 1.0E12 n/cm2.s in the inner irradiation channel will not be compromised, if the power of the LEU core is increased to 34 kW. The GHARR-1 core using LEU-U02-12.5% fuel can be operated for 23 shim cycles, with cycles length 2.5 years, for over 57 years at the 17 kW power level. All 23 LEU cycles meet the ∼ 4.0 mk excess reactivity required at the beginning of cycle . For comparison, the MNSR HEU reference core can also be operated for 23 shim cycles, but with a cycle length of 2.0 years for just over 46 years at 15.0kW power level. It is observed that the GHARR-1 core with LEU UO2 fuel enriched to 12.5% and a power level of 34 kW can be operated ∼25% longer than the current HEU core operated at 30 k

  10. 14 MW INR-TRIGA research reactor core conversion - emergency preparedness challenges

    International Nuclear Information System (INIS)

    INR-Pitesti TRIGA research reactor is basically a pool type reactor with a special design in order to fulfil the requirements for material testing, power reactor fuel and nuclear safety studies. The safety evaluation involved a several design basis accidents. For training purposes, and to exercise our ability to conduct Level-3 PSA studies, a severe accident scenario involving 14-MW INR-TRIGA research reactor has been developed. In this scenario is assumed that a large part of the reactor hall roof or a heavy object escaped from the crane hook is dropped over the 14-MW TRIGA-SSR core, resulting in mechanical damage of the core. It is assumed, also, that no core melting is occurring, but only fuel-cladding rupture being involved for several 25-pins fuel bundles. The paper evaluates the radiological consequences, both early and late consequences, from the emergency preparedness point of view. (author)

  11. Experiments about core melt stabilization. The research project 'Corium on material surfaces' (Comas)

    International Nuclear Information System (INIS)

    Experiments in controlling core meltdown accidents play a very special role in international programs of reactor safety research. The 'Corium on Material Surfaces' (Comas) research project, which is concentrated on representative studies of the dispersion behavior of prototypical core melts outside the reactor pressure vessel, has contributed important findings about the design requirements for core catchers since 1994. It has been shown that such mixed oxide - metal melts can be made to spread evenly even if the thickness of layers is small. In addition, the results obtained so far allowed conclusions to be drawn about the physico-chemical phenomena accompanying the transport of the melt as a necessary precondition for code verification. (orig.)

  12. Hard rock warehouse : Alberta's Core Research Centre unique in North America

    Energy Technology Data Exchange (ETDEWEB)

    Mahony, J.

    2005-09-01

    Alberta's Core Research Centre (CRC) is among the world's largest repositories for well cores. The facility is well known within the oil and gas industry, particularly among petroleum geologists who study the drill cuttings. The CRC, which stores samples from every Alberta well cored since 1925, has been an important aid to Canada's oil and gas industry. Canada is one of the few jurisdictions where well operators are required to provide core or drill cuttings to a public repository such as the CRC. The facility houses a collection of drill cuttings dating back to 1911, along with tour sheets going back about 50 years. Automation is the key to handling orders. Computer-equipped forklifts pull samples from the 30-foot-tall shelves that line the storage areas. The CRC's floor area measures 18,000 square metres following an expansion in 1983. In 2004, the facility received more than 50,000 boxes, of which the majority was oilsand cores, reflecting the shifting focus of Canada's oil industry. With the increased pace of development in the petroleum industry, fewer geologists have time to spend examining drill cores. Instead, they rely on well logs and seismic data due to advances in technology. However, the author emphasized that these tools only tell part of the story. The well core provides the ground-truth of the geophysical log, and must be examined to understand the reservoirs, correlations and stratigraphy. The CRC presently stores 300,000 trays of drill cuttings and about 1.2 million boxes of core. Despite its massive size, it is running out of room, particularly given the increased pace of drilling activity and steady flow of core to the centre. In response, some core is culled under certain criteria. In addition, oilsands operators are now required to submit only one core per section, a change that will greatly reduce the volume of oilsands core from surface mineable areas. 2 figs.

  13. Interfacial friction in cocurrent upward annular flow. Final report

    International Nuclear Information System (INIS)

    Cocurrent upward annular flow is investigated, with an emphasis on correlating and predicting pressure drop. Attention is given to the characteristics of the liquid flow in the film, and the interaction of the core with the film. Alternate approaches are discussed for correlating suitably defined interfacial friction factors. Both approaches are dependent on knowledge of the entrainment in order to make predictions. Dimensional analysis is used to define characteristic parameters of the flow and an effort is made to determine, to the extent possible, the influences of these parameters on the interfacial friction factor

  14. Thermal-hydraulic analysis of the MIT research reactor low enrichment uranium (LEU) Core

    International Nuclear Information System (INIS)

    The MIT research reactor (MITR) is converting from the existing high enrichment uranium (HEU) core to a low enrichment uranium (LEU) core using a high-density monolithic UMo fuel. The design of an optimum LEU core for the MIT reactor is evolving. The in-house multi-channel thermal-hydraulics code, MULCH, was developed specifically for the MITR. This code has been benchmarked against PLTEMP for steady-state analysis, and RELAP5 and temperature measurements for the loss of primary flow transient. In this paper, thermal hydraulic analyses using MULCH and RELAP5 in support of the MITR conversion tasks are described. Various fuel configurations are evaluated in order to support the LEU core design optimization study. The results show that a preferable LEU core design employs a fuel meat thickness of 20 mils with 18 plates per element with a hot channel factor less than 1.76. Simulation results also show that the LEU-fueled MITR can potentially operate at a higher power level, about 30 % higher than the current core. (authors)

  15. Performance of the WWR-M research reactor in Ukraine with a mixed-fuel core

    International Nuclear Information System (INIS)

    The in-core fuel management optimization problem was studied for the WWR-M research reactor in Ukraine. Two available types of fuel assemblies were considered: WWR-M2 with 36% enrichment and WWR-M5 with 90% enrichment. It has been demonstrated that a mixed core consisting of WWR-M2 and WWR-M5 fuel provides higher neutron flux and even less fuel expenditures in comparison with the core consisting of only WWR-M2 fuel. Utilization of mixed-fuel is especially effective for high 'neutron prices'. The optimum core configuration, as well as the number and location of fuel assemblies of different types in the core strongly depend on 'neutron price'. With growth of the 'neutron price', the total number of fuel assemblies in the core should be decreased, while the number of WWR-M5 fuel assemblies should be increased. For a not too low 'neutron price', WWR-M5 fuel assemblies should be located near the irradiation channels to make power density in these areas, limited by the maximum allowed temperature of fuel surface, as high as possible, thus increasing neutron flux. (author)

  16. Research-Based Writing Practices and the Common Core: Meta-Analysis and Meta-Synthesis

    Science.gov (United States)

    Graham, Steve; Harris, Karen R.; Santangelo, Tanya

    2015-01-01

    In order to meet writing objectives specified in the Common Core State Standards (CCSS), many teachers need to make significant changes in how writing is taught. While CCSS identified what students need to master, it did not provide guidance on how teachers are to meet these writing benchmarks. The current article presents research-supported…

  17. New Funding Opportunity: Biospecimen Core Resource - Office of Cancer Clinical Proteomics Research

    Science.gov (United States)

    The purpose of this notice is to notify the community that the National Cancer Institute's (NCI’s) Office of Cancer Clinical Proteomics Research (OCCPR) is seeking sources to establish a Biospecimen Core Resource (BCR), capable of receiving, qualifying, processing, and distributing annotated biospecimens.

  18. Research and Policy: Informational Texts and the Common Core Standards: What Are We Talking about, Anyway?

    Science.gov (United States)

    Maloch, Beth; Bomer, Randy

    2013-01-01

    This column responds to a recent push in education policy toward a curriculum that requires students to read and write more informational texts. Most evident in the now well-known Common Core State Standards, these policy moves echo, in some ways, a call by researchers for more informational texts, particularly in primary classrooms. In this…

  19. Excitational metamorphosis of surface flowfield under an impinging annular jet

    Czech Academy of Sciences Publication Activity Database

    Tesař, Václav; Trávníček, Zdeněk

    2008-01-01

    Roč. 144, č. 2 (2008), s. 312-316. ISSN 1385-8947 R&D Projects: GA ČR GA101/07/1499; GA AV ČR IAA200760705 Institutional research plan: CEZ:AV0Z20760514 Keywords : jets * impinging jets * flow topology * annular jets * stagnation points Subject RIV: BK - Fluid Dynamics Impact factor: 2.813, year: 2008 http://www.sciencedirect.com/

  20. Improving access to information – defining core electronic resources for research and wellbeing

    OpenAIRE

    2007-01-01

    Research and innovation are listed as the key success factors for the future development of Finnish prosperity and the Finnish economy. The Finnish libraries have developed a scenario to support this vision. University, polytechnic and research institute libraries as well as public libraries have defined the core electronic resources necessary to improve access to information in Finland. The primary aim of this work has been to provide information and justification for central funding for ele...

  1. Etizolam-induced superficial erythema annulare centrifugum.

    Science.gov (United States)

    Kuroda, K; Yabunami, H; Hisanaga, Y

    2002-01-01

    Erythema annulare centrifugum (EAC) is characterized by slowly enlarging annular erythematous lesions. Although the origin is not clear in most cases, EAC has been associated with infections, medications, and in rare cases, underlying malignancy. We describe a patient who developed annular erythematous lesions after etizolam administration. The eruptions were typical of the superficial form of EAC, both clinically and histopathologically. The lesions disappeared shortly after discontinuation of the medication. Patch testing with etizolam gave positive results. To our knowledge this is the first reported case of etizolam-induced superficial EAC. PMID:11952667

  2. Research progress of metal-organic framework nanomaterials with core-shell structure

    Directory of Open Access Journals (Sweden)

    ZHANG Hongwei

    2015-04-01

    Full Text Available Controllable integration of multicomponent inorganic nanoparticles (NPs and metal-organic frameworks (MOFs is leading to the creation of many new multifunctional materials.Metal-organic framework nanomaterials with core-shell structure possess synergy effect of inorganic NPs and MOFs for multifunctional applications.The outstanding advantages of NP@MOF,such as limitless selection of composition,tunable pore sizes of the shells,and multifunctional cores and shells,provide insight into their future development.The recent advances in the preparation of coordination polymer nanomaterials are reviewed.At last,the research progress in their applications in gas storage/separation and catalysis are introduced.

  3. Selecting a MAPLE research reactor core for 1-10 mW operation

    International Nuclear Information System (INIS)

    The MAPLE class of research reactors is designed so that a single reactor concept can satisfy a wide range of practical applications. This paper reports the results of physics studies performed on a number of potential core configurations fuelled with either 5 w/o or 8 w/o enriched UO2 or 20 w/o U3Si-Al and assesses the relative merits of each. Recommended core designs are given to maximize the neutron fluxes available for scientific application and isotope production

  4. The Translational Genomics Core at Partners Personalized Medicine: Facilitating the Transition of Research towards Personalized Medicine.

    Science.gov (United States)

    Blau, Ashley; Brown, Alison; Mahanta, Lisa; Amr, Sami S

    2016-01-01

    The Translational Genomics Core (TGC) at Partners Personalized Medicine (PPM) serves as a fee-for-service core laboratory for Partners Healthcare researchers, providing access to technology platforms and analysis pipelines for genomic, transcriptomic, and epigenomic research projects. The interaction of the TGC with various components of PPM provides it with a unique infrastructure that allows for greater IT and bioinformatics opportunities, such as sample tracking and data analysis. The following article describes some of the unique opportunities available to an academic research core operating within PPM, such the ability to develop analysis pipelines with a dedicated bioinformatics team and maintain a flexible Laboratory Information Management System (LIMS) with the support of an internal IT team, as well as the operational challenges encountered to respond to emerging technologies, diverse investigator needs, and high staff turnover. In addition, the implementation and operational role of the TGC in the Partners Biobank genotyping project of over 25,000 samples is presented as an example of core activities working with other components of PPM. PMID:26927185

  5. Safety and Economics of High Power Density PWR with Novel Annular Fuel

    International Nuclear Information System (INIS)

    The internally and externally cooled annular fuel is a new type of fuel for PWRs that enables an increase in core power density by 50% within the same or better safety margins as the traditional solid fuel. Each assembly of traditional side dimensions has 160 annular fuel rods arranged in a 13x13 array. Even at the much higher power density, the fuel exhibits substantially lower temperatures and a MDNBR margin comparable to that of the traditional solid fuel at nominal (100%) power. Safety analyses indicate that the new annular fuel can accommodate 50% power up-rate in a PWR and still maintain adequate safety margins for a variety of transients and accidents including Loss of Flow Accident, Main Steam Line Break, Large Break Loss of Coolant Accident and Rod Ejection Accident. An economic study of 50% up-rate of an existing 1200 MW(e) PWR using the annular fuel shows that: (1) an Internal Rate of Return (IRR) on the order of 20% or more can be expected from such projects, even when accounting for uncertainties in the fuel price, electricity price inflation and cost of equipment; (2) Gradual replacement of the solid core by annular batches prior to up-rating can improve the IRR by 2.3% to 3.5% as it allows to full use of the energy in two already paid for batches of solid fuel rather than discarding them. Mixing of annular and solid fuel assemblies in one core appears feasible due to similar pressure drop characteristics of both assemblies. (authors)

  6. Engineered safety feature, an emergency core cooling system at Pakistan research reactor-1

    International Nuclear Information System (INIS)

    In the present study effectiveness of emergency core cooling system (ECCS) has been studied in case loss of coolant accident occurs at Pakistan research reactor (PARR-1). The reactor is a swimming pool type using MTR fuel. It was converted from highly enriched uranium (HEU) to low enriched uranium (LEU) fuel in 1992. It was also upgraded from a steady-state power level of 5-10 MW. Several additional facilities were provided to satisfy the requirements of enhanced power level. For safety consideration, emergency core cooling system (ECCS) was also installed to avoid any possibility of core meltdown. Evaluation of ECCS has been carried out for which standard correlations have been employed to find peak clad temperature profile after loss of coolant accident

  7. Engineered safety feature, an emergency core cooling system at Pakistan research reactor-1

    Energy Technology Data Exchange (ETDEWEB)

    Bokhari, Ishtiaq Hussain [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)], E-mail: ishtiaq@pinstech.org.pk; Mahmood, Tariq [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)

    2008-06-15

    In the present study effectiveness of emergency core cooling system (ECCS) has been studied in case loss of coolant accident occurs at Pakistan research reactor (PARR-1). The reactor is a swimming pool type using MTR fuel. It was converted from highly enriched uranium (HEU) to low enriched uranium (LEU) fuel in 1992. It was also upgraded from a steady-state power level of 5-10 MW. Several additional facilities were provided to satisfy the requirements of enhanced power level. For safety consideration, emergency core cooling system (ECCS) was also installed to avoid any possibility of core meltdown. Evaluation of ECCS has been carried out for which standard correlations have been employed to find peak clad temperature profile after loss of coolant accident.

  8. Plutonium stabilization and storage research in the DNFSB 94-1 core technology program

    International Nuclear Information System (INIS)

    Recommendation 94-1 of the Defense Nuclear Facility Safety Board (DNFSB) addresses legacy actinide materials left in the US nuclear defense program pipeline when the production mission ended in 1989. The Department of Energy (DOE) Implementation Plan responding to this recommendation instituted a Core Technology program to augment the knowledge base about general chemical and physical processing and storage behavior and to assure safe interim nuclear material storage, until disposition policies are formulated. The Core Technology program focuses on plutonium, in concert with a complex-wide applied R/D program administered by Los Alamos National Laboratory. This paper will summarize the Core Technology program's first two years, describe the research program for FY98, and project the overall direction of the program in the future

  9. COVERING A CORE BY EXTRUSION

    Science.gov (United States)

    Karnie, A.J.

    1963-07-16

    A method of covering a cylindrical fuel core with a cladding metal ms described. The metal is forced between dies around the core from both ends in two opposing skirts, and as these meet the ends turn outward into an annular recess in the dics. By cutting off the raised portion formed by the recess, oxide impurities are eliminated. (AEC)

  10. Measurement of magnetic susceptibility on tailings cores report on cores obtained from the Ontario Research Foundation lysimeter experiment

    International Nuclear Information System (INIS)

    Bulk susceptibility and induced magnetic remanence results are reported for 40 cores obtained from the uranium tailings lysimeter experiment at the Ontario Research Foundation. Both methods indicate a broad threefold subdivision of the tailings pile. An upper zone is characterized by an enhanced susceptibility level, which is related to enhanced concentration of both magnetite and hematite. Depletion zones, where present, are of limited areal extent and strongly developed. An intermediate zone is characterized by a mixture of large areas of reduced susceptibility that separate smaller regions of slightly enhanced susceptibility. The zones of susceptibility depletion appear to define a dendritic drainage pattern. Locally in this zone magnetite is enhanced and hematite depleted. In the lowermost zone susceptibility levels are reduced over most of the tailings bed. Only in the upper most right hand corner is there any vestige of a positive susceptibility concentration. Both magnetite and hematite are strongly depleted in this lower zone. Visually it is apparent that this lowermost depleted zone correlates to the zones of strongest 'yellowcake' development

  11. Comparing neutronics codes performance in analyzing a fresh-fuelled research reactor core

    International Nuclear Information System (INIS)

    Highlights: • Calculation of neutron fluence rate with different neutronic codes is examined. • MCNP, TRIPOLI and CITATION were used for neutron fluence rate calculations. • The recently converted core of the Portuguese Research Reactor (RPI) was used. • Fresh fuel of low enrichment in U-235 was assumed. • Thermal, epithermal and fast neutron fluence rates were computed. - Abstract: In this paper the relative performance of different simulation approaches is examined, focusing on the neutron fluence rate distribution in a nuclear reactor core. The main scope of the work is to benchmark and validate the neutronics code systems utilized in the Greek Research Reactor (GRR-1) for a high-density Low Enriched Uranium (LEU) core of compact size. For this purpose the recently converted core of the Portuguese Research Reactor (RPI), fueled with fresh, low enrichment in U-235 fuel, was simulated with the stochastic code TRIPOLI and the deterministic code system XSDRN/CITATION. RPI was selected on the basis that it is a similar to GRR-1 pool-type reactor, using same fuel and control rods type, as well as same types of coolant, moderator and reflector. The neutron fluence rate in RPI was computed using each numerical approach with changed approximations. In this frame the stochastic code TRIPOLI was tested using two different nuclear data libraries, i.e., ENDF/B-VI versus JEFF3.1, and two different ways of source definition, i.e., “point sources”, placed in the center of each fuel cell, versus a “distributed source”, where each fuel volume was considered as a neutron source. The deterministic code system XSDRN/CITATION was tested with respect to the definition of the transverse leakages associated to each one-dimensional, user-defined core zone, as analyzed by the XSDRN code in order to provide the zone equivalent cross sections. Thermal, epithermal and fast neutron fluence rates were computed and local values found in a 15 cm segment immediately below the

  12. Monte Carlo simulation of core physics parameters of the Nigeria Research Reactor-1 (NIRR-1)

    International Nuclear Information System (INIS)

    The Monte Carlo N-Particle (MCNP) code, version 4C (MCNP4C) and a set of neutron cross-section data were used to develop an accurate three-dimensional computational model of the Nigeria Research Reactor-1 (NIRR-1). The geometry of the reactor core was modeled as closely as possible including the details of all the fuel elements, reactivity regulators, the control rod, all irradiation channels, and Be reflectors. The following reactor core physics parameters were calculated for the present highly enriched uranium (HEU) core: clean cold core excess reactivity (ρex), control rod (CR) and shim worth, shut down margin (SDM), neutron flux distributions in the irradiation channels, reactivity feedback coefficients and the kinetics parameters. The HEU input model was validated by experimental data from the final safety analyses report (SAR). The model predicted various key neutronics parameters fairly accurately and the calculated thermal neutron fluxes in the irradiation channels agree with the values obtained by foil activation method. Results indicate that the established Monte Carlo model is an accurate representation of the NIRR-1 HEU core and will be used to perform feasibility for conversion to low enriched uranium (LEU)

  13. Core Management and Calculation Tools for the WWR-M Research Reactor in Ukraine

    International Nuclear Information System (INIS)

    Location of fuel assemblies in the core satisfying all the safety constraints and fuel requirements, fuel types used, number of fuel assemblies of each type and their discharged burn-ups, as well as the number of beryllium blocks at the periphery of the core, are optimized for the WWR-M research reactor in Ukraine using the code PORT. To determine the best core layout providing high neutron flux and low fuel expenditures under the safety and fuel constraints, two optimization algorithms using successive integer linear programming and simulated annealing are applied. For neutronics calculation, the iterational hybrid method combining the diffusion model with higher approx- imations of the neutron transport equation is used. The thermal-hydraulics model is based on an empirical formula for the heat transfer coefficient and measurement data for hydraulic parameters, including relative coolant velocities and pressure drops. It has been shown that a mixed core consisting of WWR-M2 and WWR-M5 fuel provides higher neutron flux and even less fuel expenditures in comparison with the core consisting of only WWR-M2 fuel. Utilization of mixed fuel is especially effective for high 'neutron price'. The WWR-M reactor in Ukraine is being studied for conversion from HEU to LEU fuel. It is feasible to convert the reactor to LEU fuel using qualified LEU WWR-M2 fuel assemblies, which are currently available for this purpose. (author)

  14. High Thrust-to-Power Annular Engine Technology

    Science.gov (United States)

    Patterson, Michael J.; Thomas, Robert E.; Crofton, Mark W.; Young, Jason A.; Foster, John E.

    2015-01-01

    Gridded ion engines have the highest efficiency and total impulse of any mature electric propulsion technology, and have been successfully implemented for primary propulsion in both geocentric and heliocentric environments with excellent ground/in-space correlation of performance. However, they have not been optimized to maximize thrust-to-power, an important parameter for Earth orbit transfer applications. This publication discusses technology development work intended to maximize this parameter. These activities include investigating the capabilities of a non-conventional design approach, the annular engine, which has the potential of exceeding the thrust-to-power of other EP technologies. This publication discusses the status of this work, including the fabrication and initial tests of a large-area annular engine. This work is being conducted in collaboration among NASA Glenn Research Center, The Aerospace Corporation, and the University of Michigan.

  15. Divergent Field Annular Ion Engine Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The proposed work investigates an approach that would allow an annular ion engine geometry to achieve ion beam currents approaching the Child-Langmuir limit. In...

  16. The VVER Core Physics, Reactor Dosimetry, and Shielding Researches in the LR-0 Reactor

    International Nuclear Information System (INIS)

    Zero-power water reactor LR-0 was created by the Nuclear Research Institute Rez, Nuclear Machinery Skoda, and RRC 'Kurchatov Institute' for researches of neutron parameters of the WWER type power reactors core, fuel storages, and-first of all-for researches in the reactor pressure vessel and internals dosimetry. Suitable geometrical conditions and flexible technical arrangements of the LR-0 facility enabled to carry out the wide experimental program on several full-scale models (mock-ups) of the WWER-440 and WWER-1000 reactors. The tasks of that experiments were the measurements of the neutron (from thermal energy up to 10 MeV) and gamma (from 0.1 up to 10 MeV) spectra and integral parameters of neutron and gamma fields in the different representative points of the mock-ups from the core to the outer pressure vessel surface and the biological shielding (including channel for ex-reactor ionizing chamber), as well as the measurement of spatial power distribution in the core. Fast neutron (energy from 0.5 to 10 MeV) and gamma spectra were measured in several representative points of the mock-ups by the two-parameter spectrometer with the cylindrical stilbene scintillation detectors. Measurements in the thermal and epithermal neutron region were carried out with the activation method using a broad set of activation monitors and with the 3He(n,p) counter. Activation measurements with threshold fast neutron detectors enlarge also the proton-recoil spectra measurements, such activation measurements were carried out especially in cases, when a spectrometer couldn't be put in the necessary position. The core fission rate distribution was obtained by means of gamma-scanning of the fuel pins. The calculations were carried out by different methods (deterministic and Monte Carlo). Experimental and calculation results in the core, internals, pressure vessel and shielding are reviewed and compared. (Authors)

  17. A Coupled Calculation System for Optimal In-Core Fuel Management in Research Reactors

    International Nuclear Information System (INIS)

    The paper presents a coupled method to solve the problem of finding an optimal configuration of fuel elements in research reactor cores. Finding the optimal solution always requires a huge amount of calculations by traditional methods. Thus, in performing this work, the investigated way followed to overcome such difficulties, was a judicious combination of the artificial neural network (ANN) technique, together with the well known stochastic method which is simulated annealing (SA). It has been shown that the most distinguishing and attractive feature of such a system is the computational efficiency and an increasing probability in obtaining optimized solutions with reasonable error. Neural network offers a very fast core parameter prediction tool with reasonable accuracy, and the simulated annealing method offers a very effective searching procedure which avoids local minimum. A series of tests have been performed using a modified core configuration of the benchmark 10 MW IAEA low enriched uranium (LEU) research reactor and the result achieved is the optimum configuration of the studied core. (author)

  18. Neutronic Design of KALIMER-600 Core with Moderator Rods

    International Nuclear Information System (INIS)

    Recently, the liquid-metal reactor research team of the Korea Atomic Energy Research Institute (KAERI) designed a 600 MWe sodium-cooled, metallic fueled fast reactor meeting the goals of Generation-IV, such as economics and proliferation resistance. In this paper, the core design analysis and its performance are reported. The core is designed to have a conversion ratio slightly larger than unity with no blanket assemblies in order not to produce an excess amount of high grade plutonium and to have no need for external feeds of fissile materials. To mitigate the sodium void reactivity of the fuel-self-sufficient core with no blanket assemblies, several design changes from a reference core are tried; reduction of the active core height, annular type cores with central dummy assemblies, and the use of moderator (BeO or ZrH2) rods. As a result of the analysis, it is found that of the considered designs the use of moderator rods for the softening of the core neutron spectrum is the best choice for reducing the sodium void worth with the smallest changes from the reference fuel and assembly designs. The core analysis shows that the sodium void reactivity is reduced by ∼2$ in comparison with the reference core and the core has a much more negative fuel temperature reactivity feedback in comparison with the reference core. (authors)

  19. In core gamma dosimetry using thermoluminescence detectors (TLDs) in research reactor

    International Nuclear Information System (INIS)

    Since gamma flux co-exists with the neutrons in the reactor core of a research reactor, it becomes difficult to measure exclusively gamma dose rate. Whereas it is quite important to know the gamma dose rates while performing controlled experiments in a research reactor. With this urge experiments have been performed to measure gamma dose rate at central vertical port (CVP) of the University of Florida Training Reactor (UFTR) using thermoluminescence detectors (TLDs). It is demonstrated that among the commercially available TLDs, LiF/sub 2/ can be used to determine the in-core gamma dose rate at low rector power levels (a few hundred watts). The gamma dose rate depending upon reactor power at CVP of UFTR has been determined to be 0.66+- 0.09 Ghy/sup -1/W/sup -1/. Extrapolation of gamma dose higher power is discussed. (author)

  20. Design study of eventual core conversion for the research reactor RA

    International Nuclear Information System (INIS)

    Main options are specified for the future status of the 6.5 MW heavy water research reactor RA. Arguments pro and contra restarting the reactor are presented. When considering the option to restart the RA reactor, possibilities to improve its neutronic parameters, such as neutron flux values and irradiation capabilities are discussed, as well as the compliance with the worldwide activities of Reduced Enrichment for Research and Test Reactors (RERTR) program. Possibility of core conversion is examined. Detailed reactor physics design calculations are performed for different fuel types and uranium loading. For different fuel management schemes results are presented for the effective, multiplication factor, power distribution, fuel burnup and consumption. It is shown that, as far as reactor core parameters are considered, conversion to lower enrichment fuel could be easily accomplished. However, conversion to the lower enrichment could only be justified if combined with improvement of some other reactor attributes. (author)

  1. A 3D transport-based core analysis code for research reactors with unstructured geometry

    International Nuclear Information System (INIS)

    Highlights: • A core analysis code package based on 3D neutron transport calculation in complex geometry is developed. • The fine considerations on flux mapping, control rod effects and isotope depletion are modeled. • The code is proved to be with high accuracy and capable of handling flexible operational cases for research reactors. - Abstract: As an effort to enhance the accuracy in simulating the operations of research reactors, a 3D transport core analysis code system named REFT was developed. HELIOS is employed due to the flexibility of describing complex geometry. A 3D triangular nodal SN method transport solver, DNTR, endows the package the capability of modeling cores with unstructured geometry assemblies. A series of dedicated methods were introduced to meet the requirements of research reactor simulations. Afterwards, to make it more user friendly, a graphical user interface was also developed for REFT. In order to validate the developed code system, the calculated results were compared with the experimental results. Both the numerical and experimental results are in close agreement with each other, with the relative errors of keff being less than 0.5%. Results for depletion calculations were also verified by comparing them with the experimental data and acceptable consistency was observed in results

  2. Feasibility of maintaining natural convection mode core cooling in research reactor power upgrades

    International Nuclear Information System (INIS)

    Two operational concerns for natural convection coooled research reactors using plate type fuels are: 1) pool top 16N activity (PTNA), and 2) nucleate boiling in core channels. The feasibility assessment of a power upgrade while maintaining natural convection mode core cooling requires addressing these operational concerns. Previous studies have shown that: a) The conventional technique for reducing PTNA by plume dispersion may not be effective in a large power upgrade of research reactors with small pools. b) Currently used correlations to predict onset of nucleate boiling (ONB) in thin, rectangular core channels are not valid for low-velocity, upward flows such as encountered in natural convection cooling. The PTNA depends on the velocity distribution in the reactor pool. COMMIX-1A code is used to determine the three-dimensional velocity fields in The Ohio State University Research Reactor (OSURR) pool as a function of varying design conditions, following a power upgrade to 500 kW with LEU fuel. It is shown that a sufficiently deep stagnant water layer can be created below the pool top by properly choosing the disperser flow rate. The ONB heat flux is experimentally determined for channel gaps and upward flow velocities in the range 2mm-4mm and 3-16 cm/sec., respectively. Two alternatives to plume dispersion for reducing PTNA and a new correlation to determine the ONB heat flux in thin, rectangular channels under low-velocity, upward flow conditions are proposed. (Author)

  3. Canada's neglected tropical disease research network: who's in the core-who's on the periphery?

    Directory of Open Access Journals (Sweden)

    Kaye Phillips

    Full Text Available BACKGROUND: This study designed and applied accessible yet systematic methods to generate baseline information about the patterns and structure of Canada's neglected tropical disease (NTD research network; a network that, until recently, was formed and functioned on the periphery of strategic Canadian research funding. METHODOLOGY: MULTIPLE METHODS WERE USED TO CONDUCT THIS STUDY, INCLUDING: (1 a systematic bibliometric procedure to capture archival NTD publications and co-authorship data; (2 a country-level "core-periphery" network analysis to measure and map the structure of Canada's NTD co-authorship network including its size, density, cliques, and centralization; and (3 a statistical analysis to test the correlation between the position of countries in Canada's NTD network ("k-core measure" and the quantity and quality of research produced. PRINCIPAL FINDINGS: Over the past sixty years (1950-2010, Canadian researchers have contributed to 1,079 NTD publications, specializing in Leishmania, African sleeping sickness, and leprosy. Of this work, 70% of all first authors and co-authors (n = 4,145 have been Canadian. Since the 1990s, however, a network of international co-authorship activity has been emerging, with representation of researchers from 62 different countries; largely researchers from OECD countries (e.g. United States and United Kingdom and some non-OECD countries (e.g. Brazil and Iran. Canada has a core-periphery NTD international research structure, with a densely connected group of OECD countries and some African nations, such as Uganda and Kenya. Sitting predominantly on the periphery of this research network is a cluster of 16 non-OECD nations that fall within the lowest GDP percentile of the network. CONCLUSION/SIGNIFICANCE: The publication specialties, composition, and position of NTD researchers within Canada's NTD country network provide evidence that while Canadian researchers currently remain the overall gatekeepers

  4. The New Emergency Core Cooling (NECC) system for the National Research Universal (NRU) reactor

    International Nuclear Information System (INIS)

    The New Emergency Core Cooling (NECC) system is the penultimate of seven major safety upgrades being implemented at the National Research Universal (NRU) Reactor in Chalk River. The NECC upgrade was designed to improve the original systems for core cooling in the event of an unisolable failure within the primary cooling circuit. The NECC upgrade ensures that water is automatically made available to the emergency cooling circuit pumps in the event of a break. Reactor core cooling is achieved from the discharge of these pumps which distribute emergency coolant to the individual fuel rods. Heated water from the vessel returns to the heat exchangers within the emergency cooling circuits for heat removal to the secondary coolant. The NECC upgrade significantly improves protection for a wide range of Loss Of Coolant Accidents (LOCAs) through the use of design features such as component redundancy, automatic initiation and hazard qualification. The introduction of the NECC upgrade combined with previous improvements in liquid confinement capability provide a closed loop system that ensures stable long term reactor core cooling. CATHENA (Canadian Algorithm for THErmalhydraulic Network Analysis) analysis was performed to assess the NECC upgrade and to validate the design for credible leak scenarios. (author)

  5. Safety analysis for core conversion (from HEU to LEU) of Pakistan research reactor-2 (PARR-2)

    International Nuclear Information System (INIS)

    PARR-2 (Pakistan Research Reactor-2), an MNSR (Miniature Neutron Source Reactor) is to be converted from HEU (High Enriched Uranium) to LEU (Low Enriched Uranium) fuel, along with all current MNSRs in various other countries. The purpose of conversion is to minimize the use of HEU for non-proliferation of high-grade nuclear fuel. The present report presents thermal hydraulic and safety analyses of PARR-2 using existing HEU fuel as well as proposed LEU fuel. Presently, the core is comprised of 90.2% enriched UAl4-Al fuel. There are 344 fuel pins of 5.5 mm diameter. The core has a total of 994.8 g of U235. Standard computer code PARET/ANL (version 1992) was employed to perform steady-state and transient analyses. Various parameters were computed, which included: coolant outlet, maximum clad surface and maximum fuel centerline temperatures; and peak power and corresponding peak core temperatures resulting from a transient initiated by 4 mK positive reactivity insertion. Results were compared with the reported data in Final Safety Analysis Report (FSAR). It was found that the PARET results were in reasonable agreement with the manufacturer's results. Calculations were also carried out for the proposed LEU core with two suggested fuel pin sizes (5.5 mm and 5.1 mm diameter with 12.6% and 12.3% enrichment, respectively). Comparison of the LEU results with the existing HEU fuel has been made and discussed.

  6. Research on intelligent monitor for 3D power distribution of reactor core

    International Nuclear Information System (INIS)

    Highlights: • Core power distribution of ex-core measurement system has been reconstructed. • Building up an artificial intelligence model for 3-D core power distribution. • Error of the experiments has been reduced to 0.76%. • Methods for improving the accuracy of the model have been obtained. - Abstract: A real-time monitor for 3D reactor power distribution is critical for nuclear safety and high efficiency of NPP’s operation as well as for optimizing the control system, especially when the nuclear power plant (NPP) works at a certain power level or it works in load following operation. This paper was based on analyzing the monitor for 3D reactor power distribution technologies used in modern NPPs. Furthermore, considering the latest research outcomes, the paper proposed a method based on using an ex-core neutron detector system and a neural network to set up a real time monitor system for reactor’s 3D power distribution supervision. The results of the experiments performed on a reactor simulation machine illustrated that the new monitor system worked very well for a certain burn-up range during the fuel cycle. In addition, this new model could reduce the errors associated with the fitting of the distribution effectively, and several optimization methods were also obtained to improve the accuracy of the simulation model

  7. Design of a novel compact core with reduced enrichment for upgrading the research reactor Munich FRM

    International Nuclear Information System (INIS)

    As an important result of the international reduced enrichment programs new fuel materials with very high uranium densities have been made available to users. This progress in fuel technology should also allow the realization of advanced core concepts for research reactors. Following this idea we are presently designing a novel ''compact core'' of about 20 cm diameter and 60 cm height, which will be cooled by H2O and surrounded by a large D2O moderator tank. For a reactor power P=20 MW we expect a maximum thermal neutron flux phisub(th)sup(max) asymptotically equals 8.1014 cm-2 s-1 outside of the core, and in any case the ''quality factor'' phisub(th)sup(max)/P would be the highest ever achieved at a beam tube reactor. While we started out neutronic calculations assuming high U235-enrichment, our present results for reduced enrichment (around 45%) are very promising; further calculations are in progress. The power density radial formfactor of the compact core can be reduced significantly by varying the uranium density in the individual concentric fuel plate rings. (author)

  8. Demonstration of the reactivity constraint approach on SNL's annual core research reactor

    International Nuclear Information System (INIS)

    This paper reports on the initial demonstration of the reactivity constraint approach and its implementing algorithm, the MIT-CSDL Non-Linear Digital Controller, on the annual core research reactor (ACCR) that is operated by the Sandia National Laboratories. This demonstration constituted the first use of reactivity constraints for the closed-loop, digital control of reactor power on a facility other than the Massachusetts Institute of Technology's (MIT's) research reactor (MITR-II). Also, because the ACRR and the MITR-II are of very different design, these trials established the generic nature of the reactivity constraint approach

  9. Developments in fabrication of annular MOX fuel pellet for Indian fast reactor

    International Nuclear Information System (INIS)

    Mechanical rotary presses along with adoption of core rod feature were inducted for fabrication of intricate annular Mixed Oxide (MOX) pellets for Prototype Fast Breeder Reactor (PFBR). In the existing tooling, bottom plungers contain core rod whereas top plungers contain a central hole for the entry of core rod during compaction. Frequent manual clean up of top plungers after few operations were required due to settling of powder in the annular hole of top plungers during compaction. Delay in cleaning can also result in breakage of tooling apart from increase in the dose to extremities of personnel. New design of tooling has been introduced to clean up the top plungers online during the operation of rotary press. It leads to increase in the productivity, reduces the spillage of valuable nuclear material and also reduces man-rem to operators significantly. The present paper describes the modification in tooling design and compaction sequence established for online cleaning of top plungers. (author)

  10. Dual annular rotating open-quotes windowedclose quotes nuclear reflector reactor control system

    International Nuclear Information System (INIS)

    A nuclear reactor control system is provided in a nuclear reactor having a core operating in the fast neutron energy spectrum where criticality control is achieved by neutron leakage. The control system includes dual annular, rotatable reflector rings. There are two reflector rings: an inner reflector ring and an outer reflector ring. The reflectors are concentrically assembled, surround the reactor core, and each reflector ring includes a plurality of openings. The openings in each ring are capable of being aligned or non-aligned with each other. Independent driving means for each of the annular reflector rings is provided so that reactor criticality can be initiated and controlled by rotation of either reflector ring such that the extent of alignment of the openings in each ring controls the reflection of neutrons from the core. 4 figures

  11. Drug-resistant tuberculosis clinical trials: proposed core research definitions in adults.

    Science.gov (United States)

    Furin, J; Alirol, E; Allen, E; Fielding, K; Merle, C; Abubakar, I; Andersen, J; Davies, G; Dheda, K; Diacon, A; Dooley, K E; Dravnice, G; Eisenach, K; Everitt, D; Ferstenberg, D; Goolam-Mahomed, A; Grobusch, M P; Gupta, R; Harausz, E; Harrington, M; Horsburgh, C R; Lienhardt, C; McNeeley, D; Mitnick, C D; Nachman, S; Nahid, P; Nunn, A J; Phillips, P; Rodriguez, C; Shah, S; Wells, C; Thomas-Nyang'wa, B; du Cros, P

    2016-03-01

    Drug-resistant tuberculosis (DR-TB) is a growing public health problem, and for the first time in decades, new drugs for the treatment of this disease have been developed. These new drugs have prompted strengthened efforts in DR-TB clinical trials research, and there are now multiple ongoing and planned DR-TB clinical trials. To facilitate comparability and maximise policy impact, a common set of core research definitions is needed, and this paper presents a core set of efficacy and safety definitions as well as other important considerations in DR-TB clinical trials work. To elaborate these definitions, a search of clinical trials registries, published manuscripts and conference proceedings was undertaken to identify groups conducting trials of new regimens for the treatment of DR-TB. Individuals from these groups developed the core set of definitions presented here. Further work is needed to validate and assess the utility of these definitions but they represent an important first step to ensure there is comparability in clinical trials on multidrug-resistant TB. PMID:27046707

  12. Numerical Simulation of Fluid Mixing in Upper Annular Space of SMART during Early Stage of non-LOCA

    International Nuclear Information System (INIS)

    KAERI (Korea Atomic Energy Research Institute) is developing a passive safety injection system (PSIS) to supply cold borated water into a reactor coolant system (RCS) without any operator actions or AC power under the occurrence of postulated design basis accidents. The PSIS consists of four independent trains, each of which is furnished with a gravity drained core makeup tank (CMT) and a safety injection tank (SIT). The CMT is designed to provide makeup and boration functions to the RCS during the early stage of a loss of coolant accident (LOCA) and a non-LOCA. In this paper, we investigate numerically the fluid mixing characteristics in the upper annular space of SMART, especially when single-phase natural circulation is formed between the CMT and RCS following a non-LOCA such as a main steam line break. In this paper, the fluid mixing characteristics in the upper annular space of SMART are investigated numerically when single-phase natural circulation is formed between the RCS and CMT during the early stage of a non-LOCA

  13. Cold neutron tomography of annular coolant flow in a double subchannel model of a boiling water reactor

    International Nuclear Information System (INIS)

    Dryout of the liquid coolant film on fuel pins at the top of boiling water reactor (BWR) cores constitutes the type of heat transfer crisis relevant for the conditions of high void fractions. It is a limiting factor in the thermal power, and therefore the economy, of BWRs. Ongoing research on multiphase annular flow, specifically the liquid film thickness, is fundamental not only to nuclear reactor safety and operation but also to that of evaporators, condensers, and pipelines in a general industrial context. We have performed cold neutron tomography of adiabatic air water annular flow in a scaled up model of the subchannel geometry found in BWR fuel assemblies today. All imaging has been performed at the ICON beamline at the neutron spallation source SINQ at the Paul Scherrer Institut in Switzerland. Neutron tomography is shown to excel in investigating the interactions of air water two phase flows with spacer vanes of different geometry. The high resolution, high contrast measurements provide spatial distributions of the coolant on top of the surfaces of the spacer, including the vanes, and in the subchannel downstream of the spacers.

