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Sample records for rod bundle nuclear

  1. Large-scale Flow Pulsation in Tight Square Arrayed Rod Bundles of Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Hwan; Kim, Kyung Min; Cho, Hyung Hee [Yonsei University, Seoul (Korea, Republic of); Shin, Chang Hwan; In, Wang Kee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    As a major component of modern nuclear reactor, the nuclear fuel rod bundles with liquid coolant have been studied by a lot of researchers to understand the flow structure between the fuel rods. Recently, rod arrays with much small pitch-to-diameter ratio have been being tried to increase performance of the nuclear reactor. The liquid coolant flowing axially through these small spaces between the rods is known to show some peculiar phenomena including large-scale, quasi-periodic flow pulsation. These flow pulsation phenomena dominate mixing process in the subchannels. Thus, precise understating of the flow structure is essential to predict thermal-hydraulic phenomena in nuclear rod bundles. In this present paper, the turbulent flow in tight square arrayed rod bundles is investigated with Hot-wire anemometry. Then, the measured velocity data are analyzed by using Fast Fourier Transform analysis to find characteristic frequency of the pulsation

  2. CFD Validation Benchmark Dataset for Natural Convection in Nuclear Fuel Rod Bundles

    Science.gov (United States)

    Smith, Barton; Jones, Kyle

    2016-11-01

    The present study provide CFD validation benchmark data for coupled fluid flow/convection heat transfer on the exterior of heated rods arranged in a 2 × 2 array. The rod model incorporates grids with swirling veins to resemble a nuclear fuel bundle. The four heated aluminum rods are suspended in an open-circuit wind tunnel. Boundary conditions (BCs) are measured and uncertainties calculated to provide all quantities necessary to successfully conduct a CFD validation exercise. System response quantities (SRQs) are measured for comparing the simulation output to the experiment. Stereoscopic Particle Image Velocimetry (SPIV) is used to non-intrusively measure 3-component velocity fields. A through-plane measurement is used for the inflow while laser sheet planes aligned with the flow direction at several downstream locations are used for system response quantities. Two constant heat flux rod surface conditions are presented (400 W/m2 and 700 W/m2) achieving a peak Rayleigh number of 1010 . Uncertainty for all measured variables is reported. The boundary conditions, system response, and all material properties are now available online for download. The U.S. Department of Energy Nuclear Engineering University Program provided the funding for these experiments under Grant 00128493.

  3. Heat Transfer Enhancement By Three-Dimensional Surface Roughness Technique In Nuclear Fuel Rod Bundles

    Science.gov (United States)

    Najeeb, Umair

    This thesis experimentally investigates the enhancement of single-phase heat transfer, frictional loss and pressure drop characteristics in a Single Heater Element Loop Tester (SHELT). The heater element simulates a single fuel rod for Pressurized Nuclear reactor. In this experimental investigation, the effect of the outer surface roughness of a simulated nuclear rod bundle was studied. The outer surface of a simulated fuel rod was created with a three-dimensional (Diamond-shaped blocks) surface roughness. The angle of corrugation for each diamond was 45 degrees. The length of each side of a diamond block is 1 mm. The depth of each diamond block was 0.3 mm. The pitch of the pattern was 1.614 mm. The simulated fuel rod had an outside diameter of 9.5 mm and wall thickness of 1.5 mm and was placed in a test-section made of 38.1 mm inner diameter, wall thickness 6.35 mm aluminum pipe. The Simulated fuel rod was made of Nickel 200 and Inconel 625 materials. The fuel rod was connected to 10 KW DC power supply. The Inconel 625 material of the rod with an electrical resistance of 32.3 kO was used to generate heat inside the test-section. The heat energy dissipated from the Inconel tube due to the flow of electrical current flows into the working fluid across the rod at constant heat flux conditions. The DI water was employed as working fluid for this experimental investigation. The temperature and pressure readings for both smooth and rough regions of the fuel rod were recorded and compared later to find enhancement in heat transfer coefficient and increment in the pressure drops. Tests were conducted for Reynold's Numbers ranging from 10e4 to 10e5. Enhancement in heat transfer coefficient at all Re was recorded. The maximum heat transfer co-efficient enhancement recorded was 86% at Re = 4.18e5. It was also observed that the pressure drop and friction factor increased by 14.7% due to the increased surface roughness.

  4. Turbulet flow in a model nuclear fuel rod bundle containing partial flow blockages

    Energy Technology Data Exchange (ETDEWEB)

    Creer, J.M.; Rowe, D.S.; Bates, J.M.; Sutey, A.M.

    1977-03-01

    Local velocity and turbulence intensity measurements were obtained with a laser Doppler anemometer near flow blockages in an unheated 7 x 7 rod bundle. Sleeve blockages were positioned on the center nine rods to create area reductions of 70 and 90 percent in the center four subchannels of the bundle. Experimental results indicated that severe flow disturbances existed downstream from the blockage clusters and showed that only minor disturbances can be expected upstream from the blockages. Recirculation zones for both 70 and 90 percent blockages were detected downstream from the blockage clusters and persisted for approximately three to five subchannel hydraulic diameters depending on blockage severity. The experimental velocity results obtained with blockage clusters located midway between grid spacers were successfully predicted using the COBRA computer program.

  5. Measurement of liquid film flow on nuclear rod bundle in micro-scale by using very high speed camera system

    Science.gov (United States)

    Pham, Son; Kawara, Zensaku; Yokomine, Takehiko; Kunugi, Tomoaki

    2012-11-01

    Playing important roles in the mass and heat transfer as well as the safety of boiling water reactor, the liquid film flow on nuclear fuel rods has been studied by different measurement techniques such as ultrasonic transmission, conductivity probe, etc. Obtained experimental data of this annular two-phase flow, however, are still not enough to construct the physical model for critical heat flux analysis especially at the micro-scale. Remain problems are mainly caused by complicated geometry of fuel rod bundles, high velocity and very unstable interface behavior of liquid and gas flow. To get over these difficulties, a new approach using a very high speed digital camera system has been introduced in this work. The test section simulating a 3×3 rectangular rod bundle was made of acrylic to allow a full optical observation of the camera. Image data were taken through Cassegrain optical system to maintain the spatiotemporal resolution up to 7 μm and 20 μs. The results included not only the real-time visual information of flow patterns, but also the quantitative data such as liquid film thickness, the droplets' size and speed distributions, and the tilt angle of wavy surfaces. These databases could contribute to the development of a new model for the annular two-phase flow. Partly supported by the Global Center of Excellence (G-COE) program (J-051) of MEXT, Japan.

  6. Hydrodynamic behavior of a bare rod bundle. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Bartzis, J.G.; Todreas, N.E.

    1977-06-01

    The temperature distribution within the rod bundle of a nuclear reactor is of major importance in nuclear reactor design. However temperature information presupposes knowledge of the hydrodynamic behavior of the coolant which is the most difficult part of the problem due to complexity of the turbulence phenomena. In the present work a 2-equation turbulence model--a strong candidate for analyzing actual three dimensional turbulent flows--has been used to predict fully developed flow of infinite bare rod bundle of various aspect ratios (P/D). The model has been modified to take into account anisotropic effects of eddy viscosity. Secondary flow calculations have been also performed although the model seems to be too rough to predict the secondary flow correctly. Heat transfer calculations have been performed to confirm the importance of anisotropic viscosity in temperature predictions. All numerical calculations for flow and heat have been performed by two computer codes based on the TEACH code. Experimental measurements of the distribution of axial velocity, turbulent axial velocity, turbulent kinetic energy and radial Reynolds stresses were performed in the developing and fully developed regions. A 2-channel Laser Doppler Anemometer working on the Reference mode with forward scattering was used to perform the measurements in a simulated interior subchannel of a triangular rod array with P/D = 1.124. Comparisons between the analytical results and the results of this experiment as well as other experimental data in rod bundle array available in literature are presented. The predictions are in good agreement with the results for the high Reynolds numbers.

  7. COBRA-IV-I: an interim version of COBRA for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores

    Energy Technology Data Exchange (ETDEWEB)

    Wheeler, C.L.; Stewart, C.W.; Cena, R.J.; Rowe, D.S.; Sutey, A.M.

    1976-03-01

    The COBRA-IV-I computer code uses the subchannel analysis approach to determine the enthalpy and flow distribution in rod bundles for both steady-state and transient conditions. The steady-state and transient solution schemes used in COBRA-IIIC are still available in COBRA-IV-I as the implicit solution scheme option. In addition to these techniques, a new explicit solution scheme is now available which allows the calculation of severe transients involving flow reversals, recirculations, expulsion and reentry flows, with a pressure or flow boundary condition specified. Significant storage compaction and reduced running times have been achieved to allow the calculation of problems involving hundreds of subchannels.

  8. Thermal hydraulics of rod bundles: The effect of eccentricity

    Energy Technology Data Exchange (ETDEWEB)

    Chauhan, Amit K., E-mail: amit_fmlab@yahoo.co.in [Fluid Mechanics Laboratory, Department of Applied Mechanics, Indian Institute of Technology Madras, Chennai 600036 (India); Prasad, B.V.S.S.S., E-mail: prasad@iitm.ac.in [Thermal Turbomachines Laboratory, Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Patnaik, B.S.V., E-mail: bsvp@iitm.ac.in [Fluid Mechanics Laboratory, Department of Applied Mechanics, Indian Institute of Technology Madras, Chennai 600036 (India)

    2013-10-15

    Highlights: • Present CFD investigation explores, whole bundle eccentricity for the first time. • Fluid flow and thermal characteristics in various subchannels are analyzed. • Mass flux distribution is particularly analyzed to study eccentricity effect. • Higher eccentricity resulted in a shoot up in rod surface temperature distribution. • Both tangential and radial flow in rod bundles has resulted due to eccentricity. -- Abstract: The effect of eccentricity on the fluid flow and heat transfer through a 19-rod bundle is numerically carried out. When the whole bundle shifts downwards with respect to the outer (pressure) tube, flow redistribution happens. This in turn is responsible for changes in mass flux, pressure and differential flow development in various subchannels. The heat flux imposed on the surface of the fuel rods and the mass flux through the subchannels determines the coolant outlet temperatures. The simulations are performed for a coolant flow Reynolds number of 4 × 10{sup 5}. For an eccentricity value of 0.7, the mass flux in the bottom most subchannel (l) was found to decrease by 10%, while the surface temperature of the fuel rod in the vicinity of this subchannel increased by 250% at the outlet section. Parameters of engineering interest including skin friction coefficient, Nusselt number, etc., have been systematically explored to study the effect of eccentricity on the rod bundle.

  9. Acoustic loading effects on oscillating rod bundles

    Energy Technology Data Exchange (ETDEWEB)

    Lin, W.H.

    1980-01-01

    An analytical study of the interaction between an infinite acoustic medium and a cluster of circular rods is described. The acoustic field due to oscillating rods and the acoustic loading on the rods are first solved in a closed form. The acoustic loading is then used as a forcing function for rod responses, and the acousto-elastic couplings are solved simultaneously. Numerical examples are presented for several cases to illustrate the effects of various system parameters on the acoustic reaction force coefficients. The effect of the acoustic loading on the coupled eigenfrequencies are discussed.

  10. Studying the vibration and random hydrodynamic loads on the fuel rods bundles in the fuel assemblies of the reactor installations used at nuclear power stations equipped with VVER reactors

    Science.gov (United States)

    Solonin, V. I.; Perevezentsev, V. V.

    2012-05-01

    Random hydrodynamic loads causing vibration of fuel rod bundles in a turbulent flow of coolant are obtained from the results of pressure pulsation measurements carried out over the perimeter of the external row of fuel rods in the bundle of a full-scale mockup of a fuel assembly used in a second-generation VVER-440 reactor. It is shown that the turbulent flow structure is a factor determining the parameters of random hydrodynamic loads and the vibration of fuel rod bundles excited by these loads. The results from a calculation of random hydrodynamic loads are used for estimating the vibration levels of fuel rod bundles used in prospective designs of fuel assemblies for VVER reactors.

  11. COBRA-IV PC: A personal computer version of COBRA-IV-I for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores

    Energy Technology Data Exchange (ETDEWEB)

    Webb, B.J.

    1988-01-01

    COBRA-IV PC is a modified version of COBRA-IV-I, adapted for use with most IBM PC and PC-compatible desktop computers. Like COBRA-IV-I, COBRA-IV PC uses the subchannel analysis approach to determine the enthalpy and flow distribution in rod bundles for both steady-state and transient conditions. The steady-state and transient solution schemes used in COBRA-IIIC are still available in COBRA-IV PC as the implicit solution scheme option. An explicit solution scheme is also available, allowing the calculation of severe transients involving flow reversals, recirculations, expulsions, and reentry flows, with a pressure or flow boundary condition specified. In addition, several modifications have been incorporated into COBRA-IV PC to allow the code to run on the PC. These include a reduction in the array dimensions, the removal of the dump and restart options, and the inclusion of several code modifications by Oregon State University, most notably, a critical heat flux correlation for boiling water reactor fuel and a new solution scheme for cross-flow distribution calculations. 7 refs., 8 figs., 1 tab.

  12. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report, September 1, 1980-November 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    Todreas, N.E.; Golay, M.W.; Wolf, L.

    1981-02-01

    Four tasks are reported: bundle geometry (wrapped and bare rods), subchannel geometry (bare rods), subchannel geometry (bare rods), LMFBR outlet plenum flow mixing, and theoretical determination of local temperature fields in LMFBR fuel rod bundles. (DLC)

  13. Analysis of Subchannel and Rod Bundle PSBT Experiments with CATHARE 3

    Directory of Open Access Journals (Sweden)

    M. Valette

    2012-01-01

    Full Text Available This paper presents the assessment of CATHARE 3 against PWR subchannel and rod bundle tests of the PSBT benchmark. Noticeable measurements were the following: void fraction in single subchannel and rod bundle, multiple liquid temperatures at subchannel exit in rod bundle, and DNB power and location in rod bundle. All these results were obtained both in steady and transient conditions. Void fraction values are satisfactory predicted by CATHARE 3 in single subchannels with the pipe module. More dispersed predictions of void values are obtained in rod bundles with the CATHARE 3 3D module at subchannel scale. Single-phase liquid mixing tests and DNB tests in rod bundle are also analyzed. After calibrating the mixing in liquid single phase with specific tests, DNB tests using void mixing give mitigated results, perhaps linked to inappropriate use of CHF lookup tables in such rod bundles with many spacers.

  14. CFD analyses of flow structures in air-ingress and rod bundle problems

    Science.gov (United States)

    Wei, Hong-Chan

    Two topics from nuclear engineering field are included in this dissertation. One study is the air-ingress phenomenon during a loss of coolant accident (LOCA) scenario, and the other is a 5-by-5 bundle assembly with a PWR design. The objectives were to investigate the Kelvin-Helmholtz instability of the gravity-driven stratified flows inside a coaxial pipe and the effects caused by two types of spacers at the downstream of the rod bundle. Richardson extrapolation was used for the grid independent study. The simulation results show good agreements with the experiments. Wavelet analysis and Proper Orthogonal Decomposition (POD) were used to study the flow behaviors and flow patterns. For the air-ingress phenomenon, Brunt-Vaisala frequency, or buoyancy frequency, predicts a frequency of 2.34 Hz; this is confirmed by the dominant frequency of 2.4 Hz obtained from the wavelet analysis between times 1.2 s and 1.85 s. For the rod bundle study, the dominant frequency at the center of the subchannel was determined to be 2.4 Hz with a secondary dominant frequency of 4 Hz and a much minor frequency of 6 Hz. Generally, wavelet analysis has much better performance than POD, in the air-ingress phenomenon, for a strongly transient scenario; they are both appropriate for the rod bundle study. Based on this study, when the fluid pair in a real condition is used, the time which air intrudes into the reactor is predictable.

  15. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report, June 1, 1976--August 31, 1976

    Energy Technology Data Exchange (ETDEWEB)

    Todreas, N.E.; Golay, M.W.; Wolf, L.

    1976-01-01

    Progress is summarized in the following areas: wrapped and bare rod bundle geometry, bare rod subchannel geometry, outlet plenum flow mixing, and theoretical determination of local temperature fields in rod bundles. (DG)

  16. A thermal-hydraulic code for transient analysis in a channel with a rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Khodjaev, I.D. [Research & Engineering Centre of Nuclear Plants Safety, Electrogorsk (Russian Federation)

    1995-09-01

    The paper contains the model of transient vapor-liquid flow in a channel with a rod bundle of core of a nuclear power plant. The computer code has been developed to predict dryout and post-dryout heat transfer in rod bundles of nuclear reactor core under loss-of-coolant accidents. Economizer, bubble, dispersed-annular and dispersed regimes are taken into account. The computer code provides a three-field representation of two-phase flow in the dispersed-annular regime. Continuous vapor, continuous liquid film and entrained liquid drops are three fields. For the description of dispersed flow regime two-temperatures and single-velocity model is used. Relative droplet motion is taken into account for the droplet-to-vapor heat transfer. The conservation equations for each of regimes are solved using an effective numerical technique. This technique makes it possible to determine distribution of the parameters of flows along the perimeter of fuel elements. Comparison of the calculated results with the experimental data shows that the computer code adequately describes complex processes in a channel with a rod bundle during accident.

  17. Two-phase flow interfacial structures in a rod bundle geometry

    Science.gov (United States)

    Paranjape, Sidharth S.

    Interfacial structure of air-water two-phase flow in a scaled nuclear reactor rod bundle geometry was studied in this research. Global and local flow regimes were obtained for the rod bundle geometry. Local two-phase flow parameters were measured at various axial locations in order to understand the transport of interfacial structures. A one-dimensional two-group interfacial area transport model was evaluated using the local parameter database. Air-water two-phase flow experiments were performed in an 8 X 8 rod bundle test section to obtain flow regime maps at various axial locations. Area averaged void fraction was measured using parallel plate type impedance void meters. The cumulative probability distribution functions of the signals from the impedance void meters were used along with a self organizing neural network to identify flow regimes. Local flow regime maps revealed the cross-sectional distribution of flow regimes in the bundle. Local parameters that characterize interfacial structure, that is, void fraction alpha, interfacial area concentration, ai, bubble Sauter mean diameter, DSm and bubble velocity, vg were measured using four sensor conductivity probe technique. The local data revealed the distribution of the interfacial structure in the radial direction, as well as its development in the axial direction. In addition to this, the effect of spacer grid on the flow structure at different gas and liquid velocities was revealed by local parameter measurements across the spacer grids. A two-group interfacial area transport equation (IATE) specific to rod bundle geometry was derived. The derivation of two-group IATE required certain assumption on the bubble shapes in the subchannels and the bubbles spanning more than a subchannel. It was found that the geometrical relationship between the volume and the area of a cap bubble distorted by rods was similar to the one derived for a confined channel under a specific geometrical transformation. The one

  18. Test Facility Construction for Flow Visualization on Mixing Flow inside Subchannels of PWR Rod Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seok; Jeon, Byong-Guk; Youn, Young-Jung; Choi, Hae-Seob; Euh, Dong-Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Flow inside rod bundles has a similarity with flow in porous media. To ensure thermal performance of a nuclear reactor, detailed information of the heat transfer and turbulent mixing flow phenomena taking place within the subchannels is required. The subchannel analysis is one of the key thermal-hydraulic calculations in the safety analysis of the nuclear reactor core. At present, subchannel computer codes are employed to simulate fuel elements of nuclear reactor cores and predict the performance of cores under normal operating and hypothetical accident conditions. The ability of these subchannels codes to predict both the flow and enthalpy distribution in fuel assemblies is very important in the design of nuclear reactors. Recently, according to the modern tend of the safety analysis for the nuclear reactor, a new component scale analysis code, named CUPID, and has been developed in KAERI. The CUPID code is based on a two-fluid and three-field model, and both the open and porous media approaches are incorporated. The PRIUS experiment has addressed many key topics related to flow behaviour in a rod bundle. These issues are related to the flow conditions inside a nuclear fuel element during normal operation of the plant or in accident scenarios. From the second half of 2016, flow visualization will be performed by using a high speed camera and image analysis technique, from which detailed information for the two-dimensional movement of single phase flow is quantified.

  19. PWR FLECHT SEASET 163-Rod Bundle Flow Blockage Task data report. NRC/EPRI/Westinghouse report No. 13, August-October 1982

    Energy Technology Data Exchange (ETDEWEB)

    Loftus, M J; Hochreiter, L E; McGuire, M F; Valkovic, M M

    1983-10-01

    This report presents data from the 163-Rod Bundle Blow Blockage Task of the Full-Length Emergency Cooling Heat Transfer Systems Effects and Separate Effects Test Program (FLECHT SEASET). The task consisted of forced and gravity reflooding tests utilizing electrical heater rods with a cosine axial power profile to simulate PWR nuclear core fuel rod arrays. These tests were designed to determine effects of flow blockage and flow bypass on reflooding behavior and to aid in the assessment of computational models in predicting the reflooding behavior of flow blockage in rod bundle arrays.

  20. Uniform versus Nonuniform Axial Power Distribution in Rod Bundle CHF Experiments

    Directory of Open Access Journals (Sweden)

    Baowen Yang

    2014-01-01

    Full Text Available Rod bundle experiments with axially uniform and nonuniform heat fluxes are examined to explore the potential limitations of using uniform rod bundle CHF data for CHF correlation development of light water reactors with nonuniform axial power distribution (APD. The case of upstream burnout is presented as an example of unique phenomena associated with nonuniform rod bundle CHF experiments. It is a result from combined effect of axial nonuniform power shape and different interchannel mixing mechanisms. In addition, several key parameters are investigated with respect to their potential impacts on the thermal-hydraulic behaviors between rod bundles with uniform and nonuniform APDs. This type of misrepresentation cannot be amended or compensated through the use of correction factors due to the lack of critical information in the uniform rod bundle CHF testing as well as the fundamental difference in the underlining driving mechanisms. Other potential issues involved with the use of uniform rod bundle CHF data for nonuniform APD system applications also present strong evidence concerning the limitations and inadequacy of using uniform rod bundle CHF data for the correlation, prediction, and design limit calculation for safety analysis.

  1. Subchannel thermal-hydraulic modeling of an APT tungsten target rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.; Shadday, M.A. Jr.

    1997-09-01

    The planned target for the Accelerator Production of Tritium (APT) neutron source consists of an array of tungsten rod bundles through which D{sub 2}O coolant flows axially. Here, a scoping analysis of flow through an APT target rod bundle was conducted to demonstrate that lateral cross-flows are important, and therefore subchannel modeling is necessary to accurately predict thermal-hydraulic behavior under boiling conditions. A local reactor assembly code, FLOWTRAN, was modified to model axial flow along the rod bundle as flow through three concentric heated annular passages.

  2. Turbulent interchange in triangular array bare rod bundles

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, J.M.; Todreas, N.

    1977-07-01

    Bulk mixing coefficients were measured for single plane water flow in a simulated rod bundle with a pitch to diameter ratio of 1.10. A tracer technique employing Rhodamine B as the tracer and measuring fluorescence was used. Isokinetic sampling was achieved by using a pressure balance method. The results were corrected for both entrance effects and diversion crossflows. The results showed a change in Reynolds number behavior as the laminar sublayer began to ''choke'' the turbulent mixing. This, and a review of other mixing experiments, suggested that secondary flows do not compensate for laminarization and that turbulent mixing decreases as the pitch to diameter ratio decreases for values of P/D less than 1.05 in a manner similar to that predicted by Ramm et al. Concentration profiles were measured through the clearance gap and the values of the gradient were used to calculate the gap averaged circumferential eddy diffusivity for mass. A discussion of the eddy diffusivity concept and its applicability to turbulent mixing is presented.

  3. CFD study of isothermal water flow in rod bundle with split-type spacer grid

    Science.gov (United States)

    Batta, A.; Class, A. G.

    2014-06-01

    The design of rod bundles in nuclear application nowadays is assessed by CFD (computational fluid dynamics). The accuracy of CFD models need validation. Within the OECD/NEA benchmark MATiS-H (Measurement and Analysis of Turbulent Mixing in Sub-channels - Horizontal) a single-phase water flow in a 5x5 rod bundle is studied. In the benchmark, two types of spacer grids are tested, the swirl type and the split type, where the current study focuses on the split type spacer grid. Comparison of CFD results obtained at Karlsruhe Institut of Technology (KIT) with experimental results of KAERI (Korea Atomic Energy Research Institute) are presented. In the benchmark velocities components along selected lines downstream of the spacer grid are measured and compared to CFD results. The CFD code STAR CCM+ with the Realized k-ɛ model is used. Comparisons with experimental results show quantitative and qualitative agreement for the averaged values of velocity components. Comparisons of results to other benchmark partners using different modeling show that the selected mesh size and models for the analysis of the current case gives relatively accurate results. However, the used turbulent model (Realized k-ɛ does not capture the turbulent intensity correctly. Computation shows that the flow has very high mixing due to the spacer grid, which does not decay within the measurements domain (z/ DH =0-10 downstream of spacer grid). The same conclusion can be drawn from experimental data.

  4. Turbulent flow simulation in a wire-wrap rod bundle of an LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K. [Thermal Hydraulics Section, Reactor Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Sundararajan, T. [Department of Mechanical Engineering, Indian Institute of Technology, Madras, Chennai 600036 (India); Narasimhan, Arunn, E-mail: arunn@iitm.ac.i [Department of Mechanical Engineering, Indian Institute of Technology, Madras, Chennai 600036 (India); Velusamy, K. [Thermal Hydraulics Section, Reactor Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

    2010-05-15

    The pressure drop and heat transfer characteristics of wire-wrapped 19-pin rod bundles in a nuclear reactor subassembly of liquid metal cooled fast breeder reactor (LMFBR) have been investigated through three-dimensional turbulent flow simulations. The predicted results of eddy viscosity based turbulence models (k-epsilon, k-omega) and the Reynolds stress model are compared with those of experimental correlations for friction factor and Nusselt number. The Re is varied between 50,000 and 150,000 and the ratio of helical pitch of wire wrap to the rod diameter is varied from 15 to 45. All the three turbulence models considered yield similar results. The friction factor increases with reduction in the wire-wrap pitch while the heat transfer coefficient remains almost unaltered. However, reduction in the wire-wrap pitch also enhances the transverse flow velocity in the cross-sectional plane as well as the local turbulence intensity, thereby improving the thermal mixing of coolant. Consequently, the presence of wire wrap reduces temperature variation within each section of the subassembly. The associated reduction in differential thermal expansion of rods is expected to improve the structural integrity of the fuel subassembly.

  5. Computational fluid dynamics (CFD) round robin benchmark for a pressurized water reactor (PWR) rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Shin K., E-mail: paengki1@tamu.edu; Hassan, Yassin A.

    2016-05-15

    Highlights: • The capabilities of steady RANS models were directly assessed for full axial scale experiment. • The importance of mesh and conjugate heat transfer was reaffirmed. • The rod inner-surface temperature was directly compared. • The steady RANS calculations showed a limitation in the prediction of circumferential distribution of the rod surface temperature. - Abstract: This study examined the capabilities and limitations of steady Reynolds-Averaged Navier–Stokes (RANS) approach for pressurized water reactor (PWR) rod bundle problems, based on the round robin benchmark of computational fluid dynamics (CFD) codes against the NESTOR experiment for a 5 × 5 rod bundle with typical split-type mixing vane grids (MVGs). The round robin exercise against the high-fidelity, broad-range (covering multi-spans and entire lateral domain) NESTOR experimental data for both the flow field and the rod temperatures enabled us to obtain important insights into CFD prediction and validation for the split-type MVG PWR rod bundle problem. It was found that the steady RANS turbulence models with wall function could reasonably predict two key variables for a rod bundle problem – grid span pressure loss and the rod surface temperature – once mesh (type, resolution, and configuration) was suitable and conjugate heat transfer was properly considered. However, they over-predicted the magnitude of the circumferential variation of the rod surface temperature and could not capture its peak azimuthal locations for a central rod in the wake of the MVG. These discrepancies in the rod surface temperature were probably because the steady RANS approach could not capture unsteady, large-scale cross-flow fluctuations and qualitative cross-flow pattern change due to the laterally confined test section. Based on this benchmarking study, lessons and recommendations about experimental methods as well as CFD methods were also provided for the future research.

  6. Experimental investigation of heat transfer from a 2 × 2 rod bundle to supercritical pressure water

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Han [State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, Xi’an 710049 (China); Bi, Qincheng, E-mail: qcbi@mail.xjtu.edu.cn [State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, Xi’an 710049 (China); Wang, Linchuan; Lv, Haicai [State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, Xi’an 710049 (China); Leung, Laurence K.H. [Atomic Energy of Canada Limited, Chalk River, Ont., Canada K0J 1J0 (Canada)

    2014-08-15

    Highlights: • Heat transfer of supercritical water through a 2 × 2 rod bundles is investigated. • Circumferential wall temperature distribution is obtained. • Effects of system parameters on heat transfer characteristics are analyzed. • Heat transfer correlations are compared against the rod bundle test data. - Abstract: Heat transfer experiments with supercritical pressure water flowing vertically upward through a 2 × 2 rod bundle have been performed at Xi’an Jiaotong University. A fuel-assembly simulator with four heated rods was installed inside a square channel with rounded corner. The outer diameter of each heated rod is 8 mm with an effective heated length of 600 mm. The experiments covered the pressure range of 23–28 MPa, mass-flux range of 350–1000 kg/(m{sup 2} s) and heat-flux range on the rod surface of 200–1000 kW/m{sup 2}. Heat transfer characteristics of supercritical pressure water through the bundle were examined with respect to variations of heat flux, system pressure, and mass flux. These characteristics were shown to be similar to those previously observed in tubes or annuli. The experimental data indicate a non-uniform circumferential wall-temperature distribution around the heated rod. A maximum wall temperature was observed at the surface facing the corner gap between the heated rod and the ceramic tube, while the minimum wall temperature was observed at the surface facing the center subchannel. The difference between maximum and minimum wall temperatures varies with heat flux and/or mass flux. Eight heat transfer correlations developed for supercritical water were assessed against the current set of test data. Prediction of the Jackson correlation agrees closely with the experimental Nusselt number. A new correlation has been derived based on Jackson correlation to improve the prediction accuracy of supercritical heat transfer coefficient in a 2 × 2 rod bundle.

  7. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report, March 1, 1976--May 31, 1976

    Energy Technology Data Exchange (ETDEWEB)

    Todreas, N.E.; Golay, M.W.; Wolf, L.

    1976-01-01

    Progress is reported for the following tasks: bundle geometry studies; subchannel geometry studies; outlet plenum flow mixing studies; and the theoretical determination of local temperature fields in rod bundles.

  8. TREAT Neutronics Analysis of Water-Loop Concept Accommodating LWR 9-rod Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Hill, Connie M.; Woolstenhulme, Nicolas E.; Parry, James R.; Bess, John D.; Housley, Gregory K.

    2016-09-01

    Abstract. Simulation of a variety of transient conditions has been successfully achieved in the Transient Reactor Test (TREAT) facility during operation between 1959 and 1994 to support characterization and safety analysis of nuclear fuels and materials. A majority of previously conducted tests were focused on supporting sodium-cooled fast reactor (SFR) designs. Experiments evolved in complexity. Simulation of thermal-hydraulic conditions expected to be encountered by fuels and materials in a reactor environment was realized in the development of TREAT sodium loop experiment vehicles. These loops accommodated up to 7-pin fuel bundles and served to simulate more closely the reactor environment while safely delivering large quantities of energy into the test specimen. Some of the immediate TREAT restart operations will be focused on testing light water reactor (LWR) accident tolerant fuels (ATF). Similar to the sodium loop objectives, a water loop concept, developed and analyzed in the 1990’s, aimed at achieving thermal-hydraulic conditions encountered in commercial power reactors. The historic water loop concept has been analyzed in the context of a reactivity insertion accident (RIA) simulation for high burnup LWR 2-pin and 3-pin fuel bundles. Findings showed sufficient energy could be deposited into the specimens for evaluation. Similar results of experimental feasibility for the water loop concept (past and present) have recently been obtained using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries. The old water loop concept required only two central TREAT core grid spaces. Preparation for future experiments has resulted in a modified water loop conceptual design designated the TREAT water environment recirculating loop (TWERL). The current TWERL design requires nine TREAT core grid spaces in order to place the water recirculating pump under the TREAT core. Due to the effectiveness of water moderation, neutronics analysis shows that removal of seven additional

  9. Effects of fuel relocation on reflood in a partially-blocked rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Jae [School of Mechanical Engineering, Chungnam National University, 99 Daehak-ro, Yuseong-gu, Daejeon 34134 (Korea, Republic of); Kim, Jongrok; Kim, Kihwan; Bae, Sung Won [Thermal-Hydraulic Safety Research Division, Korea Atomic Energy Research Division, 111 Daedeok-daero, Yuseong-gu, Daejeon 34057 (Korea, Republic of); Moon, Sang-Ki, E-mail: skmoon@kaeri.re.kr [Thermal-Hydraulic Safety Research Division, Korea Atomic Energy Research Division, 111 Daedeok-daero, Yuseong-gu, Daejeon 34057 (Korea, Republic of)

    2017-02-15

    Ballooning of the fuel rods has been an important issue, since it can influence the coolability of the rod bundle in a large-break loss-of-coolant accident (LBLOCA). Numerous past studies have investigated the effect of blockage geometry on the heat transfer in a partially blocked rod bundle. However, they did not consider the occurrence of fuel relocation and the corresponding effect on two-phase heat transfer. Some fragmented fuel particles located above the ballooned region may drop into the enlarged volume of the balloon. Accordingly, the fuel relocation brings in a local power increase in the ballooned region. The present study’s objective is to investigate the effect of the fuel relocation on the reflood under a LBLOCA condition. Toward this end, experiments were performed in a 5 × 5 partially-blocked rod bundle. Two power profiles were tested: one is a typical cosine shape and the other is the modified shape considering the effect of the fuel relocation. For a typical power shape, the peak temperature in the ballooned rods was lower than that in the intact rods. On the other hand, for the modified power shape, the peak temperature in the ballooned rods was higher than that in the intact rods. Numerical simulations were also performed using the MARS code. The tendencies of the peak clad temperatures were well predicted.

  10. Double-clad nuclear fuel safety rod

    Science.gov (United States)

    McCarthy, William H.; Atcheson, Donald B.; Vaidyanathan, Swaminathan

    1984-01-01

    A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

  11. Hydrodynamic Experiments for a Flow Distribution of a 61-pin Wire-wrapped Rod Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Chang, S. K.; Euh, D. J.; Choi, H. S.; Kim, H. M.; Ko, Y. J.; Lee, D. W.; Lee, H. Y.; Choi, S. R. [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    Fuel assembly of the SFR (Sodium-cooled Fast breeder Reactor) type reactor generally has wire spacers which are wrapped around each fuel pin helically in axial direction. The configuration of a helical wire spacer guarantees the fuel rods integrity by providing the bundle rigidity, proper spacing between rods and promoting coolant mixing between subchannels. It is important to understand the flow characteristics in such a triangular array wire wrapped rod bundle in a hexagonal duct. The experimental work has been undertaken to quantify the friction and mixing parameters which characterize the flow distribution in subchannels for the KAERI's own bundle geometric configuration. This work presents the hydrodynamic experimental results for the flow distribution and the pressure drop in subchannels of a 61-pin wire wrapped rod bundle which has been fabricated considering the hydraulic similarity of the reference reactor. Hydrodynamic experiments for a 61-pin wire wrapped test assembly has been performed to provide the data of a flow distribution and pressure losses in subchannels for verifying the analysis capability of subchannel analysis codes for a KAERI's own prototype SFR reactor. Three type of sampling probes have been specially designed to conserve the shape of the flow area for each type of subchannels. All 126 subchannels have been measured to identify the characteristics of the flow distribution in a 37-pin rod assembly. Pressure drops at the interior and the edge subchannels have been also measured to recognize the friction losses of each type of subchannels.

  12. A burnout correlation for flow of boiling water in vertical rod bundles

    Energy Technology Data Exchange (ETDEWEB)

    Becker, Kurt M.

    1967-04-15

    The rod bundle burnout correlation described in the present report is a development from our earlier published rod bundle correlation for low pressures. The correlation is based on the Becker round duct correlation and is written on the form x{sub BO} = 0.68*{eta}*{eta}{sub L}*X{sub RD} where x{sub RD} is the burnout steam quality in a round duc at corresponding flow conditions, {eta} is the ratio of heated to total perimeter and {eta}{sub l} is a correction factor, which is a function of q/A only. It is demonstrated that this equation combined with the heat balance equation q/A = G/(4L/D{sub H})*({delta}h{sub SUB} + X{sub BO}*H{sub fg}) predicts the burnout heat fluxes for 312 measurements obtained in our laboratory within a scatter of {+-}7. 5 per cent and with an RMS error of 3.8 per cent. The measurements were obtained in the following ranges of variables. Number of rods n 1, 3, 6 and 7; Rod diameter d{sub i} 10.05 - 13.80 mm; Shroud diameter d{sub o} 17. 42 - 71. 0 mm; Rod clearance s 3.7 - 8.8 mm; Heated length L 608 - 4440 mm; Pressure p 20-71 kg/cm{sup 2}, Inlet sub-cooling {delta}t{sub sub} 3 - 240 deg C; Mass velocity G 80-1,500 kg/m{sup 2}; Burnout heat flux q/A 74-314 W/cm{sup 2}; Burnout steam quality x{sub BO} 0. 1 - 0.55. The correlation shows that the burnout conditions in wide ranges of variables are independent of the inlet sub-cooling and the heated length, and that the effects of mass velocity and pressure are the same in rod bundles and in round tubes. It is also demonstrated that the effects of a radial heat flux variation within the rod bundle can be handled by the correlation by modifying the {eta}-value for the bundle. The rod bundle data presented by Janssen and Kervinen, Hench, Obertelli, Matzner, Haslam, Edwards and Obertelli and Hench and Boehm were also analysed in terms of the measured and predicted burnout heat fluxes. These data covered bundles consisting of 3, 4, 6, 7, 9. 19 and 36 rods and it was found that a very good agreement

  13. AgInCd control rod failure in the QUENCH-13 bundle test

    Energy Technology Data Exchange (ETDEWEB)

    Sepold, L. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung, Nuclear Safety Program (NUKLEAR), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)], E-mail: leo.sepold@imf.fzk.de; Lind, T. [Paul Scherrer Institut, Laboratory for Thermalhydralics (LTH), Department of Nuclear Energy and Safety (NES), 5232 Villigen PSI (Switzerland); Csordas, A. Pinter [Fuel Materials Department, HAS KFKI AEKI, 1121 Budapest (Hungary); Stegmaier, U.; Steinbrueck, M.; Stuckert, J. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung, Nuclear Safety Program (NUKLEAR), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2009-09-15

    The QUENCH off-pile experiments performed at the Karlsruhe Research Center are to investigate the high-temperature behavior of Light Water Reactor (LWR) core materials under transient conditions and in particular the hydrogen source term resulting from the water injection into an uncovered LWR core. The typical LWR-type QUENCH test bundle, which is electrically heated, consists of 21 fuel rod simulators with a total length of approximately 2.5 m. The Zircaloy-4 rod claddings and the grid spacers are identical to those used in Pressurized Water Reactors (PWR) whereas the fuel is represented by ZrO{sub 2} pellets. In the QUENCH-13 experiment the single unheated fuel rod simulator in the center of the test bundle was replaced by a PWR-type control rod. The QUENCH-13 experiment consisting of pre-oxidation, transient, and quench water injection at the bottom of the test section investigated the effect of an AgInCd/stainless steel/Zircaloy-4 control rod assembly on early-phase bundle degradation and on reflood behavior. Furthermore, in the frame of the EU 6th Framework Network of Excellence SARNET, release and transport of aerosols of a failed absorber rod were to be studied in QUENCH-13, which was accomplished with help of aerosol measurements performed by PSI-Switzerland and AEKI-Hungary. Control rod failure was initiated by eutectic interaction of steel cladding and Zircaloy-4 guide tube and was indicated at about 1415 K by axial peak absorber and bundle temperature responses and additionally by the on-line aerosol monitoring system. Significant releases of aerosols and melt relocation from the control rod were observed at an axial peak bundle temperature of 1650 K. At a maximum bundle temperature of 1820 K reflood from the bottom was initiated with cold water at a flooding rate of 52 g/s. There was no noticeable temperature escalation during quenching. This corresponds to the small amount of about 1 g in hydrogen production during the quench phase (compared to 42 g

  14. Experimental investigation on anisotropic turbulent flow in a 6 × 6 rod bundle with LDV

    Energy Technology Data Exchange (ETDEWEB)

    Xiong, Jinbiao, E-mail: xiongjinbiao@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University (China); Yu, Nan [State Nuclear Power Software Development Center (China); Yu, Yang [School of Nuclear Science and Engineering, Shanghai Jiao Tong University (China); Fu, Xiaoliang [State Nuclear Power Software Development Center (China); Cheng, Xu [School of Nuclear Science and Engineering, Shanghai Jiao Tong University (China); Yang, Yanhua [School of Nuclear Science and Engineering, Shanghai Jiao Tong University (China); State Nuclear Power Software Development Center (China)

    2014-10-15

    Highlights: • Five-beam three-component LDV is applied to measure flow in a 6 × 6 rod bundle. • Three-dimension flow field is obtained at the outlet. • The effects of spacer and Reynolds number on flow are investigated. • Three components of mean velocity scale with the bulk velocity. • The Reynolds stresses scale with the square of average bulk velocity. - Abstract: The five-beam three-component laser Doppler velocimetry (LDV) is applied to investigate the turbulent flow in a 6 × 6 rod bundle installed with simple grid spacers. LDV measurement has been conducted at four cross sections downstream a grid spacer at five Reynolds numbers ranging from 6600 to 70,300. The flow evolution downstream of the grid spacer is demonstrated through the comparison of the axial mean velocity and Root Mean Square (RMS) velocity at the three cross sections downstream of the grid spacer. All the three components of the flow velocity are measured in the selected subchannels at the outlet cross section of the rod bundle which is dedicated to provide more information on the turbulence statistics in the rod bundle flow. Remarkably high ratio of axial normal stress to the turbulent kinetic energy, vv{sup ¯}/k, is observed even in the subchannel center, which indicates that the turbulence in the rod bundle is anisotropic. Comparing experiment results at the five Reynolds numbers, the low Reynolds number effect is found in the case with Re = 6.6 × 10{sup 3}. The experiment results also imply that the Reynolds number effect in the tight-lattice bundle is weak compare to that in the loose one.

  15. Effects of sleeve blockages on axial velocity and intensity of turbulence in an unheated 7 x 7 rod bundle. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Creer, J.M.; Rowe, D.S.; Bates, J.M.; Sutey, A.M.

    1976-01-01

    An experimental study is described which was performed to investigate the turbulent flow phenomena near postulated sleeve blockages in a model nuclear fuel rod bundle. The sleeve blockages were characteristic of fuel clad ''swelling'' or ''ballooning'' which could occur during loss-of-coolant accidents (LOCA) in pressurized water reactors. The study was conducted to provide information relative to the flow phenomena near postulated blockages to support detailed safety analyses of LOCAs. The results of the study are especially useful for verification of the hydraulic treatment of reactor core computer programs such as COBRA.

  16. Experimental study of laminar mixed convection in a rod bundle with mixing vane spacer grids

    Energy Technology Data Exchange (ETDEWEB)

    Mohanta, Lokanath, E-mail: lxm971@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University, University Park, PA 16802 (United States); Cheung, Fan-Bill [Department of Mechanical and Nuclear Engineering, Pennsylvania State University, University Park, PA 16802 (United States); Bajorek, Stephen M.; Tien, Kirk; Hoxie, Chris L. [Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 (United States)

    2017-02-15

    Highlights: • Investigated the heat transfer during mixed laminar convection in a rod bundle with linearly varying heat flux. • The Nusselt number increases downstream of the inlet with increasing Richardson number. • Developed an enhancement factor to account for the effects of mixed convection over the forced laminar heat transfer. - Abstract: Heat transfer by mixed convection in a rod bundle occurs when convection is affected by both the buoyancy and inertial forces. Mixed convection can be assumed when the Richardson number (Ri = Gr/Re{sup 2}) is on the order of unity, indicating that both forced and natural convection are important contributors to heat transfer. In the present study, data obtained from the Rod Bundle Heat Transfer (RBHT) facility was used to determine the heat transfer coefficient in the mixed convection regime, which was found to be significantly larger than those expected assuming purely forced convection based on the inlet flow rate. The inlet Reynolds (Re) number for the tests ranged from 500 to 1300, while the Grashof (Gr) number varied from 1.5 × 10{sup 5} to 3.8 × 10{sup 6} yielding 0.25 < Ri < 4.3. Using results from RBHT test along with the correlation from the FLECHT-SEASET test program for laminar forced convection, a new correlation ​is proposed for mixed convection in a rod bundle. The new correlation accounts for the enhancement of heat transfer relative to laminar forced convection.

  17. Reflood experiments in rod bundles with flow blockages due to clad ballooning

    Energy Technology Data Exchange (ETDEWEB)

    Moon, S.K.; Kim, J.; Kim, K.; Kim, B.J.; Park, J.K.; Youn, Y.J.; Choi, H.S.; Song, C.H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-07-15

    Clad ballooning and the resulting partial flow blockage are one of the major thermal-hydraulic concerns associated with the coolability of partially blocked cores during a loss-of-coolant accident (LOCA). Several in-pile tests have shown that fuel relocation causes a local power accumulation and a high thermal coupling between the clad and fuel debris in the ballooned regions. However, previous experiments in the 1980s did not take into account the fuel relocation phenomena and resulting local power increase in the ballooned regions. The present paper presents the results of systematic investigations on the coolability of rod bundles with flow blockages. The experiments were mainly performed in 5 x 5 rod bundles, 2 x 2 rod bundles and other test facilities. The experiments include a reflood heat transfer, single-phase convective heat transfer, flow redistributions phenomena, and droplet break-up behavior. The effects of the fuel relocation and resulting local power increase were investigated using a 5 x 5 rod bundle. The fuel relocation phenomena increase the peak cladding temperature.

  18. Turbulent mixing in a rod bundle with vaned spacer grids: OECD/NEA–KAERI CFD benchmark exercise test

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Seok-Kyu; Kim, Seok; Song, Chul-Hwa, E-mail: chsong@kaeri.re.kr

    2014-11-15

    Highlights: • Detailed velocity profiles have been examined in a rod bundle with mixing spacer grids. • Mixing characteristics strongly depend on the type of the mixing vane on a spacer grid. • The swirl in subchannels is elliptic and the cross-flow in gaps is vigorous in the split-type. • Swirl-type vanes generate a circular swirl in a subchannel and a weak cross-flow in gaps. • Mixing performance is superior in the case of the split-type compared to the swirl-type. - Abstract: An experimental study titled the 2nd International Benchmark Exercise (IBE-2) has been conducted to provide high-precision data of detailed turbulent flow mixing in a rod bundle for validating the CFD codes being used widely in the nuclear power industry. A 5 × 5 rod bundle having mixing spacer grids was adopted as a test rig, and was contained in a square flow housing with a 170 mm side length and 4670 mm length. The 25 rods in a bundle have dimensions of 25.4 mm in outer diameter and a 3863 mm length. The benchmark experiments have been performed at the MATiS-H water loop facility in KAERI. The axial bulk velocity in a rod bundle was maintained at about 1.50 m/s (equivalent to Re ∼50,000) with loop conditions of 35 °C and 1.57 bar measured upstream of the spacer during the experiments. Detailed measurements of the turbulent flow in the subchannels were accomplished using 2-D LDA at four different distances (0.5, 1, 4 and 10 D{sub H}) from the downstream of the mixing spacer grid. The upstream flow profiles also have been measured at the inlet of the mixing spacer grid for the inlet boundary condition. Precise measurements of the lateral and axial velocities in the subchannels are presented at four downstream distances, as well as the inlet from the mixing spacer grid of two types. Turbulence intensities and vorticities in the subchannels are also evaluated from the velocity measurements.

  19. Influence on rewetting temperature and wetting delay during rewetting rod bundle by various radial jet models

    Energy Technology Data Exchange (ETDEWEB)

    Debbarma, Ajoy; Pandey, Krishna Murari [National Institute of Technology, Assam (India). Dept. of Mechanical Engineering

    2016-03-15

    Numerical investigation of the rewetting of single sector fuel assembly of Advanced Heavy Water Reactor (AHWR) has been carried out to exhibit the effect of coolant jet diameters (2, 3 and 4 mm) and jet directions (Model: M, X and X2). The rewetting phenomena with various jet models are compared on the basis of rewetting temperature and wetting delay. Temperature-time curve have been evaluated from rods surfaces at different circumference, radial and axial locations of rod bundle. The cooling curve indicated the presence of vapor in respected location, where it prevents the contact between the firm and fluid phases. The peak wall temperature represents as rewetting temperature. The time period observed between initial to rewetting temperature point is wetting delay. It was noted that as improved in various jet models, rewetting temperature and wetting delay reduced, which referred the coolant stipulation in the rod bundle dominant vapor formation.

  20. Spent nuclear fuel rods encapsulated in copper

    Energy Technology Data Exchange (ETDEWEB)

    Hanes, H.D.

    1984-04-01

    Using hot isostatic pressing, spent nuclear fuel rods and other radioactive wastes can be encapsulated in solid copper. The copper capsule which is formed is free of pores and cracks, and is highly resistant to attack by reducing ground waters. Such capsules should contain radioactive materials safely for hundreds of thousands of years in underground storage.

  1. LWR nuclear fuel bundle data for use in fuel bundle handling

    Energy Technology Data Exchange (ETDEWEB)

    Weihermiller, W.B.; Allison, G.S.

    1979-09-01

    Although increasing numbers of spent light water reactor (LWR) fuel bundles are moved into storage, no handling equipment is set up to manipulate all of the various types of fuel bundles. This report summarizes fuel bundle information of interest to the designer of such handling equipment. Dimensional descriptions are included with discussions of assembly procedure and manufacturer provisions for handling equipment. No attempt is made to make a complete compilation of dimensional information; the number of fuel bundle designs and design revisions makes it impractical. Because the fuel bundle designs are so varied, any equipment intended for handling all types of bundles will have to be designed with flexibility in mind. Besides the ability to manipulate fuel bundles in space, handling equipment may be required to locate an external surface or to position a cutting operation to avoid breaking a fuel rod pressure boundary. Even with the most sophisticated and flexible handling equipment, some situations will require use of the manufacturers' as-built descriptions of individual fuel bundles.

  2. Effect of Flow Blockage on the Coolability during Reflood in a 2 × 2 Rod Bundle

    Directory of Open Access Journals (Sweden)

    Kihwan Kim

    2014-01-01

    Full Text Available During the reflood phase of a large-break loss-of-coolant accident (LBLOCA in a pressurized-water reactor (PWR, the fuel rods can be ballooned or rearranged owing to an increase in the temperature and internal pressure of the fuel rods. In this study, an experimental study was performed to understand the thermal behavior and effect of the ballooned region on the coolability using a 2 × 2 rod bundle test facility. The electrically heated rod bundle was used and the ballooning shape of the rods was simulated by superimposing hollow sleeves, which have a 90% blockage ratio. Forced reflood tests were performed to examine the transient two-phase heat transfer behavior for different reflood rates and rod powers. The droplet behaviors were also investigated by measuring the velocity and size of droplets near the blockage region. The results showed that the heat transfer was enhanced in the downstream of the blockage region, owing to the reduced flow area of the subchannel, intensification of turbulence, and deposition of the droplet.

  3. Nanofluid Applied Numerical Analysis of Subchannel in Square Rod Bundle for Fusion-Fission Hybrid System

    Energy Technology Data Exchange (ETDEWEB)

    Shamim, Jubair Ahmed; Bhowmik, Palash Kumar; Suh, Kune Y. [Seoul National Univ., Seoul (Korea, Republic of)

    2014-05-15

    Most of the traditional ways available in the literature to enhance heat transfer are mainly based on variation of structures like addition of heat surface area such as fins, vibration of heated surface, injection or suction of fluids, applying electrical or magnetic fields, and so forth. Application of these mechanical techniques to a fuel rod bundle will involve not only designing complex geometries but also using many additional mechanisms inside a nuclear reactor core which in turn will certainly increase the manufacturing cost as well as may hamper various safety features essential for sound and uninterrupted operation of a nuclear power reactor. On the other hand, traditional heat transfer fluids such as water, ethylene glycol and oils have inherently low thermal conductivity relative to metals and even metal oxides. In this study the coolant with suspended nano-sized particles in the base fluid is proposed as an alternative to increase heat transfer but minimize flow resistance inside a nuclear reactor core. Due to technical complexities most of the previous studies carried out on heat transfer of suspension of metal oxides in fluids were limited to suspensions with millimeter or micron-sized particles. Such outsized particles may lead to severe problems in heat transfer equipment including increased pressure drop and corrosion and erosion of components and pipe lines. Dramatic advancement in modern science has made it possible to produce ultrafine metallic or nonmetallic particles of nanometer dimension, which has brought a revolutionary change in the research of heat transfer enhancement methods. Due to very tiny particle size and their small volume fraction, problems such as clogging and increased pressure drop are insignificant for nanofluids. Moreover, the relatively large surface area of nanoparticles augments the stability of nanofluid solution and prevents the sedimentation of nanoparticles. Xuan and Roetzel considered two approaches to illustrate

  4. Characteristics of turbulent velocity and temperature in a wall channel of a heated rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Krauss, T.; Meyer, L. [Forschungszentrum Karlsruhe (Germany)

    1995-09-01

    Turbulent air flow in a wall sub-channel of a heated 37-rod bundle (P/D = 1.12, W/D = 1.06) was investigated. measurements were performed with hot-wire probe with X-wires and a temperature wire. The mean velocity, the mean fluid temperature, the wall shear stress and wall temperature, the turbulent quantities such as the turbulent kinetic energy, the Reynolds-stresses and the turbulent heat fluxes were measured and are discussed with respect to data from isothermal flow in a wall channel and heated flow in a central channel of the same rod bundle. Also, data on the power spectral densities of the velocity and temperature fluctuations are presented. These data show the existence of large scale periodic fluctuations are responsible for the high intersubchannel heat and momentum exchange.

  5. Laminar simulation of intersubchannel mixing in a triangular nuclear fuel bundle geometry

    Energy Technology Data Exchange (ETDEWEB)

    Zaretsky, A.; Lightstone, M.F., E-mail: lightsm@mcmaster.ca; Tullis, S.

    2015-12-15

    Highlights: • Quasi-periodic flow was observed through rod-to-wall gaps. • Triangular subchannel flows were fundamentally irregular. • Cross-gap flow was influenced both by local and adjacent cross-gap intensity. • Phase-linking between gaps induced cross-plane peripheral circulation through rod–wall gaps. • Cross-gap flow structure was dependent on subchannel geometry. - Abstract: Predicting temperature distributions in fuel rod bundles is an important component of nuclear reactor safety analysis. Intersubchannel mixing acts to homogenize coolant temperatures thus reducing the likelihood of localized regions of high fuel temperature. Previous research has shown that intersubchannel mixing in nuclear fuel rod bundles is enhanced by a large-scale quasi-periodic energetic fluid motion, which transports fluid on the cross-plane between the narrow gaps connecting subchannels. This phenomenon has also been observed in laminar flows. Unsteady laminar flow simulations were performed in a simplified bundle of three rods with a pipe. Three similar geometries of varying gap width were examined, and a thermal trace was implemented on the first geometry. Thermal mixing was driven by the advection of energy between subchannels by the cross-plane flow. Flow through the rod-to-wall gaps in the wall subchannels alternated with a dominant frequency, particularly when rod-to-wall gaps were smaller than rod-to-rod gaps. Significant phase-linking between rod-to-wall gaps was also observed such that a peripheral circulation occurred through each gap simultaneously. Cross-plane flow through the rod-to-rod gaps in the triangular subchannel was irregular in each case. This was due to the fundamental irregularity of the triangular subchannel geometry. Vortices were continually broken up by cross-plane flow from other gaps due to the odd number of fluid pathways within the central subchannel. Cross-plane flow in subchannel geometries is highly interconnected between gaps. The

  6. Numerical Simulation for Frictional Loss and Local Loss of a 5*5 SMART Rod Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong-Pil; Kim, Seong Jin; Kwon, Hyuk; Seo, Kyong-Won; Hwang, Dae-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The results showed good agreement with experimental data and/or reasonable values. However, these results were dependent on computational meshes and turbulence models and it still remains important issues in CFD analysis. The aim of present work is to assess the pressure drop in a 5*5 SMART rod bundle using 3D CFD code with various computational meshes and turbulence models. In the present work, 3D CFD code was utilized to investigate pressure drop in a SMART 5*5 rod bundle. The predicted pressure drop was strongly dependent with computational meshes and turbulence models. Based on CFD results in this study, least five of six meshes within the subchannel gap are required to get reliable result which is insensitive to the number of meshes. The friction factor predicted by k - ε model is good agreement with McAdams's correlation while SST model overestimate McAdams's correlation. However, it is difficult to judge performance of turbulence model because of lock of experimental data for a 5*5 SMART bare rod bundle. For nominal condition (Re-194,000) of SMART, SST model predict k-factor of MV and IFM grid as 1.304 and 0.748, respectively. This value is reasonable as compared with designed k-factor, 1.320 and 0.78.

