Sample records for core neutronic design

  1. Advanced neutron source final preconceptual reference core design

    Microsoft Academic Search

    G. L. Copeland; W. R. Gambill; R. M. Harrington; J. A. Johnson; F. J. Peretz; H. Reutler; J. M. Ryskamp; D. L. Selby; C. D. West; G. L. Yoder

    1989-01-01

    The preconceptual design phase of the Advanced Neutron Source (ANS) Project ended with the selection of a reference reactor core that will be used to begin conceptual design work. The new reference core consists of two involute fuel elements, of different diameters, aligned axially with a small axial gap between them. The use of different element diameters permits a separate

  2. McCARD for Neutronics Design and Analysis of Research Reactor Cores

    NASA Astrophysics Data System (ADS)

    Shim, Hyung Jin; Park, Ho Jin; Kwon, Soonwoo; Seo, Geon Ho; Hyo Kim, Chang

    2014-06-01

    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, and YALINA.

  3. PWR core design, neutronics evaluation and fuel cycle analysis for thorium-uranium breeding recycle

    SciTech Connect

    Bi, G.; Liu, C.; Si, S. [Shanghai Nuclear Engineering Research and Design Inst., No. 29, Hongcao Road, Shanghai, 200233 (China)

    2012-07-01

    This paper was focused on core design, neutronics evaluation and fuel cycle analysis for Thorium-Uranium Breeding Recycle in current PWRs, without any major change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. The fuel cycle analysis indicates that Thorium-Uranium Breeding Recycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR core utilizing 17 x 17 fuel assemblies (FAs) was taken as the model core. Two mixed cores were investigated respectively loaded with mixed reactor grade Plutonium-Thorium (PuThOX) FAs and mixed reactor grade {sup 233}U-Thorium (U{sub 3}ThOX) FAs on the basis of reference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixed core consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade {sup 233}U extracted from burnt PuThOX fuel was used to fabrication of U{sub 3}ThOX for starting Thorium-. Uranium breeding recycle. In UOX/U{sub 3}ThOX mixed core, the well designed U{sub 3}ThOX FAs with 1.94 w/o fissile uranium (mainly {sup 233}U) were located on the periphery of core as a blanket region. U{sub 3}ThOX FAs remained in-core for 6 cycles with the discharged burnup achieving 28 GWD/tHM. Compared with initially loading, the fissile material inventory in U{sub 3}ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuel assemblies were located in the inner of UOX/U{sub 3}ThOX mixed core refueling with 64 FAs at each cycle. The designed UOX/PuThOX and UOX/U{sub 3}ThOX mixed core satisfied related nuclear design criteria. The full core performance analyses have shown that mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristic parameters, such as reduced differential boron worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced control rod worth, reduced shutdown margin, etc.; while mixed core with U{sub 3}ThOX loading on the periphery of core has no visible impacts on neutronic characteristics compared with reference full UOX core. The fuel cycle analysis has shown that {sup 233}U mono-recycling with U{sub 3}ThOX fuel could save 13% of natural uranium resource compared with UOX once through fuel cycle, slightly more than that of Plutonium single-recycling with MOX fuel. If {sup 233}U multi-recycling with U{sub 3}ThOX fuel is implemented, more natural uranium resource would be saved. (authors)

  4. Neutronics methods, models, and applications at the Idaho National Engineering Laboratory for the advanced neutron source reactor three-element core design

    Microsoft Academic Search

    C. A. Wemple; B. G. Schnitzler; J. M. Ryskamp

    1995-01-01

    A summary of the methods and models used to perform neutronics analyses on the Advanced Neutron Source reactor three-element core design is presented. The applications of the neutral particle Monte Carlo code MCNP are detailed, as well as the expansion of the static role of MCNP to analysis of fuel cycle depletion calculations. Results to date of these applications are

  5. THE DEVELOPMENT OF MODERN DESIGN AND REFERENCE CORE NEUTRONICS METHODS FOR THE PBMR

    Microsoft Academic Search

    T D NEWTON

    The Pebble Bed Modular Reactor (PBMR) introduces several challenges for core neutronic methods. The particulate fuel is highly heterogeneous with a random distribution within the fuel pebbles and requires unique methods to calculate the effects of fuel resonance self shielding. In addition, the flow of fuel through the core is specific to the PBMR, again requiring specialised methods to model

  6. Advanced neutron source reactor conceptual safety analysis report, three-element-core design: Chapter 15, accident analysis

    SciTech Connect

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L.; Harrington, R.M.

    1996-02-01

    In order to utilize reduced enrichment fuel, the three-element-core design for the Advanced Neutron Source has been proposed. The proposed core configuration consists of inner, middle, and outer elements, with the middle element offset axially beneath the inner and outer elements, which are axially aligned. The three-element-core RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. To assess the impact of changes in the core region configuration and the thermal-hydraulic steady-state conditions, the safety analysis has been updated. This report gives the safety margins for the loss-of-off-site power and pressure-boundary fault accidents based on the RELAP5 results. AU margins are greater for the three-element-core simulations than those calculated for the two-element core.

  7. A Monte Carlo model system for core analysis and epithermal neutron beam design at the Washington State University Radiation Center

    SciTech Connect

    Burns, T.D. Jr.

    1996-05-01

    The Monte Carlo Model System (MCMS) for the Washington State University (WSU) Radiation Center provides a means through which core criticality and power distributions can be calculated, as well as providing a method for neutron and photon transport necessary for BNCT epithermal neutron beam design. The computational code used in this Model System is MCNP4A. The geometric capability of this Monte Carlo code allows the WSU system to be modeled very accurately. A working knowledge of the MCNP4A neutron transport code increases the flexibility of the Model System and is recommended, however, the eigenvalue/power density problems can be run with little direct knowledge of MCNP4A. Neutron and photon particle transport require more experience with the MCNP4A code. The Model System consists of two coupled subsystems; the Core Analysis and Source Plane Generator Model (CASP), and the BeamPort Shell Particle Transport Model (BSPT). The CASP Model incorporates the S({alpha}, {beta}) thermal treatment, and is run as a criticality problem yielding, the system eigenvalue (k{sub eff}), the core power distribution, and an implicit surface source for subsequent particle transport in the BSPT Model. The BSPT Model uses the source plane generated by a CASP run to transport particles through the thermal column beamport. The user can create filter arrangements in the beamport and then calculate characteristics necessary for assessing the BNCT potential of the given filter want. Examples of the characteristics to be calculated are: neutron fluxes, neutron currents, fast neutron KERMAs and gamma KERMAs. The MCMS is a useful tool for the WSU system. Those unfamiliar with the MCNP4A code can use the MCMS transparently for core analysis, while more experienced users will find the particle transport capabilities very powerful for BNCT filter design.

  8. Neutronics methods, models, and applications at the Idaho National Engineering Laboratory for the advanced neutron source reactor three-element core design

    SciTech Connect

    Wemple, C.A.; Schnitzler, B.G.; Ryskamp, J.M. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

    1995-08-01

    A summary of the methods and models used to perform neutronics analyses on the Advanced Neutron Source reactor three-element core design is presented. The applications of the neutral particle Monte Carlo code MCNP are detailed, as well as the expansion of the static role of MCNP to analysis of fuel cycle depletion calculations. Results to date of these applications are presented also. A summary of the calculations not yet performed is also given to provide a {open_quotes}to-do{close_quotes} list if the project is resurrected.

  9. RELAP5 model for advanced neutron source reactor thermal-hydraulic transients, three-element-core design

    SciTech Connect

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L.

    1996-02-01

    In order to utilize reduced enrichment fuel, the three-element-core design has been proposed. The proposed core configuration consists of inner, middle, and outer elements, with the middle element offset axially beneath the inner and outer elements, which are axially aligned. The three-element-core RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. However, the total flow rate through the core is greater and the pressure drop across the core is less so that the primary coolant pumps and heat exchangers are operating at a different point in their performance curves. This report describes the new RELAP5 input for the core components.

  10. ATW neutronics design studies.

    SciTech Connect

    Wade, D. C.; Yang, W. S.; Khalil, H.

    2000-11-10

    The Accelerator Transmutation of Waste (ATW) concept has been proposed as a transuranics (TRU) (and long-lived fission product) incinerator for processing the 87,000 metric tonnes of Light Water Reactor used fuel which will have been generated by the time the currently deployed fleet of commercial reactors in the US reach the end of their licensed lifetime. The ATW is proposed to separate the uranium from the transuranics and fission products in the LWR used fuel, to fission the transuranics, to send the LWR and ATW generated fission products to the geologic repository and to send the uranium to either a low level waste disposal site or to save it for future use. The heat liberated in fissioning the transuranics would be converted to electricity and sold to partially offset the cost of ATW construction and operations. Options for incineration of long-lived fission products are under evaluation. A six-year science-based program of ATW trade and system studies was initiated in the US FY 2000 to achieve two main purposes: (1) ''to evaluate ATW within the framework of nonproliferation, waste management, and economic considerations,'' and (2) ''to evaluate the efficacy of the numerous technical options for ATW system configuration.'' This paper summarizes the results from neutronics and thermal/hydraulics trade studies which were completed at Argonne National Laboratory during the first year of the program. Core designs were developed for Pb-Bi cooled and Na cooled 840 MW{sub th} fast spectrum transmuter designs employing recycle. Additionally, neutronics analyses were performed at Argonne for a He cooled 600 MW{sub th} hybrid thermal and fast core design proposed by General Atomics Co. which runs critical for 3/4 and subcritical for 1/4 of its four year once-thin burn cycle. The mass flows and the ultimate loss of transuranic isotopes to the waste stream per unit of heat generated during transmutation have been calculated on a consistent basis and are compared. (Long-lived fission product incineration has not been considered in the studies reported here.)

  11. Neutron Monitor Design Improvements

    Microsoft Academic Search

    Pieter H. Stoker; Lev I. Dorman; John M. Clem

    2000-01-01

    The original design by J. A. Simpson of the neutron monitor enabled continuous monitoring of the primary cosmic-ray flux by ground-based recordings of the nucleonic component with only a rather simple correction for atmospheric effects. Simpson (1957) extended the original pile to the 12 counter IGY neutron monitor which was deployed in a world wide network during the International Geophysical

  12. Neutron Monitor Design Improvements

    Microsoft Academic Search

    Pieter H. Stoker; Lev I. Dorman; John M. Clem

    2000-01-01

    The original design by J. A. Simpson of the neutron monitor enabled continuous monitoring of the primary cosmic-ray flux by\\u000a ground-based recordings of the nucleonic component with only a rather simple correction for atmospheric effects. Simpson (1957)\\u000a extended the original pile to the 12 counter IGY neutron monitor which was deployed in a world wide network during the International\\u000a Geophysical Year

  13. DANDE: a linked code system for core neutronics\\/depletion analysis

    Microsoft Academic Search

    R. J. LaBauve; T. R. England; D. C. George; R. E. MacFarlane; W. B. Wilson

    1985-01-01

    This report describes DANDE - a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of

  14. DANDE: a linked code system for core neutronics\\/depletion analysis

    Microsoft Academic Search

    R. J. LaBauve; T. R. England; D. C. George; R. E. MacFarlane; W. B. Wilson

    1986-01-01

    This report describes DANDE - a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of

  15. Shear viscosity in neutron star cores

    E-print Network

    P. S. Shternin; D. G. Yakovlev

    2008-08-21

    We calculate the shear viscosity $\\eta = \\eta_{e\\mu}+\\eta_{n}$ in a neutron star core composed of nucleons, electrons and muons ($\\eta_{e\\mu}$ being the electron-muon viscosity, mediated by collisions of electrons and muons with charged particles, and $\\eta_{n}$ the neutron viscosity, mediated by neutron-neutron and neutron-proton collisions). Deriving $\\eta_{e\\mu}$, we take into account the Landau damping in collisions of electrons and muons with charged particles via the exchange of transverse plasmons. It lowers $\\eta_{e\\mu}$ and leads to the non-standard temperature behavior $\\eta_{e\\mu}\\propto T^{-5/3}$. The viscosity $\\eta_{n}$ is calculated taking into account that in-medium effects modify nucleon effective masses in dense matter. Both viscosities, $\\eta_{e\\mu}$ and $\\eta_{n}$, can be important, and both are calculated including the effects of proton superfluidity. They are presented in the form valid for any equation of state of nucleon dense matter. We analyze the density and temperature dependence of $\\eta$ for different equations of state in neutron star cores, and compare $\\eta$ with the bulk viscosity in the core and with the shear viscosity in the crust.

  16. Diagnostics of core barrel vibrations by in-core and ex-core neutron noise

    Microsoft Academic Search

    V. Arzhanov; I. Pázsit

    2003-01-01

    Diagnostics of core-barrel vibrations has traditionally been made by use of ex-vessel neutron detector signals. We suggest that in addition to the ex-core noise, also the in-core noise, induced by core barrel vibrations, be also used. This would enhance the possibilities of diagnostics where the number of the ex-core detectors is not sufficient or their positions are disadvantageous for effective

  17. Goldstone modes in the neutron star core

    E-print Network

    Paulo F. Bedaque; Sanjay Reddy

    2013-07-31

    We formulate a theory of Goldstone bosons and their interactions in the superfluid and superconducting phase of dense nucleonic matter at densities of relevance to the neutron star core. For typical neutron star temperatures in the range T = 10^6 to 10^9 K, the Goldstone mode associated with rotational symmetry, called angulons, couple weakly to each other and to electrons. Consequently, these modes have anomalously large mean free paths and can contribute to both diffusive and ballistic transport of heat and momentum. In contrast, the two Goldstone bosons associated with density oscillations of the neutron and electron + proton fluids, called superfluid phonons, mix and couple strongly to electrons. They have shorter mean free paths, and their contribution to transport is negligible. Long-wavelength superfluid phonons and angulons can play a role in neutron star seismology, and lead to interesting phenomenology as angulons couple to magnetic fields and have anisotropic dispersion relations.

  18. Thermal mass limit of neutron cores

    NASA Astrophysics Data System (ADS)

    Roupas, Zacharias

    2015-01-01

    Static thermal equilibrium of a quantum self-gravitating ideal gas in general relativity is studied at any temperature, taking into account the Tolman-Ehrenfest effect. Thermal contribution to the gravitational stability of static neutron cores is quantified. The curve of maximum mass with respect to temperature is reported. At low temperatures the Oppenheimer-Volkoff calculation is recovered, while at high temperatures the recently reported classical gas calculation is recovered. An ultimate upper mass limit M =2.43 M? of all maximum values is found to occur at Tolman temperature T =1.27 mc2 with radius R =15.2 km .

  19. HFIR cold neutron source moderator vessel design analysis

    SciTech Connect

    Chang, S.J.

    1998-04-01

    A cold neutron source capsule made of aluminum alloy is to be installed and located at the tip of one of the neutron beam tubes of the High Flux Isotope Reactor. Cold hydrogen liquid of temperature approximately 20 degree Kelvin and 15 bars pressure is designed to flow through the aluminum capsule that serves to chill and to moderate the incoming neutrons produced from the reactor core. The cold and low energy neutrons thus produced will be used as cold neutron sources for the diffraction experiments. The structural design calculation for the aluminum capsule is reported in this paper.

  20. Air core pulse transformer design

    Microsoft Academic Search

    J. P. O'Loughlin; J. D. Sidler; Gerry J. Rohwein

    1988-01-01

    Cylindrical-air-core pulse transformers capable of passing high-voltage\\/high-energy pulse waveforms with high efficiency and low distortion require a much more delicate design balance of physical dimensions and electrical parameters than iron or ferrite core units. Special computer codes were written to evaluate their performance. The analysis includes calculation of the self and mutual inductances as determined by the dimensions and insulation

  1. A NOVEL MICROMEGAS DETECTOR FOR IN-CORE NUCLEAR REACTOR NEUTRON FLUX MEASUREMENTS

    E-print Network

    Paris-Sud XI, Université de

    1 A NOVEL MICROMEGAS DETECTOR FOR IN-CORE NUCLEAR REACTOR NEUTRON FLUX MEASUREMENTS S. ANDRIAMONJE Talence Cedex, France Future fast nuclear reactors designed for energy production and transmutation to neutron detection inside nuclear reactor is given. The advantage of this detector over conventional

  2. Core Asymmetry Evaluation using Static Measurements and Neutron Noise Analysis

    Microsoft Academic Search

    F. Adorján; T. Czibók; S. Kiss; K. Krinizs; J. Végh

    2000-01-01

    A case study is presented describing an analysis initiated by a real core event at Paks NPP Hungary. The analysis utilised the static data provided by the VERONA core monitoring system as well as the results from a specially organised series of measurements using the neutron noise signals from in-core Rh SPN detectors. It was demonstrated that the observed temperature

  3. Air core pulse transformer design

    SciTech Connect

    O'Loughlin, J.P.; Sidler, J.D.; Rohwein, G.J.

    1988-01-01

    Cylindrical air core pulse transformers capable of passing high voltage/energy pulse waveforms with high efficiency and low distortion require a much more delicate design balance of physical dimensions and electrical parameters that iron or ferrite core units. The two salient advantages of the air core transformer are a much lighter weight and a simplified high voltage insulation system. Special computer codes were written to evaluate the performance. The analysis includes calculation of the self and mutual inductances as determined by the dimensions and insulation stress, evaluation of the waveforms distortion and energy transfer efficiency. Graphical data are given for the optimization in terms of electrical parameters. The results are in agreement with experimental data. It is concluded that air core transformers are feasible operating at hundreds of kilovolts and lens of kioljoules in the microsecond region with energy transfer efficiencies of 70% to 85%. The insulation stresses required are in the 100 to 300 kV/em range. Effects of the high frequency current distribution in the windings and the use of ''slug'' type ferrite cores are also evaluated.

  4. Air core pulse transformer design

    NASA Astrophysics Data System (ADS)

    Oloughlin, J. P.; Sidler, J. D.; Rohwein, G. J.

    Cylindrical air core pulse transformers capable of passing high voltage/energy pulse waveforms with high efficiency and low distortion require a much more delicate design balance of physical dimensions and electrical parameters that iron or ferrite core units. The two salient advantages of the air core transformer are a much lighter weight and a simplified high voltage insulation system. Special computer codes were written to evaluate the performance. The analysis includes calculation of the self and mutual inductances as determined by the dimensions and insulation stress, evaluation of the waveforms distortion and energy transfer efficiency. Graphical data are given for the optimization in terms of electrical parameters. The results are in agreement with experimental data. It is concluded that air core transformers are feasible operating at hundreds of kilovolts and tens of kioljoules in the microsecond region with energy transfer efficiencies of 70 to 85 percent. The insulation stresses required are in the 100 to 300 kV/em range. Effects of the high frequency current distribution in the windings and the use of slug type ferrite cores are also evaluated.

  5. Persistent crust-core spin lag in neutron stars

    NASA Astrophysics Data System (ADS)

    Glampedakis, Kostas; Lasky, Paul D.

    2015-06-01

    It is commonly believed that the magnetic field threading a neutron star provides the ultimate mechanism (on top of fluid viscosity) for enforcing long-term corotation between the slowly spun-down solid crust and the liquid core. We show that this argument fails for axisymmetric magnetic fields with closed field lines in the core, the commonly used `twisted torus' field being the most prominent example. The failure of such magnetic fields to enforce global crust-core corotation leads to the development of a persistent spin lag between the core region occupied by the closed field lines and the rest of the crust and core. We discuss the repercussions of this spin lag for the evolution of the magnetic field, suggesting that, in order for a neutron star to settle to a stable state of crust-core corotation, the bulk of the toroidal field component should be deposited into the crust soon after the neutron star's birth.

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

    PubMed

    Lashkari, A; Khalafi, H; Kazeminejad, H

    2013-05-01

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

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

    PubMed Central

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

    2013-01-01

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

  8. Preliminary engineering design of sodium-cooled CANDLE core

    NASA Astrophysics Data System (ADS)

    Takaki, Naoyuki; Namekawa, Azuma; Yoda, Tomoyuki; Mizutani, Akihiko; Sekimoto, Hiroshi

    2012-06-01

    The CANDLE burning process is characterized by the autonomous shifting of burning region with constant reactivity and constant spacial power distribution. Evaluations of such critical burning process by using widely used neutron diffusion and burning codes under some realistic engineering constraints are valuable to confirm the technical feasibility of the CANDLE concept and to put the idea into concrete core design. In the first part of this paper, it is discussed that whether the sustainable and stable CANDLE burning process can be reproduced even by using conventional core analysis tools such as SLAROM and CITATION-FBR. As a result, it is certainly possible to demonstrate it if the proper core configuration and initial fuel composition required as CANDLE core are applied to the analysis. In the latter part, an example of a concrete image of sodium cooled, metal fuel, 2000MWt rating CANDLE core has been presented by assuming an emerging inevitable technology of recladding. The core satisfies engineering design criteria including cladding temperature, pressure drop, linear heat rate, and cumulative damage fraction (CDF) of cladding, fast neutron fluence and sodium void reactivity which are defined in the Japanese FBR design project. It can be concluded that it is feasible to design CADLE core by using conventional codes while satisfying some realistic engineering design constraints assuming that recladding at certain time interval is technically feasible.

  9. Optimization of IEC grid design for maximum neutron production

    SciTech Connect

    Miley, G.H.; DeMora, J.; Stubbers, R.; Tzonev, I.V. [Univ. of Illinois, Urbana, IL (United States); Anderl, R.A. [Lockheed Martin Idaho Technologies Company, Idaho Falls, ID (United States); Nadler, J.H. [Department of Energy, Idaho Falls, ID (United States); Nebel, R. [Los Alamos National Lab., NM (United States)

    1996-12-31

    Two different, complementary approaches were taken to determine the effects of an Inertial Electrostatic Confinement (IEC) grid`s design on the neutron production rate of the device. A semi-empirical formula developed from experimental data predicts the neutron yield of an IEC device, given the chamber size, grid radius and transparency, and operating voltage and current. Results from the IXL computer program support some of the scalings found in the semi-empirical formula. A second formula was also developed that predicts the neutron yield of an IEC device using grid design parameters and the ion core radius. The SIMION computer program was used to calculate the ion core radius. These formulas are useful tools for designing grids that will maximize the neutron yield for IEC devices. 7 refs., 9 figs.

  10. Design configuration of GCFR core assemblies

    SciTech Connect

    LaBar, M.P.; Lee, G.E.; Meyer, R.J.

    1980-05-01

    The current design configurations of the core assemblies for the gas-cooled fast reactor (GCFR) demonstration plant reactor core conceptual design are described. Primary emphasis is placed upon the design innovations that have been incorporated in the design of the core assemblies since the establishment of the initial design of an upflow GCFR core. A major feature of the design configurations is that they are prototypical of core assemblies for use in commercial plants; a larger number of the same assemblies would be used in a commercial plant.

  11. GCFR core thermal-hydralic design

    SciTech Connect

    Schleuter, G.; Baxi, C.B.; Bennett, F.O.

    1980-05-01

    The approach for developing the thermal-hydraulic core assembly designs for the gas-cooled fast reactor (GCFR) is reviewed, and key considerations for improving the core performance at all power and flow conditions are discussed. It is shown how the thermal-hydraulic core assembly designs evolve from evaluations of plant size, material limitations, safety criteria, and structural performance considerations.

  12. Nodal weighting factor method for ex-core fast neutron fluence evaluation

    SciTech Connect

    Chiang, R. T. [AREVA NP Inc., 6399 San Ignacio Ave., San Jose, CA 95119 (United States)

    2012-07-01

    The nodal weighting factor method is developed for evaluating ex-core fast neutron flux in a nuclear reactor by utilizing adjoint neutron flux, a fictitious unit detector cross section for neutron energy above 1 or 0.1 MeV, the unit fission source, and relative assembly nodal powers. The method determines each nodal weighting factor for ex-core neutron fast flux evaluation by solving the steady-state adjoint neutron transport equation with a fictitious unit detector cross section for neutron energy above 1 or 0.1 MeV as the adjoint source, by integrating the unit fission source with a typical fission spectrum to the solved adjoint flux over all energies, all angles and given nodal volume, and by dividing it with the sum of all nodal weighting factors, which is a normalization factor. Then, the fast neutron flux can be obtained by summing the various relative nodal powers times the corresponding nodal weighting factors of the adjacent significantly contributed peripheral assembly nodes and times a proper fast neutron attenuation coefficient over an operating period. A generic set of nodal weighting factors can be used to evaluate neutron fluence at the same location for similar core design and fuel cycles, but the set of nodal weighting factors needs to be re-calibrated for a transition-fuel-cycle. This newly developed nodal weighting factor method should be a useful and simplified tool for evaluating fast neutron fluence at selected locations of interest in ex-core components of contemporary nuclear power reactors. (authors)

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

    Microsoft Academic Search

    K. C. Schulz; G. T. Yahr

    1995-01-01

    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

  14. Crust-core coupling in rotating neutron stars

    SciTech Connect

    Glampedakis, Kostas; Andersson, Nils [School of Mathematics, University of Southampton, Southampton SO17 1BJ (United Kingdom)

    2006-08-15

    Motivated by their gravitational wave driven instability, we investigate the influence of the crust on r-mode oscillations in a neutron star. Using a simplistic model of an elastic neutron star crust with constant shear modulus, we carry out an analytic calculation with the main objective of deriving an expression for the slippage between the core and the crust. Our analytic estimates support previous numerical results and provide useful insights into the details of the problem.

  15. Core excitations across the neutron shell gap in 207Tl

    NASA Astrophysics Data System (ADS)

    Wilson, E.; Podolyák, Zs.; Grawe, H.; Brown, B. A.; Chiara, C. J.; Zhu, S.; Fornal, B.; Janssens, R. V. F.; Shand, C. M.; Bowry, M.; Bunce, M.; Carpenter, M. P.; Cieplicka-Ory?czak, N.; Deo, A. Y.; Dracoulis, G. D.; Hoffman, C. R.; Kempley, R. S.; Kondev, F. G.; Lane, G. J.; Lauritsen, T.; Lotay, G.; Reed, M. W.; Regan, P. H.; Rodríguez Triguero, C.; Seweryniak, D.; Szpak, B.; Walker, P. M.

    2015-07-01

    The single closed-neutron-shell, one proton-hole nucleus 207Tl was populated in deep-inelastic collisions of a 208Pb beam with a 208Pb target. The yrast and near-yrast level scheme has been established up to high excitation energy, comprising an octupole phonon state and a large number of core excited states. Based on shell-model calculations, all observed single core excitations were established to arise from the breaking of the N = 126 neutron core. While the shell-model calculations correctly predict the ordering of these states, their energies are compressed at high spins. It is concluded that this compression is an intrinsic feature of shell-model calculations using two-body matrix elements developed for the description of two-body states, and that multiple core excitations need to be considered in order to accurately calculate the energy spacings of the predominantly three-quasiparticle states.

  16. Experiment Design and Analysis Guide - Neutronics & Physics

    SciTech Connect

    Misti A Lillo

    2014-06-01

    The purpose of this guide is to provide a consistent, standardized approach to performing neutronics/physics analysis for experiments inserted into the Advanced Test Reactor (ATR). This document provides neutronics/physics analysis guidance to support experiment design and analysis needs for experiments irradiated in the ATR. This guide addresses neutronics/physics analysis in support of experiment design, experiment safety, and experiment program objectives and goals. The intent of this guide is to provide a standardized approach for performing typical neutronics/physics analyses. Deviation from this guide is allowed provided that neutronics/physics analysis details are properly documented in an analysis report.

  17. Core design investigation for a SUPERSTAR small modular lead-cooled fast reactor demonstrator

    Microsoft Academic Search

    S. Bortot; A. Moisseytsev; J. J. Sienicki; Carlo Artioli

    In this paper a preconceptual neutronics design study for a SUstainable Proliferation-resistance Enhanced Refined Secure Transportable Autonomous Reactor (SUPERSTAR) demonstrator is presented. The main goal of achieving the highest realistic power level limited by natural circulation and transportability, while providing energy security and proliferation resistance thanks to a long core lifetime design has been satisfactorily attained. A preliminary core configuration

  18. Entrainment parameters in a cold superfluid neutron star core

    SciTech Connect

    Chamel, Nicolas; Haensel, Pawel [Copernicus Astronomical Center, Polish Academy of Science, ul. Bartycka 18, PL-00-716 Warsaw (Poland); LUTH, Paris Observatory, 5 place Jules Janssen, F-92195 Meudon (France); Copernicus Astronomical Center, Polish Academy of Science, ul. Bartycka 18, PL-00-716 Warsaw (Poland)

    2006-04-15

    Hydrodynamic simulations of neutron star cores that are based on a two-fluid description in terms of a neutron-proton superfluid mixture require the knowledge of the Andreev-Bashkin entrainment matrix which relates the momentum of one constituent to the currents of both constituents. This matrix is derived for arbitrary nuclear asymmetry at zero temperature and in the limits of small relative currents in the framework of the energy density functional theory. The Skyrme energy density functional is considered as a particular case. General analytic formulas for the entrainment parameters and various corresponding effective masses are obtained. These formulas are applied to the liquid core of a neutron star composed of homogeneous plasma of nucleons, electrons, and possibly muons in {beta} equilibrium.

  19. Entrainment parameters in cold superfluid neutron star core

    E-print Network

    Nicolas Chamel; Pawel Haensel

    2006-09-13

    Hydrodynamical simulations of neutron star cores, based on a two fluid description in terms of a neutron-proton superfluid mixture, require the knowledge of the Andreev-Bashkin entrainment matrix which relates the momentum of one constituent to the currents of both constituents. This matrix is derived for arbitrary nuclear asymmetry at zero temperature and in the limits of small relative currents in the framework of the energy density functional theory. The Skyrme energy density functional is considered as a particular case. General analytic formulae for the entrainment parameters and various corresponding effective masses are obtained. These formulae are applied to the liquid core of a neutron star, composed of an homogeneous plasma of nucleons, electrons and possibly muons in beta equilibrium.

  20. AHTR Mechanical, Structural, and Neutronic Preconceptual Design

    SciTech Connect

    Varma, V.K.; Holcomb, D.E.; Peretz, F.J.; Bradley, E.C.; Ilas, D.; Qualls, A.L.; Zaharia, N.M.

    2012-09-15

    This report provides an overview of the mechanical, structural, and neutronic aspects of the Advanced High Temperature Reactor (AHTR) design concept. The AHTR is a design concept for a large output Fluoride salt cooled High-temperature Reactor (FHR) that is being developed to enable evaluation of the technology hurdles remaining to be overcome prior to FHRs becoming an option for commercial reactor deployment. This report documents the incremental AHTR design maturation performed over the past year and is focused on advancing the design concept to a level of a functional, self-consistent system. The reactor concept development remains at a preconceptual level of maturity. While the overall appearance of an AHTR design is anticipated to be similar to the current concept, optimized dimensions will differ from those presented here. The AHTR employs plate type coated particle fuel assemblies with rapid, off-line refueling. Neutronic analysis of the core has confirmed the viability of a 6-month two-batch cycle with 9 wt. % enriched uranium fuel. Refueling is intended to be performed automatically under visual guidance using dedicated robotic manipulators. The report includes a preconceptual design of the manipulators, the fuel transfer system, and the used fuel storage system. The present design intent is for used fuel to be stored inside of containment for at least six months and then transferred to local dry wells for intermediate term, on-site storage. The mechanical and structural concept development effort has included an emphasis on transportation and constructability to minimize construction costs and schedule. The design intent is that all components be factory fabricated into rail transportable modules that are assembled into subsystems at an on-site workshop prior to being lifted into position using a heavy-lift crane in an open-top style construction. While detailed accident identification and response sequence analysis has yet to be performed, the design concept incorporates fully passive responses to all identified design basis or non-very-low frequency beyond design basis accidents as well as multiple levels of radioactive material containment. Key building design elements include (1) below grade siting to minimize vulnerability to aircraft impact, (2) multiple natural circulation decay heat rejection chimneys, (3) seismic base isolation, and (4) decay heat powered back-up electricity generation.

  1. Shield Design for a Space Based Vapor Core Reactor

    SciTech Connect

    Knight, Travis; Anghaie, Samim [Innovative Nuclear Space Power and Propulsion Institute (INSPI), PO Box 116502, University of Florida, Gainesville, FL 32611-6502 (United States)

    2002-07-01

    Innovative shielding strategies were sought to reduce the mass of the required shielding for a space based vapor core reactor system with magnetohydrodynamic energy conversion. Gamma-rays directly resultant from fission were found to play no role in the dose rate, while secondary gamma-rays from fission neutron interactions were the dominant contributor to the dose rate. Hydrogen containing materials such as polyethylene were utilized to provide shielding of both radiation from the reactor complex and also solar and galactic cosmic radiation. This shield design was found to contribute 0.125 kg/kWe to the baseline vapor core reactor system specific mass. (authors)

  2. Automated Design and Optimization of Pebble-bed Reactor Cores

    SciTech Connect

    Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry

    2010-07-01

    We present a conceptual design approach for high-temperature gas-cooled reactors using recirculating pebble-bed cores. The design approach employs PEBBED, a reactor physics code specifically designed to solve for and analyze the asymptotic burnup state of pebble-bed reactors, in conjunction with a genetic algorithm to obtain a core that maximizes a fitness value that is a function of user-specified parameters. The uniqueness of the asymptotic core state and the small number of independent parameters that define it suggest that core geometry and fuel cycle can be efficiently optimized toward a specified objective. PEBBED exploits a novel representation of the distribution of pebbles that enables efficient coupling of the burnup and neutron diffusion solvers. With this method, even complex pebble recirculation schemes can be expressed in terms of a few parameters that are amenable to modern optimization techniques. With PEBBED, the user chooses the type and range of core physics parameters that represent the design space. A set of traits, each with acceptable and preferred values expressed by a simple fitness function, is used to evaluate the candidate reactor cores. The stochastic search algorithm automatically drives the generation of core parameters toward the optimal core as defined by the user. The optimized design can then be modeled and analyzed in greater detail using higher resolution and more computationally demanding tools to confirm the desired characteristics. For this study, the design of pebble-bed high temperature reactor concepts subjected to demanding physical constraints demonstrated the efficacy of the PEBBED algorithm.

  3. Radiation of Neutron Stars Produced by Superfluid Core

    NASA Astrophysics Data System (ADS)

    Svidzinsky, Anatoly A.

    2003-06-01

    We find a new mechanism of neutron star radiation wherein radiation is produced by the stellar interior. The main finding is that the neutron star interior is transparent for collisionless electron sound, the same way as it is transparent for neutrinos. In the presence of the magnetic field the electron sound is coupled with electromagnetic radiation; such collective excitation is known as a fast magnetosonic wave. At high densities such waves reduce to the zero sound in electron liquid, while near the stellar surface they are similar to electromagnetic waves in a medium. We find that zero sound is generated by superfluid vortices in the stellar core. Thermally excited helical vortex waves produce fast magnetosonic waves in the stellar crust that propagate toward the surface and transform into outgoing electromagnetic radiation. The magnetosonic waves are partially absorbed in a thin layer below the surface. The absorption is highly anisotropic; it is smaller for waves that in the absorbing layer propagate closer to the magnetic field direction. As a result, the vortex radiation is pulsed with the period of star rotation. The vortex radiation has the spectral index ?~-0.45 and can explain nonthermal radiation of middle-aged pulsars observed in the infrared, optical, and hard X-ray bands. The radiation is produced in the star interior, rather than in the magnetosphere, which allows direct determination of the core temperature. Comparing the theory with available spectra observations, we find that the core temperature of the Vela pulsar is T~8×108 K, while the core temperature of PSR B0656+14 and Geminga exceeds 2×108 K. This is the first measurement of the temperature of a neutron star core. The temperature estimate rules out equations of state incorporating Bose condensations of pions or kaons and quark matter in these objects. The estimate also allows us to determine the critical temperature of triplet neutron superfluidity in the Vela core, Tc=(7.5+/-1.5)×109 K, which agrees well with the value of critical temperature in a core of a canonical neutron star calculated based on recent data for behavior of strong interactions at high energies. We also find that in the middle-aged neutron stars the vortex radiation, rather than thermal conductivity, is the main mechanism of heat transfer from the stellar core to the surface. The core radiation opens a possibility to study composition of neutron star crust by detection of absorption lines corresponding to the low-energy excitations of crust nuclei. Bottom layers of the crust may contain exotic nuclei with the mass number up to 600, and the core radiation creates a perspective to study their properties. In principle, zero sound can also be emitted by other mechanisms, rather than vortices. In this case the spectrum of stellar radiation would contain features corresponding to such processes. As a result, zero sound opens a perspective of direct spectroscopic study of superdense matter in the neutron star interior.

  4. DANDE: a linked code system for core neutronics/depletion analysis

    SciTech Connect

    LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.

    1986-01-01

    This report describes DANDE - a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the cource of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of reactor fuel under increased burn conditions. The operation of the code system is illustrated in this report by two sample problems. 25 refs.

  5. NASA'S Chandra Finds Superfluid in Neutron Star's Core

    NASA Astrophysics Data System (ADS)

    2011-02-01

    NASA's Chandra X-ray Observatory has discovered the first direct evidence for a superfluid, a bizarre, friction-free state of matter, at the core of a neutron star. Superfluids created in laboratories on Earth exhibit remarkable properties, such as the ability to climb upward and escape airtight containers. The finding has important implications for understanding nuclear interactions in matter at the highest known densities. Neutron stars contain the densest known matter that is directly observable. One teaspoon of neutron star material weighs six billion tons. The pressure in the star's core is so high that most of the charged particles, electrons and protons, merge resulting in a star composed mostly of uncharged particles called neutrons. Two independent research teams studied the supernova remnant Cassiopeia A, or Cas A for short, the remains of a massive star 11,000 light years away that would have appeared to explode about 330 years ago as observed from Earth. Chandra data found a rapid decline in the temperature of the ultra-dense neutron star that remained after the supernova, showing that it had cooled by about four percent over a 10-year period. "This drop in temperature, although it sounds small, was really dramatic and surprising to see," said Dany Page of the National Autonomous University in Mexico, leader of a team with a paper published in the February 25, 2011 issue of the journal Physical Review Letters. "This means that something unusual is happening within this neutron star." Superfluids containing charged particles are also superconductors, meaning they act as perfect electrical conductors and never lose energy. The new results strongly suggest that the remaining protons in the star's core are in a superfluid state and, because they carry a charge, also form a superconductor. "The rapid cooling in Cas A's neutron star, seen with Chandra, is the first direct evidence that the cores of these neutron stars are, in fact, made of superfluid and superconducting material," said Peter Shternin of the Ioffe Institute in St Petersburg, Russia, leader of a team with a paper accepted in the journal Monthly Notices of the Royal Astronomical Society. Both teams show that this rapid cooling is explained by the formation of a neutron superfluid in the core of the neutron star within about the last 100 years as seen from Earth. The rapid cooling is expected to continue for a few decades and then it should slow down. "It turns out that Cas A may be a gift from the Universe because we would have to catch a very young neutron star at just the right point in time," said Page's co-author Madappa Prakash, from Ohio University. "Sometimes a little good fortune can go a long way in science." The onset of superfluidity in materials on Earth occurs at extremely low temperatures near absolute zero, but in neutron stars, it can occur at temperatures near a billion degrees Celsius. Until now there was a very large uncertainty in estimates of this critical temperature. This new research constrains the critical temperature to between one half a billion to just under a billion degrees. Cas A will allow researchers to test models of how the strong nuclear force, which binds subatomic particles, behaves in ultradense matter. These results are also important for understanding a range of behavior in neutron stars, including "glitches," neutron star precession and pulsation, magnetar outbursts and the evolution of neutron star magnetic fields. Small sudden changes in the spin rate of rotating neutron stars, called glitches, have previously given evidence for superfluid neutrons in the crust of a neutron star, where densities are much lower than seen in the core of the star. This latest news from Cas A unveils new information about the ultra-dense inner region of the neutron star. "Previously we had no idea how extended superconductivity of protons was in a neutron star," said Shternin's co-author Dmitry Yakovlev, also from the Ioffe Institute. The cooling in the Cas A

  6. NEUTRON RADIOGRAPHY (NRAD) REACTOR 64-ELEMENT CORE UPGRADE

    SciTech Connect

    John D. Bess

    2014-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA (registered) (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The interim critical configuration developed during the core upgrade, which contains only 62 fuel elements, has been evaluated as an acceptable benchmark experiment. The final 64-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (approximately +/-1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  7. Neutronic design of the ITER radial neutron camera

    Microsoft Academic Search

    L. Petrizzi; R. Barnsley; L. Bertalot; B. Esposito; H. Haskell; E. Mainardi; D. Marocco; S. Podda; C. Walker; S. Villari

    2007-01-01

    This paper summarizes the work, performed in the frame of various EFDA contracts during 2004–2005, on the design review and upgrade of the ITER radial neutron camera (RNC). The RNC, which should provide information on the spatial distribution and energy spectrum of the neutron emission, consists of an ex-vessel system (fan-like collimator with 12×3 lines of sights) and an in-vessel

  8. Neutron tube design study for boron neutron capture therapy application

    SciTech Connect

    Verbeke, J.M.; Lee, Y.; Leung, K.N.; Vujic, J.; Williams, M.D.; Wu, L.K.; Zahir, N.

    1999-05-06

    Radio-frequency (RF) driven ion sources are being developed in Lawrence Berkeley National Laboratory (LBNL) for sealed-accelerator-tube neutron generator application. By using a 5-cm-diameter RF-driven multicusp source H{sup +} yields over 95% have been achieved. These experimental findings will enable one to develop compact neutron generators based on the D-D or D-T fusion reactions. In this new neutron generator, the ion source, the accelerator and the target are all housed in a sealed metal container without external pumping. Recent moderator design simulation studies have shown that 14 MeV neutrons could be moderated to therapeutically useful energy ranges for boron neutron capture therapy (BNCT). The dose near the center of the brain with optimized moderators is about 65% higher than the dose obtained from a typical neutron spectrum produced by the Brookhaven Medical Research Reactor (BMRR), and is comparable to the dose obtained by other accelerator-based neutron sources. With a 120 keV and 1 A deuteron beam, a treatment time of {approx}35 minutes is estimated for BNCT.

  9. Neutron capture therapy beam design at Harwell.

    PubMed

    Constantine, G

    1990-01-01

    At Harwell, we have progressed from designing, building, and using small-diameter beams of epithermal neutrons for radiobiology studies to designing a radiotherapy facility for the 25-MW research reactor DIDO. The program is well into the survey phase, where the main emphasis is on tailoring the neutron spectrum. The incorporation of titanium and vanadium in an aluminium spectrum shaper in the D2O reflector has been shown to yield a significant reduction in the mean energy of neutrons incident on the patient by suppression of streaming through the cross-section window in aluminium at 25 keV. PMID:2176459

  10. R-modes of neutron stars with the superfluid core

    E-print Network

    Umin Lee; Shijun Yoshida

    2002-11-26

    We investigate the modal properties of the $r$-modes of rotating neutron stars with the core filled with neutron and proton superfluids, taking account of entrainment effects between the superfluids. The stability of the $r$-modes against gravitational radiation reaction is also examined considering viscous dissipation due to shear and a damping mechanism called mutual friction between the superfluids in the core. We find the $r$-modes in the superfluid core are split into ordinary $r$-modes and superfluid $r$-modes, which we call, respectively, $r^o$- and $r^s$-modes. The two superfluids in the core flow together for the $r^o$-modes, while they counter-move for the $r^s$-modes. For the $r^o$-modes, the coefficient $\\kappa_0\\equiv\\lim_{\\Omega\\to 0}\\omega/\\Omega$ is equal to $2m/[l^\\prime(l^\\prime+1)]$, almost independent of the parameter $\\eta$ that parameterizes the entrainment effects between the superfluids, where $\\Omega$ is the angular frequency of rotation, $\\omega$ the oscillation frequency observed in the corotating frame of the star, and $l^\\prime$ and $m$ are the indices of the spherical harmonic function representing the angular dependence of the $r$-modes. For the $r^s$-modes, on the other hand, $\\kappa_0$ is equal to $2m/[l^\\prime(l^\\prime+1)]$ at $\\eta=0$ (no entrainment), and it almost linearly increases as $\\eta$ is increased from $\\eta=0$. The mutual friction in the superfluid core is found ineffective to stabilize the $r$-mode instability caused by the $r^o$-mode except in a few narrow regions of $\\eta$. The $r$-mode instability caused by the $r^s$-modes, on the other hand, is extremely weak and easily damped by dissipative processes in the star.

  11. CHINA SPALLATION NEUTRON SOURCE DESIGN.

    SciTech Connect

    WEI,J.

    2007-01-29

    The China Spallation Neutron Source (CSNS) is an accelerator-based high-power project currently in preparation under the direction of the Chinese Academy of Sciences (CAS). The complex is based on an H- linear accelerator, a rapid cycling proton synchrotron accelerating the beam to 1.6 GeV, a solid tungsten target station, and five initial instruments for spallation neutron applications. The facility will operate at 25 Hz repetition rate with a phase-I beam power of about 120 kW. The major challenge is to build a robust and reliable user's facility with upgrade potential at a fractional of ''world standard'' cost.

  12. Design of a silicon carbide neutron sensor

    NASA Astrophysics Data System (ADS)

    Hu, Qing-Qing; Yang, Jun; Liu, GuoFu; Luo, Xiao-liang

    2013-01-01

    The radiation detectors based on the third era semiconductor material silicon carbide (SiC) with wide energy band gap are the most promising ionizing radiation detectors in high temperature and harsh radiation environment. This paper illustrated several important advantages of the SiC neutron detector and described its fabrication and detection principle briefly. Furthermore, we evaluated the neutron sensor's measurement performance when detecting the 2.5MeV and 14MeV neutrons under different conditions of sensor's active layer thickness and polyethylene converter film thickness based on MCNP simulation. According to the results of simulation, the sensor's optimal configuration was designed. For the sensor whose radius and depletion layer thickness are 3mm and 30?m respectively, the detection efficiency can reach 3.16×10-18 coulomb per neutron (c/n) and 1.80×10-17 c/n for 2.5MeV and 14MeV neutrons respectively. When adding a polyethylene converter film of 90?m thickness to the above sensor, the detection efficiency to 2.5MeV neutron will be 3.7 times that without neutron converter film; and with the converter film of 2mm thickness, the detection efficiency to 14MeV neutron will be improved by 246%.

  13. Designing avionics for terrestrial neutron environments

    Microsoft Academic Search

    Peter G. Coakley; Dennis Breuner; Randall Milanowski; Marion A. Rose; Amy L. Magnus

    2004-01-01

    This paper presents issues related to effects generated in avionic electronics by terrestrial neutron environments and methods for mitigating the effects through part selection, circuit design and system architecture design. The paper includes an explanation of the System Hardening Upset Recovery (SHUR) technology macro cell library and demonstrates how the available functions can be applied to implement robust system operation

  14. Design and simulation of a neutron facility.

    PubMed

    Studenski, Matthew T; Kearfott, Kimberlee J

    2007-02-01

    State and other regulatory entities require that for any facility housing a particle accelerator the surrounding areas must be restricted to public access unless the dose equivalent rate is less than 0.02 mSv h at 5 cm from any accessible wall surrounding the facility under conditions of maximum radiation output. A Monte Carlo radiation transport simulation code, MCNP5, was used to design a proposed facility to shield two D-T neutron generators and one D-D neutron generator. A number of different designs were simulated, but due to cost and space issues a small concrete cave proved to be the best solution for the shielding problem. With this design, all of the neutron generators could be used and all of the rooms surrounding the neutron facility could be considered unrestricted to public access. To prevent unauthorized access into the restricted area of the neutron facility, light curtains, warning lights, door interlocks, and rope barriers will be built into the facility. PMID:17228186

  15. Neutronic analysis of the Three Mile Island Unit 2 ex-core detector response

    Microsoft Academic Search

    D. J. Malloy; Y. I. Chang

    1981-01-01

    A neutronic analysis has been made with respect to the ex-core neutron detector response during the TMI-2 incident. A series of transport theory calculations quantified the impact upon the detector count rate of various core and downcomer conditions. In particular, various combinations of coolant void content and spatial distributions were investigated to yield the resulting transmission of the photoneutron source

  16. Measurement of the vortex core in sub-100 nm Fe dots using polarized neutron scattering

    E-print Network

    Roshchin, Igor V.

    OFFPRINT Measurement of the vortex core in sub-100 nm Fe dots using polarized neutron scattering neutron scattering Igor V. Roshchin1,2 , Chang-Peng Li2(a) , Harry Suhl2 , Xavier Batlle3 , S. Roy2(b diffraction and scattering Abstract ­ We use polarized neutron scattering to obtain quantitative information

  17. Advanced Neutron Sources: Plant Design Requirements

    SciTech Connect

    Not Available

    1990-07-01

    The Advanced Neutron Source (ANS) is a new, world class facility for research using hot, thermal, cold, and ultra-cold neutrons. At the heart of the facility is a 350-MW{sub th}, heavy water cooled and moderated reactor. The reactor is housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides fans out into a large guide hall, housing about 30 neutron research stations. Office, laboratory, and shop facilities are included to provide a complete users facility. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory at the end of the decade. This Plant Design Requirements document defines the plant-level requirements for the design, construction, and operation of the ANS. This document also defines and provides input to the individual System Design Description (SDD) documents. Together, this Plant Design Requirements document and the set of SDD documents will define and control the baseline configuration of the ANS.

  18. A new fuel loading design for the Advanced Neutron Source

    SciTech Connect

    Gehin, J.C.; Renier, J.P.; Worley, B.A.

    1994-06-01

    A new fuel loading design has been developed for the Advanced Neutron Source Reactor. In this reactor the combination of a small core volume and high power results in a very high power density. Using a direct optimization procedure the thermal-hydraulic margins for oxide temperature drop, centerline temperature and incipient boiling (and thus critical heat flux) were maximized to increase the limiting thermal power from 298 MW to 346 MW compared to the previous fuel grading, while maintaining the desired peak reflector thermal flux.

  19. Development of an advanced core analysis system for boiling water reactor designs

    Microsoft Academic Search

    Hiromi Maruyama; Junichi Koyama; Motoo Aoyama; Kazuya Ishii; Atsushi Zukeran; Takashi Kiguchi; Akira Nishimura

    1997-01-01

    A core analysis system has been developed for the recent advanced designs of boiling water reactors. This system consists of a fuel assembly analysis code VMONT and a three-dimensional core simulator COSNEX. To cope with heterogeneous structures found in the recent high-performance fuel, VMONT employs a Monte Carlo neutron transport calculation method. COSNEX is based on a three-group nodal expansion

  20. Advanced Neutron Source: Plant Design Requirements

    SciTech Connect

    Not Available

    1990-07-01

    The Advanced Neutron Source will be a new world-class facility for research using hot, thermal, cold, and ultra-cold neutrons. The heart of the facility will be a 330-MW (fission), heavy-water cooled and heavy-water moderated reactor. The reactor will be housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides will fan out into a large guide hall, housing about 30 neutron research stations. Appropriate office, laboratory, and shop facilities will be included to provide a complete facility for users. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory early in the next decade. This PDR document defines the plant-level requirements for the design, construction, and operation of ANS. It also defines and provides input to the individual System Design Description (SDD) documents. Together, this PDR document and the set of SDD documents will define and control the baseline configuration of ANS.

  1. Calculation of core loss in a novel transformer design

    Microsoft Academic Search

    W. L. Collett; H. R. Buswell

    2011-01-01

    Recent efforts to develop a new power transformer incorporating a novel wire core configuration have required that core magnetic flux and power loss be estimated during the development process. In this work, an innovative core concept was considered using commercial finite element simulation software, with core loss results shown to be comparable to measurements on a standard 10 kVA design.

  2. Energy Efficient Engine core design and performance report

    NASA Technical Reports Server (NTRS)

    Stearns, E. Marshall

    1982-01-01

    The Energy Efficient Engine (E3) is a NASA program to develop fuel saving technology for future large transport aircraft engines. Testing of the General Electric E3 core showed that the core component performance and core system performance necessary to meet the program goals can be achieved. The E3 core design and test results are described.

  3. Novel design of foam core junctions in sandwich panels

    Microsoft Academic Search

    E. Bozhevolnaya; A. Lyckegaard; O. T. Thomsen

    2008-01-01

    An improved design of core junctions in sandwich panels\\/beams is proposed. The improved\\/novel design consists of a specific geometrical shaping of the boundary of the adjoined core materials, which substantially diminishes the local stress concentrations at core junctions subjected to transverse shear loading. Two groups of the beams with conventional butt junctions and modified (spline shaped) junctions are manufactured and

  4. The long-term rotation dynamics of neutron stars with differentially rotating unmagnetized core

    NASA Astrophysics Data System (ADS)

    Barsukov, D. P.; Goglichidze, O. A.; Tsygan, A. I.

    2014-10-01

    We consider the pulsar long-term rotation dynamics taking into account the non-rigidity of neutron star rotation. We restrict our attention to the models with two essential assumptions: (1) crust-core interaction occurs via the viscosity (magnetic coupling is not important); (2) neutron star shape is symmetrical over the magnetic axis. The neutron star core is described by linearized quasi-stationary Newtonian hydrodynamical equations in one-fluid and two-fluid (neutron superfluidity) approximations. It is shown that in this case the pulsar inclination angle evolves to 0° or 90° very quickly. Since such fast evolution seems to contradict the observation data, either neutron stars are triaxial or the magnetic field plays the leading role in crust-core coupling.

  5. /sup 18/O as a core plus two valence neutrons: A three-body Faddeev calculation

    SciTech Connect

    Ueta, K.; Miyake, H.; Mizukami, A.

    1983-01-01

    The nucleus /sup 18/O is studied assuming a three-body model: two neutrons outside an inert core of /sup 16/O: and solving the Faddeev equations. The calculated spectrum is in good agreement with experiment.

  6. Bulk viscosity in superfluid neutron star cores. III. Effects of $?^-$ hyperons

    E-print Network

    P. Haensel; K. P. Levenfish; D. G. Yakovlev

    2001-11-15

    Bulk viscosity of neutron star cores containing hyperons is studied taking into account non-equilibrium weak process $n+n \\rightleftharpoons p+\\Sigma^-$. Rapid growth of the bulk viscosity within the neutron star core associated with switching on new reactions (modified Urca process, direct Urca process, hyperon reactions) is analyzed. The suppression of the bulk viscosity by superfluidity of baryons is considered and found out to be very important.

  7. How useful is neutron diffusion theory for nuclear rocket engine design

    Microsoft Academic Search

    T. A. Hilsmeier; S. M. Aithal; T. Aldemir

    1992-01-01

    Correct modeling of neutron leakage and geometry effects is important in the design of a nuclear rocket engine because of the need for small reactor cores in space applications. In principle, there are generalized procedures that can account for these effects in a reliable manner (e.g., a three-dimensional, continuous-energy Monte Carlo calculation with all core components explicitly modeled). However, these

  8. Development and preliminary verification of the 3D core neutronic code: COCO

    SciTech Connect

    Lu, H.; Mo, K.; Li, W.; Bai, N.; Li, J. [Reactor Design and Fuel Management Research Center, China Nuclear Power Technology Research Inst., 47F/A Jiangsu Bldg., Yitian Road, Futian District, Shenzhen (China)

    2012-07-01

    As the recent blooming economic growth and following environmental concerns (China)) is proactively pushing forward nuclear power development and encouraging the tapping of clean energy. Under this situation, CGNPC, as one of the largest energy enterprises in China, is planning to develop its own nuclear related technology in order to support more and more nuclear plants either under construction or being operation. This paper introduces the recent progress in software development for CGNPC. The focus is placed on the physical models and preliminary verification results during the recent development of the 3D Core Neutronic Code: COCO. In the COCO code, the non-linear Green's function method is employed to calculate the neutron flux. In order to use the discontinuity factor, the Neumann (second kind) boundary condition is utilized in the Green's function nodal method. Additionally, the COCO code also includes the necessary physical models, e.g. single-channel thermal-hydraulic module, burnup module, pin power reconstruction module and cross-section interpolation module. The preliminary verification result shows that the COCO code is sufficient for reactor core design and analysis for pressurized water reactor (PWR). (authors)

  9. Thermal neutron cross-section data for the design of cold and superthermal neutron sources

    NASA Astrophysics Data System (ADS)

    Bernnat, W.; Emendörfer, D.; Käfer, S.; Keinert, J.; Mattes, M.; Predel, M.

    1997-02-01

    For the relevant materials used in cold and superthermal neutron sources the database for design studies was considerably improved. For D 2O at room temperature the intra and inter molecular interaction with neutrons were included using realistic structure factors S( ?). These effects are also taken into account for the cold moderators liquid H 2 and D 2 by deriving improved S( ?). The neutron interaction in solid materials and neutron filters as Al, Pb and Bi is treated exactly at low temperatures, especially for coherent neutron scattering in the tetragonal and FCC lattices. For superfluid 4He cross-section data for ultracold neutron production were derived as well as data for neutron transport considering diffusive motions and intermolecular scattering. Validation of the generated cross-sectional data was realized by comparison with differential and integral neutron scattering experiments so far available. These thermal neutron data can be used for both multigroup SN and Monte-Carlo methods solving the neutron transport equation.

  10. Learn from the Core--Design from the Core

    ERIC Educational Resources Information Center

    Ockerse, Thomas

    2012-01-01

    The current objective, object-oriented approach to design is questioned along with design education viewed as a job-oriented endeavor. Instead relational knowledge and experience in a holistic sense, both tacit and explicit, are valued along with an appreciation of the unique character of the student. A new paradigm for design education is…

  11. Hydromagnetic and gravitomagnetic crust-core coupling in a precessing neutron star

    E-print Network

    Yuri Levin; Caroline D'Angelo

    2004-02-29

    We consider two types of mechanical coupling between the crust and the core of a precessing neutron star. First, we find that a hydromagnetic (MHD) coupling between the crust and the core strongly modifies the star's precessional modes when $t_a\\le\\sim (T_s\\times T_p)^{1/2}$; here $t_a$ is the Alfven crossing timescale, and $T_s$ and $T_p$ are the star's spin and precession periods, respectively. We argue that in a precessing pulsar PSR B1828-11 the restoring MHD stress prevents a free wobble of the crust relative to the non-precessing core. Instead, the crust and the proton-electron plasma in the core must precess in unison, and their combined ellipticity determines the period of precession. Link has recently shown that the neutron superfluid vortices in the core of PSR B1828-11 cannot be pinned to the plasma; he has also argued that this lack of pinning is expected if the proton Fermi liquid in the core is type-I superconductor. In this case, the neutron superfluid is dynamically decoupled from the precessing motion. The pulsar's precession decays due to the mutual friction between the neutron superfluid and the plasma in the core. The decay is expected to occur over tens to hundreds of precession periods and may be measurable over a human lifetime. Such a measurement would provide information about the strong n-p interaction in the neutron-star core. Second, we consider the effect of gravitomagnetic coupling between the neutron superfluid in the core and the rest of the star and show that this coupling changes the rate of precession by about 10%. The general formalism developed in this paper may be useful for other applications.

  12. Wireless, in-vessel neutron monitor for initial core-loading of advanced breeder reactors

    NASA Technical Reports Server (NTRS)

    Delorenzo, J. T.; Kennedy, E. J.; Blalock, T. V.; Rochelle, J. M.; Chiles, M. M.; Valentine, K. H.

    1981-01-01

    An experimental wireless, in-vessel neutron monitor was developed to measure the reactivity of an advanced breeder reactor as the core is loaded for the first time to preclude an accidental critically incident. The environment is liquid sodium at a temperature of approx. 220 C, with negligible gamma or neutron radiation. With ultrasonic transmission of neutron data, no fundamental limitation was observed after tests at 230 C for 2000 h. The neutron sensitivity was approx. 1 count/s-nv, and the potential data transmission rate was approx. 10,000 counts/s.