  14. Introduction of Kyoto University Research Reactor with low enriched uranium core

    International Nuclear Information System (INIS)

    Kyoto University Research Reactor (KUR) is a light water moderated / cooled tank-type reactor. The project to convert the KUR fuel from HEU to LEU was achieved on March 24, 2010. After the successful achievement of first criticality on April 15 and the reactor has been operating since May 28. The nuclear design of KUR with LEU core is calculated using SRAC code system with JENDL-3.3 and the burn-up calculations are performed using SRAC-COREBN. In this paper, the purpose of the project to convert the KUR fuel, the fuel transport, the experimental results, the burn-up characteristics and calculation of KUR LEU core are presented. (author)

  15. CFD Simulation of Annular Centrifugal Extractors

    Directory of Open Access Journals (Sweden)

    S. Vedantam

    2012-01-01

    Full Text Available Annular centrifugal extractors (ACE, also called annular centrifugal contactors offer several advantages over the other conventional process equipment such as low hold-up, high process throughput, low residence time, low solvent inventory and high turn down ratio. The equipment provides a very high value of mass transfer coefficient and interfacial area in the annular zone because of the high level of power consumption per unit volume and separation inside the rotor due to the high g of centrifugal field. For the development of rational and reliable design procedures, it is important to understand the flow patterns in the mixer and settler zones. Computational Fluid Dynamics (CFD has played a major role in the constant evolution and improvements of this device. During the past thirty years, a large number of investigators have undertaken CFD simulations. All these publications have been carefully and critically analyzed and a coherent picture of the present status has been presented in this review paper. Initially, review of the single phase studies in the annular region has been presented, followed by the separator region. In continuation, the two-phase CFD simulations involving liquid-liquid and gas-liquid flow in the annular as well as separator regions have been reviewed. Suggestions have been made for the future work for bridging the existing knowledge gaps. In particular, emphasis has been given to the application of CFD simulations for the design of this equipment.

  16. Research on Marine Sciences under Core University Marine Science Program in the Period of 2001-2005 : The Bibliometrics Approach

    OpenAIRE

    Yoganingrum, Ambar; Suestiningtyas, Nur Tri A; Royani, Yupi; Mahmudah

    2006-01-01

    This research aims to analyze the contribution of collaboration research under Core University Marine Science Program sponsored by Japan Society for Promotion of Sciences (JSPS) over a period of 2001-2005. The data were papers of JSPS seminars. We used co-word technique of bibliometrics methods to identify the research topics. Then we calculated which topics are core and prominence levels based on method of Sujit Battarcharja and Moh’d Taiyab Rashid Khan. In-depth interviews were also carried...

  17. Research of the DC discharge of He-Ne gas mixture in hollow core fiber

    Science.gov (United States)

    Wang, Xinbing; Duan, Lian

    2013-09-01

    Since the first waveguide 0.633 μm He-Ne laser from a 20 cm length of 430 μm glass capillary was reported in 1971, no smaller waveguide gas laser has ever been constructed. Recently as the development of low loss hollow core PBG fiber, it is possible to constract a He-Ne lasers based on hollow-core PBG fibers. For the small diameter of the air hole, it is necessary to do some research to obtain glow discharge in hollow core fibers. In this paper, the experimental research of DC discharge in 200 μm bore diameter hollow core fibers was reported. Stable glow discharge was obained at varioue He-Ne mixtures from 4 Torr to 18 Torr. In order to obtain the plasma parameter of the discharge, the trace gasses of N2 and H2 were added to the He-Ne mixtures, the optical emission spectroscopy of the discharge was recorded by a PI 2750 spectroscopy with a CCD camera. The gas temperature (Tg) could be obtained by matching the simulated rovibronic band of the N2 emission with the observed spectrum in the ultraviolet region. The spectral method was also used to obtained the electron density, which is based on the analysis of the wavelength profile of the 486.13 nm Hβ line, and the electron temperature was obtain by Boltzmann plot methods. Experimental results show that it is very difficult to achieve DC discharge in bore diameter less than 50 μm, and a RF discharge method was proposed. Project supported by the National Natural Science Foundation of China (61078033).

  18. Evaluating the core damage frequency of a TRIGA research reactor using risk assessment tool software

    Energy Technology Data Exchange (ETDEWEB)

    Kamyab, Shahabeddin [School of Engineering, Shiraz University, 71348-51154 Shiraz (Iran, Islamic Republic of); Nematollahi, Mohammadreza, E-mail: mrnema@yahoo.com [School of Engineering, Shiraz University, 71348-51154 Shiraz (Iran, Islamic Republic of); Safety Research Center of Shiraz University, 71348-51154 Shiraz (Iran, Islamic Republic of)

    2011-08-15

    Highlights: {center_dot} In this study, level-I PSA is performed, to reveal and modify the weak points threatening the safe operation of a typical TRIGA reactor. {center_dot} After identification of the initiating events and developing the appropriate event trees and fault trees, by the risk assessment tool interface, the core damage frequency has been estimated to be 8.368E-6 per year of reactor operation, which meets the IAEA standards. {center_dot} The results also indicate the significant effects of the common cause failures. - Abstract: After all preventive and mitigative measures considered in the design of a nuclear reactor, the installation still represents a residual risk to the outside world. Probabilistic safety assessment (PSA) is a powerful method to survey the safety of nuclear reactors. In this study the occurrence frequency of different types of core damage states (CDS) which may potentially arise in Tehran Research Reactor (TRR) is evaluated by use of the recently developed risk assessment tool (RAT) software which has been designed and represented in the Safety Research Center of Shiraz University. RAT uses event trees and fault trees to evaluate the total final core damage frequency (CDF) through studying the frequencies of initiation events, and following their consequences has resulted in one type of the CDS. The criterion must be of the order of smaller than 1E-04 through IAEA standards for research reactors (). Results show that the total final CDF for TRR is of the order of 10{sup -6}, which meets the criterion of nuclear research reactor.

  19. Design of Annular Linear Induction Pump for High Temperature Liquid Lead Transportation

    International Nuclear Information System (INIS)

    EM(Electro Magnetic) Pump is divided into two parts, which consisted of the primary one with electromagnetic core and exciting coils, and secondary one with liquid lead flow. The main geometrical variables of the pump included core length, inner diameter and flow gap while the electromagnetic ones covered pole pitch, turns of coil, number of pole pairs, input current and input frequency. The characteristics of design variables are analyzed by electrical equivalent circuit method taking into account hydraulic head loss in the narrow annular channel of the ALIP. The design program, which was composed by using MATLAB language, was developed to draw pump design variables according to input requirements of the flow rate, developing pressure and operation temperature from the analyses. The analysis on the design of ALIP for high temperature liquid lead transportation was carried for the produce of ALIP designing program based on MATLAB. By the using of ALIP designing program, we don't have to bother about geometrical relationship between each component during detail designing process because code calculate automatically. And prediction of outputs about designing pump can be done easily before manufacturing. By running the code, we also observe and analysis change of outputs caused by changing of pump factors. It will be helpful for the research about optimization of pump outputs

  20. Quantifying the uncertainty of the annular mode time scale and the role of the stratosphere

    Science.gov (United States)

    Kim, Junsu; Reichler, Thomas

    2015-10-01

    The proper simulation of the annular mode time scale may be regarded as an important benchmark for climate models. Previous research demonstrated that this time scale is systematically overestimated by climate models. As suggested by the fluctuation-dissipation theorem, this may imply that climate models are overly sensitive to external forcings. Previous research also made it clear that calculating the AM time scale is a slowly converging process, necessitating relatively long time series and casting doubts on the usefulness of the historical reanalysis record to constrain climate models in terms of the annular mode time scale. Here, we use long control simulations with the coupled and uncoupled version of the GFDL climate model, CM2.1 and AM2.1, respectively, to study the effects of internal atmospheric variability and forcing from the lower boundary on the stability of the annular mode time scale. In particular, we ask whether a model's annular mode time scale and dynamical sensitivity can be constrained from the 50-year-long reanalysis record. We find that internal variability attaches large uncertainty to the annular mode time scale when diagnosed from decadal records. Even under the fixed forcing conditions of our long control run at least 100 years of data are required in order to keep the uncertainty in the annular mode time scale of the Northern Hemisphere to 10 %; over the Southern Hemisphere, the required length increases to 200 years. If nature's annular mode time scale over the Northern Hemisphere is similarly variable, there is no guarantee that the historical reanalysis record is a fully representative target for model evaluation. Over the Southern Hemisphere, however, the discrepancies between model and reanalysis are sufficiently large to conclude that the model is unable to reproduce the observed time scale structure correctly. The effects of ocean coupling lead to a considerable increase in time scale and uncertainty in time scale, effects which

  1. The role of fission product in whole core accidents - research in the USA

    International Nuclear Information System (INIS)

    Safety of nuclear reactors has been a central concern of the nuclear energy industry from the very beginning. This concern, and the resultant excellence of design, fabrication, and operation, aided by extensive engineered safety features, has given nuclear energy its superior record of protection of the environment and of the public health and safety. With respect to the fast reactor, it was recognized early in the programme that there exists a theoretical possibility of a core compaction leading to significant energy release. Early analysis of this problem employed a number of conservative assumptions in attempting to bound the energy release. As reactors have grown in size, the suitability of such bounding calculations has diminished, and research into hypothetical accident analysis has emphasized a more mechanistic approach. In the USA, much effort has been directed towards modeling and computer code development aimed at following the course of an accident from its initiation to its ultimate conclusion with a stable, permanently subcritical, coolable core geometry, along with considerations of post-accident heat removal and radiological consequence assessment. Throughout this effort, the potential role of fission products has been recognized and account taken of the effects of fission products in determining accident progression. It is important to recognize that reactor safety is a very diverse topic, requiring consideration of a number of factors. While the major questions of public risk appear to be related to the hypothetical core disruptive accident (HCDA), it is necessary that the probability of having such an accident be extremely low In order that acceptable public risk be demonstrated. Such a demonstration requires sound engineering design and Implementation, with high standards of reliability, inspectability, maintainability, and operation, along with the requisite quality control and assurance. Tile current approach, typified by that taken by the

  2. Instability patterns in a miscible core annular flow

    Science.gov (United States)

    D'Olce, Marguerite; Martin, Jerome; Rakotomalala, Nicole; Salin, Dominique; Talon, Laurent

    2006-11-01

    Laboratoire FAST, batiment 502, campus universitaire, 91405 Orsay Cedex (France). Experiments are performed with two miscible fluids of equal density but different viscosities. The fluids are injected co-currently and concentrically into a cylindrical pipe. The so-obtained base state is an axisymmetric parallel flow, for which the ratio of the flow rates of the two fluids monitors the relative amount (and so the radius) of the fluids. Depending on this relative amount and on the total flow rate of the fluids, unstable axisymmetric patterns such as mushrooms and pearls are observed. We delineate the diagram of occurrence of the two patterns and characterize the instabilities.

  3. Annular bilayer magnetoelectric composites: theoretical analysis.

    Science.gov (United States)

    Guo, Mingsen; Dong, Shuxiang

    2010-01-01

    The laminated bilayer magnetoelectric (ME) composites consist of magnetostrictive and piezoelectric layers are known to have giant ME coefficient due to the high coupling efficiency in bending mode. In our previous report, the bar-shaped bilayer composite has been investigated by using a magnetoelectric-coupling equivalent circuit. Here, we propose an annular bilayer ME composite, which consists of magnetostrictive and piezoelectric rings. This composite has a much lower resonance frequency of bending mode compared with its radial mode. In addition, the annular bilayer ME composite is expected to respond to vortex magnetic field as well as unidirectional magnetic field. In this paper, we investigate the annular bilayer ME composite by using impedance-matrix method and predict the ME coefficients as a function of geometric parameters of the composites. PMID:20178914

  4. Radiation Characterization Summary: ACRR Central Cavity Free-Field Environment with the 32-Inch Pedestal at the Core Centerline (ACRR-FF-CC-32-cl).

    Energy Technology Data Exchange (ETDEWEB)

    Vega, Richard Manuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Parma, Edward J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Naranjo, Gerald E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lippert, Lance L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Vehar, David W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Griffin, Patrick J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-08-01

    This document presents the facilit y - recommended characteri zation o f the neutron, prompt gamma - ray, and delayed gamma - ray radiation fields in the Annular Core Research Reactor ( ACRR ) for the cen tral cavity free - field environment with the 32 - inch pedestal at the core centerline. The designation for this environmen t is ACRR - FF - CC - 32 - cl. The neutron, prompt gamma - ray , and delayed gamma - ray energy spectra , uncertainties, and covariance matrices are presented as well as radial and axial neutron and gamma - ray fluence profiles within the experiment area of the cavity . Recommended constants are given to facilitate the conversion of various dosimetry readings into radiation metrics desired by experimenters. Representative pulse operations are presented with conversion examples . Acknowledgements The authors wish to th ank the Annular Core Research Reactor staff and the Radiation Metrology Laboratory staff for their support of this work . Also thanks to David Ames for his assistance in running MCNP on the Sandia parallel machines.

  5. MTR research reactor core behavior under a loss of shutdown heat removal

    International Nuclear Information System (INIS)

    Full text of publication follows: Introduction: Heat removal during operation of medium power research reactors is assumed to be safely performed by forced convection and the adequate removal of residual decay heat after reactor shutdown need to continue forced convection removal for a certain period of time when the operating power before shut-down is above a certain power level. This is among the requirement for the overall safety of research reactor operation. Objective: The purpose of the present work is: - to estimate the maximum temperature in the core and to investigate the minimum power operating level before shutdown that needs a continuation forced convection after shutdown; - to evaluate occurrence of cladding damage following a shutdown reactor without forced convection Problem: The simulation process is undertaken using the RELAP5/Mod 3.2 code system. The IAEA 10 MW benchmark core which is a representative of medium pool type MTR research reactors was chosen here in order to investigate the cladding maximum fuel temperature without forced heat removal after shutdown of the reactor that was operating at different powers up to 10 MW. Nodalization: The benchmark core consists of 25 fuels elements placed in a 5 x 5 gird placed within pool filled by 9 m of light water. The primary loop is represented by pumps, pipes and heat exchangers. Each of the 25 fuel elements is represented individually. Results: The simulation process has shown that the cladding maximum temperature did not reach the melting point for aluminum (660 deg. C) but void is expected to be produced in the hot channels. Hence, the loss of forced heat removal after reactor scram did not induce any melting of the cladding by much deeper investigation may be undertaken because presence of void in channels could enhance corrosion phenomena and may induce some fission products release in the pools following localized fuel rupture due to corrosion. (authors)

  6. Development of Dual Cooled Annular Fuel Temperature Analysis Program

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yong Sik; Shin, C. H.; Bang, J. G.; Kim, D. H.; Kim, S. K.; Lim, I. S.; Koo Yang Hyun [KAERI, Daejeon (Korea, Republic of)

    2010-09-15

    To calculate the temperature distribution of dual cooled annular fuel, the DUOS program has been developed. Various thermal hydraulic models to determine the inner channel and outer channel flow distribution were established based on equal pressure drop condition at the top of fuel rod. The effect of gap width change was considered by employing thermal deformation model of pellet and claddings. Heat conduction model in the pellet was solved by finite difference method to consider burnup and power difference according to pellet radius. Pellet temperature model was validated by comparison with calculated temperature profile, which was determined by analytical solution of heat conduction equation under controlled input condition. Accuracy of thermal hydraulic models of DUOS were validated by core sub-channel analysis code MATRA-AF. Coolant bulk temperature of inner/outer channel and pressure drop prediction results of DUOS program show good agreement with that of MATRA-AF. Further models should be added in DUOS program to describe dual cooled annular fuel in-pile behavior, but basic thermal analysis structure has been established successfully

  7. Development of Dual Cooled Annular Fuel Temperature Analysis Program

    International Nuclear Information System (INIS)

    To calculate the temperature distribution of dual cooled annular fuel, the DUOS program has been developed. Various thermal hydraulic models to determine the inner channel and outer channel flow distribution were established based on equal pressure drop condition at the top of fuel rod. The effect of gap width change was considered by employing thermal deformation model of pellet and claddings. Heat conduction model in the pellet was solved by finite difference method to consider burnup and power difference according to pellet radius. Pellet temperature model was validated by comparison with calculated temperature profile, which was determined by analytical solution of heat conduction equation under controlled input condition. Accuracy of thermal hydraulic models of DUOS were validated by core sub-channel analysis code MATRA-AF. Coolant bulk temperature of inner/outer channel and pressure drop prediction results of DUOS program show good agreement with that of MATRA-AF. Further models should be added in DUOS program to describe dual cooled annular fuel in-pile behavior, but basic thermal analysis structure has been established successfully

  8. Neutronic analysis for core conversion (HEU-LEU) of Pakistan research reactor-2 (PARR-2)

    International Nuclear Information System (INIS)

    Neutronic analyses for the core conversion of Pakistan research reactor-2 (PARR-2) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel has been performed. Neutronic model has been verified for 90.2% enriched HEU fuel (UAl4-Al). For core conversion, UO2 fuel was chosen as an appropriate fuel option because of higher uranium density. Clad has been changed from aluminum to zircalloy-4. Uranium enrichment of 12.6% has been optimized based on the design basis criterion of excess reactivity 4 mk in miniature neutron source reactor (MNSR). Lattice calculations for cross-section generation have been performed utilizing WIMS while core modeling was carried out employing three dimensions option of CITATION. Calculated neutronic parameters were compared for HEU and LEU fuels. Comparison shows that to get same thermal neutron flux at inner irradiation sites, reactor power has to be increased from 30 to 33 kW for LEU fuel. Reactivity coefficients calculations show that doppler and void coefficient values of LEU fuel are higher while moderator coefficient of HEU fuel is higher. It is concluded that from neutronic point of view LEU fuel UO2 of 12.6% enrichment with zircalloy-4 clad is suitable to replace the existing HEU fuel provided that dimensions of fuel pin and total number of fuel pins are kept same as for HEU fuel

  9. Development of core technology for research reactors using plate type fuels

    International Nuclear Information System (INIS)

    Around 250 research reactors are under operation over the world. However, about 2/3 have been operated more than 30 years and demands for replacements are expected in the near future. The number of expected units is around 110, and around 55 units from 40 countries will be expected to be bid in the world market. In 2007, Netherlands started international bidding process to construct a new 80MW RR (named PALLAS) with the target of commercial operation in 2016, which will replace the existing HFR(45MW). KAERI consortium has been participated in that bid. Most of RRs use plate type fuels as a fuel assembly, Be and Graphite as a reflector. On the other hand, in Korea, the KAERI is operating the HANARO, which uses a rod type fuel assembly and heavy water as a reflector. Hence, core technologies for RRs using plate type fuels are in short. Therefore, core technologies should be secured for exporting a RR. In chapter 2, the conceptual design of PALLAS which use plate type fuels are described including core, cooling system and connected systems, layout of general components. Experimental verification tests for the plate type fuel and second shutdown system and the code verification for nuclear design are explained in Chapter 3 and 4, respectively

  10. Replacement of the Core Beryllium Reflector in the SAFARI-1 Research Reactor

    International Nuclear Information System (INIS)

    The SAFARI-1 Research Reactor is a 20 MW high flux MTR and has been continuously operational for more than 46 years. In this period, the core beryllium reflector had never been replaced. An ageing management action to replace the reflector received priority due to the risks involved with failure or deformation of elements. This paper elaborates on the actions taken to replace the old and manage the new reflector. To this extent a reflector replacement procedure, backed up by core neutronic calculations and a test plan, was developed for the safe replacement of the reflector. A reflector management programme will ensure that records of reflector elements are kept and used to optimally manage usage of every element. Due to the historic nature of reflector utilisation in the SAFARI-1 core, deformation of the elements was unavoidable. These deformations will be monitored in the management programme for the new reflector. Deformation measurement of the old reflector is planned and could yield interesting comparisons with analytical results. The action plan for final disposal of the old reflector, although still in development, is also mentioned in this paper. (author)

  11. Optimization of the core configuration design using a hybrid artificial intelligence algorithm for research reactors

    International Nuclear Information System (INIS)

    To successfully carry out material irradiation experiments and radioisotope productions, a high thermal neutron flux at irradiation box over a desired life time of a core configuration is needed. On the other hand, reactor safety and operational constraints must be preserved during core configuration selection. Two main objectives and two safety and operational constraints are suggested to optimize reactor core configuration design. Suggested parameters and conditions are considered as two separate fitness functions composed of two main objectives and two penalty functions. This is a constrained and combinatorial type of a multi-objective optimization problem. In this paper, a fast and effective hybrid artificial intelligence algorithm is introduced and developed to reach a Pareto optimal set. The hybrid algorithm is composed of a fast and elitist multi-objective genetic algorithm (GA) and a fast fitness function evaluating system based on the cascade feed forward artificial neural networks (ANNs). A specific GA representation of core configuration and also special GA operators are introduced and used to overcome the combinatorial constraints of this optimization problem. A software package (Core Pattern Calculator 1) is developed to prepare and reform required data for ANNs training and also to revise the optimization results. Some practical test parameters and conditions are suggested to adjust main parameters of the hybrid algorithm. Results show that introduced ANNs can be trained and estimate selected core parameters of a research reactor very quickly. It improves effectively optimization process. Final optimization results show that a uniform and dense diversity of Pareto fronts are gained over a wide range of fitness function values. To take a more careful selection of Pareto optimal solutions, a revision system is introduced and used. The revision of gained Pareto optimal set is performed by using developed software package. Also some secondary operational

  12. An intrinsically safe facility for forefront research and training on nuclear technologies — Core design

    Science.gov (United States)

    Viberti, C. M.; Ricco, G.

    2014-04-01

    The core of a subcritical, low-power research reactor in a lead matrix has been designed using the MCNPX code. The main parameters, like geometry, material composition in the fuel assembly and reflector size, have been optimized for a k eff ˜ 0.95 and a thermal power around 200 Kw. A 70 Mev, 1 mA proton beam incident on a beryllium target has been assumed as neutron source and the corresponding thermal power distribution and neutron fluxes in the reactor have been simulated.

  13. Stitching algorithm for annular subaperture interferometry

    Institute of Scientific and Technical Information of China (English)

    Xi Hou; Fan Wu; Li Yang; Shibin Wu; Qiang Chen

    2006-01-01

    @@ Annular subaperture interferometry (ASI) has been developed for low cost and flexible test of rotationally symmetric aspheric surfaces, in which accurately combining the subaperture measurement data corrupted by misalignments and noise into a complete surface figure is the key problem. By introducing the Zernike annular polynomials which are orthogonal over annulus, a method that eliminates the coupling problem in the earlier algorithm based on Zernike circle polynomials is proposed. Vector-matrix notation is used to simplify the description and calculations. The performance of this reduction method is evaluated by numerical simulation. The results prove this method with high precision and good anti-noise capability.

  14. Research and application of active hollow core slabs in building systems for utilizing low energy sources

    International Nuclear Information System (INIS)

    Highlights: • A review on the development and modeling of active hollow core slab is presented. • The applications and performance evaluation of the slab in building are reviewed. • Finite element or finite difference method is often used in multidimensional model. • Performance evaluations of building using active slabs for ventilation are limited. • More works on the active hollow core slab are worthwhile. - Abstract: The society and the building professionals have paid much concern in recent years on building energy efficiency and the development and applications of low energy technologies for buildings/green buildings allowing the elimination, or at least reduction of dependence on electricity or fossil fuel while maintaining acceptable indoor environment. Utilizations of favorable diurnal temperature difference and ground thermal source for air conditioning are among these low energy technologies. Utilization of the hollow cores in the prefabricated slab for ventilation and the mass of the slab for thermal storage is widely used in building systems in Europe by exploiting the low energy source of the ambient air. These hollow core slabs aim at enlarging the heat transfer surface between the slab mass and the air in the core, which permits substantial heat flows even for relatively small temperature differences. This, in turn, allows the use of low energy cooling or heating sources, such as the ground, outside air or recovered process heat. In this paper, we present a comprehensive review of the research and application of active hollow core slabs in building systems for utilizing low energy sources. The principle and development of active hollow core slabs in building systems for leveling the indoor temperature fluctuation by ventilation air passing the cores are described. Calculation models of the active hollow core concrete slab as well as the practical applications and performance evaluation of the slab applied in building systems for air

  15. Analysis and core-life calculation of 3 MW Triga Mark II research reactor including effects of central thimble modification

    International Nuclear Information System (INIS)

    The principal objective of this study was to formulate an effective optimal fuel management strategy for TRIGA MARK II research reactor at AERE, Savar. The core management study has been performed by utilizing four basic types of information calculated for the reactor: criticality, power peaking, neutron flux and burnup calculation. Reshuffling at 20,000 MWh step gives the longest core life of the reactor which is 64500 MWh. Central thimble modification altered the shape of the flux which increased the core reactivity by c 12 and the core-life by 500 MWh. Besides, the study gives valuable insight into the behaviour of the reactor

  16. Development of a 3-dimensional calculation model of the Danish research reactor DR3 to analyse a proposal to a new core design called ring-core

    International Nuclear Information System (INIS)

    A 3-dimensional calculation model of the Danish research reactor DR3 has been developed. Demands of a more effective utilization of the reactor and its facilities has required a more detailed calculation tool than applied so far. A great deal of attention has been devoted to the treatment of the coarse control arms. The model has been tested against measurements with satisfying results. Furthermore the model has been used to analyse a proposal to a new core design called ring-core where 4 central fuel elements are replaced by 4 dummy elements to increase the thermal flux in the center of the reactor. (author)

  17. Safety analysis for core conversion (from HEU to LEU) of Pakistan research reactor-2 (PARR-2)

    Energy Technology Data Exchange (ETDEWEB)

    Bokhari, Ishtiaq Hussain, E-mail: ishtiaq@pinstech.org.p [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan); Pervez, Showket [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)

    2010-01-15

    PARR-2 (Pakistan Research Reactor-2), an MNSR (Miniature Neutron Source Reactor) is to be converted from HEU (High Enriched Uranium) to LEU (Low Enriched Uranium) fuel, along with all current MNSRs in various other countries. The purpose of conversion is to minimize the use of HEU for non-proliferation of high-grade nuclear fuel. The present report presents thermal hydraulic and safety analyses of PARR-2 using existing HEU fuel as well as proposed LEU fuel. Presently, the core is comprised of 90.2% enriched UAl{sub 4}-Al fuel. There are 344 fuel pins of 5.5 mm diameter. The core has a total of 994.8 g of U{sup 235}. Standard computer code PARET/ANL (version 1992) was employed to perform steady-state and transient analyses. Various parameters were computed, which included: coolant outlet, maximum clad surface and maximum fuel centerline temperatures; and peak power and corresponding peak core temperatures resulting from a transient initiated by 4 mK positive reactivity insertion. Results were compared with the reported data in Final Safety Analysis Report (FSAR). It was found that the PARET results were in reasonable agreement with the manufacturer's results. Calculations were also carried out for the proposed LEU core with two suggested fuel pin sizes (5.5 mm and 5.1 mm diameter with 12.6% and 12.3% enrichment, respectively). Comparison of the LEU results with the existing HEU fuel has been made and discussed.

  18. Optimal Thrust Vectoring for an Annular Aerospike Nozzle Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Recent success of an annular aerospike flight test by NASA Dryden has prompted keen interest in providing thrust vector capability to the annular aerospike nozzle...

  19. Flow visualization study of inverted annular flow of post dryout heat transfer region

    International Nuclear Information System (INIS)

    The inverted annular flow is important in the area of LWR accident analysis in terms of the maximum cladding temperature and effectiveness of the emergency core cooling. However, the inverted annular flow thermal-hydraulics is not well understood due to its special heat transfer condition of film boiling. In view of this, the inverted flow is studied in detail experimentally. A new experimental apparatus has been constructed in which film boiling heat transfer can be established in a transparent test section. Data on liquid core stability, core break-up mechanism, and dispersed-core liquid slug and droplet sizes are obtained using F 113 as a test fluid. Both high speed movies and flash photographs are used. The inlet section consists of specially designed coaxial nozzles for gas and liquid such that the ideal inverted annular flow can be generated. The roll wave formation, droplet entrainment from wave crests, agitated sections with large interfacial areas, classical sinuous jet instability, jet break-up into multiple liquid ligaments and drop formation from liquid ligaments have been observed in detail. (orig.)

  20. Safe operation of TRIGA reactor in the situation of LEU-HEU core conversion

    International Nuclear Information System (INIS)

    Romanian TRIGA reactor was commissioned in 1980. The location of the research institute is Pitesti, 100 Km west of Bucharest. In fact there are two independent cores sharing the same pool. There are a 14 MW Steady State Reactor (SSR), high flux, and materials testing reactor and an Annular Core Pulsing Reactor (ACPR). The SSR reactor is a forced convection reactor cooled via a primary circuit with 4 pumps and 3 heat exchangers. The ACPR is natural convection reactor cooled by the pool water. The characteristics of the two reactors are presented. The reactor core configuration is shown as well as the original start-up core configuration. Fuel management of TRIGA steady state core allows obtaining the requested fluxes for experimental purposes in safe operation condition. One can firmly state that the present operation of the reactor and the HEU-LEU (High Enriched Uranium - Low Enriched Uranium), core conversion fully respect the provisions of the National Regulatory Body and IAEA. (authors)

  1. Parametric evaluation of mixed (low and high enriched) fuel core for a swimming pool type research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bakhtyar, S.; Iqbal, M.; Israr, M.; Pervez, S.; Salahuddin, A. [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)

    2004-07-01

    A study has been carried out to evaluate the performance of a swimming pool type research reactor core comprised of mixed (low and high enriched) uranium fuel. The study includes the calculations of core reactivity, worth of control rods and core criticality at the Beginning Of Life (BOL) of the core and for two operating conditions Cold Zero Power (CZP) and Hot Full Power (HFP). Further, to ensure safe and stable operation of the core from nuclear design point of view, average power densities in the fuel region, power peaking factors, axial power distribution in the hot channel and reactivity feed back coefficients have also been calculated. Two group fluxes have also been determined at different irradiation locations. All these calculations have been carried out employing reactor lattice code WIMS-D14 and reactor analysis code CITATION The calculated results show reasonably good agreement with the quoted operational data of the previous LEU cores. (Author)

  2. Parametric evaluation of mixed (low and high enriched) fuel core for a swimming pool type research reactor

    International Nuclear Information System (INIS)

    A study has been carried out to evaluate the performance of a swimming pool type research reactor core comprised of mixed (low and high enriched) uranium fuel. The study includes the calculations of core reactivity, worth of control rods and core criticality at the Beginning Of Life (BOL) of the core and for two operating conditions Cold Zero Power (CZP) and Hot Full Power (HFP). Further, to ensure safe and stable operation of the core from nuclear design point of view, average power densities in the fuel region, power peaking factors, axial power distribution in the hot channel and reactivity feed back coefficients have also been calculated. Two group fluxes have also been determined at different irradiation locations. All these calculations have been carried out employing reactor lattice code WIMS-D14 and reactor analysis code CITATION The calculated results show reasonably good agreement with the quoted operational data of the previous LEU cores. (Author)

  3. Two-phase flow characteristic of inverted bubbly, slug, and annular flow in post-critical heat flux region

    International Nuclear Information System (INIS)

    Inverted annular flow can be visualized as a liquid jet-like core surrounded by a vapor annulus. While many analytical and experimental studies of heat transfer in this regime have been performed, there is very little understanding of the basic hydrodynamics of the post-critical heat flux (CHF) flow field. However, a recent experimental study was done that was able to successfully investigate the effects of various steady-state inlet flow parameters on the post-CHF hydrodynamics of the film boiling of a single phase liquid jet. This study was carried out by means of a visual photographic analysis of an idealized single phase core inverted annular flow initial geometry (single phase liquid jet core surrounded by a coaxial annulus of gas). In order to extend this study, a subsequent flow visualization of an idealized two-phase core inverted annular flow geometry (two-phase central jet core, surrounded by a coaxial annulus of gas) was carried out. The objective of this second experimental study was to investigate the effect of steady-state inlet, pre-CHF two-phase jet core parameters on the hydrodynamics of the post-CHF flow field. In actual film boiling situations, two-phase flows with net positive qualities at the CHF point are encountered. Thus, the focus of the present experimental study was on the inverted bubbly, slug, and annular flow fields in the post dryout film boiling region. Observed post dryout hydrodynamic behavior is reported. A correlation for the axial extent of the transition flow pattern between inverted annular and dispersed droplet flow (the agitated regime) is developed. It is shown to depend strongly on inlet jet core parameters and jet void fraction at the dryout point

  4. Two-phase flow characteristic of inverted bubbly, slug and annular flow in post-critical heat flux region

    International Nuclear Information System (INIS)

    Inverted annular flow can be visualized as a liquid jet-like core surrounded by a vapor annulus. While many analytical and experimental studies of heat transfer in this regime have been performed, there is very little understanding of the basic hydrodynamics of the post-CHF flow field. However, a recent experimental study was done that was able to successfully investigate the effects of various steady-state inlet flow parameters on the post-CHF hydrodynamics of the film boiling of a single phase liquid jet. This study was carried out by means of a visual photographic analysis of an idealized single phase core inverted annular flow initial geometry (single phase liquid jet core surrounded by a coaxial annulus of gas). In order to extend this study, a subsequent flow visualization of an idealized two-phase core inverted annular flow geometry (two-phase central jet core, surrounded by a coaxial annulus of gas) was carried out. The objective of this second experimental study was to investigate the effect of steady-state inlet, pre-CHF two-phase jet core parameters on the hydrodynamics of the post-CHF flow field. In actual film boiling situations, two-phase flows with net positive qualities at the CHF point are encountered. Thus, the focus of the present experimental study was on the inverted bubbly, slug, and annular flow fields in the post dryout film boiling region. Observed post dryout hydrodynamic behavior is reported. A correlation for the axial extent of the transition flow pattern between inverted annular and dispersed droplet flow (the agitated regime) is developed. It is shown to depend strongly on inlet jet core parameters and jet void fraction at the dryout point. 45 refs., 9 figs., 4 tabs

  5. Progress in the neutronic core conversion (HEU-LEU) analysis of Ghana research reactor-1.

    Energy Technology Data Exchange (ETDEWEB)

    Anim-Sampong, S.; Maakuu, B. T.; Akaho, E. H. K.; Andam, A.; Liaw, J. J. R.; Matos, J. E.; Nuclear Engineering Division; Ghana Atomic Energy Commission; Kwame Nkrumah Univ. of Science and Technology

    2006-01-01

    The Ghana Research Reactor-1 (GHARR-1) is a commercial version of the Miniature Neutron Source Reactor (MNSR) and has operated at different power levels since its commissioning in March 1995. As required for all nuclear reactors, neutronic and thermal hydraulic analysis are being performed for the HEU-LEU core conversion studies of the Ghana Research Reactor-1 (GHARR-1) facility, which is a commercial version of the Miniature Neutron Source Reactor (MNSR). Stochastic Monte Carlo particle transport methods and tools (MCNP4c/MCNP5) were used to fine-tune a previously developed 3-D MCNP model of the GHARR-1 facility and perform neutronic analysis of the 90.2% HEU reference and candidate LEU (UO{sub 2}, U{sub 3}Si{sub 2}, U-9Mo) fresh cores with varying enrichments from 12.6%-19.75%. In this paper, the results of the progress made in the Monte Carlo neutronic analysis of the HEU reference and candidate LEU fuels are presented. In particular, a comparative performance assessment of the LEU with respect to neutron flux variations in the fission chamber and experimental irradiation channels are highlighted.

  6. Research reactor core conversion guidebook. V. 5: Operations (Appendices L-N)

    International Nuclear Information System (INIS)

    Volume 5 consists of detailed Appendices L-N, which contain a variety of useful information on the operation of research reactors with reduced enrichment fuels. Summaries of these appendices can be found in Chapters 12-14 of Volume 1 of this guidebook. Appendix L contains a summary of necessary and recommended experiments for reactor startup. Appendix M provides information on the procedures and experiences of several reactor operators with both mixed and full cores with reduced enrichment fuels. Appendix N contains information on transportation of both fresh and spent fuel elements, on spent fuel storage, and on the US Department of Energy's receipt and financial settlement provisions for nuclear research reactor fuels. Refs, figs and tabs

  7. Research reactor in-core fuel management optimisation using the multiobjective cross-entropy method

    International Nuclear Information System (INIS)

    The in-core fuel management optimisation (ICFMO) problem has been studied for several decades. Very little research has, however, been aimed at multiobjective optimisation involving the fundamental notion of Pareto optimality. In this paper, the recently developed multiobjective optimisation using the cross-entropy method (MOO CEM) algorithm is applied to a multiobjective ICFMO problem for the first time. A derivation of the MOO CEM algorithm is presented for ICFMO, along with a constraint handling technique. The algorithm is applied to a biobjective test problem for the SAFARI-1 nuclear research reactor. The Pareto set approximated by the algorithm is compared to solutions obtained by typical operational reload strategies. The results indicate that the MOO CEM algorithm for multiobjective ICFMO is a robust and efficient method which is able to obtain a good spread of trade-off solutions. The method may therefore greatly aid in the decision making of a reactor operator tasked with designing reload configurations. (author)

  8. Thermohydraulic and mechanical analysis of the research reactor Munich II Compact-Core

    International Nuclear Information System (INIS)

    The new research reactor Munich II (Forschungsreaktor Muenchen II, FRM-II), which is under construction at the Technical University of Munich, Germany, contains a compact reactor core consisting of one single fuel element, assembled by two concentric tubes between which 113 involutely bent fuel plates are located rotationally symmetric. In order to perform the hydraulic and mechanical testing of the FRM-II fuel element, two test facilities have been built at the Department for Nuclear and New Energy Systems of the Ruhr University Bochum. The first mocks up the central region of the reactor coolant system of the FRM-II in a 1:1 scale with emphasis on the fuel element and the inflow and discharge section in order to enable the analysis of the FRM-II core. In the course of the testing the vibration behaviour and the flow resistance of the core were investigated. Likewise start-up and shut down tests of the main pump unit were simulated and the flow profile at the outlet of the element as well as the flow division inside the core were determined. Furthermore an endurance test lasting 60 days (equivalent to 12 operating cycles) was performed, too. Tests including blockages of parts of the reactor cooling system cross section at the core entrance sieve proved the efficiency of the cooling capacity. No major resonances occurred during operation and an endurance test neither showed any incidents nor irregularities. In order to investigate the concept of the decay heat removal in the FRM-II a second test facility was built. This facility simulates the thermohydraulic conditions in one cooling channel of the FRM-II by means of an electrically heated test section, which enables different operating conditions of the decay heat removal system as well as enhanced safety investigations. In the FRM-II the decay heat, which is produced after a shutdown, is removed by means of decay heat removal pumps, which maintain a downward flow in the fuel element for at least three hours

  9. A New Annular Shear Piezoelectric Accelerometer

    DEFF Research Database (Denmark)

    Liu, Bin; Kriegbaum, B.

    2000-01-01

    This paper describes the construction and performance of a recently introduced Annular Shear piezoelectric accelerometer, Type 4511. The design has insulated and double-shielded case. The accelerometer housing is made of stainless steel, AISI 316L. Piezoceramic PZ23 is used. The seismic mass...

  10. Dynamic Response of Three-Layered Annular Plate with Imperfections

    Directory of Open Access Journals (Sweden)

    Pawlus Dorota

    2015-02-01

    Full Text Available This paper presents the imperfection sensitivity of annular plate with three-layered structure. The plate composed of thin elastic facings and a thicker elastic core is loaded in facing plane. The classical issue of a three-layered plate was solved for dynamic deflection problem using the approximation methods: orthogonalization and finite difference. The solution includes the axisymmetric and asymmetric plate modes of the dynamic stability loss. The evaluation of the rate of plate sensitivity to imperfection of plate preliminary geometry has been enriched by the analysis of plate models built of finite elements. The ABAQUS program has been used. The numerous calculation results in the form of deflection characteristics, buckling modes, values of critical parameters create the view of response of dynamic plate structure with different rate of imperfection and linear in time loading growth, too.

  11. MULTI-CORE AND OPTICAL PROCESSOR RELATED APPLICATIONS RESEARCH AT OAK RIDGE NATIONAL LABORATORY

    International Nuclear Information System (INIS)

    High-speed parallelization of common tasks holds great promise as a low-risk approach to achieving the significant increases in signal processing and computational performance required for next generation innovations in reconfigurable radio systems. Researchers at the Oak Ridge National Laboratory have been working on exploiting the parallelization offered by this emerging technology and applying it to a variety of problems. This paper will highlight recent experience with four different parallel processors applied to signal processing tasks that are directly relevant to signal processing required for SDR/CR waveforms. The first is the EnLight Optical Core Processor applied to matched filter (MF) correlation processing via fast Fourier transform (FFT) of broadband Dopplersensitive waveforms (DSW) using active sonar arrays for target tracking. The second is the IBM CELL Broadband Engine applied to 2-D discrete Fourier transform (DFT) kernel for image processing and frequency domain processing. And the third is the NVIDIA graphical processor applied to document feature clustering. EnLight Optical Core Processor. Optical processing is inherently capable of high-parallelism that can be translated to very high performance, low power dissipation computing. The EnLight 256 is a small form factor signal processing chip (5x5 cm2) with a digital optical core that is being developed by an Israeli startup company. As part of its evaluation of foreign technology, ORNL's Center for Engineering Science Advanced Research (CESAR) had access to a precursor EnLight 64 Alpha hardware for a preliminary assessment of capabilities in terms of large Fourier transforms for matched filter banks and on applications related to Doppler-sensitive waveforms. This processor is optimized for array operations, which it performs in fixed-point arithmetic at the rate of 16 TeraOPS at 8-bit precision. This is approximately 1000 times faster than the fastest DSP available today. The optical core

  12. 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.