  7. Turbulence Model Evaluation Study for a Secondary Flow and a Flow Pulsation in the Sub-Channels of an 18-Finned Rod Bundle by Using Computational Fluid Dynamics

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Hark; Chae, Hee Taek; Park, Cheol; Kim, Heon Il

    2008-09-15

    Since the heat flux of the rod type fuel used in the HANARO, a research reactor being operated in the KAERI, is substantially higher than the heat flux of power reactors, the HANARO fuel has 8 longitudinal fins for enhancing the heat release from the fuel rod surface. This unique shape of a nuclear fuel led us to study the flows and thermal hydraulic characteristics of it. Especially because the flows through the narrow channels built up by these finned rod fuels would be different from the flow characteristics in the coolant channels formed by bare rod fuels, some experimental studies to investigate the flow behaviors and structures in a finned rod bundle were done by other researchers. But because of the very complex geometries of the flow channels in the finned rod bundle only allowed us to obtain limited information about the flow characteristics, a numerical study by a computational fluid dynamics technique has been adopted to elucidate more about such a complicated flow in a finned rod bundle. In this study, for the development of an adequate computational model to simulate such a complex geometry, a mesh sensitivity study and the effects of various turbulence models were examined. The CFD analysis results were compared with the experimental results. Some of them have a good agreement with the experimental results. All linear eddy viscosity turbulence models could hardly predict the secondary flows near the fuel surfaces and in the sub-channel, but the RSM (Reynolds Stress Model) revealed very different results from the eddy viscosity turbulence models. In the transient analysis all turbulence model predicted flow pulsation at the center of a subchannel as well as at the gap between rods in spite of large P/D. The flow pulsation showed different results with turbulence models and the location in the sub-channels.

  8. Subchannel analysis and correlation of the Rod Bundle Heat Transfer (RBHT) steam cooling experimental data

    Energy Technology Data Exchange (ETDEWEB)

    Riley, M.P.; Mohanta, L.; Miller, D.J.; Cheung, F.B. [Pennsylvania State Univ., University Park, PA (United States); Bajorek, S.M.; Tien, K.; Hoxie, C.L. [U.S. Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Research

    2016-07-15

    A subchannel analysis of the steam cooling data obtained in the Rod Bundle Heat Transfer (RBHT) test facility has been performed in this study to capture the effect of spacer grids on heat transfer. The RBHT test facility has a 7 x 7 rod bundle with heater rods and with seven spacer grids equally spaced along the length of the rods. A method based on the concept of momentum and heat transport analogy has been developed for calculating the subchannel bulk mean temperature from the measured steam temperatures. Over the range of inlet Reynolds number, the local Nusselt number was found to exhibit a minimum value between the upstream and downstream spacer grids. The presence of a spacer grid not only affects the local Nusselt number downstream of the grid but also affects the local Nusselt number upstream of the next grid. A new correlation capturing the effect of Reynolds number on the local flow restructuring downstream as well as upstream of the spacer grids was proposed for the minimum Nusselt number. In addition, a new enhancement factor accounting for the effects of the upstream as well as downstream spacer grids was developed from the RBHT data. The new enhancement factor was found to compare well with the data from the ACHILLLES test facility.

  9. CFD Verification of 5x5 Rod Bundle with Mixing Vane Spacer Grids

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sungkew; Jang, Hyungwook; Lim, Jongseon; Park, Eungjun; Nahm, Keeyil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Results of the CHF test are used for determining the CHF correlation, which is used to evaluate the thermal margin in the reactor core. Computational fluid dynamics (CFD) has been used to save the time and cost for experimental tests, components design and complicated phenomena in all industries including the reactor coolant system. L. D. Smith et al. applied the CFD methodology in a 5x5 rod bundle with the mixing vane spacer grid using the renormalization group (RNG) k-epsilon model. This CFD model agreed reasonably well with the test data. M. E. Conner et al. conducted experiments to validate the CFD methodology for the single-phase flow conditions in PWR fuel assemblies. In this validation case, the CFD code predicted very similar flow field structures as the test data. In this study, a CFD simulation under single-phase flow condition was conducted for one specific condition in a thermal mixing flow test of 5x5 rod bundle with some mixing vane spacer grids. In this study, a CFD simulation under a single-phase flow condition was conducted for one specific condition in a thermal mixing flow test of 5x5 rod bundle with the mixing vane spacer grids to verify the applicability of the CFD model for predicting the outlet temperature distribution. FLUENT 14.5 Version was used in this CFD analysis. For the successful prediction of the wall bounded turbulent flows, the y+ with 3 prism layers was determined within 5. At this time, k-epsilon standard turbulence model was used. The temperature distribution of CFD for each sub-channel at the outlet region of test bundle showed the difference approximately within 1.1% and 0.2% while comparing to that of test and sub-channel analysis code, respectively.

  10. Evaluation of CHF experimental data for non-square lattice 7-rod bundles

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hyun; Yoo, Y. J.; Kim, K. K.; Zee, S. Q

    2001-01-01

    A series of CHF experiments are conducted for 7-rod hexagonal test bundles in order to investigate the CHF characteristics of self-sustained square finned (SSF) rod bundles. The experiments are performed in the freon-loop and water-loop located at IPPE in Russia, and 609 data of freon-12 and 229 data of water are obtained from 7 kinds of test bundles classified by the combination of heated length and axial/radial power distributions. As the result of the evaluation of four representative CHF correlations, the EPRI-1 correlation reveals good prediction capability for SSF test bundles. The inlet parameter CHF correlation suggested by IPPE calculates the mean and the standard deviation of P/M for uniformly heated test bundles as 1.002 and 0.049, respectively. In spite of its excellent accuracy, the correlation has a discontinuity point at the boundary between the low velocity and high velocity conditions. KAERI's inlet parameter correlation eliminates this defect by introducing the complete evaporation model at low velocity condition, and calculates the mean and standard deviation of P/M as 0.095 and 0.062 for uniformly heated 496 data points, respectively. The mean/standard deviation of local parameter CHF correlations suggested by IPPE and KAERI are evaluated as 1.023/0.178 and 1.002/0.158, respectively. The inlet parameter correlation developed from uniformly heated test bundles tends to under-predict CHF about 3% for axially non-uniformly heated test bundles. On the other hand, the local parameter correlation reveals large scattering of P/M, and requires re-optimization of the correlation for non-uniform axial power distributions. As the result of the analysis of experimental data, it reveals that the correction model of axial power shapes suggested by IPPE is applicable to the inlet parameter correlations. For the test bundle of radial non-uniform power distribution, the physically unexpected results are obtained at some experimental conditions. In addition

  11. Empirical models for liquid metal heat transfer in the entrance region of tubes and rod bundles

    Science.gov (United States)

    Jaeger, Wadim

    2016-10-01

    Experiments focusing on liquid metals heat transfer in pipes and rod bundles with thermally and hydraulically developing flow are reviewed. Empirical heat transfer correlations are developed for engineering applications. In the developing regions the heat transfer is in-stationary. The heat transfer at the entrance is around 100 % higher due to the developing process including the lateral exchange of energy and momentum than for developed flow. Developing flow is not physically considered in the framework of system codes, which are used for thermal-hydraulic analysis of power and process plants with a multitude of components like pipes, tanks, valves and heat exchangers. Therefore, the application to liquid metal flows is limited to developed flow, which is independent of the distance from the flow entrance. The heat transfer enhancement in developing flows is important for the optimization of components like heat exchangers and helps to reduce unnecessary conservatism. In this work, empirical models are developed to account for developing flows in pipes and rod bundles. A literature review is performed to collect available experimental data for developing flow in liquid metal heat transfer. The evaluation shows that the length for pure thermally developing pipe flow is much larger (20-30 hydraulic diameters) than for combined thermally and hydraulically developing flow (10-15 hydraulic diameters). In rod bundles, fully combined developed flow is established after 30-40 hydraulic diameters downstream of the entrance. The derived empirical models for the heat transfer enhancement in the developing regions are implemented into a best estimate system code. The validation of these models by means of post-test analyses of 16 experiments shows that they are very well able to represent the heat transfer in developing regions.

  12. Empirical models for liquid metal heat transfer in the entrance region of tubes and rod bundles

    Science.gov (United States)

    Jaeger, Wadim

    2017-05-01

    Experiments focusing on liquid metals heat transfer in pipes and rod bundles with thermally and hydraulically developing flow are reviewed. Empirical heat transfer correlations are developed for engineering applications. In the developing regions the heat transfer is in-stationary. The heat transfer at the entrance is around 100 % higher due to the developing process including the lateral exchange of energy and momentum than for developed flow. Developing flow is not physically considered in the framework of system codes, which are used for thermal-hydraulic analysis of power and process plants with a multitude of components like pipes, tanks, valves and heat exchangers. Therefore, the application to liquid metal flows is limited to developed flow, which is independent of the distance from the flow entrance. The heat transfer enhancement in developing flows is important for the optimization of components like heat exchangers and helps to reduce unnecessary conservatism. In this work, empirical models are developed to account for developing flows in pipes and rod bundles. A literature review is performed to collect available experimental data for developing flow in liquid metal heat transfer. The evaluation shows that the length for pure thermally developing pipe flow is much larger (20-30 hydraulic diameters) than for combined thermally and hydraulically developing flow (10-15 hydraulic diameters). In rod bundles, fully combined developed flow is established after 30-40 hydraulic diameters downstream of the entrance. The derived empirical models for the heat transfer enhancement in the developing regions are implemented into a best estimate system code. The validation of these models by means of post-test analyses of 16 experiments shows that they are very well able to represent the heat transfer in developing regions.

  13. Development of advanced BWR fuel bundle with spectral shift rod - BWR core characteristics with SSR

    Energy Technology Data Exchange (ETDEWEB)

    Hino, T.; Kondo, T.; Chaki, M.; Ohga, Y. [Hitachi-GE Nuclear Energy, Ltd., 1-1, Saiwai-cho, 3-chome, Hitachi-shi, Ibaraki-ken, 317-0073 (Japan); Makigami, T. [Tokyo Electric Power Company Inc., 1-1-3, Uchisaiwai-cho, Chiyoda-ku, Tokyo, 100-0011 (Japan)

    2012-07-01

    The neutron energy spectrum can be varied during an operation cycle to generate and utilize more plutonium from the non-fissile {sup 238}U by changing the void fraction in the core through control of the core coolant flow rate. This operation method, which is called a spectral shift operation, is practiced in BWRs to save natural uranium. A new component called a spectral shift rod (SSR), which is utilized instead of a conventional water rod, has been introduced to amplify the void fraction change and increase the spectral shift effect. In this study, fuel bundle design with the SSR and core design were carried out for the ABWR and the next generation BWR, HP-ABWR (High-Performance ABWR).The core characteristics with the SSR were evaluated and compared with those when using the conventional water rod. Influences of uncertainty of the water level in the SSR on the safety limit minimum critical power ratio (SLMCPR) were considered for evaluation of the uranium saving effect attained by the SSR. As a result, it was found that the amount of natural uranium needed for an operation cycle could be reduced more than 3% with 20% core coolant flow change and more than 5% with 30% core coolant flow change, in the form of increased discharge exposure by using the SSR compared with the conventional water rod use. (authors)

  14. Analysis of dynamical flow structure in a square arrayed rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Ikeno, Tsutomu, E-mail: t-ikeno@nfi.co.j [Nuclear Fuel Industries, Ltd., 950, Asashiro Nishi 1-Chome, Kumatori-Cho, Sennan-Gun, Osaka 590-0481 (Japan); Kajishima, Takeo, E-mail: kajisima@mech.eng.osaka-u.ac.j [Department of Mechanical Engineering, Osaka University (Japan)

    2010-02-15

    Large eddy simulation (LES) of turbulent flow in a bare rod bundle was performed, and a new concept about the flow structure that enhances heat transport between subchannels was proposed. To investigate the geometrical effect, the LES was performed for three different values of rod diameter over pitch ratio (D/P = 0.7, 0.8, 0.9). The computational domain containing 4 subchannels was large enough to capture large-scale structures wide across subchannels. Lateral flow obtained was unconfined in a subchannel, and some flows indicated a pulsation through the rod gap between subchannels. The gap flow became strong as D/P increased, as existing experimental studies had reported. Turbulence intensity profile in the rod gap suggested that the pulsation was caused by the turbulence energy transferred from the main flow to the wall-tangential direction. This implied that the flow pulsation was an unsteady mode of the secondary flow and arose from the same geometrical effect of turbulence. This implication was supported by the analysis results: two-points correlation functions of fluctuating velocities indicated two length-scales, P-D and D, respectively of cross-sectional and longitudinal motions; turbulence stress in the cross-sectional mean flow contained a non-potential component, which represented energy injection through the unsteady longitudinal fluid motion.

  15. Nano-mechanical characterization of tension-sensitive helix bundles in talin rod.

    Science.gov (United States)

    Maki, Koichiro; Nakao, Nobuhiko; Adachi, Taiji

    2017-03-04

    Tension-induced exposure of a cryptic signaling binding site is one of the most fundamental mechanisms in molecular mechanotransduction. Helix bundles in rod domains of talin, a tension-sensing protein at focal adhesions, unfurl under tension to expose cryptic vinculin binding sites. Although the difference in their mechanical stabilities would determine which helix bundle is tension-sensitive, their respective mechanical behaviors under tension have not been characterized. In this study, we evaluated the mechanical behaviors of residues 486-654 and 754-889 of talin, which form helix bundles with low and high tension-sensitivity, by employing AFM nano-tensile testing. As a result, residues 754-889 exhibited lower unfolding energy for complete unfolding than residues 486-654. In addition, we found that residues 754-889 transition into intermediate conformations under lower tension than residues 486-654. Furthermore, residues 754-889 showed shorter persistence length in the intermediate conformation than residues 486-654, suggesting that residues 754-889 under tension exhibit separated α-helices, while residues 486-654 assume a compact conformation with inter-helix interactions. Therefore, we suggest that residues 754-889 of talin work as a tension-sensitive domain to recruit vinculin at the early stage of focal adhesion development, while residues 486-654 contribute to rather robust tension-sensitivity by recruiting vinculin under high tension.

  16. Nuclear thermionic converter. [tungsten-thorium oxide rods

    Science.gov (United States)

    Phillips, W. M.; Mondt, J. F. (Inventor)

    1977-01-01

    Efficient nuclear reactor thermionic converter units are described which can be constructed at low cost and assembled in a reactor which requires a minimum of fuel. Each converter unit utilizes an emitter rod with a fluted exterior, several fuel passages located in the bulges that are formed in the rod between the flutes, and a collector receiving passage formed through the center of the rod. An array of rods is closely packed in an interfitting arrangement, with the bulges of the rods received in the recesses formed between the bulges of other rods, thereby closely packing the nuclear fuel. The rods are constructed of a mixture of tungsten and thorium oxide to provide high power output, high efficiency, high strength, and good machinability.

  17. Experimental study on convective heat transfer coefficient around a vertical hexagonal rod bundle

    Science.gov (United States)

    Makhmalbaf, M. H. M.

    2012-06-01

    Research on convective heat transfer coefficient around a rod bundle has many diverse applications in industry. So far, many studies have been conducted in correlations related to internal and turbulent fully-developed flow. Comparison shows that Dittus-Boelter, Sieder-Tate and Petukhov have so far been the most practical correlations in fully-developed turbulent fluid flow heat transfer. The present study conducts an experimental examination of the validity of these frequently-applied correlations and introduces a manufactured test facility as well. Due to its generalizibility, the unique geometry of this test facility (hexagonal arranged, 7 vertical rods in a hexagonal tube) can fulfil extensive applications. The paper also studies the major deviation sources in data measurements, calibrations and turbulence of fluid flow in this. Finally, regarding to sufficient number of experiments in a vast fluid mean velocity range (3,800 < Re < 40,000), a new curve and correlation are presented and the results are compared with the above mentioned commonly-applied correlations.

  18. Development of Design Technology on Thermal-Hydraulic Performance in Tight-Lattice Rod Bundles: II - Rod Bowing Effect on Boiling Transition under Transient Conditions

    Science.gov (United States)

    Liu, Wei; Tamai, Hidesada; Kureta, Masatoshi; Ohnuki, Akira; Akimoto, Hajime

    A thermal-hydraulic feasibility project for an Innovative Water Reactor for Flexible fuel cycle (FLWR) has been performed since 2002. In this R&D project, large-scale thermal-hydraulic tests, several model experiments and development of advanced numerical analysis codes have been carried out. In this paper, we describe the critical power characteristics in a 37-rod tight-lattice bundle with rod bowing under transient states. It is observed that transient Boiling Transition (BT) always occurs axially at exit elevation of upper high-heat-flux region and transversely in the central area of the bundle, which is same as that under steady state. For the postulated power increase and flow decrease cases that may be possibly met in a normal operation of the FLWR, it is confirmed that no BT occurs when Initial Critical Power Ratio (ICPR) is 1.3. Moreover, when the transients are run under severer ICPR that causes BT, the transient critical powers are generally same as the steady ones. The experiments are analyzed with a modified TRAC-BFI code, where Japan Atomic Energy Agency (JAEA) newest critical power correlation is implemented for the BT judgement. The code shows good prediction for the occurrence or the non occurrence of the BT and predicts the BT starting time conservatively. Traditional quasi-steady state prediction of the transient BT is confirmed being applicable for the postulated abnormal transient processes in the tight-lattice bundle with rod bowing.

  19. Reflooding and boil-off experiments in a VVER-440 like rod bundle and analyses with the CATHARE code

    Energy Technology Data Exchange (ETDEWEB)

    Korteniemi, V.; Haapalehto, T. [Lappeenranta Univ. of Technology (Finland); Puustinen, M. [VTT Energy, Lappeenranta (Finland)

    1995-09-01

    Several experiments were performed with the VEERA facility to simulate reflooding and boil-off phenomena in a VVER-440 like rod bundle. The objective of these experiments was to get experience of a full-scale bundle behavior and to create a database for verification of VVER type core models used with modern thermal-hydraulic codes. The VEERA facility used in the experiments is a scaled-down model of the Russian VVER-440 type pressurized water reactors used in Loviisa, Finland. The test section of the facility consists of one full-scale copy of a VVER-440 reactor rod bundle with 126 full-length electrically heated rod simulators. Bottom and top-down reflooding, different modes of emergency core cooling (ECC) injection and the effect of heating power on the heat-up of the rods was studied. In this paper the results of calculations simulating two reflood and one boil-off experiment with the French CATHARE2 thermal-hydraulic code are also presented. Especially the performance of the recently implemented top-down reflood model of the code was studied.

  20. Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang-Young; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests.

  1. A Validation of Subchannel Based CHF Prediction Model for Rod Bundles

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae-Hyun; Kim, Seong-Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    A large number of CHF data base were procured from various sources which included square and non-square lattice test bundles. CHF prediction accuracy was evaluated for various models including CHF lookup table method, empirical correlations, and phenomenological DNB models. The parametric effect of the mass velocity and unheated wall has been investigated from the experimental result, and incorporated into the development of local parameter CHF correlation applicable to APWR conditions. According to the CHF design criterion, the CHF should not occur at the hottest rod in the reactor core during normal operation and anticipated operational occurrences with at least a 95% probability at a 95% confidence level. This is accomplished by assuring that the minimum DNBR (Departure from Nucleate Boiling Ratio) in the reactor core is greater than the limit DNBR which accounts for the accuracy of CHF prediction model. The limit DNBR can be determined from the inverse of the lower tolerance limit of M/P that is evaluated from the measured-to-predicted CHF ratios for the relevant CHF data base. It is important to evaluate an adequacy of the CHF prediction model for application to the actual reactor core conditions. Validation of CHF prediction model provides the degree of accuracy inferred from the comparison of solution and data. To achieve a required accuracy for the CHF prediction model, it may be necessary to calibrate the model parameters by employing the validation results. If the accuracy of the model is acceptable, then it is applied to the real complex system with the inferred accuracy of the model. In a conventional approach, the accuracy of CHF prediction model was evaluated from the M/P statistics for relevant CHF data base, which was evaluated by comparing the nominal values of the predicted and measured CHFs. The experimental uncertainty for the CHF data was not considered in this approach to determine the limit DNBR. When a subchannel based CHF prediction model

  2. Investigation of control rod worth and nuclear end of life of BWR control rods

    Energy Technology Data Exchange (ETDEWEB)

    Magnusson, Per

    2008-01-15

    This work has investigated the Control Rod Worth (CRW) and Nuclear End of Life (NEOL) values for BWR control rods. A study of how different parameters affect NEOL was performed with the transport code PHOENIX4. It was found that NEOL, expressed in terms of {sup 10}B depletion, can be generalized beyond the conditions for which the rod is depleted, such as different power densities and void fractions, the corresponding variation in the NEOL will be about 0.2-0.4% {sup 10}B. It was also found that NEOL results for different fuel types and different fuel enrichments have a variation of about 2-3% in {sup 10}B depletion. A comparative study on NHOL and CRW was made between PHOENIX4 and the stochastic Monte Carlo code MCNP. It was found that there is a significant difference, both due to differences in the codes and to limitations in the geometrical modeling in PHOENIX4. Since MCNP is considered more physically correct, a methodology was developed to calculate the nuclear end of life of BWR control rods with MCNP. The advantages of the methodology are that it does not require other codes to perform the depletion of the absorber material, it can describe control rods of any design and it can deplete the control rod absorber material without burning the fuel. The disadvantage of the method is that is it time-consuming.

  3. Application of fast neutron radiography to three-dimensional visualization of steady two-phase flow in a rod bundle

    CERN Document Server

    Takenaka, N; Fujii, T; Mizubata, M; Yoshii, K

    1999-01-01

    Three-dimensional void fraction distribution of air-water two-phase flow in a 4x4 rod-bundle near a spacer was visualized by fast neutron radiography using a CT method. One-dimensional cross sectional averaged void fraction distribution was also calculated. The behaviors of low void fraction (thick water) two-phase flow in the rod bundle around the spacer were clearly visualized. It was shown that the void fraction distributions were visualized with a quality similar to those by thermal neutron radiography for low void fraction two-phase flow which is difficult to visualize by thermal neutron radiography. It is concluded that the fast neutron radiography is efficiently applicable to two-phase flow studies.

  4. Double-clad nuclear-fuel safety rod

    Science.gov (United States)

    McCarthy, W.H.; Atcheson, D.B.

    1981-12-30

    A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

  5. CFD modelling of supercritical water flow and heat transfer in a 2 × 2 fuel rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Podila, Krishna, E-mail: krishna.podila@cnl.ca; Rao, Yanfei, E-mail: yanfei.rao@cnl.ca

    2016-05-15

    Highlights: • Bare and wire wrapped 2 × 2 fuel rod bundles were modelled with CFD. • Sensitivity of predictions to SST k–ω, v{sup 2}–f and turbulent Prandtl number was tested. • CFD predictions were assessed with experimentally reported fuel wall temperatures. - Abstract: In the present assessment of the CFD code, two heat transfer experiments using water at supercritical pressures were selected: a 2 × 2 rod bare bundle; and a 2 × 2 rod wire-wrapped bundle. A systematic 3D CFD study of the fluid flow and heat transfer at supercritical pressures for the rod bundle geometries was performed with the key parameter being the fuel rod wall temperature. The sensitivity of the prediction to the steady RANS turbulence models of SST k–ω, v{sup 2}–f and turbulent Prandtl number (Pr{sub t}) was tested to ensure the reliability of the predicted wall temperature obtained for the current analysis. Using the appropriate turbulence model based on the sensitivity analysis, the mesh refinement, or the grid convergence, was performed for the two geometries. Following the above sensitivity analyses and mesh refinements, the recommended CFD model was then assessed against the measurements from the two experiments. It was found that the CFD model adopted in the current work was able to qualitatively capture the trends reported by the experiments but the degree of temperature rise along the heated length was underpredicted. Moreover, the applicability of turbulence models varied case-by-case and the performance evaluation of the turbulence models was primarily based on its ability to predict the experimentally reported fuel wall temperatures. Of the two turbulence models tested, the SST k–ω was found to be better at capturing the measurements at pseudo-critical and supercritical test conditions, whereas the v{sup 2}–f performed better at sub-critical test conditions. Along with the appropriate turbulence model, CFD results were found to be particularly sensitive to

  6. Measurements of Flow Mixing at Subchannels in a Wire-Wrapped 61-Rod Bundle for a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Kim, Hyungmo; Ko, Yung Joo; Choi, Hae Seob; Euh, Dong-Jin; Jeong, Ji-Young; Lee, Hyeong-Yeon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    For a safety analysis in a core thermal design of a sodium-cooled fast reactor (SFR), flow mixing characteristics at subchannels in a wire-wrapped rod bundle are crucial factor for the design code verification and validation. Wrapped wires make a cross flow in a circumference of the fuel rod, and this effect lets flow be mixed. Therefore the sub-channel analysis method is commonly used for thermal hydraulic analysis of a SFR, a wire wrapped sub-channel type. To measure flow mixing characteristics, a wire mesh sensing technique can be useful method. A wire mesh sensor has been traditionally used to measure the void fraction of a two-phase flow field, i.e. gas and liquid. However, the recent reports that the wire mesh sensor can be used successfully to recognize the flow field in liquid phase by injecting a tracing liquid with a different level of electric conductivity. The subchannel flow characteristics analysis method is commonly used for the thermal hydraulic analysis of a SFR, a wire wrapped subchannel type. In this study, mixing experiments were conducted successfully at a hexagonally arrayed 61-pin wire-wrapped fuel rod bundle test section. Wire mesh sensor was used to measure flow mixing characteristics. The developed post-processing method has its own merits, and flow mixing results were reasonable.

  7. Development of a Neutron Radiography Three-Dimensional Computed Tomography System for Void Fraction Measurement of Boiling Flow in Tight Lattice Rod Bundles

    Science.gov (United States)

    Kureta, Masatoshi

    A neutron radiography three-dimensional computed tomography (NR3DCT) system was developed to visualize the void fraction distribution of boiling flow in tight lattice heated-rod bundles. This paper chiefly reports on the data processing and the error estimation method of NR3DCT. Practical γ-ray noise reduction and image correction techniques were studied to improve the reliability of the experimental data. Using the system and a directly heated 14-rod bundle test section, the behavior of boiling flow in a tight lattice rod bundle was clearly visualized. The effect of each data processing step on the result was also discussed. By this development, the three-dimensional vapor distribution of boiling flow in a heated bundle is made clear, and void fraction databases can be provided for verification of a thermal-hydraulic simulation code.

  8. Analysis and generalization of experimental data on heat transfer to supercritical pressure water flow in annular channels and rod bundles

    Science.gov (United States)

    Deev, V. I.; Kharitonov, V. S.; Churkin, A. N.

    2017-02-01

    Experimental data on heat transfer to supercritical pressure water presented at ISSCWR-5, 6, and 7 international symposiums—which took place in 2011-2015 in Canada, China, and Finland—and data printed in recent periodical scientific publications were analyzed. Results of experiments with annular channels and three- and four-rod bundles of heating elements positioned in square or triangular grids were examined. Methodology used for round pipes was applied at generalization of experimental data and establishing of correlations suitable for engineering analysis of heat exchange coefficient in conditions of strongly changing water properties in the near-critical pressure region. Empiric formulas describing normal heat transfer to supercritical pressure water mowing in annular channels and rod bundles were obtained. As compared to existing recommendations, suggested correlations are distinguished by specified dependency of heat exchange coefficient on density of heat flux and mass flow velocity of water near pseudo-critical temperature. Differences between computed values of heat exchange coefficient and experimental data usually do not exceed ±25%. Detailed statistical analysis of deviations between computed and experimental results at different states of supercritical pressure water flow was carried out. Peculiarities of deteriorated heat exchange were considered and their existence boundaries were defined. Experimental results obtained for these regimes were generalized using criteria by J.D. Jackson that take the influence of thermal acceleration and Archimedes forces on heat exchange processes into account. Satisfactory agreement between experimental data on heat exchange at flowing of water in annular channels and rod bundles and data for round pipes was shown.

  9. Measurement of Flow Distribution at Subchannels in a Wire-Wrapped 37-Rod Bundle for a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, HYungmo; Chang, Seokkyu; Lee, Dong Won; Choi, Hae Seob; Euh, Dongjin; Lee, Hyeongyeon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In this SFR type of fuel rod, core subchannels are classified with interior, edge, and corner subchannels. Flow distribution of each subchannel is a crucial factor for the core thermal design, and experimental tests for the design code verification and validation in a temperature limitation analysis were conducted. To verify and validate computer codes for the SFR core thermal design, a hexagonally arrayed 37-pin wire-wrapped fuel rod bundle test section was fabricated. The measurement experiments were conducted using a well- designed test loop and iso-kinetic sampling probe. The developed iso-kinetic sampling method in the present study has its own merits, and flow rate results by sampling showed in good agreement with the preliminary CFD analysis results. In addition, the estimated mass balance error was only about 3% in the experiments. Therefore, the present methodology and results can be used in future experiments for design code verification and validation.

  10. Laboratory manual for static pressure drop experiments in LMFBR wire wrapped rod bundles

    Energy Technology Data Exchange (ETDEWEB)

    Burns, K.J.; Todreas, N.E.

    1980-07-01

    Purpose of this experiment is to determine both interior and edge subchannel axial pressure drops for a range of Reynolds numbers. The subchannel static pressure drop is used to calculate subchannel and bundle average friction factors, which can be used to verify existing friction factor correlations. The correlations for subchannel friction factors are used as input to computer codes which solve the coupled energy, continuity, and momentum equations, and are also used to develop flow split correlations which are needed as input to codes which solve only the energy equation. The bundle average friction factor is used to calculate the overall bundle pressure drop, which determines the required pumping power.

  11. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients

    Science.gov (United States)

    Todreas, N. E.; Cheng, S. K.; Basehore, K.

    1984-08-01

    The thermal hydraulic performance of wire wrapped fuel bundles of LMFBR configuration was investigated. Results obtained included phenomenological models for friction factors, flow split and mixing characteristics; correlations for predicting these characteristics suitable for insertion in design codes; numerical codes for analyzing bundle behavior both of the lumped subchannel and distributed parameter categories and experimental techniques for pressure velocity, flow split, salt conductivity and temperature measurement in water cooled mockups of bundles and subchannels. Flow regimes investigated included laminar, transition and turbulent flow under forced convection and mixed convection conditions. Forced convections conditions are emphasized. Outlet plenum behavior is also investigated.

  12. Large-eddy simulations of turbulent flow for grid-to-rod fretting in nuclear reactors

    CERN Document Server

    Bakosi, J; Lowrie, R B; Pritchett-Sheats, L A; Nourgaliev, R R

    2013-01-01

    The grid-to-rod fretting (GTRF) problem in pressurized water reactors is a flow-induced vibration problem that results in wear and failure of the fuel rods in nuclear assemblies. In order to understand the fluid dynamics of GTRF and to build an archival database of turbulence statistics for various configurations, implicit large-eddy simulations of time-dependent single-phase turbulent flow have been performed in 3x3 and 5x5 rod bundles with a single grid spacer. To assess the computational mesh and resolution requirements, a method for quantitative assessment of unstructured meshes with no-slip walls is described. The calculations have been carried out using Hydra-TH, a thermal-hydraulics code developed at Los Alamos for the Consortium for Advanced Simulation of Light water reactors, a United States Department of Energy Innovation Hub. Hydra-TH uses a second-order implicit incremental projection method to solve the single-phase incompressible Navier-Stokes equations. The simulations explicitly resolve the la...

  13. Development for analysis system of rods enrichment of nuclear fuels; Desarrollo de un sistema de analisis de enriquecimiento de barras de combustible nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Rojas C, E.L

    1998-11-01

    Nuclear industry is strongly regulated all over the world and quality assurance is important in every nuclear installation or process related with it. Nuclear fuel manufacture is not the exception. ININ was committed to manufacture four nuclear fuel bundles for the CFE nucleo electric station at Laguna Verde, Veracruz, under General Electric specifications and fulfilling all the requirements of this industry. One of the quality control requisites in nuclear fuel manufacture deals with the enrichment of the pellets inside the fuel bundle rods. To achieve the quality demanded in this aspect, the system described in this work was developed. With this system, developed at ININ it is possible to detect enrichment spikes since 0.4 % in a column of pellets with a 95 % confidence interval and to identify enrichment differences greater than 0.2 % e between homogeneous segments, also with a 95 % confidence interval. ININ delivered the four nuclear fuel bundles to CFE and these were introduced in the core of the nuclear reactor of Unit 1 in the fifth cycle. Nowadays they are producing energy and have shown a correct mechanical performance and neutronic behavior. (Author)

  14. Effects of axial power shapes on CHF locations in a single tube and in rod bundle assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Han, B.; Yang, B.W.; Zhang, H.; Zha, Y.; Zhang, Y. [Xi' an Jiaotong Univ. (China). School of Nuclear Science and Technology

    2016-07-15

    Currently, the prediction of rod bundle CHF is dependent on CHF correlations derived from CHF data. A simple correction factor, such as F-factor, is often used to account for the axial power shape differences based on a simple accumulated energy concept, which has totally no consideration on the impact of true local condition on CHF mechanism. Subsequently, as expected, large uncertainty is often associated with the CHF value and CHF location predictions. For the purpose of obtaining different power shapes effects on CHF, CFD calculated parameter values were used to predict the possible CHF occurrence location. The possible CHF location prediction method proposed in this paper is calculated void fraction, heat transfer coefficient (HTC), liquid temperature distribution and detailed local parameters. And the uniform and non-uniform CHF were analyzed. The prediction of possible CHF locations in a 5 x 5 rod bundle may provide useful information for the design of a full-length CHF test, enhance the accuracy of CHF and CHF location prediction, and reduce the costs of the experimentation.

  15. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Todreas, N.E.; Cheng, S.K.; Basehore, K.

    1984-08-01

    This project principally undertook the investigation of the thermal hydraulic performance of wire wrapped fuel bundles of LMFBR configuration. Results obtained included phenomenological models for friction factors, flow split and mixing characteristics; correlations for predicting these characteristics suitable for insertion in design codes; numerical codes for analyzing bundle behavior both of the lumped subchannel and distributed parameter categories and experimental techniques for pressure velocity, flow split, salt conductivity and temperature measurement in water cooled mockups of bundles and subchannels. Flow regimes investigated included laminar, transition and turbulent flow under forced convection and mixed convection conditions. Forced convections conditions were emphasized. Continuing efforts are underway at MIT to complete the investigation of the mixed convection regime initiated here. A number of investigations on outlet plenum behavior were also made. The reports of these investigations are identified.

  16. Measurements of the flow profile by means of the PIV process at a rod bundle; Stroemungsprofilmessungen mittels PIV-Verfahren an einem Stabbuendel

    Energy Technology Data Exchange (ETDEWEB)

    Franz, R.; Dominguez-Ontiveiro, E.; Barth, T.; Drapeau-Martin, S.; Hampel, U.

    2013-06-01

    Overflowed rod bundles can be used as a heat exchanger in many applications. With respect to safety aspects, the transition from nucleate boiling to film boiling at fuel assemblies in light water reactors is to be avoided. Under this aspect, the numerical flow simulation models for the description of boiling phenomenons are developed. In order to validate these models experimentally, a flow channel is constructed in which a vertical rod bundle is overflowed vertically by the refrigerant RC318 (octafluorocyclobutane). The contribution under consideration describes the test facility and measurement methodology, the process of evaluation, relevant results and error analysis.

  17. Decontamination of control rod housing from Palisades Nuclear Power Station.

    Energy Technology Data Exchange (ETDEWEB)

    Kaminski, M.D.; Nunez, L.; Purohit, A.

    1999-05-03

    Argonne National Laboratory has developed a novel decontamination solvent for removing oxide scales formed on ferrous metals typical of nuclear reactor piping. The decontamination process is based on the properties of the diphosphonic acids (specifically 1-hydroxyethane-1,1-diphosphonic acid or HEDPA) coupled with strong reducing-agents (e.g., sodium formaldehyde sulfoxylate, SFS, and hydroxylamine nitrate, HAN). To study this solvent further, ANL has solicited actual stainless steel piping material that has been recently removed from an operating nuclear reactor. On March 3, 1999 ANL received segments of control rod housing from Consumers Energy's Palisades Nuclear Plant (Covert, MI) containing radioactive contamination from both neutron activation and surface scale deposits. Palisades Power plant is a PWR type nuclear generating plant. A total of eight segments were received. These segments were from control rod housing that was in service for about 6.5 years. Of the eight pieces that were received two were chosen for our experimentation--small pieces labeled Piece A and Piece B. The wetted surfaces (with the reactor's pressurized water coolant/moderator) of the pieces were covered with as a scale that is best characterized visually as a smooth, shiny, adherent, and black/brown in color type oxide covering. This tenacious oxide could not be scratched or removed except by aggressive mechanical means (e.g., filing, cutting).

  18. Method and apparatus for monitoring the control rods of a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gravelle, A.; Marini, J.; Romy, D.

    1984-12-04

    Method and apparatus for monitoring the movement of the control rods of a nuclear reactor. The number of steps of movement in either direction of the rod from which the control rod is suspended is counted. According to the height of the step, an indication of the position of the suspension rod and of the control rod. The apparatus comprises devices for measuring the speed of movement of the control rod, for logging variations in speed higher than a given value, and for counting such variations according to their sign. The invention is particularly useful in pressurized water nuclear reactors.

  19. Non-destructive Testing Dummy Nuclear Fuel Rods by Neutron Radiography

    Institute of Scientific and Technical Information of China (English)

    WEI; Guo-hai; HAN; Song-bai; HE; Lin-feng; WANG; Yu; WANG; Hong-li; LIU; Yun-tao; CHEN; Dong-feng

    2013-01-01

    As a unique non-destructive testing technique,neutron radiography can be used to measure nuclear fuel rods with radioactivity.The images of the dummy nuclear fuel rods were obtained at the CARR.Through imaging analysis methods,the structure defections,the hydrogen accumulation in the cladding and the 235U enrichment of the pellet were studied and analyzed.Experiences for non-destructive testing real PWR nuclear fuel rods by NR

  20. CFD simulation of turbulent flow in a rod bundle with spacer grids (MATIS-H) using STAR-CCM+

    Energy Technology Data Exchange (ETDEWEB)

    Cinosi, N., E-mail: n.cinosi@imperial.ac.uk; Walker, S.P.; Bluck, M.J.; Issa, R.

    2014-11-15

    Highlights: • CDF simulation of turbulent flow generated by a typical PWR spacer grid. • Benchmarking against the MATIS-H experiments run at KAERI in Daejeon, Korea. • Deployment of various steady RANS models to compute the turbulence. • Sensitivity analysis of hardware components. - Abstract: This paper presents the CFD simulation of the turbulent flow generated by a model PWR spacer grid within a rod bundle. The investigation was part of the MATIS-H benchmark exercise, organized by the OECD-NEA, with measurements performed at the KAERI facilities in Daejeon, Korea. The study employed the CD-Adapco code Star-CCM+. An initial sensitivity study was conducted to attempt to assess the importance to the overall flow of components such as the outlet plenum and the end support grid; these were shown to be able to be safely neglected, but the tapered end portion of the rods was found to be significant, and this was incorporated in the model analyzed. A RANS model using any of K-epsilon, K-omega and Reynolds-stress turbulence models was found to be adequate for the prediction of mean velocity profiles, but they all three underestimate the time-averaged turbulent velocity components. Vorticity seems to be better predicted, although the measured values of vorticity are only presented via colored contour plots, making quantitative comparison rather difficult. Circulation, calculated via an integral for each channel, seems to be well predicted by all three models.

  1. CFD analysis of pressure drop across grid spacers in rod bundles compared to correlations and heavy liquid metal experimental data

    Energy Technology Data Exchange (ETDEWEB)

    Batta, A., E-mail: batta@kit.edu; Class, A.G., E-mail: class@kit.edu

    2017-02-15

    Early studies of the flow in rod bundles with spacer grids suggest that the pressure drop can be decomposed in contributions due to flow area variations by spacer grids and frictional losses along the rods. For these shape and frictional losses simple correlations based on theoretical and experimental data have been proposed. In the OECD benchmark study LACANES it was observed that correlations could well describe the flow behavior of the heavy liquid metal loop including a rod bundle with the exception of the core region, where different experts chose different pressure-loss correlations for the losses due to spacer grids. Here, RANS–CFD simulations provided very good data compared to the experimental data. It was observed that the most commonly applied Rehme correlation underestimated the shape losses. The available correlations relate the pressure drop across a grid spacer to the relative plugging of the spacer i.e. solidity e{sub max}. More sophisticated correlations distinct between spacer grids with round or sharp leading edge shape. The purpose of this study is to (i) show that CFD is suitable to predict pressure drop across spacer grids and (ii) to access the generality of pressure drop correlations. By verification and validation of CFD results against experimental data obtained in KALLA we show (i). The generality (ii) is challenged by considering three cases which yield identical pressure drop in the correlations. First we test the effect of surface roughness, a parameter not present in the correlations. Here we compare a simulation assuming a typical surface roughness representing the experimental situation to a perfectly smooth spacer surface. Second we reverse the flow direction for the spacer grid employed in the experiments which is asymmetric. The flow direction reversal is chosen for convenience, since an asymmetric spacer grid with given blockage ratio, may result in different flow situations depending on flow direction. Obviously blockage

  2. CFD simulations in heavy liquid metal flows for square lattice bare rod bundle geometries with a four parameter heat transfer turbulence model

    Energy Technology Data Exchange (ETDEWEB)

    Manservisi, Sandro, E-mail: sandro.manservisi@unibo.it; Menghini, Filippo, E-mail: filippo.menghini3@unibo.it

    2015-12-15

    Highlights: • Turbulent heat transfer with a κ–ϵ–κ{sub θ}–ϵ{sub θ} turbulence model is investigated. • Numerical simulations with different pitch-to-diameter ratios are performed. • The results are compared with SED model and a few available experimental correlations. - Abstract: The study of heat transfer in heavy liquid metals has gained more attention in the last several years due to their applications in new advanced nuclear reactors. These fluids are characterized by low Prandtl numbers and a peculiar heat transfer that cannot be accurately reproduced with standard turbulence approximations, such as the Simple Eddy Diffusivity model (SED), commonly used in commercial codes. In this paper we report the results obtained for the SED and a more advanced κ–ϵ–κ{sub θ}–ϵ{sub θ} four parameter turbulence model for simulations in square lattice bare rod bundle geometries with different pitch-to-diameter ratios. We compare these numerical results with the available experimental data and correlations for the prediction of the Nusselt number.

  3. Preliminary Study on Method of Quantitative Measurement of Nuclear Fuel Rod by Neutron CT at CARR

    Institute of Scientific and Technical Information of China (English)

    WEI; Guo-hai; HAN; Song-bai; WANG; Hong-li; HE; Lin-feng; WANG; Yu; WU; Mei-mei; LIU; Yun-tao; CHEN; Dong-feng

    2015-01-01

    Neutron CT technique was applied to the quantitative measurement of the key parameters of nuclear fuel rods at China Advanced Research Reactor(CARR).The sample of dummy nuclear fuel rod was rotated in 180°range,and 900neutron projections were obtained.The 3-D neutron

  4. From discovery to recognition of periodic large scale vortices in rod bundles as source of natural mixing between subchannels-A review

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, Leonhard, E-mail: leonhard.meyer@kit.ed [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies, Postfach 3640, 76021 Karlsruhe (Germany)

    2010-06-15

    The mixing of cooling fluid in rod bundles from one subchannel to another through the gaps between the rods reduces the temperature differences in the coolant as well as along the perimeter of the rods. The phenomenon of natural mixing was first intensively investigated in the 1960s and remains a topic of research up to the present time. The paper describes the main stations on the way to understand the nature of the flow in rod bundles and generally in compound channels with the focus on work performed at Research Center Karlsruhe (FZK). Earlier, it was noticed that the mixing rates where higher than could be accounted for by turbulent diffusion alone. For more than 20 years attempts were made to prove experimentally and by code application that secondary flows could account for the measured mixing rates, although the measured secondary flow velocities were much too low. Measurements of the turbulence structure by hot wire anemometry confirmed the existence of cyclic flow pulsations, which had been postulated earlier on the basis of thermocouple measurements. More sophisticated hot wire measurements revealed the nature of these pulsations as periodic, coupled to gap width and Reynolds number. Finally, the extension of the investigation to other compound channel types and flow visualization revealed the true nature of the mixing process as a vortex train moving along the gap between rods or in the narrow part of a compound channel. These findings have been confirmed by LES calculations. Based on these results CFD codes with improved turbulence models calculated successfully the flow in rod bundles including the macroscopic oscillations.

  5. Development of Design Technology on Thermal-Hydraulic Performance in Tight-Lattice Rod Bundles: I-Master Plan and Executive Summary

    Science.gov (United States)

    Ohnuki, Akira; Kureta, Masatoshi; Yoshida, Hiroyuki; Tamai, Hidesada; Liu, Wei; Misawa, Takeharu; Takase, Kazuyuki; Akimoto, Hajime

    R&D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Innovative Water Reactor for Flexible Fuel Cycle has been progressed at Japan Atomic Energy Agency in collaboration with power utilities, reactor vendors and universities since 2002. In this series-study, we will summarize the R&D achievements using large-scale test facility (37-rod bundle with full-height and full-pressure), model experiments and advanced numerical simulation technology. This first paper described the master plan for the development of design technology and showed an executive summary for this project up to FY2005. The thermal-hydraulic characteristics in the tight-lattice configuration were investigated and the feasibility was confirmed based on the experiments. We have developed the design technology including 3-D numerical simulation one to evaluate the effects of geometry/scale on the thermal-hydraulic behaviors.

  6. Development of Design Technology on Thermal-hydraulic Performance in Tight-lattice Rod Bundles: V-Estimation of Void Fraction

    Science.gov (United States)

    Kureta, Masatoshi; Tamai, Hidesada; Yoshida, Hiroyuki; Ohnuki, Akira; Akimoto, Hajime

    An estimation of the void fraction in a tight-lattice rod bundle was needed for the R&D of the Innovative Water Reactor for Flexible Fuel Cycle (FLWR). For this purpose, we measured the void fraction and studied the behaviors of boiling flow. The void fraction was measured by a neutron radiography, a quick-shut-valve technique, and an electro void fraction meter. The data were taken using the 7-, 14-, 19- and 37-rod bundle test sections with the rod gap of 1.0 or 1.3 mm under from atmospheric pressure to 7.2 MPa conditions. A spacer effect test was also carried out. The following estimations were conducted: (1) a similarity of the advanced analysis codes with the 3D void fraction data, (2) the comparisons of the TRAC-BF1 code and a drift-flux model with the 1D data. Followings were made clear: (a) The void fraction becomes lower at the peripheral and higher at the rod gap part of the lower core and at the center of the subchannel of the upper core, (b) the codes calculates the similar distribution to the data, and (c) the TRAC-BF1 and the drift-flux model tends to overestimate the void fraction at the lower quality region, on the other hand at the higher quality, those methods tend to same characteristics to the data. It was confirmed that several special features were existed in the tight-lattice rod bundle but the codes were applicable.

  7. Results of automatic system implementation for the friction control rods execution in Cofrentes nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Curiel, M. [Logistica y Acondicionamientos Industriales SAU, Sorolla Center, local 10, Av. de las Cortes Valencianas, 46015 Valencia (Spain); Palomo, M. J. [ISIRYM, Universidad Politecnica de Valencia, Camino de Vera s/n, Valencia (Spain); Urrea, M. [Iberdrola Generacion S. A., Central Nuclear Cofrentes, Carretera Almansa Requena s/n, 04662 Cofrentes, Valencia (Spain); Arnaldos, A., E-mail: m.curiel@lainsa.co [TITANIA Servicios Tecnologicos SL, Sorolla Center, local 10, Av. de las Cortes Valencianas No. 58, 46015 Valencia (Spain)

    2010-10-15

    The purpose of this presentation is to show the obtained results in Cofrentes nuclear power plant (Spain) of control rods Pcc/24 friction test procedure. In order to perform this, a control rod friction test system has been developed. Principally, this system consists on software and data acquisition hardware that obtains and analyzes the control rod pressure variation on which the test is being made. The Pcc/24 procedure objective is to detect an excessive friction in the control rod movement that could cause a control rod drive movement slower than usual. This test is necessary every time that an anomalous alteration is produced in the reactor core that could affect to a fuel rod, and it is executed before the time measure of control rods rapid scram test of the affected rods. This test has to be carried out to all the reactor control rods and takes valuable time during plant refuelling. So, by means of an automatic system to perform the test, we obtain an important time saving during refuelling. On the other hand, the on-line monitoring of the control rod insertion and changes in differential pressure, permits a control rod operation fast and safe validation. Moreover, an automatic individual report of every rod is generated by the system and a final global result report of the entire test developed in refuelling is generated. The mentioned reports can be attached directly to the procedure documents obtaining an office data processing important saving time. (Author)

  8. Results of automatic system implementation for the friction control rods execution in Cofrentes nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Palomo, M., E-mail: mpalomo@iqn.upv.es [Universidad Politecnica de Valencia (UPV) (Spain); Urrea, M., E-mail: matias.urrea@iberdrola.es [Iberdrola Generacion S.A. Valencia (Spain). C.N. Cofrentes; Curiel, M., E-mail: m.curiel@lainsa.com [Logistica y Acondicionamientos Industriales (LAINSA), Valencia (Spain); Arnaldos, A., E-mail: a.arnaldos@titaniast.com [TITANIA Servicios Teconologicos, Valencia (Spain)

    2011-07-01

    The purpose of this presentation is to show the obtained results in Cofrentes Nuclear Power Plant (Spain) of Control Rods PCC/24 Friction Test Procedure. In order to perform this, a Control Rod Friction Test System has been developed. Principally, this system consists on software and data acquisition hardware that obtains and analyzes the control rod pressure variation on which the test is being made. The PCC/24 Procedure objective is to detect an excessive friction in the control rod movement that could cause a CRD (Control Rod Drive) movement slower than usual. This test is necessary every time that an anomalous alteration is produced in the reactor core that could affect to a fuel rod, and it is executed before the time measure of control rods rapid scram test of the affected rods. This test has to be carried out to all the reactor control rods and takes valuable time during plant refuelling. So, by means of an automatic system to perform the test, we obtain an important time saving during refuelling. On the other hand, the on-line monitoring of the control rod insertion and changes in differential pressure, permits a control rod operation fast and safe validation. Moreover, an automatic individual report of every rod is generated by the system and a final global result report of the entire test developed in refuelling is generated. The mentioned reports can be attached directly to the procedure documents obtaining an office data processing important saving time.(author)

  9. GMDH-type neural network modeling and genetic algorithm-based multi-objective optimization of thermal and friction characteristics in heat exchanger tubes with wire-rod bundles

    Science.gov (United States)

    Rahimi, Masoud; Beigzadeh, Reza; Parvizi, Mehdi; Eiamsa-ard, Smith

    2016-08-01

    The group method of data handling (GMDH) technique was used to predict heat transfer and friction characteristics in heat exchanger tubes equipped with wire-rod bundles. Nusselt number and friction factor were determined as functions of wire-rod bundle geometric parameters and Reynolds number. The performance of the developed GMDH-type neural networks was found to be superior in comparison with the proposed empirical correlations. For optimization, the genetic algorithm-based multi-objective optimization was applied.

  10. Numerical evaluation of flow through a 5X5 PWR rod bundle: effect of the vane arrangement in a spacer grid

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, Moyses A. [Brazilian Nuclear Energy Commission (CNEN), Belo Horizonte, MG (Brazil)], e-mail: navarro@cdtn.br; Santos, Andre A.C. [Federal University of Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Mechanical Engineering Department], e-mail: acampagnole@yahoo.com.br

    2009-07-01

    Spacer grids along the fuel assembly of Pressurized Water Reactors (PWR) maintain rod bundles arranged in a regular square configuration. The mixing vanes present in the spacer grids promote cross and swirl flow between and within the subchannels, enhancing the heat transfer performance in the grid vicinity, but also causing an adverse increase of the pressure drop in the rod bundle due the constriction on the coolant flow area. Therefore, the thermal hydraulic design of the grid must allow for both low pressure loss and high coolant mixing, which means it is important to optimize the design of the grid in relation to the mixing vane. More recently, Computational Fluid Dynamics (CFD) using three dimensional Reynolds Averaged Navier Stokes (RANS) analysis has been used efficiently as an auxiliary tool in the development of spacer grids. The influence of some geometric characteristics of spacer grids on the flow through a rod bundle have been numerically evaluated and are still a subject of discussion. This work analyses the influence of the vanes arrangement in the spacer grid on the flow through a PWR 5 x 5 rod bundle segment. The Numerical simulations were performed with the commercial code CFX 11.0. To make the simulation possible with a limited computational capacity and acceptable mesh refinement, the computational domain was divided in 7 subdomains. The subdomains were simulated sequentially applying the outlet results of a previous subdomain as inlet condition for the next. In this study the k- turbulence model with scalable wall function was used. Five different vane arrangements were simulated at reactor level power and flow characteristics. The same grid and vane geometry were used in all simulations. The results of this study were divided in two parts. In the first part the presence of peripheral vanes on 5 x 5 rod bundle spacer grid tests were evaluated. The results showed that peripheral vanes should be avoided in experiments and simulations in order to

  11. FDD-1 System On-line Monitoring Fuel Rod Failure of Nuclear Power Plant

    Institute of Scientific and Technical Information of China (English)

    CHENPeng; ZHANGYing-chao; JISong-tao; GAOYong-guang; YINZhen-guo; HANChuan-bin

    2003-01-01

    The FDD-1 system developed by CIAE for on-line monitoring fuel rod failure of nuclear power plant consists of γ-ray detector, γ-ray spectrum analyzer, computer, and an analysis code for evaluating the status of fuel rod failure. It would be determined that the fuel rod failure occurs when a large amount of γ activity increases in the primary system measured by γ-ray detector near the CVCS.