  13. Observational constraints on neutron star crust-core coupling during glitches

    E-print Network

    William G. Newton; Sabrina Berger; Brynmor Haskell

    2015-06-04

    We demonstrate that observations of glitches in the Vela pulsar can be used to investigate the strength of the crust-core coupling in a neutron star, and suggest that recovery from the glitch is dominated by torque exerted by the re-coupling of superfluid components of the core that were decoupled from the crust during the glitch. Assuming that the recoupling is mediated by mutual friction between the superfluid neutrons and the charged components of the core, we use the observed magnitudes and timescales of the shortest timescale components of the recoveries from two recent glitches in the Vela pulsar to infer the fraction of the core that is coupled to the crust during the glitch, and hence spun up by the glitch event. Within the framework of a two-fluid hydrodynamic model of glitches, we analyze whether crustal neutrons alone are sufficient to drive the glitch activity observed in the Vela pulsar. We use two sets of neutron star equations of state (EOSs), both of which span crust and core consistently and cover a range of the slope of the symmetry energy at saturation density $30 set produces maximum masses $\\approx$2.0$M_{\\odot}$, the second $\\approx$2.6$M_{\\odot}$. We also include the effects of entrainment of crustal neutrons by the superfluid lattice. We find that for medium to stiff EOSs, observations imply $>70\\%$ of the moment of inertia of the core is coupled to the crust during the glitch, though for softer EOSs $L\\approx 30$MeV as little as $5\\%$ could be coupled. No EOS is able to reproduce the observed glitch activity with crust neutrons alone, but extending the region where superfluid vortices are strongly pinned into the core by densities as little as 0.016fm$^{-3}$ above the crust-core transition density restores agreement with the observed glitch activity.

  14. Design and Test of Processor-Core Based Systems

    Microsoft Academic Search

    Peter Marwedel

    1997-01-01

    This tutorial responds to the rapidly increasing use of various cores for implementing systems-on-a-chip. It specific- ally focusses on processor cores. We will give some examples of cores, including DSP cores and application-specific instruction- set processors (ASIPs). We will mention market trends for these components, and we will touch design procedures, in particular the use compilers. Finally, we will discuss

  15. Advanced BWR core component designs and the implications for SFD analysis

    SciTech Connect

    Ott, L.J.

    1997-02-01

    Prior to the DF-4 boiling water reactor (BWR) severe fuel damage (SFD) experiment conducted at the Sandia National Laboratories in 1986, no experimental data base existed for guidance in modeling core component behavior under postulated severe accident conditions in commercial BWRs. This paper will present the lessons learned from the DF-4 experiment (and subsequent German CORA BWR SFD tests) and the impact on core models in the current generation of SFD codes. The DF-4 and CORA BWR test assemblies were modeled on the core component designs circa 1985; that is, the 8 x 8 fuel assembly with two water rods and a cruciform control blade constructed of B{sub 4}C-filled tubelets. Within the past ten years, the state-of-the-art with respect to BWR core component development has out-distanced the current SFD experimental data base and SFD code capabilities. For example, modern BWR control blade design includes hafnium at the tips and top of each control blade wing for longer blade operating lifetimes; also water rods have been replaced by larger water channels for better neutronics economy; and fuel assemblies now contain partial-length fuel rods, again for better neutronics economy. This paper will also discuss the implications of these advanced fuel assembly and core component designs on severe accident progression and on the current SFD code capabilities.

  16. Advanced Neutron Source radiological design criteria

    SciTech Connect

    Westbrook, J.L.

    1995-08-01

    The operation of the proposed Advanced Neutron Source (ANS) facility will present a variety of radiological protection problems. Because it is desired to design and operate the ANS according to the applicable licensing standards of the Nuclear Regulatory Commission (NRC), it must be demonstrated that the ANS radiological design basis is consistent not only with state and Department of Energy (DOE) and other usual federal regulations, but also, so far as is practicable, with NRC regulations and with recommendations of such organizations as the Institute of Nuclear Power Operations (INPO) and the Electric Power Research Institute (EPRI). Also, the ANS radiological design basis is in general to be consistent with the recommendations of authoritative professional and scientific organizations, specifically the National Council on Radiation Protection and Measurements (NCRP) and the International Commission on Radiological Protection (ICRP). As regards radiological protection, the principal goals of DOE regulations and guidance are to keep occupational doses ALARA [as low as (is) reasonably achievable], given the current state of technology, costs, and operations requirements; to control and monitor contained and released radioactivity during normal operation to keep public doses and releases to the environment ALARA; and to limit doses to workers and the public during accident conditions. Meeting these general design objectives requires that principles of dose reduction and of radioactivity control by employed in the design, operation, modification, and decommissioning of the ANS. The purpose of this document is to provide basic radiological criteria for incorporating these principles into the design of the ANS. Operations, modification, and decommissioning will be covered only as they are affected by design.

  17. An open core rotator design methodology

    Microsoft Academic Search

    J. A. Kirk; G. E. Sullivan; L. P. Hromada

    1997-01-01

    In low earth orbit [LEO] satellite applications, spacecraft power is provided by photovoltaic cells and batteries. To overcome battery shortcomings, FARE, Inc., working in cooperation with the University of Maryland [UOM] and the NASA Lewis Research Center, has developed an open core magnetically-suspended graphite-epoxy flywheel for energy storage applications. This flywheel energy storage system, called the Open Core Rotator [OCR],

  18. Determination of the kinetic parameters of the CALIBAN metallic core reactor from stochastic neutron measurements

    SciTech Connect

    Casoli, P.; Authier, N.; Chapelle, A. [Commissariat a l'Energie Atomique et Aux Energies Alternatives, CEA, DAM, F-21120 Is sur Tille (France)

    2012-07-01

    Several experimental devices are operated by the Criticality and Neutron Science Research Dept. of the CEA Valduc Laboratory. One of these is the Caliban metallic core reactor. The purpose of this study is to develop and perform experiments allowing to determinate some of fundamental kinetic parameters of the reactor. The prompt neutron decay constant and particularly its value at criticality can be measured with reactor noise techniques such as Rossi-{alpha} and Feynman variance-to-mean methods. Subcritical, critical, and even supercritical experiments were performed. Fission chambers detectors were put nearby the core and measurements were analyzed with the Rossi-{alpha} technique. A new value of the prompt neutron decay constant at criticality was determined, which allows, using the Nelson number method, new evaluations of the effective delayed neutron fraction and the in core neutron lifetime. As an introduction of this paper, some motivations of this work are given in part 1. In part 2, principles of the noise measurements experiments performed at the CEA Valduc Laboratory are reminded. The Caliban reactor is described in part 3. Stochastic neutron measurements analysis techniques used in this study are then presented in part 4. Results of fission chamber experiments are summarized in part 5. Part 6 is devoted to the current work, improvement of the experimental device using He 3 neutron detectors and first results obtained with it. Finally, conclusions and perspectives are given in part 7. (authors)

  19. Conceptual design of an RFQ accelerator-based neutron source for boron neutron-capture therapy

    Microsoft Academic Search

    T. P. Wangler; J. E. Stovall; T. S. Bhatia; C. K. Wang; T. E. Blue; R. A. Gahbauer

    1989-01-01

    A conceptual design of a low-energy neutron generator for treatment of brain tumors by boron neutron capture therapy (BNCT) is presented. The concept is based on a 2.5-MeV proton beam from a radio-frequency quadrupole (RFQ) linac, and the neutrons are produced by the 7Li(p,n)7Be reaction. A liquid lithium target and modulator assembly are designed to provide a high flux of

  20. Neutronic optimization of solid breeder blankets for STARFIRE design

    SciTech Connect

    Gohar, Y.; Abdou, M.A.

    1980-01-01

    Extensive neutronic tradeoff studies were carried out to define and optimize the neutronic performance of the different solid breeder options for the STARFIRE blanket design. A set of criteria were employed to select the potential blanket materials. The basic criteria include the neutronic performance, tritium-release characteristics, material compatibility, and chemical stability. Three blanket options were analyzed. The first option is based on separate zones for each basic blanket function where the neutron multiplier is kept in a separate zone. The second option is a heterogeneous blanket type with two tritium breeder zones. In the first zone the tritium breeder is assembled in a neutron multiplier matrix behind the first wall while the second zone has a neutron moderator matrix instead of the neutron multiplier. The third blanket option is similar to the second concept except the tritium breeder and the neutron multiplier form a homogeneous mixture.

  1. Thermal conductivity due to phonons in the core of superfluid neutron stars

    E-print Network

    Cristina Manuel; Sreemoyee Sarkar; Laura Tolos

    2014-11-22

    We compute the contribution of phonons to the thermal conductivity in the core of superfluid neutron stars. We use effective field theory techniques to extract the phonon scattering rates, written as a function of the equation of state of the system. We also calculate the phonon dispersion law beyond linear order, which depends on the gap of superfluid neutron matter. With all these ingredients, we solve the Boltzmann equation numerically using a variational approach. We find that the thermal conductivity $\\kappa$ is dominated by combined small and large angle binary collisions. As in the color-flavor-locked superfluid, we find that our result can be well approximated by $\\kappa \\propto 1/ \\Delta^6$, where $\\Delta$ is the neutron gap, the constant of proportionality depending on the density. We further comment on the possible relevance of electron and superfluid phonon collisions in obtaining the total contribution to the thermal conductivity in the core of superfluid neutron stars.

  2. IRSN working program status on tools for evaluation of SFR cores static neutronics safety parameters

    SciTech Connect

    Ivanov, E.; Tiberi, V.; Ecrabet, F.; Chegrani, Y.; Canuti, E.; Bisogni, D.; Sargeni, A.; Bernard, F. [Institut de Radioprotection et de Surete Nucleaire IRSN, BP 17, 92262 Fontenay-aux-roses (France)

    2012-07-01

    As technical support of the French Nuclear Safety Authority, IRSN will be in charge of safety assessment of any future project of Sodium Fast Reactor (SFR) that could be built in France. One of the main safety topics will deal with reactivity control. Since the design and safety assessment of the last two SFR plants in France (Phenix and Superphenix, more than thirty years ago), methods, codes and safety objectives have evolved. That is why a working program on core neutronic simulations has been launched in order to be able to evaluate accuracy of future core characteristics computations. The first step consists in getting experienced with the ERANOS well-known deterministic code used in the past for Phenix and Superphenix. Then Monte-Carlo codes have been tested to help in the interpretation of ERANOS results and to define what place this kind of codes can have in a new SFR safety demonstration. This experience is based on open benchmark computations. Different cases are chosen to cover a wide range of configurations. The paper shows, as an example, criticality results obtained with ERANOS, SCALE and MORET, and the first conclusions based on these results. In the future, this work will be extended to other safety parameters such as sodium void and Doppler effects, kinetic parameters or flux distributions. (authors)

  3. Pre-conceptual design study of ASTRID core

    SciTech Connect

    Varaine, F.; Marsault, P.; Chenaud, M. S.; Bernardin, B.; Conti, A.; Sciora, P.; Venard, C.; Fontaine, B.; Devictor, N.; Martin, L. [Alternative Energies and Atomic Energy Commission, CEA, DEN DER, 13108 St Paul lez Durance (France); Scholer, A. C.; Verrier, D. [AREVA-NP, 10 rue J. Recamier, 69456 Lyon Cedex 06 (France)

    2012-07-01

    In the framework of the ASTRID project at CEA, core design studies are performed at CEA with the AREVA and EDF support. At the stage of the project, pre-conceptual design studies are conducted in accordance with GEN IV reactors criteria, in particularly for safety improvements. An improved safety for a sodium cooled reactor requires revisiting many aspects of the design and is a rather lengthy process in current design approach. Two types of cores are under evaluation, one classical derived from the SFR V2B and one more challenging called CFV (low void effect core) with a large gain on the sodium void effect. The SFR V2b core have the following specifications: a very low burn-up reactivity swing (due to a small cycle reactivity loss) and a reduced sodium void effect with regard to past designs such as the EFR (around 2$ minus). Its performances are an average burn-up of 100 GWd/t, and an internal conversion ratio equal to one given a very good behavior of this core during a control rod withdrawal transient). The CFV with its specific design offers a negative sodium void worth while maintaining core performances. In accordance of ASTRID needs for demonstration those cores are 1500 MWth power (600 MWe). This paper will focus on the CFV pre-conceptual design of the core and S/A, and the performances in terms of safety will be evaluated on different transient scenario like ULOF, in order to assess its intrinsic behavior compared to a more classical design like V2B core. The gap in term of margin to a severe accident due to a loss of flow initiator underlines the potential capability of this type of core to enhance prevention of severe accident in accordance to safety demonstration. (authors)

  4. MAGNETO-ROTATIONAL NEUTRON STAR EVOLUTION: THE ROLE OF CORE VORTEX PINNING

    SciTech Connect

    Glampedakis, Kostas [Theoretical Astrophysics, University of Tuebingen, Tuebingen D-72076 (Germany); Andersson, Nils [School of Mathematics, University of Southampton, Southampton SO17 1BJ (United Kingdom)

    2011-10-20

    We consider the pinning of superfluid (neutron) vortices to magnetic fluxtubes associated with a type II (proton) superconductor in neutron star cores. We demonstrate that core pinning affects the spin-down of the system significantly and discuss implications for regular radio pulsars and magnetars. We find that magnetars are likely to be in the pinning regime, whereas most radio pulsars are not. This suggests that the currently inferred magnetic field for magnetars may be overestimated. We also obtain a new timescale for the magnetic field evolution which could be associated with the observed activity in magnetars, provided that the field has a strong toroidal component.

  5. Neutronic evaluation of GCFR core diluents and reflectors

    E-print Network

    Yu, Kun, 1974-

    2003-01-01

    Materials are evaluated for use as in-core diluents and as peripheral reflectors for Gas-Cooled Fast Reactor (GFR) service, using coupled Monte Carlo (MCNP) and isotopics (ORIGEN) codes. The principal performance indices ...

  6. Effect of core structure irradiation on the RBMK neutron characteristics

    SciTech Connect

    Balygin, A. A., E-mail: balyg@dcnr.vver.kiae.ru; Krayushkin, A. V. [National Research Center Kurchatov Institute (Russian Federation)

    2014-12-15

    The effect of changes in the graphite density and fuel channel diameters on the RBMK neutron characteristics is estimated. It is shown that uncertainty of those quantities can lead to a noticeable difference between the calculated and experimental values of the steam coefficient of reactivity and the subcriticality of the reactor.

  7. Advanced Neutron Source design: Burnout heat flux correlation development

    SciTech Connect

    Gambill, W.R.; Mochizuki, T.

    1988-01-01

    In the Advanced Neutron Source Reactor (ANSR) fuel element region, heat fluxes will be elevated. Early designs corresponded to average and estimated hot-spot fluxes of 11-12 and 21-22 MW/m/sup 2/, respectively. Design changes under consideration may lower these values to about 9 and 17 MW/m/sup 2/. In either event, the development of a satisfactory burnout heat flux correlation is an important element among the many thermal-hydraulic design issues, since the critical power ration will depend in part on its validity. Relatively little work in the area of subcooled-flow burnout has been published over the past 12 years. We have compared seven burnout correlations and modifications thereof with several sets of experimental data, of which the most relevant to the ANS core are presently those referenced. The best overall agreement between the correlations tested and these data is currently provided by a modification of Thorgerson's correlation. 7 refs., 1 tab.

  8. Low-Enriched Fuel Design Concept for the Prismatic Very High Temperature Reactor Core

    SciTech Connect

    Sterbentz, James W

    2007-05-01

    A new non-TRISO fuel and clad design concept is proposed for the prismatic, heliumcooled Very High Temperature Reactor core. The new concept could substantially reduce the current 10-20 wt% TRISO uranium enrichments down to 4-6 wt% for both initial and reload cores. The proposed fuel form would be a high-temperature, high-density uranium ceramic, for example UO2, configured into very small diameter cylindrical rods. The small diameter fuel rods significantly increase core reactivity through improved neutron moderation and fuel lumping. Although a high-temperature clad system for the concept remains to be developed, recent success in tube fabrication and preliminary irradiation testing of silicon carbide (SiC) cladding for light water reactor applications offers good potential for this application, and for future development of other carbide clad designs. A high-temperature ceramic fuel, together with a high-temperature clad material, could also lead to higher thermal safety margins during both normal and transient reactor conditions relative to TRISO fuel. The calculated neutronic results show that the lowenrichment, small diameter fuel rods and low thermal neutron absorbing clad retain the strong negative Doppler fuel temperature coefficient of reactivity that ensures inherent safe operation of the VHTR, and depletion studies demonstrate that an 18-month power cycle can be achieved with the lower enrichment fuel.

  9. TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR.

    PubMed

    Kurosawa, Masahiko

    2005-01-01

    For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54Mn and 60Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data. PMID:16604689

  10. Neutronic analysis of boiling water reactor in-core detector noise

    Microsoft Academic Search

    H. S. Cheng; D. J. Diamond

    1979-01-01

    The response of boiling water reactor in-core detectors undergoing vibration has been calculated. A neutronic model based on calculating the fission activity at a detector position in a planar multibundle environment was employed. The model used eight energy groups and two-dimensional Cartesian geometry in a discrete-ordinates transport approximation. The in-core detector responses due to various detector displacements were calculated as

  11. On design of air-core Ethernet transformers

    NASA Astrophysics Data System (ADS)

    Bowen, D.; Mayergoyz, I.; Krafft, C.; Kroop, D.; Beyaz, M.

    2009-04-01

    The air-core designs of Ethernet transformers in which the reduction of winding magnetic coupling is compensated by increasing the cross-winding capacitance coupling are discussed. The theoretical analysis that accounts for the distributed nature of the cross-winding capacitance is presented. It is demonstrated that the capacitive cross-winding coupling may result in desirable transfer characteristics of air-core transformers that are similar to those of ferrite core Ethernet transformers.

  12. Hans A. Bethe Prize: Neutron Stars and Core-Collapse Supernovae

    NASA Astrophysics Data System (ADS)

    Lattimer, James

    2015-04-01

    Core-collapse supernovae lead to the formation of neutron stars, and both are sensitive to the dense matter equation of state. Hans Bethe first recognized that the matter in the collapsing core of a massive star has a relatively low entropy which prevents nuclear dissociation until nuclei merge near the nuclear saturation density. This recognition means that collapse continues until the core exceeds the saturation density. This prediction forms the foundation for modern simulations of supernovae. These supernovae sample matter up to about twice nuclear saturation density, but neutron stars are sensitive to the equation of state both near the saturation density and at several times higher densities. Two important recent developments are the discovery of two-solar mass neutron stars and refined experimental determinations of the behavior of the symmetry energy of nuclear matter near the saturation density. Combined with the assumption of causality, they imply that the radii of observed neutron stars are largely independent of their mass, and that this radius is in the range of 11 to 13 km. These theoretical results are not only consistent with expectations from theoretical studies of pure neutron matter, but also accumulated observations of both bursting and cooling neutron stars. In the near future, new pulsar timing data, which could lead to larger measured masses as well as measurements of moments of inertia, X-ray observations, such as from NICER, of bursting and other sources, and gravitational wave observations of neutron stars in merging compact binaries, will provide important new constraints on neutron stars and the dense matter equation of state. DOE DE-FG02-87ER-40317.

  13. Divertor Neutron Flux Monitor: Conceptual Design and Calibration

    SciTech Connect

    Kaschuck Yu, A.; Batyunin, A. V.; Egorov, O. G.; Krasil'nikov, A. V.; Tsutskih, A. Yu; Frunze, V. V. [State Research Center of the Russian Federation, Troitsk Institute for Innovation and Fusion Research, Troitsk, Moscow oblast, 142190 (Russian Federation)

    2008-03-12

    The operating conditions of a neutron diagnostic systems which are responsible for measuring of the total neutron yield in the ITER tokamak reactor are analyzed. Based on results of neutronic calculations and analysis of suitable methods for measuring the neutron yield, a concept of a system for neutron flux measurement in the divertor is proposed. The design for the neutron flux monitor located in the divertor cassette of the tokamak is selected in view of the requirements specified for the neutron diagnostic system of the ITER and its operating conditions. Several fission chambers with different sensitivities and radiator materials will be used for measurements. System is capable of neutron fluxes measuring over the entire dynamic range of the ITER neutron yield with an error of {delta}<10% and a time resolution of 1 ms that are necessary for studying the physical phenomena of ignition and burning plasma. The problems of carrying out of the divertor neutron monitor efficiency calibration with the aim to measure the absolute value of the neutron yield in the ITER tokamak reactor are also discussed.

  14. Designing systems-on-chip using cores

    Microsoft Academic Search

    Reinaldo A. Bergamaschi; William R. Lee

    2000-01-01

    Leading-edge systems-on-chip (SoC) being designed today could reach 20 Million gates and 0.5 to 1 GHz operating frequency. In order to implement such systems, designers are increasingly relying on reuse of Intellectual property (IP) blocks. Since IP blocks are pre-designed and pre-verified, the designer can concentrate on the complete system without having to worry about the correctness or performance of

  15. Sifting through the many-core design space

    E-print Network

    Mullins, Robert

    Sifting through the many-core design space Robert Mullins Computer Laboratory, University #12;A new design space · Much larger set of viable designs ­ More than a single pipeline to optimise · Disparate architectures ­ Beyond differences in simple parameters · Can't build a simple parameterised model

  16. Fuel and Core Design Experiences in Cofrentes NPP

    SciTech Connect

    Garcia-Delgado, L.; Lopez-Carbonell, M.T.; Gomez-Bernal, I. [Iberdrola Generacion, Nuclear Fuel Department, Hermosilla 3, 28001 Madrid (Spain)

    2002-07-01

    The electricity market deregulation in Spain is increasing the need for innovations in nuclear power generation, which can be achieved in the fuel area by improving fuel and core designs and by introducing vendors competition. Iberdrola has developed the GIRALDA methodology for design and licensing of Cofrentes reloads, and has introduced mixed cores with fuel from different vendors. The application of GIRALDA is giving satisfactory results, and is showing its capability to adequately reproduce the core behaviour. The nuclear design team is acquiring an invaluable experience and a deep knowledge of the core, very useful to support cycle operation. Continuous improvements are expected for the future in design strategies as well as in the application of new technologies to redesign the methodology processes. (authors)

  17. VCDS: virtual core based design system

    Microsoft Academic Search

    M. Muraoka

    1999-01-01

    The Design Productivity Crisis of LSI towards 2010 have been discussed for a few years especially in SEMATECH, USA. The innovation of LSI design methodology will be the most effective way to resolve the issues of the design crisis. SIRIJ, Semiconductor Industry Research Institute Japan, has organized `VCDS Committee' to research the next generation EDA system towards 2010. `VCDS: Virtual

  18. Neutronics Assessment of Molten Salt Breeding Blanket Design

    E-print Network

    1 Neutronics Assessment of Molten Salt Breeding Blanket Design Options Mohamed Sawan Fusion adequate tritium breeding and shielding for VV and magnet Larger margins are considered to account flow channel required to cool it #12;5 Tritium Breeding Potential If neutron coverage for double null

  19. An integrated design of an accelerator-based neutron source for boron neutron capture therapy

    Microsoft Academic Search

    Michael Christian Dobelbower

    1997-01-01

    An Accelerator Based Neutron Source (ABNS) for Boron Neutron Capture Therapy (BNCT) was first proposed at The Ohio State University (OSU). Since the conception of the ABNS for BNCT, OSU has designed and optimized a moderator assembly based on in-air and in-phantom parameters. Additionally, the fabrication of the moderator assembly has commenced along with detailed analyses of the target and

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

    SciTech Connect

    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.

  1. Verifying IP-core based system-on-chip designs

    Microsoft Academic Search

    Pankaj Chauhan; Edmund M. Clarke; Yuan Lu; Dong Wang

    1999-01-01

    We describe a methodology for verifying system-on-chip designs. In our methodology, the problem of verifying system-on-chip designs is decomposed into three tasks. First, we verify, once and for all, the standard bus interconnecting IP cores in the system. The next task is to verify the glue logic, which connects the IP cores to the buses. Finally, using the verified bus

  2. CORE ANALYSIS, DESIGN AND OPTIMIZATION OF A DEEP-BURN PEBBLE BED REACTOR

    SciTech Connect

    B. Boer; A. M. Ougouag

    2010-05-01

    Achieving a high burnup in the Deep-Burn pebble bed reactor design, while remaining within the limits for fuel temperature, power peaking and temperature reactivity feedback, is challenging. The high content of Pu and Minor Actinides in the Deep-Burn fuel significantly impacts the thermal neutron energy spectrum. This can result in power and temperature peaking in the pebble bed core in locally thermalized regions near the graphite reflectors. Furthermore, the interplay of the Pu resonances of the neutron absorption cross sections at low-lying energies can lead to a positive temperature reactivity coefficient for the graphite moderator at certain operating conditions. To investigate the aforementioned effects a code system using existing codes has been developed for neutronic, thermal-hydraulic and fuel depletion analysis of Deep-Burn pebble bed reactors. A core analysis of a Deep-Burn Pebble Bed Modular Reactor (400 MWth) design has been performed for two Deep-Burn fuel types and possible improvements of the design with regard to power peaking and temperature reactivity feedback are identified.

  3. De novo design of the hydrophobic core of ubiquitin.

    PubMed Central

    Lazar, G. A.; Desjarlais, J. R.; Handel, T. M.

    1997-01-01

    We have previously reported the development and evaluation of a computational program to assist in the design of hydrophobic cores of proteins. In an effort to investigate the role of core packing in protein structure, we have used this program, referred to as Repacking of Cores (ROC), to design several variants of the protein ubiquitin. Nine ubiquitin variants containing from three to eight hydrophobic core mutations were constructed, purified, and characterized in terms of their stability and their ability to adopt a uniquely folded native-like conformation. In general, designed ubiquitin variants are more stable than control variants in which the hydrophobic core was chosen randomly. However, in contrast to previous results with 434 cro, all designs are destabilized relative to the wild-type (WT) protein. This raises the possibility that beta-sheet structures have more stringent packing requirements than alpha-helical proteins. A more striking observation is that all variants, including random controls, adopt fairly well-defined conformations, regardless of their stability. This result supports conclusions from the cro studies that non-core residues contribute significantly to the conformational uniqueness of these proteins while core packing largely affects protein stability and has less impact on the nature or uniqueness of the fold. Concurrent with the above work, we used stability data on the nine ubiquitin variants to evaluate and improve the predictive ability of our core packing algorithm. Additional versions of the program were generated that differ in potential function parameters and sampling of side chain conformers. Reasonable correlations between experimental and predicted stabilities suggest the program will be useful in future studies to design variants with stabilities closer to that of the native protein. Taken together, the present study provides further clarification of the role of specific packing interactions in protein structure and stability, and demonstrates the benefit of using systematic computational methods to predict core packing arrangements for the design of proteins. PMID:9194177

  4. Cooling of neutron stars with color superconducting quark cores

    SciTech Connect

    Grigorian, Hovik [Institut fuer Physik, Universitaet Rostock, D-18051 Rostock (Germany); Department of Physics, Yerevan State University, Alex Manoogian Street 1, 375025 Yerevan (Armenia); Blaschke, David [Fakultaet fuer Physik, Universitaet Bielefeld, D-33615 Bielefeld (Germany); Bogoliubov Laboratory of Theoretical Physics, Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Voskresensky, Dmitri [Theory Division, GSI mbH, D-64291 Darmstadt (Germany); Moscow Institute for Physics and Engineering, 115409 Moscow (Russian Federation)

    2005-04-01

    We show that within a recently developed nonlocal, chiral quark model the critical density for a phase transition to color superconducting quark matter under neutron star conditions can be low enough for these phases to occur in compact star configurations with masses below 1.3 M{sub {center_dot}}. We study the cooling of these objects in isolation for different values of the gravitational mass. Our equation of state (EoS) allows for two-flavor color superconductivity (2SC) quark matter with a large quark gap ({approx}100 MeV) for u and d quarks of two colors that coexists with normal quark matter within a mixed phase in the hybrid star interior. We argue that, if the phases with unpaired quarks were allowed, the corresponding hybrid stars would cool too fast. If they occurred for M<1.3 M{sub {center_dot}}, as follows from our EoS, one could not appropriately describe the neutron star cooling data existing today. We discuss a ''2SC+X'' phase as a possibility for having all quarks paired in two-flavor quark matter under neutron star constraints, where the X gap is of the order of 10 keV-1 MeV. Density-independent gaps do not allow us to fit the cooling data. Only the presence of an X gap that decreases with increasing density would allow us to appropriately fit the data in a similar compact star mass interval to that following from a purely hadronic model. This scenario is suggested as an alternative explanation of the cooling data in the framework of a hybrid star model.

  5. Verification of JUPITER Standard Analysis Method for Upgrading Joyo MK-III Core Design and Management

    NASA Astrophysics Data System (ADS)

    Maeda, Shigetaka; Ito, Chikara; Sekine, Takashi; Aoyama, Takafumi

    In the experimental fast reactor Joyo, loading of irradiation test rigs causes a decrease in excess reactivity because the rigs contain less fissile materials than the driver fuel. In order to carry out duty operation cycles using as many irradiation rigs as possible, it is necessary to upgrade the core performance to increase its excess reactivity and irradiation capacity. Core modification plans have been considered, such as the installation of advanced radial reflectors and reduction of the number of control rods. To implement such core modifications, it is first necessary to improve the prediction accuracy in core design and to optimize safety margins. In the present study, verification of the JUPITER fast reactor standard analysis method was conducted through a comparison between the calculated and the measured Joyo MK-III core characteristics, and it was concluded that the accuracy for a small sodium-cooled fast reactor with a hard neutron spectrum was within 5 % of unity. It was shown that, the performance of the irradiation bed core could be upgraded by the improvement of the prediction accuracy of the core characteristics and optimization of safety margins.

  6. Neutronic Design Calculations on Moderators for the Spallation Neutron Source (SNS)

    SciTech Connect

    Murphy, D.B.

    1999-11-14

    The Spallation Neutron Source (SNS) to be built at the Oak Ridge National Laboratory will provide an intense source of neutrons for a large variety of experiments. It consists of a high-energy (1-GeV) and high-power ({approximately}1-MW) proton accelerator, an accumulator ring, together with a target station and an experimental area. In the target itself, the proton beam will produce neutrons via the spallation process and these will be converted to low-energy (<2-eV) neutrons in moderators located close to the target. Current plans are to have two liquid-hydrogen (20-K) moderators and two room-temperature H{sub 2}O moderators. Extensive engineering design work has been conducted on the moderator vessels. For our studies we have produced realistic neutronic representations of these moderators. We report on neutronic studies conducted on these representations of the moderators using Monte Carlo simulation techniques.

  7. Evaluation for 4S core nuclear design method through integration of benchmark data

    SciTech Connect

    Nagata, A.; Tsuboi, Y. [Advanced System Design and Engineering Dept., Isogo Nuclear Engineering Center, Toshiba Corporation, 8, Shinsugita-Cho, Isogo-ku, Yokohama, 235-8523 (Japan); Moriki, Y. [Power and Industrial Systems Research and Development Center, Toshiba Corporation, 8, Shinsugita-Cho, Isogo-ku, Yokohama, 235-8523 (Japan); Kawashima, M. [Nuclear Technology Application Dept., Toshiba Nuclear Engineering Services Corporation, 8, Shinsugita-Cho, Isogo-ku, Yokohama, 235-8523 (Japan)

    2012-07-01

    The 4S is a sodium-cooled small fast reactor which is reflector-controlled for operation through core lifetime about 30 years. The nuclear design method has been selected to treat neutron leakage with high accuracy. It consists of a continuous-energy Monte Carlo code, discrete ordinate transport codes and JENDL-3.3. These two types of neutronic analysis codes are used for the design in a complementary manner. The accuracy of the codes has been evaluated by analysis of benchmark critical experiments and the experimental reactor data. The measured data used for the evaluation is critical experimental data of the FCA XXIII as a physics mockup assembly of the 4S core, FCA XVI, FCA XIX, ZPR, and data of experimental reactor JOYO MK-1. Evaluated characteristics are criticality, reflector reactivity worth, power distribution, absorber reactivity worth, and sodium void worth. A multi-component bias method was applied, especially to improve the accuracy of sodium void reactivity worth. As the result, it has been confirmed that the 4S core nuclear design method provides good accuracy, and typical bias factors and their uncertainties are determined. (authors)

  8. Simultaneous measurement of neutron and gamma-ray radiation levels from a TRIGA reactor core using silicon carbide semiconductor detectors

    Microsoft Academic Search

    A. R. Dulloo; F. H. Ruddy; J. G. Seidel; C. Davison; T. Flinchbaugh; T. Daubenspeck

    1998-01-01

    The ability of a SiC detector to measure neutron and gamma radiation levels in a TRIGA reactor's mixed neutron\\/gamma field was demonstrated. Linear responses to an epicadmium neutron fluence rate (up to 3×107 cm-2 s-1) and to a gamma dose rate (0.6-234 krad-Si h-1) were obtained with the detector. Axial profiles of the reactor core's neutron and gamma-ray radiation levels

  9. The new Cold Neutron Chopper Spectrometer at the Spallation Neutron Source - Design and Performance

    E-print Network

    Ehlers, G; Niedziela, J L; Iverson, E B

    2011-01-01

    The design and performance of the new Cold Neutron Chopper Spectrometer (CNCS) at the Spallation Neutron Source in Oak Ridge are described. CNCS is a direct geometry inelastic time-of-flight spectrometer, designed to cover essentially the same energy and momentum transfer ranges as IN5 at ILL, LET at ISIS, DCS at NIST, TOFTOF at FRM2, AMATERAS at J-PARC, PHAROS at LANSCE and NEAT at HZB, at similar energy resolution. Measured values of key figures such as neutron flux at sample position and energy resolution are compared between measurements and ray tracing Monte Carlo simulations, and good agreement (better than 20% of absolute numbers) has been achieved. The instrument performs very well in the cold and thermal neutron energy ranges, and promises to become a workhorse for the neutron scattering community for quasielastic and inelastic scattering experiments.

  10. The new cold neutron chopper spectrometer at the Spallation Neutron Source: Design and performance

    NASA Astrophysics Data System (ADS)

    Ehlers, G.; Podlesnyak, A. A.; Niedziela, J. L.; Iverson, E. B.; Sokol, P. E.

    2011-08-01

    The design and performance of the new cold neutron chopper spectrometer (CNCS) at the Spallation Neutron Source in Oak Ridge are described. CNCS is a direct-geometry inelastic time-of-flight spectrometer, designed essentially to cover the same energy and momentum transfer ranges as IN5 at ILL, LET at ISIS, DCS at NIST, TOFTOF at FRM-II, AMATERAS at J-PARC, PHAROS at LANSCE, and NEAT at HZB, at similar energy resolution. Measured values of key figures such as neutron flux at sample position and energy resolution are compared between measurements and ray tracing Monte Carlo simulations, and good agreement (better than 20% of absolute numbers) has been achieved. The instrument performs very well in the cold and thermal neutron energy ranges, and promises to become a workhorse for the neutron scattering community for quasielastic and inelastic scattering experiments.

  11. Design and Demonstration of a Quasi-monoenergetic Neutron Source

    SciTech Connect

    Joshi, T.; Sangiorgio, Samuele; Mozin, Vladimir V.; Norman, E. B.; Sorensen, Peter F.; Foxe, Michael P.; Bench, G.; Bernstein, A.

    2014-03-05

    The design of a neutron source capable of producing 24 and 70 keV neutron beams with narrow energy spread is presented. The source exploits near-threshold kinematics of the 7Li(p,n)7Be reaction while taking advantage of the interference `notches' found in the scattering cross-sections of iron. The design was implemented and characterized at the Center for Accelerator Mass Spectrometry at Lawrence Livermore National Laboratory. Alternative lters such as vanadium and manganese are also explored and the possibility of studying the response of di*erent materials to low-energy nuclear recoils using the resultant neutron beams is discussed.

  12. Cooling neutron star in the Cassiopeia A supernova remnant: evidence for superfluidity in the core

    NASA Astrophysics Data System (ADS)

    Shternin, Peter S.; Yakovlev, Dmitry G.; Heinke, Craig O.; Ho, Wynn C. G.; Patnaude, Daniel J.

    2011-03-01

    According to recent results of Ho & Heinke, the Cassiopeia A supernova remnant contains a young (?330-yr-old) neutron star (NS) which has carbon atmosphere and shows notable decline of the effective surface temperature. We report a new (2010 November) Chandra observation which confirms the previously reported decline rate. The decline is naturally explained if neutrons have recently become superfluid (in triplet state) in the NS core, producing a splash of neutrino emission due to Cooper pair formation (CPF) process that currently accelerates the cooling. This scenario puts stringent constraints on poorly known properties of NS cores: on density dependence of the temperature Tcn(?) for the onset of neutron superfluidity [Tcn(?) should have a wide peak with maximum ? (7-9) × 108 K]; on the reduction factor q of CPF process by collective effects in superfluid matter (q > 0.4) and on the intensity of neutrino emission before the onset of neutron superfluidity (30-100 times weaker than the standard modified Urca process). This is serious evidence for nucleon superfluidity in NS cores that comes from observations of cooling NSs.

  13. Massive neutron stars with a hyperonic core: A case study with the IUFSU relativistic effective interaction

    NASA Astrophysics Data System (ADS)

    Bhowmick, Bipasha; Bhattacharya, Madhubrata; Bhattacharyya, Abhijit; Gangopadhyay, G.

    2014-06-01

    The recent discoveries of massive neutron stars, such as PSR J0348+0432 and PSR J1614-2230, have raised questions about the existence of exotic matter such as hyperons in the neutron star core. The validity of many established equations of states (EoSs) like GM1 and FSUGold are also questioned. We investigate the existence of hyperonic matter in the central regions of massive neutron stars by using relativistic mean field (RMF) theory with the recently proposed Indiana University Florida State University (IUFSU) model. The IUFSU model is extended by including hyperons to study the neutron star in ? equilibrium. The effect of different hyperonic potentials, namely ? and ? potentials, on the EoS and hence the maximum mass of neutron stars has been studied. We have also considered the effect of stellar rotation since the observed massive stars are pulsars. It has been found that a maximum mass of 1.93M?, which is within the 3? limit of the observed mass of PSR J0348+0432, can be obtained for rotating stars, with certain choices of the hyperonic potentials. The said star contains a fair amount of hyperons near the core.

  14. Core-collapse Supernova Equations of State Based on Neutron Star Observations

    NASA Astrophysics Data System (ADS)

    Steiner, A. W.; Hempel, M.; Fischer, T.

    2013-09-01

    Many of the currently available equations of state for core-collapse supernova simulations give large neutron star radii and do not provide large enough neutron star masses, both of which are inconsistent with some recent neutron star observations. In addition, one of the critical uncertainties in the nucleon-nucleon interaction, the nuclear symmetry energy, is not fully explored by the currently available equations of state. In this article, we construct two new equations of state which match recent neutron star observations and provide more flexibility in studying the dependence on nuclear matter properties. The equations of state are also provided in tabular form, covering a wide range in density, temperature, and asymmetry, suitable for astrophysical simulations. These new equations of state are implemented into our spherically symmetric core-collapse supernova model, which is based on general relativistic radiation hydrodynamics with three-flavor Boltzmann neutrino transport. The results are compared with commonly used equations of state in supernova simulations of 11.2 and 40 M ? progenitors. We consider only equations of state which are fitted to nuclear binding energies and other experimental and observational constraints. We find that central densities at bounce are weakly correlated with L and that there is a moderate influence of the symmetry energy on the evolution of the electron fraction. The new models also obey the previously observed correlation between the time to black hole formation and the maximum mass of an s = 4 neutron star.

  15. Automatic Design of Area-Efficient Configurable ASIC Cores

    E-print Network

    Compton, Katherine

    Automatic Design of Area-Efficient Configurable ASIC Cores Katherine Compton University for a given computation domain, we explore the design space between an ASIC and an FPGA. However, the manual to given application sets. This article discusses configurable ASIC (cASIC) architecture generation

  16. The r-mode instability of neutron stars with the superfluid core

    E-print Network

    Lee, U; Lee, Umin; Yoshida, Shijun

    2002-01-01

    We investigate the modal properties of the r-modes with l'=m in rotating neutron stars with the superfluid core, where l' and m are the indices of the spherical harmonic function representing the angular dependence of the r-modes. For the modal analysis, we employ simple neutron star models, which are composed of the normal fluid envelope and the core filled with neutron and proton superfluids and a normal fluid of electron. The stability of the r-modes against gravitational radiation reaction is examined by taking account of viscous dissipations due to shear and a damping mechanism called mutual friction in the superfluid core. At a given rotation frequency Omega, we find three r-modes, for which the values of the coefficient $kappa_0\\equiv\\omega/\\Omega$ in the limit of Omega --> 0 are numerically indistinguishable, corresponding to 2m/[l'(l'+1)], where omega is the oscillation frequency observed in the corotating frame. We find the damping due to the mutual friction in the superfluid core is effective to st...

  17. Design assumptions and bases for small D-T-fueled Sperical Tokamak (ST) fusion core

    SciTech Connect

    Peng, Y.K.M.; Galambos, J.D.; Fogarty, P.J. [and others

    1996-12-31

    Recent progress in defining the assumptions and clarifying the bases for a small D-T-fueled ST fusion core are presented. The paper covers several issues in the physics of ST plasmas, the technology of neutral beam injection, the engineering design configuration, and the center leg material under intense neutron irradiation. This progress was driven by the exciting data from pioneering ST experiments, a heightened interest in proof-of-principle experiments at the MA level in plasma current, and the initiation of the first conceptual design study of the small ST fusion core. The needs recently identified for a restructured fusion energy sciences program have provided a timely impetus for examining the subject of this paper. Our results, though preliminary in nature, strengthen the case for the potential realism and attractiveness of the ST approach.

  18. Design of Reliable Metro Core Networks

    Microsoft Academic Search

    P. Castoldi; F. Cugini; P. Ghelfil; L. Valcarenghil; G. Franzl; P. Gravey; M. Morvan; L. Rea; F. Matera; K. Wajda

    2007-01-01

    In this study the design of reliable network topologies is applied to a typical metro network scenario. Results show that, with an increasing amount of traffic, ring topologies represent a less attractive solution and, on the other hand, highly connected DWDM mesh networks are required to grant the required level of throughput for bandwidth demanding applications and services.

  19. A Tight Lattice, Epithermal Core Design for the Integral PWR

    SciTech Connect

    Saccheri, J.G.B. [Brookhaven National Laboratory, Nuclear Science and Technology Division Bldg 475, Upton, New York 11973-5000 (United States); Todreas, N.E.; Driscoll, M.J. [Massachusetts Institute of Technology, 77 Massachusetts Avenue, Cambridge, Bldg. 24-205 MA 02139-4307 (United States)

    2004-07-01

    An 8-year core design for an epithermal, water-cooled reactor has been developed based upon assessments of nuclear reactor physics, thermal-hydraulics and economics. An integral vessel configuration is adopted and self-supporting wire-wrap fuel is employed for the tight lattice of the epithermal core. A streaming path is incorporated in each assembly to ensure a negative void coefficient. A whole-core MCNP simulation of the tight core shows a negative void coefficient for any burnup with positive K{sub EFF}. The VIPRE{sup TM} code has been used to calculate the critical heat flux (CHF) by means of an appropriate wire-wrap CHF correlation, specifically introduced in the source code. Economically, the high fuel enrichment (14% w/o {sup 235}U) and the very long core life (8 ys) lead to high lifetime-levelized unit fuel cycle cost (in mills/kWhre). However, both operation and maintenance and capital-related expenditures strongly benefited from the higher electric output per unit volume, which yielded quite small lifetime-levelized unit capital and operation and maintenance costs for the overall plant. Financing costs are included and an estimate is provided for the total lifetime-levelized unit cost of the epithermal core, which is about 20% lower than that of a more open lattice thermal spectrum core fitting into the same core envelope and with 4-year lifetime. (authors)

  20. Current advances in precious metal core-shell catalyst design

    NASA Astrophysics Data System (ADS)

    Wang, Xiaohong; He, Beibei; Hu, Zhiyu; Zeng, Zhigang; Han, Sheng

    2014-08-01

    Precious metal nanoparticles are commonly used as the main active components of various catalysts. Given their high cost, limited quantity, and easy loss of catalytic activity under severe conditions, precious metals should be used in catalysts at low volumes and be protected from damaging environments. Accordingly, reducing the amount of precious metals without compromising their catalytic performance is difficult, particularly under challenging conditions. As multifunctional materials, core-shell nanoparticles are highly important owing to their wide range of applications in chemistry, physics, biology, and environmental areas. Compared with their single-component counterparts and other composites, core-shell nanoparticles offer a new active interface and a potential synergistic effect between the core and shell, making these materials highly attractive in catalytic application. On one hand, when a precious metal is used as the shell material, the catalytic activity can be greatly improved because of the increased surface area and the closed interfacial interaction between the core and the shell. On the other hand, when a precious metal is applied as the core material, the catalytic stability can be remarkably improved because of the protection conferred by the shell material. Therefore, a reasonable design of the core-shell catalyst for target applications must be developed. We summarize the latest advances in the fabrications, properties, and applications of core-shell nanoparticles in this paper. The current research trends of these core-shell catalysts are also highlighted.

  1. Core design of the upgraded TREAT reactor

    Microsoft Academic Search

    D. C. Wade; S. K. Bhattacharyya; W. C. Lipinski; C. C. Stone

    1982-01-01

    The upgrading of the TREAT reactor involves the replacement of the central 11 x 11 subzone of the 19 x 19 fuel assembly array by new, Inconel-clad, high-temperature fuel assemblies, and the additions of a new reactor control system, a safety-grade plant protection system, and an enhanced reactor filtration\\/coolant system. The final design of these modifications will be completed in

  2. Neutronic analysis of pebble-bed cores with transuranics

    E-print Network

    Pritchard, Megan Leigh

    2009-05-15

    systems designs requires this growth in code systems to accommodate current and future research. Some of the major renovations to version 5.1 are in the following areas: square4 New ORIGEN-ARP libraries square4 New covariance libraries in TSUNAMI...

  3. Safety and core design of large liquid-metal cooled fast breeder reactors

    NASA Astrophysics Data System (ADS)

    Qvist, Staffan Alexander

    In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

  4. Are Neutron Stars with Crystalline Color-Superconducting Cores Relevant for the LIGO Experiment?

    SciTech Connect

    Haskell, B.; Andersson, N.; Jones, D. I.; Samuelsson, L. [School of Mathematics, University of Southampton, Southampton SO17 1BJ (United Kingdom)

    2007-12-07

    We estimate the maximal deformation that can be sustained by a rotating neutron star with a crystalline color-superconducting quark core. Our results suggest that current gravitational-wave data from the Laser Interferometer Gravitational-Wave Observatory have already reached the level where a detection would have been possible over a wide range of the poorly constrained QCD parameters. This leads to the nontrivial conclusion that compact objects do not contain maximally strained color crystalline cores drawn from this range of parameter space. We discuss the uncertainties associated with our simple model and how it can be improved in the future.

  5. Are neutron stars with crystalline colour superconducting cores interesting for the LIGO experiment?

    E-print Network

    B. Haskell; N. Andersson; D. I. Jones; L. Samuelsson

    2008-05-16

    We estimate the maximal deformation that can be sustained by a rotating neutron star with a crystalline colour superconducting quark core. Our results suggest that current gravitational-wave data from LIGO have already reached the level where a detection would have been possible over a wide range of the poorly constrained QCD parameters. This leads to the non-trivial conclusion that compact objects \\emph{do not} contain maximally strained colour crystalline cores drawn from this range of parameter space. We discuss the uncertainties associated with our simple model and how it can be improved in the future.

  6. China Spallation Neutron Source: Design, R&D, and outlook

    NASA Astrophysics Data System (ADS)

    Wei, Jie; Chen, Hesheng; Chen, Yanwei; Chen, Yuanbo; Chi, Yunlong; Deng, Changdong; Dong, Haiyi; Dong, Lan; Fang, Shouxian; Feng, Ji; Fu, Shinian; He, Lunhua; He, Wei; Heng, Yuekun; Huang, Kaixi; Jia, Xuejun; Kang, Wen; Kong, Xiangcheng; Li, Jian; Liang, Tianjiao; Lin, Guoping; Liu, Zhenan; Ouyang, Huafu; Qin, Qing; Qu, Huamin; Shi, Caitu; Sun, Hong; Tang, Jingyu; Tao, Juzhou; Wang, Chunhong; Wang, Fangwei; Wang, Dingsheng; Wang, Qingbin; Wang, Sheng; Wei, Tao; Xi, Jiwei; Xu, Taoguang; Xu, Zhongxiong; Yin, Wen; Yin, Xuejun; Zhang, Jing; Zhang, Zong; Zhang, Zonghua; Zhou, Min; Zhu, Tao

    2009-02-01

    The China Spallation Neutron Source (CSNS) is an accelerator based multidiscipline user facility planned to be constructed in Dongguan, Guangdong, China. The CSNS complex consists of an negative hydrogen linear accelerator, a rapid cycling proton synchrotron accelerating the beam to 1.6 GeV energy, a solid tungsten target station, and instruments for spallation neutron applications. The facility operates at 25 Hz repetition rate with an initial design beam power of 120 kW and is upgradeable to 500 kW. The primary challenge is to build a robust and reliable user's facility with upgrade potential at a fraction of "world standard" cost. We report the status, design, R&D, and upgrade outlook including applications using spallation neutron, muon, fast neutron, and proton, as well as related programs including medical therapy and accelerator-driven sub-critical reactor (ADS) programs for nuclear waste transmutation.

  7. Analysis of an MTC noise measurement performed in ringhals-2 using gamma-thermometers and in-core neutron detectors

    Microsoft Academic Search

    C. Demazière; I. Pázsit; T. Andersson; B. Severinsson; T. Ranman

    2003-01-01

    A noise measurement in the Swedish Ringhals-2 PWR was performed in January 2002 by using twelve gamma-thermometers and two in-core neutron detectors, all located on the same axial level in the reactor. The gamma-thermometers are very versatile tools since they allow estimating the core-averaged moderator temperature noise throughout the core. This core-averaged temperature noise was then used to estimate the

  8. Castilla-La Mancha neutron monitor: situation and design

    NASA Astrophysics Data System (ADS)

    Medina, Jose; José Blanco, Juan; García, Oscar

    2010-05-01

    This work shows the present status of the Castilla-La Mancha Neutron Monitor (CaLMa). Thanks to the founds recently agree for the Castilla-La Mancha Comunity governement. The new neutron monitor station will be building and placed in the Alcalá University campus in Guadalajara. Being operative in a provisional site at the end of 2010. The station is based on 18NM64 of LND 25373 3He detectors integrated into the Neutron Monitor Data Base (NMDB). Details about geomagnetic conditions, design and schedule are presented.

  9. Design calculations for the ANS (Advanced Neutron Source) cold source

    SciTech Connect

    Lillie, R.A.; Alsmiller, R.G. Jr.

    1988-01-01

    The calculation procedure, based on discrete ordinates transport methods, that is being used to carry out design calculations for the Advanced Neutron Source cold source is described. Calculated results on the gain in cold neutron flux produced by a liquid deuterium cold source are compared with experimental data and with calculated data previously obtained by P. Ageron et al., at the Institute Max von Laue-Paul Langevin in Grenoble, France. Calculated results are also presented that indicated how the flux of cold neutrons vary with cold source parameters. 23 refs., 5 figs., 3 tabs.

  10. Introduction to Neutron Coincidence Counter Design Based on Boron-10

    SciTech Connect

    Kouzes, Richard T.; Ely, James H.; Lintereur, Azaree T.; Siciliano, Edward R.

    2012-01-22

    The Department of Energy Office of Nonproliferation Policy (NA-241) is supporting the project 'Coincidence Counting With Boron-Based Alternative Neutron Detection Technology' at Pacific Northwest National Laboratory (PNNL) for development of an alternative neutron coincidence counter. The goal of this project is ultimately to design, build and demonstrate a boron-lined proportional tube based alternative system in the configuration of a coincidence counter. This report, providing background information for this project, is the deliverable under Task 1 of the project.

  11. Double-core evolution and the formation of neutron-star binaries with compact companions

    E-print Network

    J. D. M. Dewi; Ph. Podsiadlowski; A. Sena

    2006-02-23

    We present the results of a systematic exploration of an alternative evolutionary scenario to form double neutron-star binaries, first proposed by Brown (1995), which does not involve a neutron star passing through a common envelope. In this scenario, the initial binary components have very similar masses, and both components have left the main sequence before they evolve into contact; preferably the primary has already developed a CO core. We have performed population synthesis simulations to study the formation of double neutron star binaries via this channel and to predict the orbital properties and system velocities of such systems. We obtain a merger rate for DNSs in this channel in the range of 0.1 - 12/Myr. These rates are still subject to substantial uncertainties such as the modelling of the contact phase.

  12. Natural Nuclear Reactor Oklo and Variation of Fundamental Constants Part 1: Computation of Neutronics of Fresh Core

    E-print Network

    Yu. V. Petrov; A. I. Nazarov; M. S. Onegin; V. Yu. Petrov; E. G. Sakhnovsky

    2005-09-15

    Using modern methods of reactor physics we have performed full-scale calculations of the natural reactor Oklo. For reliability we have used recent version of two Monte Carlo codes: Russian code MCU REA and world wide known code MCNP (USA). Both codes produce similar results. We have constructed a computer model of the reactor Oklo zone RZ2 which takes into account all details of design and composition. The calculations were performed for three fresh cores with different uranium contents. Multiplication factors, reactivities and neutron fluxes were calculated. We have estimated also the temperature and void effects for the fresh core. As would be expected, we have found for the fresh core a significant difference between reactor and Maxwell spectra, which was used before for averaging cross sections in the Oklo reactor. The averaged cross section of Sm-149 and its dependence on the shift of resonance position (due to variation of fundamental constants) are significantly different from previous results. Contrary to results of some previous papers we find no evidence for the change of the fine structure constant in the past and obtain new, most accurate limits on its variation with time: -4 10^{-17}year^{-1} < d alpha/dt/alpha < 3 10^{-17} year^{-1} A further improvement in the accuracy of the limits can be achieved by taking account of the core burnup. These calculations are in progress.

  13. The Interplay between Proto--Neutron Star Convection and Neutrino Transport in Core-Collapse Supernovae

    Microsoft Academic Search

    A. Mezzacappa; A. C. Calder; S. W. Bruenn; J. M. Blondin; M. W. Guidry; M. R. Strayer; A. S. Umar

    1998-01-01

    We couple two-dimensional hydrodynamics to realistic one-dimensional multigroup flux-limited diffusion neutrino transport to investigate proto-neutron star convection in core-collapse supernovae, and more specifically, the interplay between its development and neutrino transport. Our initial conditions, time-dependent boundary conditions, and neutrino distributions for computing neutrino heating, cooling, and deleptonization rates are obtained from one-dimensional simulations that implement multigroup flux-limited diffusion and one-dimensional

  14. Core design of the upgraded TREAT reactor

    SciTech Connect

    Wade, D.C.; Bhattacharyya, S.K.; Lipinski, W.C.; Stone, C.C.

    1982-01-01

    The upgrading of the TREAT reactor involves the replacement of the central 11 x 11 subzone of the 19 x 19 fuel assembly array by new, Inconel-clad, high-temperature fuel assemblies, and the additions of a new reactor control system, a safety-grade plant protection system, and an enhanced reactor filtration/coolant system. The final design of these modifications will be completed in early 1983. The TREAT facility is scheduled to be shut down for modification in mid-1984, and should resume the safety test program in mid-1985. The upgrading will provide a capability to conduct fast reactor safety tests on clusters of up to 37 prototypic LMFBR pins.

  15. Quark deconfinement in neutron star cores and the ground state of neutral matter

    E-print Network

    Chang-Qun Ma; Chun-Yuan Gao

    2007-06-22

    Whether or not deconfined quark phase exists in neutron star cores and represents the ground state of neutral matter at moderate densities are open questions. We use two realistic effective quark models, the three-flavor Nambu-Jona-Lasinio model and the modified quark-meson coupling model, to describe the neutron star matter. After constructing possible hybrid equations of state (EOSes) with unpaired or color superconducting quark phase, we systematically discuss the observational constraints of neutron stars on the EOSes. It is found that the neutron star with pure quark matter core is unstable and the hadronic phase with hyperons is denied, while hybrid EOSes with two-flavor color superconducting phase or unpaired quark matter phase are both allowed by the tight and most reliable constraints from two stars Ter 5 I and EXO 0748-676. And the hybrid EOS with unpaired quark matter phase is allowed even compared with the tightest constraint from the most massive pulsar star PSR J0751+1807. Therefore, we conclude that the ground state of neutral matter at moderate densities is in deconfined quark phase likely.

  16. Neutronic analysis of pebble-bed cores with transuranics 

    E-print Network

    Pritchard, Megan Leigh

    2009-05-15

    ) of current nuclear reactors in U.S., the quantity of spent fuel will reach close to 87,000 tons. The constituents of the spent fuel include about 95 wt. % of uranium (with a U-235 enrichment comparable to that of natural uranium), about 1... and enrichment. I.C TECHNICAL STATUS OF THE QUESTION An important aspect of any reactor design is the fuel type, its utilization and resulting cycle. Until recently the fuel of choice has been uranium based, typically uranium dioxide. Use of low...

  17. Influence Of Low Boron Core Design On PWR Transient Behavior

    SciTech Connect

    Aleksandrov Papukchiev, Angel; Yubo Liu [Technical University Munich, Arcisstrasse 21, 80333 Muenchen (Germany); Schaefer, Anselm [ISaR Institute for Safety and Reliability, Walther-Meissner-Str. 2 85748 Garching (Germany)

    2006-07-01

    In conventional pressurized water reactor (PWR) designs, the concentration of boron in primary coolant is limited by the requirement of having a negative moderator density coefficient. As high boron concentrations have significant impact on reactivity feedback properties, design changes to reduce boron concentration in the reactor coolant are of general interest in view of improving PWR inherent safety. In the framework of an investigation into the feasibility of low boron design, a PWR core configuration based on fuel with higher gadolinium (Gd) content has been developed which permits to reduce the natural boron concentration at begin of cycle (BOC) by approx. 50% compared to current German PWR technology. For the assessment of the potential safety advantages, a Loss-of-Feedwater Anticipated Transient Without Scram (ATWS LOFW) has been simulated with the system code ATHLET for two PWR core designs: a low boron design and a standard core design. The most significant difference in the transient performance of both designs is the total primary fluid mass released through the pressurizer (PRZ) valves. It is reduced by a factor of four for the low boron reactor, indicating its improved density reactivity feedback. (authors)

  18. Implementation of an Arm Compatible Processor Core for SOC Designs

    Microsoft Academic Search

    Ahmed A. Morgan; M. E. Allam; M. A. Salama; H. A. K. Mansour

    2005-01-01

    Hardware description languages (HDLs) are commonly used to construct hardware systems. Reuse of the design is a common practice to improve the productivity nowadays. In this paper, an implementation of a fully pipelined ARM compatible processor core, which can be embedded into system-on-chips (SOCs) is presented. The implementation aims to support research, education, and development by opening the source codes.

  19. Two stochastic optimization algorithms applied to nuclear reactor core design

    Microsoft Academic Search

    Wagner F. Sacco; Cassiano R. E. de oliveira; Cláudio M. N. A. Pereira

    2006-01-01

    Two stochastic optimization algorithms conceptually similar to Simulated Annealing are presented and applied to a core design optimization problem previously solved with Genetic Algorithms. The two algorithms are the novel Particle Collision Algorithm (PCA), which is introduced in detail, and Dueck's Great Deluge Algorithm (GDA). The optimization problem consists in adjusting several reactor cell parameters, such as dimensions, enrichment and

  20. Advanced Core Design And Fuel Management For Pebble-Bed Reactors

    SciTech Connect

    Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry

    2004-10-01

    A method for designing and optimizing recirculating pebble-bed reactor cores is presented. At the heart of the method is a new reactor physics computer code, PEBBED, which accurately and efficiently computes the neutronic and material properties of the asymptotic (equilibrium) fuel cycle. This core state is shown to be unique for a given core geometry, power level, discharge burnup, and fuel circulation policy. Fuel circulation in the pebble-bed can be described in terms of a few well?defined parameters and expressed as a recirculation matrix. The implementation of a few heat?transfer relations suitable for high-temperature gas-cooled reactors allows for the rapid estimation of thermal properties critical for safe operation. Thus, modeling and design optimization of a given pebble-bed core can be performed quickly and efficiently via the manipulation of a limited number key parameters. Automation of the optimization process is achieved by manipulation of these parameters using a genetic algorithm. The end result is an economical, passively safe, proliferation-resistant nuclear power plant.