  13. Selection and benchmarking of computer codes for research reactor core conversions

    International Nuclear Information System (INIS)

    A group of computer codes have been selected and obtained from the Nuclear Energy Agency (NEA) Data Bank in France for the core conversion study of highly enriched research reactors. ANISN, WIMSD-4, MC2, COBRA-3M, FEVER, THERMOS, GAM-2, CINDER and EXTERMINATOR were selected for the study. For the final work THERMOS, GAM-2, CINDER and EXTERMINATOR have been selected and used. A one dimensional thermal hydraulics code also has been used to calculate temperature distributions in the core. THERMOS and CINDER have been modified to serve the purpose. Minor modifications have been made to GAM-2 and EXTERMINATOR to improve their utilization. All of the codes have been debugged on both CDC and IBM computers at the University of IL. IAEA 10 MW Benchmark problem has been solved. Results of this work has been compared with the IAEA contributor's results. Agreement is very good for highly enriched fuel (HEU). Deviations from IAEA contributor's mean value for low enriched fuel (LEU) exist but they are small enough in general. Deviation of keff is about 0.5% for both enrichments at the beginning of life (BOL) and at the end of life (EOL). Flux ratios deviate only about 1.5% from IAEA contributor's mean value. (author)

  14. Lithology, fracture intensity, and fracture filling of drill core from Chalk River research area, Ontario

    International Nuclear Information System (INIS)

    In 1977, 1978, and 1979, nine inclined cored boreholes, ranging in length from 113 to 704 m, were drilled in the Chalk River Research Area in order to define the geological subsurface characteristics of the rock mass at several selected test areas. A total of 2,458 metres of NQ-3 and HQ-3 core was obtained from the nine boreholes. Orthogneiss was the most predominant rock type intersected by the boreholes. Pyroxenite, amphibolite, metagabbro and dykes of diabase, pegmatite and aplite were also encountered. The crosscutting relationships and textures within the rocks indicate that the relative ages of the rock units, from youngest to oldest, are diabase; aplite and pegmatite dykes with no defined fabric; pyroxenite; meta-ferrogabbro; amphibolite; aplite and pegmatite dykes and pegmatite pods with a defined fabric; and orthogneiss. Textural characteristics and mineral assemblages indicate that the orthogneisses in the Chalk River Area are a product of regional, medium to high-grade metamorphism and belong to the upper amphibilite to granulite facies. A total of 35,597 fractures (an average of 14.5 fractures per metre) was observed in the core. Brecciated zones and open fractures were noted in the core from all of the boreholes, and major faults were identified in four of the nine boreholes. Nearly all of the fractures have a thickness between 0.4 and 1.2 mm and contain one or more types of filling. Chlorite and calcite are the most common types of filling. Epidote, hematite, clays, sulphides, talc, sericite, and rock fragments also occur in the fractures. The crosscutting relationships between fractures and the sequence of filling layers within the fractures indicate that several episodes of fracturing have occurred and that fractures containing more than one filling have probably been reactivated. A comparison of the geological logs from one of the boreholes with natural gamma, neutron-neutron and magnetic susceptibility logs indicates that certain rock types and

  15. Laser Doppler velocimeter measurements and laser sheet imaging in an annular combustor model. M.S. Thesis, Final Report

    Science.gov (United States)

    Dwenger, Richard Dale

    1995-01-01

    An experimental study was conducted in annular combustor model to provide a better understanding of the flowfield. Combustor model configurations consisting of primary jets only, annular jets only, and a combination of annular and primary jets were investigated. The purpose of this research was to provide a better understanding of combustor flows and to provide a data base for comparison with computational models. The first part of this research used a laser Doppler velocimeter to measure mean velocity and statistically calculate root-mean-square velocity in two coordinate directions. From this data, one Reynolds shear stress component and a two-dimensional turbulent kinetic energy term was determined. Major features of the flowfield included recirculating flow, primary and annular jet interaction, and high turbulence. The most pronounced result from this data was the effect the primary jets had on the flowfield. The primary jets were seen to reduce flow asymmetries, create larger recirculation zones, and higher turbulence levels. The second part of this research used a technique called marker nephelometry to provide mean concentration values in the combustor. Results showed the flow to be very turbulent and unsteady. All configurations investigated were highly sensitive to alignment of the primary and annular jets in the model and inlet conditions. Any imbalance between primary jets or misalignment of the annular jets caused severe flow asymmetries.

  16. Management of Periocular Granuloma Annulare Using Topical Dapsone

    Science.gov (United States)

    Patel, Mayha; Shitabata, Paul; Horowitz, David

    2015-01-01

    Granuloma annulare is a disease characterized by granulomatous inflammation of the dermis. Localized granuloma annulare may resolve spontaneously, while generalized granuloma annulare may persist for decades. The authors present the case of a 41-year-old Hispanic man with a two-week history of periocular granuloma annulare. Due to previously reported success in the use of systemic dapsone for the treatment of granuloma annulare, and the periocular proximity of the patient’s lesion, topical dapsone was used for treatment. Various additional therapies for the management of granuloma annulare have been reported, such as topical and systemic steroids, isotretinoin, pentoxifylline, cyclosporine, Interferon gamma, potassium iodide, nicotinamide, niacinamide, salicylic acid, fumaric acid ester, etanercept, infliximab, and hydroxychloroquine. Additional clinical trials are necessary to further evaluate the effectiveness of topical dapsone in the management of granuloma annulare. PMID:26203321

  17. THACT-RR, Analysis of Thermal Hydraulics Transients in Research Reactor Core

    International Nuclear Information System (INIS)

    1 - Description of program or function: A Computer Program for Analyzing Thermal-Hydraulics transients in Research Reactors. THACT-RR is a channel code. It analyses the transient response of a research reactor core after power excursions or coolant flow and/or coolant temperature changes. The THACT-RR code provides a homogeneous one-dimensional compressible fluid flow capability with an optional voiding model that estimates the void produced by sub-cooled boiling. It allows flow reversal and sub-cooled nucleate boiling. It also includes a selection of flow instability, departure from nucleate boiling, single and two-phase heat transfer correlations, and a physical properties library adapted to pressures, temperatures, and flow rates encountered in research reactors. 2 - Method of solution: The conservation laws are solved by the method of Characteristics coupled with an implicit finite difference technique to insure stability and convergence of the numerical scheme. The conduction equation is solved by an implicit finite difference method. 3 - Restrictions on the complexity of the problem: The code is not adapted to very fast transient problems

  18. Activities of research-reactor-technology project in FNCA from FY2005 to FY2007. Sharing neutronics calculation technique for core management and utilization of research reactors

    International Nuclear Information System (INIS)

    RRT project (Research-Reactor-Technology Project) was carried out with the theme of 'sharing neutronics calculation technique for core management and utilization of research reactors' in the framework of FNCA (Forum for Nuclear Cooperation in Asia) from FY2005 to FY2007. The objective of the project was to improve and equalize the level of neutronics calculation technique for the reactor core management among participating countries to assure the safe and stable operation of research reactors and the promotion of the effective utilization. Neutronics calculation codes, namely SRAC code system and MVP code, were adopted as common codes. Participating countries succeeded in applying the common codes to analyzing the core of each domestic research reactor. Some participating countries succeeded in applying the common codes to analyzing for utilization of own research reactors. Activities of RRT project have improved and equalized the level of neutronics calculation technique among participating countries. (author)

  19. Research Update: Enabling ultra-thin lightweight structures: Microsandwich structures with microlattice cores

    OpenAIRE

    J. A. Kolodziejska; C. S. Roper; Yang, S S; Carter, W B; Jacobsen, A. J.

    2015-01-01

    We achieve the benefits of large-scale structural hierarchy at the micro-scale by utilizing a self-propagating photopolymer waveguide process to form ultra-thin sandwich structures. A single step forms the microlattice sandwich core and bonds the core to both facesheets, minimizing adhesive mass and manufacturing time, with core thicknesses

  20. Thermal radiation in gas core nuclear reactors for space propulsion

    International Nuclear Information System (INIS)

    A diffusive model of the radial transport of thermal radiation out of a cylindrical core of fissioning plasma is presented. The diffusion approximation is appropriate because the opacity of uranium is very high at the temperatures of interest (greater than 3000 K). We make one additional simplification of assuming constant opacity throughout the fuel. This allows the complete set of solutions to be expressed as a single function. This function is approximated analytically to facilitate parametric studies of the performance of a test module of the nuclear light bulb gas-core nuclear-rocket-engine concept, in the Annular Core Research Reactor at Sandia National Laboratories. Our findings indicate that radiation temperatures in range of 4000-6000 K are attainable, which is sufficient to test the high specific impulse potential (approximately 2000 s) of this concept. 15 refs

  1. Thermal radiation in gas core nuclear reactors for space propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Slutz, S.A.; Gauntt, R.O.; Harms, G.A.; Latham, T.; Roman, W.; Rodgers, R.J. (Sandia National Lab, Albuquerque, NM (United States))

    1994-05-01

    A diffusive model of the radial transport of thermal radiation out of a cylindrical core of fissioning plasma is presented. The diffusion approximation is appropriate because the opacity of uranium is very high at the temperatures of interest (greater than 3000 K). We make one additional simplification of assuming constant opacity throughout the fuel. This allows the complete set of solutions to be expressed as a single function. This function is approximated analytically to facilitate parametric studies of the performance of a test module of the nuclear light bulb gas-core nuclear-rocket-engine concept, in the Annular Core Research Reactor at Sandia National Laboratories. Our findings indicate that radiation temperatures in range of 4000-6000 K are attainable, which is sufficient to test the high specific impulse potential (approximately 2000 s) of this concept. 15 refs.

  2. LABORATORY AND NUMERICAL INVESTIGATIONS OF RESIDENCE TIME DISTRIBUTION OF FLUIDS IN LAMINAR FLOW STIRRED ANNULAR PHOTOREACTOR

    Science.gov (United States)

    Laboratory and Numerical Investigations of Residence Time Distribution of Fluids in Laminar Flow Stirred Annular PhotoreactorE. Sahle-Demessie1, Siefu Bekele2, U. R. Pillai11U.S. EPA, National Risk Management Research LaboratorySustainable Technology Division,...

  3. Non newtonian annular alloy solidification in mould

    Energy Technology Data Exchange (ETDEWEB)

    Moraga, Nelson O.; Garrido, Carlos P. [Universidad de La Serena, Departamento de Ingenieria Mecanica, La Serena (Chile); Castillo, Ernesto F. [Universidad de Santiago de Chile, Departamento de Ingenieria Mecanica, Santiago (Chile)

    2012-08-15

    The annular solidification of an aluminium-silicon alloy in a graphite mould with a geometry consisting of horizontal concentric cylinders is studied numerically. The analysis incorporates the behavior of non-Newtonian, pseudoplastic (n=0.2), Newtonian (n=1), and dilatant (n=1.5) fluids. The fluid mechanics and heat transfer coupled with a transient model of convection diffusion are solved using the finite volume method and the SIMPLE algorithm. Solidification is described in terms of a liquid fraction of a phase change that varies linearly with temperature. The final results make it possible to infer that the fluid dynamics and heat transfer of solidification in an annular geometry are affected by the non-Newtonian nature of the fluid, speeding up the process when the fluid is pseudoplastic. (orig.)

  4. Performance of annular high frequency thermoacoustic engines

    Science.gov (United States)

    Rodriguez, Ivan A.

    This thesis presents studies of the behavior of miniature annular thermoacoustic prime movers and the imaging of the complex sound fields using PIV inside the small acoustic wave guides when driven by a temperature gradient. Thermoacoustic engines operating in the standing wave mode are limited in their acoustic efficiency by a high degree of irreversibility that is inherent in how they work. Better performance can be achieved by using traveling waves in the thermoacoustic devices. This has led to the development of an annular high frequency thermoacoustic prime mover consisting of a regenerator, which is a random stack in-between a hot and cold heat exchanger, inside an annular waveguide. Miniature devices were developed and studied with operating frequencies in the range of 2-4 kHz. This corresponds to an average ring circumference of 11 cm for the 3 kHz device, the resonator bore being 6 mm. A similar device of 11 mm bore, length of 18 cm was also investigated; its resonant frequency was 2 kHz. Sound intensities as high as 166.8 dB were generated with limited heat input. Sound power was extracted from the annular structure by an impedance-matching side arm. The nature of the acoustic wave generated by heat was investigated using a high speed PIV instrument. Although the acoustic device appears symmetric, its performance is characterized by a broken symmetry and by perturbations that exist in its structure. Effects of these are observed in the PIV imaging; images show axial and radial components. Moreover, PIV studies show effects of streaming and instabilities which affect the devices' acoustic efficiency. The acoustic efficiency is high, being of 40% of Carnot. This type of device shows much promise as a high efficiency energy converter; it can be reduced in size for microcircuit applications.

  5. Annular Alopecia Areata: Report of Two Cases

    OpenAIRE

    Bansal, Manish; Manchanda, Kajal; Pandey, SS

    2013-01-01

    Alopecia areata (AA) is an auto-immune disorder characterized by the appearance of non-scarring bald patches affecting the hair bearing areas of the body. Scalp is the most common site of involvement. AA can affect any age group. The usual pattern of the hair loss is oval or round. We hereby, report two cases of annular and circinate pattern of AA due to its unusual morphology.

  6. Annular pancreas associated with duodenal carcinoma

    Institute of Scientific and Technical Information of China (English)

    Enrico; Bronnimann; Silke; Potthast; Tatjana; Vlajnic; Daniel; Oertli; Oleg; Heizmann

    2010-01-01

    Annular pancreas (AP) is a rare congenital anomaly. Coexisting malignancy has been reported only in a few cases. We report what is, to the best of our knowledge, the first case in the English literature of duodenal adenocarcinoma in a patient with AP. In a 55-year old woman with duodenal outlet stenosis magnetic resonance cholangiopancreatography showed an aberrant pancreatic duct encircling the duodenum. Duodenojejunostomy was performed. Eight weeks later she presented with painless jaundice. Duodenopancre...

  7. Hysteresis in annular impinging jets

    Czech Academy of Sciences Publication Activity Database

    Trávníček, Zdeněk; Tesař, Václav

    2013-01-01

    Roč. 44, Januar 2013 (2013), s. 565-570. ISSN 0894-1777 R&D Projects: GA AV ČR(CZ) IAA200760801; GA ČR(CZ) GCP101/11/J019 Institutional research plan: CEZ:AV0Z20760514 Institutional support: RVO:61388998 Keywords : impinging jet * bistability * hysteresis Subject RIV: BK - Fluid Dynamics Impact factor: 2.080, year: 2013 http://www.sciencedirect.com/science/article/pii/S0894177712002348

  8. STABILITY OF SWIRLING ANNULAR FLOW

    Czech Academy of Sciences Publication Activity Database

    Maršík, František; Trávníček, Zdeněk; Novotný, Pavel; Werner, E.

    Kaohsiung : National Pingtung University of Science and Technolog, 2009 - (Tai, C.), s. 32-32 ISBN N. [Pacific Symposium on Flow Visualization and Image Processing /7./ (PSFVIP-7 2009). Kaohsiung (TW), 16.11.2009-19.11.2009] R&D Projects: GA AV ČR(CZ) IAA200760801; GA MŠk(CZ) 1M06031 Institutional research plan: CEZ:AV0Z20760514 Keywords : flow stability * swirling jet * thermodynamic stability Subject RIV: BK - Fluid Dynamics

  9. Neutronic and thermal-hydraulic analysis of new irradiation channels inside the Moroccan TRIGA Mark II research reactor core.

    Science.gov (United States)

    Chham, E; El Bardouni, T; Benaalilou, K; Boukhal, H; El Bakkari, B; Boulaich, Y; El Younoussi, C; Nacir, B

    2016-10-01

    This study was conducted to improve the capacity of radioisotope production in the Moroccan TRIGA Mark II research reactor, which is considered as one of the most important applications of research reactors. The aim of this study is to enhance the utilization of TRIGA core in the field of neutron activation and ensure an economic use of the fuel. The main idea was to create an additional irradiation channel (IC) inside the core. For this purpose, three new core configurations are proposed, which differ according to the IC position in the core. Thermal neutron flux distribution and other neutronic safety parameters such as power peaking factors, excess reactivity, and control rods worth reactivity were calculated using the Monte Carlo N-Particle Transport (MCNP) code and neutron cross-section library based on ENDF/B-VII evaluation. The calculated thermal flux in the central thimble (CT) and in the added IC for the reconfigured core is compared with the thermal flux in the CT of the existing core, which is taken as a reference. The results show that all the obtained fluxes in CTs are very close to the reference value, while a remarkable difference is observed between the fluxes in the new ICs and reference. This difference depends on the position of IC in the reactor core. To demonstrate that the Moroccan TRIGA reactor could safely operate at 2MW, with new configurations based on new ICs, different safety-related thermal-hydraulic parameters were investigated. The PARET model was used in this study to verify whether the safety margins are met despite the new modifications of the core. The results show that it is possible to introduce new ICs safely in the reactor core, because the obtained values of the parameters are largely far from compromising the safety of the reactor. PMID:27552124

  10. Vibration analysis of annular-like plates

    Science.gov (United States)

    Cheng, L.; Li, Y. Y.; Yam, L. H.

    2003-05-01

    The existence of eccentricity of the central hole for an annular plate results in a significant change in the natural frequencies and mode shapes of the structure. In this paper, the vibration analysis of annular-like plates is presented based on numerical and experimental approaches. Using the finite element analysis code Nastran, the effects of the eccentricity, hole size and boundary condition on vibration modes are investigated systematically through both global and local analyses. The results show that analyses for perfect symmetric conditions can still roughly predict the mode shapes of "recessive" modes of the plate with a slightly eccentric hole. They will, however, lead to erroneous results for "dominant" modes. In addition, the residual displacement mode shape is verified as an effective parameter for identifying damage occurring in plate-like structures. Experimental modal analysis on a clamped-free annular-like plate is performed, and the results obtained reveal good agreement with those obtained by numerical analysis. This study provides guidance on modal analysis, vibration measurement and damage detection of plate-like structures.

  11. Fast reactor core concepts to improve transmutation efficiency

    International Nuclear Information System (INIS)

    Fast Reactor (FR) core concepts to improve transmutation efficiency were conducted. A heterogeneous MA loaded core was designed based on the 1000MWe-ABR breakeven core. The heterogeneous MA loaded core with Zr-H loaded moderated targets had a better transmutation performance than the MA homogeneous loaded core. The annular pellet rod design was proposed as one of the possible design options for the MA target. It was shown that using annular pellet MA rods mitigates the self-shielding effect in the moderated target so as to enhance the transmutation rate

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

    Science.gov (United States)

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

    1976-01-01

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

  13. Kinetic study of the Tehran research reactor core with low enriched fuel

    Energy Technology Data Exchange (ETDEWEB)

    Pazirandeh, A.; Afshar Bakeshloo, A. [Tehran Univ. (Iran, Islamic Republic of). Physics Dept.; Bartsch, G. [Technische Univ. Berlin (Germany). Inst. fuer Energietechnik

    1997-11-01

    For evaluating the performance of the newly refuelled Tehran Research Reactor core with low enriched uranium fuel (LEU) in transient states a two group time dependent diffusion equation code (COSTANZA) was used. This paper presents results of calculations of the fast transients, revealing the steady performance of the core and fuel integrity during transient for a probable reactivity insertion of less than or equal dollar 1.5/0.5 s. The temperature dependant reactivity coefficients of the Doppler resonance broadening effect and of the moderator absorption cross section change and density dilution were calculated using cell-averaged 69 energy group WIMS-D/4 for two main libraries, old library and WIMKAL88, to 13 groups. The two group parameters for the COSTANZA code were also obtained by WIMS-D/4. (orig.) [Deutsch] Zur Bewertung der Leistungsfaehigkeit des neu beladenen Teheraner Forschungsreaktors mit niedrig angereichertem Uranbrennstoff bei Reaktivitaetstransienten wurde ein 2-Gruppen zeitabhaengiges Diffusionsprogramm COSTANZA verwendet. In der vorliegenden Arbeit werden Ergebnisse der Berechnung schneller Transienten vorgestellt, die das Verhalten des Reaktorkerns bzw. die Integritaet der Brennstaebe waehrend der Transienten fuer eine Reaktivitaetsaenderung von kleiner oder gleich Dollar 1.5/0.5 s zeigen. Die temperaturabhaengigen Reaktivitaetskoeffizienten der Doppler-Verbreitung im Brennstoff sowie der Dichteaenderung und der Neutronenabsorption im Moderator wurden mit Hilfe zellengemittelter 69 Energie-Gruppen der Datenbank WIMS-D/4 und fuer 13 Energiegruppen mit der Datenbank WIMKAL 88 ermittelt. Die Zweigruppendaten fuer das COSTANZA-Programm wurden ebenfalls mit Hilfe von WIMS-D/4 bestimmt. (orig.)

  14. Stability of swirling annular flow

    Czech Academy of Sciences Publication Activity Database

    Maršík, František; Trávníček, Zdeněk; Novotný, Pavel; Werner, E.

    2010-01-01

    Roč. 17, č. 3 (2010), s. 267-279. ISSN 1065-3090 R&D Projects: GA AV ČR(CZ) IAA200760801; GA MŠk(CZ) 1M06031 Institutional research plan: CEZ:AV0Z20760514 Keywords : swirling jet * hydrodynamic stability * impinging jet Subject RIV: BK - Fluid Dynamics http://www.begellhouse.com/journals/52b74bd3689ab10b,6bfbd93509947e2e,03fca4e77476857d.html

  15. Alternative core design for the Innovative Research Reactor (RRI) from neutronics aspects

    International Nuclear Information System (INIS)

    Based on its User Requirement Document and main function, RRI shall be able to provide a maximum thermal neutron flux of 1×1015 neutron cm-2s-1. The reason is that the RRI reactor can serve targets requiring a high neutron flux. From the previous results it was obtained that RRI design using fuel of RSG-GAS type was not possible to produce that high neutron flux. One among other reasons is that the geometry dimension is the large, as the neutron flux is inversely proportional to core volume. The objective of the study is to find an alternative core for RRI which meets the high neutron flux requirement. It was chosen an alternative fuel element one like used in JMTR (Japan Material Testing Reactor) that has smaller dimension compared to that of the RSG-GAS reactor. Besides that, active core's height was also varied for 70 cm and 75 cm. Design was carried out by means of analytic codes WIMS-D5B, Batan-FUEL and Batan-3DIFF. Alternative core applied compact core configuration concept of 5×5 with 4 follower control elements. The calculations resulted 3 (three) alternative cores fulfill the requirement, including core using RSG-GAS fuel type but of 70 cm height instead of 60 cm. Through analyzing from over all aspects of core safety and efficiency as well as effectively, core using JMTR fuel type with height of 70 cm represent the best alternative core. (author)

  16. Assessment of core structural materials and surveillance programme of research reactors. Report of the consultants meeting. Working material

    International Nuclear Information System (INIS)

    A series of presentations on the assessment of core structural components and materials at their facilities were given by the experts. The different issues related to degradation mechanisms were discussed. The outputs include a more thorough understanding of the specific challenges related to Research Reactors (RRs) as well as proposals for activities which could assist RR organizations in their efforts to address the issues involved. The experts recommend that research reactor operators consider implementation of surveillance programs for materials of core structural components, as part of ageing management program (TECDOC-792 and DS-412). It is recognised by experts that adequate archived structural material data is not available for many RRs. Access to this data and extension of existing material databases could help many operating organisations extend the operation of their RRs. The experts agreed that an IAEA Technical Meeting (TM) on Assessment of Core Structural Materials should be organised in December 2009 (IAEA HQ Vienna). The proposed objectives of the TM are: (i) exchange of detailed technical information on the assessment and ageing management of core structural materials, (ii) identification of materials of interest for further investigation, (iii) proposal for a new IAEA CRP on Assessment of Core Structural Materials, and (iv) identification of RRs prepared to participate in proposed CRP. Based on the response to a questionnaire prepared for the 2008 meeting of the Technical Working Group for Research Reactors, the number of engineering capital projects related to core structural components is proportionally lower than those related to,for example, I and C or electrical power systems. This implies that many operating research reactors will be operating longer using their original core structural components and justifies the assessment and evaluation programmes and activities proposed in this report. (author)

  17. Analytical solution of neutron transport equation in an annular reactor with a rotating pulsed source; Resolucao analitica da equacao de transporte de neutrons em um reator anelar com fonte pulsada rotativa

    Energy Technology Data Exchange (ETDEWEB)

    Teixeira, Paulo Cleber Mendonca

    2002-12-01

    In this study, an analytical solution of the neutron transport equation in an annular reactor is presented with a short and rotating neutron source of the type S(x) {delta} (x- Vt), where V is the speed of annular pulsed reactor. The study is an extension of a previous study by Williams [12] carried out with a pulsed source of the type S(x) {delta} (t). In the new concept of annular pulsed reactor designed to produce continuous high flux, the core consists of a subcritical annular geometry pulsed by a rotating modulator, producing local super prompt critical condition, thereby giving origin to a rotating neutron pulse. An analytical solution is obtained by opening up of the annular geometry and applying one energy group transport theory in one dimension using applied mathematical techniques of Laplace transform and Complex Variables. The general solution for the flux consists of a fundamental mode, a finite number of harmonics and a transient integral. A condition which limits the number of harmonics depending upon the circumference of the annular geometry has been obtained. Inverse Laplace transform technique is used to analyse instability condition in annular reactor core. A regenerator parameter in conjunction with perimeter of the ring and nuclear properties is used to obtain stable and unstable harmonics and to verify if these exist. It is found that the solution does not present instability in the conditions stated in the new concept of annular pulsed reactor. (author)

  18. Analysis of the loss of coolant accident for leu cores of Pakistan research reactor-1

    Energy Technology Data Exchange (ETDEWEB)

    Khan, L.A.; Bokhari, I.H.; Raza, S.S.

    1993-12-01

    Response of LEU cores for PARR-1 to a Loss of Coolant Accident (LOCA) has been studied. It has been assumed that pool water drains out due to double ended rupture of the primary coolant pipe or complete shearing off an experimental beam tube. Results show that for an operating power level of 10 MW, both the first high power and equilibrium cores would enter into melting conditions if the pool drain time is less than 22 h and 11 h respectively. However, an Emergency Core Cooling System (ECCS) capable of spraying the core at a flow rate of 8.3 m3/h, for the above mentioned duration, would keep the peak core temperature much below the critical value. Maximum operating power levels below which melting would not occur have been assessed to be 3.4 MW and 4.8 MW, respectively, for the first high power and equilibrium cores.

  19. Analysis of the loss of coolant accident for LEU cores of Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Response of LEU cores for PARR-1 to a Loss of Coolant Accident (LOCA) has been studied. It has been assumed that pool water drains out to double ended rupture of primary coolant pipe or complete shearing of an experimental beam tube. Results show that for an operating power level of 10 MW, both the first high power and equilibrium cores would enter into melting conditions if the pool drain time is less than 22 h and 11 h respectively. However, an Emergency Core Cooling System (ECCS) capable of spraying the core at flow rate of 8.3 m/sup 3/h, for the above mentioned duration, would keep the peak core temperature much below the critical value. Maximum operating power levels below which melting would not occur have been assessed to 3.4 MW and 4.8 MW, respectively, for the first high power and equilibrium cores. (author) 5 figs

  20. Free vibration analysis of smart annular FGM plates integrated with piezoelectric layers

    International Nuclear Information System (INIS)

    In this paper, a nonlinear free vibration analysis of a thin annular functionally graded (FG) plate integrated with two uniformly distributed actuator layers made of piezoelectric (PZT4) material on the top and bottom surfaces of the annular FG plate is presented based on Kirchhoff plate theory. The material properties of the functionally graded core plate are assumed to be graded in the thickness direction according to the power law distribution in terms of the volume fractions of the constituents and the distribution of the electric potential field along the thickness direction of piezoelectric layers is simulated by a sinusoidal function such that the Maxwell static electricity equation is satisfied. The differential equations of motion are solved analytically for various boundary conditions of the plate. The analytical solutions are derived and validated by comparing the obtained resonant frequencies of the piezoelectric coupled FG annular plate with those of an isotropic core plate. In a numerical study the emphasis is placed on investigating the effect of varying the gradient index of the FG plate on the free vibration characteristics of the structure. Also the good agreement between the results of this paper and those of the finite element (FE) analyses validated the presented approach

  1. Research on the Core Competitive Power Elements Evaluation System of Green Hotel

    OpenAIRE

    Hui Liang

    2013-01-01

    Green hotel is a new type of hospitality industry development model based on the concept of circular economy and sustainable development. This paper makes an analysis and evaluation of the elements of green hotel core competence, on this basis, constructs the Green Hotel core competitive evaluation index system. The construction of the system is conducive to understand the green hotel’s own competitive advantage objectively, and explore ways to enhance its core competitiveness, providing obje...

  2. Research and development of in-core transducers at the CIAE

    International Nuclear Information System (INIS)

    In this paper, R and D of in-core transducers at the CIAE are briefly summarized. With the construction and commissioning of PWR nuclear power plant in China, fuel rod behaviour need to be studied carefully. As conventional transducers cannot meet the requirements of in-core applications, R and D of in-core transducers are developed. Since 1980's, several kinds of in-core transducers have been successfully fabricated and tested under the conditions simulating PWR. At present, in-pile tests of the transducers combining with the studies of individual behaviour of PWR fuel rod are being planned at the CIAE. (author). 11 refs, 12 figs, 4 tabs

  3. Research on the Intrinsic Relationship of Customer Value and Corporate Core Competence

    Science.gov (United States)

    Ji, Guoping

    The article defined customer value and corporate core competence using induction method and described the characteristics of these two concepts. Then the author analyzed the intrinsic relationship of customer value and corporate core competence via the survey and case analysis methods. The author found that customer value was the basic point to cultivate corporate core competence which was the platform to achieve customer value. The article is of great help to provide some ideas of cultivating and developing corporate core competence based on customer value.

  4. Effect of Annular Slit Geometry on Characteristics of Spiral Jet

    Institute of Scientific and Technical Information of China (English)

    Shigeru Matsuo; Kwon-Hee Lee; Shinsuke Oda; Toshiaki Setoguchi; Heuy-Dong Kim

    2003-01-01

    A spiral flow using an annular slit connected to a conical cylinder does not need special device to generate a tangential velocity component of the flow and differs from swirling flows. Pressurized fluid is supplied to an annular chamber and injected into the convergent nozzle through the annular slit. The annular jet develops into the spiral flow. In the present study, a spiral jet discharged out of nozzle exit was obtained by using a convergent nozzle and an annular slit set in nozzle inlet, and the effect of annular slit geometry on characteristics of the spiral jet was investigated by using a Laser Doppler Velocimeter (LDV) experimentally. Furthermore, velocity distributions of the spiral jet were compared with those of a normal jet.

  5. Mineral Physics Research on Earth's Core and UTeach Outreach Activities at UT Austin

    Science.gov (United States)

    Lin, J.; Wheat, A. J.

    2011-12-01

    Comprehension of the alloying effects of major candidate light elements on the phase diagram and elasticity of iron addresses pressing issues on the composition, thermal structures, and seismic features of the Earth's core. Integrating this mineral physics research with the educational objectives of the CAREER award was facilitated by collaboration with the University of Texas at Austin's premier teaching program, UTeach. The UTeach summer outreach program hosts three one-week summer camps every year exposing K-12th graders to university level academia, emphasizing math and science initiatives and research. Each week of the camp either focuses on math, chemistry, or geology. Many of the students were underrepresented minorities and some required simultaneous translation; this is an effect of the demographics of the region, and caused some language barrier challenges. The students' opportunity to see first-hand what it is like to be on a university campus, as well as being in a research environment, such as the mineral physics lab, helps them to visualize themselves in academia in the future. A collection of displayable materials with information about deep-Earth research were made available to participating students and teachers to disseminate accurate scientific knowledge and enthusiasm. These items included a diamond anvil cell and diagrams of the diamond crystal structure, the layers of the Earth, and the phases of carbon to show that one element can have very different physical properties purely based on differences in structure. The students learned how advanced X-ray and optical laser spectroscopies are used to study properties of planetary materials in the diamond anvil cell. Stress was greatly placed on the basic mathematical relationship between force, area, and pressure, the fundamental principle involved with diamond anvil cell research. Undergraduate researchers from the lab participated in the presentations and hands-on experiments, and answered any

  6. Ultrasonogrphic diagnosis of snapping annular ligament in the elbow

    Energy Technology Data Exchange (ETDEWEB)

    Chai, Jee Won; Kim Su Jin; Lim, Hyun Kyong; Bae, Kee Jeong [SMG-SNU Boramae Medical Center, Seoul National University College of Medicine, Seoul (Korea, Republic of)

    2015-01-15

    Elbow snapping by annular ligament is rare and may be difficult to diagnose, when this Epub ahead of print condition is not familiar. We report a case of elbow snapping by annular ligament diagnosed by ultrasonography, which was confirmed by arthroscopic observation. The ultrasonographic findings were thickening of the annular ligament and snapping in and out of the radiocapitellar joint during elbow flexion and extension on dynamic ultrasonography.

  7. Preliminary fracture analysis of the core pressure boundary tube for the Advanced Neutron Source Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schulz, K.C. [Univ. of Turabo, Gurabo, Puerto (Puerto Rico). College of Engineering; Yahr, G.T. [Oak Ridge National Lab., TN (United States)

    1995-08-01

    The outer core pressure boundary tube (CPBT) of the Advanced neutron Source (ANS) reactor being designed at Oak Ridge National Laboratory is currently specified as being composed of 6061-T6 aluminum. ASME Boiler and Pressure Vessel Code fracture analysis rules for nuclear components are based on the use of ferritic steels; the expressions, tables, charts and equations were all developed from tests and analyses conducted for ferritic steels. Because of the nature of the Code, design with thin aluminum requires analytical approaches that do not directly follow the Code. The intent of this report is to present a methodology comparable to the ASME Code for ensuring the prevention of nonductile fracture of the CPBT in the ANS reactor. 6061-T6 aluminum is known to be a relatively brittle material; the linear elastic fracture mechanics (LEFM) approach is utilized to determine allowable flaw sizes for the CPBT. A J-analysis following the procedure developed by the Electric Power Research Institute was conducted as a check; the results matched those for the LEFM analysis for the cases analyzed. Since 6061-T6 is known to embrittle when irradiated, the reduction in K{sub Q} due to irradiation is considered in the analysis. In anticipation of probable requirements regarding maximum allowable flaw size, a survey of nondestructive inspection capabilities is also presented. A discussion of probabilistic fracture mechanics approaches, principally Monte Carlo techniques, is included in this report as an introduction to what quantifying the probability of nonductile failure of the CPBT may entail.

  8. Preliminary fracture analysis of the core pressure boundary tube for the Advanced Neutron Source Research Reactor

    International Nuclear Information System (INIS)

    The outer core pressure boundary tube (CPBT) of the Advanced neutron Source (ANS) reactor being designed at Oak Ridge National Laboratory is currently specified as being composed of 6061-T6 aluminum. ASME Boiler and Pressure Vessel Code fracture analysis rules for nuclear components are based on the use of ferritic steels; the expressions, tables, charts and equations were all developed from tests and analyses conducted for ferritic steels. Because of the nature of the Code, design with thin aluminum requires analytical approaches that do not directly follow the Code. The intent of this report is to present a methodology comparable to the ASME Code for ensuring the prevention of nonductile fracture of the CPBT in the ANS reactor. 6061-T6 aluminum is known to be a relatively brittle material; the linear elastic fracture mechanics (LEFM) approach is utilized to determine allowable flaw sizes for the CPBT. A J-analysis following the procedure developed by the Electric Power Research Institute was conducted as a check; the results matched those for the LEFM analysis for the cases analyzed. Since 6061-T6 is known to embrittle when irradiated, the reduction in KQ due to irradiation is considered in the analysis. In anticipation of probable requirements regarding maximum allowable flaw size, a survey of nondestructive inspection capabilities is also presented. A discussion of probabilistic fracture mechanics approaches, principally Monte Carlo techniques, is included in this report as an introduction to what quantifying the probability of nonductile failure of the CPBT may entail

  9. E-research platform of EPOS Thematic Core Service "ANTHROPOGENIC HAZARDS"

    Science.gov (United States)

    Orlecka-Sikora, Beata; Lasocki, Stanisław; Grasso, Jean Robert; Schmittbuhl, Jean; Kwiatek, Grzegorz; Garcia, Alexander; Cassidy, Nigel; Sterzel, Mariusz; Szepieniec, Tomasz; Dineva, Savka; Biggare, Pascal; Saccorotti, Gilberto; Sileny, Jan; Fischer, Tomas

    2016-04-01

    EPOS Thematic Core Service ANTHROPOGENIC HAZARDS (TCS AH) aims to create new research opportunities in the field of anthropogenic hazards evoked by exploitation of georesources. TCS AH, based on the prototype built in the framework of the IS-EPOS project (https://tcs.ah-epos.eu/), financed from Polish structural funds (POIG.02.03.00-14-090/13-00), is being further developed within EPOS IP project (H2020-INFRADEV-1-2015-1, INFRADEV-3-2015). TCS AH is designed as a functional e-research environment to ensure a researcher the maximum possible freedom for in silico experimentation by providing a virtual laboratory in which researcher will be able to create own workspace with own processing streams. The unique integrated RI is: (i) data gathered in the so- called "episodes", comprehensively describing a geophysical process, induced or triggered by human technological activity, which under certain circumstances can become hazardous for people, infrastructure and the environment and (ii) problem-oriented, specific high-level services, with the particular attention devoted to methods analyzing correlations between technology, geophysical response and resulting hazard. Services to be implemented are grouped within six blocks: (1) Basic services for data integration and handling; (2) Services for physical models of stress/strain changes over time and space as driven by geo-resource production; (3) Services for analysing geophysical signals; (4) Services to extract the relation between technological operations and observed induced seismic/deformation; (5) Services to quantitative probabilistic assessments of anthropogenic seismic hazard - statistical properties of anthropogenic seismic series and their dependence on time-varying anthropogenesis; ground motion prediction equations; stationary and time-dependent probabilistic seismic hazard estimates, related to time-changeable technological factors inducing the seismic process; (6) Simulator for Multi

  10. Granuloma annulare localized to the shaft of the penis

    DEFF Research Database (Denmark)

    Trap, R; Wiebe, B

    1993-01-01

    A case of granuloma annulare localized to the shaft of the penis is reported. The differential diagnoses are discussed. Penile granuloma annulare is a rare disorder and it is concluded that biopsies of penile lesions are recommended to verify the correct diagnosis.......A case of granuloma annulare localized to the shaft of the penis is reported. The differential diagnoses are discussed. Penile granuloma annulare is a rare disorder and it is concluded that biopsies of penile lesions are recommended to verify the correct diagnosis....

  11. Characteristic differences of LEU and HEU cores at the German FRJ-2 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nabbi, R.; Wolters, J.; Damm, G. [Central Research Reactor Division, Forschungszentrum Juelich, 52425 Juelich (Germany)

    2002-07-01

    As a sophisticated computational method for reactor physics analysis and fuel management an MCNP model in very high fidelity was developed and coupled with a depletion code and applied to the HEU-LEU core conversion study. The analysis show that as a consequence of the high amount of U-238, the amount of U-235 in the LEU core is about 14% higher than in the HEU core. The reduction of the thermal flux varies between 16% (core) and 5% in the reflector zone. The rate of U-235 burnup in the LEU core is approx. 11.5% lower which allows an extension of irradiation time. Due to the effect of neutron spectrum the worth of the absorber system decreases in an LEU core by 17% resulting in a decrease of shutdown and excess reactivity. The kinetic parameters of the core are slightly reduced causing changes in the reactivity values and transient behavior of the core. The moderator coefficient is decreased by 18% and the Doppler coefficient is increased by 63%. Due to shortening of the absorption length of the fission neutrons the prompt neutron lifetime is reduced by 7%. (author)

  12. Core Intervention Components: Identifying and Operationalizing What Makes Programs Work. ASPE Research Brief

    Science.gov (United States)

    Blase, Karen; Fixsen, Dean

    2013-01-01

    This brief is part of a series that explores key implementation considerations. It focuses on the importance of identifying, operationalizing, and implementing the "core components" of evidence-based and evidence-informed interventions that likely are critical to producing positive outcomes. The brief offers a definition of "core components",…

  13. Effective height of the core of the nuclear research reactor Dalat

    International Nuclear Information System (INIS)

    Measurements of thermal neutron relative distributions in axial direction at different positions in the reactor core and for various control rod configurations have been carried out, and axial buckling and effective height of the core deduced. (author). 4 refs., 3 figs., 1 tab

  14. Characteristic differences of LEU and HEU cores at the German FRJ-2 research reactor

    International Nuclear Information System (INIS)

    As a sophisticated computational method for reactor physics analysis and fuel management an MCNP model in very high fidelity was developed and coupled with a depletion code and applied to the HEU-LEU core conversion study. The analysis show that as a consequence of the high amount of U-238, the amount of U-235 in the LEU core is about 14% higher than in the HEU core. The reduction of the thermal flux varies between 16% (core) and 5% in the reflector zone. The rate of U-235 burnup in the LEU core is approx. 11.5% lower which allows an extension of irradiation time. Due to the effect of neutron spectrum the worth of the absorber system decreases in an LEU core by 17% resulting in a decrease of shutdown and excess reactivity. The kinetic parameters of the core are slightly reduced causing changes in the reactivity values and transient behavior of the core. The moderator coefficient is decreased by 18% and the Doppler coefficient is increased by 63%. Due to shortening of the absorption length of the fission neutrons the prompt neutron lifetime is reduced by 7%. (author)

  15. CFD model of diabatic annular two-phase flow using the Eulerian–Lagrangian approach

    International Nuclear Information System (INIS)

    Highlights: • A CFD model of annular two-phase flow with evaporating liquid film has been developed. • A two-dimensional liquid film model is developed assuming that the liquid film is sufficiently thin. • The liquid film model is coupled to the gas core flow, which is represented using the Eulerian–Lagrangian approach. - Abstract: A computational fluid dynamics (CFD) model of annular two-phase flow with evaporating liquid film has been developed based on the Eulerian–Lagrangian approach, with the objective to predict the dryout occurrence. Due to the fact that the liquid film is sufficiently thin in the diabatic annular flow and at the pre-dryout conditions, it is assumed that the flow in the wall normal direction can be neglected, and the spatial gradients of the dependent variables tangential to the wall are negligible compared to those in the wall normal direction. Subsequently the transport equations of mass, momentum and energy for liquid film are integrated in the wall normal direction to obtain two-dimensional equations, with all the liquid film properties depth-averaged. The liquid film model is coupled to the gas core flow, which currently is represented using the Eulerian–Lagrangian technique. The mass, momentum and energy transfers between the liquid film, gas, and entrained droplets have been taken into account. The resultant unified model for annular flow has been applied to the steam–water flow with conditions typical for a Boiling Water Reactor (BWR). The simulation results for the liquid film flow rate show favorable agreement with the experimental data, with the potential to predict the dryout occurrence based on criteria of critical film thickness or critical film flow rate

  16. Substantiation of parameters of the geometric model of the research reactor core for the calculation using the Monte Carlo method

    Energy Technology Data Exchange (ETDEWEB)

    Radaev, A. I., E-mail: radaev-aleksandr@mail.ru; Schurovskaya, M. V., E-mail: mvhchurovskaya@mephi.ru [National Research Nuclear University MEPhI (Russian Federation)

    2015-12-15

    The choice of the spatial nodalization for the calculation of the power density and burnup distribution in a research reactor core with fuel assemblies of the IRT-3M and VVR-KN type using the program based on the Monte Carlo code is described. The influence of the spatial nodalization on the results of calculating basic neutronic characteristics and calculation time is investigated.