  12. An experimental investigation of supercritical heat transfer in a three-rod bundle equipped with wire-wrap and grid spacers and cooled by carbon dioxide

    Energy Technology Data Exchange (ETDEWEB)

    Eter, Ahmad, E-mail: eng.eter@yahoo.com; Groeneveld, Dé, E-mail: degroeneveld@gmail.com; Tavoularis, Stavros, E-mail: stavros.tavoularis@uottawa.ca

    2016-07-15

    Highlights: • Heat transfer at supercritical pressures was studied experimentally in a three-rod bundle equipped with wire-wrap spacers or grid spacers. • Heat transfer deterioration occurred near the heated inlet under certain conditions. • Normal heat transfer was generally comparable to that in a tube and the predictions of a correlation. - Abstract: Heat transfer measurements in a three-rod bundle equipped with wire-wrap and grid spacers were obtained at supercritical pressures in the Supercritical University of Ottawa Loop (SCUOL). The tests were performed using carbon dioxide, as a surrogate fluid for water, flowing upwards for wide ranges of conditions, including conditions equivalent to the nominal and near-normal operating conditions of the proposed Canadian Super-Critical Water-Cooled Reactor. The test section contained three heated rods and three unheated rod segments with an outer diameter of 10 mm and a pitch-to-diameter ratio of 1.14; the heated length was 1500 mm. Detailed surface temperature measurements along and around the three heated rods were collected using internally traversed thermocouples. The following ranges of test conditions were covered, with equivalent water conditions given inside parentheses: pressure from 6.6 to 8.36 MPa (19.7–25 MPa); inlet temperature from 11 to 30 °C (330–371 °C); mass flux from 200 to 1175 kg m{sup −2} s{sup −1} (340–1822 kg m{sup −2} s{sup −1}); and wall heat flux from 1 to 175 kW m{sup −2} (11–1847 kW m{sup −2}). For one set of tests, the heated rods were fitted with a 1.3 mm OD wire wrap, having an axial pitch of 200 mm along the entire heated length; for a second set, the heated rods were fitted with grid spacers having a 5.3% flow blockage and located at 500 mm axial intervals. The effects of spacer configuration on heat transfer at supercritical pressures were documented and analyzed. The observed experimental trends were compared to those obtained in a experiment in a heated

  13. Development of Welding and Instrumentation Technology for Nuclear Fuel Test Rod

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang Young; Ahn, Sung Ho; Heo, Sung Ho; Hong, Jin Tae; Kim, Ka Hye [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    It is necessary to develop various types of welding, instrumentation and helium gas filling techniques that can conduct TIG spot welding exactly at a pin-hole of the end-cap on the nuclear fuel rod to fill up helium gas. The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rod. To manufacture the nuclear fuel test rod, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rod jointing the various sensors and end-caps on a fuel cladding tube, which is charged with fuel pellets and component parts. We therefore designed and fabricated an orbital TIG welding system and a laser welding system. This paper describes not only some experiment results from weld tests for the parts of a nuclear fuel test rod, but also the contents for the instrumentation process of the dummy fuel test rod installed with the C-type T. C. A dummy nuclear fuel test rod was successfully fabricated with the welding and instrumentation technologies acquired with various tests. In the test results, the round welding has shown a good weldability at both the orbital TIG welding system and the fiber laser welding system. The spot welding to fill up helium gas has shown a good welding performance at a welding current of 30A, welding time of 0.4 sec and gap of 1 mm in a helium gas atmosphere. The soundness of the nuclear fuel test rod sealed by a mechanical sealing method was confirmed by helium leak tests and microstructural analyses.

  14. Evaluation of a numeric procedure for flow simulation of a 5X5 PWR rod bundle with a mixing vane spacer

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, Moyses A. [Brazilian Nuclear Energy Commission (CNEN), Belo Horizonte, MG (Brazil)], e-mail: navarro@cdtn.br; Santos, Andre A.C. [Federal University of Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Mechanical Engineering Department], e-mail: acampagnole@yahoo.com.br

    2009-07-01

    The fuel assemblies of the Pressurized Water Reactors (PWR) are constituted of rod bundles arranged in a regular square configuration by spacer grids placed along its length. The presence of the spacer grids promote two antagonist effects on the core: a desirable increase of the local heat transfer downstream the grids and an adverse increase of the pressure drop due the constriction on the coolant flow area. Most spacer grids are designed with mixing vanes which cause a cross and swirl flow between and within the subchannels, enhancing even more the heat transfer performance in the grid vicinity. The improvement of the heat transfer increases the departure from the nucleate boiling ratio, allowing higher operating power in the reactor. Due to these important thermal and fluid dynamic features, experimental and theoretical investigations have been carried out in the past years for the development of spacer grid design. More recently, the Computational Fluid Dynamics (CFD) using three dimensional Reynolds Averaged Navier Stokes (RANS) analysis has been used efficiently for this purpose. Many computational works have been performed, but the appropriate numerical procedure for the flow in rod bundle simulations is not yet a consensus. This work presents results of flow simulations performed with the commercial code CFX 11.0 in a PWR 5x5 rod bundle segment with a split vane spacer grid. The geometrical configuration and flow conditions used in the experimental studies performed by Karoutas et al. were assumed in the simulations. To make the simulation possible with a limited computational capacity and acceptable mesh refinement, the computational domain was divided in 7 subdomains. The subdomains were simulated sequentially applying the outlet results of a previous subdomain as inlet condition for the next. In this study the {kappa}-{epsilon} turbulence model was used. The simulations were also compared with those performed by Karoutas et al. in half a subchannel and

  15. Zirconium-based alloys, nuclear fuel rods and nuclear reactors including such alloys, and related methods

    Science.gov (United States)

    Mariani, Robert Dominick

    2014-09-09

    Zirconium-based metal alloy compositions comprise zirconium, a first additive in which the permeability of hydrogen decreases with increasing temperatures at least over a temperature range extending from 350.degree. C. to 750.degree. C., and a second additive having a solubility in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. At least one of a solubility of the first additive in the second additive over the temperature range extending from 350.degree. C. to 750.degree. C. and a solubility of the second additive in the first additive over the temperature range extending from 350.degree. C. to 750.degree. C. is higher than the solubility of the second additive in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. Nuclear fuel rods include a cladding material comprising such metal alloy compositions, and nuclear reactors include such fuel rods. Methods are used to fabricate such zirconium-based metal alloy compositions.

  16. Experimental Study on Surrogate Nuclear Fuel Rods under Reversed Cyclic Bending

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Hong [ORNL; Wang, Jy-An John [ORNL

    2017-01-01

    The mechanical behavior of spent nuclear fuel (SNF) rods under reversed cyclic bending or bending fatigue must be understood to evaluate their vibration integrity in a transportation environment. This is especially important for high-burnup fuels (>45 GWd/MTU), which have the potential for increased structural damage. It has been demonstrated that the bending fatigue of SNF rods can be effectively studied using surrogate rods. In this investigation, surrogate rods made of stainless steel (SS) 304 cladding and aluminum oxide pellets were tested under load or moment control at a variety of amplitude levels at 5 Hz using the Cyclic Integrated Reversible-Bending Fatigue Tester developed at Oak Ridge National Laboratory. The behavior of the rods was further characterized using flexural rigidity and hysteresis data, and fractography was performed on the failed rods. The proposed surrogate rods captured many of the characteristics of deformation and failure mode observed in SNF, including the linear-to-nonlinear deformation transition and large residual curvature in static tests, PPI and PCMI failure mechanisms, and large variation in the initial structural condition. Rod degradation was measured and characterized by measuring the flexural rigidity; the degradation of the rigidity depended on both the moment amplitude applied and the initial structural condition of the rods. It was also shown that a cracking initiation site can be located on the internal surface or the external surface of cladding. Finally, fatigue damage to the bending rods can be described in terms of flexural rigidity, and the fatigue life of rods can be predicted once damage model parameters are properly evaluated. The developed experimental approach, test protocol, and analysis method can be used to study the vibration integrity of SNF rods in the future.

  17. Experimental Study of Three-Dimensional Void Fraction Distribution in Heated Tight-Lattice Rod Bundles Using Three-Dimensional Neutron Tomography

    Science.gov (United States)

    Kureta, Masatoshi

    Three-dimensional (3D) void fraction distributions in a tight-lattice of heated 7- or 14-rod bundles were measured using 3D neutron tomography. The distribution was also studied parametrically from the thermal-hydraulic point of view in order to elucidate boiling phenomena in a fuel assembly of the FLWR which is being developed as an advanced BWR-type reactor. 7-rod tests were carried out to obtain high void fraction data. 14-rod tests were conducted for visualization and discussion of the 3D distribution extending from the vapor generation region to the high void fraction region at one time. Experimental data were obtained under atmospheric pressure with mass velocity, heater power and inlet quality as the test parameters. It was found from the visualization of data that the void fraction at the channel center became higher than that at the periphery, high void fraction spots appeared in narrow regions at the inlet, and a so-called 'vapor chimney' was generated at the center of a subchannel.

  18. A study on the nuclear characteristics of enriched gadolinia burnable absorber rods; the final report

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chung Chan; Zee, Quun; Song, J. S.; Cho, B. O.; Joo, H. G.; Park, S. Y.; Kim, H. Y.; Cho, J. Y.; Kim, K. S

    2001-12-01

    An analysis model using MICBURN-3/CASMO-3 is established for the enriched gadolinia burnable absorber rods. A homogenized cross section editing code, PROLOG, is modified so that it can handle such a fuel assembly that includes two different types of gadolinia rods. Study shows that Gd-155 and Gd-157 are almost same characteristics in suppressing the excess reactivity and it is recommended to enrich both odd number isotopes, Gd-155 and Gd-157. To minimize power peakings, various placements of gadolinia rods are tested. Also initial reactivity holddown and gadolinia burnout time are parametrized with respect to the number of gadolinia rods and gadolinia weight fractions. Cycle length increases by 8 days if enriched gadolinia rods are used in the commercial nuclear power plant such as YGN-3 with 2 year-cycle-length. The nuclear characteristics such as soluble boron concentrations and power distributions are almost same. The maximum power of the gadolinia rods are slightly increased but gadolinia contents are reduced. For SMART with ultra longer cycle length and soluble boron-free operation concept cycle length can be extended by maximum 90 days due to increased fuel loading.

  19. Development of nuclear fuel rod inspection technique using ultrasonic resonance phenomenon

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Myung Sun; Lee, Jong Po; Ju, Young Sang [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-11-01

    Acoustic resonance scattering from a nuclear fuel rod in water is analyzed. A new model for the background which is attributed to the interference of reflected wave and diffracted wave is found and here named {sup t}he inherent background{sup .} The resonance spectrum of a fuel rod is obtained by subtracting the inherent background from the scattered pressure. And also analyzed are the effect of material damping of cladding tube and pellet on the resonance spectrum of a fuel rod. The propagation characteristics of circumferential waves which cause the resonances of cladding tube is produced and the appropriate resonance modes for the application to the inspection of assembled fuel rods are selected. The resonance modes are experimentally measured for pre- and post-irradiated fuel rods and the validation of the fuel rod inspection using ultrasonic resonance phenomenon is examined. And thin ultrasonic sensors accessible into the narrow interval (about 2-3mm) between assembled fuel rods are designed and manufactured. 14 refs. (Author).

  20. A feasibility study for the application of enriched gadolinia burnable absorber rods in nuclear core design

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chung Chan; Zee, Sung Quun; Kim, Kang Seog; Song, Jae Seung

    2000-12-01

    An analysis model using MICBURN-3/CASMO-3 is established for the enriched gadolinia burnable absorber rods. A homogenized cross section editing code, PROLOG, is modified so that it can handle such a fuel assembly that includes two different types of gadolinia rods. Study shows that Gd-155 and Gd-157 are almost same in suppressing the excess reactivity and it is recommended to enrich both odd number isotopes, Gd-155 and Gd-157. It is estimated that the cycle length increases by 2 days if enriched gadolinia rods are used in the commercial nuclear power plant such as YGN-3 of which the cycle length is assumed 2 years. For the advanced integral reactor SMART in which ultra long cycle length and soluble boron-free operation concept is applied, natural gadolinia burnable absorber rods fail to control the excess reactivity. On the other hand, enriched gadolinia rods are successful in controling the excess reactivity. To minimize power peakings, various placements of gadolinia rods are tested. Also initial reactivity holddown and gadolinia burnout time are parametrized with respect to the number of gadolinia rods and gadolinia weight fractions.

  1. Uncertainty analysis of spent nuclear fuel isotopics and rod internal pressure

    Science.gov (United States)

    Bratton, Ryan N.

    The bias and uncertainty in fuel isotopic calculations for a well-defined radio- chemical assay benchmark are investigated with Sampler, the new sampling-based uncertainty quantification tool in the SCALE code system. Isotopic predictions are compared to measurements of fuel rod MKP109 of assembly D047 from the Calvert Cliffs Unit 1 core at three axial locations, representing a range of discharged fuel burnups. A methodology is developed which quantifies the significance of input parameter uncertainties and modeling decisions on isotopic prediction by compar- ing to isotopic measurement uncertainties. The SCALE Sampler model of the D047 assembly incorporates input parameter uncertainties for key input data such as multigroup cross sections, decay constants, fission product yields, the cladding thickness, and the power history for fuel rod MKP109. The effects of each set of input parameter uncertainty on the uncertainty of isotopic predictions have been quantified. In this work, isotopic prediction biases are identified and an investiga- tion into their sources is proposed; namely, biases have been identified for certain plutonium, europium, and gadolinium isotopes for all three axial locations. More- over, isotopic prediction uncertainty resulting from only nuclear data is found to be greatest for Eu-154, Gd-154, and Gd-160. The discharge rod internal pressure (RIP) and cladding hoop stress (CHS) distributions are quantified for Watts Bar Nuclear Unit 1 (WBN1) fuel rods by modeling core cycle design data, operation data (including modeling significant trips and downpowers), and as-built fuel enrichments and densities of each fuel rod in FRAPCON-3.5. A methodology is developed which tracks inter-cycle as- sembly movements and assembly batch fabrication information to build individual FRAPCON inputs for each considered WBN1 fuel rod. An alternate model for the amount of helium released from zirconium diboride (ZrB2) integral fuel burn- able absorber (IFBA) layers is

  2. Flow distribution and pressure loss in subchannels of a wire-wrapped 37-pin rod bundle for sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Seok Kyu; Euh, Dong Jin; Choi, Hae Seob; Kim, Hyung Mo; Choi, Sun Rock; Lee, Hyeong Yeon [Thermal-Hydraulic Safety Research Department, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-04-15

    A hexagonally arrayed 37-pin wire-wrapped rod bundle has been chosen to provide the experimental data of the pressure loss and flow rate in subchannels for validating subchannel analysis codes for the sodium-cooled fast reactor core thermal/hydraulic design. The iso-kinetic sampling method has been adopted to measure the flow rate at subchannels, and newly designed sampling probes which preserve the flow area of subchannels have been devised. Experimental tests have been performed at 20-115% of the nominal flow rate and 60 degrees C (equivalent to Re ∼ 37,100) at the inlet of the test rig. The pressure loss data in three measured subchannels were almost identical regardless of the subchannel locations. The flow rate at each type of subchannel was identified and the flow split factors were evaluated from the measured data. The predicted correlations and the computational fluid dynamics results agreed reasonably with the experimental data.

  3. Design of Testing Set-up for Nuclear Fuel Rod by Neutron Radiography at CARR

    Institute of Scientific and Technical Information of China (English)

    WEI; Guo-hai; HAN; Song-bai; WANG; Hong-li; HAO; Li-jie; WU; Mei-mei; HE; Lin-feng; WANG; Yu; LIU; Yun-tao; SUN; Kai; CHEN; Dong-feng

    2012-01-01

    <正>An experimental set-up dedicated to non-destructively test a 15 cm long pressurized water reactor (PWR) nuclear fuel rod by neutron radiography (NR) is designed and fabricated. It consists of three parts: Transport container, imaging block and steel support. The design of the transport container was optimized with Monte-Carlo simulation by the MCNP code.

  4. Results of the first nuclear blowdown test on single fuel rods (LOC-11 Series in PBF)

    Energy Technology Data Exchange (ETDEWEB)

    Larson, J.R.; Evans, D.R.; McCardell, R.K.

    1978-01-01

    This paper presents results of the first nuclear blowdown tests (LOC-11A, LOC-11B, LOC-11C) ever conducted. The Loss-of-Coolant Accident (LOCA) Test Series is being conducted in the Power Burst Facility (PBF) reactor at the Idaho National Engineering Laboratory, near Idaho Falls, Idaho, for the Nuclear Regulatory Commission. The objective of the LOC-11 tests was to obtain data on the behavior of pressurized and unpressurized rods when exposed to a blowdown similar to that expected in a pressurized water reactor (PWR) during a hypothesized double-ended cold-leg break. The data are being used for the development and verification of analytical models that are used to predict coolant and fuel rod pressure during a LOCA in a PWR.

  5. Mathematical modelling of friction-vibration interactions of nuclear fuel rods

    Directory of Open Access Journals (Sweden)

    Zeman V.

    2016-06-01

    Full Text Available Nuclear fuel rods (FRs are transverselly linked to each other by three spacer grid cells at several vertical levels inside a fuel assembly (FA. Vibration of FA components, caused by the motion of FA support plates in the reactor core, generates variable contact forces between FRs and spacer grid cells. Friction effects in contact surfaces have an influence on the expected lifetime period of nuclear FA in terms of FR cladding fretting wear. This paper introduces an original approach to mathematical modelling and simulation analysis of FR nonlinear vibrations and fretting wear taking into consideration friction forces at all levels of spacer grids.

  6. ORNL rod-bundle heat-transfer test data. Volume 6. Thermal-hydraulic test facility experimental data report for test 3. 05. 5B - double-ended cold-leg break simulation

    Energy Technology Data Exchange (ETDEWEB)

    Mullins, C.B.; Felde, D.K.; Sutton, A.G.; Gould, S.S.; Morris, D.G.; Robinson, J.J.; Schwinkendorf, K.N.

    1982-05-18

    Thermal-Hydraulic Test Facility (THTF) Test 3.05.5B was conducted by members of the ORNL PWR Blowdown Heat Transfer Separate-Effects Program on July 3, 1980. The objective of the program is to investigate heat transfer phenomena believed to occur in PWRs during accidents, including small and large break loss-of-coolant accidents. Test 3.05.5B was designed to provide transient thermal-hydraulics data in rod bundle geometry under reactor accident-type conditions. Reduced instrument responses are presented. Also included are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers.

  7. Sturdy on Orbital TIG Welding Properties for Nuclear Fuel Test Rod

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Changyoung; Hong, Jintae; Kim, Kahye; Huh, Sungho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    We developed a precision TIG welding system that is able to weld the seam between end-caps and a fuel cladding tube for the nuclear fuel test rod and rig. This system can be mainly classified into an orbital TIG welder (AMI, M-207A) and a pressure chamber. The orbital TIG welder can be independently used, and it consists of a power supply unit, a microprocessor, water cooling unit, a gas supply unit and an orbital weld head. In this welder, the power supply unit mainly supplies GTAW power for a welding specimen and controls an arc starting of high frequency, supping of purge gas, arc rotation through the orbital TIG welding head, and automatic timing functions. In addition, the pressure chamber is used to make the welded surface of the cladding specimen clean with the inert gas filled inside the chamber. To precisely weld the cladding tube, a welding process needs to establish a schedule program for an orbital TIG welding. Therefore, the weld tests were performed on a cladding tube and dummy rods under various conditions. This paper describes not only test results on parameters of the purge gas flow rates and the chamber gas pressures for the orbital TIG welding, but also test results on the program establishment of an orbital TIG welding system to weld the fuel test rods. Various welding tests were performed to develop the orbital TIG welding techniques for the nuclear fuel test rod. The width of HAZ of a cladding specimen welded with the identical power during an orbital TIG welding cycle was continuously increased from a welded start-point to a weld end-point because of heat accumulation. The welding effect of the PGFR and CGP shows a relatively large difference for FSS and LSS. Each hole on the cladding specimens was formed in the 1bar CGP with the 20L/min PGFR but not made in the case of the PGFR of 10L/min in the CGP of 2bar. The optimum schedule program of the orbital TIG welding system to weld the nuclear fuel test rod was established through the program

  8. FY15 Status Report: CIRFT Testing of Spent Nuclear Fuel Rods from Boiler Water Reactor Limerick

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jiang, Hao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-06-01

    The objective of this project is to perform a systematic study of used nuclear fuel (UNF, also known as spent nuclear fuel [SNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. The additional CIRFT was conducted on three HBR rods (R3, R4, and R5) in which two specimens failed and one specimen was tested to over 2.23 10⁷ cycles without failing. The data analysis on all the HBR UNF rods demonstrated that it is necessary to characterize the fatigue life of the UNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum of tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, ten SNF rod segments from BWR Limerick were tested using ORNL CIRFT, with one under static and nine dynamic loading conditions. Under static unidirectional loading, a moment of 85 N·m was obtained at maximum curvature 4.0 m⁻¹. The specimen did not show any sign of failure in three repeated loading cycles to almost same maximum curvature. Ten cyclic tests were conducted with amplitude varying from 15.2 to 7.1 N·m. Failure was observed in nine of the tested rod specimens. The cycles to failure were

  9. Rod internal pressure of spent nuclear fuel and its effects on cladding degradation during dry storage

    Science.gov (United States)

    Kim, Ju-Seong; Hong, Jong-Dae; Yang, Yong-Sik; Kook, Dong-Hak

    2017-08-01

    Temperature and hoop stress limits have been used to prevent the gross rupture of spent nuclear fuel during dry storage. The stress due to rod internal pressure can induce cladding degradation such as creep, hydride reorientation, and delayed hydride cracking. Creep is a self-limiting phenomenon in a dry storage system; in contrast, hydride reorientation and delayed hydride cracking are potential degradation mechanisms activated at low temperatures when the cladding material is brittle. In this work, a conservative rod internal pressure and corresponding hoop stress were calculated using FRAPCON-4.0 fuel performance code. Based on the hoop stresses during storage, a study on the onset of hydride reorientation and delayed hydride cracking in spent nuclear fuel was conducted under the current storage guidelines. Hydride reorientation is hard to occur in most of the low burn-up fuel while some high burn-up fuel can experience hydride reorientation, but their effect may not be significant. On the other hand, delayed hydride cracking will not occur in spent nuclear fuel from pressurized water reactor; however, there is a lack of confirmatory data on threshold intensity factor for delayed hydride cracking and crack size distribution in the fuel.

  10. Development of Mechanical Sealing and Laser Welding Technology to Instrument Thermocouple for Nuclear Fuel Test Rod

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang-Young; Ahn, Sung-Ho; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Zircaloy-4 of the nuclear fuel test rod, AISI 316L of the mechanical sealing parts, and the MI (mineral insulated) cable at a thermocouple instrumentation are hetero-metals, and are difficult to weld to dissimilar materials. Therefore, a mechanical sealing method to instrument the thermocouple should be conducted using two kinds of sealing process as follows: One is a mechanical sealing process using Swagelok, which is composed of sealing components that consists of an end-cap, a seal tube, a compression ring and a Swagelok nut. The other is a laser welding process used to join a seal tube, and an MI cable, which are made of the same material. The mechanical sealing process should be sealed up with the mechanical contact compressed by the strength forced between a seal tube and an end-cap, and the laser welding process should be conducted to have no defects on the sealing area between a seal tube and an MI cable. Therefore, the mechanical sealing and laser welding techniques need to be developed to accurately measure the centerline temperature of the nuclear fuel test rod in an experimental reactor. The mechanical sealing and laser welding tests were conducted to develop the thermocouple instrumentation techniques for the nuclear fuel test rod. The optimum torque value of a Swagelok nut to seal the mechanical sealing part between the end-cap and seal tube was established through various torque tests using a torque wrench. The optimum laser welding conditions to seal the welding part between a seal tube and an MI cable were obtained through various welding tests using a laser welding system.

  11. An electrical simulator of a nuclear fuel rod cooled by nucleate boiling

    Energy Technology Data Exchange (ETDEWEB)

    Costa, Antonio Carlos Lopes da [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)], e-mail: aclc@cdtn.br; Machado, Luiz; Koury, Ricardo Nicolau Nassar [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Mecanica], e-mail: luizm@demec.ufmg.br; Bonjour, Jocelyn [CETHIL, UMR5008, CNRS, INSA-Lyon (France)], e-mail: jocelyn.bonjour@insa-lyon.fr; Passos, Julio Cesar [Universidade Federal de Santa Catarina (UFSC), Florianopolis, SC (Brazil). Dept. de Engenharia Mecanica. LEPTEN/Boiling], e-mail: jpassos@emc.ufsc.br

    2009-07-01

    This study investigates an electrical heated test section designed to simulate a nuclear fuel rod. This simulator comprises a stainless steel vertical tube, with length and outside diameter of 600 mm and 10 mm, respectively, inside which there is a high power electrical resistor. The heat generated is removed by means of enhanced confined subcooled nucleate boiling of water in an annular space containing 153 small metal inclined discs. The tests were performed under electrical power and pressure up to 48 kW and 40 bar, respectively. The results show that the experimental boiling heat transfer coefficients are in good agreement with those calculated using the Jens-Lottes correlation. (author)

  12. A Mechanistic Approach for the Prediction of Critical Power in BWR Fuel Bundles

    Science.gov (United States)

    Chandraker, Dinesh Kumar; Vijayan, Pallipattu Krishnan; Sinha, Ratan Kumar; Aritomi, Masanori

    The critical power corresponding to the Critical Heat Flux (CHF) or dryout condition is an important design parameter for the evaluation of safety margins in a nuclear fuel bundle. The empirical approaches for the prediction of CHF in a rod bundle are highly geometric specific and proprietary in nature. The critical power experiments are very expensive and technically challenging owing to the stringent simulation requirements for the rod bundle tests involving radial and axial power profiles. In view of this, the mechanistic approach has gained momentum in the thermal hydraulic community. The Liquid Film Dryout (LFD) in an annular flow is the mechanism of CHF under BWR conditions and the dryout modeling has been found to predict the CHF quite accurately for a tubular geometry. The successful extension of the mechanistic model of dryout to the rod bundle application is vital for the evaluation of critical power in the rod bundle. The present work proposes the uniform film flow approach around the rod by analyzing individual film of the subchannel bounded by rods with different heat fluxes resulting in different film flow rates around a rod and subsequently distributing the varying film flow rates of a rod to arrive at the uniform film flow rate as it has been found that the liquid film has a strong tendency to be uniform around the rod. The FIDOM-Rod code developed for the dryout prediction in BWR assemblies provides detailed solution of the multiple liquid films in a subchannel. The approach of uniform film flow rate around the rod simplifies the liquid film cross flow modeling and was found to provide dryout prediction with a good accuracy when compared with the experimental data of 16, 19 and 37 rod bundles under BWR conditions. The critical power has been predicted for a newly designed 54 rod bundle of the Advanced Heavy Water Reactor (AHWR). The selected constitutive models for the droplet entrainment and deposition rates validated for the dryout in tube were

  13. Experimental study on the low flow CHF in vertical 3x3 rod bundle with non-uniform axial heat flux distribution

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Sang Ki; Cho, Seok; Chun, Se Young; Park, Jong Kuk; Kim, Bok Deuk; Youn, Young Jung; Baek, Won Pil

    2004-05-01

    An experimental study of the Critical Heat Flux (CHF) has been performed for a water flow in a non-uniformly heated vertical 3x3 rod bundle under low flow and a wide range of pressure conditions. Since most of experimental studies on the low flow CHF have been performed under low pressure conditions, present study has investigated the effects of various parameters on the CHF under low flow and a wide range of pressure conditions. Especially, these experiments are focused on the CHF under Return-To-Power (RTP) conditions that are expected to occur in a main steam line break accident of Pressurized Water Reactors (PWRs). Using present CHF data, the applicability of conventional CHF correlations are investigated in a return-to-power condition. The CHF data have been collected for system pressures ranging from 0.47 to 15.06 MPa, mass flux from 49.66 to 654.44 kg/m{sup 2}s, inlet subcooling from 67.90 to 722.70 kJ/kg and exit quality from 0.36 to 1.29. In this study, the return-to-power conditions are defined as conditions with low mass flux less than 250 kg/m{sup 2}s, intermediated pressure between 6.0 MPa and 12.0 MPa, and high inlet subcooling greater than 200 kJ/kg. Total 299 CHF data including 93 CHF data in return-to-power conditions are obtained. The effects of various parameters on the CHF are consistent with previous understandings on the round tube CHF. Conventional CHF correlations predict the present return-to-power CHF data with reasonable accuracies. However, the prediction capabilities become worse in a very low mass flux below than about 100 kg/m{sup 2}s.

  14. A New Insight into Energy Distribution of Electrons in Fuel-Rod Gap in VVER-1000 Nuclear Reactor

    Science.gov (United States)

    Fereshteh, Golian; Ali, Pazirandeh; Saeed, Mohammadi

    2015-06-01

    In order to calculate the electron energy distribution in the fuel rod gap of a VVER-1000 nuclear reactor, the Fokker-Planck equation (FPE) governing the non-equilibrium behavior of electrons passing through the fuel-rod gap as an absorber has been solved in this paper. Besides, the Monte Carlo Geant4 code was employed to simulate the electron migration in the fuel-rod gap and the energy distribution of electrons was found. As for the results, the accuracy of the FPE was compared to the Geant4 code outcomes and a satisfactory agreement was found. Also, different percentage of the volatile and noble gas fission fragments produced in fission reactions in fuel rod, i.e. Krypton, Xenon, Iodine, Bromine, Rubidium and Cesium were employed so as to investigate their effects on the electrons' energy distribution. The present results show that most of the electrons in the fuel rod's gap were within the thermal energy limitation and the tail of the electron energy distribution was far from a Maxwellian distribution. The interesting outcome was that the electron energy distribution is slightly increased due to the accumulation of fission fragments in the gap. It should be noted that solving the FPE for the energy straggling electrons that are penetrating into the fuel-rod gap in the VVER-1000 nuclear reactor has been carried out for the first time using an analytical approach.

  15. A study on the nuclear characteristics of enriched gadolinia burnable absorber rods; the first year (2000) report

    Energy Technology Data Exchange (ETDEWEB)

    Zee, Sung Quun; Lee, C.C.; Song, J. S.; Cho, B. O.; Joo, H. G.; Park, S. Y.; Kim, H. Y.; Cho, J. Y.; Kim, K. S

    2001-04-01

    An analysis model using MICBURN-3/CASMO-3 is established for the enriched gadolinia burnable absorber rods. A homogenized cross section editing code, PROLOG, is modified so that it can handle such a fuel assembly that includes two different types of gadolinia rods. Study shows that Gd-155 and Gd-157 are almost same in suppressing the excess reactivity and it is recommended to enrich both odd number isotopes, Gd-155 and Gd-157. It is estimated that the cycle length increases by 2 days if enriched gadolinia rods are used in the commercial nuclear power plant such as YGN-3 of which the cycle length is assumed 2 years. For the advanced integral reactor SMART in which ultra long cycle length and soluble boron-free operation concept is applied, natural gadolinia burnable absorber rods fail to control the excess reactivity. On the other hand, enriched gadolinia rods are successful in controling the excess reactivity. To minimize power peakings, various placements of gadolinia rods are tested. Also initial reactivity holddown and gadolinia burnout time are parametrized with respect to the number of gadolinia rods and gadolinia weight fractions.

  16. Process Management Development for Quality Monitoring on Resistance Weldment of Nuclear Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Na, Tae Hyung; Yang, Kyung Hwan; Kim, In Kyu [KEPCO, Daejeon (Korea, Republic of)

    2016-05-15

    The current, welding force, and displacement are displayed on the indicator during welding. However, real-time quality control is not performed. Due to the importance of fuel rod weldment, many studies on welding procedures have been conducted. However, there are not enough studies regarding weldment quality evaluation. On the other hand, there are continuous studies on the monitoring and control of welding phenomena. In resistance welding, which is performed in a very short time, it is important to find the process parameters that well represent the weld zone formation and the welding process. In his study, Gould attempted to analyze melt zone formation using the finite difference method. Using the artificial neural network, Javed and Sanders, Messler Jr et al., Cho and Rhee, Li and Gong et al. estimated the size of the melt zone by mapping a nonlinear functional relation between the weldment and the electrode head movement, which is a typical welding process parameter. Applications of the artificial intelligence method include fuzzy control using electrode displacement, fuzzy control using the optimal power curve, neural network control using the dynamic resistance curve, fuzzy adaptive control using the optimal electrode curve, etc. Therefore, this study induced quality factors for the real-time quality control of nuclear fuel rod end plug weldment using instantaneous dynamic resistance (IDR), which incorporates the instantaneous value of secondary current and voltage of the transformer, and using instantaneous dynamic force (IDF), obtained real-time during welding.

  17. Automatic system of welding for nuclear fuel rods; Sistema automatico de soldadura para barras de combustible nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Romero G, M; Romero C, J. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1998-07-01

    The welding process of nuclear fuel must be realized in an inert gas environment (He) and constant flow of this. In order to reach these conditions it is necessary to do vacuum at the chamber and after it is pressurized with the noble gas (purge) twice in the welding chamber. The purge eliminates impurities that can provoke oxidation in the weld. Once the conditions for initiating the welding are gotten, it is necessary to draw a graph of the flow parameters, pressure, voltage and arc current and to analyse those conditions in which have been carried out the weld. The rod weld must be free of possible pores or cracks which could provoke rod leaks, so reducing the probability of these failures should intervene mechanical and metallurgical factors. Automatizing the process it allows to do reliable welding assuring that conditions have been performed, reaching a high quality welding. Visually it can be observed the welding process by means of a mimic which represents the welding system. There are the parameters acquired such as voltage, current, pressure and flow during the welding arc to be analysed later. (Author)

  18. Detection of the Departure from Nucleate Boiling in Nuclear Fuel Rod Simulators

    Directory of Open Access Journals (Sweden)

    Amir Zacarias Mesquita

    2013-01-01

    Full Text Available In the thermal hydraulic experiments to determin parameters of heat transfer where fuel rod simulators are heated by electric current, the preservation of the simulators is essential when the heat flux goes to the critical point. One of the most important limits in the design of cooling water reactors is the condition in which the heat transfer coefficient by boiling in the core deteriorates itself. The heat flux just before deterioration is denominated critical heat flux (CHF. At this time, the small increase in heat flux or in the refrigerant inlet temperature at the core, or the small decrease in the inlet flux of cooling, results in changes in the heat transfer mechanism. This causes increases in the surface temperature of the fuel elements causing failures at the fuel (burnout. This paper describes the experiments conducted to detect critical heat flux in nuclear fuel element simulators carried out in the thermal-hydraulic laboratory of Nuclear Technology Development Centre (CDTN. It is concluded that the use of displacement transducer is the most efficient technique for detecting critical heat flux in nuclear simulators heated by electric current in open pool.

  19. Nuclear Data Library Effects on Fast to Thermal Flux Shapes Around PWR Control Rod Tips

    Science.gov (United States)

    Vasiliev, A.; Ferroukhi, H.; Zhu, T.; Pautz, A.

    2014-04-01

    The development of a high-fidelity computational scheme to estimate the accumulated fluence at the tips of PWR control rods (CR) has been initiated at the Paul Scherrer Institut (PSI). Both the fluence from high-energy (E>1 MeV) neutrons as well as for the thermal range (E<0.625 eV) are required as these affect the CR integrity through stresses/strains induced by coupled clad embrittlement / absorber swelling phenomena. The concept of the PSI scheme under development is to provide from validated core analysis models, the volumetric neutron source to a full core MCNPX model that is then used to compute the neutron fluxes. A particular aspect that needs scrutiny is the ability of the MCNPX-based calculation methodology to accurately predict the flux shapes along the control rod surfaces, especially for fully withdrawn CRs. In that case, the tip is located a short distance above the core/reflector interface and since this situation corresponds to a large part of reactor operation, the accumulated fluence will highly depend on the achieved calculation accuracy and precision in this non-fueled zone. The objective of the work presented in this paper is to quantify the influence of nuclear data on the calculated fluxes at the CR tips by (1) conducting a systematic comparison of modern neutron cross-section libraries, including JENDL-4.0, JEFF-3.1.1 and ENDF/B-VII.0, and (2) by quantifying the uncertainties in the neutron flux calculations with the help of available neutron cross-section variances/covariances data. For completeness, the magnitude of these nuclear data-based uncertainties is also assessed in relation to the influence from other typical sources of modeling uncertainties/biases.

  20. Modeling of Complex Wear Behavior Associated with Grid-to-Rod Fretting in Light Water Nuclear Reactors

    Science.gov (United States)

    Blau, P. J.; Qu, J.; Lu, R.

    2016-09-01

    Fretting wear damage to fuel cladding from flow-induced vibrations can be a significant concern in the operation of light water nuclear reactors. For years, research on the grid-to-rod fretting (GTRF) phenomena has been underway in countries where nuclear power production is a significant industry. Under the auspices of the U.S. Department of Energy Consortium for Advanced Simulation of Light Water Reactors, an effort has been underway to develop and test an engineering wear model for zirconium alloy fuel rod cladding against a supporting grid. The multi-stage model accounts for oxide layers and wear rate transitions. This paper describes the basis for a GTRF engineering wear model, the physical significance of the wear factor it contains, and recent progress toward model validation based on a fretting wear testing apparatus that accounts for coolant temperature, pressure, and the presence of periodic impacts (gaps) in grid/rod contact.

  1. Modeling of Complex Wear Behavior Associated with Grid-to-Rod Fretting in Light Water Nuclear Reactors

    Science.gov (United States)

    Blau, P. J.; Qu, J.; Lu, R.

    2016-11-01

    Fretting wear damage to fuel cladding from flow-induced vibrations can be a significant concern in the operation of light water nuclear reactors. For years, research on the grid-to-rod fretting (GTRF) phenomena has been underway in countries where nuclear power production is a significant industry. Under the auspices of the U.S. Department of Energy Consortium for Advanced Simulation of Light Water Reactors, an effort has been underway to develop and test an engineering wear model for zirconium alloy fuel rod cladding against a supporting grid. The multi-stage model accounts for oxide layers and wear rate transitions. This paper describes the basis for a GTRF engineering wear model, the physical significance of the wear factor it contains, and recent progress toward model validation based on a fretting wear testing apparatus that accounts for coolant temperature, pressure, and the presence of periodic impacts (gaps) in grid/rod contact.

  2. Single-Phase Crossflow Mixing in a Vertical Tube Bundle Geometry: An Experimental Study

    NARCIS (Netherlands)

    Mahmood, A.

    2011-01-01

    The vertical rod/tube bundle geometry has a wide variety of industrial applications. Typical examples are the core of light water nuclear reactors (LWR) and vertical tube steam generators. In the core of a LWR, primarily coolant flows upward but their also exist a flow in lateral direction, called c

  3. Velocity and turbulence distributions in wall subchannels of a road bundle in three axial planes downstream of a spacer grid

    Science.gov (United States)

    Rehme, K.

    1987-03-01

    The velocity, turbulence, and temperature distributions in nuclear fuel element bundles of nuclear reactors were investigated. The mean velocity, the wall shear stresses, and the turbulence were measured in two wall subchannels of a rod bundle of four parallel rods, arranged in a rectangular channel, for three axial planes. A spacer grid was inserted in the rod bundle, for ratios between the distance spacer grid/measuring plane and the hydraulic diameter (LIDh) of 40.4, 32.8 and 16.9. The Reynolds number was 145,000. The results show that the distributions of the velocity and the turbulence are affected by the spacer grid, already for LIDh = 40.4. The effects of the spacer grid increase with decreasing distance to the spacer grid.

  4. Two-phase flow modeling in the rod bundle subchannel analysis; Modelisation d'ecoulement a deux phases dans l'analyse du sous-canal de grappe d'assemblages

    Energy Technology Data Exchange (ETDEWEB)

    Hisashi, Ninokata [Tokyo Inst. of Tech. (Japan)

    2006-07-01

    In order to practice a design-by-analysis of thermohydraulics design of BWR fuel rod bundles, the subchannel analysis would play a major role. There, the immediate concern is improvement in its predictive capability of CHF due in particular to the film dryout (boiling transition phenomena: BT) on the fuel rod surface. Constitutive equations in the subchannel analysis formulation are responsible for the quality of calculated results. The constitutive equations are a result of integration of the local and instantaneous description of two-phase flows over the subchannel control volume. In general, they are expressed in terms of subchannel-control-volume- as well as area-averaged two-phase flow state variables. In principle the information on local and instantaneous physical phenomena taking place inside subchannels must be counted for in the algebraic form of the equations on the basis of a more mechanistic modeling approach. They should include also influences of the multi-dimensional subchannel geometry and fluid material properties. Thermohydraulics phenomena of interests in this deed are: 1) vapor-liquid re-distribution by inter-subchannel exchanges due to the diversion cross flow, turbulent mixing and void drift, 2) liquid film behaviors, 3) transition of two-phase flow regimes, 4) droplet entrainment and deposition and 5) spacer-droplet interactions. These are considered to be five key factors in understanding the BT in BWR fuel rod bundles. In Japan, a university-industry consortium has been formed under the sponsorship of the Ministry of Economics, Trade and Industry. This paper describes an outline of the on-going project and, first, an outline of the current efforts is presented in developing a new two-fluid three field subchannel code NASCA being aimed at predicting onset of BT, and post BT phenomena in advanced BWR fuel rod bundles including those of the tight lattice configuration for a higher conversion. Then the current methodology adopted to improve

  5. Simulations and measurements of adiabatic annular flows in triangular, tight lattice nuclear fuel bundle model

    Energy Technology Data Exchange (ETDEWEB)

    Saxena, Abhishek, E-mail: asaxena@lke.mavt.ethz.ch [ETH Zurich, Laboratory for Nuclear Energy Systems, Department of Mechanical and Process Engineering, Sonneggstrasse 3, 8092 Zürich (Switzerland); Zboray, Robert [Laboratory for Thermal-hydraulics, Nuclear Energy and Safety Department, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Prasser, Horst-Michael [ETH Zurich, Laboratory for Nuclear Energy Systems, Department of Mechanical and Process Engineering, Sonneggstrasse 3, 8092 Zürich (Switzerland); Laboratory for Thermal-hydraulics, Nuclear Energy and Safety Department, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland)

    2016-04-01

    High conversion light water reactors (HCLWR) having triangular, tight-lattice fuels bundles could enable improved fuel utilization compared to present day LWRs. However, the efficient cooling of a tight lattice bundle has to be still proven. Major concern is the avoidance of high-quality boiling crisis (film dry-out) by the use of efficient functional spacers. For this reason, we have carried out experiments on adiabatic, air-water annular two-phase flows in a tight-lattice, triangular fuel bundle model using generic spacers. A high-spatial-resolution, non-intrusive measurement technology, cold neutron tomography, has been utilized to resolve the distribution of the liquid film thickness on the virtual fuel pin surfaces. Unsteady CFD simulations have also been performed to replicate and compare with the experiments using the commercial code STAR-CCM+. Large eddies have been resolved on the grid level to capture the dominant unsteady flow features expected to drive the liquid film thickness distribution downstream of a spacer while the subgrid scales have been modeled using the Wall Adapting Local Eddy (WALE) subgrid model. A Volume of Fluid (VOF) method, which directly tracks the interface and does away with closure relationship models for interfacial exchange terms, has also been employed. The present paper shows first comparison of the measurement with the simulation results.

  6. Design of the Testing Set-up for a Nuclear Fuel Rod by Neutron Radiography at CARR

    Science.gov (United States)

    Wei, Guohai; Han, Songbai; Wang, Hongli; Hao, Lijie; Wu, Meimei; He, Linfeng; Wang, Yu; Liu, Yuntao; Sun, Kai; Chen, Dongfeng

    In this paper, an experimental set-up dedicated to non-destructively test a 15cm-long Pressurized Water Reactor (PWR) nuclear fuel rod by neutron radiography (NR) is described. It consists of three parts: transport container, imaging block and steel support. The design of the transport container was optimized with Monte-Carlo Simulation by the MCNP code. The material for the shell of the transport container was chosen to be lead with the thickness of 13 cm. Also, the mechanical devices were designed to control fuel rod movement inside the container. The imaging block was designed as the exposure platform, with three openings for the neutron beam, neutron converter foil, and specimen. Development and application of this experimental set-up will help gain much experience for investigating the actual irradiated fuel rod by neutron radiography at CARR in the future.

  7. Studies on sodium boiling phenomena in out of pile rod bundles for various accidental situations in Liquid Metal Fast Breeder Reactors (LMFBR) experiments and interpretations

    Science.gov (United States)

    Seiler, J. M.; Rameau, B.

    Bundle sodium boiling in nominal geometry for different accident conditions is reviewed. Voiding of a subassembly is controlled by not only hydrodynamic effects but mainly by thermal effects. There is a strong influence of the thermal inertia of the bundle material compared to the sodium thermal inertia. Flow instability, during a slow transient, can be analyzed with numerical tools and estimated using simplified approximations. Stable boiling operational conditions under bundle mixed convection (natural convection in the reactor) can be predicted. Voiding during a fast transient can be approximated from single channel calculations. The phenomenology of boiling behavior for a subassembly with inlet completely blocked, submitted to decay heat and lateral cooling; two-phase sodium flow pressure drop in a tube of large hydraulic diameter under adiabatic conditions; critical flow phenomena and voiding rate under high power, slow transient conditions; and onset of dry out under local boiling remains problematical.

  8. Characterization of a suspect nuclear fuel rod in a case of illegal international traffic of fissile material.

    Science.gov (United States)

    Capannesi, G; Vicini, C; Rosada, A; Avino, P

    2010-06-15

    This case study describes the characterization of a suspect rod of nuclear fuel seized in Italy: on request of the coroner, the characterization concerned the kind and the conditions of the rod, the amount and the specific characteristics of the species present in it, with particular attention to their possible use chemical and/or nuclear plants. The methodology used was based on radiochemical analyses (gammagraphic and gamma-spectrometry) whereas the comparison was performed by means of a fuel reference element working in the TRIGA nuclear reactor at Research Center of ENEA-Casaccia. The results show clearly how the exhibit was an element of nuclear fuel, how long it was irradiated, and the amount of (239)Pu produced and the (235)U consumed. Finally, even if the seized rod was briefly radiated at the "zero power" and traces of fission products and plutonium were found, it would be still usable as "fresh" fuel in a reactor type TRIGA if it had not been intercepted by Italian police authorities.

  9. Bending testing and characterization of surrogate nuclear fuel rods made of Zircaloy-4 cladding and aluminum oxide pellets

    Science.gov (United States)

    Wang, Hong; Wang, Jy-An John

    2016-10-01

    Behavior of surrogate nuclear fuel rods made of Zircaloy-4 (Zry-4) cladding with alumina pellets under reversed cyclic bending was studied. Tests were performed under load or moment control at 5 Hz. The surrogate rods fractured under moment amplitudes greater than 10.16 Nm with fatigue lives between 2.4 × 103 and 2.2 × 106 cycles. Fatigue response of Zry-4 cladding was characterized by using flexural rigidity. Degradation of flexural rigidity was shown to depend on the moment and the prefatigue condition of specimens. Pellet-to-pellet interface (PPI), pellet-to-cladding interface (PCI), and pellet condition affect surrogate rod failure. Both debonding of PPI/PCI and pellet fracturing contribute to surrogate rod bending fatigue. The effect of sensor spacing on curvature measurement using three-point deflections was studied; the method based on effective gauge length is effective in sensor spacing correction. The database developed and the understanding gained in this study can serve as input to analysis of SNF (spent nuclear fuel) vibration integrity.

  10. Reliability centered maintenance applied to the control rod drives of a nuclear power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lyytikaeinen, A.; Holmberg, J. (Technical Research Centre of Finland, Espoo (Finland)); Unga, E. (Teollisuuden Voima Oy, Olkiluoto (Finland))

    1990-12-01

    Reliability Centered Maintenance (RCM) offers a hybrid reliability analysis methodology to evaluate the level of maintenance and direct the resources in an effective manner. The RCM analysis consists of several steps, including a Logic Tree Analysis (LTA) for identification of applicable and efficient preventive maintenance task. As a result of the analysis the total amount of maintenance of equipment is decreased or increased, depending on whether the failures are potential having adverse effects on plant safety, availability or economics. A RCM analysis results thus in an improved preventive maintenance program, which is supposed to decrease the maintenance and outage costs and at the same time increase the system safety and reliability. A problem is the relatively large and slow analysis effort, the fact which tends to inhibit the large scale and continuous application of this useful method. A case study is described, in which the control rod equipment at the TVO I/II nuclear units was analyzed and the problematics of the RCM-method is discussed. (au).

  11. Correlation for cross-flow resistance coefficient using STAR-CCM+ simulation data for flow of water through rod bundle supported by spacer grid with split-type mixing vane

    Energy Technology Data Exchange (ETDEWEB)

    Agbodemegbe, V.Y., E-mail: vincevalt@gmail.com [Karlsruhe Institute of Technology, Institute of Fusion and Reactor Technique, Kaiserstrasse 12, Karlsruhe (Germany); Cheng, Xu, E-mail: xu.cheng@kit.edu [Karlsruhe Institute of Technology, Institute of Fusion and Reactor Technique, Kaiserstrasse 12, Karlsruhe (Germany); Akaho, E.H.K, E-mail: akahoed@yahoo.com [School of Nuclear and Allied Sciences, University of Ghana, PO Box AE 1, Kwabenya, Accra (Ghana); Allotey, F.K.A, E-mail: fkallotey@gmail.com [Institute of Mathematical Sciences, PO Box LG 197, Legon, Accra (Ghana)

    2015-04-15

    Highlights: • Investigate spacer grid with split-type mixing vanes. • Extent of predictability of experimental data by STAR-CCM+. • Reliability of two equation turbulence models. • Resistance to cross-flow through gaps. - Abstract: Mass transfer by diversion cross-flow through gaps is an important inter-subchannel interaction in fuel bundle of power reactors. It is normally due to the lateral pressure difference between adjacent sub-channels. This phenomenon is augmented in the presence of flow deflectors and is referred to as, directed cross-flow. Diversion cross-flow carries the momentum and energy of flow and hence affects the velocity and temperature profile in the rod bundle. The resistance to cross-flow in the transverse momentum equations is specified by the cross-flow resistant coefficient which is the subject of concern in the present study. In order to obtain data to correlate cross-flow resistance coefficient, computational fluid dynamic simulation using STAR-CCM+ was performed for flow of water at the bundle Reynolds number of Re1 = 3.4×10{sup 4} through a 5 × 5 rod bundle geometry supported by spacer grid with split mixing vanes for which the rod to rod pitch to diameter ratio was 1.33 and the rod to wall pitch to diameter ratio was 0.74. The two layer k-epsilon turbulence model with an all y+ automatic wall treatment function in STAR-CCM+ were adopted for an isothermal single phase (water) flow through the geometry. The objectives were to primarily investigate the extent of predictability of the experimental data by the computational fluid dynamic (CFD) simulation as a measure of reliability on the CFD code employed and also apply the simulation data to develop correlations for determining resistance coefficient to cross-flow. Validation of simulation results with experimental data showed good correlation of mean flow parameters with experimental data whiles turbulent fluctuations deviated largely from experimental trends. Generally, the

  12. Irradiation analysis, production test IP-672, HAPO 238, irradiation of impacted UO{sub 2}-PuO{sub 2} fuel rod bundles in C reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cox, J.H.

    1964-09-14

    The loss of flow is considered as far as the flow to the inlet hydraulic connector, inlet plugging and water shutoff time. A mockup revealed no vibration of the fuel element bundles and at the low temperatures present there should be no problem of corrosion. Efforts to assure safety with plutonium in the fuel elements are noted. (GHH)

  13. DIONISIO 2.0: New version of the code for simulating a whole nuclear fuel rod under extended irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Soba, Alejandro, E-mail: soba@cnea.gov.ar; Denis, Alicia

    2015-10-15

    Highlights: • A new version of the DIONISIO code is developed. • DIONISIO is devoted to simulating the behavior of a nuclear fuel rod in operation. • The formerly two-dimensional simulation of a pellet-cladding segment is now extended to the whole rod length. • An acceptable and more realistic agreement with experimental data is obtained. • The prediction range of our code is extended up to average burnup of 60 MWd/kgU. - Abstract: The version 2.0 of the DIONISIO code, that incorporates diverse new aspects, has been recently developed. One of them is referred to the code architecture that allows taking into account the axial variation of the conditions external to the rod. With this purpose, the rod is divided into a number of axial segments. In each one the program considers the system formed by a pellet and the corresponding cladding portion and solves the numerous phenomena that take place under the local conditions of linear power and coolant temperature, which are given as input parameters. To do this a bi-dimensional domain in the r–z plane is considered where cylindrical symmetry and also symmetry with respect to the pellet mid-plane are assumed. The results obtained for this representative system are assumed valid for the complete segment. The program thus produces in each rod section the values of the temperature, stress, strain, among others as outputs, as functions of the local coordinates r and z. Then, the general rod parameters (internal rod pressure, amount of fission gas released, pellet stack elongation, etc.) are evaluated. Moreover, new calculation tools designed to extend the application range of the code to high burnup, which were reported elsewhere, have also been incorporated to DIONISIO 2.0 in recent times. With these improvements, the code results are compared with some 33 experiments compiled in the IFPE data base, that cover more than 380 fuel rods irradiated up to average burnup levels of 40–60 MWd/kgU. The results of these

  14. External Attachment of Titanium Sheathed Thermocouples to Zirconium Nuclear Fuel Rods For The Loss-Of-Fluid-Test (LOFT) Reactor

    Science.gov (United States)

    Welty, Richard K.

    1980-10-01

    The Exxon Nuclear Company, Inc. acting as a Subcontractor to EG&G Idaho Inc.3 Idaho National Engineering Laboratory, Idaho Falls, Idaho, has developed a welding process to attach titanium sheathed thermocouples to the outside of the zircaloy clad fuel rods. The fuel rods and thermocouples are used to test simulated loss-of-coolant-accident (LOCA) conditions in a pressurized water reactor (LOFT Reactor, Idaho National Laboratory). The design goals were to (1) reliably attach thermocouples to the zircaloy fuel rods, (2) achieve or exceed a life expectancy of 6,000 hours of reactor operation in a borated water environment of 316°C at 2260 psi, (3) provide and sustain repeatable physical and metallurgical properties in the instrumented rods subjected to transient temperatures up to 1538°C with blowdown, shock, loading, and fast quench. A laser beam was selected as the optimum welding process because of the extremely high energy input per unit volume that can be achieved allowing local fusion of a small area irrespective of the difference in material thickness to be joined. A commercial pulsed laser and energy control system was installed along with specialized welding fixtures. Laser room facility requirements and tolerances were established. Performance qualifications and detailed welding procedures were also developed. Product performance tests were conducted to assure that engineering design requirements could be met on a production basis. Irradiation tests showed no degradation of thermocouples or weld structure. Fast thermal cycle and heater rod blowdown reflood tests were made to subject the weldments to high temperatures, high pressure steam, and fast water quench cycles. From the behavior of these tests, it was concluded that the attachment welds would survive a series of reactor safety tests.

  15. CFD analysis of rewetting vertical nuclear fuel rod by dispersed fluid jet impingement

    Directory of Open Access Journals (Sweden)

    Ajoy Debbarma

    2016-09-01

    Full Text Available Numerical analysis of cooling assessment in hot vertical fuel rod is carryout using ANSYS 14.0 – CFX Solver. Rewetting is the process of re-establishment of coolants with hot surfaces. Numerical validation exercise carried out with number of turbulence and shear stress turbulence model fairly predict the experimental data and used for further investigation. In the present paper, dispersed fluid is simulating with CFX solver to investigate the flow boiling process in emergency cooling of vertical fuel rod. When coolants come in contact on the hot surface this may not initiated the wetting patch. However, this paper introduces the unique jet impingement direction to remove the heat from the hot surface. In this report, the rewetting temperature and wetting delay also described during in progress of wetting front movement in hot vertical rod.

  16. Inspection of domestic nuclear fuel rods using neutron radiography at the Tehran research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dastjerdi, Mohammad Hosein Choopan; Khalafi, Hossein; Kasesaz, Yaser [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Movafeghi, Amir

    2016-11-01

    Three unused domestic fuel rods were investigated qualitatively and quantitatively by means of thermal neutron radiography. The neutron radiography tests were performed by the image plate method at Tehran research reactor in order to check the fuel properties. The pellets of these three fuel rods contained three different U-235 enrichments and different sizes that were filled into a zircalloy tube. In the qualitative investigations, the difference in size and enrichment between the pellets and the gaps between them were obviously recognized in the image of the fuel rods. In the quantitative investigations, data of the pellets compositions, their sizes (lengths and diameters) and the gaps between them were extracted from obtained images. It was found that the measured data and the manufacturer's specifications are in good agreement.