  1. Neutronics analyses in support of the conceptual design of the MAPS NTP reactor

    SciTech Connect

    Raepsaet, X.; Lenain, R. [Commissariat a l`Energie Atomique, DRN/DMT CEN Saclay, F-91191 Gif-sur-Yvette Cedex (France); Naury, S. [Altran Technologies, 58 Boulvard Gouvion Saint Cyr, 75858 PARIS Cedex 17 (France)

    1996-03-01

    Within the framework of the French nuclear thermal propulsion program called MAPS (Lenain 1996), several neutronics studies and analyses were performed. The aim was to determine the basic design features of a reactor based on the Pebble Bed Reactor concept (Powell 1985) and needing minimum technological developments. In the concern to further enhance the reactor safety posture and to maintain a minimum engine mass breakdown, a beryllium moderated/reflected reactor using highly enriched UO{sub 2} or UC{sub 2} as fuel has been designed with a mean hydrogen core outlet temperature of 2200 K (theoretical ISP of 859 s). The objective of this paper is to give a detailed neutronics analysis of the MAPS reactor. {copyright} {ital 1996 American Institute of Physics.}

  2. Neutron Tube Design Study for Boron Neutron Capture TherapyApplication

    SciTech Connect

    Verbeke, J.M.; Lee, Y.; Leung, K.N.; Vujic, J.; Williams, M.D.; Wu, L.K.; Zahir, N.

    1998-01-04

    Radio-frequency (RF) driven ion sources are being developed in Lawrence Berkeley National Laboratory (LBNL) for sealed-accelerator-tube neutron generator application. By using a 5-cm-diameter RF-driven multicusp source H{sup +} yields over 95% have been achieved. These experimental findings will enable one to develop compact neutron generators based on the D-D or D-T fusion reactions. In this new neutron generator, the ion source, the accelerator and the target are all housed in a sealed metal container without external pumping. Recent moderator design simulation studies have shown that 14 MeV neutrons could be moderated to therapeutically useful energy ranges for boron neutron capture therapy (BNCT). The dose near the center of the brain with optimized moderators is about 65% higher than the dose obtained from a typical neutron spectrum produced by the Brookhaven Medical Research Reactor (BMRR), and is comparable to the dose obtained by other accelerator-based neutron sources. With a 120 keV and 1 A deuteron beam, a treatment time of {approx}35 minutes is estimated for BNCT.

  3. Surface tension of the core-crust interface of neutron stars with global charge neutrality

    NASA Astrophysics Data System (ADS)

    Rueda, Jorge A.; Ruffini, Remo; Wu, Yuan-Bin; Xue, She-Sheng

    2014-03-01

    It has been shown recently that taking into account strong, weak, electromagnetic, and gravitational interactions, and fulfilling the global charge neutrality of the system, a transition layer will happen between the core and crust of neutron stars, at the nuclear saturation density. We use relativistic mean field theory together with the Thomas-Fermi approximation to study the detailed structure of this transition layer and calculate its surface and Coulomb energy. We find that the surface tension is proportional to a power-law function of the baryon number density in the core bulk region. We also analyze the influence of the electron component and the gravitational field on the structure of the transition layer and the value of the surface tension, to compare and contrast with known phenomenological results in nuclear physics. Based on the above results we study the instability against Bohr-Wheeler surface deformations in the case of neutron stars obeying global charge neutrality. Assuming the core-crust transition at nuclear density ?core?2.7×1014 g cm-3, we find that the instability sets the upper limit to the crust density, ?crustcrit?1.2×1014 g cm-3. This result implies a nonzero lower limit to the maximum electric field of the core-crust transition surface and makes inaccessible a limit of quasilocal charge neutrality in the limit ?crust=?core. The general framework presented here can be also applied to study the stability of sharp phase transitions in hybrid stars as well as in strange stars, both bare and with outer crust. The results of this work open the way to a more general analysis of the stability of these transition surfaces, accounting for other effects such as gravitational binding, centrifugal repulsion, magnetic field induced by rotating electric field, and therefore magnetic dipole-dipole interactions.

  4. Phenix Power Plant Decommissioning Project - Removal of Core and Neutron Blanket Components

    SciTech Connect

    Moitrier, C. [CEA /Marcoule DDCO/SDSP BP 17171 302078 Bagnols Sur Ceze (France)

    2008-01-15

    Phenix is a sodium-cooled fast neutron reactor located at the CEA's Rhone Valley Center where it was commissioned in 1974. It has an electric power rating of 250 MW and is operated jointly by the CEA and EDF. Its primary role today is to investigate the transmutation of long-lived radioactive waste into shorter-lived wasteform. Its final shutdown is scheduled for the beginning of 2009 in accordance with the reactor operating program with a mean availability of 80%. In this context the Phenix Power Plant Decommissioning Project was initiated in 2003. It covers the definitive cessation of plant operation and the dismantling (D and D) operations together with the final shutdown preparatory phase. The final shutdown phase corresponds to the initial operations undertaken to remove the maximum of hazardous materials. During this phase, subject to the standard operating methodological procedures, all the fuel in the plant will be removed and most of the other reactor core elements and other removable components (primary pumps, intermediate heat exchangers, control rod mechanisms, etc.) will be also dismantled and packaged in accordance with the available disposition routes. The core elements include fissile and fertile subassemblies, steel reflectors, and lateral neutron shielding rods. About half the lateral neutron shielding rods are accessible using the existing handling equipment; specific development work will be necessary to remove the others. Multiple engineering studies are in progress concerning the removal of the core elements. They address safety concerns taking into account thermal and seismic stresses, waste management including radiological characterization of the objects to be removed and the specification of suitable disposal routes (interim storage or disposal facilities), and scenario studies with the definition of each stage of the waste removal process. All these aspects are discussed in this paper. The feasibility of removing the core elements has been demonstrated with regard to handling and safety. A new handling arm is now being developed to remove the inaccessible core elements. The new earthquake safety calculations completed in 2006 did not reveal any insurmountable difficulties. Major work remains to characterize the core elements and determine the capacity of the disposition routes. The characterization program now in progress will validate the core element activation levels, but must be supplemented by a program to characterize the surface contamination from the primary sodium system. Concerning the disposition routes, several specific containers and shipping casks must be developed; engineering studies are now in progress to obtain the necessary approvals.

  5. Coupled 3D-neutronics / thermal-hydraulics analysis of an unprotected loss-of-flow accident for a 3600 MWth SFR core

    SciTech Connect

    Sun, K. [Paul Scherrer Institut PSI, 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne EPFL, 1015 Lausanne (Switzerland); Chenu, A. [Ecole Polytechnique Federale de Lausanne EPFL, 1015 Lausanne (Switzerland); Mikityuk, K.; Krepel, J. [Paul Scherrer Institut PSI, 5232 Villigen PSI (Switzerland); Chawla, R. [Paul Scherrer Institut PSI, 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne EPFL, 1015 Lausanne (Switzerland)

    2012-07-01

    The core behaviour of a large (3600 MWth) sodium-cooled fast reactor (SFR) is investigated in this paper with the use of a coupled TRACE/PARCS model. The SFR neutron spectrum is characterized by several performance advantages, but also leads to one dominating neutronics drawback - a positive sodium void reactivity. This implies a positive reactivity effect when sodium coolant is removed from the core. In order to evaluate such feedback in terms of the dynamics, a representative unprotected loss-of-flow (ULOF) transient, i.e. flow run-down without SCRAM in which sodium boiling occurs, is analyzed. Although analysis of a single transient cannot allow general conclusions to be drawn, it does allow better understanding of the underlying physics and can lead to proposals for improving the core response during such an accident. The starting point of this study is the reference core design considered in the framework of the Collaborative Project on the European Sodium Fast Reactor (CP-ESFR). To reduce the void effect, the core has been modified by introducing an upper sodium plenum (along with a boron layer) and by reducing the core height-to-diameter ratio. For the ULOF considered, a sharp increase in core power results in melting of the fuel in the case of the reference core. In the modified core, a large dryout leads to melting of the clad. It seems that, for the hypothetical event considered, fuel failure cannot be avoided with just improvement of the neutronics design; therefore, thermal-hydraulics optimization has been considered. An innovative assembly design is proposed to prevent sodium vapour blocking the fuel channel. This results in preventing a downward propagation of the sodium boiling to the core center, thus limiting it to the upper region. Such a void map introduces a negative coolant density reactivity feedback, which dominates the total reactivity change. As a result, the power level and the fuel temperature are effectively reduced, and a large dryout is prevented. As a next step, the classical sodium plenum is replaced by a fission gas plenum (with lower sodium fraction), thus improving flow stability. Stable boiling at a steady power level is achieved in this final configuration. (authors)

  6. Design of a boron neutron capture enhanced fast neutron therapy assembly

    SciTech Connect

    Wang, Zhonglu

    2006-08-01

    The use of boron neutron capture to boost tumor dose in fast neutron therapy has been investigated at several fast neutron therapy centers worldwide. This treatment is termed boron neutron capture enhanced fast neutron therapy (BNCEFNT). It is a combination of boron neutron capture therapy (BNCT) and fast neutron therapy (FNT). It is believed that BNCEFNT may be useful in the treatment of some radioresistant brain tumors, such as glioblastoma multiform (GBM). A boron neutron capture enhanced fast neutron therapy assembly has been designed for the Fermilab Neutron Therapy Facility (NTF). This assembly uses a tungsten filter and collimator near the patient's head, with a graphite reflector surrounding the head to significantly increase the dose due to boron neutron capture reactions. The assembly was designed using Monte Carlo radiation transport code MCNP version 5 for a standard 20x20 cm{sup 2} treatment beam. The calculated boron dose enhancement at 5.7-cm depth in a water-filled head phantom in the assembly with a 5x5 cm{sup 2} collimation was 21.9% per 100-ppm {sup 10}B for a 5.0-cm tungsten filter and 29.8% for a 8.5-cm tungsten filter. The corresponding dose rate for the 5.0-cm and 8.5-cm thick filters were 0.221 and 0.127 Gy/min, respectively; about 48.5% and 27.9% of the dose rate of the standard 10x10 cm{sup 2} fast neutron treatment beam. To validate the design calculations, a simplified BNCEFNT assembly was built using four lead bricks to form a 5x5 cm{sup 2} collimator. Five 1.0-cm thick 20x20 cm{sup 2} tungsten plates were used to obtain different filter thicknesses and graphite bricks/blocks were used to form a reflector. Measurements of the dose enhancement of the simplified assembly in a water-filled head phantom were performed using a pair of tissue-equivalent ion chambers. One of the ion chambers is loaded with 1000-ppm natural boron (184-ppm {sup 10}B) to measure dose due to boron neutron capture. The measured dose enhancement at 5.0-cm depth in the head phantom for the 5.0-cm thick tungsten filter is (16.6 {+-} 1.8)%, which agrees well with the MCNP simulation of the simplified BNCEFNT assembly, (16.4 {+-} 0.5)%. The error in the calculated dose enhancement only considers the statistical uncertainties. The total dose rate measured at 5.0-cm depth using the non-borated ion chamber is (0.765 {+-} 0.076) Gy/MU, about 61% of the fast neutron standard dose rate (1.255Gy/MU) at 5.0-cm depth for the standard 10x10 cm{sup 2} treatment beam. The increased doses to other organs due to the use of the BNCEFNT assembly were calculated using MCNP5 and a MIRD phantom. The activities of the activation products produced in the BNCEFNT assembly after neutron beam delivery were computed. The photon ambient dose rate due to the radioactive activation products was also estimated.

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

    SciTech Connect

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

    1998-05-01

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

  8. Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

    DOE PAGESBeta

    None

    2014-11-04

    Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental keff come from uncertainties in the manganese content and impurities in the stainless steel fuel cladding asmore »well as the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9?) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2? uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.« less

  9. Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

    SciTech Connect

    None

    2014-11-04

    Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental keff come from uncertainties in the manganese content and impurities in the stainless steel fuel cladding as well as the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9?) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2? uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

  10. Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

    DOE PAGESBeta

    None

    2014-11-04

    Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental keff come from uncertainties in the manganese content and impurities in the stainless steel fuel cladding as well as the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9?) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2? uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

  11. An integrated design of an accelerator-based neutron source for boron neutron capture therapy

    NASA Astrophysics Data System (ADS)

    Dobelbower, Michael Christian

    1997-07-01

    An Accelerator Based Neutron Source (ABNS) for Boron Neutron Capture Therapy (BNCT) was first proposed at The Ohio State University (OSU). Since the conception of the ABNS for BNCT, OSU has designed and optimized a moderator assembly based on in-air and in-phantom parameters. Additionally, the fabrication of the moderator assembly has commenced along with detailed analyses of the target and its heat removal system. In this dissertation, an integrated design of the ABNS is presented. This integrated design includes the high energy beam transport system (HEBT), the target and heat removal system (HRS), and the moderator assembly. In the integration process, a neutronic model of the HRS was developed and incorporated into the moderator assembly model. Additionally, a preliminary design of a HEBT system was developed that is compatible with both the HRS and the facility shielding. This dissertation also includes the completion of the fabrication of the moderator assembly and its experimental verification. The completion of the moderator assembly fabrication included the refabrication of the moderator and delimiter and the fabrication of the 6Li covering on the front of the moderator assembly. The experimental verification included neutron spectrum calculations and measurements in the irradiation port, and 3He detector response calculations and measurements in-phantom downstream of the moderator assembly.

  12. Physics design for the Brookhaven Medical Research Reactor epithermal neutron source.

    PubMed

    Wheeler, F J; Parsons, D K; Nigg, D W; Wessol, D E; Miller, L G; Fairchild, R G

    1990-01-01

    A collaborative effort by researchers at the Idaho National Engineering Laboratory and the Brookhaven National Laboratory has resulted in the design and implementation of an epithermal-neutron source at the Brookhaven Medical Research Reactor (BMRR). Large aluminum containers, filled with aluminum oxide tiles and aluminum spacers, were tailored to pre-existing compartments on the animal side of the reactor facility. A layer of cadmium was used to minimize the thermal-neutron component. Additional bismuth was added to the pre-existing bismuth shield to minimize the gamma component of the beam. Lead was also added to reduce gamma streaming around the bismuth. The physics design methods are outlined in this paper. Information available to date shows close agreement between calculated and measured beam parameters. The neutron spectrum is predominantly in the intermediate energy range (0.5 eV - 10 keV). The peak flux intensity is 6.4E + 12 n/(m2.s.MW) at the center of the beam on the outer surface of the final gamma shield. The corresponding neutron current is 3.8E + 12 n/(m2.s.MW). Presently, the core operates at a maximum of 3 MW. The fast-neutron KERMA is 3.6E-15 cGy/(n/m2) and the gamma KERMA is 5.0E-16 cGY/(n/m2) for the unperturbed beam. The neutron intensity falls off rapidly with distance from the outer shield and the thermal flux realized in phantom or tissue is strongly dependent on the beam-delimiter and target geometry. PMID:2268249

  13. Core polarization for the electric quadrupole moment of neutron-rich Aluminum isotopes

    E-print Network

    Kenichi Yoshida

    2009-02-18

    The core polarization effect for the electric quadrupole moment of the neutron-rich $^{31}$Al, $^{33}$Al and $^{35}$Al isotopes in the vicinity of the island of inversion are investigated by means of the microscopic particle-vibration coupling model in which the Skyrme Hartee-Fock-Bogoliubov and quasiparticle-random-phase approximation are used to calculate the single-quasiparticle wave functions and the excitation modes. It is found that the polarization charge for the proton $1d_{5/2}$ hole state in $^{33}$Al is quite sensitive to coupling to the neutrons in the $pf$-shell associated with the pairing correlations, and that the polarization charge in $^{35}$Al becomes larger due to the stronger collectivity of the low-lying quadrupole vibrational mode in the neighboring $^{36}$Si nucleus.

  14. Monte Carlo Neutronics and Thermal Hydraulics Analysis of Reactor Cores with Multilevel Grids

    NASA Astrophysics Data System (ADS)

    Bernnat, W.; Mattes, M.; Guilliard, N.; Lapins, J.; Zwermann, W.; Pasichnyk, I.; Velkov, K.

    2014-06-01

    Power reactors are composed of assemblies with fuel pin lattices or other repeated structures with several grid levels, which can be modeled in detail by Monte Carlo neutronics codes such as MCNP6 using corresponding lattice options, even for large cores. Except for fresh cores at beginning of life, there is a varying material distribution due to burnup in the different fuel pins. Additionally, for power states the fuel and moderator temperatures and moderator densities vary according to the power distribution and cooling conditions. Therefore, a coupling of the neutronics code with a thermal hydraulics code is necessary. Depending on the level of detail of the analysis, a very large number of cells with different materials and temperatures must be regarded. The assignment of different material properties to all elements of a multilevel grid is very elaborate and may exceed program limits if the standard input procedure is used. Therefore, an internal assignment is used which overrides uniform input parameters. The temperature dependency of continuous energy cross sections, probability tables for the unresolved resonance region and thermal neutron scattering laws is taken into account by interpolation, requiring only a limited number of data sets generated for different temperatures. The method is applied with MCNP6 and proven for several full core reactor models. For the coupling of MCNP6 with thermal hydraulics appropriate interfaces were developed for the GRS system code ATHLET for liquid coolant and the IKE thermal hydraulics code ATTICA-3D for gaseous coolant. Examples will be shown for different applications for PWRs with square and hexagonal lattices, fast reactors (SFR) with hexagonal lattices and HTRs with pebble bed and prismatic lattices.

  15. Neutronic design studies for an unattended, low power reactor

    NASA Astrophysics Data System (ADS)

    Palmer, R. G.; Durkee, J. W., Jr.

    The Los Alamos National Laboratory is involved in the design and demonstrations of a small, long-lived nuclear heat and electric power source for potential applications at remote sites where alternate fossil energy systems would not be cost effective. This paper describes the neutronic design analysis that was performed to arrive at two conceptual designs, one using thermoelectric conversion, the other using an organic Rankine cycle. To meet the design objectives and constraints a number of scoping and optimization studies were carried out. The results of calculations of control worths, temperature coefficients of reactivity and fuel depletion effects are reported.

  16. AIR-CORE FOIL-WOUND PULSE TRANSFORMER DESIGN CONCEPT

    Microsoft Academic Search

    K. A. Khan; R. G. Colclaser

    1993-01-01

    A high voltage output pulse can be provided by a pulse forming network (PFN) using a pulse transformer (IT) as the last output stage. The parameter values of the PFN\\/PT combination are constrained by the output pulse duration, rise time, and load impedance. An air-core foil-wound transformer design offers the advantages of light weight and quasi-uniform voltage distribution between windings,

  17. Hyper-heuristic applied to nuclear reactor core design

    NASA Astrophysics Data System (ADS)

    Domingos, R. P.; Platt, G. M.

    2013-02-01

    The design of nuclear reactors gives rises to a series of optimization problems because of the need for high efficiency, availability and maintenance of security levels. Gradient-based techniques and linear programming have been applied, as well as genetic algorithms and particle swarm optimization. The nonlinearity, multimodality and lack of knowledge about the problem domain makes de choice of suitable meta-heuristic models particularly challenging. In this work we solve the optimization problem of a nuclear reactor core design through the application of an optimal sequence of meta-heuritics created automatically. This combinatorial optimization model is known as hyper-heuristic.

  18. 5 MW pulsed spallation neutron source, Preconceptual design study

    SciTech Connect

    Not Available

    1994-06-01

    This report describes a self-consistent base line design for a 5 MW Pulsed Spallation Neutron Source (PSNS). It is intended to establish feasibility of design and as a basis for further expanded and detailed studies. It may also serve as a basis for establishing project cost (30% accuracy) in order to intercompare competing designs for a PSNS not only on the basis of technical feasibility and technical merit but also on the basis of projected total cost. The accelerator design considered here is based on the objective of a pulsed neutron source obtained by means of a pulsed proton beam with average beam power of 5 MW, in {approx} 1 {mu}sec pulses, operating at a repetition rate of 60 Hz. Two target stations are incorporated in the basic facility: one for operation at 10 Hz for long-wavelength instruments, and one operating at 50 Hz for instruments utilizing thermal neutrons. The design approach for the proton accelerator is to use a low energy linear accelerator (at 0.6 GeV), operating at 60 Hz, in tandem with two fast cycling booster synchrotrons (at 3.6 GeV), operating at 30 Hz. It is assumed here that considerations of cost and overall system reliability may favor the present design approach over the alternative approach pursued elsewhere, whereby use is made of a high energy linear accelerator in conjunction with a dc accumulation ring. With the knowledge that this alternative design is under active development, it was deliberately decided to favor here the low energy linac-fast cycling booster approach. Clearly, the present design, as developed here, must be carried to the full conceptual design stage in order to facilitate a meaningful technology and cost comparison with alternative designs.

  19. BEAM DUMP WINDOW DESIGN FOR THE SPALLATION NEUTRON SOURCE.

    SciTech Connect

    RAPARIA,D.RANK,J.MURDOCH,G.ET AL.

    2004-03-10

    The Spallation Neutron Source accelerator systems will provide a 1 GeV, 1.44 MW proton beam to a liquid mercury target for neutron production. Beam tuning dumps are provided at the end of the linac (the Linac Dump) and in the Ring-to-Target transport line (the Extraction Dump) [1]. Thin windows are required to separate the accelerator vacuum from the poor vacuum upstream of the beam dump. There are several challenging engineering issues that have been addressed in the window design. Namely, handling of the high local power density deposited by the stripped electrons from the H-beam accelerated in the linac, and the need for low-exposure removal and replacement of an activated window. The thermal design of the linac dump window is presented, as is the design of a vacuum clamp and mechanism that allows remote removal and replacement of the window.

  20. Preliminary design study of advanced multistage axial flow core compressors

    NASA Technical Reports Server (NTRS)

    Wisler, D. C.; Koch, C. C.; Smith, L. H., Jr.

    1977-01-01

    A preliminary design study was conducted to identify an advanced core compressor for use in new high-bypass-ratio turbofan engines to be introduced into commercial service in the 1980's. An evaluation of anticipated compressor and related component 1985 state-of-the-art technology was conducted. A parametric screening study covering a large number of compressor designs was conducted to determine the influence of the major compressor design features on efficiency, weight, cost, blade life, aircraft direct operating cost, and fuel usage. The trends observed in the parametric screening study were used to develop three high-efficiency, high-economic-payoff compressor designs. These three compressors were studied in greater detail to better evaluate their aerodynamic and mechanical feasibility.

  1. System design description for GCFR-core flow test loop

    SciTech Connect

    Huntley, W.R.; Grindell, A.G.

    1980-12-01

    The Core Flow Test Loop is a high-pressure, high-temperature, out-of-reactor helium circulation system that is being constructed to permit detailed study of the thermomechanical and thermal performance at prototypic steady-state and transient operating conditions of simulated segments of core assemblies for a GCFR Demonstration Plant, as designed by General Atomic Company. It will also permit the expermental verification of predictive analytical models of the GCFR core assemblies needed to reduce operational and safety uncertainties of the GCFR. Full-sized blanket assemblies and segments of fuel rod and control rod fuel assemblies will be simulated with test bundles of electrically powered fuel rod or blanket rod simulators. The loop will provide the steady-state and margin test requirements of bundle power and heat removal, and of helium coolant flow rate, pressure, and temperature for test bundles having up to 91 rods; these requirements set the maximum power, coolant helium flow, and thermal requirements for the loop. However, the size of the test vessel that contains the test bundles will be determined by the bundles that simulate a full-sized GCFR blanket assembly. The loop will also provide for power and coolant transients to simulate transient operation of GCFR core assemblies, including the capability for rapid helium depressurization to simulate the depressurization class of GCFR accidents. In addition, the loop can be used as an out-of-reactor test bed for characterizing in-reactor test bundle configurations.

  2. Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

    SciTech Connect

    Ragusa, Jean; Vierow, Karen

    2011-09-01

    The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzed advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.

  3. New approach to the design of core support structures for large LMFBR plants

    SciTech Connect

    Burelbach, J.P.; Kann, W.J.; Pan, Y.C.; Saiveau, J.G.; Seidensticker, R.W.

    1984-01-01

    The paper describes an innovative design concept for a LMFBR Core Support Structure. A hanging Core Support Structure is described and analyzed. The design offers inherent safety features, constructibility advantages, and potential cost reductions.

  4. Advanced Neutron Source: Plant Design Requirements. Revision 4

    SciTech Connect

    Not Available

    1990-07-01

    The Advanced Neutron Source will be a new world-class facility for research using hot, thermal, cold, and ultra-cold neutrons. The heart of the facility will be a 330-MW (fission), heavy-water cooled and heavy-water moderated reactor. The reactor will be housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides will fan out into a large guide hall, housing about 30 neutron research stations. Appropriate office, laboratory, and shop facilities will be included to provide a complete facility for users. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory early in the next decade. This PDR document defines the plant-level requirements for the design, construction, and operation of ANS. It also defines and provides input to the individual System Design Description (SDD) documents. Together, this PDR document and the set of SDD documents will define and control the baseline configuration of ANS.

  5. Isoscalar-vector interaction and hybrid quark core in massive neutron stars

    NASA Astrophysics Data System (ADS)

    Shao, G. Y.; Colonna, M.; Di Toro, M.; Liu, Y. X.; Liu, B.

    2013-05-01

    The hadron-quark phase transition in the core of massive neutron stars is studied with a newly constructed two-phase model. For nuclear matter, a nonlinear Walecka type model with general nucleon-meson and meson-meson couplings, recently calibrated by Steiner, Hemper and Fischer, is taken. For quark matter, a modified Polyakov-Nambu—Jona-Lasinio model, which gives consistent results with lattice QCD data, is used. Most importantly, we introduce an isoscalar-vector interaction in the description of quark matter, and we study its influence on the hadron-quark phase transition in the interior of massive neutron stars. With the constraints of neutron star observations, our calculation shows that the isoscalar-vector interaction between quarks is indispensable if massive hybrids star exist in the universe, and its strength determines the onset density of quark matter, as well as the mass-radius relations of hybrid stars. Furthermore, as a connection with heavy-ion-collision experiments we give some discussions about the strength of isoscalar-vector interaction and its effect on the signals of hadron-quark phase transition in heavy-ion collisions, in the energy range of the NICA at JINR-Dubna and FAIR at GSI-Darmstadt facilities.

  6. Seismic responses of a pool-type fast reactor with different core support designs

    SciTech Connect

    Wu, Ting-shu; Seidensticker, R.W. (Argonne National Lab., IL (USA))

    1989-01-01

    In designing the core support system for a pool-type fast reactor, there are many issues which must be considered in order to achieve an optimum and balanced design. These issues include safety, reliability, as well as costs. Several design options are possible to support the reactor core. Different core support options yield different frequency ranges and responses. Seismic responses of a large pool-type fast reactor incorporated with different core support designs have been investigated. 4 refs., 3 figs.

  7. Core and System Design of Reduced-Moderation Water Reactor with Passive Safety Features

    SciTech Connect

    Iwamura, Takamichi; Okubo, Tsutomu; Yonomoto, Taisuke [Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun, Ibaraki-ken, 319-1195 (Japan); Takeda, Renzo; Moriya, Kumiaki [Hitachi, Ltd. (Japan); Kanno, Minoru [The Japan Atomic Power Company (Japan)

    2002-07-01

    In order to ensure the sustainable energy supply in Japan, research and developments of reduced-moderation water reactor (RMWR) have been performed. The RMWR can attain the favorable characteristics such as high burn-up, long operation cycle, multiple recycling of plutonium and effective utilization of uranium resources, based on the matured LWR technologies. MOX fuel assemblies in the tight-lattice fuel rod arrangement are used to reduce the moderation of neutron, and hence, to increase the conversion ratio. The conceptual design has been accomplished for the small 330 MWe RMWR core with the discharge burn-up of 60 GWd/t and the operation cycle of 24 months, under the natural circulation cooling of the core. A breeding ratio of 1.01 and the negative void reactivity coefficient are simultaneously realized in the design. In the plant system design, the passive safety features are intended to be utilized mainly to improve the economy. At present, a hybrid one under the combination of the passive and the active components, and a fully passive one are proposed. The former has been evaluated to reduce the cost for the reactor components. (authors)

  8. Application of core structural design guidelines in conceptual fuel pin design. [LMFBR

    Microsoft Academic Search

    M. R. Patel; J. D. Stephen

    1979-01-01

    The paper describes an application of the Draft RDT Standards F9-7, -8, and -9 to conceptual design of Fast Breeder Reactor (FBR) fuel pins. The Standards are being developed to provide guidelines for structural analysis and design of the FBR core components which have limited ductility at high fluences and are not addressed by the prevalent codes. The development is

  9. Development of a neutronics calculation method for designing commercial type Japanese sodium-cooled fast reactor

    SciTech Connect

    Takeda, T.; Shimazu, Y.; Hibi, K.; Fujimura, K. [Research Inst. of Nuclear Engineering, Univ. of Fukui, 1cho-me 2gaiku 4, Kanawa-cho, Tsuruga-shi, Fukui 914-0055 (Japan)

    2012-07-01

    Under the R and D project to improve the modeling accuracy for the design of fast breeder reactors the authors are developing a neutronics calculation method for designing a large commercial type sodium- cooled fast reactor. The calculation method is established by taking into account the special features of the reactor such as the use of annular fuel pellet, inner duct tube in large fuel assemblies, large core. The Verification and Validation, and Uncertainty Qualification (V and V and UQ) of the calculation method is being performed by using measured data from the prototype FBR Monju. The results of this project will be used in the design and analysis of the commercial type demonstration FBR, known as the Japanese Sodium fast Reactor (JSFR). (authors)

  10. Transitions between turbulent and laminar superfluid vorticity states in the outer core of a neutron star

    E-print Network

    C. Peralta; A. Melatos; M. Giacobello; A. Ooi

    2006-07-08

    We investigate the global transition from a turbulent state of superfluid vorticity to a laminar state, and vice versa, in the outer core of a neutron star. By solving numerically the hydrodynamic Hall-Vinen-Bekarevich-Khalatnikov equations for a rotating superfluid in a differentially rotating spherical shell, we find that the meridional counterflow driven by Ekman pumping exceeds the Donnelly-Glaberson threshold throughout most of the outer core, exciting unstable Kelvin waves which disrupt the rectilinear vortex array, creating a vortex tangle. In the turbulent state, the torque exerted on the crust oscillates, and the crust-core coupling is weaker than in the laminar state. This leads to a new scenario for the rotational glitches observed in radio pulsars: a vortex tangle is sustained in the differentially rotating outer core by the meridional counterflow, a sudden spin-up event brings the crust and core into corotation, the vortex tangle relaxes back to a rectilinear vortex array, then the crust spins down electromagnetically until enough meridional counterflow builds up to reform a vortex tangle. The turbulent-laminar transition can occur uniformly or in patches; the associated time-scales are estimated from vortex filament theory. We calculate numerically the global structure of the flow with and without an inviscid superfluid component, for Hall-Vinen and Gorter-Mellink forms of the mutual friction. We also calculate the post-glitch evolution of the angular velocity of the crust and its time derivative, and compare the results with radio pulse timing data, predicting a correlation between glitch activity and Reynolds number.

  11. A neutronic feasibility study of the AP1000 design loaded with fully ceramic micro-encapsulated fuel

    SciTech Connect

    Liang, C.; Ji, W. [Department of Mechanical, Aerospace, and Nuclear Engineering Rensselaer, Polytechnic Institute, 110 8th Street, Troy, NY 12180 (United States)

    2013-07-01

    A neutronic feasibility study is performed to evaluate the utilization of fully ceramic microencapsulated (FCM) fuel in the AP1000 reactor design. The widely used Monte Carlo code MCNP is employed to perform the full core analysis at the beginning of cycle (BOC). Both the original AP1000 design and the modified design with the replacement of uranium dioxide fuel pellets with FCM fuel compacts are modeled and simulated for comparison. To retain the original excess reactivity, ranges of fuel particle packing fraction and fuel enrichment in the FCM fuel design are first determined. Within the determined ranges, the reactor control mechanism employed by the original design is directly used in the modified design and the utilization feasibility is evaluated. The worth of control of each type of fuel burnable absorber (discrete/integral fuel burnable absorbers and soluble boron in primary coolant) is calculated for each design and significant differences between the two designs are observed. Those differences are interpreted by the fundamental difference of the fuel form used in each design. Due to the usage of silicon carbide as the matrix material and the fuel particles fuel form in FCM fuel design, neutron slowing down capability is increased in the new design, leading to a much higher thermal spectrum than the original design. This results in different reactivity and fission power density distributions in each design. We conclude that a direct replacement of fuel pellets by the FCM fuel in the AP1000 cannot retain the original optimum reactor core performance. Necessary modifications of the core design should be done and the original control mechanism needs to be re-designed. (authors)

  12. Conceptual physics design of an epithermal-neutron facility for neutron capture therapy at the Georgia Tech research reactor

    SciTech Connect

    Nigg, D.W.; Wheeler, F.J. (Idaho National Engineering Lab., Idaho Falls (United States))

    1992-01-01

    The Idaho National Engineering Laboratory (INEL) is currently the focus for a comprehensive program directed toward the development of boron neutron capture therapy (BNCT) for certain types of refractory malignancies. One particular component of the INEL BNCT program involves a collaborative effort between the INEL and Georgia Institute of Technology to design an advanced epithermal- (neutrons in the energy range of 0.5 eV to 10 keV) neutron beam facility for BNCT research at the Georgia Tech research reactor (GTRR). The basic conceptual design and expected physics-related performance parameters for the GTRR epithermal-neutron beam are presented in this paper.

  13. Nuclear design of a vapor core reactor for space nuclear propulsion

    SciTech Connect

    Dugan, E.T.; Watanabe, Y.; Kuras, S.A.; Maya, I.; Diaz, N.J. (Innovative Nuclear Space Power and Propulsion Institute, University of Florida, Gainesville, Florida 32611 (United States))

    1993-01-15

    Neutronic analysis methodology and results are presented for the nuclear design of a vapor core reactor for space nuclear propulsion. The Nuclear Vapor Thermal Reactor (NVTR) Rocket Engine uses modified NERVA geometry and systems which the solid fuel replaced by uranium tetrafluoride vapor. The NVTR is an intermediate term gas core thermal rocket engine with specific impulse in the range of 1000--1200 seconds; a thrust of 75,000 lbs for a hydrogen flow rate of 30 kg/s; average core exit temperatures of 3100 K to 3400 K; and reactor thermal powers of 1400 to 1800 MW. Initial calculations were performed on epithermal NVTRs using ZrC fuel elements. Studies are now directed at thermal NVTRs that use fuel elements made of C-C composite. The large ZrC-moderated reactors resulted in thrust-to-weight ratios of only 1 to 2; the compact C-C composite systems yield thrust-to-weight ratios of 3 to 5.

  14. Nuclear design of a vapor core reactor for space nuclear propulsion

    NASA Astrophysics Data System (ADS)

    Dugan, Edward T.; Watanabe, Yoichi; Kuras, Stephen A.; Maya, Isaac; Diaz, Nils J.

    1993-01-01

    Neutronic analysis methodology and results are presented for the nuclear design of a vapor core reactor for space nuclear propulsion. The Nuclear Vapor Thermal Reactor (NVTR) Rocket Engine uses modified NERVA geometry and systems which the solid fuel replaced by uranium tetrafluoride vapor. The NVTR is an intermediate term gas core thermal rocket engine with specific impulse in the range of 1000-1200 seconds; a thrust of 75,000 lbs for a hydrogen flow rate of 30 kg/s; average core exit temperatures of 3100 K to 3400 K; and reactor thermal powers of 1400 to 1800 MW. Initial calculations were performed on epithermal NVTRs using ZrC fuel elements. Studies are now directed at thermal NVTRs that use fuel elements made of C-C composite. The large ZrC-moderated reactors resulted in thrust-to-weight ratios of only 1 to 2; the compact C-C composite systems yield thrust-to-weight ratios of 3 to 5.

  15. Benchmark Evaluation of the Neutron Radiography (NRAD) Reactor Upgraded LEU-Fueled Core

    SciTech Connect

    John D. Bess

    2001-09-01

    Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. The final upgraded core configuration with 64 fuel elements has been completed. Evaluated benchmark measurement data include criticality, control-rod worth measurements, shutdown margin, and excess reactivity. Dominant uncertainties in keff include the manganese content and impurities contained within the stainless steel cladding of the fuel and the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 nuclear data are approximately 1.4% greater than the benchmark model eigenvalue, supporting contemporary research regarding errors in the cross section data necessary to simulate TRIGA-type reactors. Uncertainties in reactivity effects measurements are estimated to be ~10% with calculations in agreement with benchmark experiment values within 2s. The completed benchmark evaluation de-tails are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Experiments (IRPhEP Handbook). Evaluation of the NRAD LEU cores containing 56, 60, and 62 fuel elements have also been completed, including analysis of their respective reactivity effects measurements; they are also available in the IRPhEP Handbook but will not be included in this summary paper.

  16. Impact of the symmetry energy on nuclear pasta phases and crust-core transition in neutron stars

    NASA Astrophysics Data System (ADS)

    Bao, S. S.; Shen, H.

    2015-01-01

    We study the impact of the symmetry energy on properties of nuclear pasta phases and crust-core transition in neutron stars. We perform a self-consistent Thomas-Fermi calculation employing the relativistic mean-field model. The properties of pasta phases presented in the inner crust of neutron stars are investigated and the crust-core transition is examined. It is found that the slope of the symmetry energy plays an important role in determining the pasta phase structure and the crust-core transition. The correlation between the symmetry energy slope and the crust-core transition density obtained in the Thomas-Fermi approximation is consistent with that predicted by the liquid-drop model.

  17. Design of a californium-based epithermal neutron beam for neutron capture therapy.

    PubMed

    Yanch, J C; Kim, J K; Wilson, M J

    1993-08-01

    The potential of the spontaneously fissioning isotope, 252Cf, to provide epithermal neutrons for use in boron neutron capture therapy (BNCT) has been investigated using Monte Carlo simulation. The Monte Carlo code MCNP was used to design an assembly composed of a 26 cm long, 11 cm radius cylindrical D2O moderator followed by a 64 cm long Al filter. Lithium filters are placed between the moderator and the filter and between the Al and the patient. A reflector surrounding the moderator/filter assembly is required in order to maintain adequate therapy flux at the patient position. An ellipsoidal phantom composed of skull- and brain-equivalent material was used to determine the dosimetric effect of this beam. It was found that both advantage depths and advantage ratios compare very favourably with reactor and accelerator epithermal neutron sources. The dose rate obtainable, on the other hand, is 4.1 RBE cGy min-1, based on a very large (1.0 g) source of 252Cf. This dose rate is two to five times lower than those provided by existing reactor beams and can be viewed as a drawback of using 252Cf as a neutron source. Radioisotope sources, however, do offer the advantage of in-hospital installation. PMID:8367525

  18. Spatial and spectral characteristics of a compact system neutron beam designed for BNCT facility

    Microsoft Academic Search

    J. Ghassoun; B. Chkillou; A. Jehouani

    2009-01-01

    The development of suitable neutron sources and neutron beam is critical to the success of Boron Neutron Capture Therapy (BNCT). In this work a compact system designed for BNCT is presented. The system consists of 252Cf fission neutron source and a moderator\\/reflector\\/filter\\/shield assembly. The moderator\\/reflector\\/filter arrangement has been optimized to maximize the epithermal neutron component which is useful for BNCT

  19. A three-dimensional neutronics-thermohydraulics simulation of core disruptive accident in sodium-cooled fast reactor

    Microsoft Academic Search

    Hidemasa Yamano; Yoshiharu Tobita; Satoshi Fujita

    2009-01-01

    The SIMMER-III code is a two-dimensional, multi-velocity-field, multi-phase, multi-component, Eulerian, fluid-dynamics code coupled with a fuel-pin model and a space- and energy-dependent neutron transport kinetics model. Since the three-dimensional representation of the core enables realistic distribution of the materials constituting the core, including control rods, SIMMER-IV has been developed as a direct extension of SIMMER-III to three dimensions with retaining

  20. A Study on 8-18Be Isotopes Used on Neutron Multiplier in Reactor Design

    NASA Astrophysics Data System (ADS)

    Tel, E.; Akt?, N. N.; Okuducu, ?.; Aydin, A.; ?ahan, M.; U?ur, F. A.; ?ahan, H.

    2011-02-01

    Neutronic characterization and development of structural materials, neutron multiplier materials, tritium breeders are primarily important for fusion and hybrid reactors. In order to improve neutron economy, beryllium, lead, bismuth, zirconium are considered and used as neutron multiplier materials in fusion and hybrid reactor design. In this study, rms charge radii, neutron radii, mass radii and neutron skin thickness were calculated for 8-18Be isotopes nuclei. The neutron and proton density are calculated for Be isotopes. The results obtained were compared with the experimental and theoretical results of other researchers by using Hartree-Fock method with an effective interaction with Skyrme forces.

  1. Neutron stars with hyperon cores: stellar radii and EOS near nuclear density

    E-print Network

    M. Fortin; J. L. Zdunik; P. Haensel; M. Bejger

    2015-05-23

    The existence of 2 Msun pulsars puts very strong constraints on the equation of state (EOS) of neutron stars (NSs) with hyperon cores, which can be satisfied only by special models of hadronic matter. The radius-mass relation for these models is sufficiently specific that it could be subjected to an observational test with future X-ray observatories. We want to study the impact of the presence of hyperon cores on the radius-mass relation for NS. We aim to find out how, and for which particular stellar mass range, a specific relation R(M), where M is the gravitational mass, and R is the circumferential radius, is associated with the presence of a hyperon core. We consider a set of 14 theoretical EOS of dense matter, based on the relativistic mean-field (RMF) approximation, allowing for the presence of hyperons in NSs. We seek correlations between R(M) and the stiffness of the EOS below the hyperon threshold needed to pass the 2 Msun test. For NS masses 1.013km, because of a very stiff pre-hyperon segment of the EOS. At nuclear density, the pressure is significantly higher than a robust upper bound obtained recently using chiral effective field theory. If massive NSs do have a sizable hyperon core, then according to current models the radii for M=1.0-1.6 Msun are necessarily >13km. If, on the contrary, a NS with a radius Rtest, present EOS allowing for hyperons that fulfill condition M_max>2 Msun yield a pressure at nuclear density that is too high relative to up-to-date microscopic calculations of this quantity.

  2. Monte Carlo Calculation of Core Reactivity and Fluxes for the Development of the BNCT Neutron Source at the Kyiv Research Reactor

    Microsoft Academic Search

    Olena Gritzay; Oleksandr Kalchenko; Nataliya Klimova; Volodymyr Razbudey; Andriy Sanzhur; Stephen Binney

    2005-01-01

    The presented results show our consecutive steps in developing a neutron source with parameters required by Boron Neutron Capture Therapy (BNCT) at the Kyiv Research Reactor (KRR). The main goal of this work was to analyze the influence of installation of different types of uranium converters close to the reactor core on neutron beam characteristics and on level of reactor

  3. Early Cosmic Evolution of Europium from Core Collapse Supernovae and/or Neutron Star Mergers

    E-print Network

    Vangioni, E; Daigne, F; Francois, P; Belczynski, K

    2015-01-01

    The rapid neutron-capture process is known to be of fundamental importance for explaining the origin of approximately half of the A > 60 stable nuclei observed in nature. Despite important efforts, the astrophysical site of the r process remains unidentified. The two most promising astrophysical sites of the r process, namely Core Collapse SuperNovae (CCSN) and Neutron Star Mergers (NSM) are considered in the context of the early cosmic chemical evolution through the origin and evolution of a typical r process element, Eu. The Eu abundance in very low metallicity stars is used to shed light on the possible CCSN and NSM contributions in the early Universe. Predictions are made here using a hierarchical model for structure formation for which a special attention is paid to a proper description of the stellar formation rate. Eu yields from NSM are taken from recent nucleosynthesis calculations. Observations of Eu in ultra metal poor stars are considered to constrain the model. We find that the bulk of Eu observa...

  4. Inelastic Neutron Scattering on Exchange-Biased Co/CoO Core-Shell Nanoparticles

    NASA Astrophysics Data System (ADS)

    Strycker, Glenn; Inderhees, Sue; Aronson, Meigan; Qiu, Yiming; Borchers, Julie

    2006-03-01

    We report results of inelastic neutron scattering on exchange biased Co/CoO core-shell nanoparticles. Data were taken using time-of-flight techniques at the Disk Chopper Spectrometer (DCS) at the NIST Center for Neutron Research, which allows observation of the dynamics of magnetic spin reversal over a range of energies and length scales. Above the blocking temperature (TB) the scattering is quasi-elastic, with an amplitude that peaks at the (.5ex1 -.1em/ -.15em.25ex2 .5ex1 -.1em/ -.15em.25ex2 .5ex1 -.1em/ -.15em.25ex2 ) anti-ferromagnetic CoO ordering wave vector. With decreasing temperature the quasi-elastic scattering narrows, consistent with the freezing of longitudinal moment fluctuations, and becomes resolution limited near TB. Below TB we observe a spectrum of inelastic excitations arising from a log-normal distribution of energy barriers. We will discuss in detail the length scale and temperature dependences of these features. Work at the University of Michigan performed under the auspices of the Department of Energy.

  5. Core design of long life-cycle fast reactors operating without reactivity margin

    SciTech Connect

    Aristova, E. N.; Baydin, D. F.; Gol'din, V. Y.; Pestryakova, G. A.; Stoynov, M. I. [Keldysh Inst. of Applied Mathematics RAS, Miusskaya sq., bld.4, 125047, Moscow (Russian Federation)

    2012-07-01

    In this paper we consider a possibility of designing a fast reactor core that operates without reactivity margin for a long time. This study is based on the physical principle of fast reactor operating in a self-adjustable neutron-nuclear regime (SANNR-1) introduced by L.P. Feoktistov (1988-1993) and improved by V. Ya. Gol'din SANNR-2 (1995). The mathematical modeling of active zones of fast reactors in SANNR modes is held by authors since 1992. The numerical simulation is based on solving the neutron transport equation coupled with quasi-diffusion equations. The calculations have been performed using standard 26 energy groups. We use a hierarchy of spatial models of 1D, 1.5D, 2D, and 3D geometries. The spatial models of higher dimensionality are used for verification of results. The calculations showed that operation of the reactor in this mode increases its efficiency, safety and simplifies management. It is possible to achieve continuous work of the reactor in SANNR-2 during 7-10 years without fuel overloads by means of further optimization of the mode. Small reactivity margin is used only for the reactor start up. After first 10-15 days the reactor in SANNR-2 operates without reactivity margin. (authors)

  6. Effects of lumbar artificial disc design on intervertebral mobility: in vivo comparison between mobile-core and fixed-core.

    PubMed

    Delécrin, Joël; Allain, Jérôme; Beaurain, Jacques; Steib, Jean-Paul; Huppert, Jean; Chataigner, Hervé; Ameil, Marc; Aubourg, Lucie; Nguyen, Jean-Michel

    2012-06-01

    Although in theory, the differences in design between fixed-core and mobile-core prostheses should influence motion restoration, in vivo kinematic differences linked with prosthesis design remained unclear. The aim of this study was to investigate the rationale that the mobile-core design seems more likely to restore physiological motion since the translation of the core could help to mimic the kinematic effects of the natural nucleus. In vivo intervertebral motion characteristics of levels implanted with the mobile-core prosthesis were compared with untreated levels of the same population, levels treated by a fixed-core prosthesis, and normal levels (data from literature). Patients had a single-level implantation at L4L5 or L5S1 including 72 levels with a mobile-core prosthesis and 33 levels with a fixed-core prosthesis. Intervertebral mobility characteristics included the range of motion (ROM), the motion distribution between flexion and extension, the prosthesis core translation (CT), and the intervertebral translation (VT). A method adapted to the implanted segments was developed to measure the VT: metal landmarks were used instead of the bony landmarks. The reliability assessment of the VT measurement method showed no difference between three observers (p < 0.001), a high level of agreement (ICC = 0.908) and an interobserver precision of 0.2 mm. Based on this accurate method, this in vivo study demonstrated that the mobile-core prosthesis replicated physiological VT at L4L5 levels but not at L5S1 levels, and that the fixed-core prosthesis did not replicate physiological VT at any level by increasing VT. As the VT decreased when the CT increased (p < 0.001) it was proven that the core mobility minimized the VT. Furthermore, some physiologic mechanical behaviors seemed to be maintained: the VT was higher at implanted the L4L5 level than at the implanted L5S1 level, and the CT appeared lower at the L4L5 level than at the L5S1 level. ROM and motion distribution were not different between the mobile-core prosthesis and the fixed-core prosthesis implanted levels. This study validated in vivo the concept that a mobile-core helps to restore some physiological mechanical characteristics of the VT at the implanted L4L5 level, but also showed that the minimizing effect of core mobility on the VT was not sufficient at the L5S1 level. PMID:21153595

  7. Chapter 5 Embedded Core Test Fundamentals 1 Design-for-Test for Digital IC's and Embedded Core Systems Alfred L. Crouch

    E-print Network

    Greenwood, Garrison W.

    Chapter 5 Embedded Core Test Fundamentals 1 Design-for-Test for Digital IC's and Embedded Core Access #12;Chapter 5 Embedded Core Test Fundamentals 2 Design-for-Test for Digital IC's and Embedded Core with Inserted Test Gate-Level Netlist with Mixed Test FIRM Layout GDSII with No Test Layout with Test from

  8. Neutronics assessment of stringer fuel assembly designs for the liquid-salt-cooled very high temperature reactor (LS-VHTR).

    SciTech Connect

    Szakaly, F. J.; Kim, T. K.; Taiwo, T. A.; Nuclear Engineering Division

    2007-01-01

    Neutronic studies of 18-pin and 36-pin stringer fuel assemblies have been performed to ascertain that core design requirements for the Liquid-Salt Cooled Very High Temperature Reactor (LS-VHTR) can be met. Parametric studies were performed to determine core characteristics required to achieve a target core cycle length of 18 months and fuel discharge burnup greater than 100 GWd/t under the constraint that the uranium enrichment be less than 20% in order to support non-proliferation goals. The studies were done using the WIMS9 lattice code and the linear reactivity model to estimate the core reactivity balance, fuel composition, and discharge burnup. The results show that the design goals can be met using a 1-batch fuel management scheme, uranium enrichment of 15% and a fuel packing fraction of 30% or greater for the 36-pin stringer fuel assembly design.Evaluations of a liquid-salt- (molten-salt-) cooled version of the prismatic-block type VHTR, the LS-VHTR, are ongoing at U.S. national laboratories, universities, and industry. These evaluations have included core and passive safety studies and balance of plant conceptual designs.

  9. THERMAL: A routine designed to calculate neutron thermal scattering

    SciTech Connect

    Cullen, D.E.

    1995-02-24

    THERMAL is designed to calculate neutron thermal scattering that is isotropic in the center of mass system. At low energy thermal motion will be included. At high energies the target nuclei are assumed to be stationary. The point of transition between low and high energies has been defined to insure a smooth transition. It is assumed that at low energy the elastic cross section is constant in the center of mass system. At high energy the cross section can be of any form. You can use this routine for all energies where the elastic scattering is isotropic in the center of mass system. In most materials this will be a fairly high energy.

  10. Design study for an advanced liquid-metal fast breeder reactor core with a high burnup

    Microsoft Academic Search

    T. Inagaki; H. Kuga; M. Suzuki; T. Yokoyama; M. Yamaoka; K. Kaneto; M. Ohashi; K. Kurihara

    1989-01-01

    Design studies are performed for a commercial liquid-metal fast breeder reactor core that can achieve a burnup of 200 GWd\\/t. A plutonium-type asymmetric parfait core with two different plutonium-enriched zones in the axial direction as well as in the radial direction is studied. This core concept solves core design problems related to high burnup, and it is possible to achieve

  11. New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions.

    PubMed

    Geslot, B; Vermeeren, L; Filliatre, P; Lopez, A Legrand; Barbot, L; Jammes, C; Bréaud, S; Oriol, L; Villard, J-F

    2011-03-01

    Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 × 10(20) n?cm(2). A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement. PMID:21456734

  12. New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

    NASA Astrophysics Data System (ADS)

    Geslot, B.; Vermeeren, L.; Filliatre, P.; Lopez, A. Legrand; Barbot, L.; Jammes, C.; Bréaud, S.; Oriol, L.; Villard, J.-F.

    2011-03-01

    Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 × 1020 n/cm2. A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

  13. Facing Challenges for Monte Carlo Analysis of Full PWR Cores : Towards Optimal Detail Level for Coupled Neutronics and Proper Diffusion Data for Nodal Kinetics

    NASA Astrophysics Data System (ADS)

    Nuttin, A.; Capellan, N.; David, S.; Doligez, X.; El Mhari, C.; Méplan, O.

    2014-06-01

    Safety analysis of innovative reactor designs requires three dimensional modeling to ensure a sufficiently realistic description, starting from steady state. Actual Monte Carlo (MC) neutron transport codes are suitable candidates to simulate large complex geometries, with eventual innovative fuel. But if local values such as power densities over small regions are needed, reliable results get more difficult to obtain within an acceptable computation time. In this scope, NEA has proposed a performance test of full PWR core calculations based on Monte Carlo neutron transport, which we have used to define an optimal detail level for convergence of steady state coupled neutronics. Coupling between MCNP for neutronics and the subchannel code COBRA for thermal-hydraulics has been performed using the C++ tool MURE, developed for about ten years at LPSC and IPNO. In parallel with this study and within the same MURE framework, a simplified code of nodal kinetics based on two-group and few-point diffusion equations has been developed and validated on a typical CANDU LOCA. Methods for the computation of necessary diffusion data have been defined and applied to NU (Nat. U) and Th fuel CANDU after assembly evolutions by MURE. Simplicity of CANDU LOCA model has made possible a comparison of these two fuel behaviours during such a transient.

  14. Simulation of in-core neutron noise measurements for axial void profile reconstruction in boiling water reactors

    SciTech Connect

    Dykin, V.; Pazsit, I. [Chalmers Univ. of Technology, Div. of Nuclear Engineering, Dept. of Applied Physics, SE-412 96 Gothenburg (Sweden)

    2012-07-01

    A possibility to reconstruct the axial void profile from the simulated in-core neutron noise which is caused by density fluctuations in a Boiling Water Reactor (BWR) heated channel is considered. For this purpose, a self-contained model of the two-phase flow regime is constructed which has quantitatively and qualitatively similar properties to those observed in real BWRs. The model is subsequently used to simulate the signals of neutron detectors induced by the corresponding perturbations in the flow density. The bubbles are generated randomly in both space and time using Monte-Carlo techniques. The axial distribution of the bubble production is chosen such that the mean axial void fraction and void velocity follow the actual values of BWRs. The induced neutron noise signals are calculated and then processed by the standard signal analysis methods such as Auto-Power Spectral Density (APSD) and Cross-Power Spectral Density (CPSD). Two methods for axial void and velocity profiles reconstruction are discussed: the first one is based on the change of the break frequency of the neutron auto-power spectrum with axial core elevation, while the second refers to the estimation of transit times of propagating steam fluctuations between different axial detector positions. This paper summarizes the principles of the model and presents a numerical testing of the qualitative applicability to estimate the required parameters for the reconstruction of the void fraction profile from the neutron noise measurements. (authors)

  15. Placement and Routing for Non-Rectangular Embedded Programmable Logic Cores in SoC Design

    E-print Network

    Wilton, Steve

    Placement and Routing for Non-Rectangular Embedded Programmable Logic Cores in SoC Design Tony Wong Vancouver, B.C., Canada Abstract As SoC design enters into mainstream usage, the ability to make post. These cores are like any other IP in the SoC design methodology, except that their function can be changed

  16. Media digital signal processor core design for multimedia application

    NASA Astrophysics Data System (ADS)

    Liu, Peng; Yu, Guo-jun; Cai, Wei-guang; Yao, Qing-dong

    2006-02-01

    An embedded single media processor named MediaDSP3200 core fabricated in a six-layer metal 0.18um CMOS process which implemented the RISC instruction set, DSP data processing instruction set and single-instruction-multiple-data (SIMD) multimedia-enhanced instruction set is described. MediaDSP3200 fuses RISC architecture and DSP computation capability thoroughly, which achieves RISC fundamental, DSP extended and single instruction multiple data (SIMD) instruction set with various addressing modes in a unified pipeline stage architecture. These characteristics enhance system digital signal processing performance greatly. The test processor can achieve 32x32-bit multiply-accumulate (MAC) of 320 MOPS, with 16x16-bit MAC of 1280MOPS. The test processor dissipates 600mW at 1.8v, 320MHz. Also, the implementation was primarily standard cell logic design style. MediaDSP3200 targets diverse embedded application systems, which need both powerful processing/control capability and low-cost budget, e.g. set-top-boxes, video conferencing, DTV, etc. MediaDSP3200 instruction set architecture, addressing mode, pipeline design, SIMD feature, split-ALU and MAC are described in this paper. Finally, the performance benchmark based on H.264 and MPEG decoder algorithm are given in this paper.

  17. Core excitation contributions to the breakup of the one-neutron halo nucleus {sup 11}Be on a proton

    SciTech Connect

    Crespo, R. [Centro de Fisica Nuclear, Universidade de Lisboa, Av. Prof. Gama Pinto 2, PT-1649-003 Lisboa (Portugal); Departamento de Fisica, Instituto Superior Tecnico, Taguspark, Av. Prof. Cavaco Silva, Taguspark, PT-2780-990 Porto Salvo, Oeiras (Portugal); Deltuva, A. [Centro de Fisica Nuclear, Universidade de Lisboa, Av. Prof. Gama Pinto 2, PT-1649-003 Lisboa (Portugal); Moro, A. M. [Departamento de Fisica Atomica, Molecular e Nuclear, Universidad Seville, Universidad de Sevilla, Apartado 1065, ES-41080 Sevilla (Spain)

    2011-04-15

    The effect of the core excitation in the breakup of a one-neutron halo nucleus is studied within two different reaction formalisms, namely, the core excited model and the single-scattering approximation of the three-body Faddeev-Alt-Grassberger-Sandhas equations with target-core potential allowing for the core excitation. As an example, we consider the breakup of {sup 11}Be on a proton target at 63.7 MeV/nucleon incident energy and calculate the semi-inclusive cross section in the excitation energy interval E{sub x}=3.0-5.5 MeV (E{sub rel}=2.5-5 MeV) containing the 3/2{sup +} resonance with dominant contribution of the {sup 10}Be(2{sup +}) core excited state. The effect of the core excitation to the breakup cross section integrated around this resonance is found to be very significant. Moreover, when resonant and nonresonant contributions are added, the resulting semi-inclusive cross section is in reasonable agreement with the existing data, demonstrating the relevance of the core excitation mechanism for this observable. The present calculations also show the importance of incorporating the energy dependence of the core-target transition operators in the reaction formalism.

  18. Verification of the CENTRM Module for Adaptation of the SCALE Code to NGNP Prismatic and PBR Core Designs

    SciTech Connect

    Ganapol, Barry; Maldonado, Ivan

    2014-01-23

    The generation of multigroup cross sections lies at the heart of the very high temperature reactor (VHTR) core design, whether the prismatic (block) or pebble-bed type. The design process, generally performed in three steps, is quite involved and its execution is crucial to proper reactor physics analyses. The primary purpose of this project is to develop the CENTRM cross-section processing module of the SCALE code package for application to prismatic or pebble-bed core designs. The team will include a detailed outline of the entire processing procedure for application of CENTRM in a final report complete with demonstration. In addition, they will conduct a thorough verification of the CENTRM code, which has yet to be performed. The tasks for this project are to: Thoroughly test the panel algorithm for neutron slowing down; Develop the panel algorithm for multi-materials; Establish a multigroup convergence 1D transport acceleration algorithm in the panel formalism; Verify CENTRM in 1D plane geometry; Create and test the corresponding transport/panel algorithm in spherical and cylindrical geometries; and, Apply the verified CENTRM code to current VHTR core design configurations for an infinite lattice, including assessing effectiveness of Dancoff corrections to simulate TRISO particle heterogeneity.

  19. Feasibility study on nuclear core design for soluble boron free small modular reactor

    NASA Astrophysics Data System (ADS)

    Rabir, Mohamad Hairie; Hah, Chang Joo; Ju, Cho Sung

    2015-04-01

    A feasibility study on nuclear core design of soluble boron free (SBF) core for small size (150MWth) small modular reactor (SMR) was investigated. The purpose of this study was to design a once through cycle SMR core, where it can be used to supply electricity to a remote isolated area. PWR fuel assembly design with 17×17 arrangement, with 264 fuel rods per assembly was adopted as the basis design. The computer code CASMO-3/MASTER was used for the search of SBF core and fuel assembly analysis for SMR design. A low critical boron concentration (CBC) below 200 ppm core with 4.7 years once through cycle length was achieved using 57 fuel assemblies having 170 cm of active height. Core reactivity controlled using mainly 512 number of 4 wt% and 960 12 wt% Gd rods.