  17. The Whole-Core LEU U3Si2-Al Fuel Demonstration in the 30-MW Oak Ridge Research Reactor

    International Nuclear Information System (INIS)

    The ORR Whole-Core LEU Fuel Demonstration, conducted as part of the US Reduced Enrichment Research and Test Reactor Program, has been successfully completed. Using commercially-fabricated U3Si2-Al 20%-enriched fuel elements (4.8 g U/cc) and fuel followers (3.5 g U/cc), the 30-MW Oak Ridge Research Reactor was safely converted from an all-HEU core, through a series of HEU/LEU mixed transition cores, to an all-LEU core. There were no fuel element failures and average discharge burnups were measured to be as high as 50% for the standard elements and 75% for the fuel followers. Experimental results for burnup-dependent critical configurations, cycle-averaged fuel element powers, and fuel-element-averaged 235U burnups, validated predictions based on three-dimensional depletion calculations. Calculated values for plutonium production and isotopic mass ratios as functions of 235U burnup support the corresponding measured quantities. In general, calculations for 60Co and 198Au reaction rate distributions, differential and integral control rod worths, prompt neutron decay constants, and isothermal temperature coefficients were found to agree with corresponding measured values. Experimentally determined critical configurations for fresh HEU and LEU cores radially reflected with water and with beryllium are well-predicted by both Monte Carlo and diffusion calculations. 44 refs., 57 figs., 45 tabs

  18. Thread-annular flow in vertical pipes

    Science.gov (United States)

    Frei, Ch.; Lüscher, P.; Wintermantel, E.

    2000-05-01

    Thread injection is a promising method for different minimally invasive medical applications. This paper documents an experimental study dealing with an axially moving thread in annular pipe flow. Mass flow and axial force on the thread are measured for a 0.46 mm diameter thread in pipes with diameters between 0.55 and 1.35 mm. The experiments with thread velocities of up to 1.5 ms[minus sign]1 confirm the findings of theoretical studies that for clinical requirements the radius ratio between thread and pipe is crucial for the adjustments of mass ow and force on the thread.

  19. Deep variant of Erythema Annulare Centrifugum

    OpenAIRE

    Ahu Yorulmaz; Ferda Artuz; Devrim Tuba Unal

    2014-01-01

    A 29-year-old woman came to our outpatient clinic with a several-month history of itchy red lesions over her trunk. There was no family history and past history of any other diseases or medication. Dermatological examination revealed annular and oval-shaped plaques up to several cm’s in size, one of which was polycyclic in configuration, on back of the patient (Fig. 1). It was also noticed that lesions had erythematous indurated bordes with paler central areas (Fig. 1).

  20. Deep variant of Erythema Annulare Centrifugum

    Directory of Open Access Journals (Sweden)

    Ahu Yorulmaz

    2014-10-01

    Full Text Available A 29-year-old woman came to our outpatient clinic with a several-month history of itchy red lesions over her trunk. There was no family history and past history of any other diseases or medication. Dermatological examination revealed annular and oval-shaped plaques up to several cm’s in size, one of which was polycyclic in configuration, on back of the patient (Fig. 1. It was also noticed that lesions had erythematous indurated bordes with paler central areas (Fig. 1.

  1. Wave turbulence in annular wave tank

    Science.gov (United States)

    Onorato, Miguel; Stramignoni, Ettore

    2014-05-01

    We perform experiments in an annular wind wave tank at the Dipartimento di Fisica, Universita' di Torino. The external diameter of the tank is 5 meters while the internal one is 1 meter. The tank is equipped by two air fans which can lead to a wind of maximum 5 m/s. The present set up is capable of studying the generation of waves and the development of wind wave spectra for large duration. We have performed different tests including different wind speeds. For large wind speed we observe the formation of spectra consistent with Kolmogorv-Zakharov predictions.

  2. Air-water countercurrent annular flow

    Energy Technology Data Exchange (ETDEWEB)

    Bharathan, D.

    1979-09-01

    Countercurrent annular flow of air and water in circular tubes of diameters ranging from 6.4 to 152 mm is investigated. Experimental measurements include liquid fraction, pressure gradients and countercurrent gas and liquid fluxes. Influences of tube end geometries on the countercurrent fluxes are isolated. Analogies between countercurrent flow, open channel flow, and compressible flow are established. Interfacial momentum transfer between the phases are characterized by empirical friction factors. The dependence of interfacial friction factors on tube diameter is shown to yield a basis for extending the present results to larger tubes.

  3. Annular diffraction of very unstable light nuclei

    International Nuclear Information System (INIS)

    Because they are brittle, unstable light nuclei can produce an annular diffraction pattern observed on their decay products with large cross sections. With such a simple model, the 9Li angular distribution observed in the 11Li fragmentation have been reproduced together with the reaction cross-section and the fragmentation yield provided recoil effects from neutron emission are included. It results that for this projectile and for light targets, diffraction is the main source of transverse momentum for 9Li whereas for neutrons it originates from its emission energy in the 11Li center of mass

  4. Air-water countercurrent annular flow

    International Nuclear Information System (INIS)

    Countercurrent annular flow of air and water in circular tubes of diameters ranging from 6.4 to 152 mm is investigated. Experimental measurements include liquid fraction, pressure gradients and countercurrent gas and liquid fluxes. Influences of tube end geometries on the countercurrent fluxes are isolated. Analogies between countercurrent flow, open channel flow, and compressible flow are established. Interfacial momentum transfer between the phases are characterized by empirical friction factors. The dependence of interfacial friction factors on tube diameter is shown to yield a basis for extending the present results to larger tubes

  5. Establishment of cDNA Microarray Analysis at the Genomic Medicine Research Core Laboratory (GMRCL) of Chang Gung Memorial Hospital .

    OpenAIRE

    Tzu-Hao Wang; Yun-Shien Lee; En-Shih Chen; Wei-Hsiang Kong; Lung-Kun Chen; Ding-Wei Hsueh; Min-Li Wei; Hsing-Shih Wang; Ying-Shiung Lee

    2004-01-01

    Background: Advances in molecular and computational biology have led to the developmentof powerful, high-throughput methods for analysis of differential geneexpression, which are opening up new opportunities in genomic medicine.DNA microarray technology has been enthusiastically integrated into basicbiomedical research and will eventually become a molecular monitoring toolfor various clinical courses.Methods: As a core research facility of Chang Gung University (CGU) and ChangGung Memorial Ho...

  6. Theoretical and experimental study of inverted annular film boiling and regime transition during reflood transients

    Science.gov (United States)

    Mohanta, Lokanath

    The Loss of Coolant Accident (LOCA) is a design basis accident for light water reactors that usually determines the limits on core power. During a LOCA, film boiling is the dominant mode of heat transfer prior to the quenching of the fuel rods. The study of film boiling is important because this mode of heat transfer determines if the core can be safely cooled. One important film boiling regime is the so-called Inverted Annular Film Boiling (IAFB) regime which is characterized by a liquid core downstream of the quench front enveloped by a vapor film separating it from the fuel rod. Much research have been conducted for IAFB, but these studies have been limited to steady state experiments in single tubes. In the present work, subcooled and saturated IAFB are investigated using high temperature reflood data from the experiments carried out in the Rod Bundle Heat Transfer (RBHT) test facility. Parametric effects of system parameters including the pressure, inlet subcooling, and flooding rate on the heat transfer are investigated. The heat transfer behavior during transition to Inverted Slug Film Boiling (ISFB) regime is studied and is found to be different than that reported in previous studies. The effects of spacer grids on heat transfer in the IAFB and ISFB regimes are also presented. Currently design basis accidents are evaluated with codes in which heat transfer and wall drag must be calculated with local flow parameters. The existing models for heat transfer are applicable up to a void fraction of 0.6, i.e. in the IAFB regime and there is no heat transfer correlation for ISFB. A new semi-empirical heat transfer model is developed covering the IAFB and ISFB regimes which is valid for a void fraction up to 90% using the local flow variables. The mean absolute percentage error in predicting the RBHT data is 11% and root mean square error is 15%. This new semi-empirical model is found to compare well with the reflood data of FLECHT-SEASET experiments as well as data

  7. Complex degradation and ageing phenomena of research reactor core structural materials - experience at 14 MW TRIGA reactor from INR Pitesti

    International Nuclear Information System (INIS)

    The 14 MW TRIGA Research Reactor designed in the early '70s is a relative new research reactor with an operational experience of 30 years. The specific design of reactor core objectives, were to manufacture, build and operate a flexible structure which incorporate previous experience of pool type research reactors. Aluminum alloy 6061 and stainless steel are only materials used for core structural components, which are all easily remotely removable and replaceable by simple hand tools. Properties of those categories of materials were well characterized / known for many other reactors predecessors, and no special criteria or preliminary tests were performed. The mechanical core structure is presented in the paper and designed procedure for periodic testing and inspection is also described. In spite of well known materials properties, the behavior uncertainties of those materials in each reactor case may have special aspects related to design of components, manufacturing technologies, surface finishing and processing, quality control methods, price of specific components, complex conditions in core and vicinity, history of operation, inspection and verification of components, radioactive waste characterization at the end of life of components. Limited assessment of materials properties and suitability for certain application without considering the each individual component load, exposure and life time, may produce limited information on material itself, in fact the issue is the selection criteria for a standard material suitable for a certain application and consequent failure of components. The degradation and ageing are specific to components starting from design, manufacturing technology and expected life when the component should be replaced. The paper presents the practical experience on maintenance requirements specific to TRIGA core components and some techniques of material investigations available at Institute for Nuclear Research Pitesti Post Irradiation

  8. Real time simulation research in 200 MW low temperature nuclear heating reactor core

    International Nuclear Information System (INIS)

    200 MW low temperature nuclear heating reactor is an important new-type reactor. Natural circulation is adopted in the flowage of reactor core. High precise models are built and selected, which are low temperature reactor power model, residual power releasing model, heat conductivity model in reactor core, thermo-hydraulic model, subcooling boiling model, CHF calculation model and so on. These models are solved using Gear arithmetic and Adams arithmetic, which are testified each other. Using appropriate arithmetic, the real time simulation of thermo-hydraulic process in the core is truly fulfilled. (authors)

  9. Application of Lubricant to Minimize Axial Deviation of Annular Pellet Diameter

    International Nuclear Information System (INIS)

    In the nuclear industry, the elevation of an economical efficiency for a nuclear fuel is one of the major issues. To increase the efficiency, a development of the nuclear fuel for a high burnup and extended cycle is necessary. In the development of a high performance fuel, in-reactor fuel behavior must be seriously considered. Also, a fuel fabrication and an enrichment process must be discussed. A modification and an improvement of a nuclear fuel system will be also required. The typical fuel geometry of a PWR (Pressurized Water Reactor) is composed of a cylindrical pellet with a tubular cladding. And the outer surface of the cladding is cooled with water. However, to allow for a substantial increase in the power density, an additional cooling is necessary. One of the best ways is the application of a new fuel geometry that is of an annular shape and has both an internal and external cooling. From this point of view, a double cooled fuel is being developed by KAERI (Korea Atomic Energy Research Institute), and as a part of the project, the development of a fabrication process for a UO2 annular pellet is now in progress. In developing the fabrication technology for an annular pellet, there are various methods which can be applied to the fabrication of an annular pellet. But a die pressing method was dominantly chosen, because it is profitable for a production on a large scale

  10. Development of Technology for Improving the Dual Cooling Annular Fuel Pellet Heat Transfer

    International Nuclear Information System (INIS)

    The purpose of this project is to conduct CHF experiments using nano fluid and to check the application possibility of nano fluid to annular fuel for developing high performance dual cooling annular fuel pellet. To achieve this purpose, We set the direction of research by literature survey and conducted experiments using various experimental apparatus. The main purposes of the experiments contained in the present study are understanding about effect of nano fluid on CHF and investigation of related phenomena. CHF enhancement by nano fluid can increase the the thermal margin of dual cooling annular fuel and thus increase the application possibility of annular fuel to nuclear power plant. The present study consist of two parts. First, we study about the effect of nano fluid on thermal conductivity, wettability, CHF in pool boiling condition. Second, we study about the effect of nano fluid on CHF in flow boiling condition. Part 1 : Thermal conductivity, wettability, CHF experiments using nano fluid in pool boiling condition Part 2 : CHF experiments using nano fluid in flow boiling condition

  11. Steady state thermal hydraulic analysis of LEU cores for Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Maximum operating power levels of the high power and equilibrium LEU cores for PARR-1 have been assessed. The criterion followed is that nucleate boiling should not commence at any point in the core, when reactor power approaches overpower limiting set point of 115% and simultaneously the coolant flow rate reduces low flow set point of 90%. Steady state operating conditions have been calculated for the assessed maximum power. These include coolant velocity distribution in the core, critical velocity, pressure drop, saturation temperature, temperature distribution in the core and margins to onset of nucleate boiling, onset of flow instability and departure from nucleate boiling. Cooling conditions for the end fuel plates have also been analyzed. (author)

  12. Detonation diffraction from an annular channel

    Science.gov (United States)

    Meredith, James; Ng, Hoi Dick; Lee, John H. S.

    2010-12-01

    In this study, gaseous detonation diffraction from an annular channel was investigated with a streak camera and the critical pressure for transmission of the detonation wave was obtained. The annular channel was used to approximate an infinite slot resulting in cylindrically expanding detonation waves. Two mixtures, stoichiometric acetylene-oxygen and stoichiometric acetylene-oxygen with 70% Ar dilution, were tested in a 4.3 and 14.3 mm channel width ( W). The undiluted and diluted mixtures were found to have values of the critical channel width over the cell size around 3 and 12 respectively. Comparing these results to values of the critical diameter ( d c ), in which a spherical detonation occurs, a value of critical d c / W c near 2 is observed for the highly diluted mixture. This value corresponds to the geometrical factor of the curvature term between a spherical and cylindrical diverging wave. Hence, the result is in support of Lee's proposed mechanism [Lee in Dynamics of Exothermicity, pp. 321, Gordon and Breach, Amsterdam, 1996] for failure due to diffraction based on curvature in stable mixtures such as those highly argon diluted with very regular detonation cellular patterns.

  13. Annular beam shaping and optical trepanning

    Science.gov (United States)

    Zeng, Danyong

    Percussion drilling and trepanning are two laser drilling methods. Percussion drilling is accomplished by focusing the laser beam to approximately the required diameter of the hole, exposing the material to one or a series of laser pulses at the same spot to melt and vaporize the material. Drilling by trepanning involves cutting a hole by rotating a laser beam with an optical element or an x-y galvo-scanner. Optical trepanning is a new laser drilling method using an annular beam. The annular beams allow numerous irradiance profiles to supply laser energy to the workpiece and thus provide more flexibility in affecting the hole quality than a traditional circular laser beam. Heating depth is important for drilling application. Since there are no good ways to measure the temperature inside substrate during the drilling process, an analytical model for optical trepanning has been developed by considering an axisymmetric, transient heat conduction equation, and the evolutions of the melting temperature isotherm, which is referred to as the melt boundary in this study, are calculated to investigate the influences of the laser pulse shapes and intensity profiles on the hole geometry. This mathematical model provides a means of understanding the thermal effect of laser irradiation with different annular beam shapes. To take account of conduction in the solid, vaporization and convection due to the melt flow caused by an assist gas, an analytical two-dimensional model is developed for optical trepanning. The influences of pulse duration, laser pulse length, pulse repetition rate, intensity profiles and beam radius are investigated to examine their effects on the recast layer thickness, hole depth and taper. The ray tracing technique of geometrical optics is employed to design the necessary optics to transform a Gaussian laser beam into an annular beam of different intensity profiles. Such profiles include half Gaussian with maximum intensities at the inner and outer

  14. Critical and power experiments on the low-enriched uranium core of the upgraded Pakistan Research Reactor-1

    International Nuclear Information System (INIS)

    The Pakistan Research Reactor was converted from 93% highly enriched uranium fuel to 20% low-enriched uranium fuel in October 1991. The reactor power was also upgraded from 5 to 9 MW. A series of critical and power experiments were performed on the new core for verification of design data and to determine the nuclear performance of the reactor. The characteristics tests included a criticality experiment, reactivity measurements on reflected and unreflected, critical and full-power cores, and flux distribution in and around the core, as well as thermal-hydraulic measurements. A comparison of the measured and the calculated results was also made. The results of the characteristics tests indicate that the performance of the new reactor is within design limits

  15. 75 FR 23582 - Annular Casing Pressure Management for Offshore Wells

    Science.gov (United States)

    2010-05-04

    ... Recommended Practice (RP) 90. As explained in API RP 90, Section 3, Annular Casing Pressure Management Program... Institute's Recommended Practice for managing annular casing pressure. New regulations are needed because... Continental Shelf lessees to follow best industry practices for wells with sustained casing pressure....

  16. Obtention of an empirical equation for annular channels

    International Nuclear Information System (INIS)

    Using a trial circuit, the experimental heat transfer coefficient is determined, in forced convection at one phase only within an annular channel in which water flows ascendantly and for this reason an empirical equation is determined. This work tries to contribute to the understanding of the forced convection phenomena in non tubular geometries like the annular channels. (Author)

  17. Limited Diffraction Maps for Pulsed Wave Annular Arrays

    DEFF Research Database (Denmark)

    Fox, Paul D.

    2002-01-01

    A procedure is provided for decomposing the linear field of flat pulsed wave annular arrays into an equivalent set of known limited diffraction Bessel beams. Each Bessel beam propagates with known characteristics, enabling good insight into the propagation of annular fields to be obtained...

  18. Full-Core Conversion of the WWR-M Research Reactor in Ukraine to the Use of LEU Fuel

    International Nuclear Information System (INIS)

    Full-core conversion of the WWR-M research reactor in Ukraine with simultaneous replacement of all remaining HEU fuel by fresh LEU fuel requires the new safety analysis of the reactor because of great decrease of the number of fuel assemblies in the core. Because of considerable increase of reactivity due to loading a fuel assembly into the core and reactivity worth of control rods, the following potential accidents are analysed for the new LEU core: incidental falling of a fuel assembly in a cell of the core and spontaneous withdrawal of a control rod group because of malfunction of electronic equipment. To provide the safety of the reactor, some limiting conditions for operation are revised. In particular, maximum allowed effective multiplication factor when all control rods are fully in and all safety rods are fully out is decreased from 0.988 to 0.977, and maximum allowed power of the reactor is decreased from 10 MW to 7 MW. The safety analysis shows that with the revised limiting conditions for operation, such the events with accompanying one additional equipment malfunction and one error of personnel do not lead to damage of fuel elements and release of radioactivity exceeding allowed level. For neutronics calculations, the MCNP code based on the Monte Carlo method is applied. Thermal-hydraulics is calculated with the PLTEMP code. (author)

  19. Effect of annular secondary conductor in a linear electromagnetic stirrer

    Indian Academy of Sciences (India)

    R Madhavan; V Ramanarayanan

    2008-10-01

    This paper presents the variation of average axial force density in the annular secondary conductor of a linear electromagnetic stirrer. Different geometries of secondaries are considered for numerical and experimental validation namely, 1. hollow annular ring, 2. annular ring with a solid cylinder and 3. solid cylinder. Experimental and numerical simulations are performed for a 2-pole in house built 15 kW linear electromagnetic stirrer (EMS). It is observed for a supply current of 200 A at 30 Hz the force densities in the hollow annular ring is 67% higher than the equivalent solid cylinder. The same values are 33% for annular ring with a solid cylinder. Force density variation with supply frequency and current are also reported. Numerical simulations using finite element model are validated with experimental results.

  20. Nutritional implications of organic conversion in large scale food service preliminary results from Core Organic research:Preliminary results from Core Organic research

    OpenAIRE

    Mikkelsen, Bent Egberg; He, Chen

    2010-01-01

    The discussion about nutritional advantages of organic consumption hastraditionally focused on the properties of the food it self. Studies have shownhowever that change of consumption patterns towards organic food seems toinduce changed dietary patterns. The current research was a part of the iPOPYstudy and was conducted to investigate if such changes can be found in schoolfood settings. In other words does organic food schemes at school and relatedcurricular activities help to create environ...

  1. Thermal-hydraulic design calculations for the annular fuel element with replaceable test bundles (TOAST) on the test zone position 205 of KNK II/3

    International Nuclear Information System (INIS)

    Annular fuel elements are foreseen in KNK II as carrier elements for irradiation inserts and test bundles. For the third core a reloadable annular element on position 205 is foreseen, in which replaceable 19-pin test bundles (TOAST) shall be irradiated. The present report deals with the thermal-hydraulic design of the annular carrier element and the test bundle, whereby the test bundle required additional optimization. The code CIA has been used for the calculations. Start of irradiation of the subassembly is planned at the beginning of the third core operation. After optimization of the pin-spacer geometry in the test bundle, design calculations for both bundles were performed, whereby thermal coupling between both was taken into account. The calculated mass-flows and temperature distributions are given for the nominal and the eccentric element configuration. The calculated bundle pressure losses have been corrected according to experimental results

  2. Upgrading of the coupled neutronics-fluid dynamics code SIMMER to simulate the research reactors core disruptive RIA

    Energy Technology Data Exchange (ETDEWEB)

    Biaut, Guillaume; Couturier, Jean [IRSN, Fontenay-aux-roses (France); Wilhelm, Dirk; Ping, Liu [FzK, Karlsruhe (Germany)

    2008-07-01

    Up to now, the French aluminium plate-type, water-moderated research reactors have been designed by taking into account the consequences of a core disruptive RIA with a constant bounding thermal energy released of 135 MJ during the power transient and a mechanical energy, rising from the thermodynamic interaction between molten aluminium and the liquid water, accounting for 9% of the thermal energy. Nevertheless, for the IRSN, both BORAX-I, SPERT-I destructive tests and SL-1 accident do not show restrictive phenomena on the thermal energy released which mainly depends on reactivity insertions rates and core characteristics. Consequently, the IRSN has adopted a new approach, within the framework of the 'Beyond Design Basis Accidents' (BDBA), while studying scenarios representative of large reactivity insertions sequences by upgrading the coupled neutronics-fluid dynamics code SIMMER, designed for LMFR, to treat BDBA in water-moderated research reactors. A method for taking into account the heterogeneities of the core on the resonance self-shielded fuel cross sections has been developed; a model to treat fuel plate geometry has been implemented and also a new clad-to-coolant heat transfer coefficients suitable for extremely fast transient conditions. At the present stage of the studies, it is found that, for large reactivity insertion sequences tested, the geometry around the core has a dominant influence on pressures inside the reactor. (authors)

  3. Assessment of core structural materials and surveillance programme of research reactors in Egypt

    International Nuclear Information System (INIS)

    The main structural materials to be used in the reactor core, support structures are stainless steel, aluminum and zirconium alloys (zircadyne). Other materials are also used, for example such as polymers in seals and protective coating, and hafnium (HF) as absorber materials in the control rod plates. Stainless steel is used for the reactor pool. The mechanical properties of stainless steel alloys change when they are subjected to irradiation. The main phenomena observed are swelling and irradiation - induced creep. The swelling phenomenon depends on the operating temperature and neutron fluence. For the reactor facility, components will operate at temperature below 70 o C and are expected to see a lifetime fluence of approximately 1 x 1023 n.cm-2.these conditions are well below the conditions where swelling becomes significant. Stainless steels have strong resistance to corrosion over a wide range of environments and temperature. The reactor pool and primary circuit water is demineralized water with controlled low conductivity of less than 100 μ.sm-1 no failure mechanism is known under such process conditions. Aluminum alloys will be used for the constructions of some reactor internals which working in radiation environment as their properties are well understood and show predictable behavior under such conditions. Aluminum is extensively used in water - cooled research reactors because of its low cross-section for the capture of thermal neutrons, excellent corrosion resistance and thermal conductivity. Irradiation damage of polymers strongly depends on the fluence received by the materials. Irradiation effects of polymers also depend on their compositions and molecular structure. if the content of natural rubber is high, irradiation induces an increase in the tensile strength. Where the content of polypropylene is high, irradiation reduces the strength. A materials surveillance plan has been developed and will be implemented from the commencement of reactor

  4. Research Update: Enabling ultra-thin lightweight structures: Microsandwich structures with microlattice cores

    Science.gov (United States)

    Kolodziejska, J. A.; Roper, C. S.; Yang, S. S.; Carter, W. B.; Jacobsen, A. J.

    2015-05-01

    We achieve the benefits of large-scale structural hierarchy at the micro-scale by utilizing a self-propagating photopolymer waveguide process to form ultra-thin sandwich structures. A single step forms the microlattice sandwich core and bonds the core to both facesheets, minimizing adhesive mass and manufacturing time, with core thicknesses <2 mm, facesheet thicknesses ranging from 12.7 to 300 μm, areal densities 0.030-0.041 g cm-2, and flexural rigidity per unit width up to 0.62 Nm. This work extends the lightweighting benefit of sandwich structures to lower thicknesses and areal densities that were previously the exclusive domain of monolithic materials.

  5. Research Update: Enabling ultra-thin lightweight structures: Microsandwich structures with microlattice cores

    Directory of Open Access Journals (Sweden)

    J. A. Kolodziejska

    2015-05-01

    Full Text Available We achieve the benefits of large-scale structural hierarchy at the micro-scale by utilizing a self-propagating photopolymer waveguide process to form ultra-thin sandwich structures. A single step forms the microlattice sandwich core and bonds the core to both facesheets, minimizing adhesive mass and manufacturing time, with core thicknesses <2 mm, facesheet thicknesses ranging from 12.7 to 300 μm, areal densities 0.030–0.041 g cm−2, and flexural rigidity per unit width up to 0.62 Nm. This work extends the lightweighting benefit of sandwich structures to lower thicknesses and areal densities that were previously the exclusive domain of monolithic materials.

  6. Strategies for capacity building for health research in Bangladesh: Role of core funding and a common monitoring and evaluation framework

    Directory of Open Access Journals (Sweden)

    Mahmood Shakeel

    2011-07-01

    Full Text Available Abstract Background There is increasing interest in building the capacity of researchers in low and middle income countries (LMIC to address their national priority health and health policy problems. However, the number and variety of partnerships and funding arrangements can create management problems for LMIC research institutes. This paper aims to identify problems faced by a health research institute in Bangladesh, describe two strategies developed to address these problems, and identify the results after three years of implementation. Methods This paper uses a mixture of quantitative and qualitative data collected during independent annual reviews of the International Centre for Diarrhoeal Disease Research, Bangladesh (ICDDR,B between 2006 and 2010. Quantitative data includes the number of research activities according to strategic priority areas, revenues collected and expenditure. Qualitative data includes interviews of researchers and management of ICDDR,B, and of research users and key donors. Data in a Monitoring and Evaluation Framework (MEF were assessed against agreed indicators. Results The key problems faced by ICDDR,B in 2006 were insufficient core funds to build research capacity and supporting infrastructure, and an inability to direct research funds towards the identified research priorities in its strategic plan. Two strategies were developed to address these problems: a group of donors agreed to provide unearmarked pooled core funding, and accept a single common report based on an agreed MEF. On review after three years, there had been significant increases in total revenue, and the ability to allocate greater amounts of money on capacity building and infrastructure. The MEF demonstrated progress against strategic objectives, and better alignment of research against strategic priorities. There had also been changes in the sense of ownership and collaboration between ICDDR,B's management and its core donors. Conclusions The

  7. Research reactor core conversion guidebook. V.2: Analysis (Appendices A-F)

    International Nuclear Information System (INIS)

    Volume 2 consists of detailed Appendices, covering safety analyses for generic 10 MW reactor, safety analysis - probabilistic methods, methods for preventing LOCA, radiological consequence analyses, examples of safety report amendments and safety specifications. Included in Volume 2 are example analyses for cores with with highly enriched uranium and low enriched uranium fuels showing differences that can be expected in the safety parameters and radiological consequences of postulated accidents. There are seven examples of licensing documents related to core conversion and two examples of methods for determining power limits for safety specifications in the document. Refs, figs, bibliographies and tabs

  8. Dual-Band Annular-Ring Microstrip Patch Antenna for Satellite Applications

    Directory of Open Access Journals (Sweden)

    Tvs Divakar

    2014-08-01

    Full Text Available A dual-band circularly polarized antenna fed by four apertures that covers the bands of GPS, Galileo, is introduced. The ARSAs designed using FR4 substrates in the L and S bands have 3-dB axial-ratio bandwidths (ARBWs of as large as 37% and 52%, respectively, whereas the one using an RT5880 substrate in the L band, 61%. In these 3-dB axial-ratio bands, impedance matching with VSWR<=1.8 is also achieved. Three wideband planar baluns are used to achieve good axial ratio and VSWR. The results of the annular-ring microstrip antenna show good performance of a dual-band operation, which meets the requirement of Global Navigation Satellite System (GNSS applications.

  9. A study of annular flows with bubbles in the liquid ring and entrained droplets by means of stochastic analysis techniques

    International Nuclear Information System (INIS)

    By employing stochastic analysis techniques, an experimental study of a large number of annular flows with bubbles in the liquid ring and entrained droplets has been undertaken in the experimental air-water loop FREDLI, in which the information carrier is two visible light beams crossing the diameter of the tube and modulated by the scattering of the photons at the randomly arriving interfaces; also, some earlier neutron noise measurements in the upper part of a commercial BWR core are carefully analyzed. For the BWR measurements, it is shown for the first time that in the upper part of the core, there are usually three peaks in the cross-correlation function and that all noise analytic functions look extraordinarily similar to the corresponding noise analytic functions of some of the investigated annular flows at the FREDLI loop; a plausible explanation of these findings is given. (Auth.)

  10. Optimization of geometry of annular seat valves suitable for Digital Displacement fluid power pumps/motors

    DEFF Research Database (Denmark)

    Rømer, Daniel; Johansen, Per; Pedersen, Henrik C.;

    2013-01-01

    Digital Displacement Fluid Power is an upcoming technology setting new standards for the achievable efficiency of fluid power pumps and motors. The core element of the Digital Displacement technology is high performance electronically controlled seat valves, which must exhibit very low flow...... work an annular seat valve suitable for use in Digital Displacement units is considered, and the ring geometry is optimized using finite element analysis including non-linear material behaviour, contact elements and fluid pressure penetrating load, closely reflecting the actual load of the seat valve...

  11. Interfacial friction in cocurrent upward annular flow. Final report. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hossfeld, L.M.; Bharathan, D.; Wallis, G.B.; Richter, H.J.

    1982-03-01

    Cocurrent upward annular flow is investigated, with an emphasis on correlating and predicting pressure drop. Attention is given to the characteristics of the liquid flow in the film, and the interaction of the core with the film. Alternate approaches are discussed for correlating suitably defined interfacial friction factors. Both approaches are dependent on knowledge of the entrainment in order to make predictions. Dimensional analysis is used to define characteristic parameters of the flow and an effort is made to determine, to the extent possible, the influences of these parameters on the interfacial friction factor.

  12. Detection and analysis of transition from annular to intermittent flow in vertical tubes

    International Nuclear Information System (INIS)

    In vertical co-current gas-liquid flow, the transition from annular to intermittent flow occurs when gas core becomes interrupted by liquid bridges due to the instability of the interfacial capillary waves. An analytical model is formulated to explain the liquid bridging in terms of the growth of finite amplitude interfacial capillary waves. Experimental results show that the longest wave length, which is associated with the transition, is about eight times the wave length of waves moving with the velocity of the liquid film. (author). 12 refs., 8 figs

  13. Effect of entrained liquid on turbulent mixing rate between subchannels in annular two-phase flows

    International Nuclear Information System (INIS)

    Turbulent mixing rates of gas and liquid phases between the subchannels have been measured for various air-water two-phase annular flows in a multiple channel consisting of the two identical circular subchannels. In order to study effect of entrained liquid in the gas core on the turbulent mixing rates, experiments were conducted for two types of liquid injection method, i.e., a small bore nozzle placed in the subchannel center and a porous wall, at a fixed gas injection method. The result showed that the effect of entrained liquid on the turbulent mixing rates of both phases is negligibly small. (author)

  14. Entrained liquid fraction prediction in adiabatic and evaporating annular two-phase flow

    International Nuclear Information System (INIS)

    Highlights: ► New method to predict the entrained liquid fraction in annular two-phase flow. ► Circular and non-circular tubes, adiabatic and evaporating conditions covered. ► Large underlying experimental database (2460 points). ► New method explicit and fully stand-alone. ► New method based on just 1 dimensionless group: the core flow Weber number. - Abstract: A new method to predict the entrained liquid fraction in annular two-phase flow is presented. The underlying experimental database contains 2460 data points collected from 38 different literature studies for 8 different gas–liquid or vapor–liquid combinations (R12, R113, water–steam, water–air, genklene–air, ethanol–air, water–helium, silicon–air), tube diameters from 5.0 mm to 95.3 mm, pressures from 0.1 to 20.0 MPa and covers both adiabatic and evaporating flow conditions, circular and non-circular channels and vertical upflow, vertical downflow and horizontal flow conditions. Annular flows are regarded here as a special form of a liquid atomization process, where a high velocity confined spray, composed by the gas phase and entrained liquid droplets, flows in the center of the channel dragging and atomizing the annular liquid film that streams along the channel wall. Correspondingly, the liquid film flow is assumed to be shear-driven and the energy required to drive the liquid atomization is assumed to be provided in the form of kinetic energy of the droplet-laden gas core flow, so that the liquid film–gas core aerodynamic interaction is ultimately assumed to control the liquid disintegration process. As such, the new prediction method is based on the core flow Weber number, representing the ratio of the disrupting aerodynamic force to the surface tension retaining force, a single and physically plausible dimensionless group. The new prediction method is explicit, fully stand-alone and reproduces the available data better than existing empirical correlations, including in

  15. Research Commentary: Educational Technology--An Equity Challenge to the Common Core

    Science.gov (United States)

    Kitchen, Richard; Berk, Sarabeth

    2016-01-01

    The implementation of the Common Core State Standards for Mathematics (National Governors Association Center for Best Practices & Council of Chief State School Officers, 2010) has the potential to move forward key features of standards-based reforms in mathematics that have been promoted in the United States for more than 2 decades (e.g.,…

  16. Recurrent Annular Peripheral Choroidal Detachment after Trabeculectomy

    Directory of Open Access Journals (Sweden)

    Shaohui Liu

    2013-10-01

    Full Text Available We report a challenging case of recurrent flat anterior chamber without hypotony after trabeculectomy in a 54-year-old Black male with a remote history of steroid-treated polymyositis, cataract surgery, and uncontrolled open angle glaucoma. The patient presented with a flat chamber on postoperative day 11, but had a normal fundus exam and intraocular pressure (IOP. Flat chamber persisted despite treatment with cycloplegics, steroids, and a Healon injection into the anterior chamber. A transverse B-scan of the peripheral fundus revealed a shallow annular peripheral choroidal detachment. The suprachoroidal fluid was drained. The patient presented 3 days later with a recurrent flat chamber and an annular peripheral choroidal effusion. The fluid was removed and reinforcement of the scleral flap was performed with the resolution of the flat anterior chamber. A large corneal epithelial defect developed after the second drainage. The oral prednisone was tapered quickly and the topical steroid was decreased. One week later, his vision decreased to count fingers with severe corneal stromal edema and Descemet's membrane folds that improved to 20/50 within 24 h of resumption of the oral steroid and frequent topical steroid. The patient's visual acuity improved to 20/20 following a slow withdrawal of the oral and topical steroid. Eight months after surgery, the IOP was 15 mm Hg without glaucoma medication. The detection of a shallow anterior choroidal detachment by transverse B-scan is critical to making the correct diagnosis. Severe cornea edema can occur if the steroid is withdrawn too quickly. Thus, steroids should be tapered cautiously in steroid-dependent patients.

  17. Annular MHD Physics for Turbojet Energy Bypass

    Science.gov (United States)

    Schneider, Steven J.

    2011-01-01

    The use of annular Hall type MHD generator/accelerator ducts for turbojet energy bypass is evaluated assuming weakly ionized flows obtained from pulsed nanosecond discharges. The equations for a 1-D, axisymmetric MHD generator/accelerator are derived and numerically integrated to determine the generator/accelerator performance characteristics. The concept offers a shockless means of interacting with high speed inlet flows and potentially offers variable inlet geometry performance without the complexity of moving parts simply by varying the generator loading parameter. The cycle analysis conducted iteratively with a spike inlet and turbojet flying at M = 7 at 30 km altitude is estimated to have a positive thrust per unit mass flow of 185 N-s/kg. The turbojet allowable combustor temperature is set at an aggressive 2200 deg K. The annular MHD Hall generator/accelerator is L = 3 m in length with a B(sub r) = 5 Tesla magnetic field and a conductivity of sigma = 5 mho/m for the generator and sigma= 1.0 mho/m for the accelerator. The calculated isentropic efficiency for the generator is eta(sub sg) = 84 percent at an enthalpy extraction ratio, eta(sub Ng) = 0.63. The calculated isentropic efficiency for the accelerator is eta(sub sa) = 81 percent at an enthalpy addition ratio, eta(sub Na) = 0.62. An assessment of the ionization fraction necessary to achieve a conductivity of sigma = 1.0 mho/m is n(sub e)/n = 1.90 X 10(exp -6), and for sigma = 5.0 mho/m is n(sub e)/n = 9.52 X 10(exp -6).

  18. The effect of inlet swirl on the dynamics of long annular seals in centrifugal pumps

    Science.gov (United States)

    Ismail, M.; Brown, R. D.; France, D.

    1994-01-01

    This paper describes additional results from a continuing research program which aims to identify the dynamics of long annular seals in centrifugal pumps. A seal test rig designed at Heriot-Watt University and commissioned at Weir Pumps Research Laboratory in Alloa permits the identification of mass, stiffness, and damping coefficients using a least-squares technique based on the singular value decomposition method. The analysis is carried out in the time domain using a multi-fiequency forcing function. The experimental method relies on the forced excitation of a flexibly supported stator by two hydraulic shakers. Running through the stator embodying two symmetrical balance drum seals is a rigid rotor supported in rolling element bearings. The only physical connection between shaft and stator is the pair of annular gaps filled with pressurized water discharged axially. The experimental coefficients obtained from the tests are compared with theoretical values.

  19. Technical meeting on assessment of core structural materials and surveillance programme of research reactors. Program and abstracts

    International Nuclear Information System (INIS)

    Research reactors have played and continue to play a key role in the development of the peaceful uses of nuclear energy and technology, particularly in various domains of research as, fundamental and applied science, industry, human health care and environmental studies, as well as nuclear energy applications and the development of nuclear science and technology related human resources. However, more than 50% of operating research reactors today are over 40 years old and continued operation has to be carefully assessed, especially from the structural materials point of view. In many instances data for the radiation-induced changes of research reactor core materials resulting from exposure to very high neutron fluences are not generally available. Further data is needed in order to evaluate the reliability of research reactor core components. Age-related degradation mechanisms can cause unplanned outages of the research reactors which could in many cases have been predicted by implementation of appropriate surveillance programs. Typically, neutron-based irradiation programmes are carried out at research reactors for several purposes, with particular attention to structural and moderator materials and fuel samples from conventional nuclear power plants. The aim of such experiments is to determine the neutron fluence effects on mechanical properties of materials. Research and development of new advanced materials is also carried out and many member states with research reactors are involved or interested in such R and D projects. Unfortunately, very little information from analysed structural materials can be used as inputs to evaluating research reactor structural materials because of marked differences in the materials and operating environment between power reactors and research reactors. However, the methods used in such programs could be applied to research reactors, especially in the preparation of a predictive/preventive maintenance program supporting extended

  20. Research and development of a super fast reactor. (2) Core design improvement on local void reactivity

    International Nuclear Information System (INIS)

    A 700MWe Supercritical-pressure water-cooled fast reactor (Super Fast Reactor) was designed with negative overall void reactivity. As there is no cross flow between the fuel assemblies, the local void reactivity, defined as the reactivity change when the coolant of one assembly disappears, also need to be kept negative throughout the cycle. In this study, we found out the mechanism of the local void reactivity and improved the core design to keep the local void reactivity negative for all the seed fuel assemblies. According to the theory analysis, several core configurations, including the thickness of ZrH layer, the layout of the seed fuel assembly, the layout of the core and the loading pattern, will affect the local void reactivity distribution. Sensitivity of those configurations on the local void reactivity was analyzed. 1.15cm of ZrH layer thickness is the best choice for reducing the local void reactivity for the current core design. The assembly layout has no obvious effect on the local void reactivity. It is necessary to load more blanket assemblies in the inner region of the core in order to reduce the local void reactivity of the inner seed fuel assemblies. Loading pattern is also important for flattening the local void reactivity distribution. A hybrid loading method can be employed to make the distribution of the local void reactivity more uniform. Based on those conclusions, a Super Fast Reactor is successfully designed with satisfying all of the design criteria and design goals as well as keeping the local void reactivity of all the seed fuel assemblies less than -30pcm. (author)

  1. Three-dimensional free vibration analysis of carbon nanotube reinforced composites annular plates

    OpenAIRE

    Hakimeh Zali; Fatemeh Yazdian; Meisam Omidi

    2016-01-01

    The main objective of this research work was to investigate three-dimensional free vibration of thick annular plates which are composed of carbon nanotube (CNT) reinforced composites materials using the Chebyshev–Ritz method. In order to obtain precise results, a new form of the rule of mixtures including an exponential shape function, length efficiency parameter, orientation efficiency factor, and waviness parameter was applied for predicting the mechanical properties of CNT reinforced compo...