  17. Prediction of the lifetime of the elements of the safety and control rods of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Voskoboinikov, V.V.; Emel' yanov, I. Ya.; Lineva, A.F.; Pushkin, S.N.; Semchenko, E.L.; Usov, P.P.

    1987-07-01

    The authors construct a mathematical model based on the analytical solution to such parameters as magnetic flux, alignment, sliding friction, erosion, abrasion, corrosion, and irradiation for the purpose of predicting the service life of electromagnetically driven control rods and their drives. The analytical solution was verified experimentally on a bench simulation where it was found that incorrect assembly and alignment not only serve as the largest contributors to shortened service life of the rods and drives but also render the calculations of the model invalid.

  18. Starting Point, Keys and Milestones of a Computer Code for the Simulation of the Behaviour of a Nuclear Fuel Rod

    Directory of Open Access Journals (Sweden)

    Armando C. Marino

    2011-01-01

    Full Text Available The BaCo code (“Barra Combustible” was developed at the Atomic Energy National Commission of Argentina (CNEA for the simulation of nuclear fuel rod behaviour under irradiation conditions. We present in this paper a brief description of the code and the strategy used for the development, improvement, enhancement, and validation of a BaCo during the last 30 years. “Extreme case analysis”, parametric (or sensitivity, probabilistic (or statistic analysis plus the analysis of the fuel performance (full core analysis are the tools developed in the structure of BaCo in order to improve the understanding of the burnup extension in the Atucha I NPP, and the design of advanced fuel elements as CARA and CAREM. The 3D additional tools of BaCo can enhance the understanding of the fuel rod behaviour, the fuel design, and the safety margins. The modular structure of the BaCo code and its detailed coupling of thermo-mechanical and irradiation-induced phenomena make it a powerful tool for the prediction of the influence of material properties on the fuel rod performance and integrity.

  19. Tomography on nuclear fuel rods in the nuclear power plant of Dodewaard. Tomografie aan splijtstofstaven in de centrale Dodewaard

    Energy Technology Data Exchange (ETDEWEB)

    Tanke, R.H.J.; Jaspers, J.E.; Gaalman, P.A.M. (KEMA, Arnhem (Netherlands). Division Research and Development)

    1990-09-06

    This report discusses the feasibility of using emission tomography on fuel rods in the Dodewaard reactor. The tomography can be used to increase the efficiency of the use of fissionable material. (R.A.B.). 4 refs.; 17 figs.; 1 tab.

  20. Analysis of Frictional Resistance of Two-phase Flow in Rod Bundle Channel%棒束通道内两相流动摩擦阻力特性分析

    Institute of Scientific and Technical Information of China (English)

    田齐伟; 阎昌琪; 孙立成; 闫超星

    2015-01-01

    The experimental investigation of air‐water two‐phase flow resistance charac‐teristics in a vertical channel with a 3 × 3 rod bundle was carried out under atmospheric and room temperature conditions . Eight classical correlations for predicting frictional pressure drop of two‐phase flow were evaluated against the experimental data . The experimental results show that the homogeneous model can predict the experimental data well at high flow rates ,but with relatively large deviations at low flow rates .Both the Friedel model and the Lombodi‐Pedrocchi model are not suitable any longer for the present case . The Chisholm C model , the Zhang‐Mishima model , the Chisholm B model ,the Mishima‐Hibiki model and the L .Sun model can well predict the experimen‐tal data with mean relative errors in the range of 20%‐30% . The C factor in the Chisholm C model was modified for giving a new correlation to predict the frictional pressure drop of two‐phase flow through rod bundles ,showing a good agreement with the experimental data .%常温常压下,对竖直3×3棒束通道内气液两相流动阻力特性进行了实验研究。利用所获得的实验数据,对8种典型的两相流动摩擦压降计算模型进行了评价。结果表明,均相模型在两相流速较高时精度较高,在两相流速较低时则偏差较大。分相模型中,Friedel模型和Lombodi‐Pedrocchi模型不适用于本实验条件下棒束通道内气液两相流动摩擦压降的计算。Chisholm C模型、Zhang‐M ishima模型、Chisholm B模型、Mishima‐Hibiki模型及L .Sun模型的预测值与实验值的平均相对误差介于20%~30%之间。基于实验数据,通过修正Chisholm C模型的C系数,给出一个新的修正模型,其计算值与实验值符合良好。

  1. Thermoacoustic enhancements for nuclear fuel rods and other high temperature applications

    Energy Technology Data Exchange (ETDEWEB)

    Garrett, Steven L.; Smith, James A.; Kotter, Dale K.

    2017-05-09

    A nuclear thermoacoustic device includes a housing defining an interior chamber and a portion of nuclear fuel disposed in the interior chamber. A stack is disposed in the interior chamber and has a hot end and a cold end. The stack is spaced from the portion of nuclear fuel with the hot end directed toward the portion of nuclear fuel. The stack and portion of nuclear fuel are positioned such that an acoustic standing wave is produced in the interior chamber. A frequency of the acoustic standing wave depends on a temperature in the interior chamber.

  2. 棒束燃料组件特征栅元CFD方法研究%CFD Method Research on Characteristic Cells in Rod Bundle Fuel Assembly

    Institute of Scientific and Technical Information of China (English)

    陈杰; 陈炳德; 张虹

    2011-01-01

    Two characteristic cells are in AFA-3G fuel assembly, that is typical cell and control rod guide cell. And there are some rules on the arrangement of mixing vanes. For the two characteristic cells, mixing capability is evaluated axially from the point of the first and second kind of sub-channel with CFD method.Mass mixing and heat mixing are interaction but different with each other. Although the mass mixing in the first kind of sub-channel is stronger, the thermal capability of the two is to some tune from the point of heat transfer. In the experiment research on thermal-hydraulic performance of AFA-3G fuel assembly, the arrangements of mixing vanes should refer to the two spacer grids of characteristic cells.%AFA-3G燃料组件中存在典型栅元和控制棒导向管栅元两种特征栅元,定位格架搅混翼的排列也具有一定的规律性.本文采用计算流体力学(CFD)方法,分别针对两种特征栅元,从第一类子通道和第二类子通道的角度,沿程评价其交混性能.质量交混与热交混紧密联系又相互区别,第一类子通道质量交换较强,但从传热角度,二者性能相当.AFA-3G燃料组件热工水力性能的实验研究中,格架搅混翼的排列方式应分别参照两种特征栅元格架.

  3. A preliminary approach to the extension of the Transuranus code to the fuel rod performance analysis of HLM-cooled nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Luzzi, L.; Botazzoli, P.; Devita, M.; Di Marcello, V.; Pastore, G. [Department of Energy, Politecnico di Milano, Enrico Fermi Center for Nuclear Studies - CeSNEF, via Ponzio 34/3, 20133 Milano (Italy)

    2010-07-01

    This paper briefly presents a preliminary modelling approach, aimed at the extension of the TRANSURANUS code to the fuel rod performance analysis of Heavy Liquid Metal (HLM) cooled nuclear reactors, with specific reference to the employment of the T91 steel as cladding material and of the liquid Lead-Bismuth Eutectic (LBE) as coolant. On the basis of literature indications, correlations for heat transfer to LBE, corrosion behaviour and thermo-mechanical properties of T91 are proposed, and some open issues are discussed in prospect of more reliable fuel rod performance analysis of HLM-cooled nuclear reactors. (authors)

  4. Dimensional Measurements of Fresh CANDU Fuel Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Jun, Ji Su; Jo, Chang Keun; Jung, Jong Yeob; Koo, Dae Seo; Cho, Moon Sung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    This paper intends to provide the dimensional measurements of fresh CANDU fuel (37-element) bundle for the estimation of deformation of post-irradiated (PI) bundle. It is expensive and difficult to measure the fretting wear of bearing pad, the element bowing and the waviness of endplate at the two-phase high flow condition (above 24 kg/s) of out-of-reactor test. So, it is recommended to compare the geometry of fresh bundle with that of PI bundle to estimate the integrity of fuel bundle in the CANDU-6 fuel channel with two-phase flow condition. The measurement system has been developed to provide the visual inspection and the dimensional measurements within the accuracy of 10 {mu}m. It is applicable in-air and underwater to the CANDU bundle as well as the CANFLEX bundle. The in-air measurements of the 36 fresh CANDU bundles (S/N: B400892 {approx} B400927) are done by this system from February 2004 to March 2004 in the PHWR fresh fuel storage building of KNFC. These bundles are produced by KNFC manufacturing procedure and are waiting for the delivery to the Wolsong-3 plant, and are planned to load into the proposed test channels. The detail measurements contain the outer rod profile (including the bearing pad), the diameter of bundle, the bowing of bundle, the rod length and the surface profile of end plate (waviness)

  5. Fuel enrichment and temperature distribution in nuclear fuel rod in (D-T) driven hybrid reactor system

    Energy Technology Data Exchange (ETDEWEB)

    Osman, Ypek [Suleyman Demirel Universitesi Muhendislik-Mimarlyk Fakultesi, Isparta (Turkey)

    2001-07-01

    In this study, melting point of the fuel rod and temperature distribution in nuclear fuel rod are investigated for different coolants under various first wall loads (P{sub w}, =5, 6, 7, 8, 9, and 10 MWm{sup -2}) in Fusion-Fission reactor fueled with 50%LWR +50%CANDU. The fusion source of neutrons of 14.1 MeV is simulated by a movable target along the main axis of cylindrical geometry as a line source. In addition, the fusion chamber was thought as a cylindrical cavity with a diameter of 300 cm that is comparatively small value. The fissile fuel zone is considered to be cooled with four different coolants, gas, flibe (Li{sub 2}BeF{sub 4}), natural lithium (Li), and eutectic lithium (Li{sub 17}Pb{sub 83}). Investigations are observed during 4 years for discrete time intervals of{delta}t= 0.5 month and by a plant factor (PF) of 75%. Volumetric ratio of coolant-to fuel is 1:1, 45.515% coolant, 45.515% fuel, 8.971% clad, in fuel zone. (author)

  6. Numerical and experimental study of hydraulic dashpot used in the shut-off rod drive mechanism of a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Singh, Narendra K., E-mail: nksingh_chikki@yahoo.com [Division of Remote Handling and Robotics, Bhabha Atomic Research Centre, Mumbai 400085 (India); Badodkar, Deepak N. [Division of Remote Handling and Robotics, Bhabha Atomic Research Centre, Mumbai 400085 (India); Homi Bhabha National Institute, Anushaktinagar, Mumbai 400094 (India); Singh, Manjit [Division of Remote Handling and Robotics, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2014-07-01

    Highlights: • Hydraulic dashpot performance is studied numerically as well as experimentally. • Instantaneous pressure built-up in the dashpot is mainly contributing for damping of freely falling shut-off rod at the end of its travel. • At elevated temperature, dashpot pressure does not reduce in proportion to the reduction in viscosity. • ‘C’ grove in the dashpot shaft flattens the pressure peak and shifts it toward the end of operation. - Abstract: Hydraulic dashpot design for shut-off rod drive mechanism application in a nuclear reactor has been analyzed both numerically and experimentally in this paper. Finite element commercial code COMSOL Multiphysics 4.3 has been used for numerical analysis. Experimental validation has been done at two different cases. Experimental test set-ups and hydraulic dashpot constructions have been described in detail. Various combinations of dashpot oil viscosity and clearance thickness have been analyzed. Important experimental results are also presented and discussed. Pressure distributions in the dashpot chambers obtained from COMSOL are given for both the set-ups. Numerical and experimental results are compared. Dashpot designs have been qualified after detailed analysis and testing on full-scale test stations simulating actual reactor conditions (except radiation)

  7. Model of fracture for the Zry cladding of nuclear fuel rods included in the code DIONISIO 1.0

    Energy Technology Data Exchange (ETDEWEB)

    Soba, Alejandro [Departamento Combustibles Nucleares, Comision Nacional de Energia Atomica, Av. del Libertador 8250, 1429 Buenos Aires (Argentina)], E-mail: soba@cnea.gov.ar; Denis, Alicia [Departamento Combustibles Nucleares, Comision Nacional de Energia Atomica, Av. del Libertador 8250, 1429 Buenos Aires (Argentina)], E-mail: denis@cnea.gov.ar

    2008-12-15

    The DIONISIO code describes most of the main phenomena occurring in a fuel rod during normal operation of a nuclear power reactor. Starting from the irradiation history, the code predicts the temperature distribution, elastic and plastic stress and strain, creep, swelling and densification, release of fission gases, caesium and iodine to the internal rod volume, gas mixing, pressure increase, irradiation growth of the cladding, development of an oxide layer on its surface and hydrogen uptake, restructuring and grain growth in the pellet. This work presents the model of Zircaloy fracture included in the code DIONISIO 1.0. The model of pellet-cladding mechanical interaction (PCMI) provides the forces caused by the solid-solid contact which add to the changing internal pressure and to the constant external pressure. Besides, the program evaluates the effects of a corrosive atmosphere (stress corrosion cracking, SCC) internal or external. With these data, the code calculates the J integral around the tip of an initiated crack, and proceeds to analyze, according to the quantity of corrosive substance dissolved and the cladding stress field, if the crack remains unchanged, if it grows due to the I-SCC mechanism, or if propagation is ductile, following the R curve of the material. Results corresponding to different PHWR and PWR reactors are presented and compared with code results. In particular, good agreement is obtained in the simulation of MOX experiments, where the cladding failed due to propagation of cracks originated in SCC.

  8. Simulation with DIONISIO 1.0 of thermal and mechanical pellet-cladding interaction in nuclear fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Soba, Alejandro [Departamento Combustibles Nucleares, Comision Nacional de Energia Atomica, Avenida del Libertador 8250, 1429 Buenos Aires (Argentina); Denis, Alicia [Departamento Combustibles Nucleares, Comision Nacional de Energia Atomica, Avenida del Libertador 8250, 1429 Buenos Aires (Argentina)], E-mail: denis@cnea.gov.ar

    2008-02-29

    The code DIONISIO 1.0 describes most of the main phenomena occurring in a fuel rod throughout its life under normal operation conditions of a nuclear thermal reactor. Starting from the power history, DIONISIO predicts the temperature distribution in the domain, elastic and plastic stress and strain, creep, swelling and densification, release of fission gases, caesium and iodine to the rod free volume, gas mixing, pressure increase, restructuring and grain growth in the UO{sub 2} pellet, irradiation growth of the Zircaloy cladding, oxide layer growth on its surface, hydrogen uptake and the effects of a corrosive atmosphere either internal or external. In particular, the models of thermal conductance of the gap and of pellet-cladding mechanical interaction incorporated to the code constitute two realistic tools. The possibility of gap closure (including partial contact between rough surfaces) and reopening during burnup is allowed. The non-linear differential equations are integrated by the finite element method in two-dimensions assuming cylindrical symmetry. Good results are obtained for the simulation of several irradiation tests.

  9. Simulation with DIONISIO 1.0 of thermal and mechanical pellet-cladding interaction in nuclear fuel rods

    Science.gov (United States)

    Soba, Alejandro; Denis, Alicia

    2008-02-01

    The code DIONISIO 1.0 describes most of the main phenomena occurring in a fuel rod throughout its life under normal operation conditions of a nuclear thermal reactor. Starting from the power history, DIONISIO predicts the temperature distribution in the domain, elastic and plastic stress and strain, creep, swelling and densification, release of fission gases, caesium and iodine to the rod free volume, gas mixing, pressure increase, restructuring and grain growth in the UO 2 pellet, irradiation growth of the Zircaloy cladding, oxide layer growth on its surface, hydrogen uptake and the effects of a corrosive atmosphere either internal or external. In particular, the models of thermal conductance of the gap and of pellet-cladding mechanical interaction incorporated to the code constitute two realistic tools. The possibility of gap closure (including partial contact between rough surfaces) and reopening during burnup is allowed. The non-linear differential equations are integrated by the finite element method in two-dimensions assuming cylindrical symmetry. Good results are obtained for the simulation of several irradiation tests.

  10. Development of nuclear fuel rod inspection technique using ultrasonic resonance phenomenon

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Myoung Seon; Joo, Young Sang; Jung, Hyun Kyu; Cheong, Yong Moo

    1997-02-01

    The scattering of plane acoustic waves normally incident on a multilayered cylindrical shell has been formulated using the global matrix approach. And a simple way to formulate the non-resonant background component in the field scattered by an empty elastic shell has been found. This is to replace the surface admittance for the shell with the zero-frequency limit of the surface admittance for the analogous fluid shell (i.e., the shear wave speed in the elastic shell is set to zero). It has been shown that the background thus obtained is exact and applicable to shells of arbitrary thickness and material makeup, and over all frequencies and mode numbers. This way has been also applied to obtain the expressions of the backgrounds for multilayered shells. The resonant ultrasound spectroscopy system has been constructed to measure the resonance spectrum of a single fuel rod. The leak-defective fuel rod detection system of a laboratory scale has been also constructed. Particularly, all techniques and processes necessary for manufacturing the ultrasonic probe of thin (1.2 mm) strip type have been developed. (author). 38 refs., 34 figs.

  11. CONTROL ROD

    Science.gov (United States)

    Zinn, W.H.; Ross, H.V.

    1958-11-18

    A control rod is described for a nuclear reactor. In certaln reactor designs it becomes desirable to use a control rod having great width but relatively llttle thickness. This patent is addressed to such a need. The neutron absorbing material is inserted in a triangular tube, leaving volds between the circular insert and the corners of the triangular tube. The material is positioned within the tube by the use of dummy spacers to achleve the desired absorption pattern, then the ends of the tubes are sealed with suitable plugs. The tubes may be welded or soldered together to form two flat surfaces of any desired width, and covered with sheetmetal to protect the tubes from damage. This design provides a control member that will not distort under the action of outside forces or be ruptured by gases generated within the jacketed control member.

  12. Fluid structure interaction between rods and a cross flow - Numerical approach

    Energy Technology Data Exchange (ETDEWEB)

    Simoneau, Jan-patrice, E-mail: jan-patrice.simoneau@areva.com [Areva, 10, Rue J. Recamier, F 69456 Cedex 06, Lyon (France); Sageaux, Thomas, E-mail: thomas.sageaux@areva.com [Areva, 10, Rue J. Recamier, F 69456 Cedex 06, Lyon (France); Moussallam, Nadim, E-mail: nadim.moussallam@areva.com [Areva, 10, Rue J. Recamier, F 69456 Cedex 06, Lyon (France); Bernard, Olivier, E-mail: olivier.bernard1@areva.com [Areva, 1, Place J. Millet, F 92084 Paris la Defense (France)

    2011-11-15

    This paper presents a full coupled approach between fluid dynamics and structure analysis. It is conducted in order to further improve the assessment of fluid structure interaction problems, occurring in the nuclear field such as the behavior of PWR fuel rods, steam generators and other heat exchangers tubes, fast breeder fuel assemblies. The coupling is obtained by implementing a beam mechanical model in user routines of the CFD code Star-CD, and thanks to a moving grid procedure. The configurations considered are rods in a cross flow. The model is first validated on a single rod case. The lock-in effect is pointed out and both amplitude and frequency responses of the single rod are positively compared to experimental data. Secondly, the mutual influence of two rods, either in-line or parallely set, is investigated. Different behaviors, bounded by critical distances between the rods are highlighted. Finally, the stability of a 3 Multiplication-Sign 3 bundle is calculated for different impinging velocities. Stable and unstable areas are found when varying the impinging velocity. Above a limit, the vibrations amplify up to a contact between rods, this bound is found slightly greater than literature values for close configurations. It is therefore expected that further calculations, with model refinements, will bring valuable informations about bundle stability.

  13. Determination of internal pressure and the backfill gas composition of nuclear fuel rods; Determinacion de la presion interna y la composicion del gas de llenado de barras de combustible nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Garcia C, M.A.; Cota S, G.; Merlo S, L.; Fernandez T, F. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1997-07-01

    An important consideration in the nuclear fuel manufacturing is the measurement of the helium atmosphere pressure and its composition analysis inside the nuclear fuel rod. In this work it is presented a system used to measure the internal pressure and to determine the backfill gas composition of fuel rods. The system is composed of an expansion chamber provided of a seals system to assure that when rod is drilled, the gas stays contained inside the expansion chamber. The system is connected to a pressure measurement digital system: Baratron MKS 310-AHS-1000. Range 1000 mm Hg from which the pressure readings are taken when this is stabilized in all the system. After a gas sample is sent toward a Perkin Elmer gas chromatograph, model 8410 with thermal conductivity detector to get the corresponding chromatogram and doing the necessary calculations for obtaining the backfill gas composition of the rod in matter. (Author)

  14. Development of Design Technology on Thermal-Hydraulic Performance in Tight-Lattice Rod Bundles: III - Numerical Evaluation of Fluid Mixing Phenomena using Advanced Interface-Tracking Method -

    Science.gov (United States)

    Yoshida, Hiroyuki; Nagayoshi, Takuji; Takase, Kazuyuki; Akimoto, Hajime

    Thermal-hydraulic design of the current boiling water reactor (BWR) is performed by correlations with empirical results of actual-size tests. However, for the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) core, an actual size test of an embodiment of its design is required to confirm or modify such correlations. Development of a method that enables the thermal-hydraulic design of nuclear reactors without these actual size tests is desired, because these tests take a long time and entail great cost. For this reason we developed an advanced thermal-hydraulic design method for FLWRs using innovative two-phase flow simulation technology. In this study, detailed Two-Phase Flow simulation code using advanced Interface Tracking method: TPFIT is developed to calculate the detailed information of the two-phase flow. We tried to verify the TPFIT code by comparing it with the 2-channel air-water and steam-water mixing experimental results. The predicted result agrees well the observed results and bubble dynamics through the gap and cross flow behavior could be effectively predicted by the TPFIT code, and pressure difference between fluid channels is responsible for the fluid mixing.

  15. Wavelength dependent neutron transmission and radiography investigations of the high temperature behaviour of materials applied in nuclear fuel and control rod claddings

    Science.gov (United States)

    Grosse, M.; Steinbrueck, M.; Kaestner, A.

    2011-09-01

    Neutron radiography was used for the investigation of the nuclear fuel and control rod cladding behaviour during steam oxidation under severe nuclear accident conditions. In order to verify the hypothesis that the unexpectedly high neutron cross-section found after oxidation of Zircaloy-4 in wet air containing 10% steam is caused by a strong hydrogen uptake, the wavelength dependence of the total macroscopic neutron cross-section of the specimens was measured. The characteristic dependence for hydrogen was not found, which is a proof that hydrogen is not absorbed significantly. The data agree mostly with the behaviour expected for β-Zr. Examinations of control rod simulators annealed until the failure in single-rod tests were performed. In order to separate the effect of the neutron absorber and control rod structure materials, radiographs taken with different neutron spectra were combined. This procedure clearly showed that the local melting resulting from the eutectic reaction between the stainless steel control rod cladding and the Zircaloy-4 guide tube is the reason for the failure.

  16. Wavelength dependent neutron transmission and radiography investigations of the high temperature behaviour of materials applied in nuclear fuel and control rod claddings

    Energy Technology Data Exchange (ETDEWEB)

    Grosse, M., E-mail: Mirco.Grosse@KIT.edu [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany); Steinbrueck, M. [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany); Kaestner, A. [Department of Spallation Source, Paul Scherrer Institute (PSI), CH-5232 Villigen (Switzerland)

    2011-09-21

    Neutron radiography was used for the investigation of the nuclear fuel and control rod cladding behaviour during steam oxidation under severe nuclear accident conditions. In order to verify the hypothesis that the unexpectedly high neutron cross-section found after oxidation of Zircaloy-4 in wet air containing 10% steam is caused by a strong hydrogen uptake, the wavelength dependence of the total macroscopic neutron cross-section of the specimens was measured. The characteristic dependence for hydrogen was not found, which is a proof that hydrogen is not absorbed significantly. The data agree mostly with the behaviour expected for {beta}-Zr. Examinations of control rod simulators annealed until the failure in single-rod tests were performed. In order to separate the effect of the neutron absorber and control rod structure materials, radiographs taken with different neutron spectra were combined. This procedure clearly showed that the local melting resulting from the eutectic reaction between the stainless steel control rod cladding and the Zircaloy-4 guide tube is the reason for the failure.

  17. Morphological analysis of zirconium nuclear fuel retaining rods braided with SiC: Quality assurance and defect identification

    Science.gov (United States)

    Glazoff, Michael V.; Hiromoto, Robert; Tokuhiro, Akira

    2014-08-01

    In the after-Fukushima world, the stability of materials under extreme conditions is an important issue for the safety of nuclear reactors. Among the methods explored currently to improve zircaloys’ thermal stability in off-normal conditions, using a protective coat of the SiC filaments is considered because silicon carbide is well known for its remarkable chemical inertness at high temperatures. A typical SiC fiber contains ∼50,000 individual filaments of 5-10 μm in diameter. In this paper, an effort was made to develop and apply mathematical morphology to the process of automatic defect identification in Zircaloy-4 rods braided with the protective layer of the silicon carbide filament. However, the issues of the braiding quality have to be addressed to ensure its full protective potential. We present the original mathematical morphology algorithms that allow solving this problem of quality assurance successfully. In nuclear industry, such algorithms are used for the first time, and could be easily generalized to the case of automated continuous monitoring for defect identification in the future.

  18. Hydraulic characteristics of HANARO fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Cho, S.; Chung, H. J.; Chun, S. Y.; Yang, S. K.; Chung, M. K. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    This paper presents the hydraulic characteristics measured by using LDV (Laser Doppler Velocimetry) in subchannels of HANARO, KAERI research reactor, fuel bundle. The fuel bundle consists of 18 axially finned rods with 3 spacer grids, which are arranged in cylindrical configuration. The effects of the spacer grids on the turbulent flow were investigated by the experimental results. Pressure drops for each component of the fuel bundle were measured, and the friction factors of fuel bundle and loss coefficients for the spacer grids were estimated from the measured pressure drops. Implications regarding the turbulent thermal mixing were discussed. Vibration test results measured by using laser vibrometer were presented. 9 refs., 12 figs. (Author)

  19. Thermal Hydraulic Performance of Tight Lattice Bundle

    Science.gov (United States)

    Yamamoto, Yasushi; Akiba, Miyuki; Morooka, Shinichi; Shirakawa, Kenetsu; Abe, Nobuaki

    Recently, the reduced moderation spectrum BWR has been studied. The fast neutron spectrum is obtained through triangular tight lattice fuel. However, there are few thermal hydraulic test data and thermal hydraulic correlation applicable to critical power prediction in such a tight lattice bundle. This study aims to enhance the database of the thermal hydraulic performance of the tight lattice bundle whose rod gap is about 1mm. Therefore, thermal hydraulic performance measurement tests of tight lattice bundles for the critical power, the pressure drop and the counter current flow limiting were performed. Moreover, the correlations to evaluate the thermal-hydraulic performance of the tight lattice bundle were developed.

  20. SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Benedetti, R. L.; Lords, L. V.; Kiser, D. M.

    1978-02-01

    The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocity and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage.

  1. The Recovery of the Metal Insulation Cable in the Instrumentation of Nuclear Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang Young; Ahn, Sung Ho; Sim, Bong Sik; Lee, Chul Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    Mineral-insulated (MI) cables are widely used to prolong the instrumentation cable of instruments such as a thermocouple (TC), linear variable differential transformer (LVDT) and self-powered neutron detector (SPND), which are used to measure various irradiation characteristics of nuclear fuels and materials. MI cables are expected to be helpful for instrumentation of nuclear fuel and material irradiation because of their high electrical insulation, heat resistance and mechanical strength. The MI cable used to extend thermocouple wires is classified as the following: 1) For common metal types of thermocouples, the thermocouple extension wire is of substantially the same composition as the corresponding thermocouple type and it can offer advantages in cost or mechanical properties when used for the connection between a thermocouple and instruments. 2) For noble metal types of thermocouples, the thermocouple compensation wire is an entirely different alloy formulated to match the noble metal characteristics, which is necessary due to the high cost of noble metals. During the installation of an instrument, an MI cable damaged by impact must be recovered because it is difficult to change the entire thermocouple. And for MI cable recovery, it is necessary to develop the instrumentation technology of FTL. This paper described the experimental results of MI cable recovery, which consists of a removal test of the MI cable sheath and a joining test of the compensation of the wire and MI cable/ wire/compensation wire and sheath of MI cable/bushing, for carrying out irradiation tests of nuclear fuel and materials in the FTL facility of HANARO

  2. Optimum nuclear design of target fuel rod for Mo-99 production in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myung Hyun [Kyung Hee University, Seoul (Korea)

    1998-04-01

    Nuclear target design for Mo-99 production in HANARO was performed, KAERI proposed target design was analyzed and its feasibility was shown. Three commercial target designs of Cintichem, ANL and KAERI were tested for the HANARO irradiation an d they all satisfied with design specification. A parametric study was done for target design options and Mo-99 yields ratio and surface heat flux were compared. Tested parameters were target fuel thickness, irradiation location, target axial length, packing density of powder fuel, size of target radius, target geometry, fuel enrichment, fuel composition, and cladding material. Optimized target fuel was designed for both LEU and HEU options. (author). 17 refs., 33 figs., 42 tabs.

  3. The Sloping Scram Time Simulation of the Nuclear Reactor Control Rod%核反应堆控制棒倾斜落棒时程模拟

    Institute of Scientific and Technical Information of China (English)

    许珍; 白立新; 代飞

    2015-01-01

    事故情况下控制棒组件会在自身重力作用下下落,下落时间是衡量核电站安全运行的重要参数之一。通过建立落棒物理模型,分析落棒的物理过程以及控制棒下落过程中的各阶段的受力情况,得到了落棒控制方程。采用VS2010软件对控制方程数值求解,完成了控制棒组件竖直及倾斜下落过程的时程的分析,对比了倾斜落棒角度对落棒时程的影响,得出最大倾斜落棒角度,为反应堆控制棒组件的设计提供依据。%Control rod assembly will drop because of its own gravity under accident conditions , and the dropping time is one of the important parameters of the safe operation of nuclear power plant .The mathematical model of rod dropping process is established by analyzing the physical process of rod drop and the force at each stage .U-sing VS2010 to solve control equation , the vertical and inclined dropping time of the control rod component and how the angle of tilt rod drop influences the dropping time have been analyzed .The largest angle of tilt rod drop is obtained , which can provide reference for the design of the reactor control rod components .

  4. Experimental study of water flow in nuclear fuel elements; Estudo experimental do escoamento de agua em elementos combustiveis nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, Lorena Escriche, E-mail: ler@cdtn.br [Centro Federal de Educacao Tecnologica de Minas Gerais (CEFET), Belo Horizonte, MG (Brazil); Rezende, Hugo Cesar; Mattos, Joao Roberto Loureiro de; Barros Filho, Jose Afonso; Santos, Andre Augusto Campagnole dos, E-mail: hcr@cdtn.br, E-mail: jrmattos@cdtn.br, E-mail: jabf@cdtn.br, E-mail: aacs@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2013-07-01

    This work aims to develop an experimental methodology for investigating the water flow through rod bundles after spacer grids of nuclear fuel elements of PWR type reactors. Speed profiles, with the device LDV (Laser Doppler Velocimetry), and the pressure drop between two sockets located before and after the spacer grid, using pressure transducers were measured.

  5. Coolability of ballooned VVER bundles with pellet relocation

    Energy Technology Data Exchange (ETDEWEB)

    Hozer, Z.; Nagy, I.; Windberg, P.; Vimi, A. [AEKI, P.O.box 49, Budapest, H-1525 (Hungary)

    2009-06-15

    During a LOCA incident the high pressure in the fuel rods can lead to clad ballooning and the debris of fuel pellets can fill the enlarged volume. The evaluation of the role of these two effects on the coolability of VVER type fuel bundles was the main objective of the experimental series. The tests were carried out in the modified configuration of the CODEX facility. 19-rod electrically heated VVER type bundle was used. The test section was heated up to 600 deg. C in steam atmosphere and the bundle was quenched from the bottom by cold water. Three series of tests were performed: 1. Reference bundle with fuel rods without ballooning, with uniform power profile. 2. Bundle with 86% blockage rate and with uniform power profile. The blockage rate was reached by superimposing hollow sleeves on all 19 fuel rods. 3. Bundle with 86% blockage rate and with local power peak in the ballooned area. The local power peak was produced by the local reduction the cross section of the internal heater bar inside of the fuel rods. In all three bundle configurations three different cooling water flow-rates were applied. The experimental results confirmed that a VVER bundle with even 86% blockage rate remains coolable after a LOCA event. The ballooned section creates some obstacles for the cooling water during reflood of the bundle, but this effect causes only a short delay in the cooling down of the hot fuel rods. Earlier tests on the coolability of ballooned bundles were performed only with Western type bundles with square fuel lattice. The present test series was the first confirmation of the coolability of VVER type bundles with triangular lattice. The accumulation of fuel pellet debris in the ballooned volume results in a local power peak, which leads to further slowing down of quench front. The first tests indicated that the effect of local power peak was less significant on the delay of cooling down than the effect of ballooning. (authors)

  6. Rod differentiation factor NRL activates the expression of nuclear receptor NR2E3 to suppress the development of cone photoreceptors.

    Science.gov (United States)

    Oh, Edwin C T; Cheng, Hong; Hao, Hong; Jia, Lin; Khan, Naheed Wali; Swaroop, Anand

    2008-10-21

    Neural developmental programs require a high level of coordination between the decision to exit cell cycle and acquisition of cell fate. The Maf-family transcription factor NRL is essential for rod photoreceptor specification in the mammalian retina as its loss of function converts rod precursors to functional cones. Ectopic expression of NRL or a photoreceptor-specific orphan nuclear receptor NR2E3 completely suppresses cone development while concurrently directing the post-mitotic photoreceptor precursors towards rod cell fate. Given that NRL and NR2E3 have overlapping functions and NR2E3 expression is abolished in the Nrl(-/-) retina, we wanted to clarify the distinct roles of NRL and NR2E3 during retinal differentiation. Here, we demonstrate that NRL binds to a sequence element in the Nr2e3 promoter and enhances its activity synergistically with the homeodomain protein CRX. Using transgenic mice, we show that NRL can only partially suppress cone development in the absence of NR2E3. Gene profiling of retinas from transgenic mice that ectopically express NR2E3 or NRL in cone precursors reveals overlapping and unique targets of these two transcription factors. Together with previous reports, our findings establish the hierarchy of transcriptional regulators in determining rod versus cone cell fate in photoreceptor precursors during the development of mammalian retina.

  7. Hydraulic Actuator for Ganged Control Rods

    Science.gov (United States)

    Thompson, D. C.; Robey, R. M.

    1986-01-01

    Hydraulic actuator moves several nuclear-reactor control rods in unison. Electromagnetic pump pushes liquid lithium against ends of control rods, forcing them out of or into nuclear reactor. Color arrows show lithium flow for reactor startup and operation. Flow reversed for shutdown. Conceived for use aboard spacecraft, actuator principle applied to terrestrial hydraulic machinery involving motion of ganged rods.

  8. 2D model for melt progression through rods and debris

    Energy Technology Data Exchange (ETDEWEB)

    Fichot, F. [IPSN/DRS, CEA Cadarache, St. Paul-lez-Durance (France)

    2001-07-01

    During the degradation of a nuclear core in a severe accident scenario, the high temperatures reached lead to the melting of materials. The formation of liquid mixtures at various elevations is followed by the flow of molten materials through the core. Liquid mixture may flow under several configurations: axial relocation along the rods, horizontal motion over a plane surface such as the core support plate or a blockage of material, 2D relocation through a debris bed, etc.. The two-dimensional relocation of molten material through a porous debris bed, implemented for the simulation of late degradation phases, has opened a new way to the elaboration of the relocation model for the flow of liquid mixture along the rods. It is based on a volume averaging method, where wall friction and capillary effects are taken into account by introducing effective coefficients to characterize the solid matrix (rods, grids, debris, etc.). A local description of the liquid flow is necessary to derive the effective coefficients. Heat transfers are modelled in a similar way. The derivation of the conservation equations for the liquid mixture falling flow (momentum) in two directions (axial and radial-horizontal) and for the heat exchanges (energy) are the main points of this new model for simulating melt progression. In this presentation, the full model for the relocation and solidification of liquid materials through a rod bundle or a debris bed is described. It is implemented in the ICARE/CATHARE code, developed by IPSN in Cadarache. The main improvements and advantages of the new model are: A single formulation for liquid mixture relocation, in 2D, either through a rod bundle or a porous debris bed, Extensions to complex structures (grids, by-pass, etc..), The modeling of relocation of a liquid mixture over plane surfaces. (author)

  9. Analysis of high burnup fuel behavior under control rod ejection accident in Korea standard nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bok; Lee, Chung Chan; Kim, Oh Hwan; Kim, Jong Jin [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-07-01

    Test results of high burnup fuel behavior under RIA(reactivity insertion accident) indicated that fuel might fail at the fuel enthalpy lower than that in the current fuel failure criteria was derived by the conservative assumptions and analysis of fuel failure mechanisms, and applied to the analysis of control rod ejection accident in the 1,000 MWe Korea standard PWR. Except that three dimensional core analysis was performed instead of conventional zero dimensional analysis, all the other conservative assumptions were kept. Analysis results showed that less than on percent of the fuel rods in the core has failed which was much less than the conventional fuel failure fraction, 9.8 %, even though a newly derived fuel failure criteria -Fuel failure occurs at the power level lower than that in the current fuel failure criteria. - was applied, since transient fuel rod power level was significantly decreased by analyzing the transient fuel rod power level was significantly decreased by analyzing the transient core three dimensionally. Therefore, it can be said that results of the radiological consequence analysis for the control rod ejection accident in the FSAR where fuel failure fraction was assumed 9.8 % is still bounding. 18 tabs., 48 figs., 39 refs. (Author).

  10. Aerosol behavior during SIC control rod failure in QUENCH-13 test

    Energy Technology Data Exchange (ETDEWEB)

    Lind, Terttaliisa, E-mail: terttaliisa.lind@psi.c [Paul Scherrer Institut, Villigen (Switzerland); Csordas, Anna Pinter; Nagy, Imre [HAS KFKI Atomic Energy Research Institute, Budapest (Hungary); Stuckert, Juri [Forschungszentrum Karlsruhe, Karlsruhe (Germany)

    2010-02-15

    In a nuclear reactor severe accident, radioactive fission products as well as structural materials are released from the core by evaporation, and the released gases form particles by nucleation and condensation. In addition, aerosol particles may be generated by droplet formation and fragmentation of the core. In pressurized water reactors (PWR), a commonly used control rod material is silver-indium-cadmium (SIC) covered with stainless steel cladding. The control rod elements, Cd, In and Ag, have relatively low melting temperatures, and especially Cd has also a very low boiling point. Control rods are likely to fail early on in the accident due to melting of the stainless steel cladding which can be accelerated by eutectic interaction between stainless steel and the surrounding Zircaloy guide tube. The release of the control rod materials would follow the cladding failure thus affecting aerosol source term as well as fuel rod degradation. The QUENCH experimental program at Forschungszentrum Karlsruhe investigates phenomena associated with reflood of a degrading core under postulated severe accident conditions. QUENCH-13 test was the first in this program to include a silver-indium-cadmium control rod of prototypic PWR design. To characterize the extent of aerosol release during the control rod failure, aerosol particle size distribution and concentration measurements in the off-gas pipe of the QUENCH facility were carried out. For the first time, it was possible to determine on-line the aerosol concentration and size distribution released from the core. These results are of prime importance for model development for the proper calculation of the source term resulting from control rod failure. The on-line measurement showed that the main aerosol release started at the bundle temperature maximum of T approx 1570 K at hottest bundle elevation. A very large burst of aerosols was detected 660 s later at the bundle temperature maximum of T approx 1650 K, followed by a

  11. Aerosol behavior during SIC control rod failure in QUENCH-13 test

    Science.gov (United States)

    Lind, Terttaliisa; Csordás, Anna Pintér; Nagy, Imre; Stuckert, Juri

    2010-02-01

    In a nuclear reactor severe accident, radioactive fission products as well as structural materials are released from the core by evaporation, and the released gases form particles by nucleation and condensation. In addition, aerosol particles may be generated by droplet formation and fragmentation of the core. In pressurized water reactors (PWR), a commonly used control rod material is silver-indium-cadmium (SIC) covered with stainless steel cladding. The control rod elements, Cd, In and Ag, have relatively low melting temperatures, and especially Cd has also a very low boiling point. Control rods are likely to fail early on in the accident due to melting of the stainless steel cladding which can be accelerated by eutectic interaction between stainless steel and the surrounding Zircaloy guide tube. The release of the control rod materials would follow the cladding failure thus affecting aerosol source term as well as fuel rod degradation. The QUENCH experimental program at Forschungszentrum Karlsruhe investigates phenomena associated with reflood of a degrading core under postulated severe accident conditions. QUENCH-13 test was the first in this program to include a silver-indium-cadmium control rod of prototypic PWR design. To characterize the extent of aerosol release during the control rod failure, aerosol particle size distribution and concentration measurements in the off-gas pipe of the QUENCH facility were carried out. For the first time, it was possible to determine on-line the aerosol concentration and size distribution released from the core. These results are of prime importance for model development for the proper calculation of the source term resulting from control rod failure. The on-line measurement showed that the main aerosol release started at the bundle temperature maximum of T ˜ 1570 K at hottest bundle elevation. A very large burst of aerosols was detected 660 s later at the bundle temperature maximum of T ˜ 1650 K, followed by a relatively

  12. Understanding the Atomic-Level Chemistry and Structure of Oxide Deposits on Fuel Rods in Light Water Nuclear Reactors Using First Principles Methods

    Science.gov (United States)

    Rak, Zs.; O'Brien, C. J.; Brenner, D. W.; Andersson, D. A.; Stanek, C. R.

    2016-09-01

    The results of recent studies are discussed in which first principles calculations at the atomic level have been used to expand the thermodynamic database for science-based predictive modeling of the chemistry, composition and structure of unwanted oxides that deposit on the fuel rods in pressurized light water nuclear reactors. Issues discussed include the origin of the particles that make up deposits, the structure and properties of the deposits, and the forms by which boron uptake into the deposits can occur. These first principles approaches have implications for other research areas, such as hydrothermal synthesis and the stability and corrosion resistance of other materials under other extreme conditions.

  13. Development of improved thermal hydraulics and fuel performance technology; development of turbulence model and simulation code for flow analysis in nuclear fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Myung, H. K.; Yang, S. Y.; Kim, B. H.; Song, J. H.; Oh, J. Z. [Kookmin University, Seoul (Korea)

    2002-03-01

    The flow through a nuclear rod bundle with mixing vanes is very complex and so required a suitable turbulence model for its accurate prediction. Subchannel flow in a nuclear bundle having vanes to mix flow appears complex turbulent flow. Objective of this study is to investigate performance of prediction about turbulence model contained in STAR-CD code and to develop suitable turbulence model which can predict complex flow in nuclear assembly. For several nonlinear {kappa}-{epsilon} turbulence models, their performance were investigated in the prediction of the flow in nuclear fuel assembly, and also their problems were discussed in detail. The results obtained from the present research would give a help for the development of turbulence model which can accurately predict the flow through the rod bundles with mixing vanes. 19 refs., 32 figs., 3 tabs. (Author)

  14. Application of IGBT in Rod Control System of Nuclear Power Station%IGBT在核电站棒控系统中的应用

    Institute of Scientific and Technical Information of China (English)

    许育周; 李涛; 王春生

    2014-01-01

    针对目前基于可控硅的核电站棒控系统存在输出电流纹波大、电流上升和下降时间长等特点,提出了基于绝缘栅双模型晶体管( IGBT)的捧控系统。 IGBT作为新型大功率开关器件,具有电压型控制、输入阻抗大、驱动功率小、控制电路简单、开关损耗小、工作频率高和元件容量大等优点。实际应用表明,IGBT减小了棒控系统输出的电流纹波和电流上升、下降时间,十分适合应用于核电站棒控系统中。%At present, most of the rod control system in nuclear power stations are using SCR, so the ripple of output current is big, and the rising time and calling time of the current are long, to overcome this demerits, the rod control system based on IGBT is proposed. IGBT is a new type of large power switching device; it features voltage control, high input impedance, small driving power, simple control circuit, low switching loss, high working frequency and large capacity. Practical application shows that the ripple and current rising and falling time are decreased, this shows that IGBT is suitable for application of rod control system in nuclear power station.

  15. PWR-UO{sub 2} nuclear fuel criticality study: control rod effects on infinite neutron multiplication factor and spent fuel composition

    Energy Technology Data Exchange (ETDEWEB)

    Sousa, R.V.; Pereira, C., E-mail: claubia@nuclear.ufmg.br; Silva, C.A.M.; Costa, A.L.; Veloso, M.A.F.; Oliveira, A.H. de

    2013-10-15

    Highlights: • A three-dimensional model of a PWR fuel were simulated. • Results using TRITON/T6-DEPL module in SCALE 6.0 and two libraries (238 and 44 groups) were compared. • Variations in the infinite neutron multiplication factor and the nuclides concentrations, both under control rod insertion effects were analysed. • Results show very good agreement with those published by OECD. -- Abstract: Deterministic and stochastic nuclear codes are software packages used to perform reactor physics calculations, especially in PWRs, the most common type of nuclear reactor currently in operation. The NEA Expert Group on Burn-up Credit Criticality Safety has published a Benchmark with results obtained from simulations of PWR-UO{sub 2} nuclear fuel. The same simulations were performed at DEN/UFMG with SCALE 6.0, a modular nuclear system code developed by Oak Ridge National Laboratory using two different neutron energy libraries (238 and 44 groups). The results obtained using a three-dimensional model with the T6-DEPL sequence of the TRITON module in SCALE 6.0 for spent fuel inventory and infinite neutron multiplication factor calculations show very good agreement with those published by the OECD. The main goal of this work is to validate the methodology at DEN/UFMG for future use in simulations related to Angra I, II and III Nuclear Power Plants.

  16. Subchannel and Computational Fluid Dynamic Analyses of a Model Pin Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Gairola, A.; Arif, M.; Suh, K. Y. [Seoul National Univ., Seoul (Korea, Republic of)

    2014-05-15

    The current study showed that the simplistic approach of subchannel analysis code MATRA was not good in capturing the physical behavior of the coolant inside the rod bundle. With the incorporation of more detailed geometry of the grid spacer in the CFX code it was possible to approach the experimental values. However, it is vital to incorporate more advanced turbulence mixing models to more realistically simulate behavior of the liquid metal coolant inside the model pin bundle in parallel with the incorporation of the bottom and top grid structures. In the framework of the 11{sup th} international meeting of International Association for Hydraulic Research and Engineering (IAHR) working group on the advanced reactor thermal hydraulics a standard problem was conducted. The quintessence of the problem was to check on the hydraulics and heat transfer in a novel pin bundle with different pitch to rod diameter ratio and heat flux cooled by liquid metal. The standard problem stems from the field of nuclear safety research with the idea of validating and checking the performances of computer codes against the experimental results. Comprehensive checks between the two will succor in improving the dependability and exactness of the codes used for accident simulations.

  17. On Double Vector Bundles

    Institute of Scientific and Technical Information of China (English)

    Zhuo CHEN; Zhang Ju LIU; Yun He SHENG

    2014-01-01

    In this paper, we construct a category of short exact sequences of vector bundles and prove that it is equivalent to the category of double vector bundles. Moreover, operations on double vector bundles can be transferred to operations on the corresponding short exact sequences. In particular, we study the duality theory of double vector bundles in term of the corresponding short exact sequences. Examples including the jet bundle and the Atiyah algebroid are discussed.

  18. On Double Vector Bundles

    OpenAIRE

    Chen, Zhuo; Liu, Zhangju; Sheng, Yunhe

    2011-01-01

    In this paper, we construct a category of short exact sequences of vector bundles and prove that it is equivalent to the category of double vector bundles. Moreover, operations on double vector bundles can be transferred to operations on the corresponding short exact sequences. In particular, we study the duality theory of double vector bundles in term of the corresponding short exact sequences. Examples including the jet bundle and the Atiyah algebroid are discussed.

  19. A particle assembly/constrained expansion (PACE) model for the formation and structure of porous metal oxide deposits on nuclear fuel rods in pressurized light water reactors

    Science.gov (United States)

    Brenner, Donald W.; Lu, Shijing; O'Brien, Christopher J.; Bucholz, Eric W.; Rak, Zsolt

    2015-02-01

    A new model is proposed for the structure and properties of porous metal oxide scales (aka Chalk River Unidentified Deposits (CRUD)) observed on the nuclear fuel rod cladding in Pressurized Water Reactors (PWR). The model is based on the thermodynamically-driven expansion of agglomerated octahedral nickel ferrite particles in response to pH and temperature changes in the CRUD. The model predicts that porous nickel ferrite with internal {1 1 1} surfaces is a thermodynamically stable structure under PWR conditions even when the free energy of formation of bulk nickel ferrite is positive. This explains the pervasive presence of nickel ferrite in CRUD, observed CRUD microstructures, why CRUD maintains its porosity, and variations in porosity within the CRUD observed experimentally. This model is a stark departure from decades of conventional wisdom and detailed theoretical analysis of CRUD chemistry, and defines new research directions for model validation, and for understanding and ultimately controlling CRUD formation.

  20. Principal noncommutative torus bundles

    DEFF Research Database (Denmark)

    Echterhoff, Siegfried; Nest, Ryszard; Oyono-Oyono, Herve

    2008-01-01

    In this paper we study continuous bundles of C*-algebras which are non-commutative analogues of principal torus bundles. We show that all such bundles, although in general being very far away from being locally trivial bundles, are at least locally trivial with respect to a suitable bundle version...... of bivariant K-theory (denoted RKK-theory) due to Kasparov. Using earlier results of Echterhoff and Williams, we shall give a complete classification of principal non-commutative torus bundles up to equivariant Morita equivalence. We then study these bundles as topological fibrations (forgetting the group...... action) and give necessary and sufficient conditions for any non-commutative principal torus bundle being RKK-equivalent to a commutative one. As an application of our methods we shall also give a K-theoretic characterization of those principal torus-bundles with H-flux, as studied by Mathai...

  1. Application of the porous medium heat transfer model of ICARE/CATHARE code against debris bed and 'bundle' experiments

    Energy Technology Data Exchange (ETDEWEB)

    Repetto, G. [CEA Cadarache, Institut de Radioprotection et de Surete Nucleaire, DPAM, 13 - Saint-Paul-lez-Durance (France); Ederli, St. [Ente per le Nuove Technologie, l' Energia e l' Ambiente (ENEA) (Italy)

    2007-07-01

    ICARE/CATHARE code is developed by the 'Institut de Radioprotection et de Surete Nucleaire' to simulate Nuclear Reactor behaviour during the course of a Loss of Cooling accident up to the core melting. The assessment of the heat transfer model in porous medium has been performed against experiments performed in ACRR (SNL-USA) and in Phebus reactors (at Cadarache - France). Calculation versus experiment results indicate a good agreement for the thermal behaviour. The heat transfers inside solid debris bed can be well predicted using the Imura-Yagi correlation to calculate the debris bed equivalent thermal conductivity in a wide range of particles size. In the case of 'Rod like geometry' calculations, the fuel rod assembly was modelled assuming several rings of fuel rods, with heat transfer including radiative phenomena using view factors between rods. An alternative modelling has been used considering the fuel rods as a porous medium with with pure UO{sub 2} spherical particles of 1 cm diameter and a total porosity representative of the fuel bundle inside a cylindrical shroud. With this approach (heat exchanges accounted for with the Imura-Yagi correlation), the radial gradient calculated in a small bundle was significantly increased, from a few degrees (with the previous modelling) to about 150/200 K at 2273 K. This modelling has been recently improved, to account for the heat transfer inside a fuel rod bundle, by a specific model based on an electrical analogy, considering the porous medium as a cluster of true cylinders. (authors)

  2. Results of Post Irradiation Examinations of VVER Leaky Rods

    Energy Technology Data Exchange (ETDEWEB)

    Markov, D.; Perepelkin, S.; Polenok, V.; Zhitelev, V.; Mayorshina, G. [Head of Fuel Research Department, JSC ' SSC RIAR' , 433510, Dimitrovgrad-10, Ulyanovsk region (Russian Federation)

    2009-06-15

    The most important requirement imposed on fuel elements is to maintain integrity of fuel rod claddings under operation, storage and transportation, since it is directly related to the operational safety. However, failed rod claddings are sometimes observed under reactor operation. Identification and unloading of fuel assemblies with leaky rods from VVER is available only at the time of planned preventive maintenance. An unscheduled reactor shutdown due to the excess of coolant activity limit as well as a preterm unloading of the fuel assembly cause economic damage to nuclear plant. Therefore, models and calculation codes were developed to forecast coolant contamination and failed fuel rod behavior. Criteria based on calculations were set to determine the admissible number of the failed rods in core and the opportunity to continue the reactor operation or pre-term unloading of the fuel assembly with the failed rods. Nevertheless, to prevent the fuel rod failure (for unfailing operation) it is necessary to reveal disadvantages of the design, fabrication method and fuel operation conditions, and to eliminate defects. The most complete and significant information about spent fuel assemblies may be received following the post irradiation material examinations. In order to reveal failure origins and mechanism of changes in VVER fuel and failed rod cladding condition depending on the operation, the examinations of 12 VVER-1000 fuel assemblies and 3 VVER-440 fuel assemblies, operated under normal conditions up to the fuel burnup 13..47 MWd/kgU were carried out. To evaluate the rod cladding condition, reveal defects and determine their parameters, the ultrasonic control of cladding integrity, surface visual inspection, eddy current defectoscopy, measurement of geometrical parameters were applied. In separate cases we used the metallography, measured the hydrogen percentage and carried out the mechanical tests of o-ring samples. The pellet condition was evaluated in

  3. Simulation of Thermopower Influence on Fuel Core of Power Rod in Nuclear Power Plant (NPP Active Zone

    Directory of Open Access Journals (Sweden)

    I. S. Kulikov

    2010-01-01

    Full Text Available The paper considers problems of modern methods for  calculation of designs and materials of nuclear power. A model of numerical analysis for stress-strain state of fuel pins in the NPP active zone is proposed in the paper. The paper contains simulation concerning a fuel core section of a nuclear reactor heat-generating element with subsequent solution of a temperature and thermoelastic problem in computer program complex FEA ANSYS Workbench 11.0. All the obtained results have passed through checking procedure.

  4. Laser Ultrasonic System for Surface Crack Visualization in Dissimilar Welds of Control Rod Drive Mechanism Assembly of Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Yun-Shil Choi

    2014-01-01

    Full Text Available In this paper, we propose a J-groove dissimilar weld crack visualization system based on ultrasonic propagation imaging (UPI technology. A full-scale control rod drive mechanism (CRDM assembly specimen was fabricated to verify the proposed system. An ultrasonic sensor was contacted at one point of the inner surface of the reactor vessel head part of the CRDM assembly. Q-switched laser beams were scanned to generate ultrasonic waves around the weld bead. The localization and sizing of the crack were possible by ultrasonic wave propagation imaging. Furthermore, ultrasonic spectral imaging unveiled frequency components of damage-induced waves, while wavelet-transformed ultrasonic propagation imaging enhanced damage visibility by generating a wave propagation video focused on the frequency component of the damage-induced waves. Dual-directional anomalous wave propagation imaging with adjacent wave subtraction was also developed to enhance the crack visibility regardless of crack orientation and wave propagation direction. In conclusion, the full-scale specimen test demonstrated that the multiple damage visualization tools are very effective in the visualization of J-groove dissimilar weld cracks.

  5. Morphoelastic rods

    CERN Document Server

    Tiero, Alessandro

    2014-01-01

    We propose a mechanical theory describing elastic rods which, like plant organs, can grow and can change their intrinsic curvature and torsion. The equations ruling accretion and remodeling are obtained by combining balance laws involving non-standard forces with constitutive prescriptions filtered by a dissipation principle that takes into account both standard and non-standard working.