  20. Design and implementation of a high performance matrix multiplier core for Xilinx Virtex FPGAs

    Microsoft Academic Search

    S. Belkacemi; K. Benkrid; D. Crookes; A. Benkrid

    2003-01-01

    Matrix multiplication is a core operation in digital signal processing operations with a variety of applications such as image processing, computer graphics, sonar processing and robotics. This paper presents the design and implementation of a high performance, fully parallel matrix multiplication core. The core is parameterised and scalable in terms of the matrices' dimensions (row and column number) and the

  1. Using Electronic Meeting System Support in the Design of the Graduate Core Curriculum.

    ERIC Educational Resources Information Center

    Spuck, Dennis W.; And Others

    1995-01-01

    Discusses uses of an electronic meeting system (EMS) to aid in designing graduate core curricula. Presents results from a study of a seven-member task force that used EMS for seven months to facilitate selection of core courses for master's degree programs. A diagram illustrates the task force's graduate core review process and an appendix…

  2. A new paradigm for local-global coupling in whole-core neutron transport.

    SciTech Connect

    Lewis, E.; Smith, M.; Palmiotti, G,; Nuclear Engineering Division; Northwestern Univ.; INL

    2009-01-01

    A new paradigm that increases the efficiency of whole-core neutron transport calculations without lattice homogenization is introduced. Quasi-reflected interface conditions are formulated to partially decouple periodic lattice effects from global flux gradients. The starting point is the finite subelement form of the variational nodal code VARIANT that eliminates fuel-coolant homogenization through the use of heterogeneous nodes. The interface spherical harmonics expansions that couple pin-cell-sized nodes are divided into low-order and high-order terms, and reflected interface conditions are applied to the high-order terms. Combined with an integral transport method within the node, the new approach dramatically reduces both the formation time and the dimensions of the nodal response matrices and leads to sharply reduced memory requirements and computational time. The method is applied to the two-dimensional C5G7 problem, an Organisation for Economic Co-operation and Development/Nuclear Energy Agency pressurized water reactor benchmark containing mixed oxide (MOX) and UO{sub 2} fuel assemblies, as well as to a three-dimensional MOX fuel assembly. Results indicate the new approach results in very little loss of accuracy relative to the corresponding full spherical harmonics expansions while reducing computational times by well over an order of magnitude.

  3. The role of neutron star mergers and core collapse supernovae in r process nucleosynthesis

    NASA Astrophysics Data System (ADS)

    Daigne, F.; Vangioni, E.

    2014-12-01

    Recent IR/optical/UV observations and Gamma-ray burst rate determinations at high redshift have led to significant progress in establishing the cosmic evolution of the star formation rate density (SFRD). The SFRD is then used to predict the ionization history of the Universe, and the evolution of the cosmic chemical abundances, supernova rates, etc, as a function of the redshift z. These predictions are done in the framework of the hierarchical model for structure formation. In this context, we focus here our attention on the origin and evolution of a typical r process element: Europium, in two possible sites: core collapse supernovae (SNII) or Neutron Star Mergers (NSM). In the first scenario, there is only one parameter, the yield of Eu produced in these SNII. In the second one, there are three physical parameters, Eu yield, binary star fraction and time delay before the merger. The comparison of our results with available observations of Eu in stars at various metallicities strongly favors the NSM site for the r process. In addition, it allows to put a constraint on the time delay for mergers, which is typically 0.1-0.2 Gyr, and to make an independent prediction for the expected rate of mergers in the horizon of the adv Virgo/Ligo detectors, which we find typically to be of the order of 3 to 10 events per year for NS/NS and NS/BH mergers respectively.

  4. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    SciTech Connect

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2014-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  5. Design of low-loss and highly birefringent hollow-core photonic crystal fiber

    Microsoft Academic Search

    P. J. Roberts; D. P. Williams; H. Sabert; B. J. Mangan; D. M. Bird; T. A. Birks; J. C. Knight; P. St. J. Russell

    2006-01-01

    A practical hollow-core photonic crystal fiber design suitable for attaining low-loss propagation is analyzed. The geometry involves a number of localized elliptical features positioned on the glass ring that surrounds the air core and separates the core and cladding regions. The size of each feature is tuned so that the composite core-surround geometry is antiresonant within the cladding band gap,

  6. Error Assessment of Homogenized Cross Sections Generation for Whole Core Neutronic Calculation

    SciTech Connect

    Hursin, Mathieu; Kochunas, Brendan; Downar, Thomas J. [University of California at Berkeley, 3115 Etcheverry Hall, Berkeley CA (United States)

    2007-10-26

    The objective of the work here was to assess the errors introduced by using 2D, few group homogenized cross sections to perform neutronic analysis of BWR problems with significant axial heterogeneities. The 3D method of characteristics code DeCART is used to generate 2-group assembly homogenized cross sections first using a conventional 2D lattice model and then using a full 3D solution of the assembly. A single BWR fuel assembly model based on an advanced BWR lattice design is used with a typical void distribution applied to the fuel channel coolant. This model is validated against an MCNP model. A comparison of the cross sections is performed for the assembly homogenized planar cross sections from the DeCART 3D and DeCART 2D solutions.

  7. Neutronics design of the INEL (Idaho National Engineering Laboratory) facility for boron neutron capture therapy clinical trials

    Microsoft Academic Search

    D. K. Parsons; F. J. Wheeler; B. L. Rushton; D. W. Nigg

    1988-01-01

    The PBF reactor at INEL has been redesigned for BNCT treatment of Gliobiastoma Multiforme. Analysis indicates that the design goals of 1.0E + 10 n\\/cm²{center dot}s epithermal neutron flux at the beam port can be met without exceeding the design goals of 2.6E -11 J cGy\\/(n\\/cm²) for the fast neutron KERMA and 2.0E-11 cGy\\/(n\\/cm²) for the gamma KERMA. These design

  8. Validation of the neutron and gamma fields in the JSI TRIGA reactor using in-core fission and ionization chambers.

    PubMed

    Žerovnik, Gašper; Kaiba, Tanja; Radulovi?, Vladimir; Jazbec, Anže; Rupnik, Sebastjan; Barbot, Loïc; Fourmentel, Damien; Snoj, Luka

    2015-02-01

    CEA developed fission chambers and ionization chambers were utilized at the JSI TRIGA reactor to measure neutron and gamma fields. The measured axial fission rate distributions in the reactor core are generally in good agreement with the calculated values using the Monte Carlo model of the reactor thus verifying both the computational model and the fission chambers. In future, multiple absolutely calibrated fission chambers could be used for more accurate online reactor thermal power monitoring. PMID:25479432

  9. Shielding design of a treatment room for an accelerator-based neutron source for BNCT

    Microsoft Academic Search

    J. F. Evans; T. E. Blue

    1995-01-01

    For several years, research has been ongoing in the Ohio State University (OSU) Nuclear Engineering Program toward the development of an accelerator-based irradiation facility (ANIF) neutron source for boron neutron capture therapy (BNCT). The ANIF, which is planned to be built in a hospital, has been conceptually designed and analyzed. After Qu, an OSU researcher, determined that the shielding design

  10. Design and Simulation of a Boron-loaded Neutron Spectrometer

    E-print Network

    Martin, Thomas

    2012-10-19

    The measurement of the distribution of kinetic energy carried by neutron particles is of interest to the health physics and radiation protection industry. Neutron particle spectral fluence is essential to the calculation of absorbed dose, equivalent...

  11. A Low-Overhead Design for Testability and Test Generation Technique for Core-Based Systems

    Microsoft Academic Search

    Indradeep Ghoshf; Niraj K. Jhat; Sujit Dey

    1997-01-01

    In a fundamental paradigm shift in system design, entire systems are being built on a single chip, using multiple embedded cores. Though the newest system design methodology has several advantages in terms of time-to-market and system cost, testing such core-based systems is difficult due to the problem of justifying test sequences at the inputs of a core embedded deep in

  12. Programmable logic IP cores in SoC design: opportunities and challenges

    Microsoft Academic Search

    Steven J. E. Wilton; Resve Saleh

    2001-01-01

    As SoC design enters into mainstream usage, the ability to make post-fabrication changes will become more and more attractive. This ability can be realized using programmable logic cores. These cores are like any other IP in the SoC design methodology, except that their function can be changed after fabrication. This paper outlines ways in which programmable logic cores can simplify

  13. HW\\/SW co-design for multi-core system on ESL virtual platform

    Microsoft Academic Search

    I-Yao Chuang; Tso-Yi Fan; Chi-Hung Lin; Chun-Nan Liu; Jen-Chieh Yeh

    2011-01-01

    Multi-core system and the associated software parallelization techniques have become one of the major trends of SoC design. A multi-core system with high hardware efficiency and software parallelism has the potential of achieving higher system performance and lower power consumption. This paper reveals how system performance prediction and analysis for multi-core system can be done at early design stage before

  14. Direct access test scheme-design of block and core cells for embedded ASICs

    Microsoft Academic Search

    V. Immaneni; S. Raman

    1990-01-01

    Intel requires the use of a direct-access test scheme in embedded-core or block-based ASIC (application-specific integrated-circuit) designs. This scheme provides for separate testing of individual block or core cells using proven test vectors. The authors discuss the design modifications for block cells with low pin counts, user application blocks, and large cores with high pin counts. The implementation and verification

  15. Conceptual Design of a Modular Island Core Fast Breeder Reactor \\

    Microsoft Academic Search

    Mitsuru KAMBE

    2002-01-01

    A metal fueled modular island core sodium cooled fast breeder reactor concept RAPID-M to improve reactor per- formance and proliferation resistance and to accommodate various power requirements has been demonstrated. The essential feature of the RAPID-M concept is that the reactor core consists of integrated fuel assemblies (IFAs) instead of conventional fuel subassemblies. The RAPID concept enables quick and simplified

  16. DESIGN AND OPERATION OF A WIRELINE RETRIEVABLE MOTOR DRIVEN CORE BARREL

    E-print Network

    DESIGN AND OPERATION OF A WIRELINE RETRIEVABLE MOTOR DRIVEN CORE BARREL OCEAN DRILLING PROGRAM) Ocean Research Institute of the University of Tokyo (Japan) National Science Foundation (United States of a Motor-Driven Core Barrel 9 1.0 Introduction 9 1.1 Conceptual Design 9 Chapter 2: NCB1: Development, Land

  17. Thermal Design of an Ultrahigh Temperature Vapor Core Reactor Combined Cycle Nuclear Power Plant

    Microsoft Academic Search

    Samuel E. Bays; Samim Anghaie; Blair Smith; Travis Knight

    2004-01-01

    Current work modeling high temperature compact heat exchangers may demonstrate the design feasibility of a Vapor Core Reactor (VCR) driven combined cycle power plant. For solid nuclear fuel designs, the cycle efficiency is typically limited by a metallurgical temperature limit which is dictated by fuel and structural melting points. In a vapor core, the gas\\/vapor phase nuclear fuel is uniformly

  18. Performance of truss panels with kagome cores and design of a high authority shape morphing structure

    Microsoft Academic Search

    Ju Wang

    2005-01-01

    This dissertation includes two parts: First, the performance of a light weight truss panels with Kagome cores; Second, design of a high authority morphing structure for hinging and twisting. The performance characteristics of a truss core sandwich panel design based on the 3D Kagome are measured and compared with earlier numerical simulations and the consistency is demonstrated. Panels are fabricated

  19. Conceptual design loss-of-coolant accident analysis for the Advanced Neutron Source reactor

    SciTech Connect

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr. (Oak Ridge National Lab., TN (United States))

    1994-01-01

    A RELAP5 system model for the Advanced Neutron Source Reactor has been developed for performing conceptual safety analysis report calculations. To better represent thermal-hydraulic behavior of the core, three specific changes in the RELAP5 computer code were implemented: a turbulent forced-convection heat transfer correlation, a critical heat flux (CHF) correlation, and an interfacial drag correlation. The model consists of the core region, the heat exchanger loop region, and the pressurizing/letdown system region. Results for three loss-of-coolant accident analyses are presented: (1) an instantaneous double-ended guillotine (DEG) core outlet break with a cavitating venturi installed downstream of the core, (b) a core pressure boundary tube outer wall rupture, and (c) a DEG core inlet break with a finite break-formation time. The results show that the core can survive without exceeding the flow excursion of CHF thermal limits at a 95% probability level if the proper mitigation options are provided.

  20. Monte Carlo Calculation of Core Reactivity and Fluxes for the Development of the BNCT Neutron Source at the Kyiv Research Reactor

    NASA Astrophysics Data System (ADS)

    Gritzay, Olena; Kalchenko, Oleksandr; Klimova, Nataliya; Razbudey, Volodymyr; Sanzhur, Andriy; Binney, Stephen

    2005-05-01

    The presented results show our consecutive steps in developing a neutron source with parameters required by Boron Neutron Capture Therapy (BNCT) at the Kyiv Research Reactor (KRR). The main goal of this work was to analyze the influence of installation of different types of uranium converters close to the reactor core on neutron beam characteristics and on level of reactor safety. The general Monte Carlo radiation transport code MCNP, version 4B, has been used for these calculations.

  1. Monte Carlo Calculation of Core Reactivity and Fluxes for the Development of the BNCT Neutron Source at the Kyiv Research Reactor

    SciTech Connect

    Gritzay, Olena; Kalchenko, Oleksandr; Klimova, Nataliya; Razbudey, Volodymyr; Sanzhur, Andriy [Institute for Nuclear Research, Prospekt Nauky 47, Kyiv, 03680 (Ukraine); Binney, Stephen [Oregon State University, Corvallis, OR 97331-5902 (United States)

    2005-05-24

    The presented results show our consecutive steps in developing a neutron source with parameters required by Boron Neutron Capture Therapy (BNCT) at the Kyiv Research Reactor (KRR). The main goal of this work was to analyze the influence of installation of different types of uranium converters close to the reactor core on neutron beam characteristics and on level of reactor safety. The general Monte Carlo radiation transport code MCNP, version 4B, has been used for these calculations.

  2. Charged-partricle and neutron-capture processes in the high-entropy wind of core-collapse supernovae.

    SciTech Connect

    Farouqi, K.; Kratz, K.-L.; Pfeiffer, B.; Rauscher, T.; Thielemann, F.-K.; Truran, J. W.; Physics; Univ. of Chicago; Joint Inst. for Nuclear Astrophysics; Univ. Mainz; Virtual Inst. for Nuclear Structure and Astrophysics; Max-Planck-Inst. fur Chemie; Univ. of Basel

    2010-04-01

    The astrophysical site of the r-process is still uncertain, and a full exploration of the systematics of this process in terms of its dependence on nuclear properties from stability to the neutron drip-line within realistic stellar environments has still to be undertaken. Sufficiently high neutron-to-seed ratios can only be obtained either in very neutron-rich low-entropy environments or moderately neutron-rich high-entropy environments, related to neutron star mergers (or jets of neutron star matter) and the high-entropy wind of core-collapse supernova explosions. As chemical evolution models seem to disfavor neutron star mergers, we focus here on high-entropy environments characterized by entropy S, electron abundance Y{sub e}, and expansion velocity V{sub exp}. We investigate the termination point of charged-particle reactions, and we define a maximum entropy S{sub final} for a given V{sub exp} and Y{sub e}, beyond which the seed production of heavy elements fails due to the very small matter density. We then investigate whether an r-process subsequent to the charged-particle freeze-out can in principle be understood on the basis of the classical approach, which assumes a chemical equilibrium between neutron captures and photodisintegrations, possibly followed by a {beta}-flow equilibrium. In particular, we illustrate how long such a chemical equilibrium approximation holds, how the freeze-out from such conditions affects the abundance pattern, and which role the late capture of neutrons originating from {beta}-delayed neutron emission can play. Furthermore, we analyze the impact of nuclear properties from different theoretical mass models on the final abundances after these late freeze-out phases and {beta}-decays back to stability. As only a superposition of astrophysical conditions can provide a good fit to the solar r-abundances, the question remains how such superpositions are attained, resulting in the apparently robust r-process pattern observed in low metallicity stars.

  3. CHARGED-PARTICLE AND NEUTRON-CAPTURE PROCESSES IN THE HIGH-ENTROPY WIND OF CORE-COLLAPSE SUPERNOVAE

    SciTech Connect

    Farouqi, K.; Truran, J. W. [Department of Astrophysics and Astronomy, University of Chicago, Chicago, IL 60637 (United States); Kratz, K.-L. [HGF Virtuelles Institut fuer Kernstruktur und Nukleare Astrophysik, Universitaet Mainz, D-55128 Mainz (Germany); Pfeiffer, B. [Institut fuer Kernchemie, Universitaet Mainz, D-55128 Mainz (Germany); Rauscher, T.; Thielemann, F.-K., E-mail: farouqi@uchicago.ed, E-mail: truran@nova.uchicago.ed, E-mail: BPfeiffe@uni-mainz.d, E-mail: k-l.Kratz@mpic.d, E-mail: Thomas.Rauscher@unibas.c, E-mail: F-K.Thielemann@unibas.c [Department of Physics, University of Basel, 4056 Basel (Switzerland)

    2010-04-01

    The astrophysical site of the r-process is still uncertain, and a full exploration of the systematics of this process in terms of its dependence on nuclear properties from stability to the neutron drip-line within realistic stellar environments has still to be undertaken. Sufficiently high neutron-to-seed ratios can only be obtained either in very neutron-rich low-entropy environments or moderately neutron-rich high-entropy environments, related to neutron star mergers (or jets of neutron star matter) and the high-entropy wind of core-collapse supernova explosions. As chemical evolution models seem to disfavor neutron star mergers, we focus here on high-entropy environments characterized by entropy S, electron abundance Y{sub e} , and expansion velocity V{sub exp}. We investigate the termination point of charged-particle reactions, and we define a maximum entropy S{sub final} for a given V{sub exp} and Y{sub e} , beyond which the seed production of heavy elements fails due to the very small matter density. We then investigate whether an r-process subsequent to the charged-particle freeze-out can in principle be understood on the basis of the classical approach, which assumes a chemical equilibrium between neutron captures and photodisintegrations, possibly followed by a beta-flow equilibrium. In particular, we illustrate how long such a chemical equilibrium approximation holds, how the freeze-out from such conditions affects the abundance pattern, and which role the late capture of neutrons originating from beta-delayed neutron emission can play. Furthermore, we analyze the impact of nuclear properties from different theoretical mass models on the final abundances after these late freeze-out phases and beta-decays back to stability. As only a superposition of astrophysical conditions can provide a good fit to the solar r-abundances, the question remains how such superpositions are attained, resulting in the apparently robust r-process pattern observed in low metallicity stars.

  4. Design and simulation of neutron radiography system based on 241Am-Be source

    NASA Astrophysics Data System (ADS)

    Jafari, H.; Feghhi, S. A. H.

    2012-05-01

    Neutron imaging is extended rapidly as a means of non-destructive testing (NDT) of materials. Various effective parameters on the image quality are needed to be studied for neutron radiography system with good resolution. In the present study a portable system of neutron radiography has been designed using 241Am-Be neutron source. The effective collimator parameters were calculated to obtain relatively pure, collimated and uniform neutron beam. All simulations were carried out in two stages using MCNPX Monte Carlo code. In the first stage, different collimator configurations were investigated and the appropriate design was selected based on maximum intensity and uniformity of neutron flux at the image plane in the outlet of collimator. Then, the overall system including source, collimator and sample was simulated for achieving radiographic images of standard samples. Normalized thermal neutron fluence of 2.61×10-5 cm-2 per source particle with n/? ratio of 1.92×105 cm-2 ?Sv-1 could be obtained at beam port of the designed collimator. Quality of images was assessed for two standard samples, using radiographic imaging capability in MCNPX. The collimated neutron beam in the designed system could be useful in a transportable exposure module for neutron radiography application.

  5. Current directions in core-shell nanoparticle design

    Microsoft Academic Search

    Wolfgang Schärtl

    2010-01-01

    Ten years ago I wrote a review about the important field of core-shell nanoparticles, focussing mainly on our own work about tracer systems, and briefly addressing polymer-coated nanoparticles as fillers for homogeneous polymer-colloid composites. Since then, the potential use of core-shell nanoparticles as multifunctional sensors or potential smart drug-delivery vehicles in biology and medicine has gained more and more importance,

  6. Designing a minimum-functionality neutron and gamma measurement instrument with a focus on authentication

    SciTech Connect

    Karpius, Peter J [Los Alamos National Laboratory; Williams, Richard B [Los Alamos National Laboratory

    2009-01-01

    During the design and construction of the Next-Generation Attribute-Measurement System, which included a largely commercial off-the-shelf (COTS), nondestructive assay (NDA) system, we realized that commercial NDA equipment tends to include numerous features that are not required for an attribute-measurement system. Authentication of the hardware, firmware, and software in these instruments is still required, even for those features not used in this application. However, such a process adds to the complexity, cost, and time required for authentication. To avoid these added authenticat ion difficulties, we began to design NDA systems capable of performing neutron multiplicity and gamma-ray spectrometry measurements by using simplified hardware and software that avoids unused features and complexity. This paper discusses one possible approach to this design: A hardware-centric system that attempts to perform signal analysis as much as possible in the hardware. Simpler processors and minimal firmware are used because computational requirements are kept to a bare minimum. By hard-coding the majority of the device's operational parameters, we could cull large sections of flexible, configurable hardware and software found in COTS instruments, thus yielding a functional core that is more straightforward to authenticate.

  7. Identifying and Using ‘Core Competencies’ to Help Design and Assess Undergraduate Neuroscience Curricula

    PubMed Central

    Kerchner, Michael; Hardwick, Jean C.; Thornton, Janice E.

    2012-01-01

    There has been a growing emphasis on the use of core competencies to design and inform curricula. Based on our Faculty for Undergraduate Neuroscience workshop at Pomona we developed a set of neuroscience core competencies. Following the workshop, faculty members were asked to complete an online survey to determine which core competencies are considered most essential and the results are presented. Backward Design principles are then described and we discuss how core competencies, through a backward design process, can be used to design and assess an undergraduate neuroscience curriculum. Oberlin College is used as a case study to describe the use of core competencies to help develop learning objectives, activities, and assessment measures for an undergraduate neuroscience major. PMID:23494749

  8. Neutron micro-beam design simulation by Monte Carlo

    NASA Astrophysics Data System (ADS)

    Pazirandeh, Ali; Taheri, Ali

    2007-09-01

    Over the last two decades neutron micro-beam has increasingly been developing in view of various applications in molecular activation analysis, micro-radiography in space and aviation and in radiation induced bystander effects in bio-cells. In this paper the structure and simulation of a neutron micro-beam is presented. The collimator for micro-beam is made of a polyethylene cylinder with a small hole along the centerline of the cylinder. The hole is filled with very thin needles in triangular or rectangular arrangement. The neutron source was reactor neutrons or a spontaneous Cf-252 neutron source falling on the top side of the collimator. The outgoing thermal and epithermal neutron fluxes were calculated.

  9. Design and Implementation of a Videotext Extractor on Dual-Core Platform

    Microsoft Academic Search

    Chih-Lun Fang; Tsung-Han Tsai; Ren-Chih Kuo

    2008-01-01

    Many videotexts exist in TV programs. Some videotexts provide valuable information. Thus, an efficient design to extract these videotexts is requested. Existing videotext extractors work on the PC platform and they are difficult to achieve real-time extraction and integration. Therefore, this work designs a videotext extractor on a dual-core platform. A distributed design framework for a dual-core platform is proposed.

  10. Optimization of burnable poison design for Pu incineration in fully fertile free PWR core

    SciTech Connect

    Fridman, E.; Shwageraus, E.; Galperin, A. [Dept. of Nuclear Engineering, Ben-Gurion Univ. of the Negev, Beer-Sheva (Israel)

    2006-07-01

    The design challenges of the fertile-free based fuel (FFF) can be addressed by careful and elaborate use of burnable poisons (BP). Practical fully FFF core design for PWR reactor has been reported in the past [1]. However, the burnable poison option used in the design resulted in significant end of cycle reactivity penalty due to incomplete BP depletion. Consequently, excessive Pu loading were required to maintain the target fuel cycle length, which in turn decreased the Pu burning efficiency. A systematic evaluation of commercially available BP materials in all configurations currently used in PWRs is the main objective of this work. The BP materials considered are Boron, Gd, Er, and Hf. The BP geometries were based on Wet Annular Burnable Absorber (WABA), Integral Fuel Burnable Absorber (IFBA), and Homogeneous poison/fuel mixtures. Several most promising combinations of BP designs were selected for the full core 3D simulation. All major core performance parameters for the analyzed cases are very close to those of a standard PWR with conventional UO{sub 2} fuel including possibility of reactivity control, power peaking factors, and cycle length. The MTC of all FFF cores was found at the full power conditions at all times and very close to that of the UO{sub 2} core. The Doppler coefficient of the FFF cores is also negative but somewhat lower in magnitude compared to UO{sub 2} core. The soluble boron worth of the FFF cores was calculated to be lower than that of the UO{sub 2} core by about a factor of two, which still allows the core reactivity control with acceptable soluble boron concentrations. The main conclusion of this work is that judicial application of burnable poisons for fertile free fuel has a potential to produce a core design with performance characteristics close to those of the reference PWR core with conventional UO{sub 2} fuel. (authors)

  11. Double core evolution. 7: The infall of a neutron star through the envelope of its massive star companion

    NASA Technical Reports Server (NTRS)

    Terman, James L.; Taam, Ronald E.; Hernquist, Lars

    1995-01-01

    Binary systems with properties similar to those of high-mass X-ray binaries are evolved through the common envelope phase. Three-dimensional simulations show that the timescale of the infall phase of the neutron star depends upon the evolutionary state of its massive companion. We find that tidal torques more effectively accelerate common envelope evolution for companions in their late core helium-burning stage and that the infall phase is rapid (approximately several initial orbital periods). For less evolved companions the decay of the orbit is longer; however, once the neutron star is deeply embedded within the companion's envelope the timescale for orbital decay decreases rapidly. As the neutron star encounters the high-density region surrounding the helium core of its massive companion, the rate of energy loss from the orbit increases dramatically leading to either partial or nearly total envelope ejection. The outcome of the common envelope phase depends upon the structure of the evolved companion. In particular, it is found that the entire common envelope can be ejected by the interaction of the neutron star with a red supergiant companion in binaries with orbital periods similar to those of long-period Be X-ray binaries. For orbital periods greater than or approximately equal to 0.8-2 yr (for companions of mass 12-24 solar mass) it is likely that a binary will survive the common envelope phase. For these systems, the structure of the progenitor star is characterized by a steep density gradient above the helium core, and the common envelope phase ends with a spin up of the envelope to within 50%-60% of corotation and with a slow mass outflow. The efficiency of mass ejection is found to be approximately 30%-40%. For less evolved companions, there is insufficient energy in the orbit to unbind the common envelope and only a fraction of it is ejected. Since the timescale for orbital decay is always shorter than the mass-loss timescale from the common envelope, the two cores will likely merge to form a Thorne-Zytkow object. Implications for the origin of Cyg X-3, an X-ray source consisting of a Wolf-Rayet star and a compact companion, and for the fate of the remnant binary consisting of a helium star and a neutron star are briefly discussed.

  12. Design of neutron beams at the Argonne Continuous Wave Linac (ACWL) for boron neutron capture therapy and neutron radiography

    SciTech Connect

    Zhou, X.L. [Massachusetts Inst. of Tech., Cambridge, MA (United States). Dept. of Nuclear Engineering; McMichael, G.E. [Argonne National Lab., Idaho Falls, ID (United States). Technology Development Div.

    1994-10-01

    Neutron beams are designed for capture therapy based on p-Li and p-Sc reactions using the Argonne Continuous Wave Linac (ACWL). The p-Li beam will provide a 2.5 {times} 10{sup 9} n/cm{sup 2}s epithermal flux with 7 {times} 10{sup 5} {gamma}/cm{sup 2}s contamination. On a human brain phantom, this beam allows an advantage depth (AD) of 10 cm, an advantage depth dose rate (ADDR) of 78 cGy/min and an advantage ratio (AR) of 3.2. The p-Sc beam offers 5.9 {times} 10{sup 7} n/cm{sup 2}s and a dose performance of AD = 8 cm and AR = 3.5, suggesting the potential of near-threshold (p,n) reactions such as the p-Li reaction at E{sub p} = 1.92 MeV. A thermal radiography beam could also be obtained from ACWL.

  13. A shielding design for an accelerator-based neutron source for boron neutron capture therapy

    Microsoft Academic Search

    A. E Hawk; T. E Blue; J. E Woollard

    2004-01-01

    Research in boron neutron capture therapy (BNCT) at The Ohio State University Nuclear Engineering Department has been primarily focused on delivering a high quality neutron field for use in BNCT using an accelerator-based neutron source (ABNS). An ABNS for BNCT is composed of a proton accelerator, a high-energy beam transport system, a 7Li target, a target heat removal system (HRS),

  14. Near-Core and In-Core Neutron Radiation Monitors for Real Time Neutron Flux Monitoring and Reactor Power Level Measurements

    SciTech Connect

    Douglas S. McGregor; Marvin L. Adams; Igor Carron; Paul Nelson

    2006-06-12

    MPFDs are a new class of detectors that utilize properties from existing radiation detector designs. A majority of these characteristics come from fission chamber designs. These include radiation hardness, gamma-ray background insensitivity, and large signal output.

  15. Current directions in core-shell nanoparticle design

    NASA Astrophysics Data System (ADS)

    Schärtl, Wolfgang

    2010-06-01

    Ten years ago I wrote a review about the important field of core-shell nanoparticles, focussing mainly on our own work about tracer systems, and briefly addressing polymer-coated nanoparticles as fillers for homogeneous polymer-colloid composites. Since then, the potential use of core-shell nanoparticles as multifunctional sensors or potential smart drug-delivery vehicles in biology and medicine has gained more and more importance, affording special types of multi-functionalized and bio-compatible nanoparticles. In this new review article, I try to address the most important developments during the last ten years. This overview is mainly based on frequently cited and more specialized recent review articles from leaders in their respective field. We will consider a variety of nanoscopic core-shell architectures from highly fluorescent nanoparticles (NPs), protected magnetic NPs, multifunctional NPs, thermoresponsive NPs and biocompatible systems to, finally, smart drug-delivery systems.Ten years ago I wrote a review about the important field of core-shell nanoparticles, focussing mainly on our own work about tracer systems, and briefly addressing polymer-coated nanoparticles as fillers for homogeneous polymer-colloid composites. Since then, the potential use of core-shell nanoparticles as multifunctional sensors or potential smart drug-delivery vehicles in biology and medicine has gained more and more importance, affording special types of multi-functionalized and bio-compatible nanoparticles. In this new review article, I try to address the most important developments during the last ten years. This overview is mainly based on frequently cited and more specialized recent review articles from leaders in their respective field. We will consider a variety of nanoscopic core-shell architectures from highly fluorescent nanoparticles (NPs), protected magnetic NPs, multifunctional NPs, thermoresponsive NPs and biocompatible systems to, finally, smart drug-delivery systems. Dedicated to Professor Manfred Schmidt on the occasion of his 60th birthday

  16. Coded aperture Fast Neutron Analysis: Latest design advances

    NASA Astrophysics Data System (ADS)

    Accorsi, Roberto; Lanza, Richard C.

    2001-07-01

    Past studies have showed that materials of concern like explosives or narcotics can be identified in bulk from their atomic composition. Fast Neutron Analysis (FNA) is a nuclear method capable of providing this information even when considerable penetration is needed. Unfortunately, the cross sections of the nuclear phenomena and the solid angles involved are typically small, so that it is difficult to obtain high signal-to-noise ratios in short inspection times. CAFNAaims at combining the compound specificity of FNA with the potentially high SNR of Coded Apertures, an imaging method successfully used in far-field 2D applications. The transition to a near-field, 3D and high-energy problem prevents a straightforward application of Coded Apertures and demands a thorough optimization of the system. In this paper, the considerations involved in the design of a practical CAFNA system for contraband inspection, its conclusions, and an estimate of the performance of such a system are presented as the evolution of the ideas presented in previous expositions of the CAFNA concept.

  17. Design and optimization of 6li neutron-capture pulse mode ion chamber 

    E-print Network

    Chung, Kiwhan

    2009-05-15

    The purpose of this research is to design and optimize the performance of a unique, inexpensive 6Li neutron-capture pulse-mode ion chamber (LiPMIC) for neutron detection that overcomes the fill-gas contamination stemming from outgas of detector...

  18. A Workshop on Methods for Neutron Scattering Instrument Design. Introduction and Summary

    SciTech Connect

    Hjelm, Rex P.

    1996-12-31

    The future of neutron and x-ray scattering instrument development and international cooperation was the focus of the workshop on ``Methods for Neutron Scattering Instrument Design`` September 23-25 at the E.O. Lawrence Berkeley National Laboratory. These proceedings are a collection of a portion of the invited and contributed presentations.

  19. Preliminary Neutronics Design and Analysis of D2O Cooled High Conversion PWRs

    SciTech Connect

    Hikaru Hiruta; Gilles Youinou

    2012-09-01

    This report presents a neutronics analysis of tight-pitch D2O-cooled PWRs loaded with MOX fuel and focuses essentially on the Pu breeding potential of such reactors as well as on an important safety parameter, the void coefficient, which has to be negative. It is well known that fast reactors have a better neutron economy and are better suited than thermal reactors to breed fissile material from neutron capture in fertile material. Such fast reactors (e.g. sodium-cooled reactors) usually rely on technologies that are very different from those of existing water-cooled reactors and are probably more expensive. This report investigates another possibility to obtain a fast neutron reactor while still relying mostly on a PWR technology by: (1) Tightening the lattice pitch to reduce the water-to-fuel volume ratio compared to that of a standard PWR. Water-to-fuel volume ratios of between 0.45 and 1 have been considered in this study while a value of about 2 is typical of standard PWRs, (2) Using D2O instead of H2O as a coolant. Indeed, because of its different neutron physics properties, the use of D2O hardens the neutron spectrum to an extent impossible with H2O when used in a tight-pitch lattice. The neutron spectra thus obtained are not as fast as those in sodium-cooled reactor but they can still be characterized as fast compared to that of standard PWR neutron spectra. In the phase space investigated in this study we did not find any configurations that would have, at the same time, a positive Pu mass balance (more Pu at the end than at the beginning of the irradiation) and a negative void coefficient. At this stage, the use of radial blankets has only been briefly addressed whereas the impact of axial blankets has been well defined. For example, with a D2O-to-fuel volume ratio of 0.45 and a core driver height of about 60 cm, the fissile Pu mass balance between the fresh fuel and the irradiated fuel (50 GWd/t) would be about -7.5% (i.e. there are 7.5% fewer fissile Pu isotopes at the end than at the beginning of the irradiation) and the void coefficient would be negative. The addition of 1 cm of U-238 blanket at the top and bottom of the fuel would bring the fissile Pu mass balance from -7.5% to -6.5% but would also impact the void coefficient in the wrong way. In fact, it turns out that the void coefficient is so sensitive to the presence of axial blanket that it limits its size to only a few cm for driver fuel height of about 50-60 cm. For reference, the fissile Pu mass balance is about -35% in a standard PWR MOX fuel such as those used in France. In order to reduce the fissile Pu deficit and potentially reach a true breeding regime (i.e. a positive Pu mass balance), it would be necessary to make extensive use of radial blankets, both internal and external. Even though this was not addressed in detail here, it is reasonable to believe that at least as much U-238 blanket subassemblies as MOX driver fuel subassemblies would be necessary to breed enough Pu to compensate for the Pu deficit in the driver fuel. Hence, whereas a relatively simple D2O-cooled PWR core design makes it possible to obtain a near-breeder core, it may be necessary to more than double the mass of heavy metal in the core as well as the mass of heavy metal to reprocess per unit of energy produced in order to breed the few percents of Pu missing to reach a true breeding regime. It may be interesting to quantify these aspects further in the future.

  20. Cylindrical Detector and Preamplifier Design for Detecting Neutrons 

    E-print Network

    Xia, Zhenghua

    2010-01-14

    in the tissue-equivalent walls depend on the energy of the primary neutrons. The differences in the spectra measured by different size detectors will provide additional information on the incident neutron energy. Monte Carlo N-particle extended (MCNPX) code...

  1. Ferrite core loss for power magnetic components design

    Microsoft Academic Search

    Waseem Roshen

    1991-01-01

    A practical method is presented for computing high-frequency ferrite core losses in the magnetic component for arbitrary voltage waveforms. The model presented requires only a few material parameters as input. To calculate ferrite hysteresis losses, a model based on empirical rules is employed. For high-frequency eddy current losses, a built phenomenon is assumed. It is demonstrated that the hysteresis model

  2. Development and applications of methodologies for the neutronic design of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR)

    NASA Astrophysics Data System (ADS)

    Fratoni, Massimiliano

    This study investigated the neutronic characteristics of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR), a novel nuclear reactor concept that combines liquid salt (7LiF-BeF2---flibe) cooling and TRISO coated-particle fuel technology. The use of flibe enables operation at high power density and atmospheric pressure and improves passive decay-heat removal capabilities, but flibe, unlike conventional helium coolant, is not transparent to neutrons. The flibe occupies 40% of the PB-AHTR core volume and absorbs ˜8% of the neutrons, but also acts as an effective neutron moderator. Two novel methodologies were developed for calculating the time dependent and equilibrium core composition: (1) a simplified single pebble model that is relatively fast; (2) a full 3D core model that is accurate and flexible but computationally intensive. A parametric analysis was performed spanning a wide range of fuel kernel diameters and graphite-to-heavy metal atom ratios to determine the attainable burnup and reactivity coefficients. Using 10% enriched uranium ˜130 GWd/tHM burnup was found to be attainable, when the graphite-to-heavy metal atom ratio (C/HM) is in the range of 300 to 400. At this or smaller C/HM ratio all reactivity coefficients examined---coolant temperature, coolant small and full void, fuel temperature, and moderator temperature, were found to be negative. The PB-AHTR performance was compared to that of alternative options for HTRs, including the helium-cooled pebble-bed reactor and prismatic fuel reactors, both gas-cooled and flibe-cooled. The attainable burnup of all designs was found to be similar. The PB-AHTR generates at least 30% more energy per pebble than the He-cooled pebble-bed reactor. Compared to LWRs the PB-AHTR requires 30% less natural uranium and 20% less separative work per unit of electricity generated. For deep burn TRU fuel made from recycled LWR spent fuel, it was found that in a single pass through the core ˜66% of the TRU can be transmuted; this burnup is slightly superior to that attainable in helium-cooled reactors. A preliminary analysis of the modular variant for the PB-AHTR investigated the triple heterogeneity of this design and determined its performance characteristics.

  3. Magnetic heating properties and neutron activation of tungsten-oxide coated biocompatible FePt core-shell nanoparticles.

    PubMed

    Seemann, K M; Luysberg, M; Révay, Z; Kudejova, P; Sanz, B; Cassinelli, N; Loidl, A; Ilicic, K; Multhoff, G; Schmid, T E

    2015-01-10

    Magnetic nanoparticles are highly desirable for biomedical research and treatment of cancer especially when combined with hyperthermia. The efficacy of nanoparticle-based therapies could be improved by generating radioactive nanoparticles with a convenient decay time and which simultaneously have the capability to be used for locally confined heating. The core-shell morphology of such novel nanoparticles presented in this work involves a polysilico-tungstate molecule of the polyoxometalate family as a precursor coating material, which transforms into an amorphous tungsten oxide coating upon annealing of the FePt core-shell nanoparticles. The content of tungsten atoms in the nanoparticle shell is neutron activated using cold neutrons at the Heinz Maier-Leibnitz (FRMII) neutron facility and thereby transformed into the radioisotope W-187. The sizeable natural abundance of 28% for the W-186 precursor isotope, a radiopharmaceutically advantageous gamma-beta ratio of ???30% and a range of approximately 1mm in biological tissue for the 1.3MeV ?-radiation are promising features of the nanoparticles' potential for cancer therapy. Moreover, a high temperature annealing treatment enhances the magnetic moment of nanoparticles in such a way that a magnetic heating effect of several degrees Celsius in liquid suspension - a prerequisite for hyperthermia treatment of cancer - was observed. A rise in temperature of approximately 3°C in aqueous suspension is shown for a moderate nanoparticle concentration of 0.5mg/ml after 15min in an 831kHz high-frequency alternating magnetic field of 250Gauss field strength (25mT). The biocompatibility based on a low cytotoxicity in the non-neutron-activated state in combination with the hydrophilic nature of the tungsten oxide shell makes the coated magnetic FePt nanoparticles ideal candidates for advanced radiopharmaceutical applications. PMID:25445697

  4. An Effective Instructional Design at Iowa Lakes Community College to Make Hard Core Unemployed People Employable.

    ERIC Educational Resources Information Center

    Schorzmann, Eugene F.

    In order to meet the educational needs of the hard-core unemployed, an instructional program known as the "Alternative Career Center" was designed at Iowa Lakes Community College to prepare those identified by the Comprehensive Employment and Training Act for employment. After reviewing the literature on educational programs for the hard-core…

  5. Design and Analysis of a High-Speed Claw Pole Motor With Soft Magnetic Composite Core

    Microsoft Academic Search

    Yunkai Huang; Jianguo Zhu; Youguang Guo; Zhiwei Lin; Qiansheng Hu

    2007-01-01

    Soft magnetic composite (SMC) material is formed by surface-insulated iron powder particles, generating unique properties like magnetic and thermal isotropy, and very low eddy currents. This paper presents the design and analysis of a high-speed claw pole motor with an SMC stator core for reducing core losses and cost. The analyses of magnetic and thermal fields are conducted based on

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

    SciTech Connect

    B. Boer; A. M. Ougouag

    2010-09-01

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

  7. Programmable Logic IP Cores in SoC Design: Opportunities and Challenges Steven J.E. Wilton and Resve Saleh

    E-print Network

    Wilton, Steve

    Programmable Logic IP Cores in SoC Design: Opportunities and Challenges Steven J.E. Wilton, B.C., Canada {stevew, res} @ece.ubc.ca Abstract As SoC design enters into mainstream usage be realized using programmable logic cores. These cores are like any other IP in the SoC design methodology

  8. The design, fabrication and properties of B4C/Al neutron absorbers

    NASA Astrophysics Data System (ADS)

    Zhang, Peng; Li, Yuli; Wang, Wenxian; Gao, Zhanping; Wang, Baodong

    2013-06-01

    Neutron absorber is used for the criticality safety during the storage or transportation of spent nuclear fuel. In this work, the metal matrix composite with good mechanical property and thermal neutron absorbing ability was investigated based on B4C/Al neutron radiation shielding material. The composition ratio for B4C/Al composite was firstly designed and the dependence of the neutron transmission on the thickness of the material was calculated. By vacuum hot-pressing technique at a low temperature, the neutron absorbers with high concentration of B4C were fabricated. Furthermore, the corresponding microstructure, physical, mechanical and corrosion properties as well as fracture surface were analyzed, proving that the developed composites can shield the neutron radiation as effectively as cadmium materials.

  9. Spatial and spectral characteristics of a compact system neutron beam designed for BNCT facility.

    PubMed

    Ghassoun, J; Chkillou, B; Jehouani, A

    2009-04-01

    The development of suitable neutron sources and neutron beam is critical to the success of Boron Neutron Capture Therapy (BNCT). In this work a compact system designed for BNCT is presented. The system consists of (252)Cf fission neutron source and a moderator/reflector/filter/shield assembly. The moderator/reflector/filter arrangement has been optimized to maximize the epithermal neutron component which is useful for BNCT treatment of deep seated tumors with the suitably low level of beam contamination. The MCMP5 code has been used to calculate the different components of neutrons, secondary gamma rays originating from (252)Cf source and the primary gamma rays emitted directly by this source at the exit face of the compact system. The fluence rate distributions of such particles were also computed along the central axis of a human head phantom. PMID:19168369

  10. Modeling Core Failure by the Tsai-Wu Criterion in the Design of Foam-Core Sandwich Beams

    Microsoft Academic Search

    E. E. Gdoutos; V. D. Balopoulos; P. A. Kalaitzidis; M. Konsta

    This work is part of an extensive program at the Laboratory of Applied Mechanics of D.U.Th. involving experimental and numerical\\u000a research, to expand our knowledge of composite materials and propose enhanced techniques for rational design of sandwich beams\\u000a and shells. In particular, it is a study of the importance of describing core failure in elastic sandwich beams by the Tsai–Wu

  11. DESIGN OF A LARGE-AREA FAST NEUTRON DIRECTIONAL DETECTOR.

    SciTech Connect

    VANIER, P.E.

    2006-10-29

    A large-area fast-neutron double-scatter directional detector and spectrometer is being constructed using l-meter-long plastic scintillator paddles with photomultiplier tubes at both ends. The scintillators detect fast neutrons by proton recoil and also gamma rays by Compton scattering. The paddles are arranged in two parallel planes so that neutrons can be distinguished from muons and gamma rays by time of flight between the planes. The signal pulses are digitized with a time resolution of one gigasample per second. The location of an event along each paddle can be determined from the relative amplitudes or timing of the signals at the ends. The angle of deflection of a neutron in the first plane can be estimated from the energy deposited by the recoil proton, combined with the scattered neutron time-of-flight energy. Each scattering angle can be back-projected as a cone, and many intersecting cones define the incident neutron direction from a distant point source. Moreover, the total energy of each neutron can be obtained, allowing some regions of a fission source spectrum to be distinguished from background generated by cosmic rays. Monte Carlo calculations will be compared with measurements.

  12. 76 FR 14825 - Core Principles and Other Requirements for Designated Contact Markets

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-03-18

    ...3038-AD09 Core Principles and Other Requirements for Designated Contact Markets AGENCY: Commodity Futures Trading Commission. ACTION...site at http://www.cftc.gov. FOR FURTHER INFORMATION CONTACT: Nancy Markowitz, Assistant Deputy Director,...

  13. Development of optimized core design and analysis methods for high power density BWRs

    E-print Network

    Shirvan, Koroush

    2013-01-01

    Increasing the economic competitiveness of nuclear energy is vital to its future. Improving the economics of BWRs is the main goal of this work, focusing on designing cores with higher power density, to reduce the BWR ...

  14. DESIGN AND CORE COMPETENCY, THE MISSING LINKS Eric Bonjour1

    E-print Network

    Paris-Sud XI, Université de

    competency building are key issues for managers of product development process, design skills networks Designers have to develop products with drastic performances related to cost, delay, quality, environmental competency. This framework is a first step beyond an efficient competency based management of design

  15. Design aspects of medium power double rotor radial flux air-cored PM wind generators

    Microsoft Academic Search

    J. A. Stegmann; M. J. Kamper

    2009-01-01

    The electromagnetic and mechanical design aspects of optimally designed double rotor radial flux permanent magnet wind generators with non-overlap air-cored (iron-less) stator windings are analyzed in this paper. The optimal design is based on an accurate analytical model and is confirmed with finite element analysis. It is shown, amongst other things, that the electromagnetic design and not the mechanical design

  16. Designing and testing the neutron source deployment system and calibration plan for a dark matter detector

    E-print Network

    Westerdale, Shawn (Shawn S.)

    2011-01-01

    In this thesis, we designed and tested a calibration and deployment system for the MiniCLEAN dark matter detector. The deployment system uses a computer controlled winch to lower a canister containing a neutron source into ...

  17. PUBLISHED VERSION Conceptual design of a neutron camera for MAST Upgrade

    E-print Network

    ) This paper presents two different conceptual designs of neutron cameras for Mega Ampere Spheri- cal Tokamak On the Mega Ampere Spherical Tokamak (MAST),1 fu- sion reactions between fast ions, from auxiliary heating

  18. Architecting voltage islands in core-based system-on-a-chip designs

    Microsoft Academic Search

    Jingcao Hu; Youngsoo Shin; Nagu R. Dhanwada; Radu Marculescu

    2004-01-01

    Voltage islands enable core-level power optimization for System-on-Chip (SoC) designs by utilizing a unique supply voltage for each core. Architecting voltage islands involves island partition creation, voltage level assignment and floorplanning. The task of island partition creation and level assignment have to be done simultaneously in a floorplanning context due to the physical constraints involved in the design process. This

  19. Designing accelerator-based epithermal neutron beams for boron neutron capture therapy

    Microsoft Academic Search

    D. L. Bleuel; R. J. Donahue; B. A. Ludewigt; J. Vujic

    1998-01-01

    The ⁷Li(p,n)⁷Be reaction has been investigated as an accelerator-driven neutron source for proton energies between 2.1 and 2.6 MeV. Epithermal neutron beams shaped by three moderator materials, Al\\/AlFâ, ⁷LiF, and DâO, have been analyzed and their usefulness for boron neutron capture therapy (BNCT) treatments evaluated. Radiation transport through the moderator assembly has been simulated with the Monte Carlo {ital N}-particle

  20. Multi-group helium and hydrogen production cross section libraries for fusion neutronics design

    Microsoft Academic Search

    Seiji Mori; S. Zimin; Hideyuki Takatsu

    1993-01-01

    The helium and hydrogen production cross section libraries based on the JENDL-3 data file were compiled for use in neutronics and shielding design calculation of a fusion reactor. These libraries have the same group structures as the transport cross section sets, FUSION-J3 and FUSION-40, which are often used in fusion neutronics design and can be used as the response function

  1. LMR design concepts for transuranic management in low sodium void worth cores

    SciTech Connect

    Hill, R.N.

    1991-01-01

    The fuel cycle processing techniques and hard neuron spectrum of the Integral Fast Reactor (IFR) metal fuel cycle have favorable characteristics for the management of transuranics; and the wide range of breeding characteristics available in metal fuelled cores provides for flexibility in transuranic management strategy. Previous studies indicate that most design options which decrease the breeding ratio also show a decrease in sodium void worth; therefore, low void worths are achievable in transuranic burning (low breeding ratio) core designs. This paper describes numerous trade studies assessing various design options for a low void worth transuranic burner core. A flat annular core design appears to be a promising concept; the high leakage geometry yields a low breeding ratio and small sodium void worth. To allow flexibility in breeding characteristics, alternate design options which achieve fissile self-sufficiency are also evaluated. A self-sufficient core design which is interchangeable with the burner core and maintains a low sodium void worth is developed. 13 refs., 1 fig., 4 tabs.

  2. Design of low-loss and highly birefringent hollow-core photonic crystal fiber.

    PubMed

    Roberts, P J; Williams, D P; Sabert, H; Mangan, B J; Bird, D M; Birks, T A; Knight, J C; Russell, P St J

    2006-08-01

    A practical hollow-core photonic crystal fiber design suitable for attaining low-loss propagation is analyzed. The geometry involves a number of localized elliptical features positioned on the glass ring that surrounds the air core and separates the core and cladding regions. The size of each feature is tuned so that the composite core-surround geometry is antiresonant within the cladding band gap, thus minimizing the guided mode field intensity both within the fiber material and at material/air interfaces. A birefringent design, which involves a 2-fold symmetric arrangement of the features on the core-surround ring, gives rise to wavelength ranges where the effective index difference between the polarization modes is larger than 10(-4). At such high birefringence levels, one of the polarization modes retains favorable field exclusion characteristics, thus enabling low-loss propagation of this polarization channel. PMID:19529102

  3. Design of a compact high-power neutron source—The EURISOL converter target

    NASA Astrophysics Data System (ADS)

    Samec, K.; Milenkovi?, R. Ž.; Dementjevs, S.; Ashrafi-Nik, M.; Kalt, A.

    2009-07-01

    The EURISOL project, a multi-lateral initiative supported by the EU, aims to develop a facility to achieve high yields of isotopes in radioactive beams and extend the variety of these isotopes towards more exotic types. The neutron source at the heart of the projected facility is designed to generate isotopes by fissioning uranium carbide (UC) targets arranged around a 4 MW neutron source. For reasons of efficiency, it is essential that the neutron source be as compact as possible, to avoid losing neutrons by absorption whilst maximising the escaping neutron flux, thus increasing the number of fissions in the UC targets. The resulting configuration presents a challenge in terms of absorbing heat deposition rates of up to 8 kW/cm3 in the neutron source; it has led to the selection of liquid metal for the target material. The current paper presents the design of a compact high-power liquid-metal neutron source comprising a specially optimised beam window concept. The design is based on two-dimensional (2D) and three-dimensional (3D) computational fluid dynamics (CFD) numerical simulations for thermal hydraulics and hydraulic aspects, as well as finite-element method (FEM) for assessing thermo-mechanical stability. The resulting optimised design was validated by a dedicated hydraulic test under realistic flow conditions. A full-scale mock-up was built at the Paul Scherrer Institute (PSI) and was tested at the Institute of Physics of the University of Latvia (IPUL).

  4. Integration of the FORMOSA PWR in-core fuel management optimization code into nuclear design code systems

    SciTech Connect

    Sawyer, J.M.; Murphy, P.R.; Parks, G.T.; Maldonado, G.I.; Turinsky, P.J. (North Carolina State Univ., Raleigh (United States)); Daniels, D.A. (Carolina Power and Light Co., Raleigh, NC (United States))

    1991-01-01

    North Carolina State University has, in collaboration with Duke Power and Carolina Power and Light Company (CP and L), developed an in-core fuel management code, FORMOSA, which automatically provides the much needed computational capability to determine optimal pressurized water reactor loading patterns (LPs) for a variety of objective functions. FORMOSA employs simulated annealing to control the optimization process, using the exchange of fuel assemblies between randomly selected locations to seek improvements in the LP. Candidate patterns are assessed using second-order generalized perturbation theory (GPT) expressions for the characteristics of interest based on a two-dimensional, coarse-mesh finite difference discretization of the two-group neutron diffusion equations. Before its potential can be exploited, FORMOSA must be integrated into a utility's nuclear design system. This task was undertaken at both Duke Power and CP and L, which employ the Studsvik and Scandpower methodologies, respectively.

  5. Conceptual design and neutronics analyses of a fusion reactor blanket simulation facility

    Microsoft Academic Search

    Beller

    1986-01-01

    A new conceptual design of a fusion reactor blanket simulation facility was developed. This design follows the principles that have been successfully employed in the Purdue Fast Breeder Blanket Facility (FBBR), because experiments conducted in it have resulted in the discovery of deficiencies in neutronics prediction methods. With this design, discrepancies between calculation and experimental data can be fully attributed

  6. Conceptual design and neutronics analyses of a fusion reactor blanket simulation facility

    Microsoft Academic Search

    D. E. Beller; K. O. Ott; W. K. Terry

    1987-01-01

    A new conceptual design of a fusion reactor blanket simulation facility has been developed. This design follows the principles that have been successfully employed in the Purdue Fast Breeder Blanket Facility (FBBF), where experiments have resulted in the discovery of substantial deficiencies in neutronics predictions. With this design, discrepancies between calculation and experimental data can be nearly fully attributed to

  7. Conceptual design and neutronics analyses of a fusion-reactor blanket-simulation facility. Doctoral thesis

    Microsoft Academic Search

    Beller

    1986-01-01

    A new conceptual design of a fusion-reactor blanket-simulation facility was developed. This design follows the principles that have been successfully employed in the Purdue Fast Breeder Blanket Facility (FBBF), because experiments conducted in it have resulted in the discovery of deficiencies in neutronics prediction methods. With this design, discrepancies between calculation and experimental data can be fully attributed to calculation

  8. 3He neutron detector design for active detection of cargo containers

    NASA Astrophysics Data System (ADS)

    McDevitt, Daniel B.; Eberhard, J. W.; Zelakiewicz, Scott; Maschinot, Aaron

    2008-04-01

    We report on the design of a neutron detector using industry standard 3He tubes to count delayed neutrons during the interrogation of cargo containers for the presence of Special Nuclear Material (SNM). Simulations of the detector design were run for delayed neutron spectra for a variety of cargos containing SNM using the Monte Carlo computer code COG. The simulations identified parameters crucial to optimize the detector design. These choices include moderating material type and thickness, tube spacing, tube pressure and number of tubes. An experimental prototype was also constructed based on the simulated design specifications. This paper discusses the parameters that lead up to the optimized detector design. It also compares the performance of the Monte Carlo simulated design and the experimental detector when exposed to a 239Pu-Be source.

  9. Neutron capture on Pt isotopes in iron meteorites and the Hf-W chronology of core formation in planetesimals

    NASA Astrophysics Data System (ADS)

    Kruijer, Thomas S.; Fischer-Gödde, Mario; Kleine, Thorsten; Sprung, Peter; Leya, Ingo; Wieler, Rainer

    2013-01-01

    The short-lived 182Hf-182W isotope system can provide powerful constraints on the timescales of planetary core formation, but its application to iron meteorites is hampered by neutron capture reactions on W isotopes resulting from exposure to galactic cosmic rays. Here we show that Pt isotopes in magmatic iron meteorites are also affected by capture of (epi)thermal neutrons and that the Pt isotope variations are correlated with variations in 182W/184W. This makes Pt isotopes a sensitive neutron dosimeter for correcting cosmic ray-induced W isotope shifts. The pre-exposure 182W/184W derived from the Pt-W isotope correlations of the IID, IVA and IVB iron meteorites are higher than most previous estimates and are more radiogenic than the initial 182W/184W of Ca-Al-rich inclusions (CAI). The Hf-W model ages for core formation range from +1.6±1.0 million years (Ma; for the IVA irons) to +2.7±1.3 Ma after CAI formation (for the IID irons), indicating that there was a time gap of at least ˜1 Ma between CAI formation and metal segregation in the parent bodies of some iron meteorites. From the Hf-W ages a time limit of <1.5-2 Ma after CAI formation can be inferred for the accretion of the IID, IVA and IVB iron meteorite parent bodies, consistent with earlier conclusions that the accretion of differentiated planetesimals predated that of most chondrite parent bodies.

  10. Californium252 neutron sources

    Microsoft Academic Search

    Yevgeni A. Karelin; Yan N. Gordeev; Valentin I. Karasev; Vyacheslav M. Radchenko; Yevgeni V. Schimbarev; Rostislav A. Kuznetsov

    1997-01-01

    The technologies of neutron sources production developed by RIAR are reviewed. To produce 252Cf-containing core, different techniques are used depending on the source type. They are: impregnation of foam-alundum billets, glass beads and cermet production (point sources), electroplating on the platinum cathode (linear sources).Neutron sources produced by RIAR are one- or two-capsule design. Capsules are made of stainless steel and

  11. Modified Y-TZP core design improves all-ceramic crown reliability.

    PubMed

    Silva, N R F A; Bonfante, E A; Rafferty, B T; Zavanelli, R A; Rekow, E D; Thompson, V P; Coelho, P G

    2011-01-01

    This study tested the hypothesis that all-ceramic core-veneer system crown reliability is improved by modification of the core design. We modeled a tooth preparation by reducing the height of proximal walls by 1.5 mm and the occlusal surface by 2.0 mm. The CAD-based tooth preparation was replicated and positioned in a dental articulator for core and veneer fabrication. Standard (0.5 mm uniform thickness) and modified (2.5 mm height lingual and proximal cervical areas) core designs were produced, followed by the application of veneer porcelain for a total thickness of 1.5 mm. The crowns were cemented to 30-day-aged composite dies and were either single-load-to-failure or step-stress-accelerated fatigue-tested. Use of level probability plots showed significantly higher reliability for the modified core design group. The fatigue fracture modes were veneer chipping not exposing the core for the standard group, and exposing the veneer core interface for the modified group. PMID:21057036

  12. A review of irradiation effects on LWR core internal materials - neutron embrittlement.

    SciTech Connect

    Chopra, O. K.; Rao, A. S. (Environmental Science Division); (U.S NRC)

    2011-05-01

    Austenitic stainless steels (SSs) are used extensively as structural alloys in the internal components of light water reactor (LWR) pressure vessels because of their relatively high strength, ductility, and fracture toughness. However, exposure to neutron irradiation for extended periods not only changes the microstructure and microchemistry of these steels, but also degrades their fracture properties. The existing data on irradiated austenitic SSs are reviewed to determine the effects of key parameters such as material type and condition and irradiation temperature, dose, and dose rate on neutron embrittlement. Differences in the radiation-induced degradation of fracture properties between LWR and fast-reactor irradiations are also discussed. The results are used to (a) define a threshold fluence above which irradiation effects on fracture toughness of the material are significant, (b) evaluate the potential of neutron embrittlement under LWR operating conditions, and (c) assess the potential effects of voids on fracture toughness.