  2. A New Approach to Designing the S-Shaped Annular Duct for Industrial Centrifugal Compressor

    OpenAIRE

    Ivan Yurko; German Bondarenko

    2014-01-01

    The authors propose an analytical method for designing the inlet annular duct for an industrial centrifugal compressor using high-order Bezier curves. Using the design of experiments (DOE) theory, the three-level full factorial design was developed for determination of influence of the dimensionless geometric parameters on the output criteria. Numerical research was carried out for determination of pressure loss coefficients and velocity swirl angles using the software system ANSYS CFX. Optim...

  3. Influence of the Southern Annular Mode on the sea ice-ocean system

    OpenAIRE

    W. Lefebvre; Goosse, H.; Timmermann, R.; Fichefet, T.

    2004-01-01

    [1] The global sea ice - ocean model ORCA2-LIM, driven by the NCEP/NCAR ( National Centers for Environmental Prediction-National Center for Atmospheric Research) reanalysis daily 2-m air temperatures and 10-m winds and by monthly climatologies for precipitation, cloud cover, and relative humidity, is used to investigate the impact of the Southern Annular Mode (SAM) on the Antarctic sea ice-ocean system. Our results suggest that the response of the circumpolar Southern Ocean consists of an ann...

  4. Whole core analysis of an open pool research reactor under the most severe loss of coolant accident conditions

    International Nuclear Information System (INIS)

    In the present work the accident in which either the outlet or the inlet coolant pipe connected to the bottom of the reactor tank in an open pool research reactor is completely ruptured has been analyzed. The 3-D transient computer code ThEAP-I developed at Democritus NRC has been utilized and applied to the 5 MW Greek Research Reactor (GRR-1). The results show that a partial melting of the reactor core is possible for the GRR-1, the amount of melting being roughly and conservatively estimated to be of the order of 20%. (author)

  5. Study on Reduced-Moderation Water Reactor (RMWR) core design. Joint research report (FY1998-1999)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-09-01

    The Reduce-Moderation Water Reactor (RMWR) is a next generation water-cooled reactor aiming at effective utilization of uranium resource, high burn-up and long operation cycle, and plutonium multi-recycle. Japan Atomic Energy Research Institute (JAERI) started a joint research program for conceptual design of RMWR core in collaboration with the Japan Atomic Power Company (JAPC) since 1998. The research area includes the RMWR core conceptual designs, development of analysis methods for rector physics and thermal-hydraulics to design the RMWR cores with higher accuracy and preparation of MOX critical experiment to confirm the feasibility from the reactor physics point of view. The present report describes the results of joint research program 'RMWR core design Phase 1' performed by JAERI and JAPC in FY 1998 and 1999. The results obtained from the joint research program are as follows: Conceptual design study on the RMWR core has been performed. A core concept with a conversion ratio more than about 1 is basically feasible to multiple recycling of plutonium. Investigating core characteristics at the equilibrium, some promising core concepts to satisfy above aims have been established. As for BWR-type concepts with negative void reactivity coefficients, three types of design have been obtained as follows; (1) one feasible to attain high conversion ratio about 1.1, (2) one feasible to attain operation cycle of about 2 years and burn-up of about 60 GWd/t with conversion ratio more than 1 or (3) one in simple design based on the ABWR assembly and without blanket attaining conversion ratio more than 1. And as for PWR-type concepts with negative void reactivity coefficients, two types of design have been obtained as follows; (1) one feasible to attain high conversion ratio about 1.05 by using heavy water as a coolant and (2) one feasible to attain conversion ratio about l by using light water. In the study of nuclear calculation method, a reactor analysis code

  6. Study on Reduced-Moderation Water Reactor (RMWR) core design. Joint research report (FY1998-1999)

    International Nuclear Information System (INIS)

    The Reduce-Moderation Water Reactor (RMWR) is a next generation water-cooled reactor aiming at effective utilization of uranium resource, high burn-up and long operation cycle, and plutonium multi-recycle. Japan Atomic Energy Research Institute (JAERI) started a joint research program for conceptual design of RMWR core in collaboration with the Japan Atomic Power Company (JAPC) since 1998. The research area includes the RMWR core conceptual designs, development of analysis methods for rector physics and thermal-hydraulics to design the RMWR cores with higher accuracy and preparation of MOX critical experiment to confirm the feasibility from the reactor physics point of view. The present report describes the results of joint research program 'RMWR core design Phase 1' performed by JAERI and JAPC in FY 1998 and 1999. The results obtained from the joint research program are as follows: Conceptual design study on the RMWR core has been performed. A core concept with a conversion ratio more than about 1 is basically feasible to multiple recycling of plutonium. Investigating core characteristics at the equilibrium, some promising core concepts to satisfy above aims have been established. As for BWR-type concepts with negative void reactivity coefficients, three types of design have been obtained as follows; 1) one feasible to attain high conversion ratio about 1.1, 2) one feasible to attain operation cycle of about 2 years and burn-up of about 60 GWd/t with conversion ratio more than 1 or 3) one in simple design based on the ABWR assembly and without blanket attaining conversion ratio more than 1. And as for PWR-type concepts with negative void reactivity coefficients, two types of design have been obtained as follows; 1) one feasible to attain high conversion ratio about 1.05 by using heavy water as a coolant and 2) one feasible to attain conversion ratio about l by using light water. In the study of nuclear calculation method, a reactor analysis code applicable to the

  7. RESEARCH AND PRACTICE ON NONDESTRUCTIVE FLAW DETECTION INSTALLATION FOR WIRE-CORE BELT

    Institute of Scientific and Technical Information of China (English)

    刘志河; 张海涛; 绍庆龙

    1997-01-01

    Electromagnetic self-induction theory and computer are adopted and study of online monitoring technique for wire-core belt is conducted, the study shows that there is direct proportion between distance I of broken ends and output volt V, when I≥60 mm, V keeps constantly, the running speed v of wire-core belt has no big effect on output volt V, there is inverse proportion between the height h from probe to the surface of the belt and output volt V, when h≥30mm, V tends to be zero. Based on the test result, on-line monitoring installation is developed, the practice proved that the accuracy of broken wire monitoring can be above 95%, the monitoring accuracy of joint twitch can be 0.04 V/mm.

  8. Research on burn control of core plasma with the transport code

    International Nuclear Information System (INIS)

    For the fusion reactors or experimental devices, one will be required to control several plasma parameters, like the fusion power, the heat flux, the neutron flux, the beta-value and so on. To control these parameters, many diagnostics and actuators are needed, but the diagnostics and actuators available in DEMO/commercial reactors are limited because of the high heat or neutron flux. For these reasons, to realize the fusion reactors, the construction of the reactor control logic is required. We are developing the burn control logic in the core plasma with a 1.5D transport code, and discussing on the relationship between control parameters and actuators. To demonstrate the feasibility of the core plasma control, we have demonstrated the simultaneous control of the fusion power and the safety factor profile with the gas-puff and NBI. (author)

  9. Research Information Standardization as a Wicked Problem: Possible Consequences for the Standardization Process. Case Study of the Specification Project of the German Research Core Dataset

    OpenAIRE

    Riechert, Mathias; Dees, Werner

    2014-01-01

    Delivered at the CRIS2014 Conference in Rome; published in Procedia Computer Science 33 (Jul 2014). Contains conference paper (6 pages) and presentation (24 slides) In this paper, we analyze the degree to which the standardization project “Research core dataset” (CDS) qualifies as a wicked problem. The project was initiated in 2013 by the German Science Council. We suggest three solution approaches and discuss their application to standardization processes.

  10. Analysis of facilities in OFF research in participating countries of CORE Organic

    OpenAIRE

    Nykänen, Arja; Canali, Stefano

    2006-01-01

    Report lists the following research facilities: research farms, experimental fields, on-farm studies, networks, animal research facilities, leaching fields and long-term experiments. Other facilities like facilities for laboratory analyses, food processing, greenhouses, climate chambers and growth cabinets are left out from this analysis, because they are seldom exclusively used for OFF research and because their use for OFF research does not require particular characteristics. On the other h...

  11. Research and Development for Demonstration of Fuel Performance in the BREST-OD-300 Core

    International Nuclear Information System (INIS)

    Demonstration of mixed nitride core performance is an integral part of the 300 MW(e) lead-cooled fast reactor (BREST-OD-300) design. In 2010, NIKIET carried out R and D for the reactor design with BR=CBR∼1.05, in which low- background (taken from the BN-600 core) plutonium is used for fabrication of the first core. The investigations included design studies of shrouded hexagonal fuel assemblies with gas-bonded (U-Pu)N fuel and fitted up with wire or rib spacers; development of experimental dismountable fuel assembly for testing in BOR-60 reactor; improvements in the design of BOR-60 independent lead-cooled loop channel, in which 3 fuel pins of BREST-OD-300 will be tested. In 2011, it is planned to develop BREST-OD-300 fuel with energy-grade Pu dioxide from irradiated fuel of VVERs; to analyze the startup of a large lead-cooled reactor (≥1 GW(e)) on uranium nitride fuel of moderate enrichment and the conditions of changeover to (U-Pu-MA)N to be regenerated from the reactor's own irradiated fuel; to fabricate experimental BOR-60 dismountable assembly with (U-Pu)N. (author)

  12. 科研组织的核心竞争力体系研究%Research on Core Competences System in Scientific Research Organizations

    Institute of Scientific and Technical Information of China (English)

    霍国庆; 董帅; 肖建华; 谢晔

    2011-01-01

    应用资源基础理论,系统分析了科研组织必需的关键资源和关键能力,重新探讨了核心竞争力的识别标准,并据此分析了科研组织核心竞争力的构成,提出科研组织的核心竞争力是由战略领导力、人才凝聚力、科教激发力、科研协同力与合作竞争力等构成的完整体系,最后以著名的卡文迪什实验室为案例分析了其核心竞争力体系.%Based on resource-based view, this paper concludes the key resources and key capacities in scientific research organizations.According to criteria of core competencies, five core competences are identified, namely strategic leadership competence, talent cohesion competence, research & teaching inspiration competence, research synergy competence, and co-competition competence.Finally, the paper analyzes core competences system of Cavendish laboratory.

  13. Development of Core Design Model for Small-Sized Research Reactor and Establishment of Infrastructure for Reactor Export

    International Nuclear Information System (INIS)

    Within 10 years a growing world-wide demand of new research reactor construction is expected because of obsolescence. In Korea, a new research reactor is also required in order to meet domestic demand of utilization. KAERI has been devoted to develop an export-oriented research reactors for these kinds of demand. A next generation research reactor should comply with general requirements for safety, economics, environment-friendliness and non-proliferation as well as high performance requirement of high flux level. A export-tailored reactor should be developed for the demand of developing counties or under-developed countries. A new design concept is to be developed for a long cycle length core which has excellent irradiation facility with high flux

  14. A coupled calculational system for optimal in-core fuel management in research reactors

    International Nuclear Information System (INIS)

    In-core fuel management is one of the frequently complex task faced during a useful life of a nuclear reactor, because of the huge number of possible patterns (configurations) existing for a particular fuel set and its associated decision which should be taken about the potential good pattern satisfying established safety constraints. Thus, a sensible way to handle safely such kind of problem one have to provide an automatic procedure to generate appropriate reload pattern in nuclear reactor core. In the present work, the investigated way followed to solve this problem, was the combination use of the well known stochastic method which is Simulated Annealing (SA), together with Artificial Neural Network (ANN) technique. The strategy, which is inspired from Kim's paper, was proposed to solve adequately this problem. It requires two calculational stages involved by the use of coupled method connected to each other. In the first stage, an adaptive back-propagation network (BPN), is used to predicts safety core parameters Pmax and Keff. The BPN receives the allowed configurations from a previous calculation using heuristics rules and thereafter predicts Pmax and Keff very quickly. The Simulated Annealing method, in a second stage, determines whether a current candidate is better than the reference one based on the predicted results and consequently on value of the objective function stated. The most distinguishing and attractive feature of such system is the computational efficiency in obtaining optimized loading patterns with adequate fidelity. Neural network offers very fast prediction of core parameters with reasonable accuracy, and simulated annealing method offers very effective searching procedure which avoid local minimum. An objective function was developed based on two performance parameters: cycle length which can be determined through the evaluation of the effective multiplication factor Keff and power peaking factor Pmax. The system uses optimization of

  15. Development of an emergency core cooling system for the converted IEA-R1m research reactor

    International Nuclear Information System (INIS)

    This present work describes the development program carried out in the design and construction of the Emergency Core Cooling System for the IEA-R1m Research Reactor, including the system design, the experiments performed to validate the design, manufacturing, installation and commissioning. The experiments were performed in two phases. In the first phase, the spray flow rate and distribution were measured, using a full scale mock-up of the entire core, to establish the spray header geometry and specifications. In the second phase, a test section was fitted with electrically heated plates to simulate the fuel plates. Temperature measurements were carried out to demonstrate the effectiveness of the system to keep the temperatures below the limiting value. The experimental results were shown to the licensing authorities during the certification process. The main difficulties during the system assembly are also described. (author)

  16. Compendium of ECCS [Emergency Core Cooling Systems] research for realistic LOCA [loss-of-coolant accidents] analysis: Final report

    International Nuclear Information System (INIS)

    In the United States, Emergency Core Cooling Systems (ECCS) are required for light water reactors (LWRs) to provide cooling of the reactor core in the event of a break or leak in the reactor piping or an inadvertent opening of a valve. These accidents are called loss-of-coolant accidents (LOCA), and they range from small leaks up to a postulated full break of the largest pipe in the reactor cooling system. Federal government regulations provide that LOCA analysis be performed to show that the ECCS will maintain fuel rod cladding temperatures, cladding oxidation, and hydrogen production within certain limits. The NRC and others have completed a large body of research which investigated fuel rod behavior and LOCA/ECCS performance. It is now possible to make a realistic estimate of the ECCS performance during a LOCA and to quantify the uncertainty of this calculation. The purpose of this report is to summarize this research and to serve as a general reference for the extensive research effort that has been performed. The report: (1) summarizes the understanding of LOCA phenomena in 1974; (2) reviews experimental and analytical programs developed to address the phenomena; (3) describes the best-estimate computer codes developed by the NRC; (4) discusses the salient technical aspects of the physical phenomena and our current understanding of them; (5) discusses probabilistic risk assessment results and perspectives, and (6) evaluates the impact of research results on the ECCS regulations. 736 refs., 412 figs., 66 tabs

  17. Core concepts for 'zero-sodium-void-worth core' in metal fuelled fast reactor

    International Nuclear Information System (INIS)

    Core design options to reduce the sodium void worth in metal fuelled LMRs are investigated. Two core designs which achieve a zero sodium void worth are analyzed in detail. The first design is a 'pancaked' and annular core with enhanced transuranic burning capabilities; the high leakage in this design yields a low breeding ratio and small void worth. The second design is an axially multilayered annular core which is fissile self-sufficient; in this design, the upper and lower core regions are neutronically decoupled for reduced void worth while fissile self-sufficiency is achieved using internal axial blankets plus external radial and axial blanket-zones. The neutronic performance characteristics of these low void worth designs are assessed here; their passive safety properties are discussed in a companion paper. (author)

  18. Irradiation Test Plan of the Dual Cooled UO{sub 2} Annular Pellets

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Je Geon; Kim, Dae Ho; Chun, Tae Hyun; Kim, Keon Sik; Kim, Hyung Kyu; In, Wang Ki; Yang, Yong Sik; Song, Kun Woo; Chae, Hee Taek; Seo, Chul Gyo

    2008-09-15

    In order to study the behavior of the UO{sub 2} annular pellet developed by the high performance fuel technology development project, irradiation test will be carried out in HANARO research reactor for 5 cycles up to the burnup 12 MWD/kgU. After irradiation test in HANARO, the test fuel rod will be transferred to the hot cell and examined to verify the in-pile behavior. For the irradiation test, new irradiation test rig was designed and manufactured. The out-pile verification test and safety evaluation were performed and the results showed that the test rig and test rod will maintain the integrity and satisfy all the safety requirements during irradiation test. Therefore, it is expected that UO{sub 2} annular fuel can be irradiated safely in HANARO.

  19. A New Approach to Designing the S-Shaped Annular Duct for Industrial Centrifugal Compressor

    Directory of Open Access Journals (Sweden)

    Ivan Yurko

    2014-01-01

    Full Text Available The authors propose an analytical method for designing the inlet annular duct for an industrial centrifugal compressor using high-order Bezier curves. Using the design of experiments (DOE theory, the three-level full factorial design was developed for determination of influence of the dimensionless geometric parameters on the output criteria. Numerical research was carried out for determination of pressure loss coefficients and velocity swirl angles using the software system ANSYS CFX. Optimal values of the slope for a wide range of geometric parameters, allowing minimizing losses in the duct, have been found. The study has used modern computational fluid dynamics techniques to develop a generalized technique for future development of efficient variable inlet guide vane systems. Recommendations for design of the s-shaped annular duct for industrial centrifugal compressor have been given.

  20. Annular flow of cement slurries; Escoamento anular de pastas de cimento

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Maria das Gracas Pena; Martins, Andre Leibsohn; Oliveira, Antonio Augusto J. de [PETROBRAS, Rio de Janeiro (Brazil). Centro de Pesquisas. Setor de Tecnologia de Perfuracao

    1989-12-31

    This paper considers the analysis of laminar, transitory and turbulent flow regimes of cement slurries of various compositions flowing in annular sections. It is an experimental study to evaluate the performance of dozens of equations found in the literature that reflect the rheological behavior of non-Newtonian fluids, the dimensioning of annular sections, the delimitation of the transitory zone and the estimative of friction losses in the turbulent flow regime. A large-scale physical simulator (SHS-Surface Hydraulic Simulator), was designed and constructed at the PETROBRAS Research Center in order to obtain flow parameters. A computer program capable of analysing and drawing conclusions from the behavior of non-Newtonian fluids flowing in different geometries and energetic conditions was also developed. These were considered as essential stages for the development of the project. (author) 17 refs., 9 figs., 18 tabs.

  1. Three-dimensional free vibration analysis of carbon nanotube reinforced composites annular plates

    Directory of Open Access Journals (Sweden)

    Hakimeh Zali

    2016-05-01

    Full Text Available The main objective of this research work was to investigate three-dimensional free vibration of thick annular plates which are composed of carbon nanotube (CNT reinforced composites materials using the Chebyshev–Ritz method. In order to obtain precise results, a new form of the rule of mixtures including an exponential shape function, length efficiency parameter, orientation efficiency factor, and waviness parameter was applied for predicting the mechanical properties of CNT reinforced composites. Convergence of the Chebyshev–Ritz method was also checked. Numerical results are given and compared with the available literature and finite element method (FEM analysis. Results obtained from the other well-known theories (such as: Micro-Mechanical, Halpin, etc. are compared with the new form of the rule of mixtures results. Furthermore, the effects of CNT type, structures, diameter, shape factor, density, and volume fraction on the vibration behavior of the annular plates are graphically presented.

  2. The Springtime North Asia Cyclone Activity Index and the Southern Annular Mode

    Institute of Scientific and Technical Information of China (English)

    YUE Xu; WANG Huijun

    2008-01-01

    The relationship between the North Asia cyclone (NAC) activity and the Southern Annular Mode (SAM) is documented in this research. The definition of the NAC index (NACI) is based on the atmospheric relative vorticity in North Asia. The analysis yields a significant positive correlation between previous winter Southern Annular Mode index (SAMI) and spring NACI in the interannual variability, with a correlation coefficient of 0.51 during 1948-2000. Analysis of the NAC-related and SAM-related atmospheric general circulation variability demonstrates such a relationship. The study further reveals that when the winter SAM becomes strong, the springtime atmospheric convection in tropical western Pacific will intensify and the local Hadley circulation will be strengthened. As a result, the abnormal subsiding motion over South China makes the temperature gradient intensified in the low level and strengthens the jet in the high level, both of which are beneficial to the development of NAC activity.

  3. Annular linear induction pump with an externally supported duct

    International Nuclear Information System (INIS)

    An annular linear induction pump of increased efficiency is described, capable of being readily disassembled for repair or replacement of parts, and having one pass flow of the liquid metal through the pump. (U.K.)

  4. Principle of radial transport in low temperature annular plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Yunchao, E-mail: yunchao.zhang@anu.edu.au; Charles, Christine; Boswell, Rod [Space Plasma, Power and Propulsion Laboratory, Research School of Physics and Engineering, The Australian National University, Bldg 60, Mills Road, Australian Capital Territory 2601 (Australia)

    2015-07-15

    Radial transport in low temperature annular plasmas is investigated theoretically in this paper. The electrons are assumed to be in quasi-equilibrium due to their high temperature and light inertial mass. The ions are not in equilibrium and their transport is analyzed in three different situations: a low electric field (LEF) model, an intermediate electric field (IEF) model, and a high electric field (HEF) model. The universal IEF model smoothly connects the LEF and HEF models at their respective electric field strength limits and gives more accurate results of the ion mobility coefficient and effective ion temperature over the entire electric field strength range. Annular modelling is applied to an argon plasma and numerical results of the density peak position, the annular boundary loss coefficient and the electron temperature are given as functions of the annular geometry ratio and Paschen number.

  5. Axicon-based annular laser trap for studies on sperm activity

    Science.gov (United States)

    Shao, Bing; Vinson, Jaclyn M.; Botvinick, Elliot L.; Esener, Sadik C.; Berns, Michael W.

    2005-08-01

    As a powerful and noninvasive tool, laser trapping has been widely applied for the confinement and physiological study of biological cells and organelles. Researchers have used the single spot laser trap to hold individual sperm and quantitatively evaluated the motile force generated by a sperm. Early studies revealed the relationship between sperm motility and swimming behavior and helped the investigations in medical aspects of sperm activity. As sperm chemotaxis draws more and more interest in fertilization research, the studies on sperm-egg communication may help to explain male or female infertility and provide exciting new approaches to contraception. However, single spot laser trapping can only be used to investigate an individual target, which has limits in efficiency and throughput. To study the chemotactic response of sperm to eggs and to characterize sperm motility, an annular laser trap with a diameter of several hundred microns is designed, simulated with ray tracing tool, and implemented. An axicon transforms the wavefront such that the laser beam is incident on the microscope objective from all directions while filling the back aperture completely for high efficiency trapping. A trapping experiment with microspheres is carried out to evaluate the system performance. The power requirement for annular sperm trapping is determined experimentally and compared with theoretical calculations. With a chemo-attractant located in the center and sperm approaching from all directions, the annular laser trapping could serve as a speed bump for sperm so that motility characterization and fertility sorting can be performed efficiently.

  6. Annular-intermittent flow regime transition model and its application to boil-off pattern transition and dryout model

    International Nuclear Information System (INIS)

    A model is developed to describe the transition of annular flow to intermittent flow in a vertical two-phase flow system. The instability of the disturbance wave, which is a dominant wave shape at the boundary between annular flow and intermittent flow, is considered as the governing mechanism and this instability is described by the concept of hyperbolicity breaking in the characteristic equation. The developed model is validated by comparing its predictions of gas superficial velocity for the transition with the experimental data available from the literature, and comparing those with the predictions of the other correlations. The comparison results show that the developed model gives better predictions for the transition condition than the existing correlations and the effects of fluid properties, geometry and liquid flow rate on the transition are well considered by the developed model. It is found that the predictions of the developed model have much smaller bias than those of the other correlations; the average of the prediction error is 3% for the present model. The standard deviation of the prediction errors of the present model reaches 28%, which is the smallest among the models compared here. Through the core uncovery experiments, it has been known that the low power and high power core boil-off patterns are observed in the high pressure core uncovery following a small-break loss-of-coolant accident. The developed model for the annular to intermittent flow regime transition was applied to the classification of low power boil-off and high power boil-off patterns. At first, the applicability of the developed criterion to the rod-bundle geometry is demonstrated using the flow pattern transition data taken by Bergles et al. and Venkateswararao. It is shown that the developed criterion well predicts the boundary between low power boil-off and high power boil-off through the comparisons of the predicted annular to intermittent flow transition conditions with

  7. Annular elastolytic giant cell granuloma in association with Hashimoto's thyroiditis

    OpenAIRE

    Rishi Hassan; P Arunprasath; Padmavathy, L.; K Srivenkateswaran

    2016-01-01

    Annular elastolytic giant cell granuloma (AEGCG) is a rare granulomatous skin disease characterized clinically by annular plaques with elevated borders and atrophic centers found mainly on sun-exposed skin and histologically by diffuse granulomatous infiltrates composed of multinucleated giant cells, histiocytes and lymphocytes in the dermis along with phagocytosis of elastic fibers by multinucleated giant cells. We report a case of AEGCG in a 50-year-old woman and is highlighted for the clas...

  8. Annular bright and dark field imaging of soft materials

    International Nuclear Information System (INIS)

    Here polyethylene, as an example of an important soft material, was studied by STEM annular bright and dark field. The contrast as function of the probe size/shape and the detector collection angle are discussed. The results are compared to conventional bright field transmission electron microscopy, electron energy filtered imaging and energy dispersive spectroscopy mapping. Annular bright and dark field gave a higher contrast than conventional transmission and analytical mapping techniques

  9. Modeling of debris cooling with annular gap in the lower RPV and verification based on ALPHA experiments

    International Nuclear Information System (INIS)

    For severe accident assessment in a light water reactor, heat transfer models in a narrow annular gap between the overheated core debris and the reactor pressure vessel (RPV) are important for evaluating RPV integrity and emergency procedures. Using existing data, the authors developed heat transfer models on the average critical heat flux (CHF) restricted by countercurrent flow limitation (CCFL) and local boiling heat fluxes, and showed that the average CHF depended on the steam-water flow pattern in the narrow gap and that the local heat fluxes were similar to the pool boiling curve. We evaluated the validity of heat transfer models by simple calculations for ALPHA experiments performed at Japan Atomic Energy Research Institute. Calculated results showed that heat fluxes on the crust surface were restricted mainly by thermal resistance of the crust after the crust formation, and emissivity on the crust surface did not have much effect on the heat fluxes. The calculated vessel temperature during the heat-up process and peak vessel temperature agreed well with the measurements, which confirmed the validity of the average CHF correlation. However, the vessel cooling rate was underestimated mainly due to underestimation of the gap size.

  10. Safety research on iodine plate-out during postulated HTGR core heatup events

    International Nuclear Information System (INIS)

    In support of probabilistic risk assessment (PRA) studies on the high-temperature gas-cooled reactor (HTGR), an experimental program was conducted for iodine plateout on HTGR primary circuit metals during core heatup conditions. Metal iodine formation and adsorption characteristics were measured primarily for mild steel and to a limited extent for Incoloy 800 and other alloys. Pseudoisopiestic tests indicated quantitative formation of less volatile and water soluble iodides, FeI2 or CrI2, during core heatup conditions. The rate of formation of FeI2 was limited by mass transfer at temperatures above 5700K and was proportional to the partial pressure of iodine. The rate of iodide formation on chrome-nickel alloys appeared to be temperature sensitive, indicating slower reaction kinetics. The iodides preferentially plated out on surfaces at 520 to 620 K. Plateout tests were also performed for FeI2 in helium carrier gas flowing over mild steel or quartz surfaces over which a temperature gradient was maintained. PADLOC computer program correlations of the plateout profile based on the FeI2 vapor pressure assumed in the PRA studies were in fair agreement. The temperature at which most of the plateout occurred was from 620 to 700 K, depending on the partial pressure of the FeI2 tested. (author)

  11. New mitral annular force transducer optimized to distinguish annular segments and multi-plane forces.

    Science.gov (United States)

    Skov, Søren Nielsen; Røpcke, Diana Mathilde; Ilkjær, Christine; Rasmussen, Jonas; Tjørnild, Marcell Juan; Jimenez, Jorge H; Yoganathan, Ajit P; Nygaard, Hans; Nielsen, Sten Lyager; Jensen, Morten Olgaard

    2016-03-21

    Limited knowledge exists about the forces acting on mitral valve annuloplasty repair devices. The aim of this study was to develop a new mitral annular force transducer to measure the forces acting on clinically used mitral valve annuloplasty devices. The design of an X-shaped transducer in the present study was optimized for simultaneous in- and out-of-plane force measurements. Each arm was mounted with strain gauges on four circumferential elements to measure out-of-plane forces, and the central parts of the X-arms were mounted with two strain gauges to measure in-plane forces. A dedicated calibration setup was developed to calibrate isolated forces with tension and compression for in- and out-of-plane measurements. With this setup, it was possible with linear equations to isolate and distinguish measured forces between the two planes and minimize transducer arm crosstalk. An in-vitro test was performed to verify the crosstalk elimination method and the assumptions behind it. The force transducer was implanted and evaluated in an 80kg porcine in-vivo model. Following crosstalk elimination, in-plane systolic force accumulation was found to be in average 4.0±0.1N and the out-of-plane annular segments experienced an average force of 1.4±0.4N. Directions of the systolic out-of-plane forces indicated movements towards a saddle shaped annulus, and the transducer was able to measure independent directional forces in individual annular segments. Further measurements with the new transducer coupled with clinical annuloplasty rings will provide a detailed insight into the biomechanical dynamics of these devices. PMID:26903412

  12. Fabrication and Resintering of Annular UO2 Pellet

    International Nuclear Information System (INIS)

    Nuclear fuel is one of the most important components in a PWR affecting its safety and economy. The traditional PWR fuel pellet has a shape of cylindrical tablets of about 800 μm in diameter with a chamfer and dishes. A significant reduction in its failure rate has resulted from the improvements in fuel and cladding quality. Enhanced fuel assembly design allowed appreciable power density increases. However, it is difficult to achieve a significant increase of a power density under the current fuel pin design. Recently, Massachusetts Institute of Technology (MIT) has proposed an annular UO2 fuel with an internal cooling of each fuel rod. Annular fuel pellets with a voided central region have been used in VVER reactors without an internal cooling. Annular fuels with both internal and external cooling have been proposed for high temperature gas cooled reactors. However, commercial PWR reactors have not used such annular internally and externally cooled fuel rods, yet. There must be a lot of considerations in the various fields to introduce an annular internally and externally cooled fuel to commercial PWR reactors. The dimension tolerance and the thermal stability of a pellet are very important from the viewpoint of fabrication technology, because they have an influence on the size of the gap between the pellet and the inner/outer claddings. In this study, annular UO2 pellets with various densities were fabricated and then a resintering test was conducted. The changes of dimension and density of the sintered pellets were characterized

  13. Sonographic evaluation of digital annular pulley tears

    International Nuclear Information System (INIS)

    Objective. To evaluate the sonographic (US) appearance of digital annular pulley (DAP) tears in high-level rock climbers. Design and patients. We performed a retrospective analysis of the US examinations of 16 high-level rock climbers with clinical signs of DAP lesions. MRI and surgical evaluation were performed in five and three patients respectively. The normal US and MRI appearances of DAP were evaluated in 40 and three normal fingers respectively. Results. Nine of 16 patients presented a DAP tear. In eight subjects (seven with complete tears involving the fourth finger and one the fifth finger), US diagnosis was based on the indirect sign of volar bowstringing of the flexor tendons. Injured pulleys were not appreciated by US. Tears concerned the A2 and A3 in six patients and the A3 and A4 in two patients. A2 pulley thickening and hypoechogenicity compatible with a partial tear was demonstrated in one patient. MRI and surgical data correlated well with the US findings. Four patients had tenosynovitis of the flexor tendons but no evidence of pulley disruption. US examinations of three patients were normal. In the healthy subjects US demonstrated DAP in 16 of 40 digits. Conclusion. US can diagnose DAP tears and correlates with the MRI and surgical data. Because of its low cost and non-invasiveness we suggest US as the first imaging modality in the evaluation of injuries of the digital pulley. (orig.)

  14. Annular burnout data from rod bundle experiments

    International Nuclear Information System (INIS)

    Burnout data for annular flow in a rod bundle are presented for both transient and steady-state conditions. Tests were performed at the Oak Ridge National Laboratory in the Thermal Hydraulic Test Facility (THTF), a pressurized-water loop containing an electrically heated 64-rod bundle. The bundle configuration is typical of later generation pressurized-water reactors with 17 x 17 fuel arrays. Both axial and radial power profiles are flat. All experiments were carried out in upflow with subcooled inlet conditions, insuring accurate flow measurement. Conditions within the bundle were typical of those which could be encountered during a nuclear reactor loss-of-coolant accident. Level average fluid conditions within the test section were calculated using steady-state mass and energy conservation considerations for the steady-state tests and a transient, homogeneous, equilibrium computer code for the transient tests. Unlike tube dryout, burnout within a rod bundle does not necessarily occur at one distinct axial level. The location of individual rod dryout was determined by scanning rods axially and locating the position where rod superheat increased from approx. =0 to 30 K or greater. Thermocouple instrumentation within the bundle allows the location of dryout to be determined to within approximately +.5 cm for many of the tests

  15. JSPS-CAS Core University Program seminar on summary of 10-year collaborations in plasma and nuclear fusion research area

    International Nuclear Information System (INIS)

    The JSPS-CAS Core University Program (CUP) seminar on “Summary of 10-year Collaborations in Plasma and Nuclear Fusion Research Area” was held from March 9 to March 11, 2011 in the Okinawa Prefectural Art Museum, Naha city, Okinawa, Japan. The collaboration program on plasma and nuclear fusion started from 2001 under the auspices of Japanese Society of Promotion of Science (JSPS) and Chinese Academy of Sciences (CAS). This year is the last year of the CUP. This seminar was organized in the framework of the CUP. In the seminar, 29 oral talks were presented, having 14 Chinese and 30 Japanese participants. These presentations covered key topics related to the collaboration categories: (1) improvement of core plasma properties, (2) basic research on fusion reactor technologies, and (3) theory and numerical simulation. This seminar aims at summarizing the results obtained through the collaborations for 10 years, and discussing future prospects of China-Japan collaboration in plasma and nuclear fusion research areas. (author)

  16. The numerical calculation of heat transfer performance for annular flow of liquid nitrogen in a vertical annular channel

    Science.gov (United States)

    Sun, Shufeng; Wu, Yuyuan; Zhao, Rongyi

    2001-04-01

    According to a separated phase flow model for vertical annular two-phase flow in an annular channel, the liquid film thickness, distributions of velocities and temperatures in the liquid layer are predicted in the range of heat fluxes: 6000-12000 W/m 2, mass flux: 500-1100 kg/m2 s. The pressure drop along the flow channel and heat transfer coefficient are also calculated. The liquid film thickness is in the order of micrometers and heat transfer coefficient is 2800-7800 W/m2 K of liquid nitrogen boiling in narrow annular channels. The measured heat transfer coefficient is 29% higher than the calculated values. With the mass flux increasing and the gap of the annular channel decreasing, pressure drop and heat transfer coefficient increase.

  17. Operational safety experience at 14 MW TRIGA research reactor from INR Pitesti, Romania

    International Nuclear Information System (INIS)

    The safe operation of TRIGA-14 MW Core and Annular Pulsed TRIGA Core in the assembly of Research Reactor in Pitesti, Romania for 27 years is presented from historical perspective as well in the light of evolving safety experience. The accomplishment of safety objectives and responsibilities of operating organization is described and sustained with practical examples including management responsibilities, resources of management, performance indicators, measurement analysis and monitoring. Further improvement of safety of Research Reactor trough a large refurbishment and modernization program under way is also presented in the paper. (author)

  18. In-core fuel management, safety, and thermal hydraulics studies for upgrading TRIGA MARK II research reactor

    International Nuclear Information System (INIS)

    Bangladesh Atomic Energy Commission has approved a project to upgrade the research reactor to higher flux to meet the growing demand of medical radio-isotopes production and other irradiation facilities. Preliminary studies with the various core parameters showed that it might be possible to create new irradiation flux traps, increase the neutron flux at desired location, and at the same time the fuel burn-up can be made optimal. This will need major reshuffling and reconfiguration of the core with fuel rods initially loaded. The principal objective of this study is focused to make the above improvements in the core without disturbing the safety parameters. This presentation deals with the neutronic and thermal hydraulic analysis of the 3 MW TRIGA MARK II research reactor to upgrade it to a higher flux. To realize this objective, the overall strategy followed is: (I) generation of problem dependent cross section library from basic Evaluated Nuclear Data Files such as ENDF/B-VI, JENDL 3.2 with NJOY94.10+, (ii) use WIMSD-5 package to generate cell constants for all of the materials in the core and its immediate neighborhood, (iii) use CITATION to perform 3-D global analysis of the core to study multiplication factor, neutron flux and power distribution, power peaking factors, temperature reactivity coefficients, etc., (iv) check the validity of the deterministic codes with the Monte Carlo code MCNP-4B2, (v) couple output of CITATION with PARET to study thermal hydraulic behavior to predict safety margins, and (vi) reshuffle the current core configuration to achieve the desired objectives. The computational methods, tools and techniques, customization of cross section libraries, various models for cells and super cells, and a lot of associated utilities have been standardized and established/validated for the overall core analysis. Analyses using the 4-group, and 7-group libraries of macroscopic cross sections generated from the 69-group WIMSD-5 library were performed

  19. Green Infrastructure Research Promotes Students' Deeper Interest in Core Courses of a Water Resources Program

    Science.gov (United States)

    Yerk, W.; Montalto, F. A.; Foti, R.

    2015-12-01

    As one of most innovative among low impact development technologies, Green Infrastructure (GI) is a new technology that presents a range of potential research opportunities. Inherently linked to sustainability, urban quality of life, resilience, and other such topics, GI also represents a unique opportunity to highlight the social relevance of practical STEM research to undergraduate students. The nature of research on urban GI, in fact, as well as the accessibility of the GI sites, allows students to combine hands-on experience with theoretical work. Furthermore, the range of scales of the projects is such that they can be managed within a single term, but does not preclude longer engagement. The Sustainable Water Resource Engineering lab at Drexel University is engaged in two types of GI research outside the classroom. One type is a research co-op research internship. The second is a selective university-wide faculty-mentored summer scholarship STAR (Students Tackling Advanced Research) specifically designed for freshmen. The research projects we developed for those curricula can be accomplished by undergraduate students, but also address a larger research need in this emerging field. The research tasks have included identifying and calibrating affordable instruments, designing and building experimental setups, and monitoring and evaluating performance of GI sites. The work also promoted deeper understanding of the hydrological processes and initiated learning beyond the students' current curricula. The practice of the Lab's research being embedded into the educational process receives positive feedback from the students and achieves meaningful and long-lasting learning objectives. The experience helps students to students acquire hands-on experience, improves their metacognition and evidence-based inquiring into real-world problems, and further advances decision-making and communication skills.

  20. Predatory vs. Dialogic Ethics: Constructing an Illusion or Ethical Practice as the Core of Research Methods

    Science.gov (United States)

    Cannella, Gaile S.; Lincoln, Yvonna S.

    2007-01-01

    The ethical conduct of research is addressed from two perspectives, as a regulatory enterprise that creates an illusion of ethical practice and as a philosophical concern for equity and the imposition of power within the conceptualization and practice of research itself. The authors discuss various contemporary positions that influence…

  1. Georgia Institute of Technology research on the gas core actinide transmutation reactor (GCATR)

    International Nuclear Information System (INIS)

    The Gas Core Actinide Transmutation Reactor (GCATR) offers several advantages including (1) the gaseous state of the fuel may reduce problems of processing and recycling fuel and wastes, (2) high neutron fluxes are achievable, (3) the possibility of using a molten salt in the blanket may also simplify the reprocessing problem and permit breeding, (4) the spectrum can be varied from fast to thermal by increasing the moderation in the blanket so that the trade-off of critical mass versus actinide and fission product burnup can be studied for optimization, and (5) the U233-Th cycle, which can be used, appears superior to the U235-Pu cycle in regard to actinide burnup. The program at Georgia Tech is a study of the feasibility, design, and optimization of the GCATR

  2. Possible rim crest deposits in cores 12027 and 15008 - Some interpretations and problems for future research

    Science.gov (United States)

    Nagle, J. S.

    1980-01-01

    Light colored coarse grained soils in the upper parts of cores 12027 and 15008 have a chaotic fabric and abundant bedrock-derived rock fragments. They overlie dark colored, fine grained soils with horizontal fabric and abundant regolith-derived rock fragments. The light colored soils are interpreted to be rim crest deposits associated with the craters seen in the lunar surface sampling photographs. The coarse size, poor sorting and chaotic fabric are believed to have originated by the violent mixing processes which are expected with ejecta deposition. The increase in bedrock-derived particles within the light colored soil can be explained by ejecta from craters that penetrated through or deep into the regolith. The dark soils are believed to predate the craters. Distribution of rock fragments in the material interpreted as rim crest ejecta does not fit a simple model of overturned stratigraphy.

  3. [Research Progress in the Core Proteins of the Classical Swine Fever Virus].