  6. Development of burnup dependent fuel rod model in COBRA-TF

    Science.gov (United States)

    Yilmaz, Mine Ozdemir

    The purpose of this research was to develop a burnup dependent fuel thermal conductivity model within Pennsylvania State University, Reactor Dynamics and Fuel Management Group (RDFMG) version of the subchannel thermal-hydraulics code COBRA-TF (CTF). The model takes into account first, the degradation of fuel thermal conductivity with high burnup; and second, the fuel thermal conductivity dependence on the Gadolinium content for both UO2 and MOX fuel rods. The modified Nuclear Fuel Industries (NFI) model for UO2 fuel rods and Duriez/Modified NFI Model for MOX fuel rods were incorporated into CTF and fuel centerline predictions were compared against Halden experimental test data and FRAPCON-3.4 predictions to validate the burnup dependent fuel thermal conductivity model in CTF. Experimental test cases from Halden reactor fuel rods for UO2 fuel rods at Beginning of Life (BOL), through lifetime without Gd2O3 and through lifetime with Gd 2O3 and a MOX fuel rod were simulated with CTF. Since test fuel rod and FRAPCON-3.4 results were based on single rod measurements, CTF was run for a single fuel rod surrounded with a single channel configuration. Input decks for CTF were developed for one fuel rod located at the center of a subchannel (rod-centered subchannel approach). Fuel centerline temperatures predicted by CTF were compared against the measurements from Halden experimental test data and the predictions from FRAPCON-3.4. After implementing the new fuel thermal conductivity model in CTF and validating the model with experimental data, CTF model was applied to steady state and transient calculations. 4x4 PWR fuel bundle configuration from Purdue MOX benchmark was used to apply the new model for steady state and transient calculations. First, one of each high burnup UO2 and MOX fuel rods from 4x4 matrix were selected to carry out single fuel rod calculations and fuel centerline temperatures predicted by CTF/TORT-TD were compared against CTF /TORT-TD /FRAPTRAN

  7. Operational modal analysis of flow-induced vibration of nuclear fuel rods in a turbulent axial flow

    Energy Technology Data Exchange (ETDEWEB)

    De Pauw, B., E-mail: bdepauw@vub.ac.be [Vrije Universiteit Brussel (VUB), Brussels Photonics Team (B-Phot), Brussels (Belgium); Vrije Universiteit Brussel (VUB), Department of Mechanical Engineering (AVRG), Brussels (Belgium); Belgian Nuclear Research Centre (SCK-CEN), Boeretang 200, Mol (Belgium); Weijtjens, W.; Vanlanduit, S. [Vrije Universiteit Brussel (VUB), Department of Mechanical Engineering (AVRG), Brussels (Belgium); Van Tichelen, K. [Belgian Nuclear Research Centre (SCK-CEN), Boeretang 200, Mol (Belgium); Berghmans, F. [Vrije Universiteit Brussel (VUB), Brussels Photonics Team (B-Phot), Brussels (Belgium)

    2015-04-01

    Highlights: • We describe an analysis technique to evaluate nuclear fuel pins. • We test a single fuel pin mockup subjected to turbulent axial flow. • Our analysis is based on operational modal analysis (OMA). • The accuracy and precision of our method is higher compared to traditional methods. • We demonstrate the possible onset of a fluid-elastic instability. - Abstract: Flow-induced vibration of nuclear reactor fuel pins can result in mechanical noise and lead to failure of the reactor's fuel assembly. This problem can be exacerbated in the new generation of liquid heavy metal fast reactors that use a much denser and more viscous coolant in the reactor core. An investigation of the flow-induced vibration in these particular conditions is therefore essential. In this paper, we describe an analysis technique to evaluate flow-induced vibration of nuclear reactor fuel pins subjected to a turbulent axial flow of heavy metal. We deal with a single fuel pin mockup designed for the lead–bismuth eutectic (LBE) cooled MYRRHA reactor which is subjected to similar flow conditions as in the reactor core. Our analysis is based on operational modal analysis (OMA) techniques. We show that the accuracy and precision of our OMA technique is higher compared to traditional methods and that it allows evaluating the evolution of modal parameters in operational conditions. We also demonstrate the possible onset of a fluid-elastic instability by tracking the modal parameters with increasing flow velocity.

  8. Experimental investigation of the coolability of blocked hexagonal bundles

    Energy Technology Data Exchange (ETDEWEB)

    Hózer, Zoltán, E-mail: zoltan.hozer@energia.mta.hu; Nagy, Imre; Kunstár, Mihály; Szabó, Péter; Vér, Nóra; Farkas, Róbert; Trosztel, István; Vimi, András

    2017-06-15

    Highlights: • Experiments were performed with electrically heated hexagonal fuel bundles. • Coolability of ballooned VVER-440 type bundle was confirmed up to high blockage rate. • Pellet relocation effect causes delay in the cool-down of the bundle. • The bypass line does not prevent the reflood of ballooned fuel rods. - Abstract: The CODEX-COOL experimental series was carried out in order to evaluate the effect of ballooning and pellet relocation in hexagonal bundles on the coolability of fuel rods after a LOCA event. The effects of blockage geometry, coolant flowrate, initial temperature and axial profile were investigated. The experimental results confirmed that a VVER bundle up to 80% blockage rate remains coolable after a LOCA event under design basis conditions. The ballooned section creates some obstacles for the cooling water during reflood of the bundle, but this effect causes only a short delay in the cooling down of the hot fuel rods. The accumulation of fuel pellet debris in the ballooned volume results in a local power peak, which leads to further slowing down of quench front.

  9. Natural convection heat transfer within horizontal spent nuclear fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Canaan, R.E.

    1995-12-01

    Natural convection heat transfer is experimentally investigated in an enclosed horizontal rod bundle, which characterizes a spent nuclear fuel assembly during dry storage and/or transport conditions. The basic test section consists of a square array of sixty-four stainless steel tubular heaters enclosed within a water-cooled rectangular copper heat exchanger. The heaters are supplied with a uniform power generation per unit length while the surrounding enclosure is maintained at a uniform temperature. The test section resides within a vacuum/pressure chamber in order to subject the assembly to a range of pressure statepoints and various backfill gases. The objective of this experimental study is to obtain convection correlations which can be used in order to easily incorporate convective effects into analytical models of horizontal spent fuel systems, and also to investigate the physical nature of natural convection in enclosed horizontal rod bundles in general. The resulting data consist of: (1) measured temperatures within the assembly as a function of power, pressure, and backfill gas; (2) the relative radiative contribution for the range of observed temperatures; (3) correlations of convective Nusselt number and Rayleigh number for the rod bundle as a whole; and (4) correlations of convective Nusselt number as a function of Rayleigh number for individual rods within the array.

  10. Feedback Induction of a Photoreceptor-specific Isoform of Retinoid-related Orphan Nuclear Receptor β by the Rod Transcription Factor NRL*

    Science.gov (United States)

    Fu, Yulong; Liu, Hong; Ng, Lily; Kim, Jung-Woong; Hao, Hong; Swaroop, Anand; Forrest, Douglas

    2014-01-01

    Vision requires the generation of cone and rod photoreceptors that function in daylight and dim light, respectively. The neural retina leucine zipper factor (NRL) transcription factor critically controls photoreceptor fates as it stimulates rod differentiation and suppresses cone differentiation. However, the controls over NRL induction that balance rod and cone fates remain unclear. We have reported previously that the retinoid-related orphan receptor β gene (Rorb) is required for Nrl expression and other retinal functions. We show that Rorb differentially expresses two isoforms: RORβ2 in photoreceptors and RORβ1 in photoreceptors, progenitor cells, and other cell types. Deletion of RORβ2 or RORβ1 increased the cone:rod ratio ∼2-fold, whereas deletion of both isoforms in Rorb−/− mice produced almost exclusively cone-like cells at the expense of rods, suggesting that both isoforms induce Nrl. Electroporation of either RORβ isoform into retinal explants from Rorb−/− neonates reactivated Nrl and rod genes but, in Nrl−/− explants, failed to reactivate rod genes, indicating that NRL is the effector for both RORβ isoforms in rod differentiation. Unexpectedly, RORβ2 expression was lost in Nrl−/− mice. Moreover, NRL activated the RORβ2-specific promoter of Rorb, indicating that NRL activates Rorb, its own inducer gene. We suggest that feedback activation between Nrl and Rorb genes reinforces the commitment to rod differentiation. PMID:25296752

  11. Feedback induction of a photoreceptor-specific isoform of retinoid-related orphan nuclear receptor β by the rod transcription factor NRL.

    Science.gov (United States)

    Fu, Yulong; Liu, Hong; Ng, Lily; Kim, Jung-Woong; Hao, Hong; Swaroop, Anand; Forrest, Douglas

    2014-11-21

    Vision requires the generation of cone and rod photoreceptors that function in daylight and dim light, respectively. The neural retina leucine zipper factor (NRL) transcription factor critically controls photoreceptor fates as it stimulates rod differentiation and suppresses cone differentiation. However, the controls over NRL induction that balance rod and cone fates remain unclear. We have reported previously that the retinoid-related orphan receptor β gene (Rorb) is required for Nrl expression and other retinal functions. We show that Rorb differentially expresses two isoforms: RORβ2 in photoreceptors and RORβ1 in photoreceptors, progenitor cells, and other cell types. Deletion of RORβ2 or RORβ1 increased the cone:rod ratio ∼2-fold, whereas deletion of both isoforms in Rorb(-/-) mice produced almost exclusively cone-like cells at the expense of rods, suggesting that both isoforms induce Nrl. Electroporation of either RORβ isoform into retinal explants from Rorb(-/-) neonates reactivated Nrl and rod genes but, in Nrl(-/-) explants, failed to reactivate rod genes, indicating that NRL is the effector for both RORβ isoforms in rod differentiation. Unexpectedly, RORβ2 expression was lost in Nrl(-/-) mice. Moreover, NRL activated the RORβ2-specific promoter of Rorb, indicating that NRL activates Rorb, its own inducer gene. We suggest that feedback activation between Nrl and Rorb genes reinforces the commitment to rod differentiation. © 2014 by The American Society for Biochemistry and Molecular Biology, Inc.

  12. Development of improved thermal hydraulics and fuel performance technology; development of turbulence model for flow analysis in nuclear fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Joo, W. K.; Kong, D. W.; Park, H. Z. [Yonsei University, Seoul (Korea)

    2001-04-01

    The flow through a nuclear rod bundle with mixing vanes are very complex and required a suitable turbulence model to be predicted accurately. Subchannel flow in a nuclear bundle having vanes to mix flow appears complex turbulent flow. Objective of this study is to develop turbulence model which can predict complex flow. Also, the module will be produced, which can implement the developed turbulence model in the CFX code. The selected turbulence models are k-epsilon model, non-linear k-epsilon model, Reynolds stress model and modified Reynolds stress model to test their performance in the prediction of the flow in nuclear assembly. These models are tested for a 2-D backwise step flow, square duct flow, rod bundle flow and subchannel flow using CFX. The modules, which can implement Reynolds stress model and non-linear k-epsilon odel in CFX code, are produced. The advantages and disadvantages for these turbulence models are described and the limitation of implementation of non-linear model in CFX code is discussed. The results obtained from the research would give a help for the development of turbulence model which can accurately predict the flow through the rod bundles with mixing vanes. 18 refs., 37 figs., 8 tabs. (Author)

  13. Automatic safety rod for reactors. [LMFBR

    Science.gov (United States)

    Germer, J.H.

    1982-03-23

    An automatic safety rod for a nuclear reactor containing neutron absorbing material and designed to be inserted into a reactor core after a loss-of-flow. Actuation is based upon either a sudden decrease in core pressure drop or the pressure drop decreases below a predetermined minimum value. The automatic control rod includes a pressure regulating device whereby a controlled decrease in operating pressure due to reduced coolant flow does not cause the rod to drop into the core.

  14. Parabolic k-ample bundles

    CERN Document Server

    Biswas, Indranil

    2011-01-01

    We construct projectivization of a parabolic vector bundle and a tautological line bundle over it. It is shown that a parabolic vector bundle is ample if and only if the tautological line bundle is ample. This allows us to generalize the notion of a k-ample bundle, introduced by Sommese, to the context of parabolic bundles. A parabolic vector bundle $E_*$ is defined to be k-ample if the tautological line bundle ${\\mathcal O}_{{\\mathbb P}(E_*)}(1)$ is $k$--ample. We establish some properties of parabolic k-ample bundles.

  15. Multicell slug flow heat transfer analysis of finite LMFBR bundles

    Energy Technology Data Exchange (ETDEWEB)

    Yeung, M.K.; Wolf, L.

    1978-12-01

    An analytical two-dimensional, multi-region, multi-cell technique has been developed for the thermal analysis of LMFBR rod bundles. Local temperature fields of various unit cells were obtained for 7, 19, and 37-rod bundles of different geometries and power distributions. The validity of the technique has been verified by its excellent agreement with the THTB calculational result. By comparing the calculated fully-developed circumferential clad temperature distribution with those of the experimental measurements, an axial correction factor has been derived to account for the entrance effect for practical considerations. Moreover, the knowledge of the local temperature field of the rod bundle leads to the determination of the effective mixing lengths L/sub ij/ for adjacent subchannels of various geometries. It was shown that the implementation of the accurately determined L/sub ij/ into COBRA-IIIC calculations has fairly significant effects on intersubchannel mixing. In addition, a scheme has been proposed to couple the 2-D distributed and lumped parameter calculation by COBRA-IIIC such that the entrance effect can be implanted into the distributed parameter analysis. The technique has demonstrated its applicability for a 7-rod bundle and the results of calculation were compared to those of three-dimensional analyses and experimental measurements.

  16. Rod consolidation at the West Valley Demonstration Project

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, W.J.

    1986-12-01

    A rod consolidation demonstration with irradiated pressurized water reactor fuel was recently conducted by personnel from Nuclear Assurance Corporation and West Valley Nuclear Services Company at the West Valley Demonstration Project in West Valley, New York. The rod consolidation demonstration involved pulling all of the fuel rods from six fuel Assemblies. In general, the rod pulling proceeded smoothly. The highest compaction ratio attained was 1:8:1. Among the total of 1074 fuel rods were some known degraded rods (they had collapsed cladding, a result of in-reactor fuel densification), but no rods were broken or dropped during the demonstration. One aim was to gather information on the effect of rod consolidation operations on the integrity of the fuel rods during subsequent handling and storage. Another goal was to collect information on the condition and handling of intact, damaged, and failed fuel that has been in storage for an extended period. 9 refs., 8 figs., 1 tab.

  17. Modernization of the Electric Power Systems (transformers, rods and switches) in the Laguna Verde Nuclear Power Plant (Mexico); Modernizacion de los Sistemas Electricos de Potencia (Transformadores de Principales, Interruptor de Generacion, Barras de Fase Aislada) de la Central Nuclear de Laguna Verde (Mexico)

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez Solarzano, J. J.; Gabaldon Martin, M. A.; Pallisa Nunez, J.; Florez Ordeonez, A.; Fernandez Corbeira, A.; Prieto Diez, I.

    2010-07-01

    Description of the changes made in the Electric Power Systems as a part of the power increase project in the Laguna Verde Nuclear Power Plant (Mexico). The main electrical changes to make, besides the turbo group, are the main generation transformers, the isolated rods and the generation switch.

  18. Nuclear power for commercial high-speed sea-lift

    Energy Technology Data Exchange (ETDEWEB)

    Donaldson, A.J.; Holiday, S. [Rolls-Royce, Derby (United Kingdom)

    2007-07-01

    A 1000 MW (thermal) pressurized water reactor concept design for large, fast merchant ships is presented here. The design study was undertaken to provide an input to an overall assessment of the near future economic viability of nuclear merchant shipping. The intention was to use a widely available nuclear industrial base completed with features considered necessary like: the diversity of shutdown (reactivity compensation rods, safety rods and soluble boron), a large pressurizer (to tolerate loss of steam demand without relief valve lift and to support a long response time for small leak events), the loop isolation (to manage risks of steam generator tube bundle leaks) and the passive long-term cooling to the atmosphere via duplicate systems. It appears that nuclear propulsion will have an advantage if oil prices rise by 25% or more.

  19. The ABCDEF Implementation Bundle

    Directory of Open Access Journals (Sweden)

    Annachiara Marra

    2016-08-01

    Full Text Available Long-term morbidity, long-term cognitive impairment and hospitalization-associated disability are common occurrence in the survivors of critical illness, with significant consequences for patients and for the caregivers. The ABCDEF bundle represents an evidence-based guide for clinicians to approach the organizational changes needed for optimizing ICU patient recovery and outcomes. The ABCDEF bundle includes: Assess, Prevent, and Manage Pain, Both Spontaneous Awakening Trials (SAT and Spontaneous Breathing Trials (SBT, Choice of analgesia and sedation, Delirium: Assess, Prevent, and Manage, Early mobility and Exercise, and Family engagement. The purpose of this review is to describe the core features of the ABCDEF bundle.

  20. Reactor control rod timing system. [LMFBR

    Science.gov (United States)

    Wu, P.T.K.

    1980-03-18

    A fluid driven jet-edge whistle timing system is described for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

  1. Design requirement on KALIMER control rod assembly duct

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, W.; Kang, H. Y.; Nam, C.; Kim, J. O.; Kim, Y. J

    1998-03-01

    This document establishes the design guidelines which are needs for designing the control rod assembly duct of the KALIMER as design requirements. it describes control rod assembly duct of the KALIMER and its requirements that includes functional requirements, performance requirements, interfacing systems, design limits and strength requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. The control rod system consists of three parts, which are drive mechanism, drive-line, and absorber bundle. This report deals with the absorber bundle and its outer duct only because the others are beyond the scope of fuel system design. The guidelines for design requirements intend to be used for an improved design of the control rod assembly duct of the KALIMER. (author). 19 refs.

  2. Post-test calculation of the QUENCH-17 bundle experiment with debris formation and bottom water reflood using thermal hydraulic and severe fuel damage code SOCRAT/V3

    Energy Technology Data Exchange (ETDEWEB)

    Vasiliev, A., E-mail: vasil@ibrae.ac.ru [Nuclear Safety Institute (IBRAE), B. Tulskaya 52, 115191 Moscow (Russian Federation); Stuckert, J., E-mail: juri.stuckert@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2015-03-15

    Highlights: • Modeling of processes in porous debris regions. • Analysis of coolability of massive debris bed. • Complexity of simulation of flow regime near boiling curve. - Abstract: The thermal hydraulic and SFD (Severe Fuel Damage) best estimate computer modeling code SOCRAT/V3 was used for the post-test analysis of the QUENCH-17 experiment performed at KIT on January 2013. The objective of this test was to examine the formation of a debris bed inside the completely oxidized region of the bundle without melt formation and to investigate the coolability behavior during the reflood. The test bundle for QUENCH-17 test was intentionally changed in comparison to basic QUENCH bundles (usually 21 heated rod simulators) with the emphasis to investigate debris behavior phenomena. Only 12 periphery fuel rod simulators were heated by centerline tungsten heaters. 9 unheated fuel rod simulators were located in the inner part of the test bundle. This is why the massive porous debris formation in the inner part of the bundle was not influenced by the presence of tungsten heaters. The QUENCH-17 test conditions simulated a hypothetical scenario of nuclear power plant severe accident sequence with debris bed formation in which the overheated up to 1800 K core would be flooded from the bottom by ECCS (Emergency Core Cooling System). The QUENCH-17 test included the following phases: (1) heat-up phase (heat-up rate up to 0.25 K/s); (2) oxidation phase (the cladding temperature about 1800 K in hottest region, steam mass flow rate 2 g/s); (3) bottom flood phase (characteristic cooling time about 600 s, water mass flow rate 10 g/s). SOCRAT/V3 computer modeling code was used for calculation of basic thermal hydraulic, oxidation and thermal mechanical behavior during all phases of the experiment. The calculated results are in a good agreement with experimental data which justifies the adequacy of modeling capabilities of SOCRAT code system.

  3. Composite nuclear fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Dollard, W.J.; Ferrari, H.M.

    1982-04-27

    An open lattice elongated nuclear fuel assembly including small diameter fuel rods disposed in an array spaced a selected distance above an array of larger diameter fuel rods for use in a nuclear reactor having liquid coolant flowing in an upward direction. Plenums are preferably provided in the upper portion of the upper smaller diameter fuel rods and in the lower portion of the lower larger diameter fuel rods. Lattice grid structures provide lateral support for the fuel rods and preferably the lowest grid about the upper rods is directly and rigidly affixed to the highest grid about the lower rods.

  4. Advanced gray rod control assembly

    Energy Technology Data Exchange (ETDEWEB)

    Drudy, Keith J; Carlson, William R; Conner, Michael E; Goldenfield, Mark; Hone, Michael J; Long, Jr., Carroll J; Parkinson, Jerod; Pomirleanu, Radu O

    2013-09-17

    An advanced gray rod control assembly (GRCA) for a nuclear reactor. The GRCA provides controlled insertion of gray rod assemblies into the reactor, thereby controlling the rate of power produced by the reactor and providing reactivity control at full power. Each gray rod assembly includes an elongated tubular member, a primary neutron-absorber disposed within the tubular member said neutron-absorber comprising an absorber material, preferably tungsten, having a 2200 m/s neutron absorption microscopic capture cross-section of from 10 to 30 barns. An internal support tube can be positioned between the primary absorber and the tubular member as a secondary absorber to enhance neutron absorption, absorber depletion, assembly weight, and assembly heat transfer characteristics.

  5. Subtleties Concerning Conformal Tractor Bundles

    CERN Document Server

    Graham, C Robin

    2012-01-01

    The realization of tractor bundles as associated bundles in conformal geometry is studied. It is shown that different natural choices of principal bundle with normal Cartan connection corresponding to a given conformal manifold can give rise to topologically distinct associated tractor bundles for the same inducing representation. Consequences for homogeneous models and conformal holonomy are described. A careful presentation is made of background material concerning standard tractor bundles and equivalence between parabolic geometries and underlying structures.

  6. An analytical model for the prediction of fluid-elastic forces in a rod bundle subjected to axial flow: theory, experimental validation and application to PWR fuel assemblies; Calcul des forces fluidelastiques dans les faisceaux de tubes sous ecoulement axial: theorie, validation, application aux assemblages combustibles des REP

    Energy Technology Data Exchange (ETDEWEB)

    Beaud, F. [Electricite de France (EDF), 78 - Chatou (France)

    1997-12-31

    A model predicting the fluid-elastic forces in a bundle of circular cylinders subjected to axial flow is presented in this paper. Whereas previously published models were limited to circular flow channel, the present one allows to take a rectangular flow external boundary into account. For that purpose, an original approach is derived from the standard method of images. This model will eventually be used to predict the fluid-structure coupling between the flow of primary coolant and a fuel assemblies in PWR nuclear reactors. It is indeed of major importance since the flow is shown to induce quite high damping and could therefore mitigate the incidence of an external load like a seismic excitation on the dynamics of the assemblies. The proposed model is validated on two cases from the literature but still needs further comparisons with the experiments being currently carried out on the EDF set-up. The flow has been shown to induce an approximate 12% damping on a PWR fuel assembly, at nominal reactor conditions. The possible grid effect on the fluid-structure coupling has been neglected so far but will soon be investigated at EDF. (author). 16 refs.

  7. Overview of methods to increase dryout power in CANDU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Groeneveld, D.C., E-mail: degroeneveld@gmail.com [Chalk River Laboratories, AECL, Chalk River (Canada); University of Ottawa, Department of Mechanical Engineering, Ottawa (Canada); Leung, L.K.H. [Chalk River Laboratories, AECL, Chalk River (Canada); Park, J.H. [Korean Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-06-15

    Highlights: • Small changes in bundle geometry can have noticeable effects on the bundle CHF. • Rod spacing devices can results in increases of over 200% in CHF. • CHF enhancement decays exponentially downstream from spacers. • CHF-enhancing bundle appendages also increase the post-CHF heat transfer. - Abstract: In CANDU reactors some degradation in the CCP (critical channel power, or power corresponding to the first occurrence of CHF in any fuel channel) will occur with time because of ageing effects such as pressure-tube diametral creep, increase in reactor inlet-header temperature, increased hydraulic resistance of feeders. To compensate for the ageing effects, various options for recovering the loss in CCP are described in this paper. They include: (i) increasing the bundle heated perimeter, (ii) optimizing the bundle configuration, (iii) optimizing core flow and flux distribution, (iv) reducing the bundle hydraulic resistance, (v) use of CHF-enhancing bundle appendages, (vi) more precise experimentation, and (vii) redefining CHF. The increase in CHF power has been quantified based on experiments on full-scale bundles and subchannel code predictions. The application of several of these CHF enhancement principles has been used in the development of the 43-rod CANFLEX bundle.

  8. Acoustic sensors for fission gas characterization: R and D skills devoted to innovative instrumentation in MTR, non-destructive devices in hot lab facilities and specific transducers for measurements of LWR rods in nuclear plants

    Energy Technology Data Exchange (ETDEWEB)

    Ferrandis, J.Y.; Leveque, G.; Rosenkrantz, E.; Augereau, F.; Combette, P. [University Montpellier, IES, UMR 5214, F-34000, Montpellier (France); CNRS, IES, UMR 5214, F-34000, Montpellier (France)

    2015-07-01

    First of all, we will present the main principle of the method. A piezoelectric transducer, driven by a pulse generator, generates the acoustic waves in a cavity that may be the fuel rod or a chamber connected to an instrumented rod. The composition determination consists in measuring the time of flight of the acoustic signal emitted. The pressure can be estimated by a calibration process, above the measurement of the amplitude of the signal. Two projects will then be detailed. The first project consists in the development of advanced instrumentation for in-pile experiments in Material Testing Reactor. It constitutes a main goal for the improvement of the nuclear fuel behavior knowledge. This acoustic method was tested with success during a first experiment called REMORA 3, and the results were used to differentiate helium and fission gas release kinetics under transient operating conditions. This experiment was lead at OSIRIS reactor (CEA Saclay, France). As a first step of the development program, we performed in-pile tests on the most sensitive component, i.e., the piezoelectric transducer. For this purpose, the active part of this sensor has been qualified on gamma and neutron radiations and at high temperature. Various industrial piezo-ceramics were exposed to a high activity Cobalt source for few days. The cumulated dose was ranged from 50 kGy up to 2 MGy. Next, these devices were placed inside a Material Test Reactor to investigate their reliability towards neutron fluence. The final fluence after 150 days of irradiation was up to 1.6.10{sup 21}n/cm{sup 2} (for thermal neutron). Irreversible variations have been measured. Next, a specific sensor has been implemented on an instrumented fuel rod and tested in the frame of a REMORA 3 Irradiation test. It was the first experiment under high mixed, temperature neutron and gamma flux. A first irradiation phase took place in March 2010 in the OSIRIS reactor and in November 2010 for the second step of the

  9. Control rod cluster drop time anomaly Guandong nuclear power station (Daya bay) and Electricite de France nuclear power stations (1450 MWe N4 Series); Anomalie de temps de chute des grappes de controle centrale de guang dong (daya bay) et centrales d`electricite de France (Palier N4-1450 MWE)

    Energy Technology Data Exchange (ETDEWEB)

    Olivera, J.J.; Naury, S.; Tricot, N.; Tran Dai, P.; Gama, J.M.

    1996-12-31

    The anomaly of control rod cluster drop time revealed at Guandong Nuclear Power Station in Daya Bay and in the Chooz B1 pilot unit for the N4 series, led to the replacement of the M1 type control rod cluster guide tubes with 1300 MWe PWR type guide tubes, adapted to the geometry of the Guandong reactors and the 1450 MWe reactors of the N4 series. The comparison of the drop times obtained with the 1300 MWe type control rod cluster guide 1300 MWe type control rod cluster guide tubes gave satisfactory results. These met the safety criterion for N4 series control rod cluster drop times (2.15 under hot shutdown conditions). The drop time tests which will be carried out in middle of and at the end of cycle 1 of Chooz B1 should make it possible to finally validate the solution already successfully implemented at Guandong. However, this anomaly has revealed the limits of representativeness of the experimental test loops with regard to the real reactor configuration. In view of this, it has been deemed necessary to ask Electricite de France to pursue its analysis both on the understanding of the phenomena which led to this anomaly and on the limits of the representativeness of the experimental test loops. (authors).

  10. Experiments on silver-indium-cadmium control rod failure during severe accident sequences

    Energy Technology Data Exchange (ETDEWEB)

    Steinbrueck, M.; Stegmaier, U. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany)

    2010-05-15

    Silver-indium-cadmium (SIC) alloy is used as neutron absorber material in control rods (CR) of Pressurized Water Reactors (PWR). It is the material with the lowest melting temperature (approx. 1100 K) among all metallic and ceramic materials applied in nuclear reactors. During a hypothetical severe accident the SIC melt is kept in its stainless steel (SS) cladding tube as long as this is intact. After failure of the cladding tube by eutectic interaction with the Zircaloy-4 (Zry-4) guide tube or latest by reaching the SS melting temperature SIC elements are released and may interact with other core components. Furthermore, Ag-In-Cd are one of the main contributors to aerosol release in the reactor cooling system and may strongly influence nature and transport of fission products in the primary circuit and later on in the containment. The bundle experiment QUENCH-13 with prototypical SIC control rod as well as two series of single-rod tests with 10-cm long CR segments were performed at Karlsruhe Institute of Technology (KIT, former FZK) in order to improve the data base on SIC CR degradation and aerosol release. This paper concentrates on the degradation and failure mechanisms of SIC CRs as well as on the interaction between SIC absorber melt with other core components. (orig.)

  11. Right bundle branch block

    DEFF Research Database (Denmark)

    Bussink, Barbara E; Holst, Anders Gaarsdal; Jespersen, Lasse

    2013-01-01

    AimsTo determine the prevalence, predictors of newly acquired, and the prognostic value of right bundle branch block (RBBB) and incomplete RBBB (IRBBB) on a resting 12-lead electrocardiogram in men and women from the general population.Methods and resultsWe followed 18 441 participants included.......5%/2.3% in women, P Right bundle branch block was associated with significantly...... increased all-cause and cardiovascular mortality in both genders with age-adjusted hazard ratios (HR) of 1.31 [95% confidence interval (CI), 1.11-1.54] and 1.87 (95% CI, 1.48-2.36) in the gender pooled analysis with little attenuation after multiple adjustment. Right bundle branch block was associated...

  12. Bundles of Banach algebras

    Directory of Open Access Journals (Sweden)

    J. W. Kitchen

    1994-01-01

    Full Text Available We study bundles of Banach algebras π:A→X, where each fiber Ax=π−1({x} is a Banach algebra and X is a compact Hausdorff space. In the case where all fibers are commutative, we investigate how the Gelfand representation of the section space algebra Γ(π relates to the Gelfand representation of the fibers. In the general case, we investigate how adjoining an identity to the bundle π:A→X relates to the standard adjunction of identities to the fibers.

  13. Nrl is required for rod photoreceptor development.

    Science.gov (United States)

    Mears, A J; Kondo, M; Swain, P K; Takada, Y; Bush, R A; Saunders, T L; Sieving, P A; Swaroop, A

    2001-12-01

    The protein neural retina leucine zipper (Nrl) is a basic motif-leucine zipper transcription factor that is preferentially expressed in rod photoreceptors. It acts synergistically with Crx to regulate rhodopsin transcription. Missense mutations in human NRL have been associated with autosomal dominant retinitis pigmentosa. Here we report that deletion of Nrl in mice results in the complete loss of rod function and super-normal cone function, mediated by S cones. The photoreceptors in the Nrl-/- retina have cone-like nuclear morphology and short, sparse outer segments with abnormal disks. Analysis of retinal gene expression confirms the apparent functional transformation of rods into S cones in the Nrl-/- retina. On the basis of these findings, we postulate that Nrl acts as a 'molecular switch' during rod-cell development by directly modulating rod-specific genes while simultaneously inhibiting the S-cone pathway through the activation of Nr2e3.

  14. Principal -bundles on Nodal Curves

    Indian Academy of Sciences (India)

    Usha N Bhosle

    2001-08-01

    Let be a connected semisimple affine algebraic group defined over . We study the relation between stable, semistable -bundles on a nodal curve and representations of the fundamental group of . This study is done by extending the notion of (generalized) parabolic vector bundles to principal -bundles on the desingularization of and using the correspondence between them and principal -bundles on . We give an isomorphism of the stack of generalized parabolic bundles on with a quotient stack associated to loop groups. We show that if is simple and simply connected then the Picard group of the stack of principal -bundles on is isomorphic to ⊕ , being the number of components of .

  15. Molybdenum-99-producing 37-element fuel bundle neutronically and thermal-hydraulically equivalent to a standard CANDU fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Nichita, E., E-mail: Eleodor.Nichita@uoit.ca; Haroon, J., E-mail: Jawad.Haroon@uoit.ca

    2016-10-15

    Highlights: • A 37-element fuel bundle modified for {sup 99}Mo production in CANDU reactors is presented. • The modified bundle is neutronically and thermal-hydraulically equivalent to the standard bundle. • The modified bundle satisfies all safety criteria satisfied by the standard bundle. - Abstract: {sup 99m}Tc, the most commonly used radioisotope in diagnostic nuclear medicine, results from the radioactive decay of {sup 99}Mo which is currently being produced at various research reactors around the globe. In this study, the potential use of CANDU power reactors for the production of {sup 99}Mo is investigated. A modified 37-element fuel bundle, suitable for the production of {sup 99}Mo in existing CANDU-type reactors is proposed. The new bundle is specifically designed to be neutronically and thermal-hydraulically equivalent to the standard 37-element CANDU fuel bundle in normal, steady-state operation and, at the same time, be able to produce significant quantities of {sup 99}Mo when irradiated in a CANDU reactor. The proposed bundle design uses fuel pins consisting of a depleted-uranium centre surrounded by a thin layer of low-enriched uranium. The new molybdenum-producing bundle is analyzed using the lattice transport code DRAGON and the diffusion code DONJON. The proposed design is shown to produce 4081 six-day Curies of {sup 99}Mo activity per bundle when irradiated in the peak-power channel of a CANDU core, while maintaining the necessary reactivity and power rating limits. The calculated {sup 99}Mo yield corresponds to approximately one third of the world weekly demand. A production rate of ∼3 bundles per week can meet the global demand of {sup 99}Mo.

  16. On projective space bundle with nef normalized tautological line bundle

    CERN Document Server

    Yasutake, Kazunori

    2011-01-01

    In this paper, we study the structure of projective space bundles whose relative anti-canonical line bundle is nef. As an application, we get a characterization of abelian varieties up to finite etale covering.

  17. Radiological and nuclear safety aspects in the fabrication of 1.8% enriched U O{sub 2} fuel rods for the RA-8 critical facility; Aspectos de seguridad radiologica y nuclear en la fabricacion de barras combustibles, con U O{sub 2} enriquecido al 1.8%, para la facilidad critica RA-8

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas, Hugo; Becarra, Fabian; Herrero, Jorge; Luna, Manuel; Perez, Aldo [Comision Nacional de Energia Atomica, (Argentina). Centro Atomico Constituyentes

    1997-10-01

    The neutronic behavioral study of the fuel for the future nuclear reactor CAREM required to mount critical facility with 1.8% enriched U O{sub 2} fuel rods. The present work describes the various operation and production processes, the safety and radioprotection systems, the administrative procedures and the associated radiological controls. Also, the results obtained in the area and personal monitoring and waste generation are detailed. (author). 10 refs., 4 figs., 1 tab.

  18. Kernel bundle EPDiff

    DEFF Research Database (Denmark)

    Sommer, Stefan Horst; Lauze, Francois Bernard; Nielsen, Mads

    2011-01-01

    In the LDDMM framework, optimal warps for image registration are found as end-points of critical paths for an energy functional, and the EPDiff equations describe the evolution along such paths. The Large Deformation Diffeomorphic Kernel Bundle Mapping (LDDKBM) extension of LDDMM allows scale space...

  19. Universal Lagrangian bundles

    NARCIS (Netherlands)

    Sepe, D.

    2013-01-01

    The obstruction to construct a Lagrangian bundle over a fixed integral affine manifold was constructed by Dazord and Delzant (J Differ Geom 26:223–251, 1987) and shown to be given by ‘twisted’ cup products in Sepe (Differ GeomAppl 29(6): 787–800, 2011). This paper uses the topology of universal Lagr

  20. ALUMINUM BOX BUNDLING PRESS

    Directory of Open Access Journals (Sweden)

    Iosif DUMITRESCU

    2015-05-01

    Full Text Available In municipal solid waste, aluminum is the main nonferrous metal, approximately 80- 85% of the total nonferrous metals. The income per ton gained from aluminum recuperation is 20 times higher than from glass, steel boxes or paper recuperation. The object of this paper is the design of a 300 kN press for aluminum box bundling.

  1. CANFLEX fuel bundle cross-flow endurance test (test report)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sung Deok; Chung, C. H.; Chang, S. K.; Kim, B. D.

    1997-04-01

    As part of the normal refuelling sequence of CANDU nuclear reactor, both new and irradiated bundles can be parked in the cross-flow region of the liner tubes. This situation occurs normally for a few minutes. The fuel bundle which is subjected to the cross-flow should be capable of withstanding the consequences of cross flow for normal periods, and maintain its mechanical integrity. The cross-flow endurance test was conducted for CANFLEX bundle, latest developed nuclear fuel, at CANDU-Hot Test Loop. The test was carried out during 4 hours at the inlet cross-flow region. After the test, the bundle successfully met all acceptance criteria after the 4 hours cross-flow test. (author). 2 refs., 3 tabs.

  2. Tie rod insertion test

    CERN Multimedia

    B. LEVESY

    2002-01-01

    The superconducting coil is inserted in the outer vaccum tank and supported by a set of tie rods. These tie rods are made of titanium alloy. This test reproduce the final insertion of the tie rods inside the outer vacuum tank.

  3. Simulation of bundle test Quench-12 with integral code MELCOR

    Energy Technology Data Exchange (ETDEWEB)

    Duspiva, J. [Nuclear Research Inst., Rez plc (Czech Republic)

    2011-07-01

    The past NRI analyses cover the Quench-01, Quench-03 and Quench-06 with version MELCOR 1.8.5 (including reflood model), and Quench-01 and Quench-11 tests with the latest version MELCOR 1.8.6. The Quench-12 test is specific, because it has different bundle configuration related to the VVER bundle configuration with hexagonal grid of pins and also used E110 cladding material. Specificity of Quench-12 test is also in the used material of fuel rod cladding - E110. The test specificities are a reason for the highest concern, because the VVER reactors are operated in the Czech Republic. The new input model was developed with the taking into account all experience from previous simulations of the Quench bundle tests. The recent version MELCOR 1.8.6 YU{sub 2}911 was used for the simulation with slightly modified ELHEAT package. Sensitivity studies on input parameters and oxidation kinetics were performed. (author)

  4. Proposal of a synchro panel meter instrument to replace the obsolete Synchro/Resolver reading device used as position indicator of safety rods assembly of the Brazilian IEA-R1 Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Toledo, Fabio de; Brancaccio, Franco; Cardenas, Jose Patricio N., E-mail: fatoledo@ipen.br, E-mail: fbrancac@ipen.br, E-mail: ahiru@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    IPEN (Instituto de Pesquisas Energeticas e Nucleares) was founded in 1956 (as Atomic Energy Institute - IEA) as a facility complex, for the research, development and application, in the nuclear technology field. The institute is recognized as a national leader in nuclear research and development (R and D), including the areas of reactor operation, radiopharmaceuticals, industrial and laboratory applications, materials science and laser technologies and applications. IPEN's main facility is the IEA-R1, nuclear research reactor (NRR), today, the only one in Brazil with a power level suitable for applications in physics, chemistry, biology and engineering. Some radioisotopes are also produced in IEA-R1, for medical and other applications. A common problem faced in the IEA-R1 maintenance is instrumentation obsolescence; spare parts are no more available, because of discontinued production, and an updating program is mandatory, aiming at modernization of old-aged I and C systems. In the presented context, an electronic system is here proposed, as a replacement for the reactor safety (shim) rods assembly position indicator, based on an open-source physical computing platform called Arduino, which includes a simple microcontroller board and a software-code development environment. A mathematical algorithm for the synchro-motor signal processing was developed, and the obtained resolution was better than 1.5%. (author)

  5. Parallel CFD simulations of turbulent flows inside a CANDU fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Abbasian, F.; Yu, S.D.; Cao, J. [Ryerson Univ., Dept. of Mechanical and Industrial Engineering, Toronto, Ontario (Canada)], E-mail: fabbasia@ryerson.ca

    2008-07-01

    Large Eddy Simulation (LES) is used to study the turbulent flow inside a 43-rod bundle. The two LES models developed in this paper are of dynamic Smagorinsky type, featuring a satisfactory prediction of anisotropic turbulence intensity and frequency. The first model, by taking advantage of the geometric periodicity, deals with one seventh of a rod bundle; it is developed for studying the axial, lateral turbulence intensities and frequencies in the centers of subchannels and narrow-gap regions. The second model, dealing with the full rod bundle inside a pressure tube with nominal eccentricity, is developed for studying the turbulent fluid forces acting on the bundle. In order to accelerate the solution process for the two large CFD models, the parallelized CFD technique is utilized in connection with 24 processors. The numerical results, obtained for a test case (an eight-rod bundle), are in good agreement with those experimental data available in the literature. Numerical simulations of turbulent flow phenomena within subchannels are advantageous since true flow features are difficult or costly to reveal by experiments. (author)

  6. Deformation quantization of principal bundles

    CERN Document Server

    Aschieri, Paolo

    2016-01-01

    We outline how Drinfeld twist deformation techniques can be applied to the deformation quantization of principal bundles into noncommutative principal bundles, and more in general to the deformation of Hopf-Galois extensions. First we twist deform the structure group in a quantum group, and this leads to a deformation of the fibers of the principal bundle. Next we twist deform a subgroup of the group of authomorphisms of the principal bundle, and this leads to a noncommutative base space. Considering both deformations we obtain noncommutative principal bundles with noncommutative fiber and base space as well.

  7. Helices and vector bundles

    CERN Document Server

    Rudakov, A N

    1990-01-01

    This volume is devoted to the use of helices as a method for studying exceptional vector bundles, an important and natural concept in algebraic geometry. The work arises out of a series of seminars organised in Moscow by A. N. Rudakov. The first article sets up the general machinery, and later ones explore its use in various contexts. As to be expected, the approach is concrete; the theory is considered for quadrics, ruled surfaces, K3 surfaces and P3(C).

  8. Bundling harvester; Nippukorjausharvesteri

    Energy Technology Data Exchange (ETDEWEB)

    Koponen, K. [Eko-Log Oy, Kuopio (Finland)

    1996-12-31

    The staring point of the project was to design and construct, by taking the silvicultural point of view into account, a harvesting and processing system especially for energy-wood, containing manually driven bundling harvester, automatizing of the harvester, and automatized loading. The equipment forms an ideal method for entrepreneur`s-line harvesting. The target is to apply the system also for owner`s-line harvesting. The profitability of the system promotes the utilization of the system in both cases. The objectives of the project were: to construct a test equipment and prototypes for all the project stages, to carry out terrain and strain tests in order to examine the usability and durability, as well as the capacity of the machine, to test the applicability of the Eko-Log system in simultaneous harvesting of energy and pulp woods, and to start the marketing and manufacturing of the products. The basic problems of the construction of the bundling harvester have been solved using terrain-tests. The prototype machine has been shown to be operable. Loading of the bundles to form sufficiently economically transportable loads has been studied, and simultaneously, the branch-biomass has been tried to be utilized without loosing the profitability of transportation. The results have been promising, and will promote the profitable utilization of wood-energy

  9. Motor-free actin bundle contractility driven by molecular crowding

    CERN Document Server

    Schnauß, Jörg; Schuldt, Carsten; Schmidt, B U Sebastian; Glaser, Martin; Strehle, Dan; Heussinger, Claus; Käs, Josef A

    2015-01-01

    Modeling approaches of suspended, rod-like particles and recent experimental data have shown that depletion forces display different signatures depending on the orientation of these particles. It has been shown that axial attraction of two rods yields contractile forces of 0.1pN that are independent of the relative axial shift of the two rods. Here, we measured depletion-caused interactions of actin bundles extending the phase space of single pairs of rods to a multi-particle system. In contrast to a filament pair, we found forces up to 3pN . Upon bundle relaxation forces decayed exponentially with a mean decay time of 3.4s . These different dynamics are explained within the frame of a mathematical model by taking pairwise interactions to a multi-filament scale. The macromolecular content employed for our experiments is well below the crowding of cells. Thus, we propose that arising forces can contribute to biological force generation without the need to convert chemical energy into mechanical work.

  10. Using Finite Model Analysis and Out of Hot Cell Surrogate Rod Testing to Analyze High Burnup Used Nuclear Fuel Mechanical Properties

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL; Jiang, Hao [ORNL; Wang, Hong [ORNL

    2014-07-01

    Based on a series of FEA simulations, the discussions and the conclusions concerning the impact of the interface bonding efficiency to SNF vibration integrity are provided in this report; this includes the moment carrying capacity distribution between pellets and clad, and the impact of cohesion bonding on the flexural rigidity of the surrogate rod system. As progressive de-bonding occurs at the pellet-pellet interfaces and at the pellet-clad interface, the load ratio of the bending moment carrying capacity gradually shifts from the pellets to the clad; the clad starts to carry a significant portion of the bending moment resistance until reaching the full de-bonding state at the pellet-pellet interface regions. This results in localized plastic deformation of the clad at the pellet-pellet-clad interface region; the associated plastic deformations of SS clad leads to a significant degradation in the stiffness of the surrogate rod. For instance, the flexural rigidity was reduced by 39% from the perfect bond state to the de-bonded state at the pellet-pellet interfaces.

  11. Dependence of control rod worth on fuel burnup

    Energy Technology Data Exchange (ETDEWEB)

    Savva, P., E-mail: savvapan@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Varvayanni, M., E-mail: melina@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Catsaros, N., E-mail: nicos@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece)

    2011-02-15

    Research highlights: Diffusion and MC calculations for rod worth dependence on burnup and Xe in reactors. One-step rod withdrawal/insertion are used for rod worth estimation. The study showed that when Xe is present the rods worth is significantly reduced. Rod worth variation with burnup depends on rod position in core. Rod worth obtained with MC code is higher than that obtained from deterministic. - Abstract: One important parameter in the design and the analysis of a nuclear reactor core is the reactivity worth of the control rods, i.e. their efficiency to absorb excess reactivity. The control rod worth is affected by parameters such as the fuel burnup in the rod vicinity, the Xe concentration in the core, the operational time of the rod and its position in the core. In the present work, two different computational approaches, a deterministic and a stochastic one, were used for the determination of the rods worth dependence on the fuel burnup level and the Xe concentration level in a conceptual, symmetric reactor core, based on the MTR fuel assemblies used in the Greek Research Reactor (GRR-1). For the deterministic approach the neutronics code system composed by the SCALE modules NITAWL and XSDRN and the diffusion code CITATION was used, while for the stochastic one the Monte Carlo code TRIPOLI was applied. The study showed that when Xe is present in the core, the rods worth is significantly reduced, while the rod worth variation with increasing burnup depends on the rods position in the core grid. The rod worth obtained with the use of the Monte Carlo code is higher than the one obtained from the deterministic code.

  12. A genetically encoded reporter for real-time imaging of cofilin-actin rods in living neurons.

    Directory of Open Access Journals (Sweden)

    Jianjie Mi

    Full Text Available Filament bundles (rods of cofilin and actin (1:1 form in neurites of stressed neurons where they inhibit synaptic function. Live-cell imaging of rod formation is hampered by the fact that overexpression of a chimera of wild type cofilin with a fluorescent protein causes formation of spontaneous and persistent rods, which is exacerbated by the photostress of imaging. The study of rod induction in living cells calls for a rod reporter that does not cause spontaneous rods. From a study in which single cofilin surface residues were mutated, we identified a mutant, cofilinR21Q, which when fused with monomeric Red Fluorescent Protein (mRFP and expressed several fold above endogenous cofilin, does not induce spontaneous rods even during the photostress of imaging. CofilinR21Q-mRFP only incorporates into rods when they form from endogenous proteins in stressed cells. In neurons, cofilinR21Q-mRFP reports on rods formed from endogenous cofilin and induced by all modes tested thus far. Rods have a half-life of 30-60 min upon removal of the inducer. Vesicle transport in neurites is arrested upon treatments that form rods and recovers as rods disappear. CofilinR21Q-mRFP is a genetically encoded rod reporter that is useful in live cell imaging studies of induced rod formation, including rod dynamics, and kinetics of rod elimination.

  13. Investigating hydrodynamic characteristics and peculiarities of the coolant flow behind a spacer grid of a fuel rod assembly of the floating nuclear power unit

    Science.gov (United States)

    Dmitriev, S. M.; Doronkov, D. V.; Legchanov, M. A.; Pronin, A. N.; Solncev, D. N.; Sorokin, V. D.; Hrobostov, A. E.

    2016-05-01

    The results of experimental investigations of local hydrodynamics of a coolant flow in fuel rod assembly (FA) of KLT-40C reactor behind a plate spacer grid have been presented. The investigations were carried out on an aerodynamic rig using the gas-phase diffusive tracer test. An analysis of spatial distribution of absolute flow velocity projections and distribution of tracer concentration allowed specifying a coolant flow pattern behind the plate spacer grid of the FA. On the basis of obtained experimental data the recommendations were provided to specify procedures for determining the coolant flow rates for the programs of cell-wise calculation of a core zone of KLT-40C reactor. Investigation results were accepted for the practical use in JSC "OKBM Afrikantov" to assess heat engineering reliability of cores of KLT-40C reactor and were included in a database for verification of CFD programs (CFD-codes).

  14. Telescopic drilling rod

    Energy Technology Data Exchange (ETDEWEB)

    Kagan, I.L.; Berezov, S.I.; Gavrilov, G.A.; Goykhman, Ya.A.; Makushkin, D.O.; Rachev, M.P.; Voynich, L.K.

    1981-09-07

    The telescopic drilling rod includes an inner section of the rod, in whose center cable has been passed and is attached a bearing assembly connecting it to the winch, outer section of rod along which there is pipeline connecting the working cavity formed by the inner section of rod and the housing, installed on the lower end of the outer section of rod, with cavity formed by framework of the guide swivel and end piece and connected to the hydraulic system of the machine by pipeline, as well as clamping elements. In order to drill wells to a depth greater than the length of the outer sectrion of the rod, the latter jointly with the inner section of rod is lowered into the extreme lower position until swivel rests on the feed mechanism. With further slipping of cable and the absence of pressure in the hydraulic system, clamping elements do not have an effect on the inner section of rod. It has the opportunity to freely move along the outer section of rod downwards to the face. When pressure is supplied on pipeline into cavity and further through pipeline into working cavity, the inner section of rod is clamped with feed of the outer section in the process of drilling, both sections move jointly. Because of the link between working cavity of sleeve installed on the lower end of the outer section of rod, and the hydraulic system of the machine through the swivel cavity, it is possible to fix the drilling rod in any mutual axial position of the section.

  15. Quivers For Special Fuel Rods-Disposal Of Special Fuel Rods In CASTOR V Casks

    Energy Technology Data Exchange (ETDEWEB)

    Bannani, Amin; Cebula, Wojciech; Buchmuller, Olga; Huggenberg, Roland [GNS, Essen (Germany); Helmut Kuhl [WTI, Julich (Germany)

    2015-05-15

    While GNS casks of the CASTOR family are a suitable means to transfer fuel assemblies (FA) from the NPP to an interim dry storage site, Germanys phase-out of nuclear energy has triggered the demand for an additional solution to dispose of special fuel rods (SFR), normally remaining in the fuel pond until the final shutdown of the NPP. SFR are fuel rods that had to be removed from fuel assemblies mainly due to their special condition, e. g. damages in the cladding of the fuel rods which may have occurred during reactor operations. SFR are usually stored in the spent fuel pond after they are removed from the FA. The quiver for special fuel rods features a robust yet simple design, with a high mechanical stability, a reliable leak-tightness and large safety margins for future requirements on safety analysis. The quiver for special fuel rods can be easily adapted to a large variety of different damaged fuel rods and tailored to the specific need of the customer. The quiver for special fuel rods is adaptable e.g. in length and diameter for use in other types of transport and storage casks and is applicable in other countries as well. The overall concept presented here is a first of its kind solution for the disposal of SFRs via Castor V-casks. This provides an important precondition in achieving the status 'free from nuclear fuel' of the shut down German NPPs.

  16. Managing bundled payments.

    Science.gov (United States)

    Draper, Andrew

    2011-04-01

    Results of Medicare's ACE demonstration project and Geisinger Health System's ProvenCare initiative provide insight into the challenges hospitals will face as bundled payment proliferates. An early analysis of these results suggests that hospitals would benefit from bringing full automation using clinical IT tools to bear in their efforts to meet these challenges. Other important factors contributing to success include board and physician leadership, organizational structure, pricing methodology for bidding, evidence-based medical practice guidelines, supply cost management, process efficiency management, proactive and aggressive case management, business development and marketing strategy, and the financial management system.

  17. Differential calculi on noncommutative bundles

    OpenAIRE

    Pflaum, Markus J.; Schauenburg, Peter

    1996-01-01

    We introduce a category of noncommutative bundles. To establish geometry in this category we construct suitable noncommutative differential calculi on these bundles and study their basic properties. Furthermore we define the notion of a connection with respect to a differential calculus and consider questions of existence and uniqueness. At the end these constructions are applied to basic examples of noncommutative bundles over a coquasitriangular Hopf algebra.

  18. Entrainment and deposition modeling of liquid films with applications for BWR fuel rod dryout

    Science.gov (United States)

    Ratnayake, Ruwan Kumara

    While best estimate computer codes provide the licensing basis for nuclear power facilities, they also serve as analytical tools in overall plant and component design procedures. An ideal best estimate code would comprise of universally applicable mechanistic models for all its components. However, due to the limited understanding in these specific areas, many of the models and correlations used in these codes reflect high levels of empiricism. As a result, the use of such models is strictly limited to the range of parameters within which the experiments have been conducted. Disagreements between best estimate code predictions and experimental results are often explained by the mechanistic inadequacies of embedded models. Significant mismatches between calculated and experimental critical power values are common observations in the analyses of Boiling Water Reactors (BWR). Based on experimental observations and calculations, these mismatches are attributed to the additional entrainment and deposition caused by spacer grids in BWR fuel assemblies. In COBRA-TF (Coolant Boiling in Rod Arrays-Two Fluid); a state of the art industrial best estimate code, these disagreements are hypothesized to occur due the absence of an appropriate spacer grid model. In this thesis, development of a suitably detailed spacer grid model and integrating it to COBRA-TF is documented. The new spacer grid model is highly mechanistic so that the applicability of it is not seriously affected by geometric variations in different spacer grid designs. COBRA-TF (original version) simulations performed on single tube tests and BWR rod bundles with spacer grids showed that single tube predictions were more accurate than those of the rod bundles. This observation is understood to arise from the non-availability of a suitable spacer grid model in COBRA-TF. Air water entrainment experiments were conducted in a test section simulating two adjacent BWR sub channels to visualize the flow behavior at

  19. Nefness of adjoint bundles for ample vector bundles

    Directory of Open Access Journals (Sweden)

    Hidetoshi Maeda

    1995-11-01

    Full Text Available Let E be an ample vector bundle of rank >1 on a smooth complex projective variety X of dimension n. This paper gives a classification of pairs (X,E whose adjoint bundles K_X+det E are not nef in the case when  r=n-2.

  20. Bundle Security Protocol for ION

    Science.gov (United States)

    Burleigh, Scott C.; Birrane, Edward J.; Krupiarz, Christopher

    2011-01-01

    This software implements bundle authentication, conforming to the Delay-Tolerant Networking (DTN) Internet Draft on Bundle Security Protocol (BSP), for the Interplanetary Overlay Network (ION) implementation of DTN. This is the only implementation of BSP that is integrated with ION.

  1. Vector Bundles And F Theory

    CERN Document Server

    Friedman, R; Witten, Edward

    1997-01-01

    To understand in detail duality between heterotic string and F theory compactifications, it is important to understand the construction of holomorphic G bundles over elliptic Calabi-Yau manifolds, for various groups G. In this paper, we develop techniques to describe these bundles, and make several detailed comparisons between the heterotic string and F theory.