  13. Explosively driven 1 MV output air-core pulse transformer design: Final report

    Microsoft Academic Search

    P. D. A. Champney; C. Eichenberger

    1988-01-01

    This report presents the results of a pulse transformer design program conducted by Pulse Sciences Inc., (PSI) for the Lawrence Livermore National Laboratory (LLNL). The principle program objective was the development of a 1 MV output air-core pulse transformer design compatible with an LLNL explosive driver concept. 34 figs.

  14. Explosively driven 1 MV output air-core pulse transformer design: Final report

    SciTech Connect

    Champney, P.D'A.; Eichenberger, C.

    1988-09-01

    This report presents the results of a pulse transformer design program conducted by Pulse Sciences Inc., (PSI) for the Lawrence Livermore National Laboratory (LLNL). The principle program objective was the development of a 1 MV output air-core pulse transformer design compatible with an LLNL explosive driver concept. 34 figs.

  15. Fusion Engineering and Design 82 (2007) 217236 Advanced power core system for the

    E-print Network

    California at San Diego, University of

    2007-01-01

    Fusion Engineering and Design 82 (2007) 217­236 Advanced power core system for the ARIES-AT power of Wisconsin, Fusion Technology Institute, 1500 Engineering Drive, Madison, WI 53706-1687, USA c Engineering and Design 82 (2007) 217­236 ties on the performance of advanced tokamak power plants [1

  16. Fusion Engineering and Design 80 (2006) 7998 Advanced power core system for the

    E-print Network

    California at San Diego, University of

    2006-01-01

    Fusion Engineering and Design 80 (2006) 79­98 Advanced power core system for the ARIES-AT power of Wisconsin, Fusion Technology Institute, 1500 Engineering Drive, Madison, WI 53706-1687, USA c.06.356 #12;80 A.R. Raffray et al. / Fusion Engineering and Design 80 (2006) 79­98 of new physics

  17. Neutronics analysis of an open-cycle high-impulse gas core reactor concept

    NASA Technical Reports Server (NTRS)

    Whitmarsh, C. L., Jr.

    1972-01-01

    A procedure was developed to calculate the critical fuel mass, including the effects of propellant pressure, for coaxial-flow gas-core reactors operating at 196,600 newtons thrust and 4400 seconds specific impulse. Data were generated for a range of cavity diameter, reflector-moderator thickness, and quantity of structural material. Also presented are such core characteristics as upper limits on cavity pressure, spectral hardening in very-high-temperature hydrogen, and reactivity coefficients.

  18. Dynamical analysis of innovative core designs facing unprotected transients with the MAT5 DYN code

    SciTech Connect

    Darmet, G.; Massara, S. [EDF R and D, 1 avenue du general de Gaulle, 92140 Clamart (France)

    2012-07-01

    Since 2007, advanced Sodium-cooled Fast Reactors (SFR) are investigated by CEA, AREVA and EDF in the framework of a joint French collaboration. A prototype called ASTRID, sets out to demonstrate progress made in SFR technology, is due to operate in the years 2020's. The modeling of unprotected transients by computer codes is one of the key safety issues in the design approach to such SFR systems. For that purpose, the activity on CATHARE, which is the reference code for the transient analysis of ASTRID, has been strengthened during last years by CEA. In the meantime, EDF has developed a simplified and multi-channel code, named MAT5 DYN, to analyze and validate innovative core designs facing protected and unprotected transients. First, the paper consists in a description of MAT5 DYN: a code based on the existing code MAT4 DYN including major improvements on geometry description and physical modeling. Second, two core designs based on the CFV core design developed at CEA are presented. Then, the dynamic response of those heterogeneous cores is analyzed during unprotected loss of flow (ULOF) transient and unprotected transient of power (UTOP). The results highlight the importance of the low void core effect specific to the CFV design. Such an effect, when combined with a sufficient primary pump halving time and an optimized cooling group scheme, allows to delay (or, possibly, avoid) the sodium boiling onset during ULOF accidents. (authors)

  19. Computational Benchmark Calculations Relevant to the Neutronic Design of the Spallation Neutron Source (SNS)

    SciTech Connect

    Gallmeier, F.X.; Glasgow, D.C.; Jerde, E.A.; Johnson, J.O.; Yugo, J.J.

    1999-11-14

    The Spallation Neutron Source (SNS) will provide an intense source of low-energy neutrons for experimental use. The low-energy neutrons are produced by the interaction of a high-energy (1.0 GeV) proton beam on a mercury (Hg) target and slowed down in liquid hydrogen or light water moderators. Computer codes and computational techniques are being benchmarked against relevant experimental data to validate and verify the tools being used to predict the performance of the SNS. The LAHET Code System (LCS), which includes LAHET, HTAPE ad HMCNP (a modified version of MCNP version 3b), have been applied to the analysis of experiments that were conducted in the Alternating Gradient Synchrotron (AGS) facility at Brookhaven National Laboratory (BNL). In the AGS experiments, foils of various materials were placed around a mercury-filled stainless steel cylinder, which was bombarded with protons at 1.6 GeV. Neutrons created in the mercury target, activated the foils. Activities of the relevant isotopes were accurately measured and compared with calculated predictions. Measurements at BNL were provided in part by collaborating scientists from JAERI as part of the AGS Spallation Target Experiment (ASTE) collaboration. To date, calculations have shown good agreement with measurements.

  20. Thermal neutron irradiation field design for boron neutron capture therapy of human explanted liver.

    PubMed

    Bortolussi, S; Altieri, S

    2007-12-01

    The selective uptake of boron by tumors compared to that by healthy tissue makes boron neutron capture therapy (BNCT) an extremely advantageous technique for the treatment of tumors that affect a whole vital organ. An example is represented by colon adenocarcinoma metastases invading the liver, often resulting in a fatal outcome, even if surgical resection of the primary tumor is successful. BNCT can be performed by irradiating the explanted organ in a suitable neutron field. In the thermal column of the Triga Mark II reactor at Pavia University, a facility was created for this purpose and used for the irradiation of explanted human livers. The neutron field distribution inside the organ was studied both experimentally and by means of the Monte Carlo N-particle transport code (MCNP). The liver was modeled as a spherical segment in MCNP and a hepatic-equivalent solution was used as an experimental phantom. In the as-built facility, the ratio between maximum and minimum flux values inside the phantom ((phi(max)/phi(min)) was 3.8; this value can be lowered to 2.3 by rotating the liver during the irradiation. In this study, the authors proposed a new facility configuration to achieve a uniform thermal neutron flux distribution in the liver. They showed that a phi(max)/phi(min) ratio of 1.4 could be obtained without the need for organ rotation. Flux distributions and dose volume histograms were reported for different graphite configurations. PMID:18196797

  1. Computational investigations of the dynamical characteristics of the core of a periodic pulse reactor in a system with cascade neutron multiplication

    Microsoft Academic Search

    A. V. Gulevich; P. P. D’yachenko; O. F. Kukharchuk; Yu. I. Likhachev; D. V. Razumovskii; D. A. Rogov; E. N. Kravchenko; O. G. Fokina

    2004-01-01

    The core dynamics of a fast reactor in a cascade reactor system operating in a periodic-pulse regime are examined. A model of a BN-600 fuel element is used as a computational model. Computational studies of the neutron kinetics processes in a fast rector-subcritical assembly system and the thermal dynamics of a fuel element in the core of a periodic-pulse reactor

  2. Design and performance of a pulse transformer based on Fe-based nanocrystalline core

    Microsoft Academic Search

    Liu Yi; Feng Xibo; Fuchang Lin

    2011-01-01

    A dry-type pulse transformer based on Fe-based nanocrystalline core with a load of 0.88 nF, output voltage of more than 65 kV, and winding ratio of 46 is designed and constructed. The dynamic characteristics of Fe-based nanocrystalline core under the impulse with the pulse width of several microseconds were studied. The pulse width and incremental flux density have an important

  3. Design and Analysis of a Permanent Magnet Claw Pole\\/Transverse Flux Motor with SMC Core

    Microsoft Academic Search

    YouGuang Guo; Jian Guo Zhu; Haiwei Lu

    2005-01-01

    This paper presents the design and analysis of a claw pole\\/transverse flux motor (CPTFM) with soft magnetic composite (SMC) core and permanent magnet flux-concentrating rotor. Three-dimensional magnetic field finite element analysis is conducted to accurately calculate key motor parameters such as winding flux, back electromotive force, winding inductance, and core loss. Equivalent electric circuit is derived under optimum brushless DC

  4. Design manufacture and test of a cryo-stable Offner relay using aluminum foam core optics

    Microsoft Academic Search

    Ryan S. McClelland

    2001-01-01

    Aluminum foam core optics have the desirable characteristics of being lightweight, cryo-stable, and low cost. The availability of high quality aluminum foam and a bare aluminum super-polishing process have allowed high performance foam core optics made entirely of aluminum to be produced. Mirrors with integral mounts were designed for minimum surface error induced by self-weight deflection, thermal gradients, and mounting

  5. Design and performance of a cryogenic apparatus for magnetically trapping ultracold neutrons

    NASA Astrophysics Data System (ADS)

    Huffman, P. R.; Coakley, K. J.; Doyle, J. M.; Huffer, C. R.; Mumm, H. P.; O'Shaughnessy, C. M.; Schelhammer, K. W.; Seo, P.-N.; Yang, L.

    2014-11-01

    The cryogenic design and performance of an apparatus used to magnetically confine ultracold neutrons (UCN) is presented. The apparatus is part of an effort to measure the beta-decay lifetime of the free neutron and is comprised of a high-current superconducting magnetic trap that surrounds ?21 l of isotopically pure 4He cooled to approximately 250 mK. A 0.89 nm neutron beam can enter the apparatus from one end of the magnetic trap and a light collection system allows visible light generated within the helium by decays to be transported to detectors at room temperature. Two cryocoolers are incorporated to reduce liquid helium consumption.

  6. Dependence of neutron-induced radioactivity in fusion reactors on geometric design parameters

    SciTech Connect

    Lasche, G.P.; Blink, J.A.

    1983-01-01

    Although the neutron-induced activation in a fusion reactor is a non-linear problem whose solution requires the use of neutron transport codes and neutron activation and decay codes, a number of simple arguments can be made which give useful scaling laws for the total radioactivity in a fusion reactor. Because these laws rely heavily on assumptions of linearity and the smallness of second-order effects, we have compared them to the results of computer experiments designed to investigate their validity over the range of operating parameters typical of fusion reactors.

  7. The design of a portable cosmic ray three-band neutron detector.

    PubMed

    Monk, S D; Joyce, M J

    2007-01-01

    The design of a portable three-band cosmic-ray neutron detector is reported in this article. This instrument has been designed to characterise cosmic ray neutron fields in the upper atmosphere and in cosmic reference field facilities. The design utilises a spherical moderator with a layer of spallation material covering a central (3)He proportional counter. The instrument incorporates twelve lithium-coated diodes, six on the outside of the polyethylene layer and six placed within the structure. The dimensions, materials and arrangement of these in the instrument have been optimised with MCNPX to provide a compromise between the requirements of portability and spectral response. PMID:16829509

  8. Design and execution of data acquisition systems for inelastic neutron spectrometers

    Microsoft Academic Search

    P. Klosowski; N. Maliszewskyj; H. Layer; J. Raebiger; P. D. Gallagher; M. Kirsch

    1999-01-01

    We present the design and implementation of the data acquisition and instrument control for two new state-of-the-art inelastic neutron scattering spectrometers at the NIST Center for Neutron Research. The data acquisition system for the High Flux Backscattering Spectrometer (HFBS) is a multi-processing setup of several COTS digital and analog VME modules and a Motorola 96000 digital signal processor to perform

  9. Optimization of the Ballistic Guide Design for the SNS FNPB 8.9 A Neutron Line

    E-print Network

    Takeyasu M. Ito; Christopher B. Crawford; Geoffrey L. Greene

    2006-04-28

    The optimization of the ballistic guide design for the SNS Fundamental Neutron Physics Beamline 8.9 A line is described. With a careful tuning of the shape of the curve for the tapered section and the width of the straight section, this optimization resulted in more than 75% increase in the neutron flux exiting the 33 m long guide over a straight m=3.5 guide with the same length.

  10. Gradient coil design considerations for iron core interventional magnets.

    PubMed

    Ersahin, A; Bronskill, M J; Henkelman, R M; Collick, B; Hinks, R S

    1998-01-01

    The requirements for access and imaging performance compete in the design of open-concept MR magnets and gradient coils. We conducted a theoretical and experimental investigation of gradient coil design using both solid and laminated pole piece construction to determine whether adequate eddy current control can be obtained without shielded gradient coils while maintaining good patient access and high gradient performance. Eddy currents, gradient characteristics, gradient efficiency, and magnet openness are compared and contrasted for various construction options based on a compact, .27 T iron yoke magnet. The resulting pole pieces and gradient coils have high efficiency for an interventional open-configuration magnet while taking up minimal space between the poles for improved patient access. PMID:9786154

  11. Neutronic design optimisation of modular HCPB blankets for fusion power reactors

    Microsoft Academic Search

    U. Fischer; P. Pereslavtsev; S. Hermsmeyer

    2005-01-01

    Neutronic design optimisation analyses based on three-dimensional Monte Carlo calculations have been conducted with the MCNP code as part of the development work for the modular helium cooled pebble bed (HCPB) blanket for a fusion power reactor of the model B-type of the European power plant conceptual study (PPCS). Design variants with breeder pebble beds parallel and perpendicular to the

  12. Preliminary core design studies for the advanced burner reactor over a wide range of conversion ratios

    Microsoft Academic Search

    E. A. Hoffman; W. S. Yang; R. N. Hill

    2008-01-01

    A consistent set of designs for 1000 MWt commercial-scale sodium-cooled Advance Burner Reactors (ABR) have been developed for both metal and oxide-fueled cores with conversion ratios from breakeven (CR=1.0) to fertile-free (CR=0.0). These designs are expected to satisfy thermal and irradiation damage limits based on the currently available data. The very low conversion ratio designs require fuel that is beyond

  13. The design of an intense accelerator-based epithermal neutron beam prototype for BNCT using near-threshold reactions

    Microsoft Academic Search

    Charles L. Lee

    1998-01-01

    Near-threshold boron neutron capture therapy (BNCT) uses proton energies only tens of rev above the (pan) reaction threshold in lithium in order to reduce the moderation requirements of the neutron source. The goals of this research were to prove the feasibility of this near-threshold concept for BNCT applications, using both calculation and experiment, and design a compact neutron source prototype

  14. Modeling and analysis of core debris recriticality during hypothetical severe accidents in the Advanced Neutron Source Reactor

    SciTech Connect

    Taleyarkhan, R.P.; Kim, S.H.; Slater, C.O.; Moses, D.L.; Simpson, D.B.; Georgevich, V.

    1993-05-01

    This paper discusses salient aspects of severe-accident-related recriticality modeling and analysis in the Advanced Neutron Source (ANS) reactor. The development of an analytical capability using the KENO V.A-SCALE system is described including evaluation of suitable nuclear cross-section sets to account for the effects of system geometry, mixture temperature, material dispersion and other thermal-hydraulic conditions. Benchmarking and validation efforts conducted with KENO V.A-SCALE and other neutronic codes against critical experiment data are described. Potential deviations and biases resulting from use of the 16-group Hansen-Roach library are shown. A comprehensive test matrix of calculations to evaluate the threat of a recriticality event in the ANS is described. Strong dependencies on geometry, material constituents, and thermal-hydraulic conditions are described. The introduction of designed mitigative features is described.

  15. Neutron transport with the method of characteristics for 3-D full core boiling water reactor applications

    Microsoft Academic Search

    Justin W. Thomas

    2006-01-01

    The Numerical Nuclear Reactor (NNR) is a code suite that is being developed to provide high-fidelity multi-physics capability for the analysis of light water nuclear reactors. The focus of the work here is to extend the capability of the NNR by incorporation of the neutronics module, DeCART, for Boiling Water Reactor (BWR) applications. The DeCART code has been coupled to

  16. Analysis of Stainless Steel Sandwich Panels with a Metal Foam Core for Lightweight Fan Blade Design

    NASA Technical Reports Server (NTRS)

    Min, James B.; Ghosn, Louis J.; Lerch, Bradley A.; Raj, Sai V.; Holland, Frederic A., Jr.; Hebsur, Mohan G.

    2004-01-01

    The quest for cheap, low density and high performance materials in the design of aircraft and rotorcraft engine fan and propeller blades poses immense challenges to the materials and structural design engineers. The present study investigates the use of a sandwich foam fan blade mae up of solid face sheets and a metal foam core. The face sheets and the metal foam core material were an aerospace grade precipitation hardened 17-4 PH stainless steel with high strength and high toughness. The resulting structures possesses a high stiffness while being lighter than a similar solid construction. The material properties of 17-4 PH metal foam are reviewed briefly to describe the characteristics of sandwich structure for a fan blade application. A vibration analysis for natural frequencies and a detailed stress analysis on the 17-4 PH sandwich foam blade design for different combinations of kin thickness and core volume are presented with a comparison to a solid titanium blade.

  17. Core Noise: Implications of Emerging N+3 Designs and Acoustic Technology Needs

    NASA Technical Reports Server (NTRS)

    Hultgren, Lennart S.

    2011-01-01

    This presentation is a summary of the core-noise implications of NASA's primary N+3 aircraft concepts. These concepts are the MIT/P&W D8.5 Double Bubble design, the Boeing/GE SUGAR Volt hybrid gas-turbine/electric engine concept, the NASA N3-X Turboelectric Distributed Propulsion aircraft, and the NASA TBW-XN Truss-Braced Wing concept. The first two are future concepts for the Boeing 737/Airbus A320 US transcontinental mission of 180 passengers and a maximum range of 3000 nm. The last two are future concepts for the Boeing 777 transpacific mission of 350 passengers and a 7500 nm range. Sections of the presentation cover: turbofan design trends on the N+1.5 time frame and the already emerging importance of core noise; the NASA N+3 concepts and associated core-noise challenges; the historical trends for the engine bypass ratio (BPR), overall pressure ratio (OPR), and combustor exit temperature; and brief discussion of a noise research roadmap being developed to address the core-noise challenges identified for the N+3 concepts. The N+3 conceptual aircraft have (i) ultra-high bypass ratios, in the rage of 18 - 30, accomplished by either having a small-size, high-power-density core, an hybrid design which allows for an increased fan size, or by utilizing a turboelectric distributed propulsion design; and (ii) very high OPR in the 50 - 70 range. These trends will elevate the overall importance of turbomachinery core noise. The N+3 conceptual designs specify the need for the development and application of advanced liners and passive and active control strategies to reduce the core noise. Current engineering prediction of core noise uses semi-empirical methods based on older turbofan engines, with (at best) updates for more recent designs. The models have not seen the same level of development and maturity as those for fan and jet noise and are grossly inadequate for the designs considered for the N+3 time frame. An aggressive program for the development of updated noise prediction tools for integrated core assemblies as well as and strategies for noise reduction and control is needed in order to meet the NASA N+3 noise goals. The NASA Fundamental Aeronautics Program has the principal objective of overcoming today's national challenges in air transportation. The SFW Reduced-Perceived-Noise Technical Challenge aims to develop concepts and technologies to dramatically reduce the perceived aircraft noise outside of airport boundaries. This reduction of aircraft noise is critical to enabling the anticipated large increase in future air traffic.

  18. Baseline Design Compliance Matrix for the Rotary Mode Core Sampling System

    SciTech Connect

    LECHELT, J.A.

    2000-10-17

    The purpose of the design compliance matrix (DCM) is to provide a single-source document of all design requirements associated with the fifteen subsystems that make up the rotary mode core sampling (RMCS) system. It is intended to be the baseline requirement document for the RMCS system and to be used in governing all future design and design verification activities associated with it. This document is the DCM for the RMCS system used on Hanford single-shell radioactive waste storage tanks. This includes the Exhauster System, Rotary Mode Core Sample Trucks, Universal Sampling System, Diesel Generator System, Distribution Trailer, X-Ray Cart System, Breathing Air Compressor, Nitrogen Supply Trailer, Casks and Cask Truck, Service Trailer, Core Sampling Riser Equipment, Core Sampling Support Trucks, Foot Clamp, Ramps and Platforms and Purged Camera System. Excluded items are tools such as light plants and light stands. Other items such as the breather inlet filter are covered by a different design baseline. In this case, the inlet breather filter is covered by the Tank Farms Design Compliance Matrix.

  19. Design of the 50 kW neutron converter for SPIRAL2 facility

    NASA Astrophysics Data System (ADS)

    Avilov, M. S.; Tecchio, L. B.; Titov, A. T.; Tsybulya, V. S.; Zhmurikov, E. I.

    2010-06-01

    SPIRAL2 is a facility for the study of fundamental nuclear physics and multidisciplinary research. SPIRAL2 represents a major advance for research on exotic nuclei. The radioactive ion beam (RIB) production system is comprised of a neutron converter, a target and an ion source. This paper is dedicated to the designing of the 50 kW neutron converter for the SPIRAL2 facility. Among the different variants of the neutron converter, the one based on a rotating solid disk seems quite attractive due to its safety, ease in production and relatively low cost. Dense graphite used as the converter's material allows the production of high-intensity neutron flux and, at the same time, the heat removal from the converter by means of radiation cooling. Thermo-mechanical simulations performed in order to determine the basic geometry and physical characteristics of the neutron production target for SPIRAL2 facility, to define the appropriate beam power distribution, and to predict the target behaviour under the deuteron beam of nominal parameters (40 MeV, 1.2 mA, 50 kW) are presented. To study the main physical and mechanical properties and serviceability under operating conditions, several kinds of graphite have been analyzed and tested. The paper reports the results of such measurements. Radiation damage is the most important issue for the application of graphite as neutron converter. It is well known that the thermal conductivity of the neutron-irradiated graphite is reduced by a factor of 10 from the initial value after irradiation. Difference in volume expansions between the matrix and the fiber results in serious damage of neutron-irradiated C/C composites. Calculations showed that at high temperature the effect of neutron radiation is not so critical and that the change in thermal conductivity does not prevent the use of graphite as neutron converter.

  20. Measurements of actinide-fission product yields in Caliban and Prospero metallic core reactor fission neutron fields

    SciTech Connect

    Casoli, P.; Authier, N. [CEA, Centre de Valduc, 21120 Is-sur-Tille (France); Laurec, J.; Bauge, E.; Granier, T. [CEA, Centre DIF, 91297 Arpajon (France)

    2011-07-01

    In the 1970's and early 1980's, an experimental program was performed on the facilities of the CEA Valduc Research Center to measure several actinide-fission product yields. Experiments were, in particular, completed on the Caliban and Prospero metallic core reactors to study fission-neutron-induced reactions on {sup 233}U, {sup 235}U, and {sup 239}Pu. Thick actinide samples were irradiated and the number of nuclei of each fission product was determined by gamma spectrometry. Fission chambers were irradiated simultaneously to measure the numbers of fissions in thin deposits of the same actinides. The masses of the thick samples and the thin deposits were determined by mass spectrometry and alpha spectrometry. The results of these experiments will be fully presented in this paper for the first time. A description of the Caliban and Prospero reactors, their characteristics and performances, and explanations about the experimental approach will also be given in the article. A recent work has been completed to analyze and reinterpret these measurements and particularly to evaluate the associated uncertainties. In this context, calculations have also been carried out with the Monte Carlo transport code Tripoli-4, using the published benchmarked Caliban description and a three-dimensional model of Prospero, to determine the average neutron energy causing fission. Simulation results will be discussed in this paper. Finally, new fission yield measurements will be proposed on Caliban and Prospero reactors to strengthen the results of the first experiments. (authors)

  1. The design and performance of IceCube DeepCore

    NASA Astrophysics Data System (ADS)

    Abbasi, R.; Abdou, Y.; Abu-Zayyad, T.; Ackermann, M.; Adams, J.; Aguilar, J. A.; Ahlers, M.; Allen, M. M.; Altmann, D.; Andeen, K.; Auffenberg, J.; Bai, X.; Baker, M.; Barwick, S. W.; Bay, R.; Bazo Alba, J. L.; Beattie, K.; Beatty, J. J.; Bechet, S.; Becker, J. K.; Becker, K.-H.; Benabderrahmane, M. L.; BenZvi, S.; Berdermann, J.; Berghaus, P.; Berley, D.; Bernardini, E.; Bertrand, D.; Besson, D. Z.; Bindig, D.; Bissok, M.; Blaufuss, E.; Blumenthal, J.; Boersma, D. J.; Bohm, C.; Bose, D.; Böser, S.; Botner, O.; Brown, A. M.; Buitink, S.; Caballero-Mora, K. S.; Carson, M.; Chirkin, D.; Christy, B.; Clevermann, F.; Cohen, S.; Colnard, C.; Cowen, D. F.; Cruz Silva, A. H.; D'Agostino, M. V.; Danninger, M.; Daughhetee, J.; Davis, J. C.; De Clercq, C.; Degner, T.; Demirörs, L.; Descamps, F.; Desiati, P.; de Vries-Uiterweerd, G.; DeYoung, T.; Díaz-Vélez, J. C.; Dierckxsens, M.; Dreyer, J.; Dumm, J. P.; Dunkman, M.; Eisch, J.; Ellsworth, R. W.; Engdegård, O.; Euler, S.; Evenson, P. A.; Fadiran, O.; Fazely, A. R.; Fedynitch, A.; Feintzeig, J.; Feusels, T.; Filimonov, K.; Finley, C.; Fischer-Wasels, T.; Fox, B. D.; Franckowiak, A.; Franke, R.; Gaisser, T. K.; Gallagher, J.; Gerhardt, L.; Gladstone, L.; Glüsenkamp, T.; Goldschmidt, A.; Goodman, J. A.; Góra, D.; Grant, D.; Griesel, T.; Groß, A.; Grullon, S.; Gurtner, M.; Ha, C.; Haj Ismail, A.; Hallgren, A.; Halzen, F.; Han, K.; Hanson, K.; Heinen, D.; Helbing, K.; Hellauer, R.; Hickford, S.; Hill, G. C.; Hoffman, K. D.; Hoffmann, B.; Homeier, A.; Hoshina, K.; Huelsnitz, W.; Hülß, J.-P.; Hulth, P. O.; Hultqvist, K.; Hussain, S.; Ishihara, A.; Jacobi, E.; Jacobsen, J.; Japaridze, G. S.; Johansson, H.; Kampert, K.-H.; Kappes, A.; Karg, T.; Karle, A.; Kenny, P.; Kiryluk, J.; Kislat, F.; Klein, S. R.; Köhne, J.-H.; Kohnen, G.; Kolanoski, H.; Köpke, L.; Koskinen, D. J.; Kowalski, M.; Kowarik, T.; Krasberg, M.; Kroll, G.; Kurahashi, N.; Kuwabara, T.; Labare, M.; Laihem, K.; Landsman, H.; Larson, M. J.; Lauer, R.; Lünemann, J.; Madsen, J.; Marotta, A.; Maruyama, R.; Mase, K.; Matis, H. S.; Meagher, K.; Merck, M.; Mészáros, P.; Meures, T.; Miarecki, S.; Middell, E.; Milke, N.; Miller, J.; Montaruli, T.; Morse, R.; Movit, S. M.; Nahnhauer, R.; Nam, J. W.; Naumann, U.; Nygren, D. R.; Odrowski, S.; Olivas, A.; Olivo, M.; O'Murchadha, A.; Panknin, S.; Paul, L.; Pérez de los Heros, C.; Petrovic, J.; Piegsa, A.; Pieloth, D.; Porrata, R.; Posselt, J.; Price, P. B.; Przybylski, G. T.; Rawlins, K.; Redl, P.; Resconi, E.; Rhode, W.; Ribordy, M.; Richman, M.; Rodrigues, J. P.; Rothmaier, F.; Rott, C.; Ruhe, T.; Rutledge, D.; Ruzybayev, B.; Ryckbosch, D.; Sander, H.-G.; Santander, M.; Sarkar, S.; Schatto, K.; Schmidt, T.; Schönwald, A.; Schukraft, A.; Schultes, A.; Schulz, O.; Schunck, M.; Seckel, D.; Semburg, B.; Seo, S. H.; Sestayo, Y.; Seunarine, S.; Silvestri, A.; Spiczak, G. M.; Spiering, C.; Stamatikos, M.; Stanev, T.; Stezelberger, T.; Stokstad, R. G.; Stößl, A.; Strahler, E. A.; Ström, R.; Stüer, M.; Sullivan, G. W.; Swillens, Q.; Taavola, H.; Taboada, I.; Tamburro, A.; Tepe, A.; Ter-Antonyan, S.; Tilav, S.; Toale, P. A.; Toscano, S.; Tosi, D.; van Eijndhoven, N.; Vandenbroucke, J.; Van Overloop, A.; van Santen, J.; Vehring, M.; Voge, M.; Walck, C.; Waldenmaier, T.; Wallraff, M.; Walter, M.; Weaver, Ch.; Wendt, C.; Westerhoff, S.; Whitehorn, N.; Wiebe, K.; Wiebusch, C. H.; Williams, D. R.; Wischnewski, R.; Wissing, H.; Wolf, M.; Wood, T. R.; Woschnagg, K.; Xu, C.; Xu, D. L.; Xu, X. W.; Yanez, J. P.; Yodh, G.; Yoshida, S.; Zarzhitsky, P.; Zoll, M.

    2012-05-01

    The IceCube neutrino observatory in operation at the South Pole, Antarctica, comprises three distinct components: a large buried array for ultrahigh energy neutrino detection, a surface air shower array, and a new buried component called DeepCore. DeepCore was designed to lower the IceCube neutrino energy threshold by over an order of magnitude, to energies as low as about 10 GeV. DeepCore is situated primarily 2100 m below the surface of the icecap at the South Pole, at the bottom center of the existing IceCube array, and began taking physics data in May 2010. Its location takes advantage of the exceptionally clear ice at those depths and allows it to use the surrounding IceCube detector as a highly efficient active veto against the principal background of downward-going muons produced in cosmic-ray air showers. DeepCore has a module density roughly five times higher than that of the standard IceCube array, and uses photomultiplier tubes with a new photocathode featuring a quantum efficiency about 35% higher than standard IceCube PMTs. Taken together, these features of DeepCore will increase IceCube's sensitivity to neutrinos from WIMP dark matter annihilations, atmospheric neutrino oscillations, galactic supernova neutrinos, and point sources of neutrinos in the northern and southern skies. In this paper we describe the design and initial performance of DeepCore.

  2. The Design and Performance of IceCube DeepCore

    NASA Technical Reports Server (NTRS)

    Stamatikos, M.

    2012-01-01

    The IceCube neutrino observatory in operation at the South Pole, Antarctica, comprises three distinct components: a large buried array for ultrahigh energy neutrino detection, a surface air shower array, and a new buried component called DeepCore. DeepCore was designed to lower the IceCube neutrino energy threshold by over an order of magnitude, to energies as low as about 10 GeV. DeepCore is situated primarily 2100 m below the surface of the icecap at the South Pole, at the bottom center of the existing IceCube array, and began taking pbysics data in May 2010. Its location takes advantage of the exceptionally clear ice at those depths and allows it to use the surrounding IceCube detector as a highly efficient active veto against the principal background of downward-going muons produced in cosmic-ray air showers. DeepCore has a module density roughly five times higher than that of the standard IceCube array, and uses photomultiplier tubes with a new photocathode featuring a quantum efficiency about 35% higher than standard IceCube PMTs. Taken together, these features of DeepCore will increase IceCube's sensitivity to neutrinos from WIMP dark matter annihilations, atmospheric neutrino oscillations, galactic supernova neutrinos, and point sources of neutrinos in the northern and southern skies. In this paper we describe the design and initial performance of DeepCore.

  3. Rational design of a super core promoter that enhances gene expression.

    PubMed

    Juven-Gershon, Tamar; Cheng, Susan; Kadonaga, James T

    2006-11-01

    Transcription is a critical component in the expression of genes. Here we describe the design and analysis of a potent core promoter, termed super core promoter 1 (SCP1), which directs high amounts of transcription by RNA polymerase II in metazoans. SCP1 contains four core promoter motifs-the TATA box, initiator (Inr), motif ten element (MTE) and downstream promoter element (DPE)-in a single promoter, and is distinctly stronger than the cytomegalovirus (CMV) IE1 and adenovirus major late (AdML) core promoters both in vitro and in vivo. Each of the four core promoter motifs is needed for full SCP1 activity. SCP1 is bound efficiently by TFIID and exhibits a high propensity to form productive transcription complexes. SCP1 and related super core promoters (SCPs) with multiple core promoter motifs will be useful for the biophysical analysis of TFIID binding to DNA, the biochemical investigation of the transcription process and the enhancement of gene expression in cells. PMID:17124735

  4. Modeling and design of a reload PWR core for a 48-month fuel cycle

    SciTech Connect

    McMahon, M.V.; Driscoll, M.J.; Todreas, N.E. [Massachusetts Inst. of Tech., Cambridge, MA (United States)

    1997-05-01

    The objective of this research was to use state-of-the-art nuclear and fuel performance packages to evaluate the feasibility and costs of a 48 calendar month core in existing pressurized water reactor (PWR) designs, considering the full range of practical design and economic considerations. The driving force behind this research is the desire to make nuclear power more economically competitive with fossil fuel options by expanding the scope for achievement of higher capacity factors. Using CASMO/SIMULATE, a core design with fuel enriched to 7{sup w}/{sub o} U{sup 235} for a single batch loaded, 48-month fuel cycle has been developed. This core achieves an ultra-long cycle length without exceeding current fuel burnup limits. The design uses two different types of burnable poisons. Gadolinium in the form of gadolinium oxide (Gd{sub 2}O{sub 3}) mixed with the UO{sub 2} of selected pins is sued to hold down initial reactivity and to control flux peaking throughout the life of the core. A zirconium di-boride (ZrB{sub 2}) integral fuel burnable absorber (IFBA) coating on the Gd{sub 2}O{sub 3}-UO{sub 2} fuel pellets is added to reduce the critical soluble boron concentration in the reactor coolant to within acceptable limits. Fuel performance issues of concern to this design are also outlined and areas which will require further research are highlighted.

  5. Core design and safety studies for a small modular fast reactor

    SciTech Connect

    Yang, W. S.; Cahalan, J. E.; Dunn, F. E. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2006-07-01

    The paper describes the core design and performance characteristics and the safety analysis results for a 50 MWe small modular fast reactor design that was developed jointly by ANL, CEA, and JNC as an international collaborative effort. The main goal in the core design was to achieve a 30-year lifetime with no refueling. In order to minimize the burnup reactivity swing, metal fuel with a high heavy metal volume fraction was selected. To enhance the proliferation resistance and actinide transmutation, all the transuranic (TRU) elements recovered from light water reactor spent fuel were used in a ternary alloy form of U-TRU-10Zr. A 125 MWt core design was developed, for which the burnup reactivity swing was only 1.6$ over the 30-year core lifetime. The average discharge burnup was 87 MWd/kg, and the maximum sodium void worth was 4.65$. The evaluated reactivity coefficients provided sufficient negative feedbacks. Shutdown margins of control systems were confirmed. Steady-state thermal-hydraulic analysis results showed that peak 2{sigma} cladding inner-wall and fuel centerline temperatures were less than design limits with sufficient margins. Detailed transient analyses for the total loss of power to reactor and intermediate coolant pumps showed that no fuel damage or cladding failure would occur, even when multiple safety systems were assumed to malfunction. (authors)

  6. The advanced neutron source--designing to meet the needs of the user community

    SciTech Connect

    Peretz, F.J. (Oak Ridge National Lab., TN (USA))

    1989-01-01

    The Advanced Neutron Source (ANS) is to be a multi-purpose neutron research center, constructed around a high-flux reactor now being designed at the Oak Ridge National Laboratory (ORNL). Its primary purpose is to place the United States in the forefront of neutron scattering in the twenty-first century. Other research programs include nuclear and fundamental physics, isotopes production, materials irradiation, and analytical chemistry. The Advanced Neutron Source will be a unique and invaluable research tool because of the unprecedented neutron flux available from the high intensity research reactor. But that reactor would be ineffective without world-class research facilities that allow the fullest utilization of the available neutrons. And, in turn, those research facilities will not produce new and exciting science without a broad population of users coming from all parts of the nation, and the world, placed in a simulating environment in which experiments can be effectively conducted, and in which scientific exchange is encouraged. This paper discusses the measures being taken to ensure that the design of the ANS focuses not only on the reactor, but on providing the experiment and user support facilities needed to allow its effective use. 5 refs., 4 figs.

  7. Computational Design of Strain in Core-Shell Nanoparticles for Optimizing Catalytic Activity.

    PubMed

    Moseley, Philip; Curtin, W A

    2015-06-10

    Surface strains in core-shell nanoparticles modify catalytic activity. Here, a continuum-based strategy enables accurate surface-strain-based screening and design of core-shell systems using minimal input as a means to enhance catalytic activity. The approach is validated here for Pt shells on CuxPt(1-x) cores and used to interpret experimental results on the oxygen reduction reaction in the same system. The analysis shows that precise control of particle sizes and shell thicknesses is required to achieve peak activity, rationalizing the limited increases in activity observed in experiments. The method is also applied to core-shell nanorods to demonstrate its wide applicability. PMID:25965405

  8. Research advances in polymer emulsion based on "core-shell" structure particle design.

    PubMed

    Ma, Jian-zhong; Liu, Yi-hong; Bao, Yan; Liu, Jun-li; Zhang, Jing

    2013-09-01

    In recent years, quite many studies on polymer emulsions with unique core-shell structure have emerged at the frontier between material chemistry and many other fields because of their singular morphology, properties and wide range of potential applications. Organic substance as a coating material onto either inorganic or organic internal core materials promises an unparalleled opportunity for enhancement of final functions through rational designs. This contribution provides a brief overview of recent progress in the synthesis, characterization, and applications of both inorganic-organic and organic-organic polymer emulsions with core-shell structure. In addition, future research trends in polymer composites with core-shell structure are also discussed in this review. PMID:23726300

  9. Thermal-Hydraulic Design of the Accelerator Production of Tritium Tungsten Neutron Source

    Microsoft Academic Search

    Kemal O. Pasamehmetoglu; Gordon J. Jr. Willcutt; Jay S. Elson; Donald A. Siebe

    2000-01-01

    The thermal-hydraulic design of the accelerator production of tritium (APT) tungsten neutron source is presented. A carefully engineered thermal-hydraulic design is required to remove the deposited power effectively during normal operations and remove the decay power during plant shutdown and postulated accidents. For steady-state operations and operational and anticipated transients, the design criterion is to maintain single-phase flow conditions with

  10. Design of a neutron detector for the GAMMA-400 space experiment

    NASA Astrophysics Data System (ADS)

    Taraskin, Anton

    Neutron detectors could be effectively applied to gamma astronomy increasing a factor of proton rejection in orbital gamma-ray telescopes. This article describes design and capabilities of a certain neutron detector which will be used as an additional instrument of separation between electromagnetic and nuclear cascades in the GAMMA-400 orbital gamma-ray observatory. This procedure is crucially important to exterminate proton background during any space measurements. The neutron detector operates in counting mode and is position sensitive. It contains two layers of ZnS(Tl) + (6) LiF scintillator and several layers of organic moderator. Calculated efficiency for Pu-Be neutron spectrum is about 15%. Detector uses organic glass layers and SiPMs to gather light from an event. Identification of a primordial particle is a result of time, spatial and quantitative analysis of a signal.

  11. Preliminary shielding analysis in support of the CSNS target station shutter neutron beam stop design

    NASA Astrophysics Data System (ADS)

    Zhang, Bin; Chen, Yi-Xue; Wang, Wei-Jin; Yang, Shou-Hai; Wu, Jun; Yin, Wen; Liang, Tian-Jiao; Jia, Xue-Jun

    2011-08-01

    The construction of China Spallation Neutron Source (CSNS) has been initiated in Dongguan, Guangdong, China. Thus a detailed radiation transport analysis of the shutter neutron beam stop is of vital importance. The analyses are performed using the coupled Monte Carlo and multi-dimensional discrete ordinates method. The target of calculations is to optimize the neutron beamline shielding design to guarantee personal safety and minimize cost. Successful elimination of the primary ray effects via the two-dimensional uncollided flux and the first collision source methodology is also illustrated. Two-dimensional dose distribution is calculated. The dose at the end of the neutron beam line is less than 2.5 ?Sv/h. The models have ensured that the doses received by the hall staff members are below the standard limit required.

  12. Design study for MOX fuel rod scanner for ATR fuel fabrication. Phase I: Design of active neutron scanner. Phase II: Design of passive neutron scanner. Phase III: Design of passive gamma-ray scanner

    SciTech Connect

    Griffith, G.W.; Menlove, H.O.

    1997-09-01

    An active neutron fuel-rod scanner has been designed for the assay of fissile materials in mixed oxide fuel rods. A {sup 252}Cf source is located at the center of the scanner very near the through-hole for the fuel rods. Spontaneous fission neutrons from the californium are moderated and induce fissions within the passing fuel rod. The rod continues past a combined gamma-ray and neutron shield where delayed gamma rays above 1 MeV are detected. The authors used the Monte Carlo neutron-photon (MCNP) code to design the scanner and review optimum materials and geometries. Applications of the rod scanner include accountability of fissile material for safeguards applications, quality control of the fissile content in a fuel rod, and the verification of reactivity potential for mixed oxide fuels. A passive neutron fuel-rod scanner has been designed for the assay of the plutonium in mixed oxide fuel rods. The {sup 240}Pu-effective is measured by counting the spontaneous fission neutrons using a high-efficiency thermal-neutron detector. This passive neutron detector would be combined with a high-resolution gamma-ray system (HRGS) measurement to obtain the total plutonium from the plutonium isotopic ratios. A passive gamma-ray scanner has been designed for the measurement of the {sup 241}Am and plutonium uniformity in mixed oxide fuel rods. The passive gamma-ray emissions from {sup 241}Am (60 keV) and plutonium (150-400 keV) are used to verify the unformity of the fuel enrichment zones and to check for any pellets that are out of specification. The fuel rod is moved through the interior of an NaI(Tl) or a bismuth germanate detector to measure the passive gamma-ray emissions. A tungsten sleeve collimator is used in the through-hole to improve the pellet-to-pellet spatial resolution. The same detector is used to verify the plutonium uniformity in the pellets with a 13-mm tungsten collimator. The low-resolution passive gamma system would be used in the unattended mode.

  13. Coupled Neutronics Thermal-Hydraulic Solution of a Full-Core PWR Using VERA-CS

    SciTech Connect

    Clarno, Kevin T [ORNL; Palmtag, Scott [ORNL; Davidson, Gregory G [ORNL; Salko, Robert K [ORNL; Evans, Thomas M [ORNL; Turner, John A [ORNL; Belcourt, Kenneth [Sandia National Laboratories (SNL); Hooper, Russell [Sandia National Laboratories (SNL); Schmidt, Rodney [Sandia National Laboratories (SNL)

    2014-01-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) is developing a core simulator called VERA-CS to model operating PWR reactors with high resolution. This paper describes how the development of VERA-CS is being driven by a set of progression benchmark problems that specify the delivery of useful capability in discrete steps. As part of this development, this paper will describe the current capability of VERA-CS to perform a multiphysics simulation of an operating PWR at Hot Full Power (HFP) conditions using a set of existing computer codes coupled together in a novel method. Results for several single-assembly cases are shown that demonstrate coupling for different boron concentrations and power levels. Finally, high-resolution results are shown for a full-core PWR reactor modeled in quarter-symmetry.

  14. Use of Solid Hydride Fuel for Improved long-Life LWR Core Designs

    SciTech Connect

    Greenspan, E

    2006-04-30

    The primary objective of this project was to assess the feasibility of improving the performance of PWR and BWR cores by using solid hydride fuels instead of the commonly used oxide fuel. The primary measure of performance considered is the bus-bar cost of electricity (COE). Additional performance measures considered are safety, fuel bundle design simplicity – in particular for BWR’s, and plutonium incineration capability. It was found that hydride fuel can safely operate in PWR’s and BWR’s without restricting the linear heat generation rate of these reactors relative to that attainable with oxide fuel. A couple of promising applications of hydride fuel in PWR’s and BWR’s were identified: (1) Eliminating dedicated water moderator volumes in BWR cores thus enabling to significantly increase the cooled fuel rods surface area as well as the coolant flow cross section area in a given volume fuel bundle while significantly reducing the heterogeneity of BWR fuel bundles thus achieving flatter pin-by-pin power distribution. The net result is a possibility to significantly increase the core power density – on the order of 30% and, possibly, more, while greatly simplifying the fuel bundle design. Implementation of the above modifications is, though, not straightforward; it requires a design of completely different control system that could probably be implemented only in newly designed plants. It also requires increasing the coolant pressure drop across the core. (2) Recycling plutonium in PWR’s more effectively than is possible with oxide fuel by virtue of a couple of unique features of hydride fuel – reduced inventory of U-238 and increased inventory of hydrogen. As a result, the hydride fuelled core achieves nearly double the average discharge burnup and the fraction of the loaded Pu it incinerates in one pass is double that of the MOX fuel. The fissile fraction of the Pu in the discharged hydride fuel is only ~2/3 that of the MOX fuel and the discharged hydride fuel is more proliferation resistant. Preliminary feasibility assessment indicates that by replacing some of the ZrH1.6 by ThH2 it will be possible to further improve the plutonium incineration capability of PWR’s. Other possibly promising applications of hydride fuel were identified but not evaluated in this work. A number of promising oxide fueled PWR core designs were also found as spin-offs of this study: (1) The optimal oxide fueled PWR core design features smaller fuel rod diameter of D=6.5 mm and a larger pitch-to-diameter ratio of P/D=1.39 than presently practiced by industry – 9.5mm and 1.326. This optimal design can provide a 30% increase in the power density and a 24% reduction in the cost of electricity (COE) provided the PWR could be designed to have the coolant pressure drop across the core increased from the reference 29 psia to 60 psia. (2) Using wire wrapped oxide fuel rods in hexagonal fuel assemblies it is possible to design PWR cores to operate at 54% higher power density than the reference PWR design that uses grid spacers and a square lattice, provided 60 psia coolant pressure drop across the core could be accommodated. Uprating existing PWR’s to use such cores could result in 40% reduction in the COE. The optimal lattice geometry is D = 8.08 mm and P/D = 1.41. The most notable advantages of wire wraps over grid spacers are their significant lower pressure drop, higher critical heat flux and improved vibrations characteristics.

  15. Designing high frequency ac inductors using ferrite and Molypermalloy Powder Cores (MPP)

    NASA Technical Reports Server (NTRS)

    Mclyman, W. T.; Wagner, A. P.

    1985-01-01

    The major considerations in the design of high frequency ac inductors are reviewed. Two methods for designing the inductor: the area product method and the core geometry method, are presented. The two major effects of the inductor air gap, fringing flux power loss and increase of inductance, are discussed. Equations for the inductor design and a step-by-step design procedure are given. The use of a lumped air gap or a distributed air gap are discussed and a comparison of the losses resulting from these gaps, together with experimental results are presented.

  16. Verifying IP-Core based System-On-Chip Designs Pankaj Chauhan, Edmund M. Clarke, Yuan Lu and Dong Wang

    E-print Network

    Clarke, Edmund M.

    Verifying IP-Core based System-On-Chip Designs Pankaj Chauhan, Edmund M. Clarke, Yuan Lu and Dong and for all, the standard bus interconnecting IP Cores in the system . The next task is to verify the glue logic, which connects the IP Cores to the buses. Finally, using the verified bus protocols and the IP

  17. Design and Realization of Embedded General Intelligent Controller Based on Microprocessor with ARM Core

    Microsoft Academic Search

    Qian Zhang; Qiusheng Zheng; Guoqiang Wang

    2007-01-01

    The realization method of designing a sort of controller based on microprocessor with ARM core was discussed in this paper, and the project of adopting intelligent control algorithm library as control strategy of embedded intelligent controller was proposed, a kind of improved single NN-PID control algorithm and a human-simulation intelligent control algorithm were given, too. The hardware circuit and structure

  18. Narrative Plus: Designing and Implementing the Common Core State Standards with the Gift Essay

    ERIC Educational Resources Information Center

    Chandler-Olcott, Kelly; Zeleznik, John

    2013-01-01

    The authors of this article describe their inquiry into implementation of the writing-focused Common Core State Standards in a co-taught English 9 class in an urban school. They describe instructional moves designed to increase student success with an assignment called the Gift Essay, with particular focus on planning and other organizational…

  19. Scan chain design for test time reduction in core-based ICs

    Microsoft Academic Search

    Joep Aerts; Erik Jan Marinissen

    1998-01-01

    The size of the test vector set forms a significant factor in the overall production costs of ICs, as it defines the test application time and the required pin memory size of the test equipment. Large core-based ICs often require a very large test vector set for a high test coverage. This paper deals with the design of scan chains

  20. Designing Scalable FPGA-Based Reduction Circuits Using Pipelined Floating-Point Cores

    E-print Network

    Prasanna, Viktor K.

    Designing Scalable FPGA-Based Reduction Circuits Using Pipelined Floating-Point Cores Ling Zhuo California Los Angeles, CA 90089-2562 {lzhuo,grm,prasanna}@usc.edu Abstract The use of pipelined floating associated with pipelined floating-point reduction circuits can limit the scalabil- ity or severely reduce

  1. Spring design for use in the core of a nuclear reactor

    DOEpatents

    Willard, Jr., H. James (Bethel Park, PA)

    1993-01-01

    A spring design particularly suitable for use in the core of a nuclear reactor includes one surface having a first material oriented in a longitudinal direction, and another surface having a second material oriented in a transverse direction. The respective surfaces exhibit different amounts of irraditation induced strain.

  2. Core Curriculum / 18 credits IS 6420 Database Theory and Design 3

    E-print Network

    Tipple, Brett

    Core Curriculum / 18 credits IS 6420 Database Theory and Design 3 IS 6410 Information Systems Analytics 3 IS 6571 IT Forensics 3 IS 6570 IT Security 3 ACCTG 6520 IT Risks and Controls 3 The MSIS program ­ Fall, Spring and Summer semesters MASTER OF SCIENCE IN INFORMATION SYSTEMS MSIS801-581-7785 / 877

  3. Design and Implementation of the National Institute of Environmental Health Sciences Dublin Core Metadata Schema

    Microsoft Academic Search

    W. Davenport Robertson; Ellen M. Leadem; Jed Dube; Jane Greenberg

    2001-01-01

    The National Institute of Environmental Health Sciences (NIEHS) has formed a team to design and implement a Dublin Core-based metadata schema to enhance the public's ability to retrieve pertinent public health information on the organization's Web site. The team decided to use the DC schema because it is a de facto standard and because of its flexibility. With a little

  4. Coarse-grained parallel genetic algorithm applied to a nuclear reactor core design optimization problem

    Microsoft Academic Search

    Cláudio M. N. A. Pereira; Celso M. F. Lapa

    2003-01-01

    This work extends the research related to genetic algorithms (GA) in core design optimization problems, which basic investigations were presented in previous work. Here we explore the use of the Island Genetic Algorithm (IGA), a coarse-grained parallel GA model, comparing its performance to that obtained by the application of a traditional non-parallel GA. The optimization problem consists on adjusting several

  5. AMPLIFIER DESIGN IMPLEMENTING HOLLOW-CORE PHOTONIC BANDGAP FIBER FOR FIBER-LASER BASED

    E-print Network

    Washburn, Brian

    AMPLIFIER DESIGN IMPLEMENTING HOLLOW-CORE PHOTONIC BANDGAP FIBER FOR FIBER-LASER BASED INFRARED-locked erbium-doped fiber lasers typically require an external amplifier since the pulses directly from signal to locking electronics Amplifier HNLF Mode-locked fiber laser Amplifier HNLF Mode-locked fiber

  6. AMPLIFIER DESIGN IMPLEMENTING HOLLOW-CORE PHOTONIC BANDGAP FIBER FOR FIBER-LASER BASED INFRARED

    E-print Network

    Washburn, Brian

    AMPLIFIER DESIGN IMPLEMENTING HOLLOW-CORE PHOTONIC BANDGAP FIBER FOR FIBER-LASER BASED INFRARED-doped fiber lasers typically require an external amplifier since the pulses directly from the laser have Mode-locked erbium-doped fiber lasers produce the ideal frequency comb for infrared optical frequency

  7. Design of a 64-bit, 100 MIPS microprocessor core IC for hybrid CMOS-SEED technology

    Microsoft Academic Search

    F. E. Kiamilev; J. S. Lambirth; R. G. Rozier; A. V. Krishnamoorthy

    1996-01-01

    We describe the design of a hybrid CMOS-SEED 64-bit microprocessor core IC with 192 optical I\\/Os. This 3.5 mm2 IC was fabricated and electrically tested at 100 MHz with a performance of 100 million 64-bit instructions per second (MIPS). The processor design includes a 64-bit arithmetic-logic unit (ALU) which implements 16 logic and 32 fixed-point arithmetic functions. A 1 cm2

  8. Design of a hohlraum-driven exploding pusher capsule experiment for NIF neutron diagnostic calibration

    NASA Astrophysics Data System (ADS)

    Berzak Hopkins, L.; Callahan, D.; Divol, L.; Le Pape, S.; Meezan, N.; Masse, L.

    2012-10-01

    Neutron diagnostics are a critical component of the National Ignition Facility and measure key parameters of capsule performance such as neutron yield, ion temperature, neutron bang time, and down scattered ratio. Therefore, accurate calibration of these diagnostics is essential. Such calibration requires high DT yield (greater than 1.e14) with low ?R as well as a symmetric implosion -- an implosion with small non-radial velocities remaining in the fuel during the burn phase. In order to meet these requirements, we have used the radiation-hydrodynamics code HYDRA to design an indirect-drive exploding pusher capsule experiment, driven with a 2.5 ns pulse in a vacuum hohlraum. Features of this design will be presented as well as its feasibility for symmetry control.

  9. Shielding design of a treatment room for an accelerator-based neutron source for BNCT

    SciTech Connect

    Evans, J.F.; Blue, T.E. [Ohio State Univ., Columbus, OH (United States)

    1995-12-31

    For several years, research has been ongoing in the Ohio State University (OSU) Nuclear Engineering Program toward the development of an accelerator-based irradiation facility (ANIF) neutron source for boron neutron capture therapy (BNCT). The ANIF, which is planned to be built in a hospital, has been conceptually designed and analyzed. After Qu, an OSU researcher, determined that the shielding design of a 6-MV X-ray treatment room was inadequate to protect personnel from an accelerator neutron source operating at 30 mA, we decided to analyze and determine the shielding requirements of a treatment room for an ANIF. We determined the amount of shielding that would be sufficient to protect facility personnel from excessive radiation exposure caused by operation of the accelerator at 30 mA.

  10. Design, calibration and testing of the NRCAM fast neutron spectrometry system

    Microsoft Academic Search

    F. Abbasi Davani; R. Koohi-Fayegh; H. Afarideh; G. R. Etaati; G. R. Aslani

    2003-01-01

    This paper describes the design, calibration and testing of all aspects of an NE-213 detection system which was built for measuring neutron spectra from 1 to 30MeV. To cover the whole energy range two detectors are required. A smaller detector is usually used for the lower energy range, but for the upper energy limit, in which we are interested, a

  11. Design and performance considerations for perforated semiconductor thermal-neutron detectors

    E-print Network

    Shultis, J. Kenneth

    rights reserved. 1. Introduction Semiconductor radiation detectors composed of diode devices coatedDesign and performance considerations for perforated semiconductor thermal-neutron detectors J.K. Shultis, D.S. McGregor à S.M.A.R.T. Laboratory, Department of Mechanical and Nuclear Engineering, 3002

  12. Experimental ferrite core circuit analysis and design applied to an analog/digital converter 

    E-print Network

    Hughes, Robert William

    1965-01-01

    EXPKRflIENTAL FERRITE CORE CIRCUIT ANALYSIS A/4i) CESIGN APPLI ED' TO AN ANALOG+I QITAL -CCWERTER I I 'O' Thaaka RCSERT, NI. LL'I AN HQQAKS:-. , ";, , '. . :. . '. -:. '"":, , ':-. ', . -' ". ' t, . I I I' I, ' . Sobalttad, ta tb ~ Qra... , Na'fir. Sabjaat& ' Eiaktr leal EnCInaaring . EXPERIIIENTAL FERRITE CORE CIRCUIT ANALYSIS AND DESIGN APPLIED TO 'AN AMALGG/DIGITAL CONVERTER f 1 A, . Thea I'a t ROBERT Wl LI. I Al'Jl. HUGHES (Goober, ) ' (Me@bar) ' A'yprived . a'o . . to a4...

  13. Fuel and Core Design for Long Operating Cycle Simplified BWR (LSBWR)

    SciTech Connect

    Noriyuki Yoshida; Kouji Hiraiwa; Mikihide Nakamaru; Hideaki Heki [Toshiba Corporation, Isogo Nuclear Engineering Center, 8, Shinsugita-cho, Isogo-ku, Yokohama 235-8523 (Japan)

    2002-07-01

    This paper describes an innovative core concept currently being developed for long operating cycle simplified BWR (LSBWR). The LSBWR adopts the long cycle operation (15 years) for the elimination of the fuel pool and the refueling machines and for the capacity usage ratio improvement. To achieve long cycle operation, a combination of enriched gadolinium and 0.7- times sized small bundle with peripheral-positioned gadolinium rod is adopted as a key design concept. A nuclear design for fuel bundle has been determined based on three dimensional nuclear and thermal hydraulic calculation. A core performance has been evaluated based on this bundle design and thermal performance and reactivity characteristics indicated preferable value. (authors)

  14. A low overhead design for testability and test generation technique for core-based systems-on-a-chip

    Microsoft Academic Search

    Indradeep Ghosh; Niraj K. Jha; Sujit Dey

    1999-01-01

    In a fundamental paradigm shift in system design, entire systems are being built on a single chip, using multiple embedded cores. Though the newest system design methodology has several advantages in terms of time-to-market and system cost, testing such core-based systems is difficult, mainly due to the problem of justifying test sequences at the inputs of a core embedded deep

  15. Probabilistic risk assessment in the design of the Advanced Neutron Source

    SciTech Connect

    Harrington, R.M.; Ramsey, C.T. [Oak Ridge National Lab., TN (United States); Fullwood, R.R. [Brookhaven National Lab., Upton, NY (United States)

    1994-09-01

    Probabilistic risk assessment (PRA) has been used extensively in the design of the Advanced Neutron Source (ANS) reactor to safety risk and enhance availability. Design feautures incorporated to minimize risk include submerged primary coolant piping, circulation cooling capability, and dual independent and diverse shutdown systems. The recently completed Level I, Phase 1 PRA shows that the risk dominating event sequence initiator is now blockage; a program to minimize this identified risk is described.

  16. Challenges in the development of high-fidelity LWR core neutronics tools

    SciTech Connect

    Smith, K.; Forget, B. [Massachusetts Institute of Technology, 77 Massachusetts Avenue, Cambridge MA 02139 (United States)

    2013-07-01

    Modern computing has made possible the solution of extremely large-scale reactor simulations, and the literature has numerous examples of high-resolution methods (often Monte Carlo) applied to full-core reactor problems. However, there are currently no examples in the literature of application of such 'High-Fidelity' or 'First Principles' methods to operating Light Water Reactor (LWR) analysis. This paper seeks to remind code developers, project managers, and analysts of the many important aspects of LWR simulation that must be incorporated to produce truly high-fidelity analysis tools. The authors offer a monetary prize to the first person (or group) that successfully solves a new two-cycle operational PWR depletion benchmark problem using high-fidelity tools and demonstrates acceptable accuracy by comparison with measured operational plant data (open source) provided to the reactor analysis community. (authors)

  17. Whole-core neutron transport calculations without fuel-coolant homogenization

    SciTech Connect

    Smith, M. A.; Tsoulfanidis, N.; Lewis, E. E.; Palmiotti, G.; Taiwo, T. A.