    Science.gov (United States)

    Hou, Yuzhen; Zhao, Dantong; Liu, Guoying; He, Fan; Liu, Bin; Fu, Shaoyin; Hao, Yongqing; Zhang, Wenguang

    2015-09-01

    The core protein (CP) of the classical swine fever virus (CSFV) is one of its structural proteins. Apart from forming the nucleocapsid to protect internal viral genomic RNA, this protein is involved in transcriptional regulation. Also, during viral infection, the CP is involved in interactions with many host proteins. In this review, we combine study of this protein with its disorders, structural/functional characteristics, as well as its interactions with the non-structural proteins NS3, NS5B and host proteins such as SUMO-1, UBC9, OS9 and IQGAP1. We also summarize the important part played by the CP in CSFV pathogenicity, virulence and replication of genomic RNA. We also provide guidelines for further studies in the CP of the CSFV. PMID:26738299

  4. Research on instability design method without occurring boiling transition for hyper ABWR plants of extended core power density

    International Nuclear Information System (INIS)

    The hyper ABWR (Advanced Boiling Water Reactor) project aims to develop an advanced BWR concept that is competitive in the global market with both highly economic and safety features. Expecting plant construction within the coming ten years, a research program for substantiating the basic design of a high core power density ABWR was conducted. By inheriting the conventional ABWR design, it is possible to reduce construction costs. In order to achieve the rated core power of over 1650MWe which is almost equivalent to that of the EPR (European Pressurized Water Reactor), the core power density of ABWR will be up-rated by at least 25%. Three key subjects linked to this target were recognized. They are, (1) fuel design applicable to the high power density core, (2) improvement of the evaluation method for the coupled neutronic and thermal-hydraulic instability under a wider power-flow operating range, and (3) improvement of the steam separator performance under high quality conditions. In this paper, the second subject has been focused on. In the second subject, the uncertainty approach was introduced in the instability analysis where the best-estimate plant simulator was combined with a direct prediction of boiling transition by the sub-channel code. By employing the CSAU like method, a safety evaluation system that enables to include influences of uncertainties has been developed. Based on the correlation between the time margin for reaching the boiling transition under power oscillations and the decay ratio in the power-flow operation map, an automatic power oscillation suppressing system was designed. The set-point for activating suppression mechanisms (i.e. scram or SRI) could be determined based on this correlation. It was proposed that the present conservative acceptance criterion of the deterministic decay ratio can be replaced with a more rational one of the time margin with including uncertainties. (author)

  5. A two-dimensional parabolic model for vertical annular two-phase flow

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, F.M.; Toledo, A. Alvarez; Paladino, E.E. [Graduate Program in Mechanical Engineering, Universidade Federal de Rio Grande do Norte, Natal, RN (Brazil)], e-mail: emilio@ct.ufrn.br

    2010-07-01

    This work presents a solution algorithm for predicting hydrodynamic parameters for developing and equilibrium, adiabatic, annular, vertical two-phase flow. It solves mass and momentum transport differential equations for both the core and the liquid film across their entire domains. Thus, the velocity and shear stress distributions from the tube center to the wall are obtained, together with the average film thickness and the pressure gradient, making no use of empirical closure relations nor assuming any known velocity profile to solve the triangular relationship in the liquid film. The model was developed using the Finite Volume Method and an iterative procedure is proposed to solve all flow variables for given phase superficial velocities. The procedure is validated against the analytical solution for laminar flow and experimental data for gas-liquid turbulent flow with entrainment. For the last case, an algebraic turbulence model is used for turbulent viscosity calculation for both, liquid film and gas core. (author)

  6. Archive of Geosample Data and Information from the Ohio State University Byrd Polar and Climate Research Center (BPCRC) Sediment Core Repository

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The Byrd Polar and Climate Research Center (BPCRC) Sediment Core Repository operated by the Ohio State University is a partner in the Index to Marine and Lacustrine...

  7. The action researcher as a reflective partner to a core group

    DEFF Research Database (Denmark)

    Christensen, Dorthe; Sriskandarajah, Nadarajah

    The EU suggests applying bottom-up, participative learning approaches, such as Action Research, as steering instruments to meet the challenge of multifunctionality and its links with rural development. This paper focuses on the many demanding roles placed on an action researcher when working with...... rural stakeholders to achieve normatively desirable learning. It is suggested that in order to genuinely qualify the learning process and its outcome for all, the action researcher keeps an adequate balance between being “close to” or “inside” the stakeholder arena and “distanced to” or “outside” this...

  8. X-ray diffraction from bone employing annular and semi-annular beams

    International Nuclear Information System (INIS)

    There is a compelling need for accurate, low cost diagnostics to identify osteo-tissues that are associated with a high risk of fracture within an individual. To satisfy this requirement the quantification of bone characteristics such as ‘bone quality’ need to exceed that provided currently by densitometry. Bone mineral chemistry and microstructure can be determined from coherent x-ray scatter signatures of bone specimens. Therefore, if these signatures can be measured, in vivo, to an appropriate accuracy it should be possible by extending terms within a fracture risk model to improve fracture risk prediction.In this preliminary study we present an examination of a new x-ray diffraction technique that employs hollow annular and semi-annular beams to measure aspects of ‘bone quality’. We present diffractograms obtained with our approach from ex vivo bone specimens at Mo Kα and W Kα energies. Primary data is parameterized to provide estimates of bone characteristics and to indicate the precision with which these can be determined. (paper)

  9. Full Core modeling techniques for research reactors with irregular geometries using Serpent and PARCS applied to the CROCUS reactor

    International Nuclear Information System (INIS)

    Highlights: • Modeling of research reactors. • Serpent and PARCS coupling. • Lattice physics codes modeling techniques. - Abstract: This paper summarizes the results of modeling methodologies developed for the zero-power (100 W) teaching and research reactor CROCUS located in the Laboratory for Reactor Physics and Systems Behavior (LRS) at the Swiss Federal Institute of Technology in Lausanne (EPFL). The study gives evidence that the Monte Carlo code Serpent can be used effectively as a lattice physics tool for small reactors. CROCUS’ core has an irregular geometry with two fuel zones of different lattice pitches. This and the reactor’s small size necessitate the use of nonstandard cross-section homogenization techniques when modeling the full core with a 3D nodal diffusion code (e.g. PARCS). The primary goal of this work is the development of these techniques for steady-state neutronics and future transient neutronics analyses of not only CROCUS, but research reactors in general. In addition, the modeling methods can provide useful insight for analyzing small modular reactor concepts based on light water technology. Static computational models of CROCUS with the codes Serpent and MCNP5 are presented and methodologies are analyzed for using Serpent and SerpentXS to prepare macroscopic homogenized group cross-sections for a pin-by-pin model of CROCUS with PARCS. The most accurate homogenization scheme lead to a difference in terms of keff of 385 pcm between the Serpent and PARCS model, while the MCNP5 and Serpent models differed in terms of keff by 13 pcm (within the statistical error of each simulation). Comparisons of the axial power profiles between the Serpent model as a reference and a set of PARCS models using different homogenization techniques showed a consistent root-mean-square deviation of ∼8%, indicating that the differences are not due to the homogenization technique but rather arise from the definition of the diffusion coefficients produced

  10. Progress in analytical methods to predict and control azimuthal combustion instability modes in annular chambers

    Science.gov (United States)

    Bauerheim, M.; Nicoud, F.; Poinsot, T.

    2016-02-01

    Longitudinal low-frequency thermoacoustic unstable modes in combustion chambers have been intensively studied experimentally, numerically, and theoretically, leading to significant progress in both understanding and controlling these acoustic modes. However, modern annular gas turbines may also exhibit azimuthal modes, which are much less studied and feature specific mode structures and dynamic behaviors, leading to more complex situations. Moreover, dealing with 10-20 burners mounted in the same chamber limits the use of high fidelity simulations or annular experiments to investigate these modes because of their complexity and costs. Consequently, for such circumferential acoustic modes, theoretical tools have been developed to uncover underlying phenomena controlling their stability, nature, and dynamics. This review presents recent progress in this field. First, Galerkin and network models are described with their pros and cons in both the temporal and frequency framework. Then, key features of such acoustic modes are unveiled, focusing on their specificities such as symmetry breaking, non-linear modal coupling, forcing by turbulence. Finally, recent works on uncertainty quantifications, guided by theoretical studies and applied to annular combustors, are presented. The objective is to provide a global view of theoretical research on azimuthal modes to highlight their complexities and potential.

  11. Turbulence modification in vertical upward annular flow passing through a throat section

    International Nuclear Information System (INIS)

    Experimental studies on the turbulence modification in annular two-phase flow passing through a throat section were carried out. The turbulence modification in multi-phase flow due to the interactions between two-phases is one of the most interesting scientific issues and has attracted research attention. In this study, the gas-phase turbulence modification in annular flow due to the gas-liquid phase interaction is experimentally investigated. The annular flow passing through a throat section is under the transient state due to the changing cross sectional area of the channel and resultantly the superficial velocities of both phases are changed compared with a fully developed flow in a straight pipe. The measurements for the gas-phase turbulence were precisely performed by using a constant temperature hot-wire anemometer, and made clear the turbulence structure such as velocity profiles, fluctuation velocity profiles. The behavior of the interfacial waves in the liquid film flow such as the ripple or disturbance waves was also observed. The measurements for the liquid film thickness by the electrode needle method were also performed to measure the base film thickness, mean film thickness, maximum film thickness and wave height of the ripple or the disturbance waves

  12. Archive of Geosample Information from the GEOMAR Helmholtz Centre for Ocean Research Kiel Core Repository

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The GEOMAR Helmholtz Centre for Ocean Research Kiel made a one-time contribution to the Index to Marine and Lacustrine Geological Samples (IMLGS) database of...

  13. Core calculations for the upgrading of the IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Adimir dos; Perrotta, Jose A.; Bastos, Jose Luis F.; Yamaguchi, Mitsuo; Umbehaun, Pedro E. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)]. E-mail: asantos@net.ipen.br; perrotta@net.ipen.br; mitsuo@net.ipen.br

    1998-07-01

    The IEA-R1 Research Reactor is a multipurpose reactor. It has been used for basic and applied research in the nuclear area, training and radioisotopes production since 1957. In 1995, the Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) took the decision to modernize and upgrade the power from 2 to 5 MW and increase the operational cycle. This work presents the design requirements and the calculations effectuated to reach this goal. (author)

  14. Laboratory testing of rock core samples from pre-excavation grouting area at Mizunami Underground Research Laboratory

    International Nuclear Information System (INIS)

    Japan Atomic Energy Agency has been conducting a research project on 'Grouting Technology Development for the Radioactive Waste Repository' funded by Ministry of Economy, Trade and Industry (METI), Japan. As a part of the project, various investigations were carried out in the -200m Refuge Niche where pre-excavation grouting was performed and the distribution of the injected grouting material, also the effectiveness of grouting penetration for reduction of groundwater inflow were confirmed. As the continuation of these investigations, chemical influences of grouting material on the rock mass were determined through 'Laboratory testing of rock core samples from pre-excavation grouting area at Mizunami Underground Research Laboratory'. Specifically, core samples were obtained by check boring at where infiltration solidification of the grouting material was expected, and X-ray florescent analysis and Transmission Electron Microscope observation were performed focused on the contact parts of the grouting material and rock mass in fractures. As a result, the chemical influences of grouting material on the rock mass were identified. A CD-ROM is attached as an appendix. (J.P.N.)

  15. Analysis of partial and total flow blockage of a single fuel assembly of an MTR research reactor core

    International Nuclear Information System (INIS)

    The main aim of the following study is to perform a safety analysis of the IAEA 10 MW MTR Pool type Research Reactor [IAEA-TECDOC-233, 1980. IAEA Research Reactor Core Conversion from the use of high-enriched uranium to the use of low enriched uranium fuels Guidebook] under flow blockage of a single Fuel Assembly (FA) conditions. Such event was rarely investigated in the open literature notwithstanding the fat that it constitutes a severe accident that may lead to local dryout and eventually to loss of the FA integrity. The transients herein considered are related to partial and total obstruction of the cooling channel of a single Fuel Assembly of the reactor core. This study constitutes the first step of a larger work, which consists in performing a 3D simulation using the Best Estimate coupled code technique. However, as a first approach the instantaneous reactor power is derived through the point kinetic approach of the used thermal-hydraulic system code

  16. Sea Carousel—A benthic, annular flume

    Science.gov (United States)

    Amos, Carl L.; Grant, J.; Daborn, G. R.; Black, K.

    1992-06-01

    A benthic annular flume (Sea Carousel) has been developed and tested to measure in situ the erodibility of cohesive sediments. The flume is equipped with three optical backscatter sensors, a lid rotation switch, and an electromagnetic (EM) flow meter capable of detecting azimuthal and vertical components of flow. Data are logged at rates up to 10·66 Hz. Erodibility is inferred from the rate of change in suspended sediment concentration detected in the annulus. The energy-density/wave number spectrum of azimuthal flow showed peaks in the energy spectrum at paddle rotation wave numbers (k) of 14 and 7 m -1 (macroturbulent time scales) but were not significant. Friction velocity ( U*), measured (1) at 1 Hz using a flush-mounted hot-film sensor, and (2) derived from measured velocity profiles in the inner part of the logarithmic layer gave comparable results for Ū* 0·32 m s -1. Radial velocity gradients were proportional to ( Ū y - 0·32 m s -1). Maximum radial differences in U* were 10% for Ū y = 0·5 ms -1. Suspended sediment mass concentration ( S) in the annulus resulted in a significant decrease (10·5%) in Ū* derived by method (1) over the range 0calibration with changes in S. Subaerial deployments of Sea Carousel caused severe substrate disturbance, water losses, and aeration of the annulus. Submarine deployments produced stable results, though dispersion of turbid flume water took place. Results clearly demonstrated the existence of 'Type I' and 'Type II' erosion documented from laboratory studies.

  17. Research for a multi-modal mobility and manipulation propulsion core

    Science.gov (United States)

    Edge, Harris; Collins, Jason

    2015-05-01

    There are many challenges for robotics, many of which may be placed in the context of robots acting as a teammate to Soldiers. In general one may see a robotic teammate as an unmanned system that complements a Soldier's capability, may perform some of the duties of a Soldier, or may actually protect the Soldier. There is much research that needs to be performed before robots are physically capable of performing as teammates to Soldiers in dynamic environments where speed matters, and in complex 3-D environments where navigation for today's robots is difficult. This research addresses a fundamental obstacle to addressing this issue, which is how to safely and cost effectively develop theory and controls for a new generation of robots that may operate at operations tempo (OPTEMPO) in dynamic complex 3-D environments. This paper documents design and fabrication of a research platform capable of demonstrating theory and control algorithms developed for highly dynamic robotics systems, which may need to navigate and perform a task in complex 3-D environments. The research platform has been designed to address challenging basic research in the areas of airborne manipulation, transition to and interaction with vertical surfaces, exploration of a constrained space such as urban environments (street level to rooftop), forests, and underground facilities. The platform will allow controls development and validation for a vehicle that's weight is at least partially supported by a propulsion system to perform work on the environment and/or an object within the environment.

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

    International Nuclear Information System (INIS)

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

  19. Stroke Investigative Research and Education Network: Community Engagement and Outreach within Phenomics Core

    Science.gov (United States)

    Jenkins, Carolyn; Arulogun, Oyedunni Sola; Singh, Arti; Mande, Aliyu T.; Ajayi, Eric; Benedict, Calys Tagoe; Ovbiagele, Bruce; Lackland, Daniel T.; Sarfo, Fred Stephen; Akinyemi, Rufus; Akpalu, Albert; Obiako, Reginald; Melikam, Enzinne Sylvia; Laryea, Ruth; Shidali, Vincent; Sagoe, Kwamena; Ibinaiye, Philip; Fakunle, Adekunie Gregory; Owolabi, Lukman F.; Owolabi, Mayowa O.

    2016-01-01

    Stroke is the leading cause of neurological hospital admissions in sub-Saharan Africa (SSA) and the second leading cause of death globally. The Stroke Investigative Research and Education Network [SIREN] seeks to comprehensively characterize the genomic, sociocultural, economic, and behavioral risk factors for stroke and to build effective teams…

  20. Far-field Diffraction Properties of Annular Walsh Filters

    Directory of Open Access Journals (Sweden)

    Pubali Mukherjee

    2013-01-01

    Full Text Available Annular Walsh filters are derived from the rotationally symmetric annular Walsh functions which form a complete set of orthogonal functions that take on values either +1 or −1 over the domain specified by the inner and outer radii of the annulus. The value of any annular Walsh function is taken as zero from the centre of the circular aperture to the inner radius of the annulus. The three values 0, +1, and −1 in an annular Walsh function can be realized in a corresponding annular Walsh filter by using transmission values of zero amplitude (i.e., an obscuration, unity amplitude and zero phase, and unity amplitude and phase, respectively. Not only the order of the Walsh filter but also the size of the inner radius of the annulus provides an additional degree of freedom in tailoring of point spread function by using these filters for pupil plane filtering in imaging systems. In this report, we present the far-field amplitude characteristics of some of these filters to underscore their potential for effective use in several demanding applications like high-resolution microscopy, optical data storage, microlithography, optical encryption, and optical micromanipulation.

  1. Rotordynamic Analysis of Textured Annular Seals With Multiphase (Bubbly Flow

    Directory of Open Access Journals (Sweden)

    Gérard PINEAU

    2011-09-01

    Full Text Available For some applications it must be considered that the flow in the annular seal contains a mixture of liquid and gas. The multiphase character of the flow is described by the volume fraction of gas (usually air contained in the liquid under the form of bubbles.The fluid is then a homogenous mixture of air and liquid all thru the annular seal. Its local gas volume fraction depends on the pressure field and is calculated by using a simplified form of the Rayleigh-Plesset equation.The influence of such of a multiphase (bubbly flow on the dynamic characteristics of a straight annular seal is minimal because the volume of the fluid is reduced.The situation is quite different for textured annular (damper seals provided with equally spaced deep cavities intended to increase the damping capabilities and to reduce the leakage flow rate.As a by-product, the volume of the fluid in the seal increases drastically and the compressibility effects stemming from the bubbly nature of the flow are largely increased even for a low gas volume fraction. The present work depicts the influence of the gas volume fraction on the dynamic characteristics of a textured annular seal. It is shown that variations of the gas volume fraction between 1% and 0.1% can lead to frequency dependent stiffness, damping and added mass coefficients.

  2. Benefits of research activities incorporation into the core business of smaller TSOs

    International Nuclear Information System (INIS)

    Much has been done to develop the nuclear safety technology to the current level. Nevertheless, incremental development of this knowledge based on demand-pull innovation only seems not to satisfy the future needs of nuclear industry. In order to cope with technical challenges faced by TSOs in the context of nuclear renaissance, the technology-push innovation has to be done in certain amount as well to ensure that brand-new ideas, technologies and attitudes will be implemented in order to enhance nuclear safety. For the reasons described, the technology-push innovation is accessible mainly for larger organizations. In order to cross the entrance barrier for smaller organizations, a cooperation among several TSOs is needed. Uncertainties related to investments in research can be optimalized by creating a research project portfolio. According to the traditional innovation management theory, technological innovation is a driver of competition and profitability and, therefore, the motivation is supported by economical benefits. (author)

  3. Nutritional implications of organic conversion in large scale food service preliminary results from Core Organic research

    DEFF Research Database (Denmark)

    Mikkelsen, Bent Egberg; He, Chen

    POPY study and was conducted to investigate if such changes can be found in school food settings. In other words does organic food schemes at school and related curricular activities help to create environments that are supportive for healthier eating among children? The research was carried out among school...... food coordinators in public schools in Denmark, Finland, Germany, and Italy. A questionnaire was adapted to fit the different languages and food cultures in the countries.. The data suggest that schools with organic supply tend to develop organisational environments that a more supportive for healthy......The discussion about nutritional advantages of organic consumption has traditionally focused on the properties of the food it self. Studies have shown however that change of consumption patterns towards organic food seems to induce changed dietary patterns. The current research was a part of the i...

  4. Clinical data integration model : core interoperability ontology for research using primary care data

    OpenAIRE

    Ethier, J.-F.; Curcin, V; Barton, A.; Bastiaens, H.; et al

    2015-01-01

    Abstract: Introduction: This article is part of the Focus Theme of Methods of Information in Medicine on "Managing Interoperability and Complexity in Health Systems". Background: Primary care data is the single richest source of routine health care data. However its use, both in research and clinical work, often requires data from multiple clinical sites, clinical trials databases and registries. Data integration and interoperability are therefore of utmost importance. Objectives: TRANSFoRm's...

  5. Action research in natural resource management : Marginal in the first paradigm, core in the second

    OpenAIRE

    Jiggins, J.; Roling, N.

    1997-01-01

    International audience This chapter departs {rom the assumption that action research has become an established methodology for interactive learning and the management ofcomplex processes ofchange. lt tries to go further by systematically contrasting two paradigms, positivism and constructivism. Theoretically, these paradigms are each highly internally consistent across different levels ofdiscourse, such as epîstemology, theory, and methodology. In practice, we live in an era ofparadigm cha...

  6. Nutritional implications of organic conversion in large scale food service: Preliminary results from Core Organic research

    OpenAIRE

    Mikkelsen, Bent Egberg; He, Chen

    2010-01-01

    The discussion about nutritional advantages of organic consumption has traditionally focused on the properties of the food it self. Studies have shown however that change of consumption patterns towards organic food seems to induce changed dietary patterns. The current research was a part of the iPOPY study and was conducted to investigate if such changes can be found in school food settings. In other words does organic food schemes at school and related curricular activities help to cr...

  7. The reclamation of used moulding and core sands as a part of research programme of the Foundry Research Institute in Cracow

    Directory of Open Access Journals (Sweden)

    I. Izdebska-Szanda

    2010-04-01

    Full Text Available The article presents a historical outline of studies carried out by the Foundry Research Institute in Cracow on technical and technological solutions regarding the development of a reclamation process of the used moulding and core sands with attention focused on the past twenty years. Various aspects that control studies of the sand reclamation, from the thermal process, through pneumatic reclamation, and with preferences focused on the application of a mechanical treatment of the used sand, were discussed. Particular emphasis was put on the vibration method.Examples of design solutions developed by the Foundry Research Institute in Cracow were described. They were implemented inindustrial practice both at home and abroad. Their versatility and effectiveness as satisfying BAT criteria was stressed.

  8. Review of research on CANDU-SCWR in XJTU

    International Nuclear Information System (INIS)

    The pressure channel design of SCWR can avoid a thick wall vessel, and allows, in principle, some key features for safety and performance. The present paper will review the research on CANDU-SCWR in Xi'an Jiaotong University, China. (1) Thermal-hydraulics and safety analysis: Subchannel analysis of bundle, startup procedure, radiation heat transfer capability analysis were analyzed. (2) Neutronics/thermalhydraulics coupling analysis: An improved CANDU-SCWR core design with vertical channels in hexagonal arrangement was developed. (3) Experimental work: A supercritical water heat transfer test section has been built to study heat transfer in annular flow channel. Based on the experimental results, the effects of mass flux and heat flux on heat transfer of supercritical pressure water in vertical annular channel were analyzed. (author)

  9. EPRI [Electric Power Research Institute]/ANL investigations of MCCI [molten core-concrete interactions] phenomena and aerosol release

    International Nuclear Information System (INIS)

    A program of laboratory investigations has been undertaken at Argonne National Laboratory, under sponsorship of the Electric Power Research Institute, in which the interaction between molten core materials and concrete is studied, with particular emphasis on measurements of the magnitude and chemical species present in the aerosol releases. The experiment technique used in these investigations is direct electrical heating in which a high electric current is passed through the core debris to sustain the high-temperature melt condition for potentially long periods of time. In the scoping experiments completed to date, this technique has been successfully used for corium masses of 5 and 20 kg, generating an internal heating rate of 1 kw/kg and achieving melt temperatures of 2000C. Experiments have been performed both with a concrete base and also with a cooled base with the addition of H2/CO sparging gas to represent chemical processes in a stratified layer. An aerosol and gas sampling system is being used to collect aerosol samples. Test results are now becoming available including masses of aerosols, x-ray diffraction, and scanning electron microscope analyses

  10. Analysis of Accidents at the Pakistan Research Reactor-1 Using Proposed Mixed-Fuel (HEU and LEU) Core

    International Nuclear Information System (INIS)

    The Pakistan Research Reactor-1 (PARR-1) was converted from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel in 1991. The reactor is running successfully, with an upgraded power level of 10 MW. To save money on the purchase of costly fresh LEU fuel elements, the use of less burnt HEU spent fuel elements along with the present LEU fuel elements is being considered. The proposal calls for the HEU fuel elements to be placed near the thermal column to gain the required excess reactivity. In the present study the safety analysis of a proposed mixed-fuel core has been carried out at a calculated steady-state power level of 9.8 MW. Standard computer codes and correlations were employed to compute various parameters. Initiating events in reactivity-induced accidents involve various modes of reactivity insertion, namely, start-up accident, accidental drop of a fuel element on the core, flooding of a beam tube with water, and removal of an in-pile experiment during reactor operation. For each of these transients, time histories of reactor power, energy released, temperature, and reactivity were determined

  11. Burnup analysis and in-core fuel management study of the 3 MW TRIGA MARK II research reactor

    International Nuclear Information System (INIS)

    The principal objective of this study is to formulate an effective optimal fuel management strategy for the TRIGA MARK II research reactor at AERE, Savar. The core management study has been performed by utilizing four basic types of information calculated for the reactor: criticality, power peaking, neutron flux and burnup calculation. This paper presents the results of the burnup calculations for TRIGA LEU fuel elements. The fuel element burnup for approximately 20 years of operation was calculated using the TRIGAP compute code. The calculation is performed in one-dimensional radial geometry in TRIGAP. Inter-comparison of TRIGAP results with other two calculations performed by MVP-BURN and MCNP4C-ORIGEN2.1 show very good agreement. Reshuffling at 20,000 MWh step provides the highest core lifetime of the reactor, which is 64,500 MWh. Besides, the study gives valuable insight into the behaviour of the reactor and will ensure better utilization and operation of the reactor in future

  12. Unstructured 3D core calculations with the descartes system application to the JHR research reactor

    International Nuclear Information System (INIS)

    Recent developments in the DESCARTES system enable neutronics calculations dealing with very complex unstructured geometrical configurations. The discretization can be made either by using a very fine Cartesian mesh and the fast simplified transport (SPN) solver MINOS, or a discretization based on triangles and the SP1 solver MINARET. In order to perform parallel calculations dealing with a very fine mesh in 3D, a domain decomposition with non overlapping domains has been implemented. To illustrate these capabilities, we present an application on the future European research reactor JHR dedicated to technological irradiations. (authors)

  13. Functional reliability evaluation of an MTR-pool type research reactor core using the load–capacity interference model

    International Nuclear Information System (INIS)

    Highlights: • Load–capacity concept is used for reliability evaluation. • Functional Reliability is evaluated in normal operation. • Status of reactor core in normal operation is considered as multiple states. • Latin Hypercube Sampling is used for sampling in uncertainty propagation. • Core heat transfer coefficient is identified as the dominant parameter. - Abstract: This paper presents the functional reliability evaluation of Tehran Research Reactor (TRR) core in normal operation. The concept of functional reliability, borrowed from reliability physics, uses the well-known resistance–stress or load–capacity interference model that is used in the structural reliability framework. To use the load–capacity interference model, uncertainties of significant parameters in system performance are propagated into system dynamics modeled with RELAP5/Mod 3.2 using Latin Hypercube Sampling (LHS) method and exceedance probability (EP) model is used as quantification method. The proposed method in this paper solves a common problem in reliability analysis, i.e., lack of sufficient failure data in specific operating conditions. Although defining failure criteria in normal operation are difficult, this paper focuses on the application of multiple states criteria to determine the status of a system. The status of the reactor core in normal operation is considered multiple states regarding to a performance representative parameter that is temperature in this work. Outlet temperatures of fuel hot and average channels were selected to be performance indicators in normal operation. Consulting with TRR engineers and operators as well as safety analysis report, two failure states were considered exceeding 65.1 °C and 58.9 °C for the hot channel and 50.4 °C and 45.6 °C for the average channel as upper and lower limits respectively. The calculated reliability was 9.1e−01 with 95% of confidence interval, which is in good agreement with experimental results. Using

  14. A lumped parameter core dynamics model for MTR type research reactors under natural convection regime

    International Nuclear Information System (INIS)

    Highlights: ► A model is presented to simulate the reactivity insertion transient in MTR reactors. ► Transient dynamics of IAEA 10 MW MTR type research reactor are evaluated. ► Maximum unprotected reactivity insertion for safe condition is calculated. ► The model predictions are validated with corresponding results in the literature. - Abstract: On the basis of lumped parameter modeling of both the kinetic and thermal–hydraulic effects, a reasonably accurate simplified model has been developed to predict the dynamic response of MTR reactors following to an unprotected reactivity insertion under natural convection regime. By this model the reactor transient behavior at a given initial steady-state can be solved by a set of ordinary differential equations. The model predictions have an acceptable consent with corresponding results of reactivity insertion transients analyzed in the literature. The inherent safety characteristics of MTR research reactors utilizing natural convection is clearly demonstrated by the expanded model. The safety margin of reactor operating is selected ONB condition and thereby the proposed model determines that any slight increase in the value of $0.73 for inserted reactivity will cause the maximum cladding surface temperature to exceed the ONB condition

  15. Pressure loss coefficient and flow rate of side hole in a lower end plug for dual-cooled annular nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Chang-Hwan, E-mail: shinch@kaeri.re.kr; Park, Ju-Yong, E-mail: juyong@kaeri.re.kr; In, Wang-Kee, E-mail: wkin@kaeri.re.kr

    2013-12-15

    Highlights: • A lower end plug with side flow holes is suggested to provide alternative flow paths of the inner channel. • The inlet loss coefficient of the lower end plug is estimated from the experiment. • The flow rate through the side holes is estimated in a complete entrance blockage of inner channel. • The consequence in the reactor core condition is evaluated with a subchannel analysis code. - Abstract: Dual-cooled annular nuclear fuel for a pressurized water reactor (PWR) has been introduced for a significant increase in reactor power. KAERI has been developing a dual-cooled annular fuel for a power uprate of 20% in an optimized PWR in Korea, the OPR1000. This annular fuel can help decrease the fuel temperature substantially relative to conventional cylindrical fuel at a power uprate. Annular fuel has dual flow channels around itself; however, the inner flow channel has a weakness in that it is isolated unlike the outer flow channel, which is open to other neighbouring outer channels for a coolant exchange in the reactor core. If the entrance of the inner channel is, as a hypothetical event, completely blocked by debris, the inner channel will then experience a rapid increase in coolant temperature such that a departure from nucleate boiling (DNB) may occur. Therefore, a remedy to avoid such a postulated accident is indispensable for the safety of annular fuel. A lower end plug with side flow holes was suggested to provide alternative flow paths in addition to the central entrance of the inner channel. In this paper, the inlet loss coefficient of the lower end plug and the flow rate through the side holes were estimated from the experimental results even in a complete entrance blockage of the inner channel. An optimization for the side hole was also performed, and the results are applied to a subchannel analysis to evaluate the consequence in the reactor core condition.

  16. Vibration Analysis of Annular Sector Plates under Different Boundary Conditions

    Directory of Open Access Journals (Sweden)

    Dongyan Shi

    2014-01-01

    Full Text Available An analytical framework is developed for the vibration analysis of annular sector plates with general elastic restraints along each edge of plates. Regardless of boundary conditions, the displacement solution is invariably expressed as a new form of trigonometric expansion with accelerated convergence. The expansion coefficients are treated as the generalized coordinates and determined using the Rayleigh-Ritz technique. This work allows a capability of modeling annular sector plates under a variety of boundary conditions and changing the boundary conditions as easily as modifying the material properties or dimensions of the plates. Of equal importance, the proposed approach is universally applicable to annular sector plates of any inclusion angles up to 2π. The reliability and accuracy of the current method are adequately validated through numerical examples.

  17. Fluidic Analysis in an Annular Centrifugal Contactor for Fuel Reprocessing

    International Nuclear Information System (INIS)

    An annular centrifugal contactor (ACC) is a promising device for fuel reprocessing process, because it offers several advantages—a smaller size, a smaller holdup volume, and a higher separation performance—over conventional contactors such as a mixer-settler and a pulse column. Fluid dynamics and dispersion in an ACC, which has a combined mixer/centrifuge structure, are closely related to its separation performance and capacity, and this information is useful in improving equipment design. In this paper, experimental and computational fluid dynamics (CFD) studies were conducted to analyze fluidic and dispersion behavior in ACCs. Multiphase mixing (water/TBP-dodecane/air) in the annular zone was observed by Particle Imaging Velocimetry, and the change in the fluidic and dispersion behavior was ascertained under several operational conditions. The results of the CFD studies, which considered multiphase turbulent flow in the annular and rotor interior zones, were in a good agreement with the experimental data. (author)

  18. Air entrainment into annular water flows in a vertical pipe

    International Nuclear Information System (INIS)

    An experimental investigation was carried out on air entrainment into water flowing downward in a vertical pipe. Local flow rates of water and air in a fluid layer of annular flow, formed on the pipe wall, were measured precisely by using a small tube probe. Measurements were also made of local flow rates of water and air in bubbly flow downstream of annular water flow. Distributions of local flow rates in the radial direction of the pipe for annular flow regime indicate that the fluid layer consists of a water layer adjacent to the pipe wall and a water-air (two-phase fluid) layer located inside of the water layer. The water-air layer is formed as a result of air entrainment. The departure of air bubbles from a water pool to air phase was found for bubbly flow regime. (author)

  19. Experimental and numerical investigation of an entrance blockage of an inner channel in dual-cooled annular nuclear fuel

    International Nuclear Information System (INIS)

    A dual-cooled annular nuclear fuel for a Pressurized Water Reactor (PWR) has been introduced for a significant increase in reactor power. The Korea Atomic Energy Research Institute (KAERI) has been researching the development of a dual-cooled annular fuel for a power increase in an optimized PWR in Korea, OPR-1000. The main advantage of a dual-cooled annular fuel is an increased heat transfer area and a reduction in the fuel temperature, which would result in reduced fission gas release and increased fuel melting margin and Departure from Nucleate Boiling (DNB) margin. The annular fuel rod is configured to allow the coolant flow through the inner channel as well as outer channel. Since the inner channel is isolated from the neighbor channels unlike the outer channels, an inner channel blockage is one of the principal technical issues of a dual-cooled annular fuel. Due to a partial blockage, the inner channel may be faced with a DNB accident. A conceptual design used to complement the entrance blockage of an inner channel was suggested by KAERI. The through holes in this design are formed on a cylindrical wall of the lower end plug. When the inner channel is blocked by debris, coolant for the inner channel will be supplied through the side holes. But due to very unusual shape of the lower end plug, it is difficult to estimate the flow resistance of the side flow holes using empirical correlations available in the open literatures. Experimental and Computational Fluid Dynamics (CFD) studies were performed to investigate the bypass flow through the side holes of the lower end plug to complement the entrance blockage of an inner channel. The form loss coefficient in the side holes was also estimated by using the pressure drop along the bypass flow path and DNB Ratio (DNBR) margin was estimated by a subchannel analysis code. (author)

  20. Servicing and safe operation constraints for initial and compact loading of WWR-K research reactor core

    International Nuclear Information System (INIS)

    The WWR-K research reactor with the design power of 10 MWt was put into operation in 1967 year. Its operation was ceased in 1988 year due to the absence of its safety validation under seismicity conditions (9 points by MSK-64 scale). After the activity aimed at enhancement of the reactor and technological equipment seismic resistance was completed in 1997 year, a decision to restart the reactor was made. The enhancement activity also implied a revision of the reactor core configuration and its optimization to reduce the environmental impact in case of unexpected events.There was conduction of analysis of parameters of an initial compact loading of WWR-K reactor and establishment of its reactivity margin rated for the beginning of the operating cycle. The margin allows a continuous operation of the reactor during 10 days. The nominal reactor power value is 6 MWt

  1. Sporopollen and algae research of core B106 in the northern South China Sea and its paleoenvironmental evolution

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    Based on the high-resolution sporopollen and algae research of the sediments from core B106 in the northem South China Sea,three sporopollen assemblage zones have been distinguished in ascending order:Zone 1 (294-194 cm):pinus-Quercus (evergreen)-Gramineae-Polypodia-ceae- Pterdium-Dicranopteris.Zone 2 (194-94 cm):Pinus-Quercus (evergreen)-Polypodiaceae-Pteridium- Dicranopteris.Zone 3 (94-4 cm):Pinus-Polypodiaceae-Pteridium-Quercus(evergreen)-Dicranopteris.The three sporopollen zones correspond to three stages of vegetation,climate and paleoenvironment evolution of the northern part of the South China Sea since 11 000 years ago.Combined with AMS 14C dating,the sporopollen and algae data can be a scientific basis for stratigraphic division and reconstruction of paleoclimate and paleoenvironment in the South China Sea.

  2. Study on natural convection heat transfer in vertical annular space of a double coaxial cylinder

    International Nuclear Information System (INIS)

    Water cooling panels are adopted as a vessel cooling system of a high temperature-engineering test reactor (HTTR) to cool the reactor core indirectly by natural convection and thermal radiation. In this study, we carried out experiments on natural convection heat transfer coupled with thermal radiation in vertical annular space of a double coaxial cylinder in order to investigate heat transfer characteristics in vertical annular space between the reactor pressure vessel and the cooling panels of the HTTR. In the present experiments, Rayleigh number based on the width of the vertical space was set to be 6.8 x 105 6 for helium and 4.2 x 107 8 for nitrogen. This report described about the heat transfer coefficient of natural convection in the vertical space and the effect of thermal radiation of the transferred heat. As a result, a heat transfer coefficient of natural convection coupled with thermal radiation was obtained as functions of Rayleigh number, aspect ratio of the space, temperature and emissivities on the heated and cooled walls. In addition to the experiments, numerical analyses were performed on the combined phenomena of natural convection and thermal radiation in the space. The numerical results were in good agreement with the experimental ones regarding the temperature on the heated and cooled walls. (author)

  3. Support for the Core Research Activities and Studies of the Computer Science and Telecommunications Board (CSTB)

    Energy Technology Data Exchange (ETDEWEB)

    Jon Eisenberg, Director, CSTB

    2008-05-13

    The Computer Science and Telecommunications Board of the National Research Council considers technical and policy issues pertaining to computer science (CS), telecommunications, and information technology (IT). The functions of the board include: (1) monitoring and promoting the health of the CS, IT, and telecommunications fields, including attention as appropriate to issues of human resources and funding levels and program structures for research; (2) initiating studies involving CS, IT, and telecommunications as critical resources and sources of national economic strength; (3) responding to requests from the government, non-profit organizations, and private industry for expert advice on CS, IT, and telecommunications issues; and to requests from the government for expert advice on computer and telecommunications systems planning, utilization, and modernization; (4) fostering interaction among CS, IT, and telecommunications researchers and practitioners, and with other disciplines; and providing a base of expertise in the National Research Council in the areas of CS, IT, and telecommunications. This award has supported the overall operation of CSTB. Reports resulting from the Board's efforts have been widely disseminated in both electronic and print form, and all CSTB reports are available at its World Wide Web home page at cstb.org. The following reports, resulting from projects that were separately funded by a wide array of sponsors, were completed and released during the award period: 2007: * Summary of a Workshop on Software-Intensive Systems and Uncertainty at Scale * Social Security Administration Electronic Service Provision: A Strategic Assessment * Toward a Safer and More Secure Cyberspace * Software for Dependable Systems: Sufficient Evidence? * Engaging Privacy and Information Technology in a Digital Age * Improving Disaster Management: The Role of IT in Mitigation, Preparedness, Response, and Recovery 2006: * Renewing U.S. Telecommunications

  4. Linking ice core and climate research to the K-12 and broader community in Denali National Park, Alaska

    Science.gov (United States)

    Campbell, S. W.; Williams, K.; Marston, L.; Kreutz, K. J.; Osterberg, E. C.; Wake, C. P.

    2013-12-01

    For the past six years, a multi-institution effort has undertaken a broad glaciological and climate research project in Denali National Park. Most recently, two ~208 m long surface to bedrock ice cores were recovered from the Mt. Hunter plateau with supporting geophysical and weather data collected. Twenty two individuals have participated in the field program providing thousands of person-hours towards completing our research goals. Technical and scientific results have been disseminated to the broader scientific community through dozens of professional presentations and six peer-reviewed publications. In addition, we have pursued the development of interactive computer applications that use our results for educational purposes, publically available fact sheets through Denali National Park, and most recently, with assistance from PolarTREC and other affiliations, the development of a children's book and roll-out of K-8 science curriculum based on this project. The K-8 curriculum will provide students with an opportunity to use real scientific data to meet their educational requirements through alternative, interactive, and exciting methods relative to more standard educational programs. Herein, we present examples of this diverse approach towards incorporating polar research into K-12 STEM classrooms.

  5. Annular elastolytic giant cell granuloma in association with Hashimoto's thyroiditis

    Directory of Open Access Journals (Sweden)

    Rishi Hassan

    2016-01-01

    Full Text Available Annular elastolytic giant cell granuloma (AEGCG is a rare granulomatous skin disease characterized clinically by annular plaques with elevated borders and atrophic centers found mainly on sun-exposed skin and histologically by diffuse granulomatous infiltrates composed of multinucleated giant cells, histiocytes and lymphocytes in the dermis along with phagocytosis of elastic fibers by multinucleated giant cells. We report a case of AEGCG in a 50-year-old woman and is highlighted for the classical clinical and histological findings of the disease and its rare co-existence with Hashimoto's thyroiditis.