  2. Vector Bundles And F Theory

    OpenAIRE

    Friedman, Robert; Morgan, John; Witten, Edward

    1997-01-01

    To understand in detail duality between heterotic string and F theory compactifications, it is important to understand the construction of holomorphic G bundles over elliptic Calabi-Yau manifolds, for various groups G. In this paper, we develop techniques to describe the bundles, and make several detailed comparisons between the heterotic string and F theory.

  3. Bundle Formation in Biomimetic Hydrogels

    NARCIS (Netherlands)

    Jaspers, Maarten; Pape, A C H; Voets, Ilja K; Rowan, Alan E; Portale, Giuseppe; Kouwer, Paul H J

    2016-01-01

    Bundling of single polymer chains is a crucial process in the formation of biopolymer network gels that make up the extracellular matrix and the cytoskeleton. This bundled architecture leads to gels with distinctive properties, including a large-pore-size gel formation at very low concentrations and

  4. Large-eddy simulation, fuel rod vibration and grid-to-rod fretting in pressurized water reactors

    Science.gov (United States)

    Christon, Mark A.; Lu, Roger; Bakosi, Jozsef; Nadiga, Balasubramanya T.; Karoutas, Zeses; Berndt, Markus

    2016-10-01

    Grid-to-rod fretting (GTRF) in pressurized water reactors is a flow-induced vibration phenomenon that results in wear and fretting of the cladding material on fuel rods. GTRF is responsible for over 70% of the fuel failures in pressurized water reactors in the United States. Predicting the GTRF wear and concomitant interval between failures is important because of the large costs associated with reactor shutdown and replacement of fuel rod assemblies. The GTRF-induced wear process involves turbulent flow, mechanical vibration, tribology, and time-varying irradiated material properties in complex fuel assembly geometries. This paper presents a new approach for predicting GTRF induced fuel rod wear that uses high-resolution implicit large-eddy simulation to drive nonlinear transient dynamics computations. The GTRF fluid-structure problem is separated into the simulation of the turbulent flow field in the complex-geometry fuel-rod bundles using implicit large-eddy simulation, the calculation of statistics of the resulting fluctuating structural forces, and the nonlinear transient dynamics analysis of the fuel rod. Ultimately, the methods developed here, can be used, in conjunction with operational management, to improve reactor core designs in which fuel rod failures are minimized or potentially eliminated. Robustness of the behavior of both the structural forces computed from the turbulent flow simulations and the results from the transient dynamics analyses highlight the progress made towards achieving a predictive simulation capability for the GTRF problem.

  5. Fiber bundle phase conjugate mirror

    Science.gov (United States)

    Ward, Benjamin G.

    2012-05-01

    An improved method and apparatus for passively conjugating the phases of a distorted wavefronts resulting from optical phase mismatch between elements of a fiber laser array are disclosed. A method for passively conjugating a distorted wavefront comprises the steps of: multiplexing a plurality of probe fibers and a bundle pump fiber in a fiber bundle array; passing the multiplexed output from the fiber bundle array through a collimating lens and into one portion of a non-linear medium; passing the output from a pump collection fiber through a focusing lens and into another portion of the non-linear medium so that the output from the pump collection fiber mixes with the multiplexed output from the fiber bundle; adjusting one or more degrees of freedom of one or more of the fiber bundle array, the collimating lens, the focusing lens, the non-linear medium, or the pump collection fiber to produce a standing wave in the non-linear medium.

  6. High temperature control rod assembly

    Energy Technology Data Exchange (ETDEWEB)

    Vollman, Russell E. (Solana Beach, CA)

    1991-01-01

    A high temperature nuclear control rod assembly comprises a plurality of substantially cylindrical segments flexibly joined together in succession by ball joints. The segments are made of a high temperature graphite or carbon-carbon composite. The segment includes a hollow cylindrical sleeve which has an opening for receiving neutron-absorbing material in the form of pellets or compacted rings. The sleeve has a threaded sleeve bore and outer threaded surface. A cylindrical support post has a threaded shaft at one end which is threadably engaged with the sleeve bore to rigidly couple the support post to the sleeve. The other end of the post is formed with a ball portion. A hollow cylindrical collar has an inner threaded surface engageable with the outer threaded surface of the sleeve to rigidly couple the collar to the sleeve. the collar also has a socket portion which cooperates with the ball portion to flexibly connect segments together to form a ball and socket-type joint. In another embodiment, the segment comprises a support member which has a threaded shaft portion and a ball surface portion. The threaded shaft portion is engageable with an inner threaded surface of a ring for rigidly coupling the support member to the ring. The ring in turn has an outer surface at one end which is threadably engageably with a hollow cylindrical sleeve. The other end of the sleeve is formed with a socket portion for engagement with a ball portion of the support member. In yet another embodiment, a secondary rod is slidably inserted in a hollow channel through the center of the segment to provide additional strength. A method for controlling a nuclear reactor utilizing the control rod assembly is also included.

  7. Twisted Vector Bundles on Pointed Nodal Curves

    Indian Academy of Sciences (India)

    Ivan Kausz

    2005-05-01

    Motivated by the quest for a good compactification of the moduli space of -bundles on a nodal curve we establish a striking relationship between Abramovich’s and Vistoli’s twisted bundles and Gieseker vector bundles.

  8. Vibration of the Package of Rods Linked by Spacer Grids

    Science.gov (United States)

    Zeman, V.; Hlaváč, Z.

    This paper deals with modelling and vibration analysis of the large package of identical parallel rods which are linked by transverse springs (spacer grids) placed on several level spacings. The vibration of rods is caused by the support plate motion. The rod discretization by FEM is based on Rayleigh beam theory. With respect to cyclic and central package rod symmetry, the system is decomposed to identical revolved rod segments. The modal synthesis method with condensation of the rod segments is used for modelling and determination of steady forced vibration of the whole system. The presented method is the first step to modelling of the nuclear fuel assembly vibration caused by kinematical excitation determined by motion of the support plates which are part of the reactor core.

  9. HIGH STRENGTH CONTROL RODS FOR NEUTRONIC REACTORS

    Science.gov (United States)

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

    1961-07-11

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

  10. Sensitivity study of control rod depletion coefficients

    OpenAIRE

    Blomberg, Joel

    2015-01-01

    This report investigates the sensitivity of the control rod depletion coefficients, Sg, to different input parameters and how this affects the accumulated 10B depletion, β. Currently the coefficients are generated with PHOENIX4, but the geometries can be more accurately simulated in McScram. McScram is used to calculate Control Rod Worth, which in turn is used to calculate Nuclear End Of Life, and Sg cannot be generated in the current version of McScram. Therefore, it is also analyzed whether...

  11. Actin-Interacting Protein 1 Contributes to Intranuclear Rod Assembly in Dictyostelium discoideum

    Science.gov (United States)

    Ishikawa-Ankerhold, Hellen C.; Daszkiewicz, Wioleta; Schleicher, Michael; Müller-Taubenberger, Annette

    2017-01-01

    Intranuclear rods are aggregates consisting of actin and cofilin that are formed in the nucleus in consequence of chemical or mechanical stress conditions. The formation of rods is implicated in a variety of pathological conditions, such as certain myopathies and some neurological disorders. It is still not well understood what exactly triggers the formation of intranuclear rods, whether other proteins are involved, and what the underlying mechanisms of rod assembly or disassembly are. In this study, Dictyostelium discoideum was used to examine appearance, stages of assembly, composition, stability, and dismantling of rods. Our data show that intranuclear rods, in addition to actin and cofilin, are composed of a distinct set of other proteins comprising actin-interacting protein 1 (Aip1), coronin (CorA), filactin (Fia), and the 34 kDa actin-bundling protein B (AbpB). A finely tuned spatio-temporal pattern of protein recruitment was found during formation of rods. Aip1 is important for the final state of rod compaction indicating that Aip1 plays a major role in shaping the intranuclear rods. In the absence of both Aip1 and CorA, rods are not formed in the nucleus, suggesting that a sufficient supply of monomeric actin is a prerequisite for rod formation. PMID:28074884

  12. Semiflexible Biopolymers in Bundled Arrangements

    Directory of Open Access Journals (Sweden)

    Jörg Schnauß

    2016-07-01

    Full Text Available Bundles and networks of semiflexible biopolymers are key elements in cells, lending them mechanical integrity while also enabling dynamic functions. Networks have been the subject of many studies, revealing a variety of fundamental characteristics often determined via bulk measurements. Although bundles are equally important in biological systems, they have garnered much less scientific attention since they have to be probed on the mesoscopic scale. Here, we review theoretical as well as experimental approaches, which mainly employ the naturally occurring biopolymer actin, to highlight the principles behind these structures on the single bundle level.

  13. Evaluating big deal journal bundles.

    Science.gov (United States)

    Bergstrom, Theodore C; Courant, Paul N; McAfee, R Preston; Williams, Michael A

    2014-07-01

    Large commercial publishers sell bundled online subscriptions to their entire list of academic journals at prices significantly lower than the sum of their á la carte prices. Bundle prices differ drastically between institutions, but they are not publicly posted. The data that we have collected enable us to compare the bundle prices charged by commercial publishers with those of nonprofit societies and to examine the types of price discrimination practiced by commercial and nonprofit journal publishers. This information is of interest to economists who study monopolist pricing, librarians interested in making efficient use of library budgets, and scholars who are interested in the availability of the work that they publish.

  14. Stable extensions by line bundles

    CERN Document Server

    Teixidor-i-Bigas, M

    1997-01-01

    Let C be an algebraic curve of genus g. Consider extensions E of a vector bundle F'' of rank n'' by a vector bundle F' of rank n'. The following statement was conjectured by Lange: If 0bundle. Our method uses a degeneration argument to a reducible curve.

  15. Numerical model for thermal and mechanical behaviour of a CANDU 37-element bundle

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, L.; MacKay, K. [Martec Limited, Halifax, Nova Scotia (Canada); Gibb, R. [Canadian Nuclear Safety Commission (CNSC), Ottawa, Ontario (Canada)

    2010-07-01

    Prediction of transient fuel bundle deformations is important for assessing the integrity of fuel and the surrounding structural components under different operating conditions including accidents. For numerical simulation of the interactions between fuel bundle and pressure tube, a reliable numerical bundle model is required to predict thermal and mechanical behaviour of the fuel bundle assembly under different thermal loading conditions. To ensure realistic representations of the bundle behaviour, this model must include all of the important thermal and mechanical features of the fuel bundle, such as temperature-dependent material properties, thermal viscoplastic deformation in sheath, fuel-to-sheath interactions, endplate constraints and contacts between fuel elements. In this paper, we present a finite element based numerical model for predicting macroscopic transient thermal-mechanical behaviour of a complete 37-element CANDU nuclear fuel bundle under accident conditions and demonstrate its potential for being used to investigate fuel bundle to pressure tube interaction in future nuclear safety analyses. This bundle model has been validated against available experimental and numerical solutions and applied to various simulations involving steady-state and transient loading conditions. (author)

  16. The Atiyah Bundle and Connections on a Principal Bundle

    Indian Academy of Sciences (India)

    Indranil Biswas

    2010-06-01

    Let be a ∞ manifold and a Lie a group. Let $E_G$ be a ∞ principal -bundle over . There is a fiber bundle $\\mathcal{C}(E_G)$ over whose smooth sections correspond to the connections on $E_G$. The pull back of $E_G$ to $\\mathcal{C}(E_G)$ has a tautological connection. We investigate the curvature of this tautological connection.

  17. Results of international standard problem No. 36 severe fuel damage experiment of a VVER fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Firnhaber, M. [Gesellschaft fuer Anlagen-und Reaktorsicherheit, Koeln (Germany); Yegorova, L. [Nuclear Safety Institute of Russian Research Center, Moscow (Russian Federation); Brockmeier, U. [Ruhr-Univ. of Bochum (Germany)] [and others

    1995-09-01

    International Standard Problems (ISP) organized by the OECD are defined as comparative exercises in which predictions with different computer codes for a given physical problem are compared with each other and with a carefully controlled experimental study. The main goal of ISP is to increase confidence in the validity and accuracy of analytical tools used in assessing the safety of nuclear installations. In addition, it enables the code user to gain experience and to improve his competence. This paper presents the results and assessment of ISP No. 36, which deals with the early core degradation phase during an unmitigated severe LWR accident in a Russian type VVER. Representatives of 17 organizations participated in the ISP using the codes ATHLET-CD, ICARE2, KESS-III, MELCOR, SCDAP/RELAP5 and RAPTA. Some participants performed several calculations with different codes. As experimental basis the severe fuel damage experiment CORA-W2 was selected. The main phenomena investigated are thermal behavior of fuel rods, onset of temperature escalation, material behavior and hydrogen generation. In general, the calculations give the right tendency of the experimental results for the thermal behavior, the hydrogen generation and, partly, for the material behavior. However, some calculations deviate in important quantities - e.g. some material behavior data - showing remarkable discrepancies between each other and from the experiments. The temperature history of the bundle up to the beginning of significant oxidation was calculated quite well. Deviations seem to be related to the overall heat balance. Since the material behavior of the bundle is to a great extent influenced by the cladding failure criteria a more realistic cladding failure model should be developed at least for the detailed, mechanistic codes. Regarding the material behavior and flow blockage some models for the material interaction as well as for relocation and refreezing requires further improvement.

  18. Left bundle-branch block

    DEFF Research Database (Denmark)

    Risum, Niels; Strauss, David; Sogaard, Peter

    2013-01-01

    The relationship between myocardial electrical activation by electrocardiogram (ECG) and mechanical contraction by echocardiography in left bundle-branch block (LBBB) has never been clearly demonstrated. New strict criteria for LBBB based on a fundamental understanding of physiology have recently...

  19. Bundling ecosystem services in Denmark

    DEFF Research Database (Denmark)

    Turner, Katrine Grace; Odgaard, Mette Vestergaard; Bøcher, Peder Klith;

    2014-01-01

    We made a spatial analysis of 11 ecosystem services at a 10 km × 10 km grid scale covering most of Denmark. Our objective was to describe their spatial distribution and interactions and also to analyze whether they formed specific bundle types on a regional scale in the Danish cultural landscape....... We found clustered distribution patterns of ecosystem services across the country. There was a significant tendency for trade-offs between on the one hand cultural and regulating services and on the other provisioning services, and we also found the potential of regulating and cultural services...... to form synergies. We identified six distinct ecosystem service bundle types, indicating multiple interactions at a landscape level. The bundle types showed specialized areas of agricultural production, high provision of cultural services at the coasts, multifunctional mixed-use bundle types around urban...

  20. Left bundle-branch block

    DEFF Research Database (Denmark)

    Risum, Niels; Strauss, David; Sogaard, Peter;

    2013-01-01

    The relationship between myocardial electrical activation by electrocardiogram (ECG) and mechanical contraction by echocardiography in left bundle-branch block (LBBB) has never been clearly demonstrated. New strict criteria for LBBB based on a fundamental understanding of physiology have recently...

  1. Vector bundles on toric varieties

    CERN Document Server

    Gharib, Saman

    2011-01-01

    Following Sam Payne's work, we study the existence problem of nontrivial vector bundles on toric varieties. The first result we prove is that every complete fan admits a nontrivial conewise linear multivalued function. Such functions could potentially be the Chern classes of toric vector bundles. Then we use the results of Corti\\~nas, Haesemeyer, Walker and Weibel to show that the (non-equivariant) Grothendieck group of the toric 3-fold studied by Payne is large, so the variety has a nontrivial vector bundle. Using the same computation, we show that every toric 3-fold X either has a nontrivial line bundle, or there is a finite surjective toric morphism from Y to X, such that Y has a large Grothendieck group.

  2. Fabrication of electrospun nanofibers bundles

    Science.gov (United States)

    Ye, Junjun; Sun, Daoheng

    2007-12-01

    Aligned nanofibers, filament bundle composed of large number of nanofibers have potential applications such as bio-material, composite material etc. A series of electrospinning experiments have been conducted to investigate the electrospinning process,in which some parameters such as polymer solution concentration, bias voltage, distance between spinneret and collector, solution flow rate etc have been setup to do the experiment of nanofibers bundles construction. This work firstly reports electrospun nanofiber bundle through non-uniform electrical field, and nanofibers distributed in different density on electrodes from that between them. Thinner nanofibers bundle with a few numbers of nanofiber is collected for 3 seconds; therefore it's also possible that the addressable single nanofiber could be collected to bridge two electrodes.

  3. Reconnection of superfluid vortex bundles.

    Science.gov (United States)

    Alamri, Sultan Z; Youd, Anthony J; Barenghi, Carlo F

    2008-11-21

    Using the vortex filament model and the Gross-Pitaevskii nonlinear Schroedinger equation, we show that bundles of quantized vortex lines in He II are structurally robust and can reconnect with each other maintaining their identity. We discuss vortex stretching in superfluid turbulence and show that, during the bundle reconnection process, kelvin waves of large amplitude are generated, in agreement with the finding that helicity is produced by nearly singular vortex interactions in classical Euler flows.

  4. Single Rod Vibration in Axial Flow

    Science.gov (United States)

    Weichselbaum, Noah; Wang, Shengfu; Bardet, Philippe

    2013-11-01

    Fluid structure interaction of a single rod in axial flow is a coupled dynamical system present in many application including nuclear reactors, steam generators, and towed antenna arrays. Fluid-structure response can be quantified thanks to detailed experimental data where both structure and fluid responses are recorded. Such datum deepen understanding of the physics inherent to the system and provide high-dimensionality quantitative measurements to validate coupled structural and CFD codes with various level of complexity. In this work, single rods fixed on both ends in a concentric pipe, are subjected to an axial flow with Reynolds number based on hydraulic diameter of Re =4000. Rods of varying material stiffness and diameter are utilized in the experiment resulting in a range of dimensionless U between 0.5 and 1, where U = (ρA/EI)1/2uL. Experimental measurements of the velocity field around the rod are taken with PIV from time-resolved Nd:YLF laser and a high speed CMOS camera. Three-dimensional and temporal vibration and deflection of the rod is recorded with shadowgraphy utilizing two sets of pulsed high power LED and dedicated CMOS camera. Through integration of these two diagnostics, it is possible to reconstruct the full FSI domain providing unique validation data.

  5. Evaluation of Ultrasonic Time-of-Flight Diffraction Data for Selected Control Rod Drive Nozzles from Davis Besse Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Cumblidge, Stephen E.; Cinson, Anthony D.; Anderson, Michael T.

    2011-04-25

    Pacific Northwest National Laboratory (PNNL) examined ultrasonic (UT) time-of-flight diffraction (TOFD) data from ten (10) nozzles in the Davis Besse Nuclear Power Plant reactor closure head. The TOFD data was acquired by AREVA after a bare metal visual examination of the pressure vessel head indicated potential leakage in at least one nozzle. A detailed analysis of the UT data shows that Nozzle 4 has three indications consistent with cracking in the penetration tube. One of the indications starts at the wetted side of the weld and progresses to the annulus. In addition, examination of UT data from the annulus region of Nozzle 4 displays an irregular pattern that could be associated with boric acid deposits and leakage/wastage in the interference fit. The review of TOFD data for the other nine nozzles resulted in several indications being detected in the weld region and near the inner diameter (ID) and outer diameter (OD) surfaces of the penetration tube, but no other indications that are consistent with cracking that may have resulted in leakage were observed. A review of the back-wall reflections in the other nine nozzles also did not show strong indications of leakage, although Nozzle 67 displayed an irregularly-shaped region of high ultrasonic transmission near 180 degrees on the interference fit.

  6. The Analysis of SBWR Critical Power Bundle Using Cobrag Code

    Directory of Open Access Journals (Sweden)

    Yohannes Sardjono

    2013-03-01

    Full Text Available The coolant mechanism of SBWR is similar with the Dodewaard Nuclear Power Plant (NPP in the Netherlands that first went critical in 1968. The similarity of both NPP is cooled by natural convection system. These coolant concept is very related with same parameters on fuel bundle design especially fuel bundle length, core pressure drop and core flow rate as well as critical power bundle. The analysis was carried out by using COBRAG computer code. COBRAG computer code is GE Company proprietary. Basically COBRAG computer code is a tool to solve compressible three-dimensional, two fluid, three field equations for two phase flow. The three fields are the vapor field, the continuous liquid field, and the liquid drop field. This code has been applied to analyses model flow and heat transfer within the reactor core. This volume describes the finitevolume equations and the numerical solution methods used to solve these equations. This analysis of same parameters has been done i.e.; inlet sub cooling 20 BTU/lbm and 40 BTU/lbm, 1000 psi pressure and R-factor is 1.038, mass flux are 0.5 Mlb/hr.ft2, 0.75 Mlb/hr.ft2, 1.00 Mlb/hr.ft2 and 1.25 Mlb/hr.ft2. Those conditions based on history operation of some type of the cell fuel bundle line at GE Nuclear Energy. According to the results, it can be concluded that SBWR critical power bundle is 10.5 % less than current BWR critical power bundle with length reduction of 12 ft to 9 ft.

  7. Bundle Formation in Biomimetic Hydrogels.

    Science.gov (United States)

    Jaspers, Maarten; Pape, A C H; Voets, Ilja K; Rowan, Alan E; Portale, Giuseppe; Kouwer, Paul H J

    2016-08-08

    Bundling of single polymer chains is a crucial process in the formation of biopolymer network gels that make up the extracellular matrix and the cytoskeleton. This bundled architecture leads to gels with distinctive properties, including a large-pore-size gel formation at very low concentrations and mechanical responsiveness through nonlinear mechanics, properties that are rarely observed in synthetic hydrogels. Using small-angle X-ray scattering (SAXS), we study the bundle formation and hydrogelation process of polyisocyanide gels, a synthetic material that uniquely mimics the structure and mechanics of biogels. We show how the structure of the material changes at the (thermally induced) gelation point and how factors such as concentration and polymer length determine the architecture, and with that, the mechanical properties. The correlation of the gel mechanics and the structural parameters obtained from SAXS experiments is essential in the design of future (synthetic) mimics of biopolymer networks.

  8. Treatment of defective fuel rods for interim storage

    Energy Technology Data Exchange (ETDEWEB)

    Muenchow, K.; Hummel, W. [AREVA NP GmbH, Erlangen (Germany)

    2013-07-01

    In this paper we look exclusively at the treatment of defective fuel rods for long-term dry interim storage at the nuclear power plant, in order to avoid off-site transports. AREVA has developed a technique that allows verifiably adequate drying of the defective fuel rods and reconstructs the barrier for retaining radioactive materials. This is done by individually encapsulating the defective fuel rods and achieving gas-tightness by seal welding. This guarantees the retention of radioactive materials during the storage period of at least 40 years in a transport and storage flask in an interim storage facility at site. (orig.)

  9. Extensional bundle waveguide techniques for measuring flow of hot fluids.

    Science.gov (United States)

    Lynnworth, Lawrence C; Liu, Yi; Umina, John A

    2005-04-01

    A bundle of acoustically slender metal rods, each thin compared to wavelength, tightly packed within a sheath, and welded closed at each end, provides a dispersion-free waveguide assembly that acts as a thermal buffer between a transducer and the hot fluid medium the flow of which is to be measured. Gas and steam flow applications have ranged up to 600 degrees C. Liquid applications have ranged from cryogenic (-160 degrees C) to 500 degrees C and include intermittent two-phase flows. The individual rods comprising the bundle usually are approximately one millimeter in diameter. The sheath, made of a pipe or tube, typically has an outside diameter of 12.7 to about 33 mm and usually is about 300 mm long. Materials for the sheath and bundle are selected to satisfy requirements of compatibility with the fluid as well as for acoustic properties. Corrosion-resistant alloys such as 316SS and titanium are commonly used. The buffers are used with transducers that are metal-encapsulated and certified for use in hazardous areas. They operate at a frequency in the range of 0.1 to 1 MHz. The radiating end of the buffer is usually flat and perpendicular to the buffer's main axis. In some cases the end of the buffer is stepped or angled. Angling the radiating faces at approximately 2 degrees to overcome beam drift at Mach 0.1 recently contributed to solving a high-temperature high-velocity flow measurement problem. The temperature in this situation was 300 degrees C, and the gas molecular weight was about 95, with pressure 0.9 to 1.1 bar.

  10. Principal bundles the classical case

    CERN Document Server

    Sontz, Stephen Bruce

    2015-01-01

    This introductory graduate level text provides a relatively quick path to a special topic in classical differential geometry: principal bundles.  While the topic of principal bundles in differential geometry has become classic, even standard, material in the modern graduate mathematics curriculum, the unique approach taken in this text presents the material in a way that is intuitive for both students of mathematics and of physics. The goal of this book is to present important, modern geometric ideas in a form readily accessible to students and researchers in both the physics and mathematics communities, providing each with an understanding and appreciation of the language and ideas of the other.

  11. Patterns of rod proliferation in deep-sea fish retinae.

    Science.gov (United States)

    Fröhlich, E; Negishi, K; Wagner, H J

    1995-07-01

    In a sample of 37 species of deep-sea fish species from the sea floor of the Porcupine Seabight and the Gobal spur (North Atlantic) we investigated the overall structure of the retina with special respect for the organization of rods, their length and their arrangement in multiple banks. Using an immunocytochemical marker for cell proliferation (PCNA) we studied the mechanisms of rod proliferation, and, by means of serial section reconstruction followed their integration into the existing population of rods. Furthermore, in three different species we have observed growth related changes in retinal thickness, rod density and proliferation activity. We found evidence for two different principles for the organization of rods in these deep-sea fish retinae. In the first group of species represented by Nematonurus armatus and Coryphaenoides guentheri we found rods to be rather short (20-30 microns) and arranged in three and more banks. In these species rod proliferation occurred in a single band of cells immediately vitread of the external limiting membrane, thus showing a high degree of spatial and temporal order. In these species, young rods are inserted just sclerad of the external limiting membrane and the older outer segments pushed away from the incoming light towards the back of the eye. This may be linked to a progressive loss of function of the older rods and might represent an alternative mechanism to the disk shedding in other vertebrates. In the second population (e.g. Conocara macroptera, Alepocephalus agassizii) we observed considerably longer rod outer segments (60-80 microns) forming no more than two layers. These retinae had rod precursors arranged in disseminated clusters throughout the outer nuclear layer indicating the lack of clear spatio-temporal order in mitotic activity along with a more statistical pattern of integration of the newly formed outer segments. In our sample of species both populations were of about equal size suggesting that the two

  12. Morphoelastic rods. Part I: A single growing elastic rod

    KAUST Repository

    Moulton, D.E.

    2013-02-01

    A theory for the dynamics and statics of growing elastic rods is presented. First, a single growing rod is considered and the formalism of three-dimensional multiplicative decomposition of morphoelasticity is used to describe the bulk growth of Kirchhoff elastic rods. Possible constitutive laws for growth are discussed and analysed. Second, a rod constrained or glued to a rigid substrate is considered, with the mismatch between the attachment site and the growing rod inducing stress. This stress can eventually lead to instability, bifurcation, and buckling. © 2012 Elsevier Ltd. All rights reserved.

  13. Learning with Rods: One Account.

    Science.gov (United States)

    Cherry, Donald Esha

    This paper discusses one English as a Second Language (ESL) teacher's attempts to use cuisenaire rods as a language learning tool. Cuisenaire rods (sometimes called algebricks) vary in size from 1 x 1 x 10 centimeter sticks to 1 x 1 x 1 centimeter cubes, with each of the 10 sizes a different color. Although such rods have been used to teach…

  14. Fabrication of preliminary fuel rods for SFR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sun Ki; Oh, Seok Jin; Ko, Young Mo; Woo, Youn Myung; Kim, Ki Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    Metal fuels was selected for fueling many of the first reactors in the US, including the Experimental Breeder Reactor-I (EBR-I) and the Experimental Breeder Reactor-II (EBR-II) in Idaho, the FERMI-I reactor, and the Dounreay Fast Reactor (DFR) in the UK. Metallic U.Pu.Zr alloys were the reference fuel for the US Integral Fast Reactor (IFR) program. Metallic fuel has advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant and inherent passive safety. U-Zr-Pu alloy fuels have been used for SFR (sodium-cooled fast reactor) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980s. Fabrication technology of metallic fuel for SFR has been in development in Korea as a national nuclear R and D program since 2007. For the final goal of SFR fuel rod fabrication with good performance, recently, three preliminary fuel rods were fabricated. In this paper, the preliminary fuel rods were fabricated, and then the inspection for QC(quality control) of the fuel rods was performed

  15. Creep rupture of fiber bundles

    DEFF Research Database (Denmark)

    Linga, G.; Ballone, P.; Hansen, Alex

    2015-01-01

    The creep deformation and eventual breaking of polymeric samples under a constant tensile load F is investigated by molecular dynamics based on a particle representation of the fiber bundle model. The results of the virtual testing of fibrous samples consisting of 40000 particles arranged on Nc=4...

  16. Line bundles and flat connections

    Indian Academy of Sciences (India)

    INDRANIL BISWAS; GEORG SCHUMACHER

    2017-06-01

    We prove that there are cocompact lattices $\\Gamma$ in $\\rm SL(2,\\mathbb C)$ with the property that there are holomorphic line bundles $L$ on $\\rm SL(2,\\mathbb C)/ \\Gamma$ with $c_{1}(L) = 0$ such that $L$ does not admit any unitary flat connection.

  17. Vector Bundles over Elliptic Fibrations

    CERN Document Server

    Friedman, R; Witten, Edward; Friedman, Robert; Morgan, John W.; Witten, Edward

    1997-01-01

    This paper gives various methods for constructing vector bundles over elliptic curves and more generally over families of elliptic curves. We construct universal families over generalized elliptic curves via spectral cover methods and also by extensions, and then give a relative version of the construction in families. We give various examples and make Chern class computations.

  18. Safety rod latch inspection

    Energy Technology Data Exchange (ETDEWEB)

    Leader, D.R.

    1992-02-01

    During an attempt to raise control rods from the 100 K reactor in December, one rod could not be withdrawn. Subsequent investigation revealed that a small button'' in the latch mechanism had broken off of the lock plunger'' and was wedged in a position that prevented rod withdrawal. Concern that this failure may have resulted from corrosion or some other metallurgical problem resulted in a request that SRL examine six typical latch mechanisms from the 100 L reactor by use of radiography and metallography. During the examination of the L-Area latches, a failed latch mechanism from the 100 K reactor was added to the investigation. Fourteen latches that had a history of problems were removed from K-Area and sent to SRL for inclusion in this study the week after the original seven assemblies were examined, bringing the total of latch assemblies discussed in this report to twenty one. Results of the examination of the K-Area latch that initiated this study is not included in this report.

  19. Safety rod latch inspection

    Energy Technology Data Exchange (ETDEWEB)

    Leader, D.R.

    1992-02-01

    During an attempt to raise control rods from the 100 K reactor in December, one rod could not be withdrawn. Subsequent investigation revealed that a small ``button`` in the latch mechanism had broken off of the ``lock plunger`` and was wedged in a position that prevented rod withdrawal. Concern that this failure may have resulted from corrosion or some other metallurgical problem resulted in a request that SRL examine six typical latch mechanisms from the 100 L reactor by use of radiography and metallography. During the examination of the L-Area latches, a failed latch mechanism from the 100 K reactor was added to the investigation. Fourteen latches that had a history of problems were removed from K-Area and sent to SRL for inclusion in this study the week after the original seven assemblies were examined, bringing the total of latch assemblies discussed in this report to twenty one. Results of the examination of the K-Area latch that initiated this study is not included in this report.

  20. Quantum principal bundles and corresponding gauge theories

    CERN Document Server

    Durdevic, M

    1995-01-01

    A generalization of classical gauge theory is presented, in the framework of a noncommutative-geometric formalism of quantum principal bundles over smooth manifolds. Quantum counterparts of classical gauge bundles, and classical gauge transformations, are introduced and investigated. A natural differential calculus on quantum gauge bundles is constructed and analyzed. Kinematical and dynamical properties of corresponding gauge theories are discussed.

  1. Strategic and welfare implications of bundling

    DEFF Research Database (Denmark)

    Martin, Stephen

    1999-01-01

    A standard oligopoly model of bundling shows that bundling by a firm with a monopoly over one product has a strategic effect because it changes the substitution relationships between the goods among which consumers choose. Bundling in appropriate proportions is privately profitable, reduces rival......' profits and overall welfare, and may drive rivals from the market...

  2. Sperm bundle and reproductive organs of carabid beetles tribe Pterostichini (Coleoptera: Carabidae)

    Science.gov (United States)

    Sasakawa, Kôji

    2007-05-01

    The morphological characteristics of sperm and reproductive organs may offer clues as to how reproductive systems have evolved. In this paper, the morphologies of the sperm and male reproductive organs of carabid beetles in the tribe Pterostichini (Coleoptera: Carabidae) are described, and the morphological associations among characters are examined. All species form sperm bundles in which the head of the sperm was embedded in a rod-shaped structure, i.e., spermatodesm. The spermatodesm shape (left-handed spiral, right-handed spiral, or without conspicuous spiral structure) and the condition of the sperm on the spermatodesm surface (with the tail free-moving or forming a thin, sheetlike structure) vary among species. In all species, the spiral directions of the convoluted seminal vesicles and vasa deferentia are the same on both sides of the body; that is, they show an asymmetric structure. The species in which the sperm bundle and the seminal vesicles both have a spiral structure could be classified into two types, with significant differences in sperm-bundle length between the two types. The species with a sperm-bundle spiral and seminal-vesicle spiral of almost the same diameter have longer sperm bundles than the species with a sperm-bundle spiral and seminal-vesicle tube of almost the same diameter. In the former type, the spiral directions of the sperm bundles and seminal vesicles are inevitably the same, whereas they differ in some species with the later type. Therefore, increased sperm bundle length appears to have been facilitated by the concordance of the sperm bundle’s coiling direction with the coiling direction of the seminal vesicle.

  3. Higher order jet prolongations type gauge natural bundles over vector bundles

    Directory of Open Access Journals (Sweden)

    Jan Kurek

    2004-05-01

    Full Text Available Let $rgeq 3$ and $mgeq 2$ be natural numbers and $E$ be a vector bundle with $m$-dimensional basis. We find all gauge natural bundles ``similar" to the $r$-jet prolongation bundle $J^rE$ of $E$. We also find all gauge natural bundles ``similar" to the vector $r$-tangent bundle $(J^r_{fl}(E,R_0^*$ of $E$.

  4. Multipath packet switch using packet bundling

    DEFF Research Database (Denmark)

    Berger, Michael Stubert

    2002-01-01

    The basic concept of packet bundling is to group smaller packets into larger packets based on, e.g., quality of service or destination within the packet switch. This paper presents novel applications of bundling in packet switching. The larger packets created by bundling are utilized to extend...... switching capacity by use of parallel switch planes. During the bundling operation, packets will experience a delay that depends on the actual implementation of the bundling and scheduling scheme. Analytical results for delay bounds and buffer size requirements are presented for a specific scheduling...

  5. Mathematical modelling for nanotube bundle oscillators

    Science.gov (United States)

    Thamwattana, Ngamta; Cox, Barry J.; Hill, James M.

    2009-07-01

    This paper investigates the mechanics of a gigahertz oscillator comprising a nanotube oscillating within the centre of a uniform concentric ring or bundle of nanotubes. The study is also extended to the oscillation of a fullerene inside a nanotube bundle. In particular, certain fullerene-nanotube bundle oscillators are studied, namely C60-carbon nanotube bundle, C60-boron nitride nanotube bundle, B36N36-carbon nanotube bundle and B36N36-boron nitride nanotube bundle. Using the Lennard-Jones potential and the continuum approach, we obtain a relation between the bundle radius and the radii of the nanotubes forming the bundle, as well as the optimum bundle size which gives rise to the maximum oscillatory frequency for both the fullerene and the nanotube bundle oscillators. While previous studies in this area have been undertaken through molecular dynamics simulations, this paper emphasizes the use of applied mathematical modelling techniques which provides considerable insight into the underlying mechanisms. The paper presents a synopsis of the major results derived in detail by the present authors in [1, 2].

  6. Photonic bandgap fiber bundle spectrometer

    CERN Document Server

    Hang, Qu; Syed, Imran; Guo, Ning; Skorobogatiy, Maksim

    2010-01-01

    We experimentally demonstrate an all-fiber spectrometer consisting of a photonic bandgap fiber bundle and a black and white CCD camera. Photonic crystal fibers used in this work are the large solid core all-plastic Bragg fibers designed for operation in the visible spectral range and featuring bandgaps of 60nm - 180nm-wide. 100 Bragg fibers were chosen to have complimentary and partially overlapping bandgaps covering a 400nm-840nm spectral range. The fiber bundle used in our work is equivalent in its function to a set of 100 optical filters densely packed in the area of ~1cm2. Black and white CCD camera is then used to capture spectrally "binned" image of the incoming light at the output facet of a fiber bundle. To reconstruct the test spectrum from a single CCD image we developed an algorithm based on pseudo-inversion of the spectrometer transmission matrix. We then study resolution limit of this spectroscopic system by testing its performance using spectrally narrow test peaks (FWHM 5nm-25nm) centered at va...

  7. Report on the evaluation of the tritium producing burnable absorber rod lead test assembly. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-03-01

    This report describes the design and fabrication requirements for a tritium-producing burnable absorber rod lead test assembly and evaluates the safety issues associated with tritium-producing burnable absorber rod irradiation on the operation of a commercial light water reactor. The report provides an evaluation of the tritium-producing burnable absorber rod design and concludes that irradiation can be performed within U.S. Nuclear Regulatory Commission regulations applicable to a commercial pressurized light water reactor.

  8. The results of postirradiation examinations of VVER-1000 and VVER-440 fuel rods

    Science.gov (United States)

    Dubrovin, K. P.; Ivanov, E. G.; Strijov, P. N.; Yakovlev, V. V.

    1991-02-01

    The paper presents the results of postirradiation examination of the fuel rods having different fuel-cladding gaps, pellet densities, pellet inner diameters and so on. The fuel rods were irradiated in the material science reactor (MR) of the Kurchatov Institute of Atomic Energy and at 4 unit of the Novo-Voronezh nuclear powerplant. Some data on fission gas release and rod geometry and compared with computer code predictions.

  9. Cone rod dystrophies

    Directory of Open Access Journals (Sweden)

    Hamel Christian P

    2007-02-01

    Full Text Available Abstract Cone rod dystrophies (CRDs (prevalence 1/40,000 are inherited retinal dystrophies that belong to the group of pigmentary retinopathies. CRDs are characterized by retinal pigment deposits visible on fundus examination, predominantly localized to the macular region. In contrast to typical retinitis pigmentosa (RP, also called the rod cone dystrophies (RCDs resulting from the primary loss in rod photoreceptors and later followed by the secondary loss in cone photoreceptors, CRDs reflect the opposite sequence of events. CRD is characterized by primary cone involvement, or, sometimes, by concomitant loss of both cones and rods that explains the predominant symptoms of CRDs: decreased visual acuity, color vision defects, photoaversion and decreased sensitivity in the central visual field, later followed by progressive loss in peripheral vision and night blindness. The clinical course of CRDs is generally more severe and rapid than that of RCDs, leading to earlier legal blindness and disability. At end stage, however, CRDs do not differ from RCDs. CRDs are most frequently non syndromic, but they may also be part of several syndromes, such as Bardet Biedl syndrome and Spinocerebellar Ataxia Type 7 (SCA7. Non syndromic CRDs are genetically heterogeneous (ten cloned genes and three loci have been identified so far. The four major causative genes involved in the pathogenesis of CRDs are ABCA4 (which causes Stargardt disease and also 30 to 60% of autosomal recessive CRDs, CRX and GUCY2D (which are responsible for many reported cases of autosomal dominant CRDs, and RPGR (which causes about 2/3 of X-linked RP and also an undetermined percentage of X-linked CRDs. It is likely that highly deleterious mutations in genes that otherwise cause RP or macular dystrophy may also lead to CRDs. The diagnosis of CRDs is based on clinical history, fundus examination and electroretinogram. Molecular diagnosis can be made for some genes, genetic counseling is

  10. Analysis of High Temperature Reactor Control Rod Worth for the Initial and Full Core

    Science.gov (United States)

    Oktajianto, Hammam; Setiawati, Evi; Anam, Khoirul; Sugito, Heri

    2017-01-01

    Control rod is one important component in a nuclear reactor. In nuclear reactor operations the control rod functions to shut down the reactor. This research analyses ten control rods worth of HTR (High Temperature Reactor) at initial and full core. The HTR in this research adopts HTR-10 China and HTR- of pebble bed. Core calculations are performed by using MCNPX code after modelling the entire parts of core in condition of ten control rods fully withdrawn, all control rods in with 20 cm ranges of depth and the use of one control rod. Pebble bed and moderator balls are distributed in the core zone using a Body Centred Cubic (BCC) lattice by ratio of 57:43. The research results are obtained that the use of one control rod will decrease the reactor criticality of 2.04±0.12 %Δk/k at initial core and 1.57±0.10 %Δk/k at full core. The deeper control rods are in, the lesser criticality of reactor is with reactivity of ten control rods of 16.41±0.11 %Δk/k at initial core and 15.43±0.11 %Δk/k at full core. The results show that the use of ten control rods at full core will keep achieving subcritical condition even though the reactivity is smaller than reactivity at initial core.

  11. Study of heat transfer in a 7-element bundle cooled with the upward flow of supercritical Freon-12

    Science.gov (United States)

    Richards, Graham

    Experimental data on SuperCritical-Water (SCW) cooled bundles are very limited. Major problems with performing such experiments are: 1) small number of operating SCW experimental setups and 2) difficulties in testing and experimental costs at very high pressures, temperatures and heat fluxes. However, SuperCritical Water-cooled nuclear Reactor (SCWRs) designs cannot be finalized without such data. Therefore, as a preliminary approach experiments in SCW-cooled bare tubes and in bundles cooled with SC modeling fluids can be used. One of the SC modeling fluids typically used is Freon- 12 (R-12) where the critical pressure is 4.136 MPa and the critical temperature is 111.97ºC. These conditions correspond to a critical pressure of 22.064 MPa and critical temperature of 373.95ºC in water. A set of experimental data obtained in a Freon-12 cooled vertical bare bundle at the Institute of Physics and Power Engineering (IPPE, Obninsk, Russia) was analyzed. This set consisted of 20 cases of a vertically oriented 7-element bundle installed in a hexagonal flow channel. To secure the bundle in the flow channel 3 thin spacers were used. The dataset was obtained at equivalent parameters of the proposed SCWR concepts. Data was collected at pressures of about 4.65 MPa for several different combinations of wall and bulk-fluid temperatures that were below, at, or above the pseudocritical temperature. Heat fluxes ranged from 9 kW/m2 to 120 kW/m2 and mass fluxes ranged from 440 kg/m2s to 1320 kg/m2s. Also inlet temperatures ranged from 70ºC -- 120ºC. The test section consisted of fuel elements that were 9.5 mm in diameter with the total heated length of 1 m. Bulk-fluid and wall temperature profiles were recorded using a combination of 8 different thermocouples. The data was analyzed with respect to its temperature profile and heat transfer coefficient along the heated length of the test section. In a previous study it was confirmed that there is the existence of three distinct

  12. High temperature control rod assembly

    Energy Technology Data Exchange (ETDEWEB)

    Vollman, R.E.

    1991-12-24

    This patent describes a control rod assembly for use in nuclear reactor control. It comprises segments, each the segment being made of a graphite composite material, each the segment having a chamber for containing neutron-absorbing material, wherein the chamber compromises a hollow cylindrical sleeve having a first end formed with an opening for receiving the neutron-absorbing material, and having a second end formed with a sleeve bore and an outer sleeve surface; a cylindrical weight-bearing support post positioned substantially centrally of the sleeve, the support post having a first end formed as a ball surface portion and a second end formed as a ball surface portion and a second end formed as a shaft, the shaft being engageable with the sleeve bore for rigidly coupling the support post axially within the hollow sleeve, a hollow cylindrical collar having a socket lip portion correspondingly shaped to receive the ball surface portion of an adjacent support post, and having an inner surface for engaging the outer sleeve surface on the second end of the sleeve to rigidly couple the collar to the sleeve.

  13. Active Brownian rods

    Science.gov (United States)

    Peruani, Fernando

    2016-11-01

    Bacteria, chemically-driven rods, and motility assays are examples of active (i.e. self-propelled) Brownian rods (ABR). The physics of ABR, despite their ubiquity in experimental systems, remains still poorly understood. Here, we review the large-scale properties of collections of ABR moving in a dissipative medium. We address the problem by presenting three different models, of decreasing complexity, which we refer to as model I, II, and III, respectively. Comparing model I, II, and III, we disentangle the role of activity and interactions. In particular, we learn that in two dimensions by ignoring steric or volume exclusion effects, large-scale nematic order seems to be possible, while steric interactions prevent the formation of orientational order at large scales. The macroscopic behavior of ABR results from the interplay between active stresses and local alignment. ABR exhibit, depending on where we locate ourselves in parameter space, a zoology of macroscopic patterns that ranges from polar and nematic bands to dynamic aggregates.

  14. Fall simulation of a control rod in Trillo Nuclear Power Plant with CASMO-4/SIMULATE-3; Simulacion de la caida de un dedo de una barra de control en C.N. Trillo con CASMO-4/SIMULATE-3

    Energy Technology Data Exchange (ETDEWEB)

    Bermejo, J. A.; Crespo, A.; Novo, M.; Ruiz, J. M.

    2010-07-01

    As of the general model CASMO-4/SIMULATE-3, two new models has been developed to analyze in detail the effect of a finger fall of a control rod in the power element and in the adjacent elements. It is also possible to analyze the theoretical response of LVD and KMS systems before such events.

  15. Control of Rod-Rod Interactions in Poly(3-alkylthiophenes)

    Science.gov (United States)

    Ho, Victor; Boudouris, Bryan W.; Segalman, Rachel A.

    2010-03-01

    Poly(3-hexylthiophene) is a commonly used semiconducting polymer because of its relatively high charge transport ability, low band gap, and solution processiblity. Strong intermolecular interactions lead to the formation of nanofibers during crystallization, which prevents long-range microstructural ordering. We show rod-rod interactions, parameterized by the Maier-Saupe parameter, can be controlled by rational polythiophene side chain design. Effects of side chain passivation are evidenced by a depressed melting temperature and the presence of a liquid crystalline region. Additionally, the Maier-Saupe parameters are estimated for poly(3-dodecylthiophene) and poly(3-ethylhexylthiophene); the relative magnitudes of each are related to the interchain spacings obtained by x-ray diffraction experiments. The systematic tuning of the rod-rod interactions in polythiophenes allows for manipulation of the ratio of Maier-Saupe to the Flory-Huggins parameter, a crucial value in obtaining long-range order in rod-coil block copolymer morphologies.

  16. Effect of Entry/Exit Length on Flow Distribution in the Test Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Byeong Il; Jang, Beom Jun; Kim, Hong Ju; Kim, Kanghoon; Nahm, Kee Yil; Park, Sang Weon [KEPCO Nuclear Fuel, Daejeon (Korea, Republic of)

    2014-05-15

    In data analysis, the geometric information within the heated section of the rod bundle is important because the CHF occurs in the heated section. To ensure a constant geometry and to prevent adverse flow effects, it is required to extend the same geometry beyond the heated section of rod bundle geometry. Regarding to evaluate the validity of inlet boundary condition of subchannel analysis code, the effect of the entry and exit length on the flow distribution is evaluated under the various inlet flow conditions which could be produced without flow distributor or strainer. To evaluate the validation of the inlet and outlet boundary conditions used in the subchannel analysis code, a study on the effect of the entry and exit length on the flow distribution is conducted. Even though the non-uniform flow is entered inside the test bundle, the flow gets more saturated by the simple supports and frictions. Through the code calculation under various flow conditions, it is concluded that the flow is to be fully developed flow over about 40∼80 inches of the entry length. If the exit length is about 30∼40 inches, the effect of the exit pressure can be negligible. The entry and exit length in this paper is calculated based on only rod bundles and simple supports. By installing the flow distributors or strainer, these lengths can get shorter and the flow difference between the subchannels become smaller. This study could be very useful in order to confirm the validation of the boundary conditions used in the subchannel analysis code.

  17. Flow mechanism and heat transfer enhancement in longitudinal-flow tube bundle of shell-and-tube heat exchanger

    Institute of Scientific and Technical Information of China (English)

    2009-01-01

    The flow disturbance and heat transfer mechanism in the tube bundle of rod baffle shell-and-tube heat exchanger were analyzed, on the basis of which and combined with the concept of heat transfer enhancement in the core flow, a new type of shell-and-tube heat exchanger with combination of rod and van type spoiler was designed. Corresponding mathematical and physical models on the shell side about the new type heat exchanger were established, and fluid flow and heat transfer characteristics were numerically analyzed. The simulation results showed that heat transfer coefficient of the new type of heat exchanger approximated to that of rod baffle heat exchanger, but flow pressure drop was much less than the latter, indicating that comprehensive performance of the former is superior to that of the latter. Compared with rod baffle heat exchanger, heat transfer coefficient of the heat exchanger under investigation is higher under same pressure drop, especially under the high Reynolds numbers.

  18. Cuisenaire Rods Go to College.

    Science.gov (United States)

    Chinn, Phyllis; And Others

    1992-01-01

    Presents examples of questions and answers arising from a hands-on and exploratory approach to discrete mathematics using cuisenaire rods. Combinatorial questions about trains formed of cuisenaire rods provide the setting for discovering numerical patterns by experimentation and organizing the results using induction and successive differences.…

  19. General frame structures on quantum principal bundles

    CERN Document Server

    Durdevic, M

    1996-01-01

    A noncommutative-geometric generalization of the classical formalism of frame bundles is developed, incorporating into the theory of quantum principal bundles the concept of the Levi-Civita connection. The construction of a natural differential calculus on quantum principal frame bundles is presented, including the construction of the associated differential calculus on the structure group. General torsion operators are defined and analyzed. Illustrative examples are presented.

  20. ACM Bundles on Del Pezzo surfaces

    Directory of Open Access Journals (Sweden)

    Joan Pons-Llopis

    2009-11-01

    Full Text Available ACM rank 1 bundles on del Pezzo surfaces are classified in terms of the rational normal curves that they contain. A complete list of ACM line bundles is provided. Moreover, for any del Pezzo surface X of degree less or equal than six and for any n ≥ 2 we construct a family of dimension ≥ n − 1 of non-isomorphic simple ACM bundles of rank n on X.

  1. Rod internal pressure quantification and distribution analysis using Frapcon

    Energy Technology Data Exchange (ETDEWEB)

    Jessee, Matthew Anderson [ORNL; Wieselquist, William A [ORNL; Ivanov, Kostadin [Pennsylvania State University, University Park

    2015-09-01

    This report documents work performed supporting the Department of Energy (DOE) Office of Nuclear Energy (NE) Fuel Cycle Technologies Used Fuel Disposition Campaign (UFDC) under work breakdown structure element 1.02.08.10, ST Analysis. In particular, this report fulfills the M4 milestone M4FT- 15OR0810036, Quantify effects of power uncertainty on fuel assembly characteristics, within work package FT-15OR081003 ST Analysis-ORNL. This research was also supported by the Consortium for Advanced Simulation of Light Water Reactors (http://www.casl.gov), an Energy Innovation Hub (http://www.energy.gov/hubs) for Modeling and Simulation of Nuclear Reactors under U.S. Department of Energy Contract No. DE-AC05-00OR22725. The discharge rod internal pressure (RIP) and cladding hoop stress (CHS) distributions are quantified for Watts Bar Nuclear Unit 1 (WBN1) fuel rods by modeling core cycle design data, operation data (including modeling significant trips and downpowers), and as-built fuel enrichments and densities of each fuel rod in FRAPCON-3.5. A methodology is developed which tracks inter-cycle assembly movements and assembly batch fabrication information to build individual FRAPCON inputs for each evaluated WBN1 fuel rod. An alternate model for the amount of helium released from the zirconium diboride (ZrB2) integral fuel burnable absorber (IFBA) layer is derived and applied to FRAPCON output data to quantify the RIP and CHS for these types of fuel rods. SCALE/Polaris is used to quantify fuel rodspecific spectral quantities and the amount of gaseous fission products produced in the fuel for use in FRAPCON inputs. Fuel rods with ZrB2 IFBA layers (i.e., IFBA rods) are determined to have RIP predictions that are elevated when compared to fuel rod without IFBA layers (i.e., standard rods) despite the fact that IFBA rods often have reduced fill pressures and annular fuel pellets. The primary contributor to elevated RIP predictions at burnups less than and greater than 30 GWd

  2. Entropy for frame bundle systems and Grassmann bundle systems induced by a diffeomorphism

    Institute of Scientific and Technical Information of China (English)

    SUN; Weniang(孙文祥)

    2002-01-01

    ALiao hyperbolic diffeomorphism has equal measure entropy and topological entropy to that ofits induced systems on frame bundles and Grassmann bundles. This solves a problem Liao posed in 1996 forLiao hyperbolic diffeomorphisms.

  3. Principal $G$-bundles over elliptic curves

    CERN Document Server

    Friedman, R; Witten, Edward; Friedman, Robert; Morgan, John W.; Witten, Edward

    1997-01-01

    Let $G$ be a simple and simply connected complex Lie group. We discuss the moduli space of holomorphic semistable principal $G$-bundles over an elliptic curve $E$. In particular, we give a new proof of a theorem of Looijenga and Bernshtein-Shvartsman, that the moduli space is a weighted projective space. The method of proof is to study the deformations of certain unstable bundles coming from special maximal parabolic subgroups of $G$. We also discuss the associated automorphism sheaves and universal bundles, as well as the relation between various universal bundles and spectral covers.

  4. Statistical Constitutive Equation of Aramid Fiber Bundles

    Institute of Scientific and Technical Information of China (English)

    熊杰; 顾伯洪; 王善元

    2003-01-01

    Tensile impact tests of aramid (Twaron) fiber bundles were carried om under high strain rates with a wide range of 0. 01/s~1000/s by using MTS and bar-bar tensile impact apparatus. Based on the statistical constitutive model of fiber bundles, statistical constitutive equations of aramid fiber bundles are derived from statistical analysis of test data at different strain rates. Comparison between the theoretical predictions and experimental data indicates statistical constitutive equations fit well with the experimental data, and statistical constitutive equations of fiber bundles at different strain rates are valid.

  5. Jacobi Structures on Affine Bundles

    Institute of Scientific and Technical Information of China (English)

    J. GRABOWSKI; D. IGLESIAS; J. C. MARRERO; E. PADR(O)N; P. URBA(N)SKI

    2007-01-01

    We study affine Jacobi structures (brackets) on an affine bundle π: A→M, i.e. Jacobi brackets that close on affine functions. We prove that if the rank of A is non-zero, there is a one-to- one correspondence between affine Jacobi structures on A and Lie algebroid structures on the vector bundle A+=∪p∈M Aff(Ap, R) of affine functionals. In the case rank A = 0, it is shown that there is a one-to-one correspondence between affins Jacobi structures on A and local Lie algebras on A+. Some examples and applications, also for the linear case, are discussed. For a special type of affine Jacobi structures which are canonically exhibited (strongly-affine or affine-homogeneous Jacobi structures) over a real vector space of finite dimension, we describe the leaves of its characteristic foliation as the orbits of an affine representation. These afline Jacobi structures can be viewed as an analog of the Kostant-Arnold-LiouviUe linear Poisson structure on the dual space of a real finite-dimensional Lie algebra.

  6. A Multi-Dimensional Heat Transfer Model of a Tie-Tube and Hexagonal Fuel Element for Nuclear Thermal Propulsion

    Science.gov (United States)

    Gomez, C. F.; Mireles, O. R.; Stewart, E.