    2000-02-10

    The variational nodal method implemented in the VARIANT code is generalized to perform full core transport calculations without spatial homogenization of cross sections at either the fuel-pin cell or fuel assembly level. The node size is chosen to correspond to one fuel-pin cell in the radial plane. Each node is divided into triangular finite subelements, with the interior spatial flux distribution represented by piecewise linear trial functions. The step change in the cross sections at the fuel-coolant interface can thus be represented explicitly in global calculations while retaining the fill spherical harmonics capability of VARIANT. The resulting method is applied to a two-dimensional seven-group representation of a LWR containing MOX fuel assemblies. Comparisons are made of the accuracy of various space-angle approximations and of the corresponding CPU times.

  18. Design of the radiation shielding for the time of flight enhanced diagnostics neutron spectrometer at Experimental Advanced Superconducting Tokamak

    SciTech Connect

    Du, T. F.; Chen, Z. J.; Peng, X. Y.; Yuan, X.; Zhang, X.; Hu, Z. M.; Cui, Z. Q.; Xie, X. F.; Ge, L. J.; Li, X. Q.; Zhang, G. H.; Chen, J. X.; Fan, T. S., E-mail: tsfan@pku.edu.cn [School of Physics, State Key Lab of Nuclear Physics and Technology, Peking University, Beijing (China); Gorini, G.; Nocente, M. [Dipartimento di Fisica, Università di Milano-Bicocca, Milano (Italy); Istituto di Fisicadel Plasma “P. Caldirola,” Milano (Italy); Tardocchi, M. [Istituto di Fisicadel Plasma “P. Caldirola,” Milano (Italy); Hu, L. Q.; Zhong, G. Q.; Lin, S. Y.; Wan, B. N. [Institute of Plasma Physics, CAS, Hefei (China)

    2014-11-15

    A radiation shielding has been designed to reduce scattered neutrons and background gamma-rays for the new double-ring Time Of Flight Enhanced Diagnostics (TOFED). The shielding was designed based on simulation with the Monte Carlo code MCNP5. Dedicated model of the EAST tokamak has been developed together with the emission neutron source profile and spectrum; the latter were simulated with the Nubeam and GENESIS codes. Significant reduction of background radiation at the detector can be achieved and this satisfies the requirement of TOFED. The intensities of the scattered and direct neutrons in the line of sight of the TOFED neutron spectrometer at EAST are studied for future data interpretation.

  19. Physics Analyses in the Design of the HFIR Cold Neutron Source

    SciTech Connect

    Bucholz, J.A.

    1999-09-27

    Physics analyses have been performed to characterize the performance of the cold neutron source to be installed in the High Flux Isotope Reactor at the Oak Ridge National Laboratory in the near future. This paper provides a description of the physics models developed, and the resulting analyses that have been performed to support the design of the cold source. These analyses have provided important parametric performance information, such as cold neutron brightness down the beam tube and the various component heat loads, that have been used to develop the reference cold source concept.

  20. Insert Design and Manufacturing for Foam-Core Composite Sandwich Structures

    NASA Astrophysics Data System (ADS)

    Lares, Alan

    Sandwich structures have been used in the aerospace industry for many years. The high strength to weight ratios that are possible with sandwich constructions makes them desirable for airframe applications. While sandwich structures are effective at handling distributed loads such as aerodynamic forces, they are prone to damage from concentrated loads at joints or due to impact. This is due to the relatively thin face-sheets and soft core materials typically found in sandwich structures. Carleton University's Uninhabited Aerial Vehicle (UAV) Project Team has designed and manufactured a UAV (GeoSury II Prototype) which features an all composite sandwich structure fuselage structure. The purpose of the aircraft is to conduct geomagnetic surveys. The GeoSury II Prototype serves as the test bed for many areas of research in advancing UAV technologies. Those areas of research include: low cost composite materials manufacturing, geomagnetic data acquisition, obstacle detection, autonomous operations and magnetic signature control. In this thesis work a methodology for designing and manufacturing inserts for foam-core sandwich structures was developed. The results of this research work enables a designer wishing to design a foam-core sandwich airframe structure, a means of quickly manufacturing optimized inserts for the safe introduction of discrete loads into the airframe. The previous GeoSury II Prototype insert designs (v.1 & v.2) were performance tested to establish a benchmark with which to compare future insert designs. Several designs and materials were considered for the new v.3 inserts. A plug and sleeve design was selected, due to its ability to effectively transfer the required loads to the sandwich structure. The insert material was chosen to be epoxy, reinforced with chopped carbon fibre. This material was chosen for its combination of strength, low mass and also compatibility with the face-sheet material. The v.3 insert assembly is 60% lighter than the previous insert designs. A casting process for manufacturing the v.3 inserts was developed. The developed casting process, when producing more than 13 inserts, becomes more economical than machining. An exploratory study was conducted looking at the effects of dynamic loading on the v.3 insert performance. The results of this study highlighted areas for improving dynamic testing of foam-core sandwich structure inserts. Correlations were developed relating design variables such as face-sheet thickness and insert diameter to a failure load for different load cases. This was done through simulations using Computer Aided Engineering (CAE) software, and experimental testing. The resulting correlations were integrated into a computer program which outputs the required insert dimensions given a set of design parameters, and load values.

  1. Performance of truss panels with kagome cores and design of a high authority shape morphing structure

    NASA Astrophysics Data System (ADS)

    Wang, Ju

    This dissertation includes two parts: First, the performance of a light weight truss panels with Kagome cores; Second, design of a high authority morphing structure for hinging and twisting. The performance characteristics of a truss core sandwich panel design based on the 3D Kagome are measured and compared with earlier numerical simulations and the consistency is demonstrated. Panels are fabricated by investment casting and tested in compression, shear and 3-point bending. The isotropic nature of this core design is confirmed. The superior performance relative to truss designs based on the tetrahedron is demonstrated and attributed to the greater resistance to plastic buckling at the equivalent core density. The failed samples are examined in the scanning electron microscope and imperfections are identified to have caused the premature failures. A concept for a high authority shape morphing plate is described. The design incorporates an active Kagome back-plane capable of changing the shape of a solid face by transmitting loads through a tetrahedral core. The two shape deformations to be achieved and demonstrated consist of hinging and twisting. The design is performed by a combination of analytic estimation and numerical simulation, guided by previous assessments of the Kagome configuration. The objective is to ascertain designs that provide the maximum edge displacement subjected to specified passive load. An optimization is used to ascertain the largest displacement achievable within the force capacity of the actuators. These displacements have been demonstrated and shown to correspond with values predicted by numerical simulation. Non-linear effects, such as face wrinkling, are probed by using a finite element method and the fidelity of the results assessed through comparison with measurements. The numerical results are used to validate a dimensional analysis of trends in the actuation resistance of the structure with geometry, as well as the passive load capacity. The forces determined by such analysis are combined with the failure mechanisms for all sub-systems to establish the constraints. The important domains are visualized using mechanism maps. An optimization is used to generate load capacity maps that guide geometric design and provide actuator capacity requirements.

  2. A new design of fission detector for prompt fission neutron investigation

    NASA Astrophysics Data System (ADS)

    Zeynalov, Sh.; Zeynalova, O.; Nazarenko, M. A.; Hambsch, F.-J.; Oberstedt, S.

    2012-10-01

    In this work we report recent achievements in design of twin back-to-back ionization chamber (TIC) for fission fragment (FF) mass and kinetic energy spectroscopy. Correlated FF kinetic energies, their masses and the angle of the fission axes in 3D Cartesian coordinates can be determined from analysis of the heights and shapes of the pulses induced by the fission fragments on the anodes of TIC. Anodes of TIC were designed as consisting of isolated strips each having independent electronic circuitry and special multi-channel pulse processing apparatus. Mathematical algorithms were provided along with formulae derived for fission axis angles determination. It was shown how the point of fission fragments origin on the target plane may be determined using the same measured data. The last feature made the TIC a rather powerful tool for prompt fission neutron (PFN) emission investigation in event by event analysis of individual fission reactions from non point fissile source. Position sensitive neutron induced fission detector for neutron imaging applications with both thermal and low energy neutrons was found as another possible implementation of the designed TIC.

  3. Conceptual design of thorium-fuelled Mitrailleuse accelerator-driven subcritical reactor using D-Be neutron source

    SciTech Connect

    Kokubo, Y. [Quan Japan Company Limited, 3-9-15 Sannomiya-cho, Chuo-ku, Kobe, Hyogo, 650-0021 (Japan); Kamei, T. [Research Inst. for Applied Sciences, 49 Tanaka Ohicho, Sakyo-ku, Kyoto-shi, Kyoto, 606-8202 (Japan)

    2012-07-01

    A distributed accelerator is a charged-particle accelerator that uses a new acceleration method based on repeated electrostatic acceleration. This method offers outstanding benefits not possible with the conventional radio-frequency acceleration method, including: (1) high acceleration efficiency, (2) large acceleration current, and (3) lower failure rate made possible by a fully solid-state acceleration field generation circuit. A 'Mitrailleuse Accelerator' is a product we have conceived to optimize this distributed accelerator technology for use with a high-strength neutron source. We have completed the conceptual design of a Mitrailleuse Accelerator and of a thorium-fuelled subcritical reactor driven by a Mitrailleuse Accelerator. This paper presents the conceptual design details and approach to implementing the subcritical reactor core. We will spend the next year or so on detailed design work, and then will start work on developing a prototype for demonstration. If there are no obstacles in setting up a development organization, we expect to finish verifying the prototype's performance by the third quarter of 2015. (authors)

  4. The design of a multisource americium-beryllium (Am-Be) neutron irradiation facility using MCNP for the neutronic performance calculation.

    PubMed

    Sogbadji, R B M; Abrefah, R G; Nyarko, B J B; Akaho, E H K; Odoi, H C; Attakorah-Birinkorang, S

    2014-08-01

    The americium-beryllium neutron irradiation facility at the National Nuclear Research Institute (NNRI), Ghana, was re-designed with four 20 Ci sources using Monte Carlo N-Particle (MCNP) code to investigate the maximum amount of flux that is produced by the combined sources. The results were compared with a single source Am-Be irradiation facility. The main objective was to enable us to harness the maximum amount of flux for the optimization of neutron activation analysis and to enable smaller sample sized samples to be irradiated. Using MCNP for the design construction and neutronic performance calculation, it was realized that the single-source Am-Be design produced a thermal neutron flux of (1.8±0.0007)×10(6) n/cm(2)s and the four-source Am-Be design produced a thermal neutron flux of (5.4±0.0007)×10(6) n/cm(2)s which is a factor of 3.5 fold increase compared to the single-source Am-Be design. The criticality effective, k(eff), of the single-source and the four-source Am-Be designs were found to be 0.00115±0.0008 and 0.00143±0.0008, respectively. PMID:24792123

  5. Core-crust transition in neutron stars: Predictivity of density developments

    NASA Astrophysics Data System (ADS)

    Ducoin, Camille; Margueron, Jérôme; Providência, Constança; Vidaña, Isaac

    2011-04-01

    The possibility to draw links between the isospin properties of nuclei and the structure of compact stars is a stimulating perspective. In order to pursue this objective on a sound basis, the correlations from which such links can be deduced have to be carefully checked against model dependence. Using a variety of nuclear effective models and a microscopic approach, we study the relation between the predictions of a given model and those of a Taylor density development of the corresponding equation of state: this establishes to what extent a limited set of phenomenological constraints can determine the core-crust transition properties. From a correlation analysis, we show that (a) the transition density ?t is mainly correlated with the symmetry energy slope L, (b) the proton fraction Yp,t with the symmetry energy and symmetry energy slope (J,L) defined at saturation density, or, even better, with the same quantities defined at ?=0.1 fm-3, and (c) the transition pressure Pt with the symmetry energy slope and curvature (L,Ksym) defined at ?=0.1 fm-3.

  6. Core-crust transition in neutron stars: Predictivity of density developments

    SciTech Connect

    Ducoin, Camille; Providencia, Constanca; Vidana, Isaac [CFC, Department of Physics, University of Coimbra, PT-3004-516, Coimbra (Portugal); Margueron, Jerome [Institut de Physique Nucleaire, Universite Paris-Sud, IN2P3-CNRS, FR-91406 Orsay Cedex (France)

    2011-04-15

    The possibility to draw links between the isospin properties of nuclei and the structure of compact stars is a stimulating perspective. In order to pursue this objective on a sound basis, the correlations from which such links can be deduced have to be carefully checked against model dependence. Using a variety of nuclear effective models and a microscopic approach, we study the relation between the predictions of a given model and those of a Taylor density development of the corresponding equation of state: this establishes to what extent a limited set of phenomenological constraints can determine the core-crust transition properties. From a correlation analysis, we show that (a) the transition density {rho}{sub t} is mainly correlated with the symmetry energy slope L, (b) the proton fraction Y{sub p,t} with the symmetry energy and symmetry energy slope (J,L) defined at saturation density, or, even better, with the same quantities defined at {rho}=0.1 fm{sup -3}, and (c) the transition pressure P{sub t} with the symmetry energy slope and curvature (L,K{sub sym}) defined at {rho}=0.1 fm{sup -3}.

  7. Core-crust transition in neutron stars: predictivity of density developments

    E-print Network

    Camille Ducoin; Jérôme Margueron; Constança Providência; Isaac Vidaña

    2011-05-20

    The possibility to draw links between the isospin properties of nuclei and the structure of compact stars is a stimulating perspective. In order to pursue this objective on a sound basis, the correlations from which such links can be deduced have to be carefully checked against model dependence. Using a variety of nuclear effective models and a microscopic approach, we study the relation between the predictions of a given model and those of a Taylor density development of the corresponding equation of state: this establishes to what extent a limited set of phenomenological constraints can determine the core-crust transition properties. From a correlation analysis we show that a) the transition density $\\rho_t$ is mainly correlated with the symmetry energy slope $L$, b) the proton fraction $Y_{p,t}$ with the symmetry energy and symmetry energy slope $(J,L)$ defined at saturation density, or, even better, with the same quantities defined at $\\rho=0.1$ fm$^{-3}$, and c) the transition pressure $P_t$ with the symmetry energy slope and curvature $(J,K_{\\rm sym})$ defined at $\\rho=0.1$ fm$^{-3}$.

  8. The design and installation of a core discharge monitor for CANDU-type reactors

    SciTech Connect

    Halbig, J.K. (Los Alamos National Lab., NM (USA)); Monticone, A.C.; Ksiezak, L. (International Atomic Energy Agency, Vienna (Austria)); Smiltnieks, V. (International Atomic Energy Agency, Toronto, ON (Canada). Regional Office)

    1990-01-01

    A new type of surveillance systems that monitors neutron and gamma radiation in a reactor containment is being installed at the Ontario Hydro Darlington Nuclear Generating Station A, Unit 2. Unlike video or film surveillance that monitors mechanical motion, this system measures fuel-specific radiation emanating from irradiated fuel as it is pushed from the core of CANDU-type reactors. Proof-of-principle measurements have been carried out at Bruce Nuclear Generating Station A, Unit 3. The system uses ({gamma},n) threshold detectors and ionization detectors. A microprocessor-based electronics package, GRAND-II (Gamma Ray and Neutron Detector electronics package), provides detector bias, preamplifier power, and signal processing. Firmware in the GRAND-2 controls the surveillance activities, including data acquisition and a level of detector authentication, and it handles authenticated communication with a central data logging computer. Data from the GRAND-II are transferred to an MS-DOS-compatible computer and stored. These data are collected and reviewed for fuel-specific radiation signatures from the primary detector and proper ratios of signals from secondary detectors. 5 figs.

  9. Nuclear safety analyses and core design calculations to convert the Texas A & M University Nuclear Science Center reactor to low enrichment uranium fuel. Final report

    SciTech Connect

    Parish, T.A.

    1995-03-02

    This project involved performing the nuclear design and safety analyses needed to modify the license issued by the Nuclear Regulatory Commission to allow operation of the Texas A& M University Nuclear Science Center Reactor (NSCR) with a core containing low enrichment uranium (LEU) fuel. The specific type of LEU fuel to be considered was the TRIGA 20-20 fuel produced by General Atomic. Computer codes for the neutronic analyses were provided by Argonne National Laboratory (ANL) and the assistance of William Woodruff of ANL in helping the NSCR staff to learn the proper use of the codes is gratefully acknowledged. The codes applied in the LEU analyses were WIMSd4/m, DIF3D, NCTRIGA and PARET. These codes allowed full three dimensional, temperature and burnup dependent calculations modelling the NSCR core to be performed for the first time. In addition, temperature coefficients of reactivity and pulsing calculations were carried out in-house, whereas in the past this modelling had been performed at General Atomic. In order to benchmark the newly acquired codes, modelling of the current NSCR core with highly enriched uranium fuel was also carried out. Calculated results were compared to both earlier licensing calculations and experimental data and the new methods were found to achieve excellent agreement with both. Therefore, even if an LEU core is never loaded at the NSCR, this project has resulted in a significant improvement in the nuclear safety analysis capabilities established and maintained at the NSCR.

  10. Design of broadband dispersion compensating fiber with the small core photonic crystal fiber

    NASA Astrophysics Data System (ADS)

    Wu, Ming; Liu, Hairong; Huang, Dexiu; Tong, Weijun; Wei, Huifeng

    2007-11-01

    This paper has presented a design of dispersion compensating fiber with small core photonic crystal fiber (PCF) based on pure silica. The designed small core PCF can be controlled the dispersion properties in terms of the structural parameters, the pitch ? and the air-filling fraction d/?. The negative chromatic dispersion coefficient can be achieved - 586.5ps/(nmÂ.km) at ?=1550nm with ?=0.9?m and d/?=0.9. This kind of PCF can be used for broadband dispersion compensation in S+C+L band (1460-1565nm) and the chromatic dispersion coefficient is lower than -450ps/(nmÂ.km) in S+C+L band. It can realize the dispersion slope compensation because that it exhibits a negative dispersion slope.

  11. Evaluation of D(d,n) 3 He reaction neutron source models for BNCT irradiation system design

    Microsoft Academic Search

    YAO Ze'en; LUO Peng; Tooru KOBAYASHI; Gerard BENGUA

    2007-01-01

    A mathematical method was developed to calculate the yield, energy spectrum and angular distribution of neutrons from D(d,n)3 He (D-D) reaction in a thick deuterium-titanium target for incident deuterons in energies lower than 1.0MeV. The data of energy spectrum and angular distribution were applied to set up the neutron source model for the beam-shaping-assembly (BSA) design of Boron-Neutron-Capture-Therapy (BNCT) using

  12. Design and simulation of a neutron source based on an electron linear accelerator for BNCT of skin melanoma

    Microsoft Academic Search

    Ali Pazirandeh; Ali Torkamani; Ali Taheri

    2011-01-01

    The PNS project is devoted to the design and simulation of a Photo-Neutron Source for BNCT using a Varian 2300C\\/D electron accelerator. This paper describes the production of the high-energy gamma-rays followed by neutron production in (gamma,n) reaction. To optimize the whole setup and maximize the neutron flux, the FLUKA code is employed to simulate the system. The results show

  13. Nonlinear CA Based Scalable Design of On-Chip TPG for Multiple Cores

    Microsoft Academic Search

    Sukanta Das; Biplab K. Sikdar; Parimal Pal Chaudhuri

    2004-01-01

    This paper reports an efficient design of test pattern generators (TPGs) for a chip having multiple cores. It is built around nonlinear cellular automata (CA) based pseudo-random pattern generator (PRPG). The modular and cascadable structure of proposed n-cell PRPG can be utilized to construct the (n+1)-cell PRPG without sacrificing the pseudo-randomness quality. The efficiency of such a scalable PRPG structure

  14. Test Planning and Design Space Exploration in a Core-Based Environment

    Microsoft Academic Search

    Erika Cota; Luigi Carro; Marcelo Lubaszewski; Alex Orailoglu

    2002-01-01

    This paper proposes a comprehensive model for testplanning in a core-based environment. The main contributionof this work is the use of several types of TAMs and theconsideration of different optimization factors (area, pinsand test time) during the global TAM and test schedule definition.This expansion of concerns makes possible an efficientyet fine-grained search in the huge design space ofa reuse-based environment.

  15. Design optimization of radially magnetized, iron-cored, tubular permanent-magnet machines and drive systems

    Microsoft Academic Search

    Jiabin Wang; David Howe

    2004-01-01

    In this paper we deduce, from analytical field solutions, the influence of leading design parameters on the performance of a radially magnetized, iron-cored, tubular permanent-magnet machine and its drive system. We derive analytical formulas for predicting the open-circuit electromotive force, the thrust force, the iron loss, and the winding resistance and inductances, as well as the converter losses. The force

  16. Design, synthesis and photochemical properties of the first examples of iminosugar clusters based on fluorescent cores

    PubMed Central

    Lepage, Mathieu L; Mirloup, Antoine; Ripoll, Manon; Stauffert, Fabien; Bodlenner, Anne

    2015-01-01

    Summary The synthesis and photophysical properties of the first examples of iminosugar clusters based on a BODIPY or a pyrene core are reported. The tri- and tetravalent systems designed as molecular probes and synthesized by way of Cu(I)-catalysed azide–alkyne cycloadditions are fluorescent analogues of potent pharmacological chaperones/correctors recently reported in the field of Gaucher disease and cystic fibrosis, two rare genetic diseases caused by protein misfolding.

  17. Design and measurement of all-rod terahertz photonic crystal fiber with air-core

    NASA Astrophysics Data System (ADS)

    Zhang, Le; Li, Jiu-sheng

    2015-06-01

    An all-rod terahertz wave air-core photonic crystal fiber is designed and fabricated. Transmission spectra through the sample are measured by using terahertz time-domain spectroscopy system. Periodic transmission band and low loss of the fiber are experimentally proved. Measurement results show that the power loss of the present terahertz wave photonic crystal fiber is less than 1.9 dB/cm in transmission bands. The wave confinement ability of the fiber is also demonstrated.

  18. Parametric design and manufacture of transfer port cores for experimental cylinder block castings

    Microsoft Academic Search

    J. Mccartney; R. Kenny

    1989-01-01

    The parametric approach to creating designs from a generalised model is discussed. The technique is extended to the manufacture of transfer port cores for two-stroke internal combustion engines. Bi-cubic patches representing each of the three-dimensional sides of the transfer port are constructed from a parametric model using the minimum amount of information. A postprocessor then creates a CNC program from

  19. Low-Power Design of 90-nm SuperH Processor Core

    Microsoft Academic Search

    Tetsuya Yamada; Masahide Abe; Yusuke Nitta; Kenji Ogura; Manabu Kusaoke; Makoto Ishikawa; Motokazu Ozawa; Kiwamu Takada; Fumio Arakawa; Osamu Nishii; Toshihiro Hattori

    2005-01-01

    A low-power SuperH™ embedded processor core, the SH-X2, has been designed in 90-nm CMOS technology. The power consumption was reduced by using hierarchical fine-grained clock gating to reduce the power consumption of the flip-flops and clock-tree, synthesis and a layout that support implementation of the clock gating, and several-level power evaluations for RTL refinement. With this clock gating and RTL

  20. Energy Efficient Engine integrated core/low spool design and performance report

    NASA Technical Reports Server (NTRS)

    Stearns, E. Marshall

    1985-01-01

    The Energy Efficient Engine (E3) is a NASA program to create fuel saving technology for future transport aircraft engines. The E3 technology advancements were demonstrated to operate reliably and achieve goal performance in tests of the Integrated Core/Low Spool vehicle. The first build of this undeveloped technology research engine set a record for low fuel consumption. Its design and detailed test results are herein presented.

  1. Resource allocation and test scheduling for concurrent test of core-based SOC design

    Microsoft Academic Search

    Yu Huang; Wu-Tung Cheng; Chien-Chung Tsai; Nilanjan Mukherjee; Omer Samman; Yahya Zaidan; Sudhakar M. Reddy

    2001-01-01

    A method to solve the resource allocation and test scheduling problems together in order to achieve concurrent test for core-based system-on-chip (SOC) designs is presented in this paper. The primary objective for concurrent SOC test is to reduce test application time. The methodology used in this paper is not limited to any specific test access mechanism (TAM). Additionally, it can

  2. A Metropolis algorithm combined with Nelder–Mead Simplex applied to nuclear reactor core design

    Microsoft Academic Search

    Wagner F. Sacco; Hermes Alves Filho; Nélio Henderson; Cassiano R. E. de Oliveira

    2008-01-01

    A hybridization of the recently introduced Particle Collision Algorithm (PCA) and the Nelder–Mead Simplex algorithm is introduced and applied to a core design optimization problem which was previously attacked by other metaheuristics. The optimization problem consists in adjusting several reactor cell parameters, such as dimensions, enrichment and materials, in order to minimize the average peak-factor in a three-enrichment-zone reactor, considering

  3. Preliminary design report for SCDAP/RELAP5 lower core plate model

    SciTech Connect

    Coryell, E.W. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab.; Griffin, F.P. [Oak Ridge National Lab., TN (United States)

    1998-07-01

    The SCDAP/RELAP5 computer code is a best-estimate analysis tool for performing nuclear reactor severe accident simulations. Under primary sponsorship of the US Nuclear Regulatory Commission (NRC), Idaho National Engineering and Environmental Laboratory (INEEL) is responsible for overall maintenance of this code and for improvements for pressurized water reactor (PWR) applications. Since 1991, Oak Ridge National Laboratory (ORNL) has been improving SCDAP/RELAP5 for boiling water reactor (BWR) applications. The RELAP5 portion of the code performs the thermal-hydraulic calculations for both normal and severe accident conditions. The structures within the reactor vessel and coolant system can be represented with either RELAP5 heat structures or SCDAP/RELAP5 severe accident structures. The RELAP5 heat structures are limited to normal operating conditions (i.e., no structural oxidation, melting, or relocation), while the SCDAP portion of the code is capable of representing structural degradation and core damage progression that can occur under severe accident conditions. DCDAP/RELAP5 currently assumes that molten material which leaves the core region falls into the lower vessel head without interaction with structural materials. The objective of this design report is to describe the modifications required for SCDAP/RELAP5 to treat the thermal response of the structures in the core plate region as molten material relocates downward from the core, through the core plate region, and into the lower plenum. This has been a joint task between INEEL and ORNL, with INEEL focusing on PWR-specific design, and ORNL focusing upon the BWR-specific aspects. Chapter 2 describes the structures in the core plate region that must be represented by the proposed model. Chapter 3 presents the available information about the damage progression that is anticipated to occur in the core plate region during a severe accident, including typical SCDAP/RELAP5 simulation results. Chapter 4 provides a description of the implementation of the recommended model and Chapter 5 discusses the testing which could be done to verify the design and implementation of the model.

  4. A Los Alamos design study for a high-power spallation-neutron-source driver

    SciTech Connect

    Jason, A.J.; Hardekopf, R.A.; Macek, R.W.; Schriber, S.O.; Thiessen, H.A.; Woods, R.

    1993-06-01

    A design study for an accelerator-driven spallation-neutron source is underway at Los Alamos. The driver, based on the LAMPF facility, produces a 1-MW proton beam and is upgradable to 5 MW. After linear acceleration to full energy, an H-beam is accumulated for approximately I ms in a compressor ring and then extracted to produce an intense proton burst, less than 1 {mu}s long, onto a spallation target system with a 60-Hz cycle rate. The design uses existing infrastructure insofar as possible while maintaining project goals. This paper summarizes the system specifications and design status.

  5. A Los Alamos design study for a high-power spallation-neutron-source driver

    SciTech Connect

    Jason, A.J.; Hardekopf, R.A.; Macek, R.W.; Schriber, S.O.; Thiessen, H.A.; Woods, R.

    1993-01-01

    A design study for an accelerator-driven spallation-neutron source is underway at Los Alamos. The driver, based on the LAMPF facility, produces a 1-MW proton beam and is upgradable to 5 MW. After linear acceleration to full energy, an H-beam is accumulated for approximately I ms in a compressor ring and then extracted to produce an intense proton burst, less than 1 [mu]s long, onto a spallation target system with a 60-Hz cycle rate. The design uses existing infrastructure insofar as possible while maintaining project goals. This paper summarizes the system specifications and design status.

  6. Design and operation of a passive neutron monitor for assaying the TRU content of solid wastes

    SciTech Connect

    Brodzinski, R.L.; Brown, D.P.; Rieck, H.G. Jr.; Rogers, L.A.

    1984-02-01

    A passive neutron monitor has been designed and built for determining the residual transuranic (TRU) and plutonium content of chopped leached fuel hulls and other solid wastes from spent Fast Flux Test Facility (FFTF) fuel. The system was designed to measure as little as 8 g of plutonium or 88 mg of TRU in a waste package as large as a 208-l drum which could be emitting up to 220,000 R/hr of gamma radiation. For practical purposes, maximum assay times were chosen to be 10,000 sec. The monitor consists of 96 /sup 10/BF/sub 3/ neutron sensitive proportional counting tubes each 5.08 cm in diameter and 183 cm in active length. Tables of neutron emission rates from both spontaneous fission and (..cap alpha..,n) reactions on oxygen are given for all contributing isotopes expected to be present in spent FFTF fuel. Tables of neutron yeilds from isotopic compositions predicted for various exposures and cooling times are also given. Methods of data reduction and sources, magnitude, and control of errors are discussed. Backgrounds and efficiencies have been measured and are reported. A section describing step-by-step operational procedures is included. Guidelines and procedures for quality control and troubleshooting are also given. 13 references, 15 figures, 4 tables.

  7. Design and Construction of the Ultracold Neutron Source at the NC State PULSTAR Research Reactor

    NASA Astrophysics Data System (ADS)

    Palmquist, Grant R.

    An ultracold neutron (UCN) source using solid deuterium is being constructed at the 1MWPULSTAR nuclear reactor on the campus of North Carolina State University. The final stages of assembly and commissioning are underway. The overall design, status of construction, and benchmarking measurements are presented. The UCN source design is based on detailed simulations including MCNP, UCN transport Monte Carlo, and computational fluid dynamics (CFD) simulation of the cryogenic systems. The source will be useful for developing UCN technologies, including guides and detectors, and in support of current projects including measurements of neutron beta-decay asymmetry coefficients and the electric dipolemoment of the neutron. The facility will also be available for testing new techniques using UCN in material and surface physics, as well as new fundamental physics measurements such as neutron lifetime and beta decay measurements. The expected experimental density of UCN/cm3 in a storage volume will be competitive with currently available sources, including those at significantly more powerful reactors.

  8. Design of a neutron penumbral-aperture microscope with 10-. mu. m resolution

    SciTech Connect

    Ress, D.; Lerche, R.A.; Ellis, R.J.; Lane, S.M.

    1990-05-01

    We are currently designing a 10-{mu}m resolution neutron penumbral-aperture microscope to diagnose high-convergence targets at the Nova laser facility. To achieve such high resolution, the new microscope will require substantial improvements in three areas. First, we have designed thick penumbral apertures with extremely sharp cutoffs over a useful ({approx}100 {mu}m) field of view; fabrication of such apertures appears feasible using gold electroplating techniques. Second, the limited field of view and required close proximity of the aperture to the target (2 cm) necessitates a durable mounting and alignment system with {plus}25 {mu}m accuracy. Finally, a neutron detector containing 160,000 scintillator elements is required; readout and optimization of this large array are outstanding issues. 5 refs., 3 figs.

  9. Multi-group helium and hydrogen production cross section libraries for fusion neutronics design

    NASA Astrophysics Data System (ADS)

    Mori, Seiji; Zimin, S.; Takatsu, Hideyuki

    1993-09-01

    The helium and hydrogen production cross section libraries based on the JENDL-3 data file were compiled for use in neutronics and shielding design calculation of a fusion reactor. These libraries have the same group structures as the transport cross section sets, FUSION-J3 and FUSION-40, which are often used in fusion neutronics design and can be used as the response function libraries for the reaction rate calculation code, APPLE-3. These libraries were processed from the JENDL gas production cross section file which is one of the JENDL special purpose files. Some sample calculations using the discrete ordinate code, ANISN, with these libraries were performed and the results were compared with the existing results. Consequently it was found that the appropriate results can be obtained with these libraries. The generated multi-group cross sections for helium and hydrogen production are presented in graphs and tables.

  10. Validation of the MCNPX-PoliMi Code to Design a Fast-Neutron Multiplicity Counter

    SciTech Connect

    J. L. Dolan; A. C. Kaplan; M. Flaska; S. A. Pozzi; D. L. Chichester

    2012-07-01

    Many safeguards measurement systems used at nuclear facilities, both domestically and internationally, rely on He-3 detectors and well established mathematical equations to interpret coincidence and multiplicity-type measurements for verifying quantities of special nuclear material. Due to resource shortages alternatives to these existing He-3 based systems are being sought. Work is also underway to broaden the capabilities of these types of measurement systems in order to improve current multiplicity analysis techniques. As a part of a Material Protection, Accounting, and Control Technology (MPACT) project within the U.S. Department of Energy's Fuel Cycle Technology Program we are designing a fast-neutron multiplicity counter with organic liquid scintillators to quantify important quantities such as plutonium mass. We are also examining the potential benefits of using fast-neutron detectors for multiplicity analysis of advanced fuels in comparison with He-3 detectors and testing the performance of such designs. The designs are being developed and optimized using the MCNPX-PoliMi transport code to study detector response. In the full paper, we will discuss validation measurements used to justify the use of the MCNPX-PoliMi code paired with the MPPost multiplicity routine to design a fast neutron multiplicity counter with liquid scintillators. This multiplicity counter will be designed with the end goal of safeguarding advanced nuclear fuels. With improved timing qualities associated with liquid scintillation detectors, we can design a system that is less limited by nuclear materials of high activities. Initial testing of the designed system with nuclear fuels will take place at Idaho National Laboratory in a later stage of this collaboration.

  11. Design, Assembly, and Testing of the Neutron Imaging Lens for the National Ignition Facility

    SciTech Connect

    Malone, Robert M; Fatherley, Valerie E; Frogget, Brent C; Grim, Gary P; Kaufman, Morris I; McGillivray, Kevin D; Oertel, John A; Palagi, Martin J; Skarda, William K; Tibbitts, Aric; Wilde, Carl H

    2010-09-01

    The National Ignition Facility will begin testing DT fuel capsules yielding greater than 10^13 neutrons during 2010. Neutron imaging is an important diagnostic for understanding capsule behavior. Neutrons are imaged at a scintillator after passing through a pinhole. The pixelated, 160-mm square scintillator is made up of ¼ mm diameter rods 50 mm long. Shielding and distance (28 m) are used to preserve the recording diagnostic hardware. Neutron imaging is light starved. We designed a large nine-element collecting lens to relay as much scintillator light as reasonable onto a 75 mm gated microchannel plate (MCP) intensifier. The image from the intensifier’s phosphor passes through a fiber taper onto a CCD camera for digital storage. Alignment of the pinhole and tilting of the scintillator is performed before the relay lens and MCP can be aligned. Careful tilting of the scintillator is done so that each neutron only passes through one rod (no crosstalk allowed). The 3.2 ns decay time scintillator emits light in the deep blue, requiring special glass materials. The glass within the lens housing weighs 26 lbs, with the largest element being 7.7 inches in diameter. The distance between the scintillator and the MCP is only 27 inches. The scintillator emits light with 0.56 NA and the lens collects light at 0.15 NA. Thus, the MCP collects only 7% of the available light. Baffling the stray light is a major concern in the design of the optics. Glass cost considerations, tolerancing, and alignment of this lens system will be discussed.

  12. Dosimetric comparison of four new design 103Pd brachytherapy sources: optimal design using silver and copper rod cores.

    PubMed

    Hosseini, S Hamed; Sadeghi, Mahdi; Ataeinia, Vahideh

    2009-07-01

    Four new brachytherapy sources, IRA1-103Pd, IRA2-103Pd, IRA3-103Pd, and IRA4-103Pd, have been developed at Agricultural, Medical, and Industrial Research School and are designed for permanent implant application. With the goal of determining an optimal design for a 103Pd source, this article compares the dosimetric properties of these sources with reference to the authors' earlier IRA-103Pd source. The four new sources differ in end cap configuration and thickness and in the core material, silver or copper, that carries the adsorbed 103Pd. Dosimetric data derived from the authors' Monte Carlo simulation results are reported in accordance with the updated AAPM Task Group No. 43 report (TG-43U1). For each source, the authors obtained detailed results for the dose rate constant lambda, the radial dose function g(r), the anisotropy function F(r, theta), and the anisotropy factor phi(an)(r). In this study, the optimal source IRA3-103Pd provides the most isotropic dose distribution in water with the dose rate constant of 0.678(+/-0.1%) cGy h(-1) U(-1). The IRA3-103Pd design has a silver rod core combined with thin-wall, concave end caps. Finally, the authors compared the results for their optimal source with published results for those of other source manufacturers. PMID:19673207

  13. Dosimetric comparison of four new design {sup 103}Pd brachytherapy sources: Optimal design using silver and copper rod cores

    SciTech Connect

    Hosseini, S. Hamed; Sadeghi, Mahdi; Ataeinia, Vahideh [Agricultural, Medical and Industrial Research School, Nuclear Science and Technology Research Institute, P.O. Box 31485-498, Karaj, Tehran (098)21 (Iran, Islamic Republic of)

    2009-07-15

    Four new brachytherapy sources, IRA1-{sup 103}Pd, IRA2-{sup 103}Pd, IRA3-{sup 103}Pd, and IRA4-{sup 103}Pd, have been developed at Agricultural, Medical, and Industrial Research School and are designed for permanent implant application. With the goal of determining an optimal design for a {sup 103}Pd source, this article compares the dosimetric properties of these sources with reference to the authors' earlier IRA-{sup 103}Pd source. The four new sources differ in end cap configuration and thickness and in the core material, silver or copper, that carries the adsorbed {sup 103}Pd. Dosimetric data derived from the authors' Monte Carlo simulation results are reported in accordance with the updated AAPM Task Group No. 43 report (TG-43U1). For each source, the authors obtained detailed results for the dose rate constant {Lambda}, the radial dose function g(r), the anisotropy function F(r,{theta}), and the anisotropy factor {phi}{sub an}(r). In this study, the optimal source IRA3-{sup 103}Pd provides the most isotropic dose distribution in water with the dose rate constant of 0.678({+-}0.1%) cGy h{sup -1} U{sup -1}. The IRA3-{sup 103}Pd design has a silver rod core combined with thin-wall, concave end caps. Finally, the authors compared the results for their optimal source with published results for those of other source manufacturers.

  14. Design of a boiling water reactor equilibrium core using thorium-uranium fuel

    SciTech Connect

    Francois, J-L.; Nunez-Carrera, A.; Espinosa-Paredes, G.; Martin-del-Campo, C.

    2004-10-06

    In this paper the design of a Boiling Water Reactor (BWR) equilibrium core using thorium is presented; a heterogeneous blanket-seed core arrangement concept was adopted. The design was developed in three steps: in the first step two different assemblies were designed based on the integrated blanket-seed concept, they are the blanket-dummy assembly and the blanket-seed assembly. The integrated blanketseed concept comes from the fact that the blanket and the seed rods are located in the same assembly, and are burned-out in a once-through cycle. In the second step, a core design was developed to achieve an equilibrium cycle of 365 effective full power days in a standard BWR with a reload of 104 fuel assemblies designed with an average 235U enrichment of 7.5 w/o in the seed sub-lattice. The main operating parameters, like power, linear heat generation rate and void distributions were obtained as well as the shutdown margin. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The shutdown margin design criterion was fulfilled by addition of a burnable poison region in the assembly. In the third step an in-house code was developed to evaluate the thorium equilibrium core under transient conditions. A stability analysis was also performed. Regarding the stability analysis, five operational states were analyzed; four of them define the traditional instability region corner of the power-flow map and the fifth one is the operational state for the full power condition. The frequency and the boiling length were calculated for each operational state. The frequency of the analyzed operational states was similar to that reported for BWRs; these are close to the unstable region that occurs due to the density wave oscillation phenomena in some nuclear power plants. Four transient analyses were also performed: manual SCRAM, recirculation pumps trip, main steam isolation valves closure and loss of feed water. The results of these transients are similar to those obtained with the traditional UO2 nuclear fuel.

  15. Shielding analysis and design of the KIPT experimental neutron source facility of Ukraine.

    SciTech Connect

    Zhong, Z.; Gohar, M. Y. A.; Naberezhnev, D.; Duo, J.; Nuclear Engineering Division

    2008-10-31

    Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an experimental neutron source facility based on the use of an electron accelerator driven subcritical (ADS) facility [1]. The facility uses the existing electron accelerators of KIPT in Ukraine. The neutron source of the sub-critical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The electron beam has a uniform spatial distribution and the electron energy in the range of 100 to 200 MeV, [2]. The main functions of the facility are the production of medical isotopes and the support of the Ukraine nuclear power industry. Reactor physics experiments and material performance characterization will also be carried out. The subcritical assembly is driven by neutrons generated by the electron beam interactions with the target material. A fraction of these neutrons has an energy above 50 MeV generated through the photo nuclear interactions. This neutron fraction is very small and it has an insignificant contribution to the subcritical assembly performance. However, these high energy neutrons are difficult to shield and they can be slowed down only through the inelastic scattering with heavy isotopes. Therefore the shielding design of this facility is more challenging relative to fission reactors. To attenuate these high energy neutrons, heavy metals (tungsten, iron, etc.) should be used. To reduce the construction cost, heavy concrete with 4.8 g/cm{sup 3} density is selected as a shielding material. The iron weight fraction in this concrete is about 0.6. The shape and thickness of the heavy concrete shield are defined to reduce the biological dose equivalent outside the shield to an acceptable level during operation. At the same time, special attention was give to reduce the total shield mass to reduce the construction cost. The shield design is configured to maintain the biological dose equivalent during operation {le} 0.5 mrem/h inside the subcritical hall, which is five times less than the allowable dose for working forty hours per week for 50 weeks per year. This study analyzed and designed the thickness and the shape of the radial and top shields of the neutron source based on the biological dose equivalent requirements inside the subcritical hall during operation. The Monte Carlo code MCNPX is selected because of its capabilities for transporting electrons, photons, and neutrons. Mesh based weight windows variance reduction technique is utilized to estimate the biological dose outside the shield with good statistics. A significant effort dedicated to the accurate prediction of the biological dose equivalent outside the shield boundary as a function of the shield thickness without geometrical approximations or material homogenization. The building wall was designed with ordinary concrete to reduce the biological dose equivalent to the public with a safety factor in the range of 5 to 20.

  16. Design of an RFQ-Based Neutron Source for Cargo ContainerInterrogation

    SciTech Connect

    Staples, John W.; Hoff, M.D.; Kwan, J.W.; Li, D.; Ludewigt, B.A.; Ratti, A.; Virostek, S.P.; Wells, R.P.

    2006-08-01

    An RFQ-based neutron generator system is described that produces pulsed neutrons for the active screening of sea-land cargo containers for the detection of shielded special nuclear materials (SNM).A microwave-driven deuteron source is coupled to an electrostatic LEBT that injects a 40 mA D+ beam into a 6 MeV, 5.1 meter-long 200 MHz RFQ.The RFQ has a unique beam dynamics design and is capable of operating at duty factors of 5 to 10 percent accelerating a D+ time-averaged current of up to 1.5 mA at 5 percent duty factor, including species and transmission loss. The beam is transported through a specially-designed thin window into a 2.5-atmosphere deuterium gas target. A high-frequency dipole magnet is used to scan the beam over the long dimension of the 5by 35 cm target window. The source will deliver a neutron flux of 1 cdot107 n/(cm2s) to the center of an empty cargo container. Details of the ion source, LEBT, RFQ beam dynamics and gas target design are presented.

  17. Design of a highly nonlinear liquid-core photonic crystal fiber used for fiber optical parametric amplifiers

    Microsoft Academic Search

    Chao-yang Lee; Dai Juan; Yong-gang Wang; Bo-jun Yang

    2008-01-01

    In this paper, we design a new nonlinear fiber by filling a highly nonlinear liquid into hollow-core photonic crystal fibers. The liquid-core photonic crystal fiber with carbon disulfide exhibits an extremely high nonlinear parameter gamma which can be more than 20 times larger than that of a conventional PCF, which is desired for FOPA. By using Full Vector Finite Element

  18. Design and performance of horizontal-type neutron reflectometer SOFIA at J-PARC/MLF

    NASA Astrophysics Data System (ADS)

    Yamada, N. L.; Torikai, N.; Mitamura, K.; Sagehashi, H.; Sato, S.; Seto, H.; Sugita, T.; Goko, S.; Furusaka, M.; Oda, T.; Hino, M.; Fujiwara, T.; Takahashi, H.; Takahara, A.

    2011-11-01

    Neutron reflectometry is a powerful method for investigating the surface and interfacial structures of materials in the spatial range from nanometers to sub-micrometers. At the Japan Proton Accelerator Research Complex (J-PARC), a high-intensity pulsed neutron beam is produced with a proton accelerator at 220kW, which will be upgraded to 1MW in future. Beamline 16 (BL16) at the Materials and Life Science Experimental Facility (MLF) in J-PARC is dedicated to a horizontal-type reflectometer, and in this beamline, neutrons are transported downward at two different angles, 2.2° and 5.7° , relative to the horizontal. In December 2008, we started to accept the neutron beam at BL16 with the old ARISA reflectometer relocated from the KENS facility, KEK, Japan; and we have now replaced it with the brand-new reflectometer SOFIA (SOFt Interface Analyzer). With a high-flux beam and instrumental upgrade, the observable reflectivity of SOFIA reaches around 10-7 within a few hours for specimens on 3" substrates. In this paper, we will present the design and performance of the SOFIA reflectometer, and discuss some preliminary results on the device development for further upgrade.

  19. The ARIES-RS power core -- Recent development in Li/V designs

    SciTech Connect

    Sze, D.K.; Billone, M.C.; Hua, T.Q. [and others

    1997-04-01

    The ARIES-RS fusion power plant design study is based on reversed-shear (RS) physics with a Li/V (lithium breeder and vanadium structure) blanket. The reversed-shear discharge has been documented in many large tokamak experiments. The plasma in the RS mode has a high beta, low current, and low current drive requirements. Therefore, it is an attractive physics regime for a fusion power plant. The blanket system based on a Li/V has high temperature operating capability, good tritium breeding, excellent high heat flux removal capability, long structural life time, low activation, low after heat and good safety characteristics. For these reasons, the ARIES-RS reactor study selected Li/V as the reference blanket. The combination of attractive physics and attractive blanket engineering is expected to result in a superior power plant design. This paper summarizes the power core design of the ARIES-RS power plant study.

  20. Preliminary core design studies for the advanced burner reactor over a wide range of conversion ratios.

    SciTech Connect

    Hoffman, E. A.; Yang, W. S.; Hill, R. N.; Nuclear Engineering Division

    2008-05-05

    A consistent set of designs for 1000 MWt commercial-scale sodium-cooled Advance Burner Reactors (ABR) have been developed for both metal and oxide-fueled cores with conversion ratios from breakeven (CR=1.0) to fertile-free (CR=0.0). These designs are expected to satisfy thermal and irradiation damage limits based on the currently available data. The very low conversion ratio designs require fuel that is beyond the current fuel database, which is anticipated to be qualified by and for the Advanced Burned Test Reactor. Safety and kinetic parameters were calculated, but a safety analysis was not performed. Development of these designs was required to achieve the primary goal of this study, which was to generate representative fuel cycle mass flows for system studies of ABRs as part of the Global Nuclear Energy Partnership (GNEP). There are slight variations with conversion ratio but the basic ABR configuration consists of 144 fuel assemblies and between 9 and 22 primary control assemblies for both the metal and oxide-fueled cores. Preliminary design studies indicated that it is feasible to design the ABR to accommodate a wide range of conversion ratio by employing different assembly designs and including sufficient control assemblies to accommodate the large reactivity swing at low conversion ratios. The assemblies are designed to fit within the same geometry, but the size and number of fuel pins within each assembly are significantly different in order to achieve the target conversion ratio while still satisfying thermal limits. Current irradiation experience would allow for a conversion ratio of somewhat below 0.75. The fuel qualification for the first ABR should expand this experience to allow for much lower conversion ratios and higher bunrups. The current designs were based on assumptions about the performance of high and very high enrichment fuel, which results in significant uncertainty about the details of the designs. However, the basic fuel cycle performance trends such as conversion ratio and mass flow parameters are less sensitive to these parameters and the current results should provide a good basis for static and dynamic system analysis. The conversion ratio is fundamentally a ratio of the macroscopic cross section of U-238 capture to that of TRU fission. Since the microscopic cross sections only change moderately with fuel design and isotopic concentration for the fast reactor, a specific conversion ratio requires a specific enrichment. The approximate average charge enrichment (TRU/HM) is 14%, 21%, 33%, 56%, and 100% for conversion ratios of 1.0, 0.75, 0.50, 0.25, and 0.0 for the metal-fueled cores. The approximate average charge enrichment is 17%, 25%, 38%, 60%, and 100% for conversion ratios of 1.0, 0.75, 0.50, 0.25, and 0.0 for the oxide-fueled core. For the split batch cores, the maximum enrichment will be somewhat higher. For both the metal and oxide-fueled cores, the reactivity feedback coefficients and kinetics parameters seem reasonable. The maximum single control assembly reactivity faults may be too large for the low conversion ratio designs. The average reactivity of the primary control assemblies was increased, which may cause the maximum reactivity of the central control assembly to be excessive. The values of the reactivity coefficients and kinetics parameters show that some values appear to improve significantly at lower conversion ratios while others appear far less favorable. Detailed safety analysis is required to determine if these designs have adequate safety margins or if appropriate design modifications are required. Detailed system analysis data has been generated for both metal and oxide-fueled core designs over the entire range of potential burner reactors. Additional data has been calculated for a few alternative fuel cycles. The systems data has been summarized in this report and the detailed data will be provided to the systems analysis team so that static and dynamic system analyses can be performed.

  1. Thermal-hydraulic design of tungsten rod bundles for the APT 3He neutron spallation target

    NASA Astrophysics Data System (ADS)

    Willcutt, Gordon J. E.

    1995-01-01

    A preconceptual design has been developed for the 3He Target/Blanket System for the Accelerator Production of Tritium Project. The design use tungsten wire-wrapped rods to produce neutrons when the rods are struck by a proton beam. The rods are contained in bundles inside hexagonal Inconel ducts and cooled by D2O. Rod bundles are grouped in patterns in the proton beam inside a chamber filled with 3He that is transmuted to tritium by the neutrons coming from the tungsten rods. Additional 3He is transmuted in a blanket region surrounding the helium chamber. This paper describes the initial thermal-hydraulic design and testing that has been completed to confirm the designed calculations for pressure drop through the bundle and heat transfer in the bundle. Heat transfer tests were run to verify steady-state operation. These tests were followed by increasing power until nucleate boiling occurs to determine operating margins. Changes that improve the initial design are described.

  2. Design and Analysis of 3D-MAPS: A Many-core 3D Processor with Stacked Memory

    E-print Network

    Lim, Sung Kyu

    Design and Analysis of 3D-MAPS: A Many-core 3D Processor with Stacked Memory Michael B. Healy, Krit--We describe the design and analysis of 3D-MAPS, a 64- core 3D-stacked memory-on-processor running at 277 MHz with 63 GB/s memory bandwidth, sent for fabrication using Tezzaron's 3D stacking technology. We also

  3. Design of a Resistively Heated Thermal Hydraulic Simulator for Nuclear Rocket Reactor Cores

    NASA Technical Reports Server (NTRS)

    Litchford, Ron J.; Foote, John P.; Ramachandran, Narayanan; Wang, Ten-See; Anghaie, Samim

    2007-01-01

    A preliminary design study is presented for a non-nuclear test facility which uses ohmic heating to replicate the thermal hydraulic characteristics of solid core nuclear reactor fuel element passages. The basis for this testing capability is a recently commissioned nuclear thermal rocket environments simulator, which uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce high-temperature pressurized hydrogen flows representative of reactor core environments, excepting radiation effects. Initially, the baseline test fixture for this non-nuclear environments simulator was configured for long duration hot hydrogen exposure of small cylindrical material specimens as a low cost means of evaluating material compatibility. It became evident, however, that additional functionality enhancements were needed to permit a critical examination of thermal hydraulic effects in fuel element passages. Thus, a design configuration was conceived whereby a short tubular material specimen, representing a fuel element passage segment, is surrounded by a backside resistive tungsten heater element and mounted within a self-contained module that inserts directly into the baseline test fixture assembly. With this configuration, it becomes possible to create an inward directed radial thermal gradient within the tubular material specimen such that the wall-to-gas heat flux characteristics of a typical fuel element passage are effectively simulated. The results of a preliminary engineering study for this innovative concept are fully summarized, including high-fidelity multi-physics thermal hydraulic simulations and detailed design features.

  4. Design of a nuclear reactor system for lunar base applications 

    E-print Network

    Griffith, Richard Odell

    1986-01-01

    to help me when I needed it. TABLE OF CONTENTS CHAPTER Page INTRODUCTION REAC. OR CORE DESIGN General Neutron Spectrum Fuel Type Structural Materials Reactor Coolant Operating Temperatures System Pressure Core Geometry Cor e Modeling Diffusion... APPENDIX C PROGRAM FLUX LISTING AND OUTPUT 145 VITA 151 vii LIST OF TABLES TABLE PAGE Fuel Pin Dimensions and Core Geometry Parameters . . . 28 III IV Cor e Par ameter s LASL Neutron Energy Groups P r ogr am CORE Out put 31 45 47 Burnup Values...

  5. Fluence-limited burnup as a function of fast reactor core parameters

    E-print Network

    Kersting, Alyssa (Alyssa Rae)

    2011-01-01

    The limiting factor in current designs for fast reactors is not only the reactivity, but also the maximum permissible fast-neutron fluence in the cladding, especially for reduced uranium enrichment cores using high-albedo ...

  6. Design and implementation of a multiaxial loading capability during heating on an engineering neutron diffractometer.

    PubMed

    Benafan, O; Padula, S A; Skorpenske, H D; An, K; Vaidyanathan, R

    2014-10-01

    A gripping capability was designed, implemented, and tested for in situ neutron diffraction measurements during multiaxial loading and heating on the VULCAN engineering materials diffractometer at the spallation neutron source at Oak Ridge National Laboratory. The proposed capability allowed for the acquisition of neutron spectra during tension, compression, torsion, and/or complex loading paths at elevated temperatures. The design consisted of age-hardened, Inconel(®) 718 grips with direct attachment to the existing MTS load frame having axial and torsional capacities of 100 kN and 400 N·m, respectively. Internal cooling passages were incorporated into the gripping system for fast cooling rates during high temperature experiments up to ?1000 K. The specimen mounting couplers combined a threaded and hexed end-connection for ease of sample installation/removal without introducing any unwanted loads. Instrumentation of this capability is documented in this work along with various performance parameters. The gripping system was utilized to investigate deformation in NiTi shape memory alloys under various loading/control modes (e.g., isothermal, isobaric, and cyclic), and preliminary results are presented. The measurements facilitated the quantification of the texture, internal strain, and phase fraction evolution in NiTi shape memory alloys under various loading/control modes. PMID:25362410

  7. Design and implementation of a multiaxial loading capability during heating on an engineering neutron diffractometer

    NASA Astrophysics Data System (ADS)

    Benafan, O.; Padula, S. A.; Skorpenske, H. D.; An, K.; Vaidyanathan, R.

    2014-10-01

    A gripping capability was designed, implemented, and tested for in situ neutron diffraction measurements during multiaxial loading and heating on the VULCAN engineering materials diffractometer at the spallation neutron source at Oak Ridge National Laboratory. The proposed capability allowed for the acquisition of neutron spectra during tension, compression, torsion, and/or complex loading paths at elevated temperatures. The design consisted of age-hardened, Inconel® 718 grips with direct attachment to the existing MTS load frame having axial and torsional capacities of 100 kN and 400 N.m, respectively. Internal cooling passages were incorporated into the gripping system for fast cooling rates during high temperature experiments up to ˜1000 K. The specimen mounting couplers combined a threaded and hexed end-connection for ease of sample installation/removal without introducing any unwanted loads. Instrumentation of this capability is documented in this work along with various performance parameters. The gripping system was utilized to investigate deformation in NiTi shape memory alloys under various loading/control modes (e.g., isothermal, isobaric, and cyclic), and preliminary results are presented. The measurements facilitated the quantification of the texture, internal strain, and phase fraction evolution in NiTi shape memory alloys under various loading/control modes.

  8. Design and implementation of a multiaxial loading capability during heating on an engineering neutron diffractometer

    SciTech Connect

    Benafan, O., E-mail: othmane.benafan@nasa.gov [NASA Glenn Research Center, Structures and Materials Division, Cleveland, Ohio 44135 (United States); Advanced Materials Processing and Analysis Center, Materials Science and Engineering Department, University of Central Florida, Orlando, Florida 32816 (United States); Padula, S. A. [NASA Glenn Research Center, Structures and Materials Division, Cleveland, Ohio 44135 (United States); Skorpenske, H. D.; An, K. [Spallation Neutron Source, Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Vaidyanathan, R. [Advanced Materials Processing and Analysis Center, Materials Science and Engineering Department, University of Central Florida, Orlando, Florida 32816 (United States)

    2014-10-01

    A gripping capability was designed, implemented, and tested for in situ neutron diffraction measurements during multiaxial loading and heating on the VULCAN engineering materials diffractometer at the spallation neutron source at Oak Ridge National Laboratory. The proposed capability allowed for the acquisition of neutron spectra during tension, compression, torsion, and/or complex loading paths at elevated temperatures. The design consisted of age-hardened, Inconel{sup ®} 718 grips with direct attachment to the existing MTS load frame having axial and torsional capacities of 100 kN and 400 N·m, respectively. Internal cooling passages were incorporated into the gripping system for fast cooling rates during high temperature experiments up to ~1000 K. The specimen mounting couplers combined a threaded and hexed end-connection for ease of sample installation/removal without introducing any unwanted loads. Instrumentation of this capability is documented in this work along with various performance parameters. The gripping system was utilized to investigate deformation in NiTi shape memory alloys under various loading/control modes (e.g., isothermal, isobaric, and cyclic), and preliminary results are presented. The measurements facilitated the quantification of the texture, internal strain, and phase fraction evolution in NiTi shape memory alloys under various loading/control modes.

  9. Design of a magnetic field mapping rover system for a neutron lifetime experiment

    NASA Astrophysics Data System (ADS)

    Libersky, Matthew; UCNTau Collaboration

    2014-09-01

    The beta decay lifetime of the free neutron is an important input to the Standard Model of particle physics, but values measured using different methods have exhibited substantial disagreement. The UCN ? experiment in development at Los Alamos National Laboratory (LANL) plans to explore better methods of measuring the neutron lifetime using ultracold neutrons (UCNs). In this experiment, UCNs are confined in a magneto-gravitational trap formed by a curved, asymmetric Halbach array placed inside a vacuum vessel and surrounded by holding field coils. If any defects present in the Halbach array are sufficient to reduce the local field near the surface below that needed to repel the desired energy level UCNs, loss by material interaction can occur at a rate similar to the loss by beta decay. A map of the magnetic field near the surface of the array is necessary to identify any such defects, but the array's curved geometry and placement in a vacuum vessel make conventional field mapping methods difficult. A system consisting of computer vision-based tracking and a rover holding a Hall probe has been designed to map the field near the surface of the array, and construction of an initial prototype has begun at LANL. A description of the design and prototype will be presented.