  6. Annular elastolytic giant cell granuloma in association with Hashimoto's thyroiditis.

    Science.gov (United States)

    Hassan, Rishi; Arunprasath, P; Padmavathy, L; Srivenkateswaran, K

    2016-01-01

    Annular elastolytic giant cell granuloma (AEGCG) is a rare granulomatous skin disease characterized clinically by annular plaques with elevated borders and atrophic centers found mainly on sun-exposed skin and histologically by diffuse granulomatous infiltrates composed of multinucleated giant cells, histiocytes and lymphocytes in the dermis along with phagocytosis of elastic fibers by multinucleated giant cells. We report a case of AEGCG in a 50-year-old woman and is highlighted for the classical clinical and histological findings of the disease and its rare co-existence with Hashimoto's thyroiditis. PMID:27057492

  7. Portal annular pancreas: the pancreatic duct ring sign on MRCP.

    Science.gov (United States)

    Lath, Chinar O; Agrawal, Dilpesh S; Timins, Michael E; Wein, Melissa M

    2015-12-01

    Portal annular pancreas is a rare pancreatic variant in which the uncinate process of the pancreas extends and fuses to the dorsal surface of the body of the pancreas by surrounding the portal vein. It is asymptomatic, but it can be mistaken for a pancreatic head mass on imaging and could also have serious consequences during pancreatic surgery, if unrecognized. We report this case of a 53-year-old female patient who was diagnosed to have portal annular pancreas on the basis of an unusual course (ring appearance) of the main pancreatic duct on magnetic resonance cholangiopancreatography, not described earlier in the radiology literature. PMID:26649117

  8. Portal annular pancreas: the pancreatic duct ring sign on MRCP

    Directory of Open Access Journals (Sweden)

    Chinar O. Lath, MD

    2015-12-01

    Full Text Available Portal annular pancreas is a rare pancreatic variant in which the uncinate process of the pancreas extends and fuses to the dorsal surface of the body of the pancreas by surrounding the portal vein. It is asymptomatic, but it can be mistaken for a pancreatic head mass on imaging and could also have serious consequences during pancreatic surgery, if unrecognized. We report this case of a 53-year-old female patient who was diagnosed to have portal annular pancreas on the basis of an unusual course (ring appearance of the main pancreatic duct on magnetic resonance cholangiopancreatography, not described earlier in the radiology literature.

  9. Flow Visualisation of Annular Liquid Sheet Instability & Atomisation

    CERN Document Server

    Duke, Daniel; Soria, Julio

    2012-01-01

    Fluid dynamics videos of unstable thin annular liquid sheets are presented in this short paper. These videos are to be presented in the Gallery of Fluid Motion for the American Physical Society 65th Annual Meeting of the Division of Fluid Dynamics in San Diego, CA, 18-20 November 2012. An annular sheet of thickness h=1mm and mean radius R=18.9mm is subjected to aerodynamic axial shear from co-flowing air at various shear rates on both the inner and outer surface at a liquid sheet Reynolds Number of Re=500.

  10. Core transfer

    Science.gov (United States)

    Good news for all petroleum geoscientists, mining and environmental scientists, university researchers, and the like: Shell Oil Company has deeded its Midland core and sample repository to the Bureau of Economic Geology (BEG) at the University of Texas at Austin. The Midland repository includes more than 1 million linear meters of slab, whole core, and prepared cuttings. Data comprising one of the largest U.S. core collections—the geologic samples from wells drilled in Texas and 39 other states—are now public data and will be incorporated into the existing BEG database. Both Shell and the University of Texas at Austin are affiliated with the American Geological Institute, which assisted in arranging the transfer as part of its goal to establish a National Geoscience Data Repository System at regional centers across the United States.

  11. Fluid-elastic instability in a confined annular flow: an experimental and analytical approach

    International Nuclear Information System (INIS)

    Self excitation of slender structures under axial flow have been reported in a large variety of local flow configurations. This paper reports the result of a research program, both experimental and analytical, aimed at the result of the basic phenomena leading to such instabilities. A cylindrical body with a diffuser is put in a confined annular flow of water. A case of flutter is observed and analysed with a classical potential flow method and with a friction based model. Closed-form solutions are proposed and the origin of the flutter instability is discussed. (authors). 25 refs., 6 figs., 5 tabs

  12. TWO-PHASE ANNULAR FLOW IN A VERTICALLY MOUNTED VENTURI FLOW METER

    OpenAIRE

    Panella, Bruno; Salve, Mario De; Monni, Grazia

    2014-01-01

    In the present research work, the experimental investigation of a vertical upward annular two-phase flow in a Venturi Flow Meter (VFM) is performed. The pressure drops between the inlet and throat section and between inlet and outlet (irreversible pressure drops) are measured and analyzed. The flow meter is characterized by an inlet diameter of 80 mm and a throat diameter of 40 mm (β=0.5), with equal convergent and divergent angles (θ=21°). The instrument has been tested in a test section, ha...

  13. Fluid-Structure Interaction Analysis on Turbulent Annular Seals of Centrifugal Pumps during Transient Process

    OpenAIRE

    Dazhuan Wu; Leqin Wang; Qinglei Jiang; Lulu Zhai

    2011-01-01

    The current paper studies the influence of annular seal flow on the transient response of centrifugal pump rotors during the start-up period. A single rotor system and three states of annular seal flow were modeled. These models were solved using numerical integration and finite difference methods. A fluid-structure interaction method was developed. In each time step one of the three annular seal models was chosen to simulate the annular seal flow according to the state of rotor systems. The ...

  14. A droplet entrainment model based on the force balance of an interfacial wave in two-phase annular flow

    International Nuclear Information System (INIS)

    Highlights: → Knowledge of the interfacial wave structure is essential for making an accurate prediction of the amount of entrained droplets. → A new droplet entrainment model based on the force balance of interfacial waves in vertical annular flow. → An analytic wave shape function was developed. → A new droplet entrainment model was validated using the experimental data reported by Hewitt and Pulling and by Keeys et al. - Abstract: Droplets are generated at the interface of annular flow due to an interaction between a liquid film and gas core flow. Therefore, knowledge of the interfacial wave structure is essential for making an accurate prediction of the amount of entrained droplets. A new droplet entrainment model was proposed based on the force balance of interfacial waves in vertical annular flow. An analytic wave shape function was developed reflecting the detailed experimental findings, and was used in the development of a new model. The model was validated using the experimental data reported by Hewitt and Pulling at low pressures and by Keeys et al. at high pressures, which had been performed in adiabatic vertical tubes. The root-mean-square error of the prediction of the amount of entrainment was approximately 27% when the model was implemented into COBRA-TF code, which is approximately 23% less than that determined by the Wuertz model. The models proposed by Okawa et al. and Stevanovic et al. were also implemented into COBRA-TF and compared with the proposed model.

  15. IAEA Coordinated Research Project on the Establishment of a Material Properties Database for Irradiated Core Structural Components for Continued Safe Operation and Lifetime Extension of Ageing Research Reactors

    International Nuclear Information System (INIS)

    Today more than 50% of operating Research Reactors (RRs) are over 45 years old. Thus, ageing management is one of the most important issues to face in order to ensure availability (including life extension), reliability and safe operation of these facilities for the future. Management of the ageing process requires, amongst others, the predictions for the behavior of structural materials of primary components subjected to irradiation such as reactor vessel and core support structures, many of which are extremely difficult or impossible to replace. In fact, age-related material degradation mechanisms resulted in high profile, unplanned and lengthy shutdowns and unique regulatory processes of relicensing the facilities in recent years. These could likely have been prevented by utilizing available data for the implementation of appropriate maintenance and surveillance programmes. This IAEA Coordinated Research Project (CRP) will provide an international forum to establish a material properties Database for irradiated core structural materials and components. It is expected that this Database will be used by research reactor operators and regulators to help predict ageing related degradation. This would be useful to minimize unpredicted outages due to ageing processes of primary components and to mitigate lengthy and costly shutdowns. The Database will be a compilation of data from RRs operators' inputs, comprehensive literature reviews and experimental data from RRs. Moreover, the CRP will specify further activities needed to be addressed in order to bridge the gaps in the new created Database, for potential follow-on activities. As per today, 13 Member States (MS) confirmed their agreement to contribute to the development of the Database, covering a wide number of materials and properties. The present publication incorporates two parts: the first part includes details on the pre-CRP Questionnaire, including the conclusions drawn from the answers received from the MS

  16. Uninstrumented assembly airflow testing in the Annular Flow Distribution facility

    Energy Technology Data Exchange (ETDEWEB)

    Kielpinski, A.L.

    1992-02-01

    During the Emergency Cooling System phase of a postulated large-break loss of coolant accident (ECS-LOCA), air enters the primary loop and is pumped down the reactor assemblies. One of the experiments performed to support the analysis of this accident was the Annular Flow Distribution (AFD) experiment, conducted in a facility built for this purpose at Babcock and Wilcox Alliance Research Center in Alliance, Ohio. As part of this experiment, a large body of airflow data were acquired in a prototypical mockup of the Mark 22 reactor assembly. This assembly was known as the AFD (or the I-AFD here) reference assembly. The I-AFD assembly was fully prototypical, having been manufactured in SRS`s production fabrication facility. Similar Mark 22 mockup assemblies were tested in several test facilities in the SRS Heat Transfer Laboratory (HTL). Discrepancies were found. The present report documents further work done to address the discrepancy in airflow measurements between the AFD facility and HTL facilities. The primary purpose of this report is to disseminate the data from the U-AFD test, and to compare these test results to the I-AFD data and the U-AT data. A summary table of the test data and the B&W data transmittal letter are included as an attachment to this report. The full data transmittal volume from B&W (including time plots of the various instruments) is included as an appendix to this report. These data are further analyzed by comparing them to two other HTL tests, namely, SPRIHTE 1 and the Single Assembly Test Stand (SATS).

  17. 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.

  18. Thermal Hydraulic Analysis Of Thorium-Based Annular Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Han, Kyu Hyun [Korea Institute of Nuclear Safety, 19, Guseong-dong, Yuseong-gu, Daejeon, 305-338 (Korea, Republic of)

    2008-07-01

    Thermal hydraulic characteristics of thorium-based fuel assemblies loaded with annular seed pins have been analyzed using AMAP combined with MATRA, and compared with those of the existing thorium-based assemblies. MATRA and AMAP showed good agreements for the pressure drops at the internal sub-channels. The pressure drop generally increased in the cases of the assemblies loaded with annular seed pins due to the larger wetted perimeter, but an exception existed. In the inner sub-channels of the seed pins, mass fluxes were high due to the grid form losses in the outer sub-channels. About 43% of the heat generated from the seed pin flowed into the inner sub-channel and the rest into the outer sub-channel, which implies the inner to outer wall heat flux ratio was approximately 1.2. The maximum temperatures of the annular seed pins were slightly above 500 deg. C. The MDNBRs of the assemblies loaded with annular seed pins were higher than those of the existing assemblies. Due to the fact that inter-channel mixing cannot occur in the inner sub-channels, temperatures and enthalpies were higher in the inner sub-channels. (author)

  19. Thermal Hydraulic Analysis Of Thorium-Based Annular Fuel Assemblies

    International Nuclear Information System (INIS)

    Thermal hydraulic characteristics of thorium-based fuel assemblies loaded with annular seed pins have been analyzed using AMAP combined with MATRA, and compared with those of the existing thorium-based assemblies. MATRA and AMAP showed good agreements for the pressure drops at the internal sub-channels. The pressure drop generally increased in the cases of the assemblies loaded with annular seed pins due to the larger wetted perimeter, but an exception existed. In the inner sub-channels of the seed pins, mass fluxes were high due to the grid form losses in the outer sub-channels. About 43% of the heat generated from the seed pin flowed into the inner sub-channel and the rest into the outer sub-channel, which implies the inner to outer wall heat flux ratio was approximately 1.2. The maximum temperatures of the annular seed pins were slightly above 500 deg. C. The MDNBRs of the assemblies loaded with annular seed pins were higher than those of the existing assemblies. Due to the fact that inter-channel mixing cannot occur in the inner sub-channels, temperatures and enthalpies were higher in the inner sub-channels. (author)

  20. Annular linear induction pump with an externally supported duct

    International Nuclear Information System (INIS)

    Several embodiments of an annular linear induction pump for pumping liquid metals are disclosed having the features of generally one pass flow of the liquid metal through the pump and an increased efficiency resulting from the use of thin duct walls to enclose the stator. The stator components of this pump are removable for repair and replacement. 15 claims

  1. Localized granuloma annulare and autoimmune thyroid disease. Are they associated?

    OpenAIRE

    Moran, J; Lamb, J.

    1995-01-01

    This case report identifies a temporal relationship between the diagnosis of localized granuloma annulare and the subsequent development of primary hypothyroidism in a previously healthy 10-year-old girl. We suspect these disorders are associated, but any association between them requires further study.

  2. Fluxon dynamics in long annular Josephson tunnel junctions

    DEFF Research Database (Denmark)

    Martucciello, N.; Mygind, Jesper; Koshelets, V.P.; Shchukin, A.V.; Filippenko, L.; Monaco, R

    1998-01-01

    Single-fluxon dynamics has been experimentally investigated in high-quality Nb/Al-AlOx/Nb annular Josephson tunnel junctions having a radius much larger than the Josephson penetration depth. Strong evidence of self-field effects is observed. An external magnetic field in the barrier plane acts on...

  3. Effect of liquid entrainment on turbulent mixing rates between subchannels in gas-liquid annular two-phase flows

    International Nuclear Information System (INIS)

    Turbulent mixing rates of gas and liquid phases between adjacent subchannels have been measured for various air-water two-phase annular flows in a multiple channel consisting of two identical circular subchannels. In order to study effect of liquid entrainment in the gas core on the turbulent mixing rates, experiments were conducted for two types of liquid injection method, i.e., a small bore nozzle placed in the subchannel center and a porous wall. The result showed that the effect of liquid entrainment on the turbulent mixing rates of both phases is negligibly small. (author)

  4. The review of fuel types for Russian research reactors. Their fabrication and quality control

    International Nuclear Information System (INIS)

    The design of tubular fuel elements (FEs) for research reactor fuel assemblies (FAs) is considered. Commercial extrusion and annular-type technologies for tubular FE fabrication are described. 'Extrusion' technology is based on fabrication of tubular billet of fuel core by means of powder metallurgy followed by hot extrusion of fuel core tubular billet and tubular cladding billet. The process is completed with FE assembly operation. 'Annular' technology is based on fuel core fabrication using powder metallurgy followed by chemical treatment of fuel core surface and fuel core insertion into the cladding. The list of FE and FA control operations to check their conformance to the required quality level is given. The most common FA designs (WWR-M2, WWR-M5, IRT-2M, IRT-3M, MR, MIR, WWR-TS, IVV-2M, IVV-10, TWR-S, IR-100) for research reactors built according to the Russian projects are described. The Quality Assurance System in operation at 'Novosibirsk' Chemical Concentrates Plant' is presented. (author)

  5. Physical start up of the Dalat nuclear research reactor with the core configuration having a central neutron trap

    International Nuclear Information System (INIS)

    After the reactor has reached physical criticality with the core configuration exempt from central neutron trap on 1 November 1983, the core configuration with a central neutron trap has been arranged in the reactor and the reactor has reached physical criticality with this core configuration at 17h48 on 18 December 1983. The integral worths of different control rods are determined with accuracy. 2 refs., 24 figs., 18 tabs

  6. Study of film boiling dispersed two phase in narrow annular gap

    International Nuclear Information System (INIS)

    Experimental investigation on film boiling dispersed two phase friction pressure drop in narrow annular gap with deionized water was performed in three types of narrow annular gap. The friction pressure drop differences were compared between narrow annular gap and circular channel was compared in the paper. The influence of narrow annular gap on friction pressure drop was examined in this paper. Results showed that the modified Sadatomi's correlation can be used to calculate film boiling dispersed two-phase friction pressure drop in narrow annular gap for engineering application

  7. Safety core parameters prediction in research reactors using artificial neural networks: A comparative study of various learning algorithms

    International Nuclear Information System (INIS)

    In recent years, Artificial Neural Networks (ANNs) were applied successfully as an advanced and promising tool for simulating several reactor physics parameters in nuclear engineering applications. The main objective in using such Artificial Intelligent (AI) methods, in the field of nuclear engineering, is to develop simple and 1st estimate models capable of simulating adequately, with reasonable error, important reactor physics parameters in relatively short time comparatively to time consuming and cumbersome reactor physics computer codes. The feasibility of this application has been demonstrated through a previous work done for a typical benchmark 10 Mw IAEA LEU (Low Enriched Uranium) core research reactor, using an adaptive learning rate procedure in a typical back-propagation algorithm in the training process. However, even tough the predictive results achieved are within ±0.7% for Keff and within ±8.5% for Pmax, the convergence time spent during the training phase were of about 36 and 24 hours, respectively for both cited parameters, on a small computational system (300 Mhz Pentium II PC). Hence, this paper suggests one of the suitable ways explored to speed up the training process and to improve neural networks performances by carrying out a comprehensive sensitivity studies on an iterative and multistage calculation process using Neural Network MATLAB Toolbox

  8. A core design study for 'zero-sodium-void-worth' cores

    International Nuclear Information System (INIS)

    Recently, a number of low sodium-void-worth metal-fueled core design concepts have been proposed; to provide for flexibility in transuranic nuclide management strategy, core designs which exhibit a wide range of breeding characteristics have been developed. Two core concepts, a flat annular (transuranic burning) core and an absorber-type parfait (transuranic self-sufficient) core, are selected for this study. In this paper, the excess reactivity management schemes applied in the two designs are investigated in detail. In addition, the transient effect of reactivity insertions on the parfait core design is assessed. The upper and lower core regions in the parfait design are neutronically decoupled; however, the common coolant channel creates thermalhydraulic coupling. This combination of neutronic and thermalhydraulic characteristics leads to unique behavior in anticipated transient overpower events. (author)

  9. Neutronic analysis for core conversion (HEU–LEU of the low power research reactor using the MCNP4C code

    Directory of Open Access Journals (Sweden)

    Aldawahra Saadou

    2015-06-01

    Full Text Available Comparative studies for conversion of the fuel from HEU to LEU in the miniature neutron source reactor (MNSR have been performed using the MCNP4C code. The HEU fuel (UAl4-Al, 90% enriched with Al clad and LEU (UO2 12.6% enriched with zircaloy-4 alloy clad cores have been analyzed in this study. The existing HEU core of MNSR was analyzed to validate the neutronic model of reactor, while the LEU core was studied to prove the possibility of fuel conversion of the existing HEU core. The proposed LEU core contained the same number of fuel pins as the HEU core. All other structure materials and dimensions of HEU and LEU cores were the same except the increase in the radius of control rod material from 0.195 to 0.205 cm and keeping the outer diameter of the control rod unchanged in the LEU core. The effective multiplication factor (keff, excess reactivity (ρex, control rod worth (CRW, shutdown margin (SDM, safety reactivity factor (SRF, delayed neutron fraction (βeff and the neutron fluxes in the irradiation tubes for the existing and the potential LEU fuel were investigated. The results showed that the safety parameters and the neutron fluxes in the irradiation tubes of the LEU fuels were in good agreements with the HEU results. Therefore, the LEU fuel was validated to be a suitable choice for fuel conversion of the MNSR in the future.

  10. Neutronic simulation of a research reactor core of (232Th, 235U)O2 fuel using MCNPX2.6 code

    Indian Academy of Sciences (India)

    Seyed Amir Hossein Feghhi; Marzieh Rezazadeh; Yachine Kadi; Claudio Tenreiro; Morteza Aref; Zohreh Gholamzadeh

    2013-01-01

    The small reactor design for the remote and less developed areas of the user countries should have simple features in view of the lack of infra-structure and resources. Many researchers consider long core life with no on-site refuelling activity as a primary feature for the small reactor design. Long core life can be achieved by enhancing internal conversion rate of fertile to fissile materials. For that purpose, thorium cycle can be adopted because a high fissile production rate of 233U converted from 232Th can be expected in the thermal energy region. A simple nuclear reactor core arranged 19 assemblies in hexagonal structure, using thorium-based fuel and heavy water as coolant and moderator was simulated using MCNPX2.6 code, aiming an optimized critical assembly. Optimized reflector thickness and gap between assemblies were determined to achieve minimum neutron leakage and void reactivity. The result was a more compact core, where assemblies were designed having 19-fuel pins in 1.25 pitch-to-diameter ratio. Optimum reflector thickness of 15 cm resulted in minimal neutron leakage in view of economic limitations. A 0.5 cm gap between assembles achieved more safety and 2.2% enrichment requirements. The present feasibility study suggests a thermal core of acceptable neutronic parameters to achieve a simple and safe core.

  11. Connecting pre-marketing clinical research and medical practice : opinion-based study of core issues and possible changes in drug regulation

    NARCIS (Netherlands)

    Wieringa, N.F; Peschar, J.L.; Denig, P; de Graeff, P.A.; Vos, R

    2003-01-01

    Objectives: To identify core issues that contribute to the gap between pre-marketing clinical research and practice as seen from the perspective of medical practice, as well as possible changes and potential barriers for closing this gap. Methods: Interviews with 47 physicians and pharmacists who we

  12. Establishing a quality assurance program for in-core fuel management of the Dalat Nuclear Research Reactor using low enriched fuel

    International Nuclear Information System (INIS)

    Quality assurance program for calculating of in-core fuel management of research reactor plays very important role in safety operation and effective utilization. The main objective of the program is to ensure the safe, reliable and optimum use of nuclear fuel and to meet the reactor utilization, which remains reactor operation within the limits imposed by the design safety considerations and the operational limits and conditions (OLCs) on the basis of safety analysis. The management of reactor core and nuclear fuel must be organized in a coherent way and comply with safety requirements. After successfully converting from HEU to LEU fuel for Dalat Research Reactor, a work to be in place is to study and implement the management of reactor core and nuclear fuel. This not only helps to ensure safety operation and efficient utilization but also contributes to build the safety culture and to be valuable experience for other nuclear projects. In addition, the application of the quality assurance program for in-core fuel management will contribute to avoid subjective mistakes, to clearly define responsibilities and to ensure legacy of expertise, which is also an urgent requirement. The selected computer code systems, data libraries and computation models must be fully met the requirements for analyzing status and characteristics of reactor core as well as the requirements for selecting, verifying and evaluating according to the regulations of the IAEA. (author)

  13. Phase separation of dispersed annular (rivulet or thin film) flow in a TEE

    International Nuclear Information System (INIS)

    An experimental and analytical investigation of dispersed annular (rivulet or thin film) flow phase separation in a tee is described in this paper. The research is directed at determining flow conditions, following a loss of coolant accident, in the large rectangular passageways leading to vacuum buildings in the containment envelope of some CANDU reactors. The reported research is part of a larger study, sponsored by Ontario Hydro of Canada, which also investigates dispersed mist flow and transient phase separation effects. The primary objective of the research was to develop mechanistic analytical models and incorporate them in a computer code which predicts phase separation from upstream flow and pressure and downstream and side branch pressure boundary conditions

  14. ANNULAR PANCREAS CAUSING DUODENAL OBSTRUCTION: A CASE REPORT

    Directory of Open Access Journals (Sweden)

    Swish Kumar

    2016-01-01

    Full Text Available Annular pancreas is a rare congenital anomaly characterized by the band of pancreatic tissue of variable width partially or completely encircling the duodenum. This abnormality, although at times clinically silent or may be the cause of a broad spectrum of diseases. Complications range from neonatal intestinal obstruction to more complex pathologies in the adult such as pancreatitis, duodenal stenosis or duodenal or gastric ulceration. This condition is important to recognise, because radiologists are usually the first person to diagnose such condition. We report the case of a young patient of 10 years old female hospitalized for epigastric pain and repeated episodes of vomiting, in whom radiological investigations showed an annular pancreas. No other congenital anomaly of the intra-abdominal organs was noted. Both the rarity of this congenital abnormality and its probability of successful correction by surgical means have prompted us to make the following presentation.

  15. Analysis of a Low-Angle Annular Expander Nozzle

    Directory of Open Access Journals (Sweden)

    Kyll Schomberg

    2015-01-01

    Full Text Available An experimental and numerical analysis of a low-angle annular expander nozzle is presented to observe the variance in shock structure within the flow field. A RANS-based axisymmetric numerical model was used to evaluate flow characteristics and the model validated using experimental pressure readings and schlieren images. Results were compared with an equivalent converging-diverging nozzle to determine the capability of the wake region in varying the effective area of a low-angle design. Comparison of schlieren images confirmed that shock closure occurred in the expander nozzle, prohibiting the wake region from affecting the area ratio. The findings show that a low angle of deflection is inherently unable to influence the effective area of an annular supersonic nozzle design.

  16. Energy and Exergy Analysis of an Annular Thermoelectric Heat Pump

    Science.gov (United States)

    Kaushik, S. C.; Manikandan, S.; Hans, Ranjana

    2016-07-01

    In this paper, the concept of an annular thermoelectric heat pump (ATEHP) has been introduced. An exoreversible thermodynamic model of the ATEHP considering the Thomson effect in conjunction with Peltier, Joule and Fourier heat conduction has been investigated using exergy analysis. New expressions for dimensionless heating power, optimum current at the maximum energy, exergy efficiency conditions and dimensionless irreversibilities in the ATEHP are derived. The results show that the heating power, energy and exergy efficiency of the ATEHP are lower than the flat-plate thermoelectric heat pump. The effects of annular shape parameter ( S r = r 2 /r 1), dimensionless temperature ratio ( θ = T h /T c) and the electrical contact resistances on the heating power, energy/exergy efficiency of an ATEHP have been studied. This study will help in the designing of actual ATEHP systems.

  17. Axisymmetric buckling of laminated thick annular spherical cap

    Science.gov (United States)

    Dumir, P. C.; Dube, G. P.; Mallick, A.

    2005-03-01

    Axisymmetric buckling analysis is presented for moderately thick laminated shallow annular spherical cap under transverse load. Buckling under central ring load and uniformly distributed transverse load, applied statically or as a step function load is considered. The central circular opening is either free or plugged by a rigid central mass or reinforced by a rigid ring. Annular spherical caps have been analysed for clamped and simple supports with movable and immovable inplane edge conditions. The governing equations of the Marguerre-type, first order shear deformation shallow shell theory (FSDT), formulated in terms of transverse deflection w, the rotation ψ of the normal to the midsurface and the stress function Φ, are solved by the orthogonal point collocation method. Typical numerical results for static and dynamic buckling loads for FSDT are compared with the classical lamination theory and the dependence of the effect of the shear deformation on the thickness parameter for various boundary conditions is investigated.

  18. Energy and Exergy Analysis of an Annular Thermoelectric Heat Pump

    Science.gov (United States)

    Kaushik, S. C.; Manikandan, S.; Hans, Ranjana

    2016-04-01

    In this paper, the concept of an annular thermoelectric heat pump (ATEHP) has been introduced. An exoreversible thermodynamic model of the ATEHP considering the Thomson effect in conjunction with Peltier, Joule and Fourier heat conduction has been investigated using exergy analysis. New expressions for dimensionless heating power, optimum current at the maximum energy, exergy efficiency conditions and dimensionless irreversibilities in the ATEHP are derived. The results show that the heating power, energy and exergy efficiency of the ATEHP are lower than the flat-plate thermoelectric heat pump. The effects of annular shape parameter (S r = r 2 /r 1), dimensionless temperature ratio (θ = T h /T c) and the electrical contact resistances on the heating power, energy/exergy efficiency of an ATEHP have been studied. This study will help in the designing of actual ATEHP systems.

  19. Development of annular targets for 99Mo production

    International Nuclear Information System (INIS)

    During 1999, significant progress was made in the development of a low-enriched uranium (LEU) target for production of 99Mo. Successful conversion requires an inexpensive, reliable target. To keep the target geometry the same when changing from high-enriched uranium (HEU) to LEU targets, a denser form of uranium is required in order to increase the amount of uranium per target by a factor of approximately five. Targets containing LEU in the form of a metal foil are being developed for producing 99Mo from the fissioning of 235U. A new annular target was developed this year, and seven targets were irradiated in the Indonesian RSG-GAS reactor. Results of development of this annular target and its performance during irradiation are described. (author)

  20. Production of annular flat-topped vortex beams

    Institute of Scientific and Technical Information of China (English)

    Jiannong Chen; Yongjiang Yu; Feifei Wang

    2011-01-01

    @@ A model of an annular flat-topped vortex beam based on multi-Gaussian superimposition is proposed. We experimentally produce this beam with a computer-generated hologram (CGH) displayed on a spatial light modulator (SLM). The power of the beam is concentrated on a single-ring structure and has an extremely strong radial intensity gradient. This beam facilitates various applications ranging from Sisyphus atom cooling to micro-particle trapping.%A model of an annular fiat-topped vortex beam based on multi-Gaussian superimposition is proposed. We experimentally produce this beam with a computer-generated hologram (CGH) displayed on a spatial light modulator (SLM). The power of the beam is concentrated on a single-ring structure and has an extremely strong radial intensity gradient. This beam facilitates various applications ranging from Sisyphus atom cooling to micro-particle trapping.

  1. Electroosmotic flow and Joule heating in preparative continuous annular electrochromatography.

    Science.gov (United States)

    Laskowski, René; Bart, Hans-Jörg

    2015-09-01

    An openFOAM "computational fluid dynamic" simulation model was developed for the description of local interaction of hydrodynamics and Joule heating in annular electrochromatography. A local decline of electrical conductivity of the background eluent is caused by an electrokinetic migration of ions resulting in higher Joule heat generation. The model equations consider the Navier-Stokes equation for incompressible fluids, the energy equation for stationary temperature fields, and the mass transfer equation for the electrokinetic flow. The simulations were embedded in commercial ANSYS Fluent software and in open-source environment openFOAM. The annular gap (1 mm width) contained an inorganic C8 reverse-phase monolith as stationary phase prepared by an in situ sol-gel process. The process temperature generated by Joule heating was determined by thermal camera system. The local hydrodynamics in the prototype was detected by a gravimetric contact-free measurement method and experimental and simulated values matched quite well. PMID:25997390

  2. Enhancing VVER annular proliferation resistance fuel with minor actinides

    International Nuclear Information System (INIS)

    Key aspects of the Global Nuclear Energy Partnership (GNEP) are to significantly advance the science and technology of nuclear energy systems and the Advanced Fuel Cycle (AFC) program. It consists of both innovative nuclear reactors and innovative research in separation and transmutation. To accomplish these goals, international cooperation is very important and public acceptance is crucial. The merits of nuclear energy are high-density energy, with low environmental impacts (i.e. almost zero greenhouse gas emission). Planned efforts involve near term and intermediate-term improvements in fuel utilization and recycling in current light water reactors (LWRs) as well as the longer-term development of new nuclear energy systems that offer much improved fuel utilization and proliferation resistance, along with continued advances in operational safety. The challenges are solving the energy needs of the world, protection against nuclear proliferation, the problem of nuclear waste, and the global environmental problem. To reduce spent fuel for storage and enhance the proliferation resistance for the intermediate-term, there are two major approaches (a) increase the discharged spent fuel burnup in the advanced LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the 238Pu and 240Pu isotopes ratio to enhance the proliferation resistance, and (b) use of transuranic nuclides (237Np and 241Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope ratio of 238Pu /Pu. For future advanced nuclear systems, the minor actinides (MA) are viewed more as a resource to be recycled, or transmuted to less hazardous and possibly more useful forms, rather than simply as a waste stream to be disposed of in expensive repository facilities. As a result, MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the

  3. Heat transfer to liquid sodium flowing through annular channel, (4)

    International Nuclear Information System (INIS)

    An experimental study was carried out to clarify the heat transfer characteristics of liquid sodium flowing turbulently through an annular channel. For a concentric condition, average psi(=average epsilonH/epsilonM) was found to agree with that proposed by Aoki or Ramm for circular tube. For eccentric conditions, circumferential temperature variations around the inner wall were measured and Nusselt numbers were evaluated. Numerical calculations were also made for temperature fields and compared with the measurements. (author)

  4. Unusual Presentation of Acute Annular Urticaria: A Case Report

    OpenAIRE

    Gilles Guerrier; Jean-Marc Daronat; Roger Deltour

    2011-01-01

    Acute urticarial lesions may display central clearing with ecchymotic or haemorrhagic hue, often misdiagnosed as erythema multiforme, serum-sickness-like reactions, or urticarial vasculitis. We report a case of acute annular urticaria with unusual presentation occurring in a 20-month-old child to emphasize the distinctive morphologic manifestations in a single disease. Clinicians who care for children should be able to differentiate acute urticaria from its clinical mimics. A directed history...

  5. Fluxon dynamics in long annular Josephson tunnel junctions

    OpenAIRE

    Martucciello, N.; Mygind, Jesper; Koshelets, V. P.; Shchukin, A. V.; Filippenko, L.; Monaco, R.

    1998-01-01

    Single-fluxon dynamics has been experimentally investigated in high-quality Nb/Al-AlOx/Nb annular Josephson tunnel junctions having a radius much larger than the Josephson penetration depth. Strong evidence of self-field effects is observed. An external magnetic field in the barrier plane acts on the fluxon as a periodic potential and lowers its average speed. Further, the results of perturbative calculations do not fit the experimental current-voltage profile and, provided the temperature is...

  6. Ignition sequence of an annular multi-injector combustor

    OpenAIRE

    Philip, Maxime; Boileau, Matthieu; Vicquelin, Ronan; Schmitt, Thomas; Durox, Daniel; Bourgoin, Jean-François; Candel, Sébastien

    2013-01-01

    Ignition is a critical process in combustion systems. In aeronautical combustors, altitude relight capacities are required in case of accidental extinction of the chamber. A simultaneous study of light-round ignition in an annular multi-injector combustor has been performed on the experimental and numerical sides. This effort allows a unique comparison to assess the reliability of Large-Eddy Simulation (LES) in such a configuration. Results are presented in fluid dynamics videos.

  7. Large Eddy Simulation of thermoacoustic instabilities in annular combustion chambers

    OpenAIRE

    Wolf, Pierre

    2011-01-01

    Increasingly stringent regulations and the need to tackle rising fuel prices have placed great emphasis on the design of aeronautical gas turbines. This drive towards innovation has resulted sometimes in new concepts being prone to combustion instabilities. Combustion instabilities arise from the coupling of acoustics and combustion. In the particular field of annular combustion chambers, these instabilities often take the form of azimuthal modes. To predict these modes, one must consider the...

  8. Thermohydraulic analysis of smooth and finned annular ducts

    International Nuclear Information System (INIS)

    The present work is concerned with the turbulent heat transfer and pressure drop in smooth and finned annular ducts overage heat transfer coefficients have been obtained by means of the heat exchanger theory. In addition, friction factors have also been determined. The experiments were performed by utilizing four double-pipe heat exchangers. The flowing fluids, in the heat exchangers, were air and water. The average heat transfer coefficients, for air flowing in the annular section, were determined by measuring the overall heat transfer coefficients of the heat exchangers. In order to attain fully developed conditions, the heat exchangers had a starting length of 30 hydraulic diameters. The thermal boundary conditions consisted of uniform temperature on the inner surface, the outer surface being insulated. The heat transfer coefficients and friction factors are presented in dimensionaless forms, as functions of the Reynolds number of the flow. The results for the smooth and finned annular ducts were compared. The purpose of such comparison was to study the influence of the fins on the pressure drop and heat transfer rate. In the case of the finned nular ducts, it is shown that the fin efficiency has some fluence on the heat transfer rates. The, a two-dimensional at transfer analysis was performed in order to obtain the n efficiency and the annular region efficiency. It is also shown that the overall thermal performance of finned surfaces epends mainly on the Nusselt number and on the region eficiency. These parameters are presented as functions of the Reynolds number of the flow and the geometry of the problem. (author)

  9. Hydraulic forces caused by annular pressure seals in centrifugal pumps

    Science.gov (United States)

    Iino, T.; Kaneko, H.

    1980-01-01

    The hydraulic forces caused by annular pressure seals were investigated. The measured inlet and exit loss coefficients of the flow through the seals were much smaller than the conventional values. The results indicate that the damping coefficient and the inertia coefficient of the fluid film in the seal are not affected much by the rotational speed or the eccentricity of the rotor, though the stiffness coefficient seemed to be influenced by the eccentricity.

  10. Hysteretic behavior of annular impinging jets

    Czech Academy of Sciences Publication Activity Database

    Trávníček, Zdeněk; Tesař, Václav

    Eindhoven: Eindhoven University of Technology , 2008 - (Stoffels, G.; van der Meer, T.; van Steenhoven, A.), s. 186-187 ISBN 978-90-386-1274-4. [European Thermal-Sciences Conference /5./. Eindhoven (NL), 18.05.2008-22.05.2008] R&D Projects: GA ČR GA101/05/2681; GA AV ČR(CZ) IAA200760504 Institutional research plan: CEZ:AV0Z20760514 Keywords : impinging jet * visualization * hysteresis Subject RIV: BK - Fluid Dynamics

  11. Guided Wave Annular Array Sensor Design for Improved Tomographic Imaging

    Science.gov (United States)

    Koduru, Jaya Prakash; Rose, Joseph L.

    2009-03-01

    Guided wave tomography for structural health monitoring is fast emerging as a reliable tool for the detection and monitoring of hotspots in a structure, for any defects arising from corrosion, crack growth etc. To date guided wave tomography has been successfully tested on aircraft wings, pipes, pipe elbows, and weld joints. Structures practically deployed are subjected to harsh environments like exposure to rain, changes in temperature and humidity. A reliable tomography system should take into account these environmental factors to avoid false alarms. The lack of mode control with piezoceramic disk sensors makes it very sensitive to traces of water leading to false alarms. In this study we explore the design of annular array sensors to provide mode control for improved structural tomography, in particular, addressing the false alarm potential of water loading. Clearly defined actuation lines in the phase velocity dispersion curve space are calculated. A dominant in-plane displacement point is found to provide a solution to the water loading problem. The improvement in the tomographic images with the annular array sensors in the presence of water traces is clearly illustrated with a series of experiments. An annular array design philosophy for other problems in NDE/SHM is also discussed.

  12. Entrainment in vertical annular two-phase flow

    International Nuclear Information System (INIS)

    Prediction of amount of entrained droplets or entrainment fraction in annular two-phase flow is essential for the estimation of dryout condition and analysis of post dryout heat transfer in light water nuclear reactors and steam boilers. In this study, air-water and organic fluid (Freon-113) annular flow entrainment experiments have been carried out in 9.4 and 10.2 mm diameter test sections, respectively. Both the experiments covered three distinct pressure conditions and wide range of liquid and gas flow conditions. The organic fluid experiments simulated high pressure steam-water annular flow conditions. In each of the experiments, measurements of entrainment fraction, droplet entrainment rate and droplet deposition rate have been performed by using a liquid film extraction method. A simple, explicit and non-dimensional correlation developed by Sawant et al. (2008a) for the prediction of entrainment fraction is further improved in this study in order to account for the existence of critical gas and liquid flow rates below which no entrainment is possible. Additionally, a new correlation is proposed for the estimation of minimum liquid film flow rate at the maximum entrainment fraction condition. The improved correlation successfully predicted the newly collected air-water and Freon-113 entrainment fraction data. Furthermore, the correlations satisfactorily compared with the air-water, helium-water and air-genklene experimental data measured by Willetts (1987). (author)

  13. Treatment of generalized granuloma annulare - a systematic review.

    Science.gov (United States)

    Lukács, J; Schliemann, S; Elsner, P

    2015-08-01

    Granuloma annulare (GA) is a benign inflammatory skin disease. Localized GA is likely to resolve spontaneously, while generalized GA (GGA) is rare and may persist for decades. GGA usually is resistant to a variety of therapeutic modalities and takes a chronic course. The objective of this study was to summarize all reported treatments of generalized granuloma annulare. This is a systematic review based on MEDLINE, Embase and Cochrane Central Register search of articles in English and German and a manual search, between 1980 and 2013, to summarize the treatment of generalized granuloma annulare. Most medical literature on treatment of GGA is limited to individual case reports and small series of patients treated without a control group. Randomized controlled clinical studies are missing. Multiple treatment modalities for GGA were reported including topical and systemic steroids, PUVA, isotretinoin, dapsone, pentoxifylline, hydroxychloroquine, cyclosporine, IFN-γ, potassium iodide, nicotinamide, niacinamide, salicylic acid, dipyridamole, PDT, fumaric acid ester, etanercept, infliximab, adalimumab. While there are numerous case reports of successful treatments in the literature including surgical, medical and phototherapy options, well-designed, randomized, controlled clinical trials are required for an evidence-based treatment of GGA. PMID:25651003

  14. Experimental study on particles mixing in an annular spouted bed

    Energy Technology Data Exchange (ETDEWEB)

    Hao, Huang; Guoxin, Hu [School of Mechanical Engineering, Shanghai Jiao Tong University, Shanghai 200030 (China); Fengchao, Wang [Science and Technology Development Office, Shanghai Jiao Tong University, Shanghai 200030 (China)

    2008-02-15

    A novel annular spouted bed was developed and studied experimentally. The experiments were performed to examine the effects of feeding mode, air velocity and static bed height as well as particle size on particle mixing for different conditions in this device. The results show that feeding by a rotating cone greatly improves particle mixing by homogeneously projecting the particles into the annular bed. For feeding by a rotating cone, the time required to get uniform mixing laterally is shorten almost 10 times less than that for feeding at a fixed point. With increasing air velocity, the axial mixing speed increases more significantly than the lateral mixing speed. The static bed height has important effects on the uniformity of the final admixtures. With increasing static bed height, the degree of mixing of the final mixture (FDM) axially first decreases, then increases, but laterally, the FDM is monotone decreasing. The particles of small size are helpful to raise the mixing speed. In addition, the dead zone in the annular spouted bed was analyzed. (author)

  15. Experimental study on particles mixing in an annular spouted bed

    Energy Technology Data Exchange (ETDEWEB)

    Huang Hao [School of Mechanical Engineering, Shanghai Jiao Tong University, Shanghai 200030 (China); Hu Guoxin [School of Mechanical Engineering, Shanghai Jiao Tong University, Shanghai 200030 (China)], E-mail: hugx@sjtu.edu.cn; Wang Fengchao [Science and Technology Development Office, Shanghai Jiao Tong University, Shanghai 200030 (China)

    2008-02-15

    A novel annular spouted bed was developed and studied experimentally. The experiments were performed to examine the effects of feeding mode, air velocity and static bed height as well as particle size on particle mixing for different conditions in this device. The results show that feeding by a rotating cone greatly improves particle mixing by homogeneously projecting the particles into the annular bed. For feeding by a rotating cone, the time required to get uniform mixing laterally is shorten almost 10 times less than that for feeding at a fixed point. With increasing air velocity, the axial mixing speed increases more significantly than the lateral mixing speed. The static bed height has important effects on the uniformity of the final admixtures. With increasing static bed height, the degree of mixing of the final mixture (FDM) axially first decreases, then increases, but laterally, the FDM is monotone decreasing. The particles of small size are helpful to raise the mixing speed. In addition, the dead zone in the annular spouted bed was analyzed.