    2016-01-01

    The Space Capable Cryogenic Thermal Engine (SCCTE) effort considers a nuclear thermal rocket design based around a Low-Enriched Uranium (LEU) design fission reactor. The reactor core is comprised of bundled hexagonal fuel elements that directly heat hydrogen for expansion in a thrust chamber and hexagonal tie-tubes that house zirconium hydride moderator mass for the purpose of thermalizing fast neutrons resulting from fission events. Created 3D steady state Hex fuel rod model with 1D flow channels. Hand Calculation were used to set up initial conditions for fluid flow. The Hex Fuel rod uses 1D flow paths to model the channels using empirical correlations for heat transfer in a pipe. Created a 2-D axisymmetric transient to steady state model using the CFD turbulent flow and Heat Transfer module in COMSOL. This model was developed to find and understand the hydrogen flow that might effect the thermal gradients axially and at the end of the tie tube where the flow turns and enters an annulus. The Hex fuel rod and Tie tube models were made based on requirements given to us by CSNR and the SCCTE team. The models helped simplify and understand the physics and assumptions. Using pipe correlations reduced the complexity of the 3-D fuel rod model and is numerically more stable and computationally more time-efficient compared to the CFD approach. The 2-D axisymmetric tie tube model can be used as a reference "Virtual test model" for comparing and improving 3-D Models.

  7. Literature search on Light Water Reactor (LWR) fuel and absorber rod fabrication, 1960--1976

    Energy Technology Data Exchange (ETDEWEB)

    Sample, C R [comp.

    1977-02-01

    A literature search was conducted to provide information supporting the design of a conceptual Light Water Reactor (LWR) Fuel Fabrication plant. Emphasis was placed on fuel processing and pin bundle fabrication, effects of fuel impurities and microstructure on performance and densification, quality assurance, absorber and poison rod fabrication, and fuel pin welding. All data have been taken from publicly available documents, journals, and books. This work was sponsored by the Finishing Processes-Mixed Oxide (MOX) Fuel Fabrication Studies program at HEDL.

  8. Eulerian formulation of elastic rods

    Science.gov (United States)

    Huynen, Alexandre; Detournay, Emmanuel; Denoël, Vincent

    2016-06-01

    In numerous biological, medical and engineering applications, elastic rods are constrained to deform inside or around tube-like surfaces. To solve efficiently this class of problems, the equations governing the deflection of elastic rods are reformulated within the Eulerian framework of this generic tubular constraint defined as a perfectly stiff normal ringed surface. This reformulation hinges on describing the rod-deformed configuration by means of its relative position with respect to a reference curve, defined as the axis or spine curve of the constraint, and on restating the rod local equilibrium in terms of the curvilinear coordinate parametrizing this curve. Associated with a segmentation strategy, which partitions the global problem into a sequence of rod segments either in continuous contact with the constraint or free of contact (except for their extremities), this re-parametrization not only trivializes the detection of new contacts but also transforms these free boundary problems into classic two-points boundary-value problems and suppresses the isoperimetric constraints resulting from the imposition of the rod position at the extremities of each rod segment.

  9. Status of rod consolidation, 1988

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, W.J.

    1989-01-01

    It is estimated that the spent fuel storage pools at some domestic light-water reactors will run out of space before 2003, the year that the US Department of Energy currently predicts it will have a repository available. Of the methods being studied to alleviate the problem, rod consolidation is one of the leading candidates for achieving more efficient use of existing space in spent fuel storage pools. Rod consolidation involves mechanically removing all the fuel rods from the fuel assembly hardware (i.e., the structural components) and placing the fuel rods in a close-packed array in a canister without space grids. A typical goal of rod consolidation systems is to insert the fuel rods from two fuel assemblies into a canister that has the same exterior dimensions as one standard fuel assembly (i.e., to achieve a consolidation or compaction ratio of 2:1) and to compact the nonfuel-bearing structural components from those two fuel assemblies by a factor of 10 to 20. This report provides an overview of the current status of rod consolidation in the United States and a small amount of information on related activities in other countries. 85 refs., 36 figs., 5 tabs.

  10. Principal Bundles on the Projective Line

    Indian Academy of Sciences (India)

    V B Mehta; S Subramanian

    2002-08-01

    We classify principal -bundles on the projective line over an arbitrary field of characteristic ≠ 2 or 3, where is a reductive group. If such a bundle is trivial at a -rational point, then the structure group can be reduced to a maximal torus.

  11. Anatomic Double-bundle ACL Reconstruction

    NARCIS (Netherlands)

    V.M. Schreiber; C.F. van Eck; F.H. Fu

    2010-01-01

    Rupture of the anterior cruciate ligament (ACL) is one of the most frequent forms of knee trauma. The traditional surgical treatment for ACL rupture is single-bundle reconstruction. However, during the past few years there has been a shift in interest toward double-bundle reconstruction to closely r

  12. The Verlinde formula for Higgs bundles

    CERN Document Server

    Andersen, Jørgen Ellegaard; Pei, Du

    2016-01-01

    We propose and prove the Verlinde formula for the quantization of the Higgs bundle moduli spaces and stacks for any simple and simply-connected group. This generalizes the equivariant Verlinde formula for the case of $SU(n)$ proposed previously by the second and third author. We further establish a Verlinde formula for the quantization of parabolic Higgs bundle moduli spaces and stacks.

  13. Line bundle embeddings for heterotic theories

    Energy Technology Data Exchange (ETDEWEB)

    Groot Nibbelink, Stefan [Muenchen Univ. (Germany). Arnold Sommerfeld Center for Theoretical Physics; Ruehle, Fabian [Deutsches Elektronen-Synchrotron (DESY), Hamburg (Germany)

    2016-03-15

    In heterotic string theories consistency requires the introduction of a non-trivial vector bundle. This bundle breaks the original ten-dimensional gauge groups E{sub 8} x E{sub 8} or SO(32) for the supersymmetric heterotic string theories and SO(16) x SO(16) for the non-supersymmetric tachyon-free theory to smaller subgroups. A vast number of MSSM-like models have been constructed up to now, most of which describe the vector bundle as a sum of line bundles. However, there are several different ways of describing these line bundles and their embedding in the ten-dimensional gauge group. We recall and extend these different descriptions and explain how they can be translated into each other.

  14. Requirements for disordered actomyosin bundle contractility

    CERN Document Server

    Lenz, Martin

    2011-01-01

    Actomyosin contractility is essential for biological force generation, and is well understood in highly ordered structures such as striated muscle. In vitro experiments have shown that non-sarcomeric bundles comprised only of F-actin and myosin thick filaments can also display contractile behavior, which cannot be described by standard muscle models. Here we investigate the microscopic symmetries underlying this process in large non-sarcomeric bundles with long actin filaments. We prove that contractile behavior requires non-identical motors that generate large enough forces to probe the nonlinear elastic behavior of F-actin. A simple disordered bundle model demonstrates a contraction mechanism based on these assumptions and predicts realistic bundle deformations. Recent experimental observations of F-actin buckling in in vitro contractile bundles support our model.

  15. Line bundle embeddings for heterotic theories

    CERN Document Server

    Nibbelink, Stefan Groot

    2016-01-01

    In heterotic theories consistency requires the introduction of a non-trivial vector bundle. This bundle breaks the original ten-dimensional gauge groups E_8 x E_8 or SO(32) for the supersymmetric heterotic theories and SO(16) x SO(16) for the non-supersymmetric tachyon-free theory to smaller subgroups. A vast number of MSSM-like models have been constructed up to now, most of which describe the vector bundle as a sum of line bundles. However, there are several different ways of describing these line bundles and their embedding in the ten-dimensional gauge group. We recall and extend these different descriptions and explain how they can be translated into each other.

  16. Line bundle embeddings for heterotic theories

    Science.gov (United States)

    Nibbelin, Stefan Groot; Ruehle, Fabian

    2016-04-01

    In heterotic string theories consistency requires the introduction of a non-trivial vector bundle. This bundle breaks the original ten-dimensional gauge groups E8 × E8 or SO(32) for the supersymmetric heterotic string theories and SO(16) × SO(16) for the non-supersymmetric tachyon-free theory to smaller subgroups. A vast number of MSSM-like models have been constructed up to now, most of which describe the vector bundle as a sum of line bundles. However, there are several different ways of describing these line bundles and their embedding in the ten-dimensional gauge group. We recall and extend these different descriptions and explain how they can be translated into each other.

  17. Composite spinor bundles in gravitation theory

    CERN Document Server

    Sardanashvily, G

    1995-01-01

    In gravitation theory, the realistic fermion matter is described by spinor bundles associated with the cotangent bundle of a world manifold X. In this case, the Dirac operator can be introduced. There is the 1:1 correspondence between these spinor bundles and the tetrad gravitational fields represented by sections of the quotient \\Si of the linear frame bundle over X by the Lorentz group. The key point lies in the fact that different tetrad fields imply nonequivalent representations of cotangent vectors to X by the Dirac's matrices. It follows that a fermion field must be regarded only in a pair with a certain tetrad field. These pairs can be represented by sections of the composite spinor bundle S\\to\\Si\\to X where values of tetrad fields play the role of parameter coordinates, besides the familiar world coordinates.

  18. Double Fell bundles and Spectral triples

    CERN Document Server

    Martins, Rachel A D

    2007-01-01

    As a natural and canonical extension of Kumjian's Fell bundles over groupoids \\cite{fbg}, we give a definition for a double Fell bundle (a double category) over a double groupoid. We show that finite dimensional double category Fell line bundles tensored with their dual with $S^o$-reality satisfy the finite real spectral triples axioms but not necessarily orientability. This means that these product bundles with noncommutative algebras can be regarded as noncommutative compact manifolds more general than real spectral triples as they are not necessarily orientable. By construction, they unify the noncommutative geometry axioms and hence provide an algebraic enveloping structure for finite spectral triples to give the Dirac operator $D$ new algebraic and geometric structures that are otherwise missing in the transition from Fredholm operator to Dirac operator. The Dirac operator in physical applications as a result becomes less ad hoc. The new noncommutative space we present is a complex line bundle over a dou...

  19. On Harder–Narasimhan Reductions for Higgs Principal Bundles

    Indian Academy of Sciences (India)

    Arijit Dey; R Parthasarathi

    2005-05-01

    The existence and uniqueness of – reduction for the Higgs principal bundles over nonsingular projective variety is shown. We also extend the notion of – reduction for (, )-bundles and ramified -bundles over a smooth curve.

  20. Functional bundles of the medial patellofemoral ligament.

    Science.gov (United States)

    Kang, Hui Jun; Wang, Fei; Chen, Bai Cheng; Su, Yan Ling; Zhang, Zhan Chi; Yan, Chang Bao

    2010-11-01

    The purpose of this study was to explore the anatomy and evaluate the function of the medial patellofemoral ligament (MPFL). Anatomical dissection was performed on 12 fresh-frozen knee specimens. The MPFL is a condensation of capsular fibers, which originates at the medial femoral condyle. It runs transversely and inserts to the medial edge of the patella. With the landmark of the medial femur epicondyle (MFE), the femoral origination was located: just 8.90 ± 3.27 mm proximally and 13.47 ± 3.68 mm posteriorly to the MFE. The most interesting finding in present study was functional bundles of its patellar insertion. Approximately from the femoral origination point, fibers of the MPFL form two relatively concentrated fiber bundles: the inferior-straight bundle and the superior-oblique bundle. The whole length of each was 71.78 ± 5.51 and 73.67 ± 5.40 mm, respectively. The included angle between bundles was 15.1° ± 2.1°. Although the superior-oblique bundle and the inferior-straight bundle run on the patellar MPFL inferiorly and superiorly, respectively, as their name indicates, the two bundles are not entirely separated, which make MPFL one intact structure. The inferior-straight bundle is the main static soft tissue restraints where the superior-oblique bundle associated with vastus medialis obliquus (VMO) is to serve as the main dynamic soft tissue restraints. So this finding may provide the theoretical foundation for the anatomical reconstruction of the MPFL and shed lights on the future researchers.

  1. Topological mixing with ghost rods

    Science.gov (United States)

    Gouillart, Emmanuelle; Thiffeault, Jean-Luc; Finn, Matthew D.

    2006-03-01

    Topological chaos relies on the periodic motion of obstacles in a two-dimensional flow in order to form nontrivial braids. This motion generates exponential stretching of material lines, and hence efficient mixing. Boyland, Aref, and Stremler [J. Fluid Mech. 403, 277 (2000)] have studied a specific periodic motion of rods that exhibits topological chaos in a viscous fluid. We show that it is possible to extend their work to cases where the motion of the stirring rods is topologically trivial by considering the dynamics of special periodic points that we call “ghost rods”, because they play a similar role to stirring rods. The ghost rods framework provides a new technique for quantifying chaos and gives insight into the mechanisms that produce chaos and mixing. Numerical simulations for Stokes flow support our results.

  2. Manufacturing of 37-element fuel bundles for PHWR 540 - new approach

    Energy Technology Data Exchange (ETDEWEB)

    Arora, U.K.; Sastry, V.S.; Banerjee, P.K.; Rao, G.V.S.H.; Jayaraj, R.N. [Nuclear Fuel Complex, Dept. Atomic Energy, Government of India, Hyderabad (India)

    2003-07-01

    Nuclear Fuel Complex (NFC), established in early seventies, is a major industrial unit of Department of Atomic Energy. NFC is responsible for the supply of fuel bundles to all the 220 MWe PHWRs presently in operation. For supplying fuel bundles for the forthcoming 540 MWe PHWRs, NEC is dovetailing 37-element fuel bundle manufacturing facilities in the existing plants. In tune with the philosophy of self-reliance, emphasis is given to technology upgradation, higher customer satisfaction and application of modern quality control techniques. With the experience gained over the years in manufacturing 19-element fuel bundles, NEC has introduced resistance welding of appendages on fuel tubes prior to loading of UO{sub 2} pellets, use of bio-degradable cleaning agents, simple diagnostic tools for checking the equipment condition, on line monitoring of variables, built-in process control methods and total productive maintenance concepts in the new manufacturing facility. Simple material handling systems have been contemplated for handling of the fuel bundles. This paper highlights the flow-sheet adopted for the process, design features of critical equipment and the methodology for fabricating the 37-element fuel bundles, 'RIGHT FIRST TIME'. (author)

  3. Heat transfer on HLM cooled wire-spaced fuel pin bundle simulator in the NACIE-UP facility

    Energy Technology Data Exchange (ETDEWEB)

    Di Piazza, Ivan, E-mail: ivan.dipiazza@enea.it [Italian National Agency for New Technologies, Energy and Sustainable Economic Development, C.R. ENEA Brasimone, Camugnano (Italy); Angelucci, Morena; Marinari, Ranieri [University of Pisa, Dipartimento di Ingegneria Civile e Industriale, Pisa (Italy); Tarantino, Mariano [Italian National Agency for New Technologies, Energy and Sustainable Economic Development, C.R. ENEA Brasimone, Camugnano (Italy); Forgione, Nicola [University of Pisa, Dipartimento di Ingegneria Civile e Industriale, Pisa (Italy)

    2016-04-15

    Highlights: • Experiments with a wire-wrapped 19-pin fuel bundle cooled by LBE. • Wall and bulk temperature measurements at three axial positions. • Heat transfer and error analysis in the range of low mass flow rates and Péclet number. • Comparison of local and section-averaged Nusselt number with correlations. - Abstract: The NACIE-UP experimental facility at the ENEA Brasimone Research Centre (Italy) allowed to evaluate the heat transfer coefficient of a wire-spaced fuel bundle cooled by lead-bismuth eutectic (LBE). Lead or lead-bismuth eutectic are very attractive as coolants for the GEN-IV fast reactors due to the good thermo-physical properties and the capability to fulfil the GEN-IV goals. Nevertheless, few experimental data on heat transfer with heavy liquid metals (HLM) are available in literature. Furthermore, just a few data can be identified on the specific topic of wire-spaced fuel bundle cooled by HLM. Additional analysis on thermo-fluid dynamic behaviour of the HLM inside the subchannels of a rod bundle is necessary to support the design and safety assessment of GEN. IV/ADS reactors. In this context, a wire-spaced 19-pin fuel bundle was installed inside the NACIE-UP facility. The pin bundle is equipped with 67 thermocouples to monitor temperatures and analyse the heat transfer behaviour in different sub-channels and axial positions. The experimental campaign was part of the SEARCH FP7 EU project to support the development of the MYRRHA irradiation facility (SCK-CEN). Natural and mixed circulation flow regimes were investigated, with subchannel Reynolds number in the range Re = 1000–10,000 and heat flux in the range q″ = 50–500 kW/m{sup 2}. Local Nusselt numbers were calculated for five sub-channels in different ranks at three axial positions. Section-averaged Nusselt number was also defined and calculated. Local Nusselt data showed good consistency with some of the correlation existing in literature for heat transfer in liquid metals

  4. New Catalytic Proportions for Syntheses of SWNT Bundles (Ropes) and Its Characterization

    Institute of Scientific and Technical Information of China (English)

    DAI Tong; DAI Jian-feng

    2006-01-01

    The single-walled carbon nanotube(SWNT) bundles and ropes have been prepared by using the anode arc discharge plasma to evaporate the graphite rods which contain Fe,Co and Ni powders as catalyst in He atmosphere. Many purifying methods are used for the products. It indicates that the synthesis of SWNTs has been greatly affected by the preparation parameters of catalyzer,the buffer gas and its pressure,the arc current intensity,etc. The optimal condition for preparing SWNTs in our case has been proposed. The forming mechanism of the SWNTs bundles and ropes is also studied qualitatively. The evaporated single graphite sheet tends to reduce its active energy.

  5. Three dimensional considerations in thermal-hydraulics of helical cruciform fuel rods for LWR power uprates

    Energy Technology Data Exchange (ETDEWEB)

    Shirvan, Koroush, E-mail: kshirvan@mit.edu; Kazimi, Mujid S.

    2014-04-01

    Highlights: • We benchmarked the 4 × 4 helical cruciform fuel (HCF) bundle pressure drop experimental data with CFD. • We also benchmarked the 4 × 4 HCF mixing experimental data with CFD. • We derived new friction factors for PWR and BWR designs at PWR and BWR operating conditions from CFD. • We showed the importance of modeling the 3D conduction in HCF in steady state and transient conditions. - Abstract: In order to increase the power density of current and new light water reactor designs, the helical cruciform fuel (HCF) rods have been proposed. The HCF rod is equivalent to a thin cylindrical rod, with 4 fuel containing vanes, wrapped around it. The HCF rods increase the surface area to volume ratio of the fuel and enhance the inter-subchannel mixing due to their helical shape. The rods do not need supporting grids, as they are packed to periodically contact their neighbors along the flow direction, enabling a higher power density in the core. The HCF rods were reported to have the potential to uprate existing PWRs by 45% and BWRs by 20%. In order to quantify the mixing behavior of the HCF rods based on their twist pitch, experiments were previously performed at atmospheric pressures with single phase water in a 4 by 4 HCF and cylindrical rod bundles. In this paper, the experimental results on pressure drop and mixing are benchmarked with computational fluid dynamic (CFD) using steady state the Reynolds average Navier–Stokes (RANS) turbulence model. The sensitivity of the CFD approach to computational domain, mesh size, mesh shape and RANS turbulence models are examined against the experimental conditions. Due to the refined radial velocity profile from the HCF rods twist, the turbulence models showed little sensitivity to the domain. Based on the CFD simulations, the total pressure drops under the PWR and BWR conditions are expected to be about 10% higher than the values previously reported solely from an empirical correlation based on the

  6. Changes of surface structure and elemental composition of Pd rod and collector of nuclear reaction products irradiated with 10-MeV γ-quanta in dense deuterium gas

    Science.gov (United States)

    Didyk, A. Yu.; Wiśniewski, R.; Wilczynska-Kitowska, T.

    2013-12-01

    A high-pressure chamber is filled with 1.2 kbar molecular deuterium (DHPC). The palladium rod saturated with deuterium is loaded inside the DHPC and irradiated with 10-MeV bremsstrahlung γ-quanta for 18 h with a 11-13 μA electron beam using the MT-25 electron accelerator. The elemental compositions of all surfaces of the DHPC element inside the dense deuterium gas were studied using scanning electronic microscopes with X-ray microelement microprobe analysis. It is found that all surfaces, including the surface of a high-purity palladium rod (99.995%), are covered with a partly homogeneous layer or large microparticles of lead. Also, light elements such as 6C, 8O, 11Na, 12Mg, 13Al, 14Si, 22Ti, 25Mn, 26Fe, 29Cu, and 30Zn and heavy metals such as 47Ag, 73Ta, 74W, 78Pt, 79Au, and 82Pb are observed. Possible processes that can cause the anomalies observed in the newly formed chemical elements are briefly discussed.

  7. Future plans for performance analysis and maintenance/inspection optimization of shutoff rods based on the case study of Bruce Power Unit-3 Shutoff Rod 5 inspection

    Energy Technology Data Exchange (ETDEWEB)

    Nasimi, E. [Bruce Power LLP, Tiverton, Ontario (Canada); Univ. of Ontario Inst. of Technology, Oshawa, Ontario (Canada); Gabbar, H.A. [Univ. of Ontario Inst. of Technology, Oshawa, Ontario (Canada)

    2011-07-01

    Shutdown System 1 (SDS1) is a preferred method for a quick shutdown of nuclear fission process in CANDU (CANada Deuterium Uranium) reactor units. Failure of a routine SDS1 safety test during Fall 2009 outage resulted in the need to develop and execute a new methodology for Shutoff Rod inspection and re-evaluate the known degradation mechanisms and failure modes. This paper describes the development of this methodology and the obtained results. It also proposes several alternative solutions for the future performance analysis and maintenance/inspection optimization for SDS1 Shutoff Rods based on the Bruce Power Unit-3 Shutoff Rod 5 case study. (author)

  8. Control rod worth calculations using deterministic and stochastic methods

    Energy Technology Data Exchange (ETDEWEB)

    Varvayanni, M. [NCSR ' DEMOKRITOS' , PO Box 60228, 15310 Aghia Paraskevi (Greece); Savva, P., E-mail: melina@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PO Box 60228, 15310 Aghia Paraskevi (Greece); Catsaros, N. [NCSR ' DEMOKRITOS' , PO Box 60228, 15310 Aghia Paraskevi (Greece)

    2009-11-15

    Knowledge of the efficiency of a control rod to absorb excess reactivity in a nuclear reactor, i.e. knowledge of its reactivity worth, is very important from many points of view. These include the analysis and the assessment of the shutdown margin of new core configurations (upgrade, conversion, refuelling, etc.) as well as several operational needs, such as calibration of the control rods, e.g. in case that reactivity insertion experiments are planned. The control rod worth can be assessed either experimentally or theoretically, mainly through the utilization of neutronic codes. In the present work two different theoretical approaches, i.e. a deterministic and a stochastic one are used for the estimation of the integral and the differential worth of two control rods utilized in the Greek Research Reactor (GRR-1). For the deterministic approach the neutronics code system SCALE (modules NITAWL/XSDRNPM) and CITATION is used, while the stochastic one is made using the Monte Carlo code TRIPOLI. Both approaches follow the procedure of reactivity insertion steps and their results are tested against measurements conducted in the reactor. The goal of this work is to examine the capability of a deterministic code system to reliably simulate the worth of a control rod, based also on comparisons with the detailed Monte Carlo simulation, while various options are tested with respect to the deterministic results' reliability.

  9. Higgs bundles and the real symplectic group

    CERN Document Server

    Gothen, Peter B

    2011-01-01

    We give an overview of the work of Corlette, Donaldson, Hitchin and Simpson leading to the non-abelian Hodge theory correspondence between representations of the fundamental group of a surface and the moduli space of Higgs bundles. We then explain how this can be generalized to a correspondence between character varieties for representations of surface groups in real Lie groups G and the moduli space of G-Higgs bundles. Finally we survey recent joint work with Bradlow, Garc\\'ia-Prada and Mundet i Riera on the moduli space of maximal Sp(2n,R)-Higgs bundles.

  10. Rolls-Royce digital Rod Control System

    Energy Technology Data Exchange (ETDEWEB)

    Pouillot, M. [Rolls-Royce Civil Nuclear SAS (France)

    2010-07-01

    -graph representing min-max monitoring thresholds, and grouped faults. Plugs allow output current checking per coil. - From the cycler, the output current can be forced on a rack for test, while ensuring that rods are held by another available gripper rack. - Low power dissipation: no need for multiplexing or LC-inhibition thyristors - Commissioning: 3-thyristor well-tried half bridges (no energy storage), fuses for protection of feed-through penetrations. The paper will present in details the technical features of the Rolls-Royce Rod Control System and the main benefits to use it. About Rolls-Royce Rolls-Royce is a global business providing integrated power systems for use on land, at sea and in the air. The Group has a balanced business portfolio with leading market positions. Rolls-Royce has a broad range of civil nuclear expertise, including work related to licensing and safety reviews, engineering design, supply chain management, manufacturing, installation and commissioning of the nuclear island systems and equipments, as well as operational management and through life support. Rolls-Royce benefits of a strong reputation worldwide and of a unique experience in the nuclear market. (author)

  11. Results from studies of surface deposits on the claddings of fuel rods used in RBMK-1000 reactors

    Science.gov (United States)

    Smirnova, I. M.; Markov, D. V.

    2010-07-01

    The results of studies on analyzing the element composition of deposits on the cladding surfaces of fuel rods used in a fuel assembly at the Leningrad nuclear power station are presented. The distribution of elements in deposits over the fuel rod height is analyzed, and the zones of their concentration are revealed. It is shown that deposits of copper penetrating into cracks in the surface layer of zirconium oxide introduce an essential contribution in the development of nodular corrosion of fuel rod claddings.

  12. Subchannel void-fraction measurements in a 6 by 6 rod tube bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kok, H.V.; van der Hagen, T.H.J.J.; Adams, B.T. [Interfaculty Reactor Inst., Delf Univ. of Technology, Delft (Netherlands); Mudde, R.F.

    1997-12-31

    Using gamma-absorption and tomographic reconstruction techniques the void-fraction in each subchannel of a 6 by 6 scaled BWR fuel assembly could be measured at different axial positions along the assembly. The measurements were performed on the DESIRE facility at the Interfaculty Reactor Institute, Delft. The DESIRE facility is a scaled natural circulation loop that uses Freon-12 as a coolant. The fuel assembly is scaled for correct representation of the void-fraction and flow patterns, except at the bubbly flow regime. The scaling has been verified using the MONA code. A clear transition from bubbly to annular flow was observed in the experiments. Experiments using a tilted power profile show that there is no significant lateral transport of vapour across subchannels. (author)

  13. The Third ATLAS ROD Workshop

    CERN Multimedia

    Poggioli, L.

    A new-style Workshop After two successful ATLAS ROD Workshops dedicated to the ROD hardware and held at the Geneva University in 1998 and in 2000, a new style Workshop took place at LAPP in Annecy on November 14-15, 2002. This time the Workshop was fully dedicated to the ROD-TDAQ integration and software in view of the near future integration activities of the final RODs for the detector assembly and commissioning. More precisely, the aim of this workshop was to get from the sub-detectors the parameters needed for T-DAQ, as well as status and plans from ROD builders. On the other hand, what was decided and assumed had to be stated (like EB decisions and URDs), and also support plans. The Workshop gathered about 70 participants from all ATLAS sub-detectors and the T-DAQ community. The quite dense agenda allowed nevertheless for many lively discussions, and for a dinner in the old town of Annecy. The Sessions The Workshop was organized in five main sessions: Assumptions and recommendations Sub-de...

  14. Mobility of Taxol in Microtubule Bundles

    Science.gov (United States)

    Ross, J.

    2003-06-01

    Mobility of taxol inside microtubules was investigated using fluorescence recovery after photobleaching (FRAP) on flow-aligned bundles. Bundles were made of microtubules with either GMPCPP or GTP at the exchangeable site on the tubulin dimer. Recovery times were sensitive to bundle thickness and packing, indicating that taxol molecules are able to move laterally through the bundle. The density of open binding sites along a microtubule was varied by controlling the concentration of taxol in solution for GMPCPP samples. With > 63% sites occupied, recovery times were independent of taxol concentration and, therefore, inversely proportional to the microscopic dissociation rate, k_{off}. It was found that 10*k_{off} (GMPCPP) ~ k_{off} (GTP), consistent with, but not fully accounting for, the difference in equilibrium constants for taxol on GMPCPP and GTP microtubules. With taxol along the microtubule interior is hindered by rebinding events when open sites are within ~7 nm of each other.

  15. Noncommutative principal bundles through twist deformation

    CERN Document Server

    Aschieri, Paolo; Pagani, Chiara; Schenkel, Alexander

    2016-01-01

    We construct noncommutative principal bundles deforming principal bundles with a Drinfeld twist (2-cocycle). If the twist is associated with the structure group then we have a deformation of the fibers. If the twist is associated with the automorphism group of the principal bundle, then we obtain noncommutative deformations of the base space as well. Combining the two twist deformations we obtain noncommutative principal bundles with both noncommutative fibers and base space. More in general, the natural isomorphisms proving the equivalence of a closed monoidal category of modules and its twist related one are used to obtain new Hopf-Galois extensions as twists of Hopf-Galois extensions. A sheaf approach is also considered, and examples presented.

  16. Design requirements of ACR-1000 fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Gossain, D.; Reid, P. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)

    2008-07-01

    The design process for ACR-1000 fuel bundle is being undertaken in accordance with the CSA standard N286.2. As an element of the process, the design requirements were established early in the design phase and compiled in the ACR-1000 Fuel Design Requirements (DR) document. The ACR-1000 fuel bundle design is being developed to meet these requirements. This paper discusses the sources for the requirements such as the ACR project requirements, the plant specifications and regulatory requirements. It also discusses considerations of reactor design decisions and operational decisions in establishing functional, performance, safety and other design requirements for the fuel bundle. The design requirements for the ACR-1000 fuel bundle are summarized and the relationship of the requirements to the plant states of Normal Operation, Anticipated Operational Occurrences (AOOs) and Design Basis Accidents (DBAs) are discussed. Structure of the document to capture all the requirements in addition to functional, performance and safety requirements is presented. (author)

  17. Bundled Hybrid Offset Riser Global Strength Analysis

    Institute of Scientific and Technical Information of China (English)

    William C.Webster; Zhuang Kang; Wenzhou Liang; Youwei Kang; Liping Sun

    2011-01-01

    Bundled hybrid offset riser(BHOR)global strength analysis,which is more complex than single line offset riser global strength analysis,was carried out in this paper.At first,the equivalent theory is used to deal with BHOR,and then its global strength in manifold cases was analyzed,along with the use of a three-dimensional nonlinear time domain finite element program.So the max bending stress,max circumferential stress,and max axial stress in the BHOR bundle main section(BMS)were obtained,and the values of these three stresses in each riser were obtained through the "stress distribution method".Finally,the Max Von Mises stress in each riser was given and a check was made whether or not they met the demand.This paper provides a reference for strength analysis of the bundled hybrid offset riser and some other bundled pipelines.

  18. Liquid Flow in Shaped Fiber Bundle

    Institute of Scientific and Technical Information of China (English)

    ZHANG Yan; WANG Hua-ping; CHEN Yue-hua

    2006-01-01

    By computation and comparison of the critical spreading coefficient parameter, it was found that shaped fiber bundle is better for wetting. Liquid-air interface tension of liquid arising the shaped fiber bundle body is considered as one critical factor besides liquid viscosity, inertia force and liquid-fiber interface tension. Experimental result simulation demonstrated that the liquid-air interface tension is correlated with the geometric size of the liquid arising in body, φ0 (x) and which is affected by the cross sectional shape of fiber and the radius of single fiber. The shaped fiber bundle model is introduced to investigate liquid flow in fiber assembly. The model is generated based on a random function for stochastic forming of fibers in bundle and it is necessary to combine this fundamental model with physical explanation for investigation of liquid flow in fiber assembly.

  19. Dynamic bi-product bundle pricing problem

    Directory of Open Access Journals (Sweden)

    Rafiei Hamed

    2014-01-01

    Full Text Available This paper addresses bundle pricing problem of two products in a stochastic environment so as to maximize net profit of a retailer. In the considered problem, it is assumed that customers are received upon a Poisson distribution and their demands follow a bi-variant distribution function. Also, it is assumed that products are sold individually or in the form of a bundle, which are offered from an initial stock of the products. To tackle the problem, a stochastic dynamic program is developed in which optimum values of the initial stock and order quantities of every planning period are determined. Moreover, prices of the individual products and their bundle are optimized. Also, the proposed dynamic program tackles bundling/ unbundling decisions taken in every planning period. A numerical example of a two planning period horizon is considered to validate the proposed model.

  20. Effect of Testing Conditions on Fibre-Bundle Tensile Properties Part Ⅰ: Sample Preparation, Bundle Mass and Fibre Alignment of Wool Bundles

    Institute of Scientific and Technical Information of China (English)

    YU Wei-dong; YAN Hao-jing; Ron Postle; Yang Shouren

    2002-01-01

    Due to the effects of samples and testing conditions on fibre-bundle tensile behaviour, it is necessary to investigate the relationships between experimental factors and tensile properties for the fibre-bumdle tensile tester (TENSOR). The effects of bundle sample preparation, fibre bundle mass and fibre alignment have been tested. The experimental results indicated that (1) the low damage in combing and no free-end fibres in the cut bundle are most important for the sample preparation; (2) the reasonable bundle mass is 400- 700tex, but the tensile properties measured should bemodified with the bundle mass because a small amount of bundle mass causes the scatter results, while the larger is the bundle mass, the more difficult to comb fibres parallel and to clamp fibre evenly; and (3) the fibre irregular arrangement forms a slack bundle resulting in interaction between fibres, which will affect the reproducibility and accuracy of the tensile testing.

  1. Amyloid-β and proinflammatory cytokines utilize a prion protein-dependent pathway to activate NADPH oxidase and induce cofilin-actin rods in hippocampal neurons.

    Directory of Open Access Journals (Sweden)

    Keifer P Walsh

    Full Text Available Neurites of neurons under acute or chronic stress form bundles of filaments (rods containing 1∶1 cofilin∶actin, which impair transport and synaptic function. Rods contain disulfide cross-linked cofilin and are induced by treatments resulting in oxidative stress. Rods form rapidly (5-30 min in >80% of cultured hippocampal or cortical neurons treated with excitotoxic levels of glutamate or energy depleted (hypoxia/ischemia or mitochondrial inhibitors. In contrast, slow rod formation (50% of maximum response in ∼6 h occurs in a subpopulation (∼20% of hippocampal neurons upon exposure to soluble human amyloid-β dimer/trimer (Aβd/t at subnanomolar concentrations. Here we show that proinflammatory cytokines (TNFα, IL-1β, IL-6 also induce rods at the same rate and within the same neuronal population as Aβd/t. Neurons from prion (PrP(C-null mice form rods in response to glutamate or antimycin A, but not in response to proinflammatory cytokines or Aβd/t. Two pathways inducing rod formation were confirmed by demonstrating that NADPH-oxidase (NOX activity is required for prion-dependent rod formation, but not for rods induced by glutamate or energy depletion. Surprisingly, overexpression of PrP(C is by itself sufficient to induce rods in over 40% of hippocampal neurons through the NOX-dependent pathway. Persistence of PrP(C-dependent rods requires the continuous activity of NOX. Removing inducers or inhibiting NOX activity in cells containing PrP(C-dependent rods causes rod disappearance with a half-life of about 36 min. Cofilin-actin rods provide a mechanism for synapse loss bridging the amyloid and cytokine hypotheses for Alzheimer disease, and may explain how functionally diverse Aβ-binding membrane proteins induce synaptic dysfunction.

  2. A Geometric Approach to Noncommutative Principal Bundles

    CERN Document Server

    Wagner, Stefan

    2011-01-01

    From a geometrical point of view it is, so far, not sufficiently well understood what should be a "noncommutative principal bundle". Still, there is a well-developed abstract algebraic approach using the theory of Hopf algebras. An important handicap of this approach is the ignorance of topological and geometrical aspects. The aim of this thesis is to develop a geometrically oriented approach to the noncommutative geometry of principal bundles based on dynamical systems and the representation theory of the corresponding transformation group.

  3. Supporting the Secure Deployment of OSGi Bundles

    OpenAIRE

    Parrend, Pierre; Frénot, Stéphane

    2007-01-01

    International audience; The OSGi platform is a lightweight management layer over a Java virtual machine that makes runtime extensi- bility and multi-application support possible in mobile and constraint environments. This powerfull capability opens a particular attack vector against mobile platforms: the in- stallation of malicious OSGi bundles. The first countermea- sure is the digital signature of the bundles. We developed a tool suite that supports the signature, the publication and the va...

  4. HLM fuel pin bundle experiments in the CIRCE pool facility

    Energy Technology Data Exchange (ETDEWEB)

    Martelli, Daniele, E-mail: daniele.martelli@ing.unipi.it [University of Pisa, Department of Civil and Industrial Engineering, Pisa (Italy); Forgione, Nicola [University of Pisa, Department of Civil and Industrial Engineering, Pisa (Italy); Di Piazza, Ivan; Tarantino, Mariano [Italian National Agency for New Technologies, Energy and Sustainable Economic Development, C.R. ENEA Brasimone (Italy)

    2015-10-15

    Highlights: • The experimental results represent the first set of values for LBE pool facility. • Heat transfer is investigated for a 37-pin electrical bundle cooled by LBE. • Experimental data are presented together with a detailed error analysis. • Nu is computed as a function of the Pe and compared with correlations. • Experimental Nu is about 25% lower than Nu derived from correlations. - Abstract: Since Lead-cooled Fast Reactors (LFR) have been conceptualized in the frame of GEN IV International Forum (GIF), great interest has focused on the development and testing of new technologies related to HLM nuclear reactors. In this frame the Integral Circulation Experiment (ICE) test section has been installed into the CIRCE pool facility and suitable experiments have been carried out aiming to fully investigate the heat transfer phenomena in grid spaced fuel pin bundles providing experimental data in support of European fast reactor development. In particular, the fuel pin bundle simulator (FPS) cooled by lead bismuth eutectic (LBE), has been conceived with a thermal power of about 1 MW and a uniform linear power up to 25 kW/m, relevant values for a LFR. It consists of 37 fuel pins (electrically simulated) placed on a hexagonal lattice with a pitch to diameter ratio of 1.8. The FPS was deeply instrumented by several thermocouples. In particular, two sections of the FPS were instrumented in order to evaluate the heat transfer coefficient along the bundle as well as the cladding temperature in different ranks of sub-channels. Nusselt number in the central sub-channel was therefore calculated as a function of the Peclet number and the obtained results were compared to Nusselt numbers obtained from convective heat transfer correlations available in literature on Heavy Liquid Metals (HLM). Results reported in the present work, represent the first set of experimental data concerning fuel pin bundle behaviour in a heavy liquid metal pool, both in forced and

  5. Is It Complete Left Bundle Branch Block? Just Ablate the Right Bundle.

    Science.gov (United States)

    Ali, Hussam; Lupo, Pierpaolo; Foresti, Sara; De Ambroggi, Guido; Epicoco, Gianluca; Fundaliotis, Angelica; Cappato, Riccardo

    2017-03-01

    Complete left bundle branch block (LBBB) is established according to standard electrocardiographic criteria. However, functional LBBB may be rate-dependent or can perpetuate during tachycardia due to repetitive concealed retrograde penetration of impulses through the contralateral bundle "linking phenomenon." In this brief article, we present two patients with basal complete LBBB in whom ablating the right bundle unmasked the actual antegrade conduction capabilities of the left bundle. These cases highlight intriguing overlap between electrophysiological concepts of complete block, linking, extremely slow, and concealed conduction.

  6. Topological Optimization of Rod Mixers

    Science.gov (United States)

    Finn, Matthew D.; Thiffeault, Jean-Luc

    2006-11-01

    Stirring of fluid with moving rods is necessary in many practical applications to achieve homogeneity. These rods are topological obstacles that force stretching of fluid elements. The resulting stretching and folding is commonly observed as filaments and striations, and is a precursor to mixing. In a space-time diagram, the trajectories of the rods form a braid [1], and the properties of this braid impose a minimal complexity in the flow. We discuss how optimal mixing protocols can be obtained by a judicious choice of braid, and how these protocols can be implemented using simple gearing [2].[12pt] [1] P. L. Boyland, H. Aref, and M. A. Stremler, JFM 403, 277 (2000).[8pt] [2] J.-L. Thiffeault and M. D. Finn, http://arxiv.org/nlin/0603003

  7. Morphological Diversity of the Rod Spherule: A Study of Serially Reconstructed Electron Micrographs.

    Directory of Open Access Journals (Sweden)

    Shuai Li

    spherules lacking an axon (R2 were also described and characterized. This rod spherule group consists of a specific spatial organization that is strictly located at the apical OPL-facing layer of the Outer Nuclear Layer (ONL. The R2 spherule displays a large bowl-shaped synaptic terminal that hugs the rod soma. Retracted spherules in the DscamGoF retina were also reconstructed to test if they are structurally similar to R2 spherules. The misplaced rod spherules in DscamGoF have a gross morphology that is similar to R2 spherules but have significant disruption in internal synapse organization.We described a morphological diversity within Mus musculus rod spherules. This diversity is correlated with rod location in the ONL and contributes to the intracellular differences within spherules. Analysis of the DscamGoF retina indicated that their R2 spherules are not significantly different than wild type R2 spherules, but that their retracted rod spherules have abnormal synaptic organization.

  8. Control rod drop surveillance using two friction coefficients

    OpenAIRE

    Blázquez, Juan; Vallejo, I.; García-Berrocal Sánchez, Agustin; Balbás Antón, Miguel

    2011-01-01

    In the case of large burnup, a control rod (CR) guide tube in the pressurized water reactor of a commercial nuclear power plant might bend. As a consequence, a CR drop experiment may indicate an event of a CR partially inserted and whether the CR should be deemed inoperable. Early prevention of such an event can be achieved by measuring two friction coefficients: the hydraulic coefficient and the sliding coefficient. The hydraulic coefficient hardly changes, so that the curvature of the guide...

  9. Twisted Bundle on Noncommutative Space and U(1) Instanton

    CERN Document Server

    Ho, P M

    2000-01-01

    We study the notion of twisted bundles on noncommutative space. Due to theexistence of projective operators in the algebra of functions on thenoncommutative space, there are twisted bundles with non-constant dimension.The U(1) instanton solution of Nekrasov and Schwarz is such an example. As amathematical motivation for not excluding such bundles, we find gaugetransformations by which a bundle with constant dimension can be equivalent toa bundle with non-constant dimension.

  10. NUCLEAR REACTOR

    Science.gov (United States)

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  11. Control rods in LMFBRs: a physics assessment

    Energy Technology Data Exchange (ETDEWEB)

    McFarlane, H.F.; Collins, P.J.

    1982-08-01

    This physics assessment is based on roughly 300 control rod worth measurements in ZPPR from 1972 to 1981. All ZPPR assemblies simulated mixed-oxide LMFBRs, representing sizes of 350, 700, and 900 MWe. Control rod worth measurements included single rods, various combinations of rods, and Ta and Eu rods. Additional measurements studied variations in B/sub 4/C enrichment, rod interaction effects, variations in rod geometry, neutron streaming in sodium-filled channels, and axial worth profiles. Analyses were done with design-equivalent methods, using ENDF/B Version IV data. Some computations for the sensitivities to approximations in the methods have been included. Comparisons of these analyses with the experiments have allowed the status of control rod physics in the US to be clearly defined.

  12. The histology of retinal nerve fiber layer bundles and bundle defects.

    Science.gov (United States)

    Radius, R L; Anderson, D R

    1979-05-01

    The fiber bundle striations recognized clinically in normal monkey eyes appear to be bundles of axons compartmentalized within glial tunnels formed by Müller's-cell processes, when viewed histologically. The dark boundaries that separate individual bundles are the broadened foot endings of these cells near the inner surface of the retina. Within one week after focal retinal photocoagulation, characteristic fundus changes could be seen in experimental eyes. In histologic sections of the involved retina, there was marked cystic degeneration of the retinal nerve fiber layer. Within one month, atrophy of distal axon segments was complete. With the drop-out of damaged axons and thinning of individual fiber bundles, retinal striations became less prominent. The resulting fundus picture in these experimental eyes is similar to fiber bundle defects that can be seen clinically in various neuro-ophthalmic disorders.

  13. Analysis of Dynamic Insertion of Control Rod of BWR under Seismic Excitation

    Science.gov (United States)

    Koide, Yuichi; Nakagawa, Masaki; Fukushi, Naoki; Ishigaki, Hirokuni; Okumura, Kazue

    The dynamic characteristics of control rod for boiling water reactor being inserted under seismic excitation were investigated using non-linear analytical models. The capability of managing the insertion of control rod is one of the most important factors affecting the safety of nuclear power plant undergoing seismic events. Predicting the behavior of control rod being inserted during earthquakes is important when designing how rod should be controlled during seismic events. We developed analytical models using the finite element method (FEM). The effect of the interaction force between the control rod and the fuel assemblies is considered in non-linear analysis. This interaction force causes resistance force to be applied to the control rod when they are being inserted. The validity of the analytical models was confirmed by comparing the analytical results with the experimental ones. The effects of input seismic motion and structural parameters on the insertion time ware investigated using the analytical models. These analytical methods can be used to predict the time to insert the control rod into the core region of reactor, and are useful for designing control rod system that can survive seismic events.

  14. Solid-state-laser-rod holder

    Science.gov (United States)

    Gettemy, D.J.; Barnes, N.P.; Griggs, J.E.

    1981-08-11

    The disclosure relates to a solid state laser rod holder comprising Invar, copper tubing, and epoxy joints. Materials and coefficients of expansion of the components of the holder combine with the rod to produce a joint which will give before the rod itself will. The rod may be lased at about 70 to 80/sup 0/K and returned from such a temperature to room temperature repeatedly without its or the holder's destruction.

  15. 关于秦山一期棒控系统电流调节回路软件化的设想%The Assumption of the Current Regulating Circuit of Control rod Control System in Qin Shan Nuclear Power Plant Phase I Achieved by Software

    Institute of Scientific and Technical Information of China (English)

    谭平; 黄程

    2016-01-01

    The paper takes the control rod control system of Qin Shan nuclear power phase I as anexample, describe thecontrol rod driven mechanism、the principle of current regulating circuit and control scheme briefly, discuss the function of theset point function module, currentregulating module,phase shift trigger module, anddo the software requirement analysis for them. The paper is based on the C8051F(micro control unit)toachieve the function of these three module,improving the integrity of the current regulating circuit, saving the space of cabinet, conforming to the Miniaturized design.%本文以秦山一期棒控系统为例,简要介绍了秦山一期控制棒控制系统(下述简称棒控系统)的控制对象控制棒驱动机构、电流调节回路原理及控制方案,论述了棒控系统定值、电流调节、移相触发器组件的功能,并对这三类模块功能进行软件分析。本文采用以C8051F单片机为控制核心,将原来采用硬件实现的功能改由软件实现,提高了电流调节回路的集成度,节省机柜空间,符合微型化设计理念。

  16. Evaluation of the thermal-hydraulic response and fuel rod thermal and mechanical deformation behavior during the power burst facility test LOC-3. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Yackle, T.R.; MacDonald, P.E.; Broughton, J.M.

    1980-01-01

    An evaluation of the results from the LOC-3 nuclear blowdown test conducted in the Power Burst Facility is presented. The test objective was to examine fuel and cladding behavior during a postulated cold leg break accident in a pressurized water reactor (PWR). Separate effects of rod internal pressure and the degree of irradiation were investigated in the four-rod test. Extensive cladding deformation (ballooning) and failure occurred during blowdown. The deformation of the low and high pressure rods was similar; however, the previously irradiated test rod deformed to a greater extent than a similar fresh rod exposed to identical system conditions.

  17. 21 CFR 876.4270 - Colostomy rod.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Colostomy rod. 876.4270 Section 876.4270 Food and... GASTROENTEROLOGY-UROLOGY DEVICES Surgical Devices § 876.4270 Colostomy rod. (a) Identification. A colostomy rod is a device used during the loop colostomy procedure. A loop of colon is surgically brought out...

  18. Solitary waves on nonlinear elastic rods. II

    DEFF Research Database (Denmark)

    Sørensen, Mads Peter; Christiansen, Peter Leth; Lomdahl, P. S.

    1987-01-01

    In continuation of an earlier study of propagation of solitary waves on nonlinear elastic rods, numerical investigations of blowup, reflection, and fission at continuous and discontinuous variation of the cross section for the rod and reflection at the end of the rod are presented. The results...

  19. Phase behavior of colloidal silica rods

    NARCIS (Netherlands)

    Kuijk, A.; Byelov, D.; Petukhov, A.V.; van Blaaderen, A.; Imhof, A.

    2012-01-01

    Recently, a novel colloidal hard-rod-like model system was developed which consists of silica rods [Kuijk et al., JACS, 2011, 133, 2346]. Here, we present a study of the phase behavior of these rods, for aspect ratios ranging from 3.7 to 8.0. By combining real-space confocal laser scanning microscop

  20. Finite-element procedure for calculating the three-dimensional inelastic bowing of fuel rods (AWBA development program)

    Energy Technology Data Exchange (ETDEWEB)

    Martin, S E

    1982-05-01

    An incremental finite element procedure is developed for calculating the in-pile lateral bowing of nuclear fuel rods. The fuel rod is modeled as a viscoelastic beam whose material properties are derived as perturbations of the results of an axisymmetric stress analysis of the fuel rod. The effects which are taken into account in calculating the rod's lateral bowing include: (a) lateral, axial, and rotational motions and forces at the rod supports, (b) transverse gradients of temperature, fast-neutron flux, and fissioning rate, and (c) cladding circumferential wall thickness variation. The procedure developed in this report could be used to form the basis for a computer program to calculate the time-dependent bowing as a function of the fuel rod's operational and environmental history.

  1. Stress Analysis of Single Spacer Grid Support considering Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Y. G.; Jung, D. H.; Kim, J. H. [Chungnam National University, Daejeon (Korea, Republic of); Park, J. K.; Jeon, K. L. [Korea Nuclear Fuel, Daejeon (Korea, Republic of)

    2010-10-15

    Pressurized water reactor (PWR) nuclear fuel assembly is mainly composed of a top-end piece, a bottom-end piece, lots of fuel rods, and several spacer grids. Among them, the main function of spacer grid is protecting fuel rods from Fluid Induced Vibration (FIV). The cross section of spacer grid assembled by laser welding in upper and lower point. When the fuel rod inserted in spacer gird, spring and dimple and around of welded area got a stresses. The main hypothesis of this analysis is the boundary area of HAZ and base metal can get a lot of damage than other area by FIV. So, design factors of spacer grid mainly considered to preventing the fatigue failure in HAZ and spring and dimple of spacer grid. From previous researching, the environment in reactor verified. Pressure and temperature of light water observed 15MPa and 320 .deg. C, and vibration of the fuel rod observed within 0 {approx} 50Hz. In this study, mechanical properties of zirconium alloy that extracted from the test and the spacer grid model which used in the PWR were applied in stress analyzing. General-purpose finite element analysis program was used ANSYS Workbench 12.0.1 version. 3-D CAD program CATIA was used to create spacer grid model

  2. Coexistence of central nucleus, cores, and rods: Diagnostic relevance

    Science.gov (United States)

    Dhinakaran, Sathiyabama; Kumar, Rashmi Santhosh; Thakkar, Ravindra; Narayanappa, Gayathri

    2016-01-01

    Background: Congenital myopathies (CMs) though considered distinct disorders, simultaneous occurrence of central nucleus, nemaline rods, and cores in the same biopsy are scarcely reported. Objective: A retrospective reassessment of cases diagnosed as CMs to look for multiple pathologies missed, if any, during the initial diagnosis. Materials and Methods: Enzyme histochemical, and immunohistochemical-stained slides from 125 cases diagnosed as congenital myopathy were reassessed. Results: The study revealed 15 cases (12%) of congenital myopathy with more than one morphological feature. Central nucleus with cores (n = 11), central nucleus, nemaline rods and cores (n = 3), and nemaline rods with cores (n = 1). 4/11 cases were diagnosed as centronuclear myopathy (CNM) in the first instance; in addition, cores were revealed on reassessment. Discussion: The prevalence of CMs of all neuromuscular disorders is approximately 6 in 100,000 live births, with regional variations. Three main defined CMs include centro nuclear myopathy (CNM), nemaline rod myopathy (NRM), and central core disease (CCD). However, they are more diverse with overlapping clinical and histopathological features, thus broadening the spectra within each category of congenital myopathy. Conclusion: Identification of cases with overlap of pathological features has diagnostic relevance. PMID:27293330

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

    Energy Technology Data Exchange (ETDEWEB)

    Sease, J.D.

    1998-03-01

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

  4. Coexistence of central nucleus, cores, and rods: Diagnostic relevance

    Directory of Open Access Journals (Sweden)

    Sathiyabama Dhinakaran

    2016-01-01

    Full Text Available Background: Congenital myopathies (CMs though considered distinct disorders, simultaneous occurrence of central nucleus, nemaline rods, and cores in the same biopsy are scarcely reported. Objective: A retrospective reassessment of cases diagnosed as CMs to look for multiple pathologies missed, if any, during the initial diagnosis. Materials and Methods: Enzyme histochemical, and immunohistochemical-stained slides from 125 cases diagnosed as congenital myopathy were reassessed. Results: The study revealed 15 cases (12% of congenital myopathy with more than one morphological feature. Central nucleus with cores (n = 11, central nucleus, nemaline rods and cores (n = 3, and nemaline rods with cores (n = 1. 4/11 cases were diagnosed as centronuclear myopathy (CNM in the first instance; in addition, cores were revealed on reassessment. Discussion: The prevalence of CMs of all neuromuscular disorders is approximately 6 in 100,000 live births, with regional variations. Three main defined CMs include centro nuclear myopathy (CNM, nemaline rod myopathy (NRM, and central core disease (CCD. However, they are more diverse with overlapping clinical and histopathological features, thus broadening the spectra within each category of congenital myopathy. Conclusion: Identification of cases with overlap of pathological features has diagnostic relevance.

  5. Tangent bundle formulation of a charged gas

    CERN Document Server

    Sarbach, Olivier

    2013-01-01

    We discuss the relativistic kinetic theory for a simple, collisionless, charged gas propagating on an arbitrary curved spacetime geometry. Our general relativistic treatment is formulated on the tangent bundle of the spacetime manifold and takes advantage of its rich geometric structure. In particular, we point out the existence of a natural metric on the tangent bundle and illustrate its role for the development of the relativistic kinetic theory. This metric, combined with the electromagnetic field of the spacetime, yields an appropriate symplectic form on the tangent bundle. The Liouville vector field arises as the Hamiltonian vector field of a natural Hamiltonian. The latter also defines natural energy surfaces, called mass shells, which turn out to be smooth Lorentzian submanifolds. A simple, collisionless, charged gas is described by a distribution function which is defined on the mass shell and satisfies the Liouville equation. Suitable fibre integrals of the distribution function define observable fie...

  6. RodPilot. Control rod control arriving in the 21st century. Higher operational safety and availability, more information, greater ease of operation; RodPilot. Die Steuerstabsteuerung kommt im 21. Jahrhundert an. Hoehere Betriebssicherheit und Verfuegbarkeit, mehr Information und Bedienkomfort

    Energy Technology Data Exchange (ETDEWEB)

    Baron, Clemens [AREVA NP GmbH, Erlangen (Germany)

    2010-01-15

    Areva has successfully launched RodPilot {sup registered}, its new digital control rod control system. After several years of development and qualification as well as testing in experimental facilities, the system was ready for worldwide installation and use in nuclear power plants equipped with pressurized water reactors. It is for this reason that the South African operator, Eskom, decided in favor of Areva and RodPilot when upgrading the reactor control system of its Koeberg nuclear power station. After implementation of RodPilot, the first of two units resumed operation smoothly in early April 2009. RodPilot offers clear increases in operational safety, availability, information, and ease of operation in reactor control especially to the operators of older reactors. (orig.)