  10. Preliminary Neutronics Design Studies for a Molten Salt Blanket LIFE Engine

    SciTech Connect

    Powers, J

    2008-10-23

    The Laser Inertial Confinement Fusion Fission Energy (LIFE) Program being developed at Lawrence Livermore National Laboratory (LLNL) aims to design a hybrid fission-fusion subcritical nuclear engine that uses a laser-driven Inertial Confinement Fusion (ICF) system to drive a subcritical fission blanket. This combined fusion-fission hybrid system could be used for generating electricity, material transmutation or incineration, or other applications. LIFE does not require enriched fuel since it is a sub-critical system and LIFE can sustain power operation beyond the burnup levels at which typical fission reactors need to be refueled. In light of these factors, numerous options have been suggested and are being investigated. Options being investigated include fueling LIFE engines with spent nuclear fuel to aid in disposal/incineration of commercial spent nuclear fuel or using depleted uranium or thorium fueled options to enhance proliferation resistance and utilize non-fissile materials [1]. LIFE engine blanket designs using a molten salt fuel system represent one area of investigation. Possible applications of a LIFE engine with a molten salt blanket include uses as a spent nuclear fuel burner, fissile fuel breeding platform, and providing a backup alternative to other LIFE engine blanket designs using TRISO fuel particles in case the TRISO particles are found to be unable to withstand the irradiation they will be subjected to. These molten salts consist of a mixture of LiF with UF{sub 4} or ThF{sub 4} or some combination thereof. Future systems could look at using PuF{sub 3} or PuF{sub 4} as well, though no work on such system with initial plutonium loadings has been performed for studies documented in this report. The purpose of this report is to document preliminary neutronics design studies performed to support the development of a molten salt blanket LIFE engine option, as part of the LIFE Program being performed at Lawrence Livermore National laboratory. Preliminary design studies looking at fast ignition and hot spot ignition fusion options are documented, along with limited scoping studies performed to investigate other options of interest that surfaced during the main design effort. Lastly, side studies that were not part of the main design effort but may alter future work performed on LIFE engine designs are shown. The majority of all work reported in this document was performed during the Molten Salt Fast Ignition Moderator Study (MSFIMS) which sought to optimize the amount of moderator mixed into the molten salt region in order to produce the most compelling design. The studies in this report are of a limited scope and are intended to provide a preliminary neutronics analysis of the design concepts described herein to help guide decision processes and explore various options that a LIFE engine with a molten salt blanket might enable. None of the designs shown in this report, even reference cases selected for detailed description and analysis, have been fully optimized. The analyses were performed primarily as a neutronics study, though some consultation was made regarding thermal-hydraulic and structural concerns during both scoping out an initial model and subsequent to identifying a neutronics-based reference case to ensure that the design work contained no glaring mechanical or thermal issues that would preclude its feasibility. Any analyses and recommendations made in this report are either primarily or solely from the point of view of LIFE neutronics and ignore other fundamental issues related to molten salt fuel blankets such as chemical processing feasibility and political feasibility of a molten salt system.

  11. Comparison of a NuScale SMR conceptual core design using CASMO5/simulate5 and MCNP5

    SciTech Connect

    Haugh, B. [Studsvik Scandpower Inc., 1015 Ashes Drive, Wilmington, NC 28405 (United States); Mohamed, A. [NuScale Power Inc., 1100 NE Circle Blvd, Corvallis, OR 97330 (United States)

    2012-07-01

    A key issue during the initial start-ups of new Small Modular Reactors (SMRs) is the lack of operational data for reactor model validation. To help better understand the accuracy of the reactor analysis codes CASMO5 and SIMULATE5, higher order comparisons to MCNP5 have been performed. These comparisons are for an initial core conceptual design of the NuScale reactor. The data have been evaluated at Hot Zero Power (HZP) conditions. Comparisons of core reactivity, fuel temperature coefficient (FTC), and moderator temperature coefficients (MTC) have been performed. Comparison results show good agreement between CASMO5/SIMULATE5 and MCNP5 for the conceptual initial core design. (authors)

  12. Thermal-hydraulic criteria for the APT tungsten neutron source design

    SciTech Connect

    Pasamehmetoglu, K.

    1998-03-01

    This report presents the thermal-hydraulic design criteria (THDC) developed for the tungsten neutron source (TNS). The THDC are developed for the normal operations, operational transients, and design-basis accidents. The requirements of the safety analyses are incorporated into the design criteria, consistent with the integrated safety management and the safety-by-design philosophy implemented throughout the APT design process. The phenomenology limiting the thermal-hydraulic design and the confidence level requirements for each limit are discussed. The overall philosophy of the uncertainty analyses and the confidence level requirements also are presented. Different sets of criteria are developed for normal operations, operational transients, anticipated accidents, unlikely accidents, extremely unlikely accidents, and accidents during TNS replacement. In general, the philosophy is to use the strictest criteria for the high-frequency events. The criteria is relaxed as the event frequencies become smaller. The THDC must be considered as a guide for the design philosophy and not as a hard limit. When achievable, design margins greater than those required by the THDC must be used. However, if a specific event sequence cannot meet the THDC, expensive design changes are not necessary if the single event sequence results in sufficient margin to safety criteria and does not challenge the plant availability or investment protection considerations.

  13. A new design for post and core restorations implementing positive locking.

    PubMed

    Richter, Ernst-Jürgen; Boldt, Julian; Groth, Sybille; Proff, Peter; Gredes, Tomasz; Rottner, Kurt

    2008-10-01

    The design of a post and core restoration is a trade-off between a series of requirements to achieve stability of the post itself, the surrounding root dentine and the joint between tooth and post, while maintaining a sufficient apical seal of the remaining root canal filling. Post and core restoration systems come in a variety of different designs and dimensions, where each has its specific strength and weakness. With the exception of threaded versions, posts normally rely on either chemical and/or frictional locking between the post and the remaining root. Failure due to fatigue of the joint or root fracture due to overloading of the dentine is a frequent failure mode, especially for posts anchoring removable prostheses. Perforation of the root in an attempt to maximize the post length is a main cause for failure, too. A new design is proposed which uses a short but large diameter post. The risk of decementation is reduced by positive locking. A cavity with an undercut is prepared into the root, into which the post is fitted. Once joined, the post cannot be separated from the tooth without destruction of either the root or the post. The principle of the new design uses preparation tools and a post which is spread at the bottom. A cylindrically prepared hole is re-shaped to a defined inverse taper with the wider diameter at the bottom of the hole. A cylindrical post is inserted and spread at the bottom to a matching shape after placement. A first in vitro test of the stability showed that the positive locking provides at least as good extraction resistance as conventional post without the critical reliance on the luting/bonding agent. PMID:18840064

  14. The response of ex-core neutron detectors to large- and small-break loss-of-coolant accidents in pressurized water reactors

    SciTech Connect

    Okyere, E.W. (City Univ. of New York, Staten Island, NY (US)); Baratta, A.J.; Jester, W.A. (Pennsylvania State Univ., University Park, PA (United States). Dept. of Nuclear Engineering)

    1991-12-01

    This paper reports on a variety of water level measurement systems that are proposed to resolve the problem of reactor vessel level measurement. Two such systems, the heated thermocouple and the multiple differential pressure cell system, are used commercially. A third system based on ex-core neutron detectors was tested at the Pennsylvania State University Breazeale nuclear reactor facility and at the Idaho National Engineering Laboratory Loss-of-Fluid Test Facility. Results of these tests show that such a system is sensitive to both large- and small-break loss-of-coolant accidents and to voiding in the upper plenum of the vessel.

  15. Methodology for worker neutron exposure evaluation in the PDCF facility design.

    PubMed

    Scherpelz, R I; Traub, R J; Pryor, K H

    2004-01-01

    A project headed by Washington Group International is meant to design the Pit Disassembly and Conversion Facility (PDCF) to convert the plutonium pits from excessed nuclear weapons into plutonium oxide for ultimate disposition. Battelle staff are performing the shielding calculations that will determine appropriate shielding so that the facility workers will not exceed target exposure levels. The target exposure levels for workers in the facility are 5 mSv y(-1) for the whole body and 100 mSv y(-1) for the extremity, which presents a significant challenge to the designers of a facility that will process tons of radioactive material. The design effort depended on shielding calculations to determine appropriate thickness and composition for glove box walls, and concrete wall thicknesses for storage vaults. Pacific Northwest National Laboratory (PNNL) staff used ORIGEN-S and SOURCES to generate gamma and neutron source terms, and Monte Carlo (computer code for) neutron photon (transport) (MCNP-4C) to calculate the radiation transport in the facility. The shielding calculations were performed by a team of four scientists, so it was necessary to develop a consistent methodology. There was also a requirement for the study to be cost-effective, so efficient methods of evaluation were required. The calculations were subject to rigorous scrutiny by internal and external reviewers, so acceptability was a major feature of the methodology. Some of the issues addressed in the development of the methodology included selecting appropriate dose factors, developing a method for handling extremity doses, adopting an efficient method for evaluating effective dose equivalent in a non-uniform radiation field, modelling the reinforcing steel in concrete, and modularising the geometry descriptions for efficiency. The relative importance of the neutron dose equivalent compared with the gamma dose equivalent varied substantially depending on the specific shielding conditions and lessons were learned from this effect. This paper addresses these issues and the resulting methodology. PMID:15353738

  16. Conceptual design of a high-intensity positron source for the Advanced Neutron Source

    SciTech Connect

    Hulett, L.D.; Eberle, C.C.

    1994-12-01

    The Advanced Neutron Source (ANS) is a planned new basic and applied research facility based on a powerful steady-state research reactor that provides neutrons for measurements and experiments in the fields of materials science and engineering, biology, chemistry, materials analysis, and nuclear science. The useful neutron flux will be at least five times more than is available in the world`s best existing reactor facility. Construction of the ANS provides a unique opportunity to build a positron spectroscopy facility (PSF) with very-high-intensity beams based on the radioactive decay of a positron-generating isotope. The estimated maximum beam current is 1000 to 5000 times higher than that available at the world`s best existing positron research facility. Such an improvement in beam capability, coupled with complementary detectors, will reduce experiment durations from months to less than one hour while simultaneously improving output resolution. This facility will remove the existing barriers to the routine use of positron-based analytical techniques and will be a giant step toward realization of the full potential of the application of positron spectroscopy to materials science. The ANS PSF is based on a batch cycle process using {sup 64}Cu isotope as the positron emitter and represents the status of the design at the end of last year. Recent work not included in this report, has led to a proposal for placing the laboratory space for the positron experiments outside the ANS containment; however, the design of the positron source is not changed by that relocation. Hydraulic and pneumatic flight tubes transport the source material between the reactor and the positron source where the beam is generated and conditioned. The beam is then transported through a beam pipe to one of several available detectors. The design presented here includes all systems necessary to support the positron source, but the beam pipe and detectors have not been addressed yet.

  17. METHODOLOGY FOR WORKER NEUTRON EXPOSURE EVALUATION IN THE PDCF FACILITY DESIGN

    SciTech Connect

    Scherpelz, Robert I.; Traub, Richard J.; Pryor, Kathryn H.

    2004-08-01

    A project headed by Washington Group International is meant to design the Pit Disassembly and Conversion Facility (PDCF) to convert the plutonium pits from excessed nuclear weapons into plutonium oxide for ultimate disposition. Battelle staff are performing the shielding calculations that will determine appropriate shielding so that the facility workers will not exceed target exposure levels. The target exposure levels for workers in the facility are 5 mSv y?1 for the whole body and 100 mSv y?1 for the extremity, which presents a significant challenge to the designers of a facility that will process tons of radioactive material. The design effort depended on shielding calculations to determine appropriate thickness and composition for glove box walls, and concrete wall thicknesses for storage vaults. Pacific Northwest National Laboratory (PNNL) staff used ORIGEN-S and SOURCES to generate gamma and neutron source terms, and Monte Carlo (computer code for) neutron photon (transport) (MCNP-4C) to calculate the radiation transport in the facility. The shielding calculations were performed by a team of four scientists, so it was necessary to develop a consistent methodology. There was also a requirement for the study to be cost-effective, so efficient methods of evaluation were required. The calculations were subject to rigorous scrutiny by internal and external reviewers, so acceptability was a major feature of the methodology. Some of the issues addressed in the development of the methodology included selecting appropriate dose factors, developing a method for handling extremity doses, adopting an efficient method for evaluating effective dose equivalent in a non-uniform radiation field, modeling the reinforcing steel in concrete, and modularizing the geometry descriptions for efficiency. The relative importance of the neutron dose equivalent compared with the gamma dose equivalent varied substantially depending on the specific shielding conditions and lessons were learned from this effect. This paper addresses these issues and the resulting methodology.

  18. Conceptual design for a scintillating-fiber neutron detector for fusion reactor plasma diagnostics

    SciTech Connect

    Sailor, W.C.; Barnes, C.W.; Chrien, R.E.; Wurden, G.A.

    1994-07-01

    A conceptual design for a ``pointing`` neutron detector that is capable of delivering 10{sup 4} - 10{sup 5} Hz countrate of T(D,n) events from triton burnup at a deuterium-burning tokamak is described. The detector consists of collimated bundles of scintillating fibers that are separated by metal or polyethylene. These bundles in turn are set into a larger collimator that has some of the bundles set in ``unplugged`` holes and others in ``plugged`` holes whose countrate difference gives the net countrate. It is computed that the use of a 6 MeV{sub ee} (electron equivalent) discriminator will allow 14-MeV neutron countrates of 2xl0{sup 4} Hz in a DD machine or 3 MHz in a DT machine, while effectively rejecting the gamma background. The efficiency-area product for 14-MeV neutrons will be {similar_to} 0.014 cm{sup 2}. The angular resolution is computed to be 4.5{degree} HWHM for a 35 cm long collimator.

  19. Reliability Design for Neutron Induced Single-Event Burnout of IGBT

    NASA Astrophysics Data System (ADS)

    Shoji, Tomoyuki; Nishida, Shuichi; Ohnishi, Toyokazu; Fujikawa, Touma; Nose, Noboru; Hamada, Kimimori; Ishiko, Masayasu

    Single-event burnout (SEB) caused by cosmic ray neutrons leads to catastrophic failures in insulated gate bipolar transistors (IGBTs). It was found experimentally that the incident neutron induced SEB failure rate increases as a function of the applied collector voltage. Moreover, the failure rate increased sharply with an increase in the applied collector voltage when the voltage exceeded a certain threshold value (SEB cutoff voltage). In this paper, transient device simulation results indicate that impact ionization at the n-drift/n+ buffer boundary is a crucially important factor in the turning-on of the parasitic pnp transistor, and eventually latch-up of the parasitic thyristor causes SEB. In addition, the device parameter dependency of the SEB cutoff voltage was analytically derived from the latch-up condition of the parasitic thyristor. As a result, it was confirmed that reducing the current gain of the parasitic transistor, such as by increasing the n-drift region thickness d was effective in increasing the SEB cutoff voltage. Furthermore, `white' neutron-irradiation experiments demonstrated that suppressing the inherent parasitic thyristor action leads to an improvement of the SEB cutoff voltage. It was confirmed that current gain optimization of the parasitic transistor is a crucial factor for establishing highly reliable design against chance failures.

  20. Optimal design at inner core of the shaped pyramidal truss structure

    NASA Astrophysics Data System (ADS)

    Lee, Sung-Uk; Yang, Dong-Yol

    2013-12-01

    Sandwich material is a type of composite material with lightweight, high strength, good dynamic properties and high bending stiffness-to-weight ratio. This can be found well such structures in the nature (for example, internal structure of bones, plants, etc.). New trend which prefers eco-friendly products and energy efficiency is emerging in industries recently. Demand for materials with high strength and light weight is also increasing. In line with these trends, researches about manufacturing methods of sandwich material have been actively conducted. In this study, a sandwich structure named as "Shaped Pyramidal Truss Structure" is proposed to improve mechanical strength and to apply a manufacturing process suitable for massive production. The new sandwich structure was designed to enhance compressive strength by changing the cross-sectional shape at the central portion of the core. As the next step, optimization of the shape was required. Optimization technique used here was the SZGA(Successive Zooming Genetic Algorithm), which is one of GA(Genetic Algorithm) methods gradually reducing the area of design variable. The objective function was defined as moment of inertia of the cross-sectional shape of the strut. The control points of cubic Bezier curve, which was assumed to be the shape of the cross section, were used as design variables. By using FEM simulation, it was found that the structure exhibited superior mechanical properties compared to the simple design of the prior art.

  1. Preliminary probabilistic design accident evaluation of the cold source facilities of the advanced neutron source

    SciTech Connect

    Harrington, R.M.; Ramsey, C.T.

    1995-08-01

    Consistent with established Advanced Neutron Source (ANS) project policy for the use of probabilistic risk assessment (PRA) in design, a task has been established to use PRA techniques to help guide the design and safety analysis of the ANS cold sources. The work discussed in this report is the first formal output of the cold source PRA task. The major output at this stage is a list of design basis accidents, categorized into approximate frequency categories. This output is expected to focus attention on continued design work to define and optimize the design such that design basis accidents are better defined and have acceptable outcomes. Categorizing the design basis events (DBEs) into frequency categories should prove helpful because it will allow appropriate acceptance criteria to be applied. Because the design of the cold source is still proceeding, it is beyond the scope of this task to produce detailed event probability calculations or even, in some cases, detailed event sequence definitions. That work would take place as a logically planned follow-on task, to be completed as the design matures. Figure 1.1 illustrates the steps that would typically be followed in selecting design basis accidents with the help of PRA. Only those steps located above the dashed line on Fig. 1.1 are included in the scope of the present task. (Only an informal top-level failure modes and effects analysis was done.) With ANS project closeout expected in the near future, the scope of this task has been abbreviated somewhat beyond the state of available design information on the ANS cold sources, or what could be achieved in a reasonable time. This change was necessary to ensure completion before the closeout and because the in-depth analytical support necessary to define fully some of the accidents has already been curtailed.

  2. Melt spreading code assessment, modifications, and application to the EPR core catcher design.

    SciTech Connect

    Farmer, M. T .; Nuclear Engineering Division

    2009-03-30

    The Evolutionary Power Reactor (EPR) is under consideration by various utilities in the United States to provide base load electrical production, and as a result the design is undergoing a certification review by the U.S. Nuclear Regulatory Commission (NRC). The severe accident design philosophy for this reactor is based upon the fact that the projected power rating results in a narrow margin for in-vessel melt retention by external cooling of the reactor vessel. As a result, the design addresses ex-vessel core melt stabilization using a mitigation strategy that includes: (1) an external core melt retention system to temporarily hold core melt released from the vessel; (2) a layer of 'sacrificial' material that is admixed with the melt while in the core melt retention system; (3) a melt plug in the lower part of the retention system that, when failed, provides a pathway for the mixture to spread to a large core spreading chamber; and finally, (4) cooling and stabilization of the spread melt by controlled top and bottom flooding. The overall concept is illustrated in Figure 1.1. The melt spreading process relies heavily on inertial flow of a low-viscosity admixed melt to a segmented spreading chamber, and assumes that the melt mass will be distributed to a uniform height in the chamber. The spreading phenomenon thus needs to be modeled properly in order to adequately assess the EPR design. The MELTSPREAD code, developed at Argonne National Laboratory, can model segmented, and both uniform and nonuniform spreading. The NRC is thus utilizing MELTSPREAD to evaluate melt spreading in the EPR design. MELTSPREAD was originally developed to support resolution of the Mark I containment shell vulnerability issue. Following closure of this issue, development of MELTSPREAD ceased in the early 1990's, at which time the melt spreading database upon which the code had been validated was rather limited. In particular, the database that was utilized for initial validation consisted of: (1) comparison to an analytical solution for the dam break problem, (2) water spreading tests in a 1/10 linear scale model of the Mark I containment by Theofanous et al., and (3) steel spreading tests by Suzuki et al. that were also conducted in a geometry similar to the Mark I. The objective of this work was to utilize the MELTSPREAD code to check the assumption of uniform melt spreading in the EPR core catcher design. As a starting point for the project, the code was validated against the worldwide melt spreading database that emerged after the code was originally written in the very early 1990's. As part of this exercise, the code was extensively modified and upgraded to incorporate findings from these various analytical and experiment programs. In terms of expanding the ability of the code to analyze various melt simulant experiments, the options to input user-specified melt and/or substrate material properties was added. The ability to perform invisicid and/or adiabatic spreading analysis was also added so that comparisons with analytical solutions and isothermal spreading tests could be carried out. In terms of refining the capability to carry out reactor material melt spreading analyses, the code was upgraded with a new melt viscosity model; the capability was added to treat situations in which solid fraction buildup between the liquidus-solidus is non-linear; and finally, the ability to treat an interfacial heat transfer resistance between the melt and substrate was incorporated. This last set of changes substantially improved the predictive capability of the code in terms of addressing reactor material melt spreading tests. Aside from improvements and upgrades, a method was developed to fit the model to the various melt spreading tests in a manner that allowed uncertainties in the model predictions to be statistically characterized. With these results, a sensitivity study was performed to investigate the assumption of uniform spreading in the EPR core catcher that addressed parametric variations in: (1) melt pour mass, (2) melt composition, (3) me

  3. Dynamic analysis of sandwich panels and topological design of cores considering the size effect

    Microsoft Academic Search

    K. Qiu; W. Zhang; P. Duysinx

    A typical sandwich panel is composed of the upper and lower skins separated by a lightweight core, for instance, foams, trusshoneycombs and corrugated cores (1-4). And each kind of cores has various structural forms. For example, the foam cores have polymer and metallic ones with open-cell and close- cell ones; the truss cores have tetrahedral, pyramidal, 3D Kagome and diamond

  4. An Area-Efficient Variable Length Decoder IP Core Design for MPEG1\\/2\\/4 Video Coding Applications

    Microsoft Academic Search

    Chih-Da Chien; Keng-Po Lu; Yu-Min Chen; Jiun-In Guo; Yuan-Sun Chu; Ching-Lung Su

    2006-01-01

    This paper proposes an area-efficient variable length decoder (VLD) IP core design for MPEG-1\\/2\\/4 video coding applications. The proposed IP core exploits the parallel numerical matching in the MPEG-1\\/2\\/4 entropy decoding to achieve high data throughput rate in terms of limited hardware cost. This feature not only improves the performance of VLD, but also facilitates reducing the power consumption through

  5. The features of neutronic calculations for fast reactors with hybrid cores on the basis of BFS-62-3A critical assembly experiments

    SciTech Connect

    Mitenkova, E. F.; Novikov, N. V. [Nuclear Safety Inst. of Russian Academy of Sciences, B. Tulskaya 52, Moscow, 115119 (Russian Federation); Blokhin, A. I. [State Scientific Center of Russian Federation, Inst. of Physics and Power Engineering Named after A.I. Leypunsky, Bondarenko Square 1, Obninsk, Kaluga Region, 249030 (Russian Federation)

    2012-07-01

    The different (U-Pu) fuel compositions are considered for next generation of sodium fast breeder reactors. The considerable discrepancies in axial and radial neutron spectra for hybrid reactor systems compared to the cores with UO{sub 2} fuel cause increasing uncertainty of generating the group nuclear constants in those reactor systems. The calculation results of BFS-62-3A critical assembly which is considered as full-scale model of BN-600 hybrid core with steel reflector specify quite different spectra in local areas. For those systems the MCNP 5 calculations demonstrate significant sensitivity of effective multiplication factor K{sub eff} and spectral indices to nuclear data libraries. For {sup 235}U, {sup 238}U, {sup 239}Pu the results of calculated radial fission rate distributions against the reconstructed ones are analyzed. Comparative analysis of spectral indices, neutron spectra and radial fission rate distributions are performed using the different versions of ENDF/B, JENDL-3.3, JENDL-4, JEFF-3.1.1 libraries and BROND-3 for Fe, Cr isotopes. For analyzing the fission rate sensitivity to the plutonium presence in the fuel {sup 239}Pu is substituted for {sup 235}U (enrichment 90%) in the FA areas containing the plutonium. For {sup 235}U, {sup 238}U, {sup 239}Pu radial fission rate distributions the explanation of pick values discrepancies is based on the group fission constants analyses and possible underestimation of some features at the experimental data recovery method (Westcott factors, temperature dependence). (authors)

  6. Advanced Neutron Source (ANS) Project progress report

    SciTech Connect

    McBee, M.R.; Chance, C.M. (eds.) (Oak Ridge National Lab., TN (USA)); Selby, D.L.; Harrington, R.M.; Peretz, F.J. (Oak Ridge National Lab., TN (USA))

    1990-04-01

    This report discusses the following topics on the advanced neutron source: quality assurance (QA) program; reactor core development; fuel element specification; corrosion loop tests and analyses; thermal-hydraulic loop tests; reactor control concepts; critical and subcritical experiments; material data, structural tests, and analysis; cold source development; beam tube, guide, and instrument development; hot source development; neutron transport and shielding; I C research and development; facility concepts; design; and safety.

  7. Core design study of a supercritical light water reactor with double row fuel rods

    SciTech Connect

    Zhao, C.; Wu, H.; Cao, L.; Zheng, Y. [School of Nuclear Science and Technology, Xi'an Jiaotong Univ., No. 28, Xianning West Road, Xi'an, ShannXi, 710049 (China); Yang, J.; Zhang, Y. [China Nuclear Power Technology Research Inst., Yitian Road, ShenZhen, GuangDong, 518026 (China)

    2012-07-01

    An equilibrium core for supercritical light water reactor has been designed. A novel type of fuel assembly with dual rows of fuel rods between water rods is chosen and optimized to get more uniform assembly power distributions. Stainless steel is used for fuel rod cladding and structural material. Honeycomb structure filled with thermal isolation is introduced to reduce the usage of stainless steel and to keep moderator temperature below the pseudo critical temperature. Water flow scheme with ascending coolant flow in inner regions is carried out to achieve high outlet temperature. In order to enhance coolant outlet temperature, the radial power distributions needs to be as flat as possible through operation cycle. Fuel loading pattern and control rod pattern are optimized to flatten power distribution at inner regions. Axial fuel enrichment is divided into three parts to control axial power peak, which affects maximum cladding surface temperature. (authors)

  8. Computer simulations for rf design of a Spallation Neutron Source external antenna H- ion source.

    PubMed

    Lee, S W; Goulding, R H; Kang, Y W; Shin, K; Welton, R F

    2010-02-01

    Electromagnetic modeling of the multicusp external antenna H(-) ion source for the Spallation Neutron Source (SNS) has been performed in order to optimize high-power performance. During development of the SNS external antenna ion source, antenna failures due to high voltage and multicusp magnet holder rf heating concerns under stressful operating conditions led to rf characteristics analysis. In rf simulations, the plasma was modeled as an equivalent lossy metal by defining conductivity as sigma. Insulation designs along with material selections such as ferrite and Teflon could be included in the computer simulations to compare antenna gap potentials, surface power dissipations, and input impedance at the operating frequencies, 2 and 13.56 MHz. Further modeling and design improvements are outlined in the conclusion. PMID:20192395

  9. Computer simulations for rf design of a Spallation Neutron Source external antenna H ion source

    SciTech Connect

    Lee, Sung-Woo [ORNL] [ORNL; Goulding, Richard Howell [ORNL] [ORNL; Kang, Yoon W [ORNL] [ORNL; Shin, Ki [ORNL] [ORNL; Welton, Robert F [ORNL] [ORNL

    2010-01-01

    Electromagnetic modeling of the multicusp external antenna H ion source for the Spallation Neutron Source SNS has been performed in order to optimize high-power performance. During development of the SNS external antenna ion source, antenna failures due to high voltage and multicusp magnet holder rf heating concerns under stressful operating conditions led to rf characteristics analysis. In rf simulations, the plasma was modeled as an equivalent lossy metal by defining conductivity as . Insulation designs along with material selections such as ferrite and Teflon could be included in the computer simulations to compare antenna gap potentials, surface power dissipations, and input impedance at the operating frequencies, 2 and 13.56 MHz. Further modeling and design improvements are outlined in the conclusion.

  10. Soft error rate simulation and initial design considerations of neutron intercepting silicon chip (NISC)

    NASA Astrophysics Data System (ADS)

    Celik, Cihangir

    Advances in microelectronics result in sub-micrometer electronic technologies as predicted by Moore's Law, 1965, which states the number of transistors in a given space would double every two years. The most available memory architectures today have submicrometer transistor dimensions. The International Technology Roadmap for Semiconductors (ITRS), a continuation of Moore's Law, predicts that Dynamic Random Access Memory (DRAM) will have an average half pitch size of 50 nm and Microprocessor Units (MPU) will have an average gate length of 30 nm over the period of 2008-2012. Decreases in the dimensions satisfy the producer and consumer requirements of low power consumption, more data storage for a given space, faster clock speed, and portability of integrated circuits (IC), particularly memories. On the other hand, these properties also lead to a higher susceptibility of IC designs to temperature, magnetic interference, power supply, and environmental noise, and radiation. Radiation can directly or indirectly affect device operation. When a single energetic particle strikes a sensitive node in the micro-electronic device, it can cause a permanent or transient malfunction in the device. This behavior is called a Single Event Effect (SEE). SEEs are mostly transient errors that generate an electric pulse which alters the state of a logic node in the memory device without having a permanent effect on the functionality of the device. This is called a Single Event Upset (SEU) or Soft Error . Contrary to SEU, Single Event Latchup (SEL), Single Event Gate Rapture (SEGR), or Single Event Burnout (SEB) they have permanent effects on the device operation and a system reset or recovery is needed to return to proper operations. The rate at which a device or system encounters soft errors is defined as Soft Error Rate (SER). The semiconductor industry has been struggling with SEEs and is taking necessary measures in order to continue to improve system designs in nano-scale technologies. Prevention of SEEs has been studied and applied in the semiconductor industry by including radiation protection precautions in the system architecture or by using corrective algorithms in the system operation. Decreasing 10B content (20%of natural boron) in the natural boron of Borophosphosilicate glass (BPSG) layers that are conventionally used in the fabrication of semiconductor devices was one of the major radiation protection approaches for the system architecture. Neutron interaction in the BPSG layer was the origin of the SEEs because of the 10B (n,alpha) 7Li reaction products. Both of the particles produced have the capability of ionization in the silicon substrate region, whose thickness is comparable to the ranges of these particles. Using the soft error phenomenon in exactly the opposite manner of the semiconductor industry can provide a new neutron detection system based on the SERs in the semiconductor memories. By investigating the soft error mechanisms in the available semiconductor memories and enhancing the soft error occurrences in these devices, one can convert all memory using intelligent systems into portable, power efficient, directiondependent neutron detectors. The Neutron Intercepting Silicon Chip (NISC) project aims to achieve this goal by introducing 10B-enriched BPSG layers to the semiconductor memory architectures. This research addresses the development of a simulation tool, the NISC Soft Error Analysis Tool (NISCSAT), for soft error modeling and analysis in the semiconductor memories to provide basic design considerations for the NISC. NISCSAT performs particle transport and calculates the soft error probabilities, or SER, depending on energy depositions of the particles in a given memory node model of the NISC. Soft error measurements were performed with commercially available, off-the-shelf semiconductor memories and microprocessors to observe soft error variations with the neutron flux and memory supply voltage. Measurement results show that soft errors in the memories increase proportionally with the neutron flux, wherea

  11. Neutron Stars

    NASA Technical Reports Server (NTRS)

    Cottam, J.

    2007-01-01

    Neutron stars were discovered almost 40 years ago, and yet many of their most fundamental properties remain mysteries. There have been many attempts to measure the mass and radius of a neutron star and thereby constrain the equation of state of the dense nuclear matter at their cores. These have been complicated by unknown parameters such as the source distance and burning fractions. A clean, straightforward way to access the neutron star parameters is with high-resolution spectroscopy. I will present the results of searches for gravitationally red-shifted absorption lines from the neutron star atmosphere using XMM-Newton and Chandra.

  12. Large Core Code Evaluation Working Group Benchmark Problem Four: neutronics and burnup analysis of a large heterogeneous fast reactor. Part 1. Analysis of benchmark results. [LMFBR

    Microsoft Academic Search

    C. L. Cowan; R. Protsik; J. W. Lewellen

    1984-01-01

    The Large Core Code Evaluation Working Group Benchmark Problem Four was specified to provide a stringent test of the current methods which are used in the nuclear design and analyses process. The benchmark specifications provided a base for performing detailed burnup calculations over the first two irradiation cycles for a large heterogeneous fast reactor. Particular emphasis was placed on the

  13. Neutron cross-sections above 20 MeV for design and modeling of accelerator driven systems

    NASA Astrophysics Data System (ADS)

    Blomgren, J.

    2007-02-01

    One of the outstanding new developments in the field of partitioning and transmutation (P{&}T) concerns accelerator-driven systems (ADS) which consist of a combination of a high-power, high-energy accelerator, a spallation target for neutron production and a sub-critical reactor core. The development of the commercial critical reactors of today motivated a large effort on nuclear data up to about 20 MeV, and presently several million data points can be found in various data libraries. At higher energies, data are scarce or even non-existent. With the development of nuclear techniques based on neutrons at higher energies, nowadays there is a need also for higher-energy nuclear data. To provide alternative to this lack of data, a wide program on neutron-induced data related to ADS for P{&}T is running at the 20-180 MeV neutron beam facility at `The Svedberg Laboratory' (TSL), Uppsala. The programme encompasses studies of elastic scattering, inelastic neutron production, i.e., (n, xn') reactions, light-ion production, fission and production of heavy residues. Recent results are presented and future program of development is outlined.

  14. Design, construction, and characterization of a facility for neutron capture gamma ray analysis of sulfur in coal using californium-252

    SciTech Connect

    Layfield, J.R.

    1980-03-01

    A study of neutron capture gamma ray analysis of sulfur in coal using californium-252 as a neutron source is reported. Both internal and external target geometries are investigated. The facility designed for and used in this study is described. The external target geometry is found to be inappropriate because of the low thermal neutron flux at the sample location, which must be outside the biological shielding. The internal target geometry is found to have a sufficient thermal neutron flux, but an excessive gamma ray background. A water filled plastic facility, rather than the paraffin filled steel one used in this study, is suggested as a means of increasing flexibility and decreasing the beackground in the internal target geometry.

  15. Design, synthesis and biological evaluation of FLT3 covalent inhibitors with a resorcylic acid core.

    PubMed

    Xu, Jin; Ong, Esther H Q; Hill, Jeffrey; Chen, Anqi; Chai, Christina L L

    2014-12-01

    A series of simplified ring-opened resorcylic acid lactone (RAL) derivatives were conveniently synthesized to target FLT3 and its mutants either irreversibly or reversibly. Our design of covalent FLT3 inhibitors is based on cis-enone RALs (e.g., L-783,277) that have a ?-resorcylic acid as the core structure. The designed compounds contain three types of Michael acceptors (acrylamide, vinylsulfonamide and maleimide) as potential covalent traps of a cysteine residue at the binding site of kinases. A variety of functional substitutions were also introduced to maximize the binding interactions. Biological evaluations revealed that compound 17, despite the presence of a highly reactive maleimide Michael acceptor, is a potent covalent FLT3 inhibitor which shows some specificity in cellular assays. On the other hand, compounds 2 and 6 containing acrylamide or vinylsulfonamide groups are reversible towards FLT3 binding, and are potent and selective inhibitors of mutant FLT3-ITD versus wt-FLT3. They also inhibit cell proliferation in FLT3-ITD expressing cell line MV-4-11 as compared to wt-FLT3 expressing cell line THP-1 and non-FLT3 cell lines (K562, HL60 and Hek-293T). PMID:25456387

  16. The design and implementation of the parallel out-of-core ScaLAPACK LU, QR and Cholesky factorization routines

    SciTech Connect

    D`Azevedo, E.F.; Dongarra, J.J.

    1997-04-01

    This paper describes the design and implementation of three core factorization routines--LU, QR and Cholesky--included in the out-of-core extension of ScaLAPACK. These routines allow the factorization and solution of a dense system that is too large to fit entirely in physical memory. An image of the full matrix is maintained on disk and the factorization routines transfer sub-matrices into memory. The left-looking column-oriented variant of the factorization algorithm is implemented to reduce the disk I/O traffic. The routines are implemented using a portable I/O interface and utilize high performance ScaLAPACK factorization routines as in-core computational kernels. The authors present the details of the implementation for the out-of-core ScaLAPACK factorization routines, as well as performance and scalability results on the Intel Paragon.

  17. Laser Intertial Fusion Energy: Neutronic Design Aspects of a Hybrid Fusion-Fission Nuclear Energy System

    SciTech Connect

    Kramer, K

    2010-04-08

    This study investigates the neutronics design aspects of a hybrid fusion-fission energy system called the Laser Fusion-Fission Hybrid (LFFH). A LFFH combines current Laser Inertial Confinement fusion technology with that of advanced fission reactor technology to produce a system that eliminates many of the negative aspects of pure fusion or pure fission systems. When examining the LFFH energy mission, a significant portion of the United States and world energy production could be supplied by LFFH plants. The LFFH engine described utilizes a central fusion chamber surrounded by multiple layers of multiplying and moderating media. These layers, or blankets, include coolant plenums, a beryllium (Be) multiplier layer, a fertile fission blanket and a graphite-pebble reflector. Each layer is separated by perforated oxide dispersion strengthened (ODS) ferritic steel walls. The central fusion chamber is surrounded by an ODS ferritic steel first wall. The first wall is coated with 250-500 {micro}m of tungsten to mitigate x-ray damage. The first wall is cooled by Li{sub 17}Pb{sub 83} eutectic, chosen for its neutron multiplication and good heat transfer properties. The {sub 17}Pb{sub 83} flows in a jacket around the first wall to an extraction plenum. The main coolant injection plenum is immediately behind the Li{sub 17}Pb{sub 83}, separated from the Li{sub 17}Pb{sub 83} by a solid ODS wall. This main system coolant is the molten salt flibe (2LiF-BeF{sub 2}), chosen for beneficial neutronics and heat transfer properties. The use of flibe enables both fusion fuel production (tritium) and neutron moderation and multiplication for the fission blanket. A Be pebble (1 cm diameter) multiplier layer surrounds the coolant injection plenum and the coolant flows radially through perforated walls across the bed. Outside the Be layer, a fission fuel layer comprised of depleted uranium contained in Tristructural-isotropic (TRISO) fuel particles having a packing fraction of 20% in 2 cm diameter fuel pebbles. The fission blanket is cooled by the same radial flibe flow that travels through perforated ODS walls to the reflector blanket. This reflector blanket is 75 cm thick comprised of 2 cm diameter graphite pebbles cooled by flibe. The flibe extraction plenum surrounds the reflector bed. Detailed neutronics designs studies are performed to arrive at the described design. The LFFH engine thermal power is controlled using a technique of adjusting the {sup 6}Li/{sup 7}Li enrichment in the primary and secondary coolants. The enrichment adjusts system thermal power in the design by increasing tritium production while reducing fission. To perform the simulations and design of the LFFH engine, a new software program named LFFH Nuclear Control (LNC) was developed in C++ to extend the functionality of existing neutron transport and depletion software programs. Neutron transport calculations are performed with MCNP5. Depletion calculations are performed using Monteburns 2.0, which utilizes ORIGEN 2.0 and MCNP5 to perform a burnup calculation. LNC supports many design parameters and is capable of performing a full 3D system simulation from initial startup to full burnup. It is able to iteratively search for coolant {sup 6}Li enrichments and resulting material compositions that meet user defined performance criteria. LNC is utilized throughout this study for time dependent simulation of the LFFH engine. Two additional methods were developed to improve the computation efficiency of LNC calculations. These methods, termed adaptive time stepping and adaptive mesh refinement were incorporated into a separate stand alone C++ library name the Adaptive Burnup Library (ABL). The ABL allows for other client codes to call and utilize its functionality. Adaptive time stepping is useful for automatically maximizing the size of the depletion time step while maintaining a desired level of accuracy. Adaptive meshing allows for analysis of fixed fuel configurations that would normally require a computationally burdensome number of depletion zones. Alternatively, Adaptive M

  18. Reactivity studies on the advanced neutron source. [Advanced Neutron Source

    SciTech Connect

    Ryskamp, J.M.; Redmond, E.L. II; Fletcher, C.D.

    1990-01-01

    An Advanced Neutron Source (ANS) with a peak thermal neutron flux of about 8.5 {times} 10{sup 19} m{sup {minus}2}s{sup {minus}1} is being designed for condensed matter physics, materials science, isotope production, and fundamental physics research. The ANS is a new reactor-based research facility being planned by Oak Ridge National Laboratory (ORNL) to meet the need for an intense steady-state source of neutrons. The design effort is currently in the conceptual phase. A reference reactor design has been selected in order to examine the safety, performance, and costs associated with this one design. The ANS Project has an established, documented safety philosophy, and safety-related design criteria are currently being established. The purpose of this paper is to present analyses of safety aspects of the reference reactor design that are related to core reactivity events. These analyses include control rod worth, shutdown rod worth, heavy water voiding, neutron beam tube flooding, light water ingress, and single fuel element criticality. Understanding these safety aspects will allow us to make design modifications that improve the reactor safety and achieve the safety related design criteria. 8 refs., 3 tabs.

  19. Space nuclear-power reactor design based on combined neutronic and thermal-fluid analyses

    SciTech Connect

    Koenig, D.R.; Gido, R.G.; Brandon, D.I.

    1985-01-01

    The design and performance analysis of a space nuclear-power system requires sophisticated analytical capabilities such as those developed during the nuclear rocket propulsion (Rover) program. In particular, optimizing the size of a space nuclear reactor for a given power level requires satisfying the conflicting requirements of nuclear criticality and heat removal. The optimization involves the determination of the coolant void (volume) fraction for which the reactor diameter is a minimum and temperature and structural limits are satisfied. A minimum exists because the critical diameter increases with increasing void fraction, whereas the reactor diameter needed to remove a specified power decreases with void fraction. The purpose of this presentation is to describe and demonstrate our analytical capability for the determination of minimum reactor size. The analysis is based on combining neutronic criticality calculations with OPTION-code thermal-fluid calculations.

  20. The fuzzy clearing approach for a niching genetic algorithm applied to a nuclear reactor core design optimization problem

    Microsoft Academic Search

    Wagner F. Sacco; Marcelo D. Machado; Cláudio M. N. A. Pereira; Roberto Schirru

    2004-01-01

    This article extends previous efforts on genetic algorithms (GAs) applied to a core design optimization problem. We introduce the application of a new Niching Genetic Algorithm (NGA) to this problem and compare its performance to these previous works. The optimization problem consists in adjusting several reactor cell parameters, such as dimensions, enrichment and materials, in order to minimize the average

  1. Design of a boiling water reactor core based on an integrated blanket–seed thorium–uranium concept

    Microsoft Academic Search

    Alejandro Núñez-Carrera; Juan Luis François; Cecilia Martín-del-Campo; Gilberto Espinosa-Paredes

    2005-01-01

    This paper is concerned with the design of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material in an integrated blanket–seed (BS) assembly. The integrated BS concept comes from the fact that the blanket and the seed rods are located in the same assembly, and are burned out in a once-through cycle. The idea behind the

  2. Enhancing the Practicum Experience for Pre-Service Chemistry Teachers through Collaborative CoRe Design with Mentor Teachers

    ERIC Educational Resources Information Center

    Hume, Anne; Berry, Amanda

    2013-01-01

    This paper reports findings from an ongoing study exploring how the Content Representation (CoRe) design can be used as a tool to help chemistry student teachers begin acquiring the professional knowledge required to become expert chemistry teachers. Phase 2 of the study, reported in this paper, investigated how collaboration with school-based…

  3. How to Reduce Power in 3D IC Designs: A Case Study with OpenSPARC T2 Core

    E-print Network

    Lim, Sung Kyu

    packaging and cooling costs. In addition, the power of an IC has a significant impact on its reliabilityHow to Reduce Power in 3D IC Designs: A Case Study with OpenSPARC T2 Core Moongon Jung1 , Taigon as the driving force for 3D ICs, yet there have been few thorough design studies on how to reduce power in 3D ICs

  4. Design of a horizontal neutron reflectometer for the European Spallation Source

    NASA Astrophysics Data System (ADS)

    Nekrassov, D.; Trapp, M.; Lieutenant, K.; Moulin, J.-F.; Strobl, M.; Steitz, R.

    2014-08-01

    A design study of a horizontal neutron reflectometer adapted to the general baseline of the long pulse European Spallation Source (ESS) is presented. The instrument layout comprises advanced solutions for the neutron guide, high-resolution pulse shaping and beam bending onto a sample surface being thoroughly adjusted to the properties of the ESS. The length of this instrument is roughly 55 m, enabling ??/? resolutions from 0.5% to 10%. The incident beam is focused in horizontal plane to boost measurements of sample sizes of 1×1 cm2 and smaller with potential beam deflection in both downward and upward directions. The primary range of neutron wavelengths utilized by the instrument is 2-7.1 Å. If the wavelength range needs to be extended, then this is possible by utilizing only every second (third, fourth) pulse by suppressing all other pulses by the chopper system and thus increase the longest usable wavelength to 12.2 (17.3, 22.4) Å. Angles of incidence can be set between 0° and 9° with a total accessible q-range from 4×10-3 Å-1 up to 1 Å-1, while the ??/? resolution can be freely set. The instrument operates in both ?/? (free liquid surfaces) and ?/2? (solid-liquid, air-solid interfaces) geometries. The experimental setup will in particular enable direct studies on ultrathin films (d ?10 Å) and buried monolayers to multilayered structures of up to 3000 Å total thickness. The horizontal reflectometer will further foster investigations of hierarchical systems from nanometer to micrometer length scale (the latter by off-specular scattering), as well as their kinetics and dynamical properties, in particular under load (shear, pressure, external fields). Polarization and polarization analysis as well as the GISANS option are designed as potential modules to be implemented in the generic instrument layout. The instrument is highly flexible and offers a variety of different measurement modes. With respect to its mechanical components the instrument is exclusively based on current technology. Risks of failure for the chosen setup are minimum.

  5. Design of moderator and multiplier systems for D–T neutron source in BNCT

    Microsoft Academic Search

    M. R. Eskandari; S. Kashian

    2009-01-01

    Boron Neutron Capture Therapy (BNCT) is an outstanding way to treat Glioblastoma Multiforme. Epithermal neutrons with energy from 1eV to 10keV represent the most effective range for brain tumor therapy. In this research we have focused on 3H(d,n)4He reaction as a neutron source using Cock Craft Walton accelerator. High neutron yield with 14.1MeV energy can be generated via accelerating a

  6. Advanced MOX Core Design Study of Sodium Cooled Reactors in Current Feasibility Study on Commercialized Fast Reactor Cycle Systems in Japan

    SciTech Connect

    Mizuno, T.; Niwa, H. [Japan Nuclear Cycle development institute, O-arai Engineering Center, 4002 Narita-cho, O-arai-machi, Higashi-ibaraki-gun, Ibaraki-ken, 311-1393 (Japan)

    2002-07-01

    The Sodium cooled MOX core design studies are performed with the target burnup of 150 GWd/t and measures against the recriticality issues in core disruptive accidents (CDAs). Four types of core are comparatively studied in viewpoints of core performance and reliability. Result shows that all the types of core satisfy the target and that the homogeneous core with axial blanket partial elimination subassembly is the most superior concept in case the effectiveness of measures against recriticality issues by the axial blanket partial elimination is assured. (authors)

  7. Lead-Cooled Fast Reactor (LFR) Design: Safety, Neutronics, Thermal Hydraulics, Structural Mechanics, Fuel, Core, and Plant Design

    SciTech Connect

    Smith, C

    2010-02-22

    The idea of developing fast spectrum reactors with molten lead (or lead alloy) as a coolant is not a new one. Although initially considered in the West in the 1950s, such technology was not pursued to completion because of anticipated difficulties associated with the corrosive nature of these coolant materials. However, in the Soviet Union, such technology was actively pursued during the same time frame (1950s through the 1980s) for the specialized role of submarine propulsion. More recently, there has been a renewal of interest in the West for such technology, both for critical systems as well as for Accelerator Driven Subcritical (ADS) systems. Meanwhile, interest in the former Soviet Union, primarily Russia, has remained strong and has expanded well beyond the original limited mission of submarine propulsion. This section reviews the past and current status of LFR development.

  8. Measurements of the impedance matrix of a thermoacoustic core: applications to the design of thermoacoustic engines.

    PubMed

    Bannwart, Flávio C; Penelet, Guillaume; Lotton, Pierrick; Dalmont, Jean-Pierre

    2013-05-01

    The successful design of a thermoacoustic engine depends on the appropriate description of the processes involved inside the thermoacoustic core (TAC). This is a difficult task when considering the complexity of both the heat transfer phenomena and the geometry of the porous material wherein the thermoacoustic amplification process occurs. An attempt to getting round this difficulty consists in measuring the TAC transfer matrix under various heating conditions, the measured transfer matrices being exploited afterward into analytical models describing the complete apparatus. In this paper, a method based on impedance measurements is put forward, which allows the accurate measurement of the TAC transfer matrix, contrarily to the classical two-load method. Four different materials are tested, each one playing as the porous element allotted inside the TAC, which is submitted to different temperature gradients to promote thermoacoustic amplification. The experimental results are applied to the modeling of basic standing-wave and traveling-wave engines, allowing the prediction of the engine operating frequency and thermoacoustic amplification gain, as well as the optimum choice of the components surrounding the TAC. PMID:23654373

  9. Modular Approach to Launch Vehicle Design Based on a Common Core Element

    NASA Technical Reports Server (NTRS)

    Creech, Dennis M.; Threet, Grady E., Jr.; Philips, Alan D.; Waters, Eric D.; Baysinger, Mike

    2010-01-01

    With a heavy lift launch vehicle as the centerpiece of our nation's next exploration architecture's infrastructure, the Advanced Concepts Office at NASA's Marshall Space Flight Center initiated a study to examine the utilization of elements derived from a heavy lift launch vehicle for other potential launch vehicle applications. The premise of this study is to take a vehicle concept, which has been optimized for Lunar Exploration, and utilize the core stage with other existing or near existing stages and boosters to determine lift capabilities for alternative missions. This approach not only yields a vehicle matrix with a wide array of capabilities, but also produces an evolutionary pathway to a vehicle family based on a minimum development and production cost approach to a launch vehicle system architecture, instead of a purely performance driven approach. The upper stages and solid rocket booster selected for this study were chosen to reflect a cross-section of: modified existing assets in the form of a modified Delta IV upper stage and Castor-type boosters; potential near term launch vehicle component designs including an Ares I upper stage and 5-segment boosters; and longer lead vehicle components such as a Shuttle External Tank diameter upper stage. The results of this approach to a modular launch system are given in this paper.

  10. Supercool Neutrons (Ultracold Neutrons)

    E-print Network

    Martin, Jeff

    Supercool Neutrons (Ultracold Neutrons) Jeff Martin University of Winnipeg Skywalk 2007 Manitoba Research & Innovation Fund #12; Ultracold Neutrons What are neutrons? Why are they important? How to make lots of neutrons. Interesting properties of ultracold neutrons (UCN) Supercool

  11. The design of an intense accelerator-based epithermal neutron beam prototype for BNCT using near-threshold reactions

    NASA Astrophysics Data System (ADS)

    Lee, Charles L.

    Near-threshold boron neutron capture therapy (BNCT) uses proton energies only tens of rev above the (pan) reaction threshold in lithium in order to reduce the moderation requirements of the neutron source. The goals of this research were to prove the feasibility of this near-threshold concept for BNCT applications, using both calculation and experiment, and design a compact neutron source prototype from these results. This required a multidisciplinary development of methods for calculation of neutron yields, head phantom dosimetry, and accelerator target heat removal. First, a method was developed to accurately calculate thick target neutron yields for both near-threshold and higher energy proton beams, in lithium metal as well as lithium compounds. After these yields were experimentally verified, they were used as neutron sources for Monte Carlo (MCNP) simulations of neutron and photon transport in head phantoms. The theoretical and experimental determination of heat removal from a target backing with multiple fins, as well as numerical calculations of heat deposition profiles based on proton energy loss in target and backing materials, demonstrated that lithium integrity can be maintained for proton beam currents up to 2.5 mA. The final design uses a proton beam energy of 1.95 MeV and has a centerline epithermal neutron flux of 2.2 × 108 n/cm2- sec/mA, an advantage depth of 5.7 cm, an advantage ratio of 4.3, and an advantage depth dose rate of 6.7 RBE- cGy/min/mA, corresponding to an irradiation time of 38 minutes with a 5 mA beam. Moderator, reflector, and shielding weigh substantially less than other accelerator BNCT designs based on higher proton energies, e.g. 2.5 MeV. The near-threshold concept is useful as a portable neutron source for hospital settings, with applications ranging from glioblastomas to melanomas and synovectomy. (Copies available exclusively from MIT Libraries, Rm. 14- 0551, Cambridge, MA 02139-4307. Ph. 617-253-5668; Fax 617-253-1690.)

  12. PHISICS multi-group transport neutronic capabilities for RELAP5

    SciTech Connect

    Epiney, A.; Rabiti, C.; Alfonsi, A.; Wang, Y.; Cogliati, J.; Strydom, G. [Idaho National Laboratory (INL), 2525 N. Fremont Ave., Idaho Falls, ID 83402 (United States)

    2012-07-01

    PHISICS is a neutronic code system currently under development at INL. Its goal is to provide state of the art simulation capability to reactor designers. This paper reports on the effort of coupling this package to the thermal hydraulic system code RELAP5. This will enable full prismatic core and system modeling and the possibility to model coupled (thermal-hydraulics and neutronics) problems with more options for 3D neutron kinetics, compared to the existing diffusion theory neutron kinetics module in RELAP5 (NESTLE). The paper describes the capabilities of the coupling and illustrates them with a set of sample problems. (authors)

  13. An overview of the planned advanced neutron source facility

    SciTech Connect

    West, C.D.

    1990-01-01

    The Advanced Neutron Source (ANS), now in the conceptual design stage, will be a new user facility for neutron research, including neutron beam experiments, materials irradiation testing and materials analysis capabilities, and production facilities for transuranic and lighter isotopes. The neutron source is to be the world's highest flux beam reactor and is based on existing reactor technology to minimize safety issues. The preferred fuel, U{sub 3}Si{sub 2}, has been tested in operating reactors in the United States, Japan, and Europe. The core is cooled, moderated, and reflected by heavy water, common practice for research reactors. 3 refs., 9 figs., 3 tabs.

  14. Design of hollow-core photonic bandgap fibers for CH4 and C2H2 optical fiber sensors

    Microsoft Academic Search

    Qiu-guo Wang; Bo-jun Yang

    2009-01-01

    A hollow-core photonic bandgap fiber for optical fiber sensors is designed. According to the calculation, we get the parameters of the photonic bandgap fiber as d=0.94*2.60?m, R=?=2.60?m for C2H2, and d=0.94*2.41?m, R=?=2.41?m for CH4. Around the wavelength of ?C2 H2 = 1.53?m and ?CH4 = 1.66?m, 98% of the energy is propagated in the core, the fiber is suitable to

  15. Strong volume reduction of common mode choke for RFI filters with the help of nanocrystalline cores design and experiments

    NASA Astrophysics Data System (ADS)

    Thierry, Waeckerlé; Thierry, Save; Benoît, Vachey; Dominique, Gautard

    2006-09-01

    The design of common mode chokes (CMC) for radio frequency interference filter is considered, explained and applied to the compared magnetic core characterization and computation between ferrite and nanocrystalline materials. The latter shows some better insertion losses at low frequencies when comparison is made with the same component characteristics. The volume of the component can be reduced by 50-80% when an appropriate ferrite is replaced by a well-chosen nanocrystalline core, as a result of its superimposed advantages of high and tailored permeabilitys, high saturation and inductive behavior near CMC resonance.

  16. The influence of various core designs on stress distribution in the veneered zirconia crown: a finite element analysis study

    PubMed Central

    Ha, Seung-Ryong; Kim, Sung-Hun; Yoo, Seung-Hyun; Jeong, Se-Chul; Lee, Jai-Bong; Yeo, In-Sung

    2013-01-01

    PURPOSE The purpose of this study was to evaluate various core designs on stress distribution within zirconia crowns. MATERIALS AND METHODS Three-dimensional finite element models, representing mandibular molars, comprising a prepared tooth, cement layer, zirconia core, and veneer porcelain were designed by computer software. The shoulder (1 mm in width) variations in core were incremental increases of 1 mm, 2 mm and 3 mm in proximal and lingual height, and buccal height respectively. To simulate masticatory force, loads of 280 N were applied from three directions (vertical, at a 45° angle, and horizontal). To simulate maximum bite force, a load of 700 N was applied vertically to the crowns. Maximum principal stress (MPS) was determined for each model, loading condition, and position. RESULTS In the maximum bite force simulation test, the MPSs on all crowns observed around the shoulder region and loading points. The compressive stresses were located in the shoulder region of the veneer-zirconia interface and at the occlusal region. In the test simulating masticatory force, the MPS was concentrated around the loading points, and the compressive stresses were located at the 3 mm height lingual shoulder region, when the load was applied horizontally. MPS increased in the shoulder region as the shoulder height increased. CONCLUSION This study suggested that reinforced shoulder play an essential role in the success of the zirconia restoration, and veneer fracture due to occlusal loading can be prevented by proper core design, such as shoulder. PMID:23755346

  17. Optimization algorithms in boiling water reactor lattice design

    E-print Network

    Burns, Chad (Chad D.), III

    2013-01-01

    Given the highly complex nature of neutronics and reactor physics, efficient methods of optimizing are necessary to effectively design the core reloading pattern and operate a nuclear reactor. The current popular methods ...

  18. Efficient transformer design by computing core loss using a novel approach

    Microsoft Academic Search

    A. Basak; Chi-Hang Yu; G. Lloyd

    1994-01-01

    Manufacturers and users of transformers are nowadays capitalising the core losses while considering the costings. Therefore a software package which enables an engineer to predict within a short period the approximate core loss in a transformer of any rating and geometry will be very useful. This paper describes how such a 2-D package has been developed using the finite element

  19. Conceptual Design of a Modular Island Core Fast Breeder Reactor “RAPID-M”

    Microsoft Academic Search

    Mitsuru KAMBE

    2002-01-01

    A metal fueled modular island core sodium cooled fast breeder reactor concept RAPID-M to improve reactor performance and proliferation resistance and to accommodate various power requirements has been demonstrated. The essential feature of the RAPID-M concept is that the reactor core consists of integrated fuel assemblies (IFAs) instead of conventional fuel subassemblies. The RAPID concept enables quick and simplified refueling

  20. Effective design of a sandwich beam with a metal foam core

    Microsoft Academic Search

    E. Magnucka-Blandzi; K. Magnucki

    2007-01-01

    The subject of the paper is a simply supported sandwich beam with a metal foam core. Mechanical properties of the core vary through its depth. A nonlinear hypothesis of deformation of a plane cross section of the beam is assumed and described. The elastic potential energy and the work of the load are formulated. The system of differential equations of

  1. Designing fast LTL model checking algorithms for many-core GPUs

    Microsoft Academic Search

    Ji?í Barnat; Petr Bauch; Luboš Brim; Milan ?eška

    Recent technological developments made various many-core hardware platforms widely accessible. These massively parallel architectures have been used to significantly accelerate many computation demanding tasks. In this paper, we show how the algorithms for LTL model checking can be redesigned in order to accelerate LTL model checking on many-core GPU platforms. Our detailed experimental evaluation demonstrates that using the NVIDIA CUDA

  2. Designing the core zone in a biosphere reserve based on suitable habitats: Yancheng Biosphere Reserve and the red crowned crane ( Grus japonensis)

    Microsoft Academic Search

    Wenjun Li; Zijian Wang; Zhijun Ma; Hongxiao Tang

    1999-01-01

    Although much research has been undertaken to design nature reserves, there are few practical methods to determine the interior structure of a reserve. A procedure for design of the core zone in reserves is proposed. As a case study, the core zone in Yancheng Biosphere Reserve, People's Republic of China, which was established to preserve the endangered red crowned crane

  3. LOW LOSS DESIGN OF THE LINAC AND ACCUMULATOR RING FOR THE SPALLATION NEUTRON SOURCE.

    SciTech Connect

    RAPARIA,D.

    2003-02-03

    The Spallation Neutron Source (SNS) is a second generation pulsed neutron source and is presently in the fourth year of a seven-year construction cycle at Oak Ridge National Laboratory. A collaboration of six national laboratories (ANL, BNL, LANL, LBNL, ORNL, TJNAF) is responsible for the design and construction of the various subsystems. The operation of the facility will begin in 2006 and deliver a 1.0 GeV, 1.4 MW proton beam with pulse length of 650 nanosecond at a repetition rate of 60 Hz, on a liquid mercury target. It consists of an RF volume H{sup -} source of 50 mA peak current at 6% duty; an all electrostatic Low-Energy Beam Transport (LEBT) which also serves as a first stage beam chopper with {+-} 25 ns rise/fall time; a 402.5 MHz, 4-vane Radio-Frequency Quadrupole (RFQ) for acceleration up to 2.5 MeV; a Medium Energy Beam Transport (MEBT) housing a second stage chopper (<{+-} 10ns rise/fall), an adjustable beam halo scraper, and diagnostics devices; a 6-tank Drift Tube Linac (DTL) with permanent magnet quadrupoles up to 87 MeV; an 805 MHz, 4-module, Side Coupled Cavity Linac (CCL) up to 186 MeV; an 805 MHz, superconducting RF (SRF) linac with eleven medium beta ({beta} = 0.61) cryo-modules and twelve high beta ({beta} = 0.81) cryo-modules accelerating the beam to the full energy; a High Energy Beam transport (HEBT) for diagnostics, transverse and longitudinal collimation, energy correction, painting and matching; an accumulator ring compressing the 1 GeV, 1 ms pulse to 650 ns for delivery onto the target through a Ring to Target Beam Transport (RTBT) with transverse collimators.