  16. A novel reactor concept for boron neutron capture therapy: annular low-low power reactor (ALLPR)

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, B.; Levine, S.H. [Department of Nuclear Engineering, Pennsylvania State University, University Park, PA 16802 (United States)

    1998-07-01

    Boron Neutron Capture Therapy (BNC), originally proposed in 50's, has been getting renewed attention over the last {approx}10 years. This is in particular due to its potential for treating deep-seated brain tumors by employing epithermal neutron beams. Large (several MW) research reactors are currently used to obtain epithermal beams for BNCT, but because of cost and licensing issues it is not likely that such high-power reactors can be placed in regular medical centers. This paper describes a novel reactor concept for BNCT devised to overcome this obstacle. The design objective was to produce a beam of epithermal neutrons of sufficient intensity for BNCT at <50 kW using low enriched uranium. It is achieved by the annular reactor design, which is called Annular Low-Low Power Reactor (ALLPR). Preliminary studies using Monte Carlo simulations are summarized in this paper. The ALLPR should be relatively economical to build, and safe and easy to operate. This novel concept may increase the viability of using BNCT in medical centers worldwide. (author)

  17. Investigation of primary cooling water chemistry following the partial meltdown of Pu-Be neutron source in Tehran Research Reactor Core (TRR)

    International Nuclear Information System (INIS)

    Research highlights: → Effect of Pu-Be neutron source meltdown in core on reactor water chemistry. → Water chemistry of primary cooling before, during and after of above incident was compared. → Training importance. → Management of nuclear incident and accident. - Abstract: Effect of Pu-Be neutron source meltdown in core on reactor water chemistry was main aim of this study. Leaving the neutron source in the core after reactor power exceeds a few hundred Watts was the main reason for its partial meltdown. Water chemistry of primary cooling before, during and after of above incident was compared. Activity of some radio-nuclides such as Ba-140, La-140, I-131, I-132, Te-132 and Xe-135 increased. Other radio-nuclides such as Nd-147, Xe-133, Sr-91, I-133 and I-135 are also detected which were not existed before this incident.

  18. 武警学院核心竞争力培育研究%The Research on Core Competence of the Armed Police Academy

    Institute of Scientific and Technical Information of China (English)

    高士杰; 刘永泉

    2012-01-01

    The core competence of a university is an ability to maintain a sustainable edge. Based on the analysis and definition of core competence, this pager discusses the way to enforce the core competence of the Armed Police Academy from discipline construction, talent cultivation, academic research, campus culture.%核心竞争力是指高校具有持续竞争优势的一种能力。通过对学院核心竞争力的分析与界定,从学科建设、人才培养、学术研究、校园文化等方面对如何增强核心竞争力进行论述,对学院的建设与发展具有一定的理论探索意义。

  19. Structural assessment of the IEA-R1 research reactor core matrix plate under loads from a new pneumatic irradiation system

    International Nuclear Information System (INIS)

    One of the most recent actions related to the continuous modernization of the IEA-R1 research reactor is the replacement of the old pneumatic irradiation system by a new one. The new system has two positions supported by the core matrix plate introducing loads up to five times greater than the usual loads over it in those positions. From the safety point of view, a structural assessment of the core matrix plate under the new loads is strongly recommended. So, this paper presents this assessment considering that the core matrix plate is simply supported on its corners by the core support frame and its holes in several positions to support fuel elements, control elements, reflectors, etc., and to provide the coolant flow through it. To obtain a realistic structural behavior of the plate a three dimensional finite element model was developed and processed considering its support conditions in the corners and the usual loads from a typical core configuration plus the loads from the new pneumatic system in the most unfavorable positions in terms of stress. The obtained results from the finite element analysis show that there are adequate safety margins under the applied loads described above indicating that no modification or reinforcement of the plate is necessary. (author)

  20. Youngster Badmindon Athletes’ Core Strength Training Research%青少年业余羽毛球选手核心力量训练研究

    Institute of Scientific and Technical Information of China (English)

    王秋云

    2012-01-01

    The badmindon net separated competition project,regardless of is the cannonball serve or vigorously smash all requires the athlete to have good ability to control body,core strength training can improve the core strength and control of body balance function,to the athlete in the process of completing the technical action to provide strong support.Through the literature material law,expert interview law of core strength,core stability and core strength training method of badmindon has carried on the analysis and research.%羽毛球作为一项典型的隔网对抗性项目,无论是炮弹式跳起劈杀,还是底线大力拉高球都要求青少年业余选手有良好身体控制能力,核心力量训练可以提高核心部位的力量以及控制身体稳定平衡的作用,为业余选手在运动过程中完成技术动作提供强有力的支撑。通过文献资料法、专家访谈法对核心力量、核心稳定性及其羽毛球核心力量训练方法应用进行了分析与研究。

  1. Proteomics Core

    Data.gov (United States)

    Federal Laboratory Consortium — Proteomics Core is the central resource for mass spectrometry based proteomics within the NHLBI. The Core staff help collaborators design proteomics experiments in...

  2. Evaluation of loss coefficient for an end plug with side holes in dual cooled annular nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Chang Hwan; Chun, Tae Hyun; Oh, Dong Seok; In, Wang Kee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    The Korea Atomic Energy Research Institute (KAERI) has been developing a dual cooled annular fuel for a power uprate of 20% in an optimized pressurized water reactor (PWR) in Korea, OPR1000. The dual cooled annular fuel is configured to allow coolant flow through the inner channel as well as the outer channel. Several thermal hydraulic issues exist for the application of dual cooled annular fuel to OPR1000. One is the hypothetical event of inner channel blockage because the inner channel is an isolated flow channel without the coolant mixing between the neighboring flow channels. The inner channel blockage could cause a departure from nucleate boiling (DNB) in the inner channel that eventually results in fuel failure. A long lower end plug for the annular fuel was invented to provide flow holes by perforating the side surface of the end plug body. The side holes in the lower end plug are expected to supply a minimum coolant in the inner channel to prevent the DNB occurrence in the event of partial or even complete blockage of the inner channel entrance. But due to the very unusual shape of the lower end plug, it is difficult to estimate the flow resistance of the side flow holes using empirical equations available in the open literature. An experiment and computational fluid dynamics (CFD) analysis were performed to investigate the bypass flow through the side holes of the end plug in the case of complete entrance blockage of the inner channel. The form loss coefficient in the side holes was also estimated using the pressure drop along the bypass flow path.

  3. Steady thermocapillary-buoyant convection in a shallow annular pool.Part 2: Two immiscible fluids

    Institute of Scientific and Technical Information of China (English)

    You-Rong Li; Shuang-Cheng Wang; Chun-Mei Wu

    2011-01-01

    This work is devoted to the study of steady thermocapillary-buoyant convection in a system of two horizontal superimposed immiscible liquid layers filling a lateral heated thin annular pool.The governing equations are solved using an asymptotic theory for the aspect ratios e → 0.Asymptotic solutions of the velocity and temperature fields are obtained in the core region away from the cylinder walls.In order to validate the asymptotic solutions,numerical simulations are also carried out and the results are compared to each other.It is found that the present asymptotic solutions are valid in most of the core region.And the applicability of the obtained asymptotic solutions decreases with the increase of the aspect ratio and the thickness ratio of the two layers.For a system of gallium arsenide (lower layer) and boron oxide (upper layer),the buoyancy slightly weakens the thermocapillary convection in the upper layer and strengthens it in the lower layer.

  4. Using 6- and 8-tube IRT-4M fuel assemblies in WWR-SM research reactor core

    International Nuclear Information System (INIS)

    The WWR-SM reactor at the Institute of Nuclear Physics of Academy of Sciences (INP AS) in Uzbekistan was converted to 6-tube IRT-4M LEU (19.7%) fuel. Presently, INP intends to also use IRT-4M 8-tube FA, and a safety analysis for these 'mixed' (8-tube and 6-tube FA) cores is required by the regulatory authorities. This paper presents results of the safety analysis being performed for the 'mixed' cores. Neutronics analysis (burnup, power distributions and shutdown margin), steady-state thermal hydraulics analysis, kinetics parameters for these mixed cores are discussed in this paper. These results will be used to amend the present SAR. (authors)

  5. Optimization of radially heterogeneous 1000-MW(e) LMFBR core configurations. Appendixes D and E. Research project 620-25

    Energy Technology Data Exchange (ETDEWEB)

    Barthold, W.P.; Orechwa, Y.; Su, S.F.; Hutter, E.; Batch, R.V.; Beitel, J.C.; Turski, R.B.; Lam, P.S.K.

    1979-11-01

    A parameter study was conducted to determine the interrelated effects of: loosely or tightly coupled fuel regions separated by internal blanket assemblies, number of fuel regions, core height, number and arrangement of internal blanket subassemblies, number and size of fuel pins in a subassembly, etc. the effects of these parameters on sodium void reactivity, Doppler, incoherence, breeding gain, and thermohydraulics were of prime interest. Trends were established and ground work laid for optimization of a large, radially-heterogeneous, LMFBR core that will have low energetics in an HCDA and will have good thermal and breeding performance.

  6. Optimization of radially heterogenous 1000-MW(e) LMFBR core configurations. Appendix C. Research project 620-25

    International Nuclear Information System (INIS)

    A parameter study was conducted to determine the interrelated effects of: loosely or tightly coupled fuel regions separated by internal blanket assemblies, number of fuel regions, core height, number and arrangement of internal blanket subassemblies, number and size of fuel pins in a subassembly, etc. The effects of these parameters on sodium void reactivity, Doppler, incoherence, breeding gain, and thermohydraulics were of prime interest. Trends were established and ground work laid for optimization of a large, radially-heterogeneous, LMFBR core that will have low energetics in an HCDA and will have good thermal and breeding performance

  7. Optimization of radially heterogeneous 1000-MW(e) LMFBR core configurations. Appendixes D and E. Research project 620-25

    International Nuclear Information System (INIS)

    A parameter study was conducted to determine the interrelated effects of: loosely or tightly coupled fuel regions separated by internal blanket assemblies, number of fuel regions, core height, number and arrangement of internal blanket subassemblies, number and size of fuel pins in a subassembly, etc. the effects of these parameters on sodium void reactivity, Doppler, incoherence, breeding gain, and thermohydraulics were of prime interest. Trends were established and ground work laid for optimization of a large, radially-heterogeneous, LMFBR core that will have low energetics in an HCDA and will have good thermal and breeding performance

  8. Core thermal-hydraulics behaviour within the framework of the feasibility studies of the RJH research reactor

    International Nuclear Information System (INIS)

    The purpose of this work is to give a preliminary evaluation of the thermal-hydraulic behaviour of the reactor Jules Horowitz (RJH project). These performances are evaluated by using computer codes. First a calculation procedure was drawn then the output data have been elaborated. The calculations are performed by using the operating code SIRENE which provides us with the boundary limits of the core. The 3-dimension thermal-hydraulic code FLICA-4 has allowed us to get an accurate behaviour of the core in various operating modes (nominal, accidental, natural convection). This work is only an introduction and further studies have to be led. (A.C.)

  9. Pancreaticoduodenectomy for pancreas carcinoma occurring in the annular pancreas: report of a case

    OpenAIRE

    Kawaida, Hiromichi; KONO, Hiroshi; Watanabe, Mitsuaki; Maki, Akira; Amemiya, Hidetake; Matsuda, Masanori; Fujii, Hideki; Fukasawa, Mitsuharu; Takahashi, Ei; Sano, Katsuhiro; Inoue, Tomohiro

    2015-01-01

    The annular pancreas is a rare congenital anomaly in which a ring of the pancreas parenchyma surrounds the second part of the duodenum. Malignant tumors are extremely rare in patients with an annular pancreas. A 64-year-old man presented with appetite loss and vomiting. Abdominal contrast-enhanced computed tomography (CT) indicated pancreas parenchyma surrounding the second part of the duodenum, and a hypovascular area occupying lesion in the annular pancreas. Subtotal stomach-preserving panc...

  10. Repeated mitral valve replacement in a patient with extensive annular calcification

    OpenAIRE

    Kitamura Tadashi; Fukuda Sachito; Sawada Takahiro; Miura Sumio; Kigawa Ikutaro; Miyairi Takeshi

    2011-01-01

    Abstract Background Mitral valve replacement in the presence of severe annular calcification is a technical challenge. Case report A 47-year-old lady who had undergone mitral and aortic valve replacement for rheumatic disease 27 years before presented with dyspnea. At reoperation, extensive mitral annular calcification was hindering the disc motion of the Starr-Edwards mitral prosthesis. The old prosthesis was removed and a St Jude Medical mechanical valve was implanted after thorough annular...

  11. Study of startup conditions of a pulsed annular reactor; Estudo das reacoes de partida de um reator anelar pulsado

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Mario Augusto Bezerra da

    2003-10-15

    A new concept of reactor, which combines features of pulsed and stationary reactors, was proposed so as to produce intense neutronic fluxes. Such a reactor, known as VICHFPR (Very Intense Continuous High Flux Pulsed Reactor), consists of a subcritical core with an annular geometry and pulsed by a rotating reflector which acts as a reactivity modulator as it produces a short pulse (approximately equal to 1 ms) of high intensity, guiding the region near the pulser to super-prompt critical state. This dissertation intends to analyze the startup conditions of a Pulsed Annular Reactor. The evolution of the neutron pulse intensity is analyzed when the reactivity modulator is brought upwards according to a helicoidal path from its initial position (far away from the core), when the multiplication factor has a subcritical value, up to the final position (near the core), in which a super-prompt critical state is reached. Part of the analysis is based on the variation of neutron reflection, which is a uniform function of the exit and reflection angles between the core and the modulator. It must be emphasized that this work is an approximation of the real situation. As the initial and final reactor parameters are known, a programming code in Fortran is worked out to provide the multiplication factor and the flux intensity evolution. According to the results obtained with this code, the conditions under which the modulator must be lifted up during the startup are established. Basically, these conditions are related to the analysis of the rising and the rotation velocities, the reflector saving and the initial distance between the reactor and the modulator. The Pulsed Annular Reactor startup was divided into three stages. Because of its negative reactivity in the first two stages, the neutron multiplication is not large, while the last one, having a positive reactivity, shows an intense multiplication as is usually expected when handling pulsed systems. This last stage is quite

  12. Development of probabilistic design method for annular fuel. Development of BORNFREE-CEPTAR code

    International Nuclear Information System (INIS)

    The increase of linear power and burn-up during the reactor operation is considered as one of measures for the utility of fast reactor in future, and then the application of annular fuels is under consideration. In order to make a design for thus annular fuels, annular fuel design code 'CEPTAR' has been developed in Japan Atomic Energy Agency (JAEA). In addition, probabilistic fuel design code 'BORNFREE' has been also developed for the reasonable fuel design with safety and the quantitative evaluation of design margin. In this study, aiming at the development of probabilistic design method, we developed BORNFREE-CEPTAR code to develop the reasonable design method for annular fuels. As the results of probability evaluation of fuel melting at the transient at the initial power increase, by using the probabilistic annular fuel design code 'BORNFREE-CEPTAR', the melting probability for annular fuels was estimated to be approximately two figures lower than that for solid fuels, and the remarkable decrease of melting probability, which was caused by the fuel restructuring effect, was seen in the estimation results for solid fuels, on the other hand, the results for annular fuels indicated that this effect was comparably small. In addition, the permissive linear power for annular fuels tended to enhance from that for solid fuels with the increase of initial central-hole diameter under the similar fuel melting probability condition. This indicated the possibility of higher linear power operation for high-density annular fuels than low-density solid fuels. (author)

  13. Electron beam diagnostic system using computed tomography and an annular sensor

    Energy Technology Data Exchange (ETDEWEB)

    Elmer, John W.; Teruya, Alan T.

    2015-08-11

    A system for analyzing an electron beam including a circular electron beam diagnostic sensor adapted to receive the electron beam, the circular electron beam diagnostic sensor having a central axis; an annular sensor structure operatively connected to the circular electron beam diagnostic sensor, wherein the sensor structure receives the electron beam; a system for sweeping the electron beam radially outward from the central axis of the circular electron beam diagnostic sensor to the annular sensor structure wherein the electron beam is intercepted by the annular sensor structure; and a device for measuring the electron beam that is intercepted by the annular sensor structure.

  14. Electron beam diagnostic system using computed tomography and an annular sensor

    Science.gov (United States)

    Elmer, John W.; Teruya, Alan T.

    2014-07-29

    A system for analyzing an electron beam including a circular electron beam diagnostic sensor adapted to receive the electron beam, the circular electron beam diagnostic sensor having a central axis; an annular sensor structure operatively connected to the circular electron beam diagnostic sensor, wherein the sensor structure receives the electron beam; a system for sweeping the electron beam radially outward from the central axis of the circular electron beam diagnostic sensor to the annular sensor structure wherein the electron beam is intercepted by the annular sensor structure; and a device for measuring the electron beam that is intercepted by the annular sensor structure.

  15. Neutronics analysis of the initial core of the TRIGA Mark II reactor

    Energy Technology Data Exchange (ETDEWEB)

    Khan, R., E-mail: rustamzia@yahoo.co [Vienna University of Technology, Atominstitute (ATI), Stadion allee 2, A-1020, Vienna (Austria); Stummer, T.; Boeck, H.; Villa, M. [Vienna University of Technology, Atominstitute (ATI), Stadion allee 2, A-1020, Vienna (Austria)

    2011-05-15

    Highlights: The TRIGA Mark II Vienna is modeled employing MCNP5. The model is confirmed through three different experiments. Initial critical, reactivity distribution and flux mapping experiment. - Abstract: The Atominstitute (ATI) of Vienna University of Technology (VUT) operates a TRIGA Mark II research reactor since March 1962. Its initial criticality was achieved on 7th March 1962 when 57th Fuel Element (FE) was loaded to the core. This paper describes the development of the MCNP model of the TRIGA reactor and its validation through three different experiments i.e. initial criticality, reactivity distribution and a thermal flux mapping experiment in the reactor core. All these experiments were performed on the initial core configuration. The MCNP model includes all necessary core components i.e. FE, Graphite Element GE, neutron Source Element (SE), Central IRradiation channel (CIR) etc. Outside the core, this model simulates the annular grooved graphite reflector, the thermal and thermalizing column, four beam tubes and the reactor water tank up to 100 cm in radial and +60 and -60 cm in axial direction. Each grid position at its exact location is modeled. This model employs the ENDF/B-VI data library except for the Sm-isotopes which are taken from JEFF 3.1 because ENDF/B-VI lacks samarium (Sm) cross sections. For the first experiment, the model predicts an effective multiplication factor ({kappa}{sub eff}) of 1.00183 with an estimated standard deviation 0.00031 which is very close to the experimental value 1.00114. The second experiment measures the reactivity values of four FE and one GE. In comparison to the MCNP results, the percent difference ranges from 4 to 22. The third experiment verifies the model at a local level with the radial and axial thermal flux density distribution in the core. Though the trends are similar, the MCNP model overestimates the radial thermal flux density in the core and underestimates these results at the core periphery.

  16. Neutronics analysis of the initial core of the TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    Highlights: → The TRIGA Mark II Vienna is modeled employing MCNP5. → The model is confirmed through three different experiments. → Initial critical, reactivity distribution and flux mapping experiment. - Abstract: The Atominstitute (ATI) of Vienna University of Technology (VUT) operates a TRIGA Mark II research reactor since March 1962. Its initial criticality was achieved on 7th March 1962 when 57th Fuel Element (FE) was loaded to the core. This paper describes the development of the MCNP model of the TRIGA reactor and its validation through three different experiments i.e. initial criticality, reactivity distribution and a thermal flux mapping experiment in the reactor core. All these experiments were performed on the initial core configuration. The MCNP model includes all necessary core components i.e. FE, Graphite Element GE, neutron Source Element (SE), Central IRradiation channel (CIR) etc. Outside the core, this model simulates the annular grooved graphite reflector, the thermal and thermalizing column, four beam tubes and the reactor water tank up to 100 cm in radial and +60 and -60 cm in axial direction. Each grid position at its exact location is modeled. This model employs the ENDF/B-VI data library except for the Sm-isotopes which are taken from JEFF 3.1 because ENDF/B-VI lacks samarium (Sm) cross sections. For the first experiment, the model predicts an effective multiplication factor (κeff) of 1.00183 with an estimated standard deviation 0.00031 which is very close to the experimental value 1.00114. The second experiment measures the reactivity values of four FE and one GE. In comparison to the MCNP results, the percent difference ranges from 4 to 22. The third experiment verifies the model at a local level with the radial and axial thermal flux density distribution in the core. Though the trends are similar, the MCNP model overestimates the radial thermal flux density in the core and underestimates these results at the core periphery.

  17. The Research and Development of an International Core Curriculum for Information and Communications Technology in Teacher Training.

    Science.gov (United States)

    Davis, Niki; Tearle, Penni

    This paper outlines progress of the European Commission (EC) supported Telematics for Teacher Training (T3) project in the development and dissemination of a European Core Curriculum for Information and Communications Technology (ICT) in Teacher Training. National and European governments have recognized the importance of training teachers in the…

  18. Final Report, Nuclear Energy Research Initiative (NERI) Project: An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model

    International Nuclear Information System (INIS)

    OAK (B204) Final Report, NERI Project: ''An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model'' The present generation of reactor analysis methods uses few-group nodal diffusion approximations to calculate full-core eigenvalues and power distributions. The cross sections, diffusion coefficients, and discontinuity factors (collectively called ''group constants'') in the nodal diffusion equations are parameterized as functions of many variables, ranging from the obvious (temperature, boron concentration, etc.) to the more obscure (spectral index, moderator temperature history, etc.). These group constants, and their variations as functions of the many variables, are calculated by assembly-level transport codes. The current methodology has two main weaknesses that this project addressed. The first weakness is the diffusion approximation in the full-core calculation; this can be significantly inaccurate at interfaces between different assemblies. This project used the nodal diffusion framework to implement nodal quasidiffusion equations, which can capture transport effects to an arbitrary degree of accuracy. The second weakness is in the parameterization of the group constants; current models do not always perform well, especially at interfaces between unlike assemblies. The project developed a theoretical foundation for parameterization and homogenization models and used that theory to devise improved models. The new models were extended to tabulate information that the nodal quasidiffusion equations can use to capture transport effects in full-core calculations

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

    International Nuclear Information System (INIS)

    Separate abstracts have been prepared for each paper presented in the following areas of interest: (1) fuel development; (2) post-irradiation examinations; (3) reprocessing; (4) thermite reaction; (5) fuel fabrication; (6) element tests; (7) core tests; (8) criticals; (9) shipping; and (10) reactors and methods

  20. Numerical simulation on dimension decrease for annular casing of one centrifugal boiler circulation pump

    International Nuclear Information System (INIS)

    Primary formulation derivation indicates that the dimension of one existing centrifugal boiler circulation pump casing is too large. As great manufacture cost can be saved by dimension decrease, a numerical simulation research is developed in this paper on dimension decrease for annular casing of this pump with a specific speed equaling to 189, which aims at finding an appropriately smaller dimension of the casing while hydraulic performance and strength performance will hardly be changed according to the requirements of the cooperative company. The research object is one existing centrifugal pump with a diffuser and a semi-spherical annular casing, working as the boiler circulation pump for (ultra) supercritical units in power plants. Dimension decrease, the modification method, is achieved by decreasing the existing casing's internal radius (marked as Ri0) while keeping the wall thickness. The research analysis is based on primary formulation derivation, CFD (Computational Fluid Dynamics) simulation and FEM (Finite Element Method) simulation. Primary formulation derivation estimates that a design casing's internal radius should be less than 0.75 Ri0. CFD analysis indicates that smaller casing with 0.75 Ri0 has a worse hydraulic performance when working at large flow rates and a better hydraulic performance when working at small flow rates. In consideration of hydraulic performance and dimension decrease, an appropriate casing's internal radius is determined, which equals to 0.875 Ri0. FEM analysis then confirms that modified pump casing has nearly the same strength performance as the existing pump casing. It is concluded that dimension decrease can be an economical method as well as a practical method for large pumps in engineering fields.

  1. Modeling of annular film dryout with Cobra-TF

    International Nuclear Information System (INIS)

    The COBRA-TF computer code uses a two-fluid, three-field and three-dimensional formulation to model a two-phase flow field in a specific geometry. The liquid phase is divided in a continuous liquid field and a separate dispersed field, which is used to describe the entrained liquid drops. For each space dimension, the code solves three momentum equations, three mass conservation equations and two energy conservation equations. Entrainment and depositions models are implemented into the code to model the mass transfer between the two liquid fields. In annular flow condition critical heat flux is caused by annular film dryout. Film dryout is a complex function of the film flow rate, the applied heat flux, and the entrainment from the liquid film to the continuous vapor region, and the deposition of entrained droplets back to the liquid film. Because of the three-field approach, COBRA-TF hydrodynamic equations are able to predict dry-out by solving directly the film dry-out as a hydrodynamic process rather than using an empirical dry-out correlation. The dry-out is driven by the hydraulic calculation and the prediction is the result of the combined effect of the entrainment, the deposition models and interfacial heat transfer. The paper discusses the annular film entrainment and deposition models used in the code as well as the logic, which is used to determine the dry-out phenomena as the film thickness decreases. The obtained results with COBRA-TF are compared with the test data from the Bennett Tube Dry-out Experiments. In general, the COBRA-TF prediction of the dry-out location is in good agreement with Bennett test data. In particular, results show that the predicted dry-out length tends to be longer than the measured value and in the post dry-out region the wall temperature, which is dependent on vapor superheat, tends to be underestimated by the code. (authors)

  2. Radiation Characterization Summary: ACRR Polyethylene-Lead-Graphite (PLG) Bucket Located in the Central Cavity on the 32-Inch Pedestal at the Core Centerline (ACRR-PLG-CC-32-cl).

    Energy Technology Data Exchange (ETDEWEB)

    Parma, Edward J.,; Vehar, David W.; Lippert, Lance L.; Griffin, Patrick J.; Naranjo, Gerald E.; Luker, Spencer M.

    2015-06-01

    This document presents the facility-recommended characterization of the neutron, prompt gamma-ray, and delayed gamma-ray radiation fields in the Annular Core Research Reactor (ACRR) for the polyethylene-lead-graphite (PLG) bucket in the central cavity on the 32-inch pedestal at the core centerline. The designation for this environment is ACRR-PLG-CC-32-cl. The neutron, prompt gamma-ray, and delayed gamma-ray energy spectra, uncertainties, and covariance matrices are presented as well as radial and axial neutron and gamma-ray fluence profiles within the experiment area of the bucket. Recommended constants are given to facilitate the conversion of various dosimetry readings into radiation metrics desired by experimenters. Representative pulse operations are presented with conversion examples. Acknowledgements The authors wish to thank the Annular Core Research Reactor staff and the Radiation Metrology Laboratory staff for their support of this work. Also thanks to David Ames for his assistance in running MCNP on the Sandia parallel machines.

  3. Characteristic analysis of a double stator annular linear electromagnetic pump

    International Nuclear Information System (INIS)

    A annular linear induction electromagnetic pump (ALIP) is generally used to transport liquid sodium coolants for liquid metal reactors. In the present study, the theoretical induction of a developing equation has been carried out for a double stator version of the ALIP which is noticebly employed for the sodium circulation of a large flowrate. The computerzed P-Q relation, which is represented by the pump geometrical and electrical variables, has been applied to a double stator version of the ALMR EM pump. An induced equation was verified by the compared analysis with the known data on the P-Q characteristic according to the input currents

  4. A high efficiency annular dark field detector for STEM.

    Science.gov (United States)

    Kirkland, E J; Thomas, M G

    1996-01-01

    A new high efficiency annular dark field (ADF) detector for an HB501 STEM (Scanning Transmission Electron Microscope) has been constructed and tested. This detector uses a single crystal YAP scintillator and a solid quartz light pipe extending from the scintillator (inside the vacuum) to the photomultiplier tube (outside the vacuum). A factor of approximately 100 improvement in signal relative to the original detector has been obtained. This has substantially improved the signal to noise ratio in the recorded high resolution ADF-STEM images. PMID:22666919

  5. Unusual Presentation of Acute Annular Urticaria: A Case Report

    Directory of Open Access Journals (Sweden)

    Gilles Guerrier

    2011-01-01

    Full Text Available Acute urticarial lesions may display central clearing with ecchymotic or haemorrhagic hue, often misdiagnosed as erythema multiforme, serum-sickness-like reactions, or urticarial vasculitis. We report a case of acute annular urticaria with unusual presentation occurring in a 20-month-old child to emphasize the distinctive morphologic manifestations in a single disease. Clinicians who care for children should be able to differentiate acute urticaria from its clinical mimics. A directed history and physical examination can reliably orientate necessary diagnostic testing and allow for appropriate treatment.

  6. Analytic vortex dynamics in an annular Bose-Einstein condensate

    Science.gov (United States)

    Toikka, L. A.; Suominen, K.-A.

    2016-05-01

    We consider analytically the dynamics of an arbitrary number and configuration of vortices in an annular Bose-Einstein condensate obtaining expressions for the free energy and vortex precession rates to logarithmic accuracy. We also obtain lower bounds for the lifetime of a single vortex in the annulus. Our results enable a closed-form analytic treatment of vortex-vortex interactions in the annulus that is exact in the incompressible limit. The incompressible hydrodynamics that is developed here paves the way for more general analytical treatments of vortex dynamics in non-simply-connected geometries.

  7. New fluxon resonant mechanism in annular Josephson tunnel structures

    International Nuclear Information System (INIS)

    A novel dynamical state has been observed in the dynamics of a perturbed sine-Gordon system. This resonant state has been experimentally observed as a singularity in the dc current-voltage characteristic of an annular Josephson tunnel junction, excited in the presence of a magnetic field. In this respect it can be assimilated to self-resonances known as Fiske steps. Differently from these, however, we demonstrate, on the basis of numerical simulations, that its detailed dynamics involves rotating fluxon pairs, a mechanism associated, so far, to self-resonances known as zero-field steps. This occurs because the size of nonlinear excitations is comparable with that of the system

  8. Exhaust emissions of a double annular combustor: Parametric study

    Science.gov (United States)

    Schultz, D. F.

    1974-01-01

    A full scale double-annular ram-induction combustor designed for Mach 3.0 cruise operation was tested. Emissions of oxides of nitrogen, carbon monoxide, unburned hydrocarbons, and smoke were measured over a range of combustor operating variables including reference velocity, inlet air temperature and pressure, and exit average temperature. ASTM Jet-A fuel was used for these tests. An equation is provided relating oxides of nitrogen emissions as a function of the combustor, operating variables. A small effect of radial fuel staging on reducing exhaust emissions (which were originally quite low) is demonstrated.

  9. Annular burnout data from rod-bundle experiments

    International Nuclear Information System (INIS)

    Burnout data for annular flow in a rod bundle are presented for both transient and steady-state conditions. Tests were performed at the Oak Ridge National Laboratory in the Thermal Hydraulic Test Facility (THTF), a pressurized-water loop containing an electrically heated 64-rod bundle. The bundle configuration is typical of later generation pressurized-water reactors with 17 x 17 fuel arrays. Both axial and radial power profiles are flat. All experiments were carried out in upflow with subcooled inlet conditions, insuring accurate flow measurement. Conditions within the bundle were typical of those which could be encountered during a nuclear reactor loss-of-coolant accident

  10. Steady-state thermal hydraulic and safety analyses of a proposed mixed fuel (HEU and LEU) core for Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Pakistan Research Reactor (PARR-1) was converted from Highly Enriched Uranium (HEU) to Low Enriched Uranium (LEU) fuel, in 1992. The reactor is running successfully with an upgraded power level of 10 MW. In order to save money on the purchase of costly fresh LEU fuel elements, it is being thought to use some of the less burnt HEU spent fuel elements along with the present LEU fuel elements. In the present study steady-state thermal hydraulics of a proposed mixed fuel core (see Fig. 2) has been carried out. Results show that the proposed core, comprising of 24 LEU and 5 HEU standard fuel elements, with 4 LEU and one HEU control fuel elements, can be safely operated at a power level of 9.86 MW without compromising on safety. Standard computer codes and correlations were employed to compute various parameters, which include: coolant velocity distribution in the core; critical velocity; pressure drop; saturation temperature; temperature distribution in the core and margins to Onset of Nucleate Boiling (ONB), Onset of Flow Instability (OFI) and Departure from Nucleate Boiling (DNB)

  11. Steady-state thermal hydraulic and safety analyses of a proposed mixed fuel (HEU and LEU) core for Pakistan research reactor-1

    Energy Technology Data Exchange (ETDEWEB)

    Bokhari, I.H. E-mail: ishtiaq@pinstech.org.pk

    2004-07-01

    Pakistan Research Reactor (PARR-1) was converted from Highly Enriched Uranium (HEU) to Low Enriched Uranium (LEU) fuel, in 1992. The reactor is running successfully with an upgraded power level of 10 MW. In order to save money on the purchase of costly fresh LEU fuel elements, it is being thought to use some of the less burnt HEU spent fuel elements along with the present LEU fuel elements. In the present study steady-state thermal hydraulics of a proposed mixed fuel core (see Fig. 2) has been carried out. Results show that the proposed core, comprising of 24 LEU and 5 HEU standard fuel elements, with 4 LEU and one HEU control fuel elements, can be safely operated at a power level of 9.86 MW without compromising on safety. Standard computer codes and correlations were employed to compute various parameters, which include: coolant velocity distribution in the core; critical velocity; pressure drop; saturation temperature; temperature distribution in the core and margins to Onset of Nucleate Boiling (ONB), Onset of Flow Instability (OFI) and Departure from Nucleate Boiling (DNB)

  12. The effect of ozone depletion on the Southern Annular Mode and stratosphere-troposphere coupling

    Science.gov (United States)

    Dennison, Fraser W.; McDonald, Adrian J.; Morgenstern, Olaf

    2015-07-01

    The aim of this study is to investigate the influence of ozone depletion and recovery on the Southern Annular Mode (SAM) and stratosphere-troposphere coupling. Using the National Institute of Water and Atmospheric Research-United Kingdom Chemistry and Aerosols chemistry-climate model, we compare reference runs that include forcing due to greenhouse gases and ozone-depleting substances to sensitivity simulations in which ozone-depleting substances are fixed at their 1960 levels. We find that ozone depletion leads to an increased frequency of extreme anomalies and increased persistence of the SAM in the stratosphere as well as stronger, more persistent stratosphere-troposphere coupling. Currently, the stratosphere provides an appreciable amount of predictability to the troposphere on timescales of 1 or 2 months; however, we find that this effect reduces over time as stratospheric ozone recovers to preozone hole levels toward the latter part of this century.

  13. Nanda-gikendaasowin Naawij Gaa-izhiwebakin Manoomini-zaaga'iganiing: Core-based research by Native students on wild rice lakes in northern Minnesota

    Science.gov (United States)

    Myrbo, A.; Howes, T.; Defoe, R.; Dalbotten, D. M.; Pellerin, H.; McEathron, M.; Ito, E.

    2011-12-01

    Little is known about how local and global environmental changes affect the habitat of wild rice (manoomin in Ojibwe; Zizania sp.). Using transects of sediment cores from wild rice lakes on the Fond du Lac Band of Lake Superior Chippewa Reservation (FDL) in Minnesota, undergraduate student researchers are working to reconstruct the lakes' ecological history in order to better manage future change. Reservation Resource Management personnel and University science mentors work together to develop research questions and mentor small groups of college-age students during short (two-week) and long (ten-week) summer internships. Cores are collected during the winter from the frozen lake surface with "Lake Teams" of mainly Native junior high and high school students attending weekend science camps, who also visit LacCore (the National Lacustrine Core Facility) in Minneapolis to conduct initial core description and basic analyses. At the same time as the Fond du Lac Band gains information about the long-term history and variability of the Reservation's lakes, young Native people are exposed to primary research, natural resources management and academia as occupations, and scientists as people. Scientific results, as well as the results of program evaluation, show clearly that this approach has so far been successful and eye-opening for both students and mentors. Lead-210 dated records of the past ~150 years cover the period of European settlement, logging, and the massive ditching of FDL lakes to convert wetlands to agricultural land. Phytolith, pollen, plant macrofossil, and diatom studies by interns, as well as sediment composition and mass accumulation rate data, show anthropogenic lake level and vegetation fluctuations associated with these activities. Earlier in the record (~10,000 years to ~100 years before present), the natural and slow processes of lake infilling and encroachment of shallow-water vegetation are the dominant processes controlling the ecology of the

  14. Continuum-kinetic-microscopic model of lung clearance due to core-annular fluid entrainment

    Energy Technology Data Exchange (ETDEWEB)

    Mitran, Sorin, E-mail: mitran@unc.edu

    2013-07-01

    The human lung is protected against aspirated infectious and toxic agents by a thin liquid layer lining the interior of the airways. This airway surface liquid is a bilayer composed of a viscoelastic mucus layer supported by a fluid film known as the periciliary liquid. The viscoelastic behavior of the mucus layer is principally due to long-chain polymers known as mucins. The airway surface liquid is cleared from the lung by ciliary transport, surface tension gradients, and airflow shear forces. This work presents a multiscale model of the effect of airflow shear forces, as exerted by tidal breathing and cough, upon clearance. The composition of the mucus layer is complex and variable in time. To avoid the restrictions imposed by adopting a viscoelastic flow model of limited validity, a multiscale computational model is introduced in which the continuum-level properties of the airway surface liquid are determined by microscopic simulation of long-chain polymers. A bridge between microscopic and continuum levels is constructed through a kinetic-level probability density function describing polymer chain configurations. The overall multiscale framework is especially suited to biological problems due to the flexibility afforded in specifying microscopic constituents, and examining the effects of various constituents upon overall mucus transport at the continuum scale.

  15. Continuum-kinetic-microscopic model of lung clearance due to core-annular fluid entrainment

    International Nuclear Information System (INIS)

    The human lung is protected against aspirated infectious and toxic agents by a thin liquid layer lining the interior of the airways. This airway surface liquid is a bilayer composed of a viscoelastic mucus layer supported by a fluid film known as the periciliary liquid. The viscoelastic behavior of the mucus layer is principally due to long-chain polymers known as mucins. The airway surface liquid is cleared from the lung by ciliary transport, surface tension gradients, and airflow shear forces. This work presents a multiscale model of the effect of airflow shear forces, as exerted by tidal breathing and cough, upon clearance. The composition of the mucus layer is complex and variable in time. To avoid the restrictions imposed by adopting a viscoelastic flow model of limited validity, a multiscale computational model is introduced in which the continuum-level properties of the airway surface liquid are determined by microscopic simulation of long-chain polymers. A bridge between microscopic and continuum levels is constructed through a kinetic-level probability density function describing polymer chain configurations. The overall multiscale framework is especially suited to biological problems due to the flexibility afforded in specifying microscopic constituents, and examining the effects of various constituents upon overall mucus transport at the continuum scale

  16. LMFBR in-core thermal-hydraulics: the state of the art and US research and development needs

    Energy Technology Data Exchange (ETDEWEB)

    Khan, E.U.

    1980-04-01

    A detailed critical review is presented of the literature relevant to predicting coolant flow and temperature fields in LMFBR core assemblies for nominal and non-nominal rod bundle geometries and reactor operating conditions. The review covers existing thermal-hydraulic models, computational methods, and experimental data useful for the design of an LMFBR core. The literature search made for this review included publications listed by Nuclear Science Abstracts and Energy Data Base as well as papers presented at key nuclear conferences. Based on this extensive review, the report discusses the accuracy with which the models predict flow and temperature fields in rod assemblies, identifying areas where analytical, experimental, and model development needs exist.

  17. Exploitation limits and safety operation limits for initial compact loading of the WWR-K research reactor core

    International Nuclear Information System (INIS)

    Analysis of initial compact loading of the WWR-K reactor is carried out. It is defined, that reactor reactivity supply should not exceed of 5,24 % (ΔK/K) under calculated efficiencies values of reactor control and protection control systems units in operation cycle start. Value of reactivity supply one should be corrected with taking into account real values of protection control system units efficiencies. Values of normal operation limits are defined by reactor power level in dependence of water expense in first contour and water temperature on entrance of reactor core. Values of reactor operation limits norms are chosen by reactor power (precautionary signalling (PS) setting). The PS setting of reactor core for thermo-hydraulic parameters monitoring is chosen as well

  18. LMFBR in-core thermal-hydraulics: the state of the art and US research and development needs

    International Nuclear Information System (INIS)

    A detailed critical review is presented of the literature relevant to predicting coolant flow and temperature fields in LMFBR core assemblies for nominal and non-nominal rod bundle geometries and reactor operating conditions. The review covers existing thermal-hydraulic models, computational methods, and experimental data useful for the design of an LMFBR core. The literature search made for this review included publications listed by Nuclear Science Abstracts and Energy Data Base as well as papers presented at key nuclear conferences. Based on this extensive review, the report discusses the accuracy with which the models predict flow and temperature fields in rod assemblies, identifying areas where analytical, experimental, and model development needs exist

  19. Research and practice on the training mode of the core competence about the application-oriented electrical information engineering undergraduate

    OpenAIRE

    Bai Jing; Song Yanxia

    2016-01-01

    Nowadays, the method of cultivating talents in colleges and universities is out of line with the social demand on the training goal of electrical information talents in local application-oriented colleges and universities. To solve this problem, this paper put forward the concept of “five view” about the engineering education (practice) and the engineering core competence, systematically solved the key problems such as the training mode of application-oriented undergraduate, positioning, teac...

  20. Design research of the possible replacement of IVG.1M reactor core fuel into U - Mo alloy

    International Nuclear Information System (INIS)

    Results of neutron-physics calculations, characteristics of IVG. 1M reactor in case of core fuel replacement to low enrichment fuel from U-Mo alloy was represented. By means of MCNP/4B design code, assigned for radioactive transport in three-dimensional geometry task solving, effective multiplication factor for reactor, initial reactivity margin, relative neutron flux density, specific energy release were defined. (author)