  7. Classical Higgs fields on gauge gluon bundles

    Directory of Open Access Journals (Sweden)

    Palese Marcella

    2016-01-01

    Full Text Available Classical Higgs fields and related canonical conserved quantities are defined by invariant variational problems on suitably defined gauge gluon bundles. We consider Lagrangian field theories which are assumed to be invariant with respect to the action of a gauge-natural group. As an illustrative example we exploit the ‘gluon Lagrangian’, i.e. a Yang-Mills Lagrangian on the (1, 1-order gauge-natural bundle of SU(3-principal connections. The kernel of the gauge-natural Jacobi morphism for such a Lagrangian, by inducing a reductive split structure, canonically defines a ‘gluon classical Higgs field’.

  8. Abelian conformal field theory and determinant bundles

    DEFF Research Database (Denmark)

    Andersen, Jørgen Ellegaard; Ueno, K.

    2007-01-01

    Following [10], we study a so-called bc-ghost system of zero conformal dimension from the viewpoint of [14, 16]. We show that the ghost vacua construction results in holomorphic line bundles with connections over holomorphic families of curves. We prove that the curvature of these connections...... are up to a scale the same as the curvature of the connections constructed in [14, 16]. We study the sewing construction for nodal curves and its explicit relation to the constructed connections. Finally we construct preferred holomorphic sections of these line bundles and analyze their behaviour near...

  9. ELECTROMAGNETIC APPARATUS FOR MOVING A ROD

    Science.gov (United States)

    Young, J.N.

    1958-04-22

    An electromagnetic apparatus for moving a rod-like member in small steps in either direction is described. The invention has particular application in the reactor field where the reactor control rods must be moved only a small distance and where the use of mechanical couplings is impractical due to the high- pressure seals required. A neutron-absorbing rod is mounted in a housing with gripping uaits that engage the rod, and coils for magnetizing the gripping units to make them grip, shift, and release the rod are located outside the housing.

  10. Exploiting rod technology. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1990-06-01

    ROD development was proceeding apace until recent budgetary decisions caused funding support for ROD development to be drastically reduced. The funding which was originally provided by DARPA and the Balanced Technology Initiative (BTI) Office has been cut back to zero from $800K. To determine the aeroballistic coefficients of a candidate dart, ARDEC is currently supporting development out of its own 6.2 funds at about $100K. ARDEC has made slow progress toward achieving this end because of failures in the original dart during testing. It appears that the next dart design to be tested will diverge from the original concept visualized by DARPA and Science and Technology Associates (STA). STA, the design engineer, takes exception to these changes on the basis of inappropriate test conditions and insufficient testing. At this time, the full resolution of this issue will be difficult because of the current management structure, which separates the developer (ARDEC) from the designer (STA).

  11. Balance Ability and Proprioception after Single-Bundle, Single-Bundle Augmentation, and Double-Bundle ACL Reconstruction

    Directory of Open Access Journals (Sweden)

    Yubao Ma

    2014-01-01

    Full Text Available Purpose. The present study sought to determine the influences of single-bundle (SB, single-bundle augmentation (SBA, and double-bundle (DB reconstructions on balance ability and proprioceptive function. Methods. 67 patients who underwent a single- or double-bundle ACL reconstruction or a SBA using multistranded autologous hamstring tendons were included in this study with a 1-year follow-up. Body sway and knee kinesthesia (using the threshold to detect passive motion test (TTDPM were measured to indicate balance ability and proprioceptive function, respectively. Additionally, within-subject differences in anterior-posterior stability of the tibia and lower extremity muscle strength were evaluated before and after surgery. Results. At 6 and 12 months after surgery, DB reconstruction resulted in better balance and proprioceptive function than SB reconstruction (P<0.05. Although no significant difference was observed in balance ability or proprioceptive function between the SBA and DB reconstructions, knee stability was significantly better with SBA and DB reconstructions than SB reconstruction (P<0.05. No significant differences were found in quadriceps and hamstrings strength among the three reconstruction techniques. Conclusions. Our findings consider that joint stability, proprioceptive function, and balance ability were superior with SBA and DB reconstructions compared to SB reconstruction at 6 and 12 months after surgery.

  12. Subchannel Code Benchmarking to Columbia University 4x4 and Pacific Northwest Laboratory 2x6 Bundle Test Data

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Kang Hoon; Oezdemir, Erdal; Oh, Seung Jong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2014-10-15

    The subchannel code is used to assess the safety of a reactor core at the steady-state and transient conditions. KEPCO Nuclear Fuel (KNF) has been developed new subchannel code, THALES, for PWR core design application. In this study, we are comparing the THALES result with VIPRE-01 code result utilizing bundle test data. VIPRE-01 was developed under EPRI sponsorship and has been used by U.S. PWR commercial nuclear utilities, historically. THALES and VIPRE-01 codes were benchmarked to two kind of bundle test data which were at the steady-state and transient conditions. THALES predicted fluid velocity and temperature profile of bundle test data well and the error rate between THALES and VIPRE-01 was very small.

  13. Application of hafnium hydride control rod to large sodium cooled fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ikeda, Kazumi, E-mail: kazumi_ikeda@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Moriwaki, Hiroyuki, E-mail: hiroyuki_moriwaki@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Ohkubo, Yoshiyuki, E-mail: yoshiyuki_okubo@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Iwasaki, Tomohiko, E-mail: tomohiko.iwasaki@qse.tohoku.ac.jp [Department of Quantum Science and Energy Engineering, Tohoku University, Aoba, Aramaki, Aoba-ku, Sendai-shi, Miyagi-ken 980-8579 (Japan); Konashi, Kenji, E-mail: konashi@imr.tohoku.ac.jp [Institute for Materials Research, Tohoku University, Narita-cho, Oarai-machi, Higashi-Ibaraki-gun, Ibaraki-ken 311-1313 (Japan)

    2014-10-15

    Highlights: • Application of hafnium hydride control rod to large sodium cooled fast breeder reactor. • This paper treats application of an innovative hafnium hydride control rod to a large sodium cooled fast breeder reactor. • Hydrogen absorption triples the reactivity worth by neutron spectrum shift at H/Hf ratio of 1.3. • Lifetime of the control rod quadruples because produced daughters of hafnium isotopes are absorbers. • Nuclear and thermal hydraulic characteristics of the reactor are as good as or better than B-10 enriched boron carbide. - Abstract: This study treats the feasibility of long-lived hafnium hydride control rod in a large sodium-cooled fast breeder reactor by nuclear and thermal analyses. According to the nuclear calculations, it is found that hydrogen absorption of hafnium triples the reactivity by the neutron spectrum shift at the H/Hf ratio of 1.3, and a hafnium transmutation mechanism that produced daughters are absorbers quadruples the lifetime due to a low incineration rate of absorbing nuclides under irradiation. That is to say, the control rod can function well for a long time because an irradiation of 2400 EFPD reduces the reactivity by only 4%. The calculation also reveals that the hafnium hydride control rod can apply to the reactor in that nuclear and thermal characteristics become as good as or better than 80% B-10 enriched boron carbide. For example, the maximum linear heat rate becomes 3% lower. Owing to the better power distribution, the required flow rate decreases approximately by 1%. Consequently, it is concluded on desk analyses that the long lived hafnium hydride control rod is feasible in the large sodium-cooled fast breeder reactor.

  14. Computations in intersection rings of flag bundles

    CERN Document Server

    Grayson, Daniel R; Stillman, Michael E

    2012-01-01

    Intersection rings of flag varieties and of isotropic flag varieties are generated by Chern classes of the tautological bundles modulo the relations coming from multiplicativity of total Chern classes. In this paper we describe the Groebner bases of the ideals of relations and give applications to computation of intersections, as implemented in Macaulay2.

  15. Capacity efficiency of recovery request bundling

    DEFF Research Database (Denmark)

    Ruepp, Sarah Renée; Dittmann, Lars; Berger, Michael Stübert

    2010-01-01

    This paper presents a comparison of recovery methods in terms of capacity efficiency. In particular, a method where recovery requests are bundled towards the destination (Shortcut Span Protection) is evaluated against traditional recovery methods. Our simulation results show that Shortcut Span...... Protection uses more capacity than the unbundled related methods, but this is compensated by easier control and management of the recovery actions....

  16. η-Invariant and Flat Vector Bundles

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    We present an alternate definition of the mod Z component of the AtiyahPatodi-Singer η invariant associated to (not necessary unitary) fiat vector bundles, which identifies explicitly its real and imaginary parts. This is done by combining a deformation of flat connections introduced in a previous paper with the analytic continuation procedure appearing in the original article of Atiyah, Parodi and Singer.

  17. Lazarsfeld-Mukai bundles and applications

    CERN Document Server

    Aprodu, Marian

    2012-01-01

    We survey the development of the notion of Lazarsfeld-Mukai bundles together with various applications, from the classification of Mukai manifolds to Brill-Noether theory and syzygies of $K3$ sections. To see these techniques at work, we present a short proof of a result of M. Reid on the existence of elliptic pencils.

  18. Meromorphic Higgs bundles And Related Geometries

    OpenAIRE

    Dalakov, Peter

    2016-01-01

    The present note is mostly a survey on the generalised Hitchin integrable system and moduli spaces of meromorphic Higgs bundles. We also fill minor gaps in the existing literature, outline a calculation of the infinitesimal period map and review briefly some related geometries.

  19. Meromorphic Higgs bundles and related geometries

    Science.gov (United States)

    Dalakov, Peter

    2016-11-01

    The present note is mostly a survey on the generalised Hitchin integrable system and moduli spaces of meromorphic G-Higgs bundles. We also fill minor gaps in the existing literature, outline a calculation of the infinitesimal period map and review some related geometries.

  20. The Hodge bundle on Hurwitz spaces

    NARCIS (Netherlands)

    van der Geer, G.; Kouvidakis, A.

    2011-01-01

    In 2009 Kokotov, Korotkin and Zograf gave in [7] a formula for the class of the Hodge bundle on the Hurwitz space of admissible covers of genus g and degree d of the projective line. They gave an analytic proof of it. In this note we give an algebraic proof and an extension of the result.

  1. Capacity efficiency of recovery request bundling

    DEFF Research Database (Denmark)

    Ruepp, Sarah Renée; Dittmann, Lars; Berger, Michael Stübert

    2010-01-01

    This paper presents a comparison of recovery methods in terms of capacity efficiency. In particular, a method where recovery requests are bundled towards the destination (Shortcut Span Protection) is evaluated against traditional recovery methods. Our simulation results show that Shortcut Span...

  2. Strategy for Fuel Rod Receipt, Characterization, Sample Allocation for the Demonstration Sister Rods

    Energy Technology Data Exchange (ETDEWEB)

    Marschman, Steven C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Warmann, Stephan A. [Portage, Inc., Idaho Falls, ID (United States); Rusch, Chris [NAC International, Inc., Norcross, GA (United States)

    2014-03-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technology, has established the Used Fuel Disposition Campaign (UFDC) to conduct the research and development activities related to storage, transportation, and disposal of used nuclear fuel and high-level radioactive waste. The mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel (UNF) and wastes generated by existing and future nuclear fuel cycles. The UFDC Storage and Transportation staffs are responsible for addressing issues regarding the extended or long-term storage of UNF and its subsequent transportation. The near-term objectives of the Storage and Transportation task are to use a science-based approach to develop the technical bases to support the continued safe and secure storage of UNF for extended periods, subsequent retrieval, and transportation. While low burnup fuel [that characterized as having a burnup of less than 45 gigawatt days per metric tonne uranium (GWD/MTU)] has been stored for nearly three decades, the storage of high burnup used fuels is more recent. The DOE has funded a demonstration project to confirm the behavior of used high burnup fuel under prototypic conditions. The Electric Power Research Institute (EPRI) is leading a project team to develop and implement the Test Plan to collect this data from a UNF dry storage system containing high burnup fuel. The Draft Test Plan for the demonstration outlines the data to be collected; the high burnup fuel to be included; the technical data gaps the data will address; and the storage system design, procedures, and licensing necessary to implement the Test Plan. To provide data that is most relevant to high burnup fuel in dry storage, the design of the test storage system must closely mimic real conditions high burnup SNF experiences during all stages of dry storage: loading, cask drying

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

  4. Active Hair-Bundle Motility by the Vertebrate Hair Cell

    Science.gov (United States)

    Tinevez, J.-Y.; Martin, P.; Jülicher, F.

    2009-02-01

    The hair bundle is both a mechano-sensory antenna and a force generator that might help the vertebrate hair cell from the inner ear to amplify its responsiveness to small stimuli. To study active hair-bundle motility, we combined calcium iontophoresis with mechanical stimulation of single hair bundles from the bullfrog's sacculus. A hair bundle could oscillate spontaneously, or be quiescent but display non-monotonic movements in response to abrupt force steps. Extracellular calcium changes or static biases to the bundle's position at rest could affect the kinetics of bundle motion and evoke transitions between the different classes of motility. The calcium-dependent location of a bundle's operating point within its nonlinear force-displacement relation controlled the type of movements observed. A unified theoretical description, in which mechanical activity stems from myosin-based adaptation and electro-mechanical feedback by Ca2+, could account for the fast and slow manifestations of active hair-bundle motility.

  5. Model predictive load following control for APR+ reactors through discrete control rod speed optimization

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jae Hwan; Kim, Ju Hyun; Park, Sun Ho; Yoo, Kwae Hwan; Kim, Dae Seop; Na, Man Gyun [Chosun Univ., Gwanggu (Korea, Republic of)

    2012-10-15

    Until now, nuclear power has been only used for the base load power operation. However, current nuclear power plants are recognized as the most reasonable energy source. As a result, the proportion of nuclear power has being grown increasingly. Therefore, load following operation of a nuclear power plant should be an essential option. Most of the existing nuclear power plants perform reactor operation by varying the boron concentration in the coolant. But it is hard to respond quickly to demands for the power changes. In case of using the control rods, reactivity control is easy, but axial power distribution control is very hard because it has very complex and nonlinear dynamic characteristics. In this study, we have introduced a Model Predictive Control (MPC) method to control the average coolant temperature and Axial Shape Index (ASI) automatically at the same time, and we have improved the performance of controller by applying the Genetic Algorithm (GA) to optimize the control rod movement.

  6. LUSTERNIK-S CHNIRELMANN CATEGORY AND EMBEDDING FINITE COVERING MAPS, PRINCIPAL G-BUNDLES INTO BUNDLES

    Institute of Scientific and Technical Information of China (English)

    LIULUOFEI

    1996-01-01

    The author proves several embedding theorems for finite covering maps,principal G-bundies into bundles.The main results are 1. Let π:E→X be a finite covering map, and X a connected locally path-connected paracompact space. If cat X≤k, then the finite covering space π:E→X can be embedded into the trivial real k-plane bundle. 2. Let π:E→X be a principal G-bundle over a paracompact space. If there exists a linera action of Gon F(F=R or C)and cat X≤k ,then π:E→X can be embedded into ξ1 … ξn for any F-vector bundles ξi,i=1,…k.

  7. Interplanetary Overlay Network Bundle Protocol Implementation

    Science.gov (United States)

    Burleigh, Scott C.

    2011-01-01

    The Interplanetary Overlay Network (ION) system's BP package, an implementation of the Delay-Tolerant Networking (DTN) Bundle Protocol (BP) and supporting services, has been specifically designed to be suitable for use on deep-space robotic vehicles. Although the ION BP implementation is unique in its use of zero-copy objects for high performance, and in its use of resource-sensitive rate control, it is fully interoperable with other implementations of the BP specification (Internet RFC 5050). The ION BP implementation is built using the same software infrastructure that underlies the implementation of the CCSDS (Consultative Committee for Space Data Systems) File Delivery Protocol (CFDP) built into the flight software of Deep Impact. It is designed to minimize resource consumption, while maximizing operational robustness. For example, no dynamic allocation of system memory is required. Like all the other ION packages, ION's BP implementation is designed to port readily between Linux and Solaris (for easy development and for ground system operations) and VxWorks (for flight systems operations). The exact same source code is exercised in both environments. Initially included in the ION BP implementations are the following: libraries of functions used in constructing bundle forwarders and convergence-layer (CL) input and output adapters; a simple prototype bundle forwarder and associated CL adapters designed to run over an IPbased local area network; administrative tools for managing a simple DTN infrastructure built from these components; a background daemon process that silently destroys bundles whose time-to-live intervals have expired; a library of functions exposed to applications, enabling them to issue and receive data encapsulated in DTN bundles; and some simple applications that can be used for system checkout and benchmarking.

  8. Full Scale Component Test Facility KOPRA - Qualification Test of EPR Control Rod Drive Mechanism

    Energy Technology Data Exchange (ETDEWEB)

    Sykora, Alexander; Herr, Wolfgang [AREVA NP GmbH, P.O. Box 3220, 91050 Erlangen (Germany); Champomier, Francois [AREVA NP SAS, Tour AREVA - Cedex 16, 92084 Paris-La Defense (France)

    2008-07-01

    The test facility KOPRA is designed for full scale-tests on nuclear components under operational conditions. One part of it is the component test loop for developing and qualifying nuclear core components respecting temperature, pressure and mass flow of pressurized water reactor conditions. The KOPRA test facility and its measuring equipment is presented through qualification tests for the control rod drive mechanism and the control rod drive line of the new European Pressurized Water Reactor (EPR). The control rod drive mechanism qualification test program is split into three different test phases. At first, performance tests are conducted to verify the adequate performance of the new equipment, e.g. measurement of rod cluster control assembly drop time under different thermal hydraulic conditions, impact velocity of drive rod on CRDM latch tips and drive rod acceleration during stepping operation by means of strain gauges or through direct measurement. After these functional tests follow the stability tests to ensure that proper functioning is reliably achieved over an appreciable amount of time and the endurance tests to quantify the amount of time and/or the number of steps during which no appreciable wear, that could possibly alter the correct behaviour, is to be expected. (authors)

  9. Bundling Revisited: Substitute Products and Inter-Firm Discounts

    OpenAIRE

    Armstrong, Mark

    2011-01-01

    This paper extends the standard model of bundling to allow products to be substitutes and for products to be supplied by separate sellers. Whether integrated or separate, firms have an incentive to introduce bundling discounts when demand for the bundle is elastic relative to demand for stand-alone products. When products are partial substitutes, this typically gives an integrated firm a greater incentive to offer a bundle discount (relative to the standard model with additive preferences), w...

  10. Holomorphic Vector Bundle on Hopf Manifolds with Abelian Fundamental Groups

    Institute of Scientific and Technical Information of China (English)

    Xiang Yu ZHOU; Wei Ming LIU

    2004-01-01

    Let X be a Hopf manifolds with an Abelian fundamental group. E is a holomorphic vector bundle of rank r with trivial pull-back to W = Cn - {0}. We prove the existence of a non-vanishing section of L(×) E for some line bundle on X and study the vector bundles filtration structure of E. These generalize the results of D. Mall about structure theorem of such a vector bundle E.

  11. Internal hydriding in irradiated defected Zircaloy fuel rods: A review (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Clayton, J C

    1987-10-01

    Although not a problem in recent commercial power reactors, including the Shippingport Light Water Breeder Reactor, internal hydriding of Zircaloy cladding was a persistent cause of gross cladding failures during the 1960s. It occurred in the fuel rods of water-cooled nuclear power reactors that had a small cladding defect. This report summarizes the experimental findings, causes, mechanisms, and methods of minimizing internal hydriding in defected Zircaloy-clad fuel rods. Irradiation test data on the different types of defected fuel rods, intentionally fabricated defected and in-pile operationally defected rods, are compared. Significant factors affecting internal hydriding in defected Zircaloy-clad fuel rods (defect hole size, internal and external sources of hydrogen, Zircaloy cladding surface properties, nickel alloy contamination of Zircaloy, the effect of heat flux and fluence) are discussed. Pertinent in-pile and out-of-pile test results from Bettis and other laboratories are used as a data base in constructing a qualitative model which explains hydrogen generation and distribution in Zircaloy cladding of defected water-cooled reactor fuel rods. Techniques for minimizing internal hydride failures in Zircaloy-clad fuel rods are evaluated.

  12. 核电厂控制棒导向筒半方管真空电子束焊接的试验研究%Manufacturing Process of Electric Beam Welding on Control Rod Guide Tube of Million-kilowatt-class Nuclear Power Station

    Institute of Scientific and Technical Information of China (English)

    张斌; 吴义党; 刘鸣宇; 张其先; 陈洪

    2013-01-01

    百万千瓦级核电站控制棒导向筒中半方管采用真空电子束进行焊接.本文制定了合理的试验方案,对真空电子束焊接的热循环曲线与焊缝及附近区域残余应力的分布进行了试验测量,并分析了其热循环与残余应力的特征,为更好的控制焊接质量与制造精度提供依据.%The control rod guide tube of million-kilowatt-class nuclear power station was welded by electric beam welding. A test plan for the measurement on welding thermal circulation and residual stress was provided. The character and relationship of welding thermal circulation and residual stress were analyzed in order to provide a practical basis for improving the welding quality and manufacturing precision.

  13. A Comparison between Clinical Results of Selective Bundle and Double Bundle Anterior Cruciate Ligament Reconstruction

    Science.gov (United States)

    Yoo, Yon-Sik; Song, Si Young; Yang, Cheol Jung; Ha, Jong Mun; Kim, Yoon Sang

    2016-01-01

    Purpose The purpose of this study was to compare the clinical outcomes of arthroscopic anatomical double bundle (DB) anterior cruciate ligament (ACL) reconstruction with either selective anteromedial (AM) or posterolateral (PL) bundle reconstruction while preserving a relatively healthy ACL bundle. Materials and Methods The authors evaluated 98 patients with a mean follow-up of 30.8±4.0 months who had undergone DB or selective bundle ACL reconstructions. Of these, 34 cases underwent DB ACL reconstruction (group A), 34 underwent selective AM bundle reconstruction (group B), and 30 underwent selective PL bundle reconstructions (group C). These groups were compared with respect to Lysholm and International Knee Documentation Committee (IKDC) score, side-to-side differences of anterior laxity measured by KT-2000 arthrometer at 30 lbs, and stress radiography and Lachman and pivot shift test results. Pre- and post-operative data were objectively evaluated using a statistical approach. Results The preoperative anterior instability measured by manual stress radiography at 90° of knee flexion in group A was significantly greater than that in groups B and C (all pACL tears offers comparable clinical results to DB reconstruction in complete ACL tears. PMID:27401652

  14. Heat transfer in a vertical 7-element bundle cooled with supercritical Freon-12

    Energy Technology Data Exchange (ETDEWEB)

    Richards, G.; Milner, A.; Pascoe, C.; Patel, H.; Peiman, W.; Saltanov, Eu. [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada); Pometko, R.S.; Opanasenko, A.N. [Inst. of Physics and Power Engineering, Obninsk (Russian Federation); Shelegov, A.S. [IATE, Obninsk (Russian Federation); Kirillov, P.L. [Inst. of Physics and Power Engineering, Obninsk (Russian Federation); Pioro, I. [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada)

    2010-07-01

    Currently, SuperCritical Water-cooled nuclear Reactor (SCWR) concepts are being developed worldwide with an objective to increase thermal efficiencies of future Nuclear Power Plants (NPPs) on 10 -15% compared to those of current water-cooled NPPs. With such an increase in the thermal efficiencies, SCW NPPs will be at the current level of the most advanced thermal power plants: coal-fired SCW NPPs and combined-cycle gas-fired NPPs. However, to be able to develop SCWRs at least several key technical problems should be resolved. One of these problems is limited amount of experimental data on heat transfer in fuel bundles and based on that SCW heat-transfer correlations. Experiments in SCW are very complicated and expensive due to high critical parameters of water (pressure 22.064 MPa and temperature 374.95°C). Moreover, there are only few SCW test rigs, which capable to perform experiments in full-scale bundles. As a preliminary approach supercritical-pressure heat-transfer experiments in bundles can be performed in modeling fluids such as Freons or carbon dioxide. Therefore, a set of experimental data was obtained in Freon-12-cooled bundle simulator at the Institute of Physics and Power Engineering (IPPE, Obninsk, Russia). A vertical 7-element bundle was installed in a hexagonal flow channel. The test section consisted of elements that were 9.5 mm in diameter with the total heated length of 1 m. Bulk-fluid and wall temperature profiles were recorded using thermocouples. Several heat-transfer regimes were tested. Also, this paper references thermophysical properties of supercritical Freon-12 at the critical pressure (4.1361 MPa) and test pressure of 4.65 MPa. (author)

  15. VECTOR BUNDLE, KILLING VECTOR FIELD AND PONTRYAGIN NUMBERS

    Institute of Scientific and Technical Information of China (English)

    周建伟

    1991-01-01

    Let E be a vector bundle over a compact Riemannian manifold M. We construct a natural metric on the bundle space E and discuss the relationship between the killing vector fields of E and M. Then we give a proof of the Bott-Baum-Cheeger Theorem for vector bundle E.

  16. Heat exchanger with helical bundles of finned tubes

    Energy Technology Data Exchange (ETDEWEB)

    Eyking, H.J.

    1975-01-23

    The invention applies to a heat exchanger with helical bundles of tubes consisting of finned tubes separated by spacers. The spacers are designed as closed holding cylinders with holding devices for the tube bundles, each ot which surrounds a bundle of tubes. This construction serves to simplify the production process and to enable the use of the heat exchanger at higher loads.

  17. Gauge bundles and Born-Infeld on the noncommutative torus

    NARCIS (Netherlands)

    Hofman, C.; Verlinde, E.

    1998-01-01

    In this paper, we describe non-abelian gauge bundles with magnetic and electric uxes on higher dimensional noncomm utative tori. We give an explicit construction of a large class of bundles with nonzero magnetic 't Hooft uxes. W e discuss Morita equiv alence between these bundles. The action of

  18. QTLs analysis of rice peduncle vascular bundle and panicle traits

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    @@The vascular bundle in plants plays an important role in transportation of photosynthetic products, mineral nutrients, water, and so on. Significant positive correlations were found between grain yield, panicle traits and the No. Of peduncle vascular bundles. So, it is very important to study the inheritance of peduncle vascular bundle, which is a quantitative trait.

  19. The sperm of Matsucoccus feytaudi (Insecta, Coccoidea): Can the microtubular bundle be considered as a true flagellum?

    Science.gov (United States)

    Paoli, Francesco; Roversi, Pio Federico; Benassai, Daniele; Squarcini, Michele; Mercati, David; Dallai, Romano

    2015-03-01

    In the present work the spermiogenesis and sperm structure of Matsucoccus feytaudi, a primary pest of the maritime pine in southern eastern Europe, is studied. In addition to the already known characteristics of coccid sperm, such as the absence of the acrosome and mitochondria, and the presence of a bundle of microtubules responsible for sperm motility, a peculiar structure from which the microtubule bundle takes origin is described. Such a structure--a short cylinder provided with a central hub surrounded by several microtubules with a dense wall--is regarded as a Microtubule Organizing Centre (MTOC). During spermiogenesis, quartets of fused spermatids are formed; from each spermatid, a bundle of microtubules, generated by the MTOC, projects from the cell surface. Each cell has two centrioles, suggesting the lack of a meiotic process and the occurrence of parthenogenesis. At the end of the spermiogenesis, when the cysts containing bundles of sperm are formed, part of the nuclear material together with the MTOC structure is eliminated. Based on the origin of the microtubular bundle from the MTOC, the nature of the bundle as a flagellum is discussed.

  20. Stability of Picard Bundle Over Moduli Space of Stable Vector Bundles of Rank Two Over a Curve

    Indian Academy of Sciences (India)

    Indranil Biswas; Tomás L Gómez

    2001-08-01

    Answering a question of [BV] it is proved that the Picard bundle on the moduli space of stable vector bundles of rank two, on a Riemann surface of genus at least three, with fixed determinant of odd degree is stable.

  1. Core Quadrant Power Tilt Induced by Control Rods Axial Position Mismatch

    Energy Technology Data Exchange (ETDEWEB)

    Son, Seungbeom; Lee, Changhee; Woo, Haeseuk; Jung, Yilsup [KEPCO Nuclear Fuel, Daejeon (Korea, Republic of); Kim, Yonghee [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    If the limits are not met, plant operation should be restricted at various conditions. Actually, core quadrant power tilt in many nuclear power plants has been occurred frequently. This study deals with the various general causes of the core quadrant power tilt, including the control rods axial position mismatch induced quadrant power tilt problem. Based on the simulation results in each control rods axial position mismatch in a same control bank, it was found that some control rods axial position mismatch (simply around 10 steps difference) can induce quadrant power tilt relatively high (around 1.5%). With the other causes induced QPT, it may exceed the technical specification limit (2%). So it is needed to drive the control rods in detail during the operation including power change mode.

  2. CALCULATION OF STRESS AND DEFORMATION IN FUEL ROD CLADDING DURING PELLET-CLADDING INTERACTION

    Directory of Open Access Journals (Sweden)

    Dávid Halabuk

    2015-12-01

    Full Text Available The elementary parts of every fuel assembly, and thus of the reactor core, are fuel rods. The main function of cladding is hermetic separation of nuclear fuel from coolant. The fuel rod works in very specific and difficult conditions, so there are high requirements on its reliability and safety. During irradiation of fuel rods, a state may occur when fuel pellet and cladding interact. This state is followed by changes of stress and deformations in the fuel cladding. The article is focused on stress and deformation analysis of fuel cladding, where two fuels are compared: a fresh one and a spent one, which is in contact with cladding. The calculations are done for 4 different shapes of fuel pellets. It is possible to evaluate which shape of fuel pellet is the most appropriate in consideration of stress and deformation forming in fuel cladding, axial dilatation of fuel, and radial temperature distribution in the fuel rod, based on the obtained results.

  3. Growth Factor Receptor-Bound Protein 14 Undergoes Light-Dependent Intracellular Translocation in Rod Photoreceptors: Functional Role on Retinal Insulin Receptor Activation

    OpenAIRE

    Rajala, Ammaji; Roger J. Daly; Tanito, Masaki; Allen, Dustin T.; Lowenna J Holt; Lobanova, Ekaterina; Arshavsky, Vadim Y; Rajala, Raju V.S.

    2009-01-01

    Growth factor receptor-bound protein 14 (Grb14) is involved in growth factor receptor tyrosine kinase signaling. Here we report that light causes a major redistribution of Grb14 among the individual subcellular compartments of the retinal rod photoreceptor. Grb14 is localized predominantly to the inner segment, nuclear layer and synapse in dark-adapted rods, whereas in the light-adapted rods, Grb14 redistributed throughout the entire cell, including the outer segment. The translocation of Grb...

  4. Topological Mixing with Ghost Rods

    OpenAIRE

    2005-01-01

    Topological chaos relies on the periodic motion of obstacles in a two-dimensional flow in order to form nontrivial braids. This motion generates exponential stretching of material lines, and hence efficient mixing. Boyland et al. [P. L. Boyland, H. Aref, and M. A. Stremler, J. Fluid Mech. 403, 277 (2000)] have studied a specific periodic motion of rods that exhibits topological chaos in a viscous fluid. We show that it is possible to extend their work to cases where the motion of the stirring...

  5. Control rod calibration and reactivity effects at the IPEN/MB-01 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pinto, Letícia Negrão; Gonnelli, Eduardo; Santos, Adimir dos [Nuclear Engineering Center, Nuclear and Energy Research Institute- IPEN/CNEN-SP, Av. Lineu Prestes 2242 - Cidade Universitária - 05508-000 - São Paulo - SP (Brazil)

    2014-11-11

    Researches that aim to improve the performance of neutron transport codes and quality of nuclear cross section databases are very important to increase the accuracy of simulations and the quality of the analysis and prediction of phenomena in the nuclear field. In this context, relevant experimental data such as reactivity worth measurements are needed. Control rods may be made of several neutron absorbing materials that are used to adjust the reactivity of the core. For the reactor operation, these experimental data are also extremely important: with them it is possible to estimate the reactivity worth by the movement of the control rod, understand the reactor response at each rod position and to operate the reactor safely. This work presents a temperature correction approach for the control rod calibration problem. It is shown the control rod calibration data of the IPEN/MB-01 reactor, the integral and differential reactivity curves and a theoretical analysis, performed by the MCNP-5 reactor physics code, developed and maintained by Los Alamos National Laboratory, using the ENDF/B-VII.0 nuclear data library.

  6. Control rod calibration and reactivity effects at the IPEN/MB-01 reactor

    Science.gov (United States)

    Pinto, Letícia Negrão; Gonnelli, Eduardo; dos Santos, Adimir

    2014-11-01

    Researches that aim to improve the performance of neutron transport codes and quality of nuclear cross section databases are very important to increase the accuracy of simulations and the quality of the analysis and prediction of phenomena in the nuclear field. In this context, relevant experimental data such as reactivity worth measurements are needed. Control rods may be made of several neutron absorbing materials that are used to adjust the reactivity of the core. For the reactor operation, these experimental data are also extremely important: with them it is possible to estimate the reactivity worth by the movement of the control rod, understand the reactor response at each rod position and to operate the reactor safely. This work presents a temperature correction approach for the control rod calibration problem. It is shown the control rod calibration data of the IPEN/MB-01 reactor, the integral and differential reactivity curves and a theoretical analysis, performed by the MCNP-5 reactor physics code, developed and maintained by Los Alamos National Laboratory, using the ENDF/B-VII.0 nuclear data library.

  7. Test Methodology of Reproducing Fuel Rod Failure by Debris Fretting Wear

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Oh Joon; Park, Nam Gyu; Kim, Jae Ik [KEPCO NF, Daejeon (Korea, Republic of)

    2015-10-15

    A test was conducted with simple debris to reproduce debris fretting wear. 68% of fuel rod cladding thickness is worn out by Inconel debris in 75 hours. The test result shows that a simple link system is useful to accommodate debris oscillation, and mid grid mixing vanes could be a source of debris forcing. Additional tests will be conducted with various debris such as wire brush, metal chip, etc which are suspected to generate actual debris fretting wear in future works. Debris fretting is one of the most common cause of the nuclear fuel rod failure. Even the most of the nuclear fuels has debris protection system, debris still cause fuel rod failure. From 1994 to 2006, debris fretting failure is around 11% of the total fuel failure. In 2006-2010, the portion of debris rises to over 13%. The total number of fuel rods failure is decreasing, but the portion of the debris fretting wear is growing with time. Therefore reproducing and identifying the mechanism of fuel rod failure by debris fretting wear is needed to improve reliability of the nuclear fuel.

  8. Productivity and costs of slash bundling in Nordic conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kaerhae, K.; Vartiamaeki, T. [Metsaeteho Oy, P.O. Box 101, FI-00171 Helsinki (Finland)

    2006-12-15

    The number of slash bundlers and the volume of slash bundling have been rapidly increasing during the last few years in Finland. However, no comprehensive time or follow-up studies have been carried out on slash bundling technology in Finland or in any other country. Metsateho Oy carried out studies on the productivity and costs of slash bundling in different Nordic recovering conditions. The study methods included both time and follow-up studies. Data were collected during the summer and winter period primarily in Norway spruce (Picea abies L. Karst.) dominated clear cutting sites. The bundling techniques performed by different types of bundler (Fiberpac 370, Timberjack 1490D, Pika RS 2000, Valmet WoodPac) were studied. The average productivity of slash bundling was 18.1 bundles per operating (E{sub 15}, including delays shorter than 15min) hour with the Timberjack 1490D and Fiberpac 370 bundlers in the follow-up study. The operator of the slash bundler had the greatest effect on the productivity of bundling. The prerequisite for increased bundling volumes is a reduction in the costs of the most expensive sub-stage of the bundling supply chain, i.e. bundling itself. This requires improved recovery conditions at bundling sites, increased bundling productivity, larger sized bundles, and the execution of bundling operations in two work shifts using an efficient bundler and effective operator working methods. Implementation of these development measures will bring the bundling supply chain up to a speed that makes it the most competitive supply chain for forest chips in terms of total supply costs for long-distance transportation distances of more than 60km. (author)

  9. Development of a suitable weld geometry for pressure resistance welding of the leader test assembly (LTA's) 16NGF fuel assembly fuel rod at Angra-1 Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Junqueira, Fabio da Silva; Silva, Josue Ribeiro, E-mail: fabiojunqueira@inb.gov.br, E-mail: josueribeiro@inb.gov.br [Industrias Nucleares do Brasil (GEPRDN/INB), Rio de Janeiro, RJ (Brazil). Gerencia do Produto

    2013-07-01

    The purpose of this work is to develop suitable weld geometry for pressure resistance welding of the zircaloy-4 end plug to the special zirconium alloy cladding tube, Ø 9,14mm, for demonstration at Angra-1 Nuclear Plant. Weld geometry development was carried out in two steps: at the first one, the influence caused by the variation of the welding process key parameters, the axial compression strength of the end plug against the cladding tube, projection of the cladding tube into the welding chamber and the welding current have been evaluated; at the second step, the influence of the variation of end-plug weld geometry area was checked. For the combination of welding parameters, the technique of factorial design was used. Results from mechanical and metallographic tests have indicated a strong and direct influence of weld geometry dimensional variation on the weld mechanical resistance, and a modest influence in relation to the range of key parameters used to carry out tests. (author)

  10. Bundling of harvesting residues and whole-trees and the treatment of bundles; Hakkuutaehteiden ja kokopuiden niputus ja nippujen kaesittely

    Energy Technology Data Exchange (ETDEWEB)

    Kaipainen, H.; Seppaenen, V.; Rinne, S.

    1996-12-31

    The conditions on which the bundling of the harvesting residues from spruce regeneration fellings would become profitable were studied. The calculations showed that one of the most important features was sufficient compaction of the bundle, so that the portion of the wood in the unit volume of the bundle has to be more than 40 %. The tests showed that the timber grab loader of farm tractor was insufficient for production of dense bundles. The feeding and compression device of the prototype bundler was constructed in the research and with this device the required density was obtained.The rate of compaction of the dry spruce felling residues was about 40 % and that of the fresh residues was more than 50 %. The comparison between the bundles showed that the calorific value of the fresh bundle per unit volume was nearly 30 % higher than that of the dry bundle. This means that the treatment of the bundles should be done of fresh felling residues. Drying of the bundles succeeded well, and the crushing and chipping tests showed that the processing of the bundles at the plant is possible. The treatability of the bundles was also excellent. By using the prototype, developed in the research, it was possible to produce a bundle of the fresh spruce harvesting residues, the diameter of which was about 50 cm and the length about 3 m, and the rate of compaction over 50 %. By these values the reduction target of the costs is obtainable

  11. ASSERT-PV 3.2: Advanced subchannel thermalhydraulics code for CANDU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Rao, Y.F., E-mail: raoy@aecl.ca; Cheng, Z., E-mail: chengz@aecl.ca; Waddington, G.M., E-mail: waddingg@aecl.ca; Nava-Dominguez, A., E-mail: navadoma@aecl.ca

    2014-08-15

    Highlights: • Introduction to a new version of the Canadian subchannel code, ASSERT-PV 3.2. • Enhanced models for flow-distribution, CHF and post-dryout heat transfer prediction. • Model changes focused on unique features of horizontal CANDU bundles. • Detailed description of model changes for all major thermalhydraulics models. • Discussion on rationale and limitation of the model changes. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The most recent release version, ASSERT-PV 3.2 has enhanced phenomenon models for improved predictions of flow distribution, dryout power and CHF location, and post-dryout (PDO) sheath temperature in horizontal CANDU fuel bundles. The focus of the improvements is mainly on modeling considerations for the unique features of CANDU bundles such as horizontal flows, small pitch to diameter ratios, high mass fluxes, and mixed and irregular subchannel geometries, compared to PWR/BWR fuel assemblies. This paper provides a general introduction to ASSERT-PV 3.2, and describes the model changes or additions in the new version to improve predictions of flow distribution, dryout power and CHF location, and PDO sheath temperatures in CANDU fuel bundles.

  12. Reactor physics analysis for the design of nuclear fuel lattices with burnable poisons

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa-Paredes, G. [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico, D.F. (Mexico); Guzman, Juan R., E-mail: maestro_juan_rafael@hotmail.com [Departamento de Fisica y Matematicas, Instituto Politecnico Nacional, Adolfo Lopez Mateos, San Pedro Zacatenco, 07738 Mexico, D.F. (Mexico)

    2011-12-15

    Highlights: Black-Right-Pointing-Pointer A fuel rod optimization for the coupled bundle-core design in a BWR is developed. Black-Right-Pointing-Pointer An algorithm to minimize the rod power peaking factor is used. Black-Right-Pointing-Pointer The fissile content is divided in two factors. Black-Right-Pointing-Pointer A reactor physics analysis of these factors is performed. Black-Right-Pointing-Pointer The algorithm is applied to a typical BWR fuel lattice. - Abstract: The main goals in nuclear fuel lattice design are: (1) minimizing the rod power peaking factor (PPF) in order that the power level distribution is the most uniform; (2) obtaining a prescribed target value for the multiplication factor (k) at the end of the irradiation in order that the fuel lattice reaches the desired reactivity; and (3) obtaining a prescribed target value for the k at the beginning of the irradiation in order that the reactivity excess is neither a high value (to ease the maneuvering of the control systems) nor a low value (to avoid the penalization of the high cost of the burnable poison content). In this work a simple algorithm to design the burnable poison bearing nuclear fuel lattice is presented. This algorithm is based on a reactor physics analysis. The algorithm is focused on finding the radial distribution of the fuel rods having different fissile and burnable poison contents in order to obtain: (1) an adequate minimum PPF; (2) a prescribed target value of the k at the end of the irradiation; and (3) a prescribed target value of the k at the beginning of the irradiation. This algorithm is based on the factorization of the fissile and burnable poison contents of each fuel rod and on the application of the first-order perturbation theory. The performance of the algorithm is demonstrated with the design of a fuel lattice composed of uranium dioxide (UO{sub 2}) and gadolinium dioxide (Gd{sub 2}O{sub 3}) for boiling water reactors (BWR). This algorithm has been accomplished

  13. Phase Slips in Oscillatory Hair Bundles

    Science.gov (United States)

    Roongthumskul, Yuttana; Shlomovitz, Roie; Bruinsma, Robijn; Bozovic, Dolores

    2013-01-01

    Hair cells of the inner ear contain an active amplifier that allows them to detect extremely weak signals. As one of the manifestations of an active process, spontaneous oscillations arise in fluid immersed hair bundles of in vitro preparations of selected auditory and vestibular organs. We measure the phase-locking dynamics of oscillatory bundles exposed to low-amplitude sinusoidal signals, a transition that can be described by a saddle-node bifurcation on an invariant circle. The transition is characterized by the occurrence of phase slips, at a rate that is dependent on the amplitude and detuning of the applied drive. The resultant staircase structure in the phase of the oscillation can be described by the stochastic Adler equation, which reproduces the statistics of phase slip production. PMID:25167040

  14. Care bundles reduce readmissions for COPD.

    Science.gov (United States)

    Matthews, Healther; Tooley, Cathy; Nicholls, Carol; Lindsey-Halls, Anna

    In 2011, the respiratory nursing team at the James Paget University Hospital Foundation Trust were considering introducing a discharge care bundle for patients admitted with an acute exacerbation of chronic obstructive pulmonary disease. At the same time, the trust was asking for applications for Commissioning for Quality and Innovation schemes (CQUINs). These are locally agreed packages of quality improvement goals and indicators, which, if achieved in total, enable the provider to earn its full CQUIN payment. A CQUIN scheme should address the three domains of quality, safety and effectiveness, patient experience and also show innovation. This article discusses how the care bundle was introduced and how, over a 12-month period, it showed tangible results in improving the care pathway for COPD patients as well as reducing readmissions and saving a significant amount of money.

  15. Emitters of N-photon bundles.

    Science.gov (United States)

    Muñoz, C Sánchez; Del Valle, E; Tudela, A González; Müller, K; Lichtmannecker, S; Kaniber, M; Tejedor, C; Finley, J J; Laussy, F P

    2014-07-01

    Controlling the ouput of a light emitter is one of the basic tasks of photonics, with landmarks such as the laser and single-photon sources. The development of quantum applications makes it increasingly important to diversify the available quantum sources. Here, we propose a cavity QED scheme to realize emitters that release their energy in groups, or "bundles" of N photons, for integer N. Close to 100% of two-photon emission and 90% of three-photon emission is shown to be within reach of state of the art samples. The emission can be tuned with system parameters so that the device behaves as a laser or as a N-photon gun. The theoretical formalism to characterize such emitters is developed, with the bundle statistics arising as an extension of the fundamental correlation functions of quantum optics. These emitters will be useful for quantum information processing and for medical applications.

  16. Client Provider Collaboration for Service Bundling

    Directory of Open Access Journals (Sweden)

    LETIA, I. A.

    2008-04-01

    Full Text Available The key requirement for a service industry organization to reach competitive advantages through product diversification is the existence of a well defined method for building service bundles. Based on the idea that the quality of a service or its value is given by the difference between expectations and perceptions, we draw the main components of a frame that aims to support the client and the provider agent in an active collaboration meant to co-create service bundles. Following e3-value model, we structure the supporting knowledge around the relation between needs and satisfying services. We deal with different perspectives about quality through an ontological extension of Value Based Argumentation. The dialog between the client and the provider takes the form of a persuasion whose dynamic object is the current best configuration. Our approach for building service packages is a demand driven approach, allowing progressive disclosure of private knowledge.

  17. Non-abelian higher gauge theory and categorical bundle

    Science.gov (United States)

    Viennot, David

    2016-12-01

    A gauge theory is associated with a principal bundle endowed with a connection permitting to define horizontal lifts of paths. The horizontal lifts of surfaces cannot be defined into a principal bundle structure. An higher gauge theory is an attempt to generalize the bundle structure in order to describe horizontal lifts of surfaces. A such attempt is particularly difficult for the non-abelian case. Some structures have been proposed to realize this goal (twisted bundle, gerbes with connection, bundle gerbe, 2-bundle). Each of them uses a category in place of the total space manifold of the usual principal bundle structure. Some of them replace also the structure group by a category (more precisely a Lie crossed module viewed as a category). But the base space remains still a simple manifold (possibly viewed as a trivial category with only identity arrows). We propose a new principal categorical bundle structure, with a Lie crossed module as structure groupoid, but with a base space belonging to a bigger class of categories (which includes non-trivial categories), that we called affine 2-spaces. We study the geometric structure of the categorical bundles built on these categories (which are a more complicated structure than the 2-bundles) and the connective structures on these bundles. Finally we treat an example interesting for quantum dynamics which is associated with the Bloch wave operator theory.

  18. Quantum principal bundles and their characteristic classes

    CERN Document Server

    Durdevic, M

    1996-01-01

    A brief exposition of the general theory of characteristic classes of quantum principal bundles is given. The theory of quantum characteristic classes incorporates ideas of classical Weil theory into the conceptual framework of non-commutative differential geometry. A purely cohomological interpretation of the Weil homomorphism is given, together with a standard geometrical interpretation via quantum invariant polynomials. A natural spectral sequence is described. Some quantum phenomena appearing in the formalism are discussed.

  19. Uncovering ecosystem service bundles through social preferences.

    Directory of Open Access Journals (Sweden)

    Berta Martín-López

    Full Text Available Ecosystem service assessments have increasingly been used to support environmental management policies, mainly based on biophysical and economic indicators. However, few studies have coped with the social-cultural dimension of ecosystem services, despite being considered a research priority. We examined how ecosystem service bundles and trade-offs emerge from diverging social preferences toward ecosystem services delivered by various types of ecosystems in Spain. We conducted 3,379 direct face-to-face questionnaires in eight different case study sites from 2007 to 2011. Overall, 90.5% of the sampled population recognized the ecosystem's capacity to deliver services. Formal studies, environmental behavior, and gender variables influenced the probability of people recognizing the ecosystem's capacity to provide services. The ecosystem services most frequently perceived by people were regulating services; of those, air purification held the greatest importance. However, statistical analysis showed that socio-cultural factors and the conservation management strategy of ecosystems (i.e., National Park, Natural Park, or a non-protected area have an effect on social preferences toward ecosystem services. Ecosystem service trade-offs and bundles were identified by analyzing social preferences through multivariate analysis (redundancy analysis and hierarchical cluster analysis. We found a clear trade-off among provisioning services (and recreational hunting versus regulating services and almost all cultural services. We identified three ecosystem service bundles associated with the conservation management strategy and the rural-urban gradient. We conclude that socio-cultural preferences toward ecosystem services can serve as a tool to identify relevant services for people, the factors underlying these social preferences, and emerging ecosystem service bundles and trade-offs.

  20. Uncontrolled inexact information within bundle methods

    OpenAIRE

    Malick, Jérôme; Welington De Oliveira, ·; Zaourar-Michel, Sofia

    2016-01-01

    International audience; We consider convex nonsmooth optimization problems where additional information with uncontrolled accuracy is readily available. It is often the case when the objective function is itself the output of an optimization solver, as for large-scale energy optimization problems tackled by decomposition. In this paper, we study how to incorporate the uncontrolled linearizations into (proximal and level) bundle algorithms in view of generating better iterates and possibly acc...

  1. On Complex Supermanifolds with Trivial Canonical Bundle

    CERN Document Server

    Groeger, Josua

    2016-01-01

    We give an algebraic characterisation for the triviality of the canonical bundle of a complex supermanifold in terms of a certain Batalin-Vilkovisky superalgebra structure. As an application, we study the Calabi-Yau case, in which an explicit formula in terms of the Levi-Civita connection is achieved. Our methods include the use of complex integral forms and the recently developed theory of superholonomy.

  2. Uncovering Ecosystem Service Bundles through Social Preferences

    Science.gov (United States)

    Martín-López, Berta; Iniesta-Arandia, Irene; García-Llorente, Marina; Palomo, Ignacio; Casado-Arzuaga, Izaskun; Amo, David García Del; Gómez-Baggethun, Erik; Oteros-Rozas, Elisa; Palacios-Agundez, Igone; Willaarts, Bárbara; González, José A.; Santos-Martín, Fernando; Onaindia, Miren; López-Santiago, Cesar; Montes, Carlos

    2012-01-01

    Ecosystem service assessments have increasingly been used to support environmental management policies, mainly based on biophysical and economic indicators. However, few studies have coped with the social-cultural dimension of ecosystem services, despite being considered a research priority. We examined how ecosystem service bundles and trade-offs emerge from diverging social preferences toward ecosystem services delivered by various types of ecosystems in Spain. We conducted 3,379 direct face-to-face questionnaires in eight different case study sites from 2007 to 2011. Overall, 90.5% of the sampled population recognized the ecosystem’s capacity to deliver services. Formal studies, environmental behavior, and gender variables influenced the probability of people recognizing the ecosystem’s capacity to provide services. The ecosystem services most frequently perceived by people were regulating services; of those, air purification held the greatest importance. However, statistical analysis showed that socio-cultural factors and the conservation management strategy of ecosystems (i.e., National Park, Natural Park, or a non-protected area) have an effect on social preferences toward ecosystem services. Ecosystem service trade-offs and bundles were identified by analyzing social preferences through multivariate analysis (redundancy analysis and hierarchical cluster analysis). We found a clear trade-off among provisioning services (and recreational hunting) versus regulating services and almost all cultural services. We identified three ecosystem service bundles associated with the conservation management strategy and the rural-urban gradient. We conclude that socio-cultural preferences toward ecosystem services can serve as a tool to identify relevant services for people, the factors underlying these social preferences, and emerging ecosystem service bundles and trade-offs. PMID:22720006

  3. Uncovering ecosystem service bundles through social preferences

    OpenAIRE

    Berta Martín-López; Irene Iniesta-Arandia; Marina García-Llorente; Ignacio Palomo; Izaskun Casado-Arzuaga; David García Del Amo; Erik Gómez-Baggethun; Elisa Oteros-Rozas; Igone Palacios-Agundez; Bárbara Willaarts; González, José A.; Fernando Santos-Martín; Miren Onaindia; Cesar López-Santiago; Carlos Montes

    2012-01-01

    11 p. Ecosystem service assessments have increasingly been used to support environmental management policies, mainly based on biophysical and economic indicators. However, few studies have coped with the social-cultural dimension of ecosystem services, despite being considered a research priority. We examined how ecosystem service bundles and trade-offs emerge from diverging social preferences toward ecosystem services delivered by various types of ecosystems in Spain. We conducted 3,379 d...

  4. Deformations of Fell bundles and twisted graph algebras

    Science.gov (United States)

    Raeburn, Iain

    2016-11-01

    We consider Fell bundles over discrete groups, and the C*-algebra which is universal for representations of the bundle. We define deformations of Fell bundles, which are new Fell bundles with the same underlying Banach bundle but with the multiplication deformed by a two-cocycle on the group. Every graph algebra can be viewed as the C*-algebra of a Fell bundle, and there are are many cocycles of interest with which to deform them. We thus obtain many of the twisted graph algebras of Kumjian, Pask and Sims. We demonstate the utility of our approach to these twisted graph algebras by proving that the deformations associated to different cocycles can be assembled as the fibres of a C*-bundle.

  5. Bundling harvester; Harvennuspuun automaattisen nippukorjausharvesterin kehittaeminen

    Energy Technology Data Exchange (ETDEWEB)

    Koponen, K. [Eko-Log Oy, Kuopio (Finland)

    1997-12-01

    The starting point of the project was to design and construct, by taking the silvicultural point of view into account, a harvesting and processing system especially for energy-wood, containing manually driven bundling harvester, automating of the harvester, and automated loading. The equipment forms an ideal method for entrepreneur`s-line harvesting. The target is to apply the system also for owner`s-line harvesting. The profitability of the system promotes the utilisation of the system in both cases. The objectives of the project were: to construct a test equipment and prototypes for all the project stages, to carry out terrain and strain tests in order to examine the usability and durability, as well as the capacity of the machine, to test the applicability of the Eko-Log system in simultaneous harvesting of energy and pulp woods, and to start the marketing and manufacturing of the products. The basic problems of the construction of the bundling harvester have been solved using terrain-tests. The prototype machine has been shown to be operable. Loading of the bundles to form sufficiently economically transportable loads has been studied, and simultaneously, the branch-biomass has been tried to be utilised without loosing the profitability of transportation. The results have been promising, and will promote the profitable utilisation of wood-energy. (orig.)

  6. Development of Strengthened Bundle High Temperature Superconductors

    Energy Technology Data Exchange (ETDEWEB)

    Lue, J.W.; Lubell, M.S. [Oak Ridge National Lab., TN (United States); Demko, J.A. [Oak Ridge Inst. for Science and Education, TN (United States); Tomsic, M. [Plastronic, Inc., Troy, OH (United States); Sinha, U. [Southwire Company, Carollton, GA (United States)

    1997-12-31

    In the process of developing high temperature superconducting (HTS) transmission cables, it was found that mechanical strength of the superconducting tape is the most crucial property that needs to be improved. It is also desirable to increase the current carrying capacity of the conductor so that fewer layers are needed to make the kilo-amp class cables required for electric utility usage. A process has been developed by encapsulating a stack of Bi-2223/Ag tapes with a silver or non-silver sheath to form a strengthened bundle superconductor. This process was applied to HTS tapes made by the Continuous Tube Forming and Filling (CTFF) technique pursued by Plastronic Inc. and HTS tapes obtained from other manufacturers. Conductors with a bundle of 2 to 6 HTS tapes have been made. The bundled conductor is greatly strengthened by the non-silver sheath. No superconductor degradation as compared to the sum of the original critical currents of the individual tapes was seen on the finished conductors.

  7. An analytical fiber bundle model for pullout mechanics of root bundles

    Science.gov (United States)

    Cohen, D.; Schwarz, M.; Or, D.

    2011-09-01

    Roots in soil contribute to the mechanical stability of slopes. Estimation of root reinforcement is challenging because roots form complex biological networks whose geometrical and mechanical characteristics are difficult to characterize. Here we describe an analytical model that builds on simple root descriptors to estimate root reinforcement. Root bundles are modeled as bundles of heterogeneous fibers pulled along their long axes neglecting root-soil friction. Analytical expressions for the pullout force as a function of displacement are derived. The maximum pullout force and corresponding critical displacement are either derived analytically or computed numerically. Key model inputs are a root diameter distribution (uniform, Weibull, or lognormal) and three empirical power law relations describing tensile strength, elastic modulus, and length of roots as functions of root diameter. When a root bundle with root tips anchored in the soil matrix is pulled by a rigid plate, a unique parameter