  4. Design and fabrication of hollow-core photonic crystal fibers for high-power ultrashort pulse transportation and pulse compression.

    PubMed

    Wang, Y Y; Peng, Xiang; Alharbi, M; Dutin, C Fourcade; Bradley, T D; Gérôme, F; Mielke, Michael; Booth, Timothy; Benabid, F

    2012-08-01

    We report on the recent design and fabrication of kagome-type hollow-core photonic crystal fibers for the purpose of high-power ultrashort pulse transportation. The fabricated seven-cell three-ring hypocycloid-shaped large core fiber exhibits an up-to-date lowest attenuation (among all kagome fibers) of 40 dB/km over a broadband transmission centered at 1500 nm. We show that the large core size, low attenuation, broadband transmission, single-mode guidance, and low dispersion make it an ideal host for high-power laser beam transportation. By filling the fiber with helium gas, a 74 ?J, 850 fs, and 40 kHz repetition rate ultrashort pulse at 1550 nm has been faithfully delivered at the fiber output with little propagation pulse distortion. Compression of a 105 ?J laser pulse from 850 fs down to 300 fs has been achieved by operating the fiber in ambient air. PMID:22859102

  5. System design specification for rotary mode core sample trucks No. 2, 3, and 4 programmable logic controller

    SciTech Connect

    Dowell, J.L.; Akers, J.C.

    1995-12-31

    The system this document describes controls several functions of the Core Sample Truck(s) used to obtain nuclear waste samples from various underground storage tanks at Hanford. The system will monitor the sampling process and provide alarms and other feedback to insure the sampling process is performed within the prescribed operating envelope. The intended audience for this document is anyone associated with rotary or push mode core sampling. This document describes the Alarm and Control logic installed on Rotary Mode Core Sample Trucks (RMCST) {number_sign}2, 3, and 4. It is intended to define the particular requirements of the RMCST alarm and control operation (not defined elsewhere) sufficiently for detailed design to implement on a Programmable Logic Controller (PLC).

  6. Optimum design of a moderator system based on dose calculation for an accelerator driven Boron Neutron Capture Therapy.

    PubMed

    Inoue, R; Hiraga, F; Kiyanagi, Y

    2014-06-01

    An accelerator based BNCT has been desired because of its therapeutic convenience. However, optimal design of a neutron moderator system is still one of the issues. Therefore, detailed studies on materials consisting of the moderator system are necessary to obtain the optimal condition. In this study, the epithermal neutron flux and the RBE dose have been calculated as the indicators to look for optimal materials for the filter and the moderator. As a result, it was found that a combination of MgF2 moderator with Fe filter gave best performance, and the moderator system gave a dose ratio greater than 3 and an epithermal neutron flux over 1.0×10(9)cm(-2)s(-1). PMID:24440538

  7. Thermal Hydraulics Design and Analysis Methodology for a Solid-Core Nuclear Thermal Rocket Engine Thrust Chamber

    NASA Technical Reports Server (NTRS)

    Wang, Ten-See; Canabal, Francisco; Chen, Yen-Sen; Cheng, Gary; Ito, Yasushi

    2013-01-01

    Nuclear thermal propulsion is a leading candidate for in-space propulsion for human Mars missions. This chapter describes a thermal hydraulics design and analysis methodology developed at the NASA Marshall Space Flight Center, in support of the nuclear thermal propulsion development effort. The objective of this campaign is to bridge the design methods in the Rover/NERVA era, with a modern computational fluid dynamics and heat transfer methodology, to predict thermal, fluid, and hydrogen environments of a hypothetical solid-core, nuclear thermal engine the Small Engine, designed in the 1960s. The computational methodology is based on an unstructured-grid, pressure-based, all speeds, chemically reacting, computational fluid dynamics and heat transfer platform, while formulations of flow and heat transfer through porous and solid media were implemented to describe those of hydrogen flow channels inside the solid24 core. Design analyses of a single flow element and the entire solid-core thrust chamber of the Small Engine were performed and the results are presented herein

  8. Design of a GEM-based detector for the measurement of fast neutrons

    Microsoft Academic Search

    B. Esposito; F. Murtas; R. Villari; M. Angelone; D. Marocco; M. Pillon; S. Puddu

    2010-01-01

    A novel neutron detector has been developed and tested in collaboration between LNF-INFN and ENEA-Frascati. The aim is to obtain a versatile system that can be employed for the simultaneous measurement of the neutron flux in various energy bands from 1 to 20MeV. The main drive for this development is the need of neutron detectors with low sensitivity to ?-rays

  9. Oyster Creek fuel thermal margin during core thermal-hydraulic oscillations

    SciTech Connect

    Dougher, J.D.

    1990-01-01

    The Oyster Creek nuclear facility, a boiling water reactor (BWR)-2 plant type, has never experienced core thermal-hydraulic instability. Power oscillations, however, have been observed in other BWR cores both domestically and internationally. Two modes of oscillations have been observed, core wide and regional half-core. During core wide oscillations, the neutron flux in the core oscillates in the radial fundamental mode. During regional half-core oscillations, higher order harmonics in the radial plane result in out-of-phase oscillations with the neutron flux in one half of the core oscillating 180 deg out-of-phase with the neutron flux in the other half of the core. General Design Criteria 12 requires either prevention or detection and suppression of power oscillations which could result in violations of fuel design limits. Analyses performed by General Electric have demonstrated that for large-magnitude oscillations the potential exists for violation of the safety limit minimum critical power ratio (MCPR). However, for plants with a flow-biased neutron flux scram automatic mitigation of oscillations may be provided at an oscillation magnitude below that at which the safety limit is challenged. Plant-specific analysis for Oyster Creek demonstrates that the existing average power range monitor (APRM) system will sense and suppress power oscillations prior to violation of any safety limits.

  10. Fusion-neutron-yield, activation measurements at the Z accelerator: Design, analysis, and sensitivity

    SciTech Connect

    Hahn, K. D., E-mail: kdhahn@sandia.gov; Ruiz, C. L.; Fehl, D. L.; Chandler, G. A.; Knapp, P. F.; Smelser, R. M.; Torres, J. A. [Sandia National Laboratories, Diagnostics and Target Physics, Albuquerque, New Mexico 87123 (United States)] [Sandia National Laboratories, Diagnostics and Target Physics, Albuquerque, New Mexico 87123 (United States); Cooper, G. W.; Nelson, A. J. [Department of Chemical and Nuclear Engineering, University of New Mexico, Albuquerque, New Mexico 87131 (United States)] [Department of Chemical and Nuclear Engineering, University of New Mexico, Albuquerque, New Mexico 87131 (United States); Leeper, R. J. [Los Alamos National Laboratories, Plasma Physics Group, Los Alamos, New Mexico 87545 (United States)] [Los Alamos National Laboratories, Plasma Physics Group, Los Alamos, New Mexico 87545 (United States)

    2014-04-15

    We present a general methodology to determine the diagnostic sensitivity that is directly applicable to neutron-activation diagnostics fielded on a wide variety of neutron-producing experiments, which include inertial-confinement fusion (ICF), dense plasma focus, and ion beam-driven concepts. This approach includes a combination of several effects: (1) non-isotropic neutron emission; (2) the 1/r{sup 2} decrease in neutron fluence in the activation material; (3) the spatially distributed neutron scattering, attenuation, and energy losses due to the fielding environment and activation material itself; and (4) temporally varying neutron emission. As an example, we describe the copper-activation diagnostic used to measure secondary deuterium-tritium fusion-neutron yields on ICF experiments conducted on the pulsed-power Z Accelerator at Sandia National Laboratories. Using this methodology along with results from absolute calibrations and Monte Carlo simulations, we find that for the diagnostic configuration on Z, the diagnostic sensitivity is 0.037% ± 17% counts/neutron per cm{sup 2} and is ? 40% less sensitive than it would be in an ideal geometry due to neutron attenuation, scattering, and energy-loss effects.

  11. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

    SciTech Connect

    Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

    2010-09-01

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.

  12. Hybrid wireless Network on Chip: a new paradigm in multi-core design

    Microsoft Academic Search

    Partha Pratim Pande; Amlan Ganguly; Kevin Chang; Christof Teuscher

    2009-01-01

    The performance benefits of conventional Network-on-Chip (NoC) architectures are limited by the high latency and energy dissipation in long distance multihop communication between embedded cores. This limitation of conventional NoC architectures can be addressed by introducing long-range, high bandwidth and low power wireless links between the distant cores. Using miniaturized on-chip antennas as an enabling technology, wireless NoCs (WiNoCs) can

  13. Design of a single-polarization single-mode photonic crystal fiber double-core coupler

    Microsoft Academic Search

    Jianfeng Li; Kailiang Duan; Yishan Wang; Xiangjie Cao; Yongkang Guo; Xiangdi Lin

    2009-01-01

    A simple single-polarization single-mode (SPSM) photonic crystal fiber (PCF) coupler with two cores is introduced. The full-vector finite-element method (FEM) is applied to analyze the modal interference phenomenon of the even and odd modes of two orthogonal polarizations and the power propagation within the two cores. Meanwhile, the SPSM coupling wavelength range and its corresponding coupling length for different structure

  14. Designing Scalable FPGA-Based Reduction Circuits Using Pipelined Floating-Point Cores

    Microsoft Academic Search

    Ling Zhuo; Gerald R. Morris; Viktor K. Prasanna

    2005-01-01

    The use of pipelined floating-point arithmetic cores to create high-performance FPGA-based computational ker- nels has introduced a new class of problems that do not exist when using single-cycle arithmetic cores. In par- ticular, the data hazards associated with pipelined floating-point reduction circuits can limit the scalabil- ity or severely reduce the performance of an otherwise high-performancecomputationalkernel. The inability to ef-

  15. Design of a TOF-SANS instrument for the proposed Long Wavelength Target Station at the Spallation Neutron Source.

    SciTech Connect

    Thiyagarajan, P.; Littrell, K.; Seeger, P. A.

    2000-11-28

    We have designed a versatile high-throughput SANS instrument [Broad Range Intense Multipurpose SANS (BRIMS)] for the proposed Long Wavelength Target Station at the SNS by using acceptance diagrams and the Los Alamos NISP Monte Carlo simulation package. This instrument has been fully optimized to take advantage of the 10 Hz source frequency (broad wavelength bandwidth) and the cold neutron spectrum from a tall coupled solid methane moderator (12 cm x 20 cm). BRIMS has been designed to produce data in a Q range spanning from 0.001 to 0.7 {angstrom}{sup {minus}1} in a single measurement by simultaneously using neutrons with wavelengths ranging from 1 to 14.5 {angstrom} in a time of flight mode. A supermirror guide and bender assembly is employed to separate and redirect the useful portion of the neutron spectrum with {lambda} > 1 {angstrom}, by 2.3{degree} away from the direct beam containing high energy neutrons and {gamma} rays. The effects of the supermirror coating of the guide, the location of the bender assembly with respect to the source, the bend angle, and various collimation choices on the flux, resolution and Q{sub min} have been characterized using spherical particle and delta function scatterers. The overall performance of BRIMS has been compared with that of the best existing reactor-based SANS instrument D22 at ILL.

  16. All Metal Iron Core For A Low Aspect Ratio Tokamak

    SciTech Connect

    D.A. Gates, C. Jun, I. Zatz, A. Zolfaghari

    2010-06-02

    A novel concept for incorporating a iron core transformer within a axisymmetric toroidal plasma containment device with a high neutron flux is described. This design enables conceptual design of low aspect ratio devices which employ standard transformer-driven plasma startup by using all-metal high resistance separators between the toroidal field windings. This design avoids the inherent problems of a multiturn air core transformer which will inevitably suffer from strong neutron bombardment and hence lose the integrity of its insulation, both through long term material degradation and short term neutron- induced conductivity.. A full 3-dimensional model of the concept has been developed within the MAXWELL program and the resultant loop voltage calculated. The utility of the result is found to be dependent on the resistivity of the high resistance separators. Useful loop voltage time histories have been obtained using achievable resistivities.

  17. Preliminary neutronics design of china lead-alloy cooled demonstration reactor (CLEAR-III) for nuclear waste transmutation

    SciTech Connect

    Chen, Z. [Inst. of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Southwest Science and Technology Univ., No.350 Shushanhu Road, Shushan District, Hefei, Anhui, 230031 (China); Chen, Y.; Bai, Y.; Wang, W.; Chen, Z.; Hu, L.; Long, P. [Inst. of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); School of Nuclear Science and Technology, Univ. of Science and Technology of China, Hefei, Anhui, 230031 (China)

    2012-07-01

    China Lead-Alloy cooled Demonstration Reactor (CLEAR-III), which is the concept of lead-bismuth cooled accelerator driven sub-critical reactor for nuclear waste transmutation, was proposed and designed by FDS team in China. In this study, preliminary neutronics design studies have primarily focused on three important performance parameters including Transmutation Support Ratio (TSR), effective multiplication factor and blanket thermal power. The constraint parameters, such as power peaking factor and initial TRU loading, were also considered. In the specific design, uranium-free metallic dispersion fuel of (TRU-Zr)-Zr was used as one of the CLEAR-III fuel types and the ratio between MA and Pu was adjusted to maximize transmutation ratio. In addition, three different fuel zones differing in the TRU fraction of the fuel were respectively employed for this subcritical reactor, and the zone sizes and TRU fractions were determined such that the linear powers of these zones were close to each other. The neutronics calculations and analyses were performed by using Multi-Functional 4D Neutronics Simulation System named VisualBUS and nuclear data library HENDL (Hybrid Evaluated Nuclear Data Library). In the preliminary design, the maximum TSRLLMA was {approx}11 and the blanket thermal power was {approx}1000 MW when the effective multiplication factor was 0.98. The results showed that good performance of transmutation could be achieved based on the subcritical reactor loaded with uranium-free fuel. (authors)

  18. The physics experimental study for in-hospital neutron irradiator

    SciTech Connect

    Li Yiguo; Xia Pu; Zou Shuyun; Zhang Yongbao; Zheng Iv; Zheng Wuqing; Shi Yongqian; Gao Jijin; Zhou Yongmao [China Institute of Atomic Energy, Beijing 102413, P.O. Box 275-75 (China)

    2008-07-15

    MNSRs (Miniature Neutron Source Reactor) are low power research reactors designed and manufactured by China Institute of Atomic Energy (CIAE). MNSRs are mainly used for NAA, training and teaching, testing of nuclear instrumentation. The first MNSR, the prototype MNSR, was put into operation in 1984, later, eight other MNSRs had been built both at home and abroad. For MNSRs, highly enriched uranium (90%) is used as the fuel material. The In-Hospital Neutron Irradiator (IHNI) is designed for Boron Neutron Capture Therapy (BNCT) based on Miniature Neutron Source Reactor(MNSR). On both sides of the reactor core, there are two neutron beams, one is thermal neutron beam, and the other opposite to the thermal beam, is epithermal neutron beam. A small thermal neutron beam is specially designed for the measurement of blood boron concentration by the prompt gamma neutron activation analysis (PGNAA). In this paper, the experimental results of critical mass worth of the top Be reflectors worth of the control rod, neutron flux distribution and other components worth were measured, the experiment was done on the Zero Power Experiment equipment of MNSR. (author)

  19. Neutronics and Thermal Hydraulics Study for Using a Low-Enriched Uranium Core in the Advanced Test Reactor -- 2008 Final Report

    SciTech Connect

    G. S. Chang; M. A. Lillo; R. G. Ambrosek

    2008-06-01

    The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuel cycle burnup comparison analysis. Using the current HEU U 235 enrichment of 93.0 % as a baseline, an analysis was performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff versus effective full power days (EFPDs) between the HEU and the LEU cores. The MCNP ATR 1/8th core model was used to optimize the U 235 loading in the LEU core, such that the differences in K-eff and heat flux profiles between the HEU and LEU cores were minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the ATR reference HEU case study. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm (20 mil). In this work, the proposed LEU (U-10Mo) core conversion case with nominal fuel meat thickness of 0.330 mm (13 mil) and U-235 enrichment of 19.7 wt% is used to optimize the radial heat flux profile by varying the fuel meat thickness from 0.191 mm (7.0 mil) to 0.330 mm (13.0 mil) at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). A 0.8g of Boron-10, a burnable absorber, was added in the inner and outer plates to reduce the initial excess reactivity, and the peak to average ratio of the inner/outer heat flux more effectively. Because the B-10 (n,a) reaction will produce Helium-4 (He-4), which might degrade the LEU foil type fuel performance, an alternative absorber option is proposed. The proposed LEU case study will have 6.918 g of Cadmium (Cd) mixed with the LEU at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19) as a burnable absorber to achieve peak to average ratios similar to those for the ATR reference HEU case study.

  20. The Landscape of the Neutrino Mechanism of Core-collapse Supernovae: Neutron Star and Black Hole Mass Functions, Explosion Energies, and Nickel Yields

    NASA Astrophysics Data System (ADS)

    Pejcha, Ond?ej; Thompson, Todd A.

    2015-03-01

    If the neutrino luminosity from the proto-neutron star formed during a massive star core collapse exceeds a critical threshold, a supernova (SN) results. Using spherical quasi-static evolutionary sequences for hundreds of progenitors over a range of metallicities, we study how the explosion threshold maps onto observables, including the fraction of successful explosions, the neutron star (NS) and black hole (BH) mass functions, the explosion energies (E SN) and nickel yields (M Ni), and their mutual correlations. Successful explosions are intertwined with failures in a complex pattern that is not simply related to initial progenitor mass or compactness. We predict that progenitors with initial masses of 15 ± 1, 19 ± 1, and ~21-26 M ? are most likely to form BHs, that the BH formation probability is non-zero at solar-metallicity and increases significantly at low metallicity, and that low luminosity, low Ni-yield SNe come from progenitors close to success/failure interfaces. We qualitatively reproduce the observed E SN-M Ni correlation, we predict a correlation between the mean and width of the NS mass and E SN distributions, and that the means of the NS and BH mass distributions are correlated. We show that the observed mean NS mass of ~= 1.33 M ? implies that the successful explosion fraction is higher than 0.35. Overall, we show that the neutrino mechanism can in principle explain the observed properties of SNe and their compact objects. We argue that the rugged landscape of progenitors and outcomes mandates that SN theory should focus on reproducing the wide ranging distributions of observed SN properties.

  1. The Neutron Imaging Diagnostic at NIF

    SciTech Connect

    Merrill, F E; Buckles, R; Clark, D; Danly, C R; Drury, O B; Dzenitis, J M; Fatherly, V E; Fittinghoff, D N; Gallegos, R; Grim, G P; Guler, N; Loomis, E N; Lutz, S; Malone, R M; Martinson, D D; Mares, D; Morley, D J; Morgan, G L; Oertel, J A; Tregillis, I L; Volegov, P L; Weiss, P B; Wilde, C H

    2012-10-01

    A neutron imaging diagnostic has recently been commissioned at the National Ignition Facility (NIF). This new system is an important diagnostic tool for inertial fusion studies at the NIF for measuring the size and shape of the burning DT plasma during the ignition stage of ICF implosions. The imaging technique utilizes a pinhole neutron aperture, placed between the neutron source and a neutron detector. The detection system measures the two dimensional distribution of neutrons passing through the pinhole. This diagnostic has been designed to collect two images at two times. The long flight path for this diagnostic, 28 m, results in a chromatic separation of the neutrons, allowing the independently timed images to measure the source distribution for two neutron energies. Typically the first image measures the distribution of the 14 MeV neutrons and the second image of the 6-12 MeV neutrons. The combination of these two images has provided data on the size and shape of the burning plasma within the compressed capsule, as well as a measure of the quantity and spatial distribution of the cold fuel surrounding this core.

  2. Nuclear waste disposal utilizing a gaseous core reactor

    NASA Technical Reports Server (NTRS)

    Paternoster, R. R.

    1975-01-01

    The feasibility of a gaseous core nuclear reactor designed to produce power to also reduce the national inventories of long-lived reactor waste products through nuclear transmutation was examined. Neutron-induced transmutation of radioactive wastes is shown to be an effective means of shortening the apparent half life.

  3. Vacuum seals design and testing for a linear accelerator of the National Spallation Neutron Source

    SciTech Connect

    Z. Chen; C. Gautier; F. Hemez; N. K. Bultman

    2000-02-01

    Vacuum seals are very important to ensure that the Spallation Neutron Source (SNS) Linac has an optimum vacuum system. The vacuum joints between flanges must have reliable seals to minimize the leak rate and meet vacuum and electrical requirements. In addition, it is desirable to simplify the installation and thereby also simplify the maintenance required. This report summarizes an investigation of the metal vacuum seals that include the metal C-seal, Energized Spring seal, Helcoflex Copper Delta seal, Aluminum Delta seal, delta seal with limiting ring, and the prototype of the copper diamond seals. The report also contains the material certifications, design, finite element analysis, and testing for all of these seals. It is a valuable reference for any vacuum system design. To evaluate the suitability of several types of metal seals for use in the SNS Linac and to determine the torque applied on the bolts, a series of vacuum leak rate tests on the metal seals have been completed at Los Alamos Laboratory. A copper plated flange, using the same type of delta seal that was used for testing with the stainless steel flange, has also been studied and tested. A vacuum seal is desired that requires significantly less loading than a standard ConFlat flange with a copper gasket for the coupling cavity assembly. To save the intersegment space the authors use thinner flanges in the design. The leak rate of the thin ConFlat flange with a copper gasket is a baseline for the vacuum test on all seals and thin flanges. A finite element analysis of a long coupling cavity flange with a copper delta seal has been performed in order to confirm the design of the long coupling cavity flange and the welded area of a cavity body with the flange. This analysis is also necessary to predict a potential deformation of the cavity under the combined force of atmospheric pressure and the seating load of the seal. Modeling of this assembly has been achieved using both HKS/Abaqus and COSMOS/M, which are finite element packages for analysis of coupled, nonlinear problems. From these studies, the appropriate seals that are reliable for SNS long coupling cavities and beamline joints were determined.

  4. Optimization study for an epithermal neutron beam for boron neutron capture therapy at the University of Virginia Research Reactor

    SciTech Connect

    Burns, T.D. Jr.

    1995-05-01

    The non-surgical brain cancer treatment modality, Boron Neutron Capture Therapy (BNCT), requires the use of an epithermal neutron beam. This purpose of this thesis was to design an epithermal neutron beam at the University of Virginia Research Reactor (UVAR) suitable for BNCT applications. A suitable epithermal neutron beam for BNCT must have minimal fast neutron and gamma radiation contamination, and yet retain an appreciable intensity. The low power of the UVAR core makes reaching a balance between beam quality and intensity a very challenging design endeavor. The MCNP monte carlo neutron transport code was used to develop an equivalent core radiation source, and to perform the subsequent neutron transport calculations necessary for beam model analysis and development. The code accuracy was validated by benchmarking output against experimental criticality measurements. An epithermal beam was designed for the UVAR, with performance characteristics comparable to beams at facilities with cores of higher power. The epithermal neutron intensity of this beam is 2.2 {times} 10{sup 8} n/cm{sup 2} {center_dot} s. The fast neutron and gamma radiation KERMA factors are 10 {times} 10{sup {minus}11}cGy{center_dot}cm{sup 2}/n{sub epi} and 20 {times} 10{sup {minus}11} cGy{center_dot}cm{sup 2}/n{sub epi}, respectively, and the current-to-flux ratio is 0.85. This thesis has shown that the UVAR has the capability to provide BNCT treatments, however the performance characteristics of the final beam of this study were limited by the low core power.

  5. Supernova Neutrino Light Curves and Spectra for Various Progenitor Stars: From Core Collapse to Proto-neutron Star Cooling

    NASA Astrophysics Data System (ADS)

    Nakazato, Ken'ichiro; Sumiyoshi, Kohsuke; Suzuki, Hideyuki; Totani, Tomonori; Umeda, Hideyuki; Yamada, Shoichi

    2013-03-01

    We present a new series of supernova neutrino light curves and spectra calculated by numerical simulations for a variety of progenitor stellar masses (13-50 M ?) and metallicities (Z = 0.02 and 0.004), which would be useful for a broad range of supernova neutrino studies, e.g., simulations of future neutrino burst detection by underground detectors or theoretical predictions for the relic supernova neutrino background. To follow the evolution from the onset of collapse to 20 s after the core bounce, we combine the results of neutrino-radiation hydrodynamic simulations for the early phase and quasi-static evolutionary calculations of neutrino diffusion for the late phase, with different values of shock revival time as a parameter that should depend on the still unknown explosion mechanism. We describe the calculation methods and basic results, including the dependence on progenitor models and the shock revival time. The neutrino data are publicly available electronically.

  6. Design of Gas-phase Synthesis of Core-Shell Particles by Computational Fluid – Aerosol Dynamics

    PubMed Central

    Buesser, B.; Pratsinis, S.E.

    2013-01-01

    Core-shell particles preserve the bulk properties (e.g. magnetic, optical) of the core while its surface is modified by a shell material. Continuous aerosol coating of core TiO2 nanoparticles with nanothin silicon dioxide shells by jet injection of hexamethyldisiloxane precursor vapor downstream of titania particle formation is elucidated by combining computational fluid and aerosol dynamics. The effect of inlet coating vapor concentration and mixing intensity on product shell thickness distribution is presented. Rapid mixing of the core aerosol with the shell precursor vapor facilitates efficient synthesis of hermetically coated core-shell nanoparticles. The predicted extent of hermetic coating shells is compared to the measured photocatalytic oxidation of isopropanol by such particles as hermetic SiO2 shells prevent the photocatalytic activity of titania. Finally the performance of a simpler, plug-flow coating model is assessed by comparisons to the present detailed CFD model in terms of coating efficiency and silica average shell thickness and texture. PMID:23729817

  7. Thermal-hydraulic analysis for wire-wrapped PWR cores

    Microsoft Academic Search

    P. Diller; N. Todreas; P. Hejzlar

    2009-01-01

    This work focuses on the steady-state and transient thermal-hydraulic analyses for PWR cores using wire wraps in a hexagonal array with either U (45% w\\/o)–ZrH1.6 (referred to as U–ZrH1.6) or UO2 fuels. Equivalences (thermal-hydraulic and neutronic) were created between grid spacer and wire wrap designs, and were used to apply results calculated for grid spacers to wire wrap designs. Design

  8. Monte Carlo-based treatment planning for boron neutron capture therapy using custom designed models automatically generated from CT data

    Microsoft Academic Search

    R. Zamenhof; E. Redmond; G. Solares; D. Katz; K. Riley; S. Kiger; O. Harling

    1996-01-01

    Purpose: A Monte Carlo-based treatment planning code for boron neutron capture therapy (BNCT), called NCTPLAN, has been developed in support of the NeW England Medical Center-Massachusetts Institute of Technology program in BNCT. This code has been used to plan BNCT irradiations in an ongoing peripheral melanoma BNCT protocol. The concept and design of the code is described and illustrative applications

  9. Optimized design and predicted performance of a deep ocean 50 m piston coring system

    SciTech Connect

    Karnes, C. H.; Burchett, S. N.; Dzwilewski, P. T.

    1980-01-01

    Calculational techniques are described which were developed or adapted for the purpose of analyzing the mechanical response of a proposed piston coring system capable of recovering high quality 50 m long cores. The analysis includes the effects of barrel geometry on the mass required to penetrate 50 m of an assumed sediment, the effects of non-vertical entry and pullout on the stresses within the barrel, and the effects of steel cable or parachute piston restraints on the resulting core sample distortion. The results show that a wall thickness of 50 mm in the upper section is necessary to survive an entry of up to 1.5/sup 0/ from vertical or a recovery angle of up to 5/sup 0/. They also show that a mass of 15,400 kg and a pullout force of 330 kN are required. It is shown that active piston control is necessary to eliminate piston motion during penetration.

  10. Replacing a 252Cf source with a neutron generator in a shuffler - a conceptual design performed with MCNPX

    SciTech Connect

    Schear, Melissa A [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory

    2009-01-01

    The {sup 252}Cf shuffler has been widely used in nuclear safeguards and radioactive waste management to assay fissile isotopes, such as {sup 235}U or {sup 239}Pu, present in a variety of samples, ranging from small cans of uranium waste to metal samples weighing several kilograms. Like other non-destructive assay instruments, the shuffler uses an interrogating neutron source to induce fissions in the sample. Although shufflers with {sup 252}Cf sources have been reliably used for several decades, replacing this isotopic source with a neutron generator presents some distinct advantages. Neutron generators can be run in a continuous or pulsed mode, and may be turned off, eliminating the need for shielding and a shuffling mechanism in the shuffler. There is also essentially no dose to personnel during installation, and no reliance on the availability of {sup 252}Cf. Despite these advantages, the more energetic neutrons emitted from the neutron generator (141 MeV for D-T generators) present some challenges for certain material types. For example when the enrichment of a uranium sample is unknown, the fission of {sup 238}U is generally undesirable. Since measuring uranium is one of the main uses of a shuffler, reducing the delayed neutron contribution from {sup 238}U is desirable. Hence, the shuffler hardware must be modified to accommodate a moderator configuration near the source to tailor the interrogating spectrum in a manner which promotes sub-threshold fissions (below 1 MeV) but avoids the over-moderation of the interrogating neutrons so as to avoid self-shielding. In this study, where there are many material and geometry combinations, the Monte Carlo N-Particle eXtended (MCNPX) transport code was used to model, design, and optimize the moderator configuration within the shuffler geometry. The code is then used to evaluate and compare the assay performances of both the modified shuffler and the current {sup 252}Cf shuffler designs for different test samples. The matrix effect and the non-uniformity of the interrogating flux are investigated and quantified in each case. The modified geometry proposed by this study can serve s a guide in retrofitting shufflers that are already in use.

  11. Design and implementation of a dynamic neutron radiographic imaging system: by John Winston Wright. 

    E-print Network

    Wright, John Winston

    1990-01-01

    . Rudolph, R. G. , Henry, E. B. Jr?"Real-Time Radiographic Imaging for Submerged-Arc Welded Pipe", ASTM STP 716, Ibid. , pp 128-155, 6. Rant, J. , Ilic, R. , Pregl, G. , Leskovar, P. , Znidar, B. , "The Sensitivity of Neutron Radiography for Detection...&Abs Abs=S cat 200 keV Photon o 1 . 01 0 ~ ts 8 t ~ ~ gs~ ty ts ~ ~ ~ ~ SS ~ 20 40 "+B 8 8 B pB ~ ~ is JIP ~ 'l ~ 60 80 100 Atomic Number Figure 2- Neutron and photon attenuation coefficients. As stated earlier, the advantage of neutrons...

  12. Nuclear Data Needs for the Neutronic Design of MYRRHA Fast Spectrum Research Reactor

    NASA Astrophysics Data System (ADS)

    Stankovskiy, A.; Malambu, E.; Van den Eynde, G.; Díez, C. J.

    2014-04-01

    A global sensitivity analysis of effective neutron multiplication factor to the change of nuclear data library has been performed. It revealed that the test version of JEFF-3.2 neutron-induced evaluated data library produces closer results to ENDF/B-VII.1 than JEFF-3.1.2 does. The analysis of contributions of individual evaluations into keff sensitivity resulted in the priority list of nuclides, uncertainties on cross sections and fission neutron multiplicities of which have to be improved by setting up dedicated differential and integral experiments.

  13. Ten Core Principles for Designing Effective Learning Environments: Insights from Brain Research and Pedagogical Theory

    ERIC Educational Resources Information Center

    Boettcher, Judith V.

    2007-01-01

    In this article, Judith V. Boettcher provides ten core learning principles that can guide technology-enhanced teaching as well as more traditional forms of instruction. Drawn from both traditional pedagogical theory as well as current research about how people learn, the ten principles integrate these findings in a helpful set of guidelines that…

  14. A core-theoretic solution for the design of cooperative agreements on transfrontier pollution

    Microsoft Academic Search

    Parkash Chander; Henry Tulkens

    1995-01-01

    For a simple economic model of transfrontier pollution, widely used in theoretical studies of international treaties bearing on joint abatement, we offer in this paper a scheme for sharing national abatement costs through international financial transfers that is inspired by a classical solution concept from the theory of cooperative games—namely, the core of a game. The scheme has the following

  15. A Core-Theoretic Solution for the Design of Cooperative Agreements on Transfrontier Pollution

    Microsoft Academic Search

    Parkash Chander; Henry Tulkens

    1994-01-01

    For a simple economic model of transfrontier pollution, widely used in theoretical studies of international treaties bearing on joint abatement, we offer in this paper a scheme for sharing national abatement costs through international financial transfers that is inspired by a classical solution concept from the theory of cooperative games?namely, the core of a game. The scheme has the following

  16. Hardware-Assisted Reliability Enhancement for Embedded Multi-core Virtualization Design

    Microsoft Academic Search

    Tsung-Han Lin; Yuki Kinebuchi; Alexandre Courbot; Hiromasa Shimada; Takushi Morita; Hitoshi Mitake; Chen-Yi Lee; Tatsuo Nakajima

    2011-01-01

    In this paper, we propose a virtualization ar- chitecture for the multi-core embedded system to provide more system reliability and security while maintaining the same performance without introducing additional special hard- ware supports or having to implement complex protection mechanism in the virtualization layer. Virtualization has been widely used in embedded systems, especially in consumer electronics, albeit itself is not

  17. New multi-core Intel Xeon processors help design energy efficient solution for high performance computing

    Microsoft Academic Search

    Pawel Gepner; David L. Fraser; Michal Filip Kowalik; R. Tylman

    2009-01-01

    The second generation of Intel® Xeon¿ processors based on core microarchitecture and 45 nm process technology bring not only a new level of performance but also significant improvement in power characteristics. Continuous performance improvement and power efficiency are the paradigms for most data centers today and are also the challenges that will not go away anytime soon. The increasing energy

  18. Designing a VH-mode core/L-mode edge discharge

    SciTech Connect

    Staebler, G.M.; Hinton, F.L. [General Atomics, San Diego, CA (United States); Wiley, J.C. [Univ. of Texas, Austin, TX (United States). Fusion Research Center

    1995-12-01

    An operating mode with a very high confinement core like the VH-mode but a very low power flow to the divertor plates and low edge particle confinement like an L-mode would be beneficial. For a large tokamak like the proposed ITER, the power density at the separatrix is not that far above the scaled H-mode power threshold so not much of the power can be radiated inside of the separatrix without causing a return to L-mode. The thicker scrape-off layer of an L-mode increases the radiating volume of the scrape-off layer and helps shield impurities from the core. This is especially important if the first wall is metallic. In this paper an H-mode transport model based on E x B velocity shear suppression of turbulence will be used to show that it is possible to have a strongly radiating mantle near the separatrix, which keeps the edge in L-mode, while having a VH-mode core with a broad region of suppressed turbulence. The existing results of enhanced L-mode confinement during impurity injection on a number of tokamaks will be surveyed. The operating conditions which will most likely result in the further improvement of the core confinement by control of the heating, fueling, and torque profiles will be identified.

  19. Design of a magnetic shielding system for the time of flight enhanced diagnostics neutron spectrometer at Experimental Advanced Superconducting Tokamak

    SciTech Connect

    Cui, Z. Q.; Chen, Z. J.; Xie, X. F.; Peng, X. Y.; Hu, Z. M.; Du, T. F.; Ge, L. J.; Zhang, X.; Yuan, X.; Fan, T. S.; Chen, J. X.; Li, X. Q., E-mail: lixq2002@pku.edu.cn, E-mail: guohuizhang@pku.edu.cn; Zhang, G. H., E-mail: lixq2002@pku.edu.cn, E-mail: guohuizhang@pku.edu.cn [School of Physics, State Key Lab of Nuclear Physics and Technology, Peking University, Beijing 100871 (China); Xia, Z. W. [Southwestern Institute of Physics, Chengdu 610225 (China); Hu, L. Q.; Zhong, G. Q.; Lin, S. Y.; Wan, B. N. [Institute of Plasma Physics, CAS, Hefei 230031 (China)

    2014-11-15

    The novel neutron spectrometer TOFED (Time of Flight Enhanced Diagnostics), comprising 90 individual photomultiplier tubes coupled with 85 plastic scintillation detectors through light guides, has been constructed and installed at Experimental Advanced Superconducting Tokamak. A dedicated magnetic shielding system has been constructed for TOFED, and is designed to guarantee the normal operation of photomultiplier tubes in the stray magnetic field leaking from the tokamak device. Experimental measurements and numerical simulations carried out employing the finite element method are combined to optimize the design of the magnetic shielding system. The system allows detectors to work properly in an external magnetic field of 200 G.

  20. Design and commissioning of a high magnetic field muon spin relaxation spectrometer at the ISIS pulsed neutron and muon source

    SciTech Connect

    Lord, J. S.; McKenzie, I.; Baker, P. J.; Cottrell, S. P.; Giblin, S. R.; Hillier, A. D.; Holsman, B. H.; King, P. J. C.; Nightingale, J. B.; Pratt, F. L.; Rhodes, N. J. [ISIS Facility, STFC Rutherford Appleton Laboratory, Chilton, Oxon OX11 0QX (United Kingdom); Blundell, S. J.; Lancaster, T. [Clarendon Laboratory, Department of Physics, Oxford University, Parks Road, Oxford OX1 3PU (United Kingdom); Good, J.; Mitchell, R.; Owczarkowski, M.; Poli, S. [Cryogenic Limited, 30 Acton Park Industrial Estate, The Vale, Acton, London W3 7QE (United Kingdom); Scheuermann, R. [Laboratory for Muon Spin Spectroscopy, Paul Scherrer Institut, CH-5232 Villigen PSI (Switzerland); Salman, Z. [ISIS Facility, STFC Rutherford Appleton Laboratory, Chilton, Oxon OX11 0QX (United Kingdom); Clarendon Laboratory, Department of Physics, Oxford University, Parks Road, Oxford OX1 3PU (United Kingdom)

    2011-07-15

    The high magnetic field (HiFi) muon instrument at the ISIS pulsed neutron and muon source is a state-of-the-art spectrometer designed to provide applied magnetic fields up to 5 T for muon studies of condensed matter and molecular systems. The spectrometer is optimised for time-differential muon spin relaxation studies at a pulsed muon source. We describe the challenges involved in its design and construction, detailing, in particular, the magnet and detector performance. Commissioning experiments have been conducted and the results are presented to demonstrate the scientific capabilities of the new instrument.

  1. Powder Cores s Molypermalloy

    E-print Network

    Software q Current Transformer Design Software q Inductor Design Software q Mag Amp Design Software POWDER.mag-inc.com PRODUCT LITERATURE AND DESIGN SOFTWARE CD CONTAINS q All Product Literature q Common Mode FIlter Design the Proper Core for Saturating Transformers q TWC-S3 Inverter Transformer Core Design and Material Selection

  2. Journal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 39, No. 11, p. 11691181 (November 2002) Conceptual Design of a Modular Island Core Fast Breeder Reactor "RAPID-M"

    E-print Network

    Laughlin, Robert B.

    Journal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 39, No. 11, p. 1169­1181 (November 2002) Conceptual Design of a Modular Island Core Fast Breeder Reactor "RAPID-M" Mitsuru KAMBE Central Research Institute and accepted September 10, 2002) A metal fueled modular island core sodium cooled fast breeder reactor concept

  3. An Empirical Determination of Tasks Essential to Successful Performance as a Floral Designer. Determination of a Common Core of Basic Skills in Agribusiness and Natural Resources.

    ERIC Educational Resources Information Center

    Miller, Daniel R.; And Others

    To improve vocational educational programs in agriculture, occupational information on a common core of basic skills within the occupational area of the floral designer is presented in the revised task inventory survey. The purpose of the occupational survey was to identify a common core of basic skills which are performed and are essential for…

  4. SCC500: next-generation infrared imaging camera core products with highly flexible architecture for unique camera designs

    NASA Astrophysics Data System (ADS)

    Rumbaugh, Roy N.; Grealish, Kevin; Kacir, Tom; Arsenault, Barry; Murphy, Robert H.; Miller, Scott

    2003-09-01

    A new 4th generation MicroIR architecture is introduced as the latest in the highly successful Standard Camera Core (SCC) series by BAE SYSTEMS to offer an infrared imaging engine with greatly reduced size, weight, power, and cost. The advanced SCC500 architecture provides great flexibility in configuration to include multiple resolutions, an industry standard Real Time Operating System (RTOS) for customer specific software application plug-ins, and a highly modular construction for unique physical and interface options. These microbolometer based camera cores offer outstanding and reliable performance over an extended operating temperature range to meet the demanding requirements of real-world environments. A highly integrated lens and shutter is included in the new SCC500 product enabling easy, drop-in camera designs for quick time-to-market product introductions.

  5. Nuclear data uncertainty propagation for neutronic key parameters of CEA's SFR V2B and CFV sodium fast reactor designs

    SciTech Connect

    Archier, P.; Buiron, L.; De Saint Jean, C.; Dos Santos, N. [CEA, DEN, DER/SPRC, F-13108 Saint Paul-lez-Durance (France)

    2012-07-01

    This paper presents a nuclear data uncertainty propagation analysis for two CEA's Sodium-cooled Fast Reactor designs: the SFR V2B and CFV cores. The nuclear data covariance matrices are provided by the DER/SPRC/LEPh's nuclear data team (see companion paper) for several major isotopes. From the current status of this analysis, improvements on certain nuclear data reactions are highlighted as well as the need for new specific integral experiments in order to meet the technological breakthroughs proposed by the CFV core. (authors)

  6. Final Stage in the Design of a Boron Neutron Capture Therapy facility at CEADEN, Cuba

    SciTech Connect

    Cabal, F. Padilla [Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC) (Cuba); Martin, G. [Centro de Aplicaciones Tecnologicas y Desarrollo Nuclear (CEADEN) (Cuba)

    2008-08-11

    A neutron beam simulation study is carried out to determine the most suitable neutron energy for treatment of shallow and deep-seated brain tumors in the context of Boron Neutron Capture Therapy (BNCT). Two figures-of-merit, the therapeutic gain and the neutron fluence are utilized as beam assessment parameters. An irradiation cavity is used instead of a parallel beam port for the therapy. Calculations are performed using the MCNP5 code. After the optimization of our beam-shaper a study of the dose distribution in the head, neck, tyroids, lungs and upper and middle spine had been made. The therapeutic gain is increased while the current required for one hour treatment is decreased in comparison with the trading prototypes of NG used for BNCT.

  7. NEUTRONIC AND THERMAL HYDRAULIC DESIGNS OF ANNULAR FUEL FOR HIGH POWER DENSITY BWRS

    E-print Network

    Morra, P.

    As a promising new fuel for high power density light water reactors, the feasibility of using annular fuel for BWR services is explored from both thermal hydraulic and neutronic points of view. Keeping the bundle size ...

  8. Design study on an accelerator-based facility for BNCT and low energy neutron source

    Microsoft Academic Search

    Makoto Sasaki; Toshiyuki Yamanaka

    2000-01-01

    We have challenged to reduce an accelerator beam power for an accelerator-based BNCT facility. The required neutron source strength at the target has been estimated so as to make the epithermal neutron flux in the patient irradiation field exceed 1.7 × 109 n\\/cm2s. The energy of the incident proton and the arrangement of the moderator assemblies are optimized. The beam

  9. ASPUN: design for an Argonne super-intense pulsed neutron source

    Microsoft Academic Search

    T. K. Khoe; R. L. Kustom

    1983-01-01

    Argonne pioneered the pulsed spallation neutron source with the ZING-P and IPNS-I concepts. IPNS-I is now a reliable and actively used source for pulsed spallation neutrons. The accelerator is a 500-MeV, 8 to 9 ..mu..a, 30-Hz rapid-cycling proton synchrotron. Other proton spallation sources are now in operation or in construction. These include KENS-I at the National Laboratory for High Energy

  10. Space Station Furnace Facility Core. Requirements definition and conceptual design study. Volume 2: Technical report. Appendix 6: Technical summary reports

    NASA Technical Reports Server (NTRS)

    1992-01-01

    The Space Station Furnace Facility (SSFF) is a modular facility for materials research in the microgravity environment of the Space Station Freedom (SSF). The SSFF is designed for crystal growth and solidification research in the fields of electronic and photonic materials, metals and alloys, and glasses and ceramics and will allow for experimental determination of the role of gravitational forces in the solidification process. The facility will provide a capability for basic scientific research and will evaluate the commercial viability of low-gravity processing of selected technologically important materials. The facility is designed to support a complement of furnace modules as outlined in the Science Capabilities Requirements Document (SCRD). The SSFF is a three rack facility that provides the functions, interfaces, and equipment necessary for the processing of the furnaces and consists of two main parts: the SSFF Core Rack and the two Experiment Racks. The facility is designed to accommodate two experimenter-provided furnace modules housed within the two experiment racks, and is designed to operate these two furnace modules simultaneously. The SCRD specifies a wide range of furnace requirements and serves as the basis for the SSFF conceptual design. SSFF will support automated processing during the man-tended operations and is also designed for crew interface during the permanently manned configuration. The facility is modular in design and facilitates changes as required, so the SSFF is adept to modifications, maintenance, reconfiguration, and technology evolution.

  11. IEEE TRANSACTIONS ON CIRCUITS AND SYSTEMS FOR VIDEO TECHNOLOGY, VOL. 10, NO. 3, APRIL 2000 439 A Simple Processor Core Design for DCT/IDCT

    E-print Network

    Ha, Dong S.

    . This problem is not affective in this AU-based design due to the software-like controller design. Furthermore A Simple Processor Core Design for DCT/IDCT Tian-Sheuan Chang, Student Member, IEEE, Chin-Sheng Kung that features the simplest hardware and is suitable for discrete cosine transform/indiscrete cosine transform

  12. Neutronic and severe safety aspects of 1500 MWth lead and sodium cooled fast reactors

    SciTech Connect

    Tucek, Kamil; Carlsson, Johan; Vidovic, Dragan; Wider, Hartmut [Joint Research Center of the European Commission Institute for Energy, Postbus 2, 1755 ZG Petten (Netherlands)

    2007-07-01

    In this paper, neutronics and severe safety characteristics of Lead-cooled Fast Reactor (LFR) and Sodium-cooled Fast Reactor (SFR) cores concurrently breeding plutonium and burning minor actinides (MAs) are investigated. For LFR, two core variants were modeled: with active core part 90 cm and 120 cm tall. Monte Carlo code was used for neutronics and European Accident Code EAC-2 for severe safety studies. It is shown that both 1500 MW{sub th} LFR and SFR start-up cores can transmute on average 70 kg of MAs annually in the homogeneous mode. In this case, 5% of MAs were admixed to the core fuel. More than 110 kg of MAs can be burned per year when 10% of MAs are additionally added to axial and radial blankets. LFR core designs show advantages over SFR cores regarding severe safety behavior due to higher thermal inertia, better natural circulation behavior and the higher boiling point of lead. (authors)

  13. Vibration Characteristics Determined for Stainless Steel Sandwich Panels With a Metal Foam Core for Lightweight Fan Blade Design

    NASA Technical Reports Server (NTRS)

    Ghosn, Louis J.; Min, James B.; Raj, Sai V.; Lerch, Bradley A.; Holland, Frederic A., Jr.

    2004-01-01

    The goal of this project at the NASA Glenn Research Center is to provide fan materials that are safer, weigh less, and cost less than the currently used titanium alloy or polymer matrix composite fans. The proposed material system is a sandwich fan construction made up of thin solid face sheets and a lightweight metal foam core. The stiffness of the sandwich structure is increased by separating the two face sheets by the foam layer. The resulting structure has a high stiffness and lighter weight in comparison to the solid facesheet material alone. The face sheets carry the applied in-plane and bending loads (ref. 1). The metal foam core must resist the transverse shear and transverse normal loads, as well as keep the facings supported and working as a single unit. Metal foams have ranges of mechanical properties, such as light weight, impact resistance, and vibration suppression (ref. 2), which makes them more suitable for use in lightweight fan structures. Metal foams have been available for decades (refs. 3 and 4), but the difficulties in the original processes and high costs have prevented their widespread use. However, advances in production techniques and cost reduction have created a new interest in this class of materials (ref. 5). The material chosen for the face sheet and the metal foam for this study was the aerospace-grade stainless steel 17-4PH. This steel was chosen because of its attractive mechanical properties and the ease with which it can be made through the powder metallurgy process (ref. 6). The advantages of a metal foam core, in comparison to a typical honeycomb core, are material isotropy and the ease of forming complex geometries, such as fan blades. A section of a 17-4PH sandwich structure is shown in the following photograph. Part of process of designing any blade is to determine the natural frequencies of the particular blade shape. A designer needs to predict the resonance frequencies of a new blade design to properly identify a useful operating range. Operating a blade at or near the resonance frequencies leads to high-cycle fatigue, which ultimately limits the blade's durability and life. So the aim of this study is to determine the variation of the resonance frequencies for an idealized sandwich blade as a function of its face-sheet thickness, core thickness, and foam density. The finite element method is used to determine the natural frequencies for an idealized rectangular sandwich blade. The proven Lanczos method (ref. 7) is used in the study to extract the natural frequency.

  14. Laser inertial fusion-based energy: Neutronic design aspects of a hybrid fusion-fission nuclear energy system

    NASA Astrophysics Data System (ADS)

    Kramer, Kevin James

    This study investigates the neutronics design aspects of a hybrid fusion-fission energy system called the Laser Fusion-Fission Hybrid (LFFH). A LFFH combines current Laser Inertial Confinement fusion technology with that of advanced fission reactor technology to produce a system that eliminates many of the negative aspects of pure fusion or pure fission systems. When examining the LFFH energy mission, a significant portion of the United States and world energy production could be supplied by LFFH plants. The LFFH engine described utilizes a central fusion chamber surrounded by multiple layers of multiplying and moderating media. These layers, or blankets, include coolant plenums, a beryllium (Be) multiplier layer, a fertile fission blanket and a graphite-pebble reflector. Each layer is separated by perforated oxide dispersion strengthened (ODS) ferritic steel walls. The central fusion chamber is surrounded by an ODS ferritic steel first wall. The first wall is coated with 250-500 mum of tungsten to mitigate x-ray damage. The first wall is cooled by Li17Pb83 eutectic, chosen for its neutron multiplication and good heat transfer properties. The Li17Pb 83 flows in a jacket around the first wall to an extraction plenum. The main coolant injection plenum is immediately behind the Li17Pb83, separated from the Li17Pb83 by a solid ODS wall. This main system coolant is the molten salt flibe (2LiF-BeF2), chosen for beneficial neutronics and heat transfer properties. The use of flibe enables both fusion fuel production (tritium) and neutron moderation and multiplication for the fission blanket. A Be pebble (1 cm diameter) multiplier layer surrounds the coolant injection plenum and the coolant flows radially through perforated walls across the bed. Outside the Be layer, a fission fuel layer comprised of depleted uranium contained in Tristructural-isotropic (TRISO) fuel particles having a packing fraction of 20% in 2 cm diameter fuel pebbles. The fission blanket is cooled by the same radial flibe flow that travels through perforated ODS walls to the reflector blanket. This reflector blanket is 75 cm thick comprised of 2 cm diameter graphite pebbles cooled by flibe. The flibe extraction plenum surrounds the reflector bed. Detailed neutronics designs studies are performed to arrive at the described design. The LFFH engine thermal power is controlled using a technique of adjusting the 6Li/7Li enrichment in the primary and secondary coolants. The enrichment adjusts system thermal power in the design by increasing tritium production while reducing fission. To perform the simulations and design of the LFFH engine, a new software program named LFFH Nuclear Control (LNC) was developed in C++ to extend the functionality of existing neutron transport and depletion software programs. Neutron transport calculations are performed with MCNP5. Depletion calculations are performed using Monteburns 2.0, which utilizes ORIGEN 2.0 and MCNP5 to perform a burnup calculation. LNC supports many design parameters and is capable of performing a full 3D system simulation from initial startup to full burnup. It is able to iteratively search for coolant 6Li enrichments and resulting material compositions that meet user defined performance criteria. LNC is utilized throughout this study for time dependent simulation of the LFFH engine. Two additional methods were developed to improve the computation efficiency of LNC calculations. These methods, termed adaptive time stepping and adaptive mesh refinement were incorporated into a separate stand alone C++ library name the Adaptive Burnup Library (ABL). The ABL allows for other client codes to call and utilize its functionality. Adaptive time stepping is useful for automatically maximizing the size of the depletion time step while maintaining a desired level of accuracy. Adaptive meshing allows for analysis of fixed fuel configurations that would normally require a computationally burdensome number of depletion zones. Alternatively, Adaptive Mesh Refinement (AMR) adjusts the depletion zone size according to the vari

  15. Thermal hydraulic method for whole core design analysis of an HTGR

    SciTech Connect

    Huning, A. J.; Garimella, S. [George W. Woodruff School of Mechanical Engineering, Georgia Institute of Technology, Atlanta, GA (United States)

    2013-07-01

    A new thermal hydraulic method and initial results are presented for core-wide steady state analysis of prismatic High Temperature Gas-Cooled Reactors (HTGR). The method allows for the complete solution of temperature and coolant mass flow distribution by solving quasi-steady energy balances for the discretized core. Assembly blocks are discretized into unit cells for which the average temperature of each unit cell is determined. Convective heat removal is coupled to the unit cell energy balances by a 1-D axial flow model. The flow model uses established correlations for friction factor and Nusselt number. Bypass flow is explicitly calculated by using an initial guess for mass flow distribution and determining the exit pressure of each flow channel. The mass flow distribution is updated until a uniform core exit pressure condition is reached. Results are obtained for the MHTGR-350 with emphasis on the change in thermal hydraulic parameters due to various steady state power profiles and bypass gap widths. Steady state temperature distribution and its variations are discussed. (authors)

  16. Responses to solar cosmic rays of neutron monitors of a various design

    NASA Astrophysics Data System (ADS)

    Vashenyuk, E. V.; Balabin, Yu. V.; Stoker, P. H.

    The modeled and observed responses of neutron monitors of two various types: the standard 3-NM-64 and a leadless neutron moderated detector 4NMD at the SANAE South African Antarctic station during a number of large GLE events were compared to precise the specific yield of the NMD at low rigidity range. The parameters of primary relativistic solar protons outside magnetosphere: rigidity spectrum, anisotropy direction and pitch angle distribution were determined on data of the worldwide NM-64 neutron monitor network by modeling technique. The modeling included: definition of asymptotic viewing cones of the neutron monitor (NM) stations under study by the particle trajectory computations in a model magnetosphere [Tsyganenko, N.A. A model of the near magnetosphere with a down-dusk asymmetry: 1. Mathematical structure. Geophys. Res. 107(A8) 1176, doi: 10.101029/2001JA000219, 2002a; Tsyganenko, N.A. A model of the near magnetosphere with a down-dusk asymmetry: 2. Parameterization and fitting to observations. J. Geophys. Res. 107(A8) 1179, doi: 10.1029/2001JA000220, 2002b.]; calculation of the NM responses at variable primary solar proton flux parameters; determination of primary solar proton parameters outside the magnetosphere by a least square procedure at comparison of computed NM responses with observations. Then the response of both neutron monitors NM-64 and leadless NMD was calculated using the specific yield functions obtained earlier in the latitude and high-altitude survey of both instruments [Stoker, P.H. Spectra of solar proton ground level events using neutron monitor and neutron moderated detector recordings. in: Proc. 19th ICRC La Jolla, vol. 4, pp. 114-117, 1985; Stoker, P.H. Relativistic solar proton events, Space Sci. Rev. 73, 327-385, 1994.]. By fitting modeled responses to observations in a number of large GLEs the specific yield function for the NMD detector was adjusted so that it precisely described the response to solar cosmic rays.

  17. Design and synthesis of novel ROR inverse agonists with a dibenzosilole scaffold as a hydrophobic core structure.

    PubMed

    Toyama, Hirozumi; Nakamura, Masaharu; Hashimoto, Yuichi; Fujii, Shinya

    2015-07-01

    Molecular structure calculations indicated that the dibenzosilole skeleton could be well superposed on phenanthridinone, which is a structural component of ligands of retinoic acid receptor-related orphan receptors (RORs). Therefore, we designed, synthesized and biologically evaluated a series of novel ROR ligands based on the dibenzosilole scaffold as a hydrophobic core structure. Dibenzosilole derivatives bearing a hexafluoro-2-hydroxypropyl group on the benzene ring exhibited significant ROR-inhibitory activity, comparable to that of the lead phenanthridinone derivative 5. Our results indicate that the dibenzosilole skeleton would be a useful scaffold for developing novel biologically active compounds, and that cis-amide structure can be replaced by an alkylsilyl functionality. PMID:26014484

  18. The ITER Radial Neutron Camera Detection System

    SciTech Connect

    Marocco, D.; Belli, F.; Esposito, B.; Petrizzi, L.; Riva, M. [Associazione ENEA-Euratom sulla Fusione, C.R. Frascati, C.P. 65, I-00044 Frascati (Roma) (Italy); Bonheure, G. [Laboratory for Plasma Physics, Association 'Euratom-Belgian State', Royal Military Academy, Avenue dela Renaissance, 30, B-1000 Brussels (Belgium); Kaschuck, Y. [TRINITI, Troitsk 142190, Moscow Region (Russian Federation)

    2008-03-12

    A multichannel neutron detection system (Radial Neutron Camera, RNC) will be installed on the ITER equatorial port plug 1 for total neutron source strength, neutron emissivity/ion temperature profiles and n{sub t}/n{sub d} ratio measurements [1]. The system is composed by two fan shaped collimating structures: an ex-vessel structure, looking at the plasma core, containing tree sets of 12 collimators (each set lying on a different toroidal plane), and an in-vessel structure, containing 9 collimators, for plasma edge coverage. The RNC detecting system will work in a harsh environment (neutron fiux up to 10{sup 8}-10{sup 9} n/cm{sup 2} s, magnetic field >0.5 T or in-vessel detectors), should provide both counting and spectrometric information and should be flexible enough to cover the high neutron flux dynamic range expected during the different ITER operation phases. ENEA has been involved in several activities related to RNC design and optimization [2,3]. In the present paper the up-to-date design and the neutron emissivity reconstruction capabilities of the RNC will be described. Different options for detectors suitable for spectrometry and counting (e.g. scintillators and diamonds) focusing on the implications in terms of overall RNC performance will be discussed. The increase of the RNC capabilities offered by the use of new digital data acquisition systems will be also addressed.

  19. GPU Based General-Purpose Parallel computing to Solve Nuclear Reactor In-Core fuel Management Design and Operation Problem

    SciTech Connect

    Prayudhatama, D.; Waris, A.; Kurniasih, N.; Kurniadi, R. [Bosscha Laboratory, Department of Physics, Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia)

    2010-06-22

    In-core fuel management study is a crucial activity in nuclear power plant design and operation. Its common problem is to find an optimum arrangement of fuel assemblies inside the reactor core. Main objective for this activity is to reduce the cost of generating electricity, which can be done by altering several physical properties of the nuclear reactor without violating any of the constraints imposed by operational and safety considerations. This research try to address the problem of nuclear fuel arrangement problem, which is, leads to the multi-objective optimization problem. However, the calculation of the reactor core physical properties itself is a heavy computation, which became obstacle in solving the optimization problem by using genetic algorithm optimization.This research tends to address that problem by using the emerging General Purpose Computation on Graphics Processing Units (GPGPU) techniques implemented by C language for CUDA (Compute Unified Device Architecture) parallel programming. By using this parallel programming technique, we develop parallelized nuclear reactor fitness calculation, which is involving numerical finite difference computation. This paper describes current prototype of the parallel algorithm code we have developed on CUDA, that performs one hundreds finite difference calculation for nuclear reactor fitness evaluation in parallel by using GPU G9 Hardware Series developed by NVIDIA.

  20. Mixed-clock issue queue design for energy aware, high-performance cores

    Microsoft Academic Search

    Venkata Syam P. Rapaka; Emil Talpes; Diana Marculescu

    2004-01-01

    Globally-Asynchronous, Locally-Synchronous (GALS) design style has started to gain interest recently as a possible solution to the increased design complexity, power and thermal costs, as well as an enabler for allowing fine grain speed and voltage management. Due to its inherent complexity, a possible driver application for such a design style is the case of superscalar, out-of-order processors. This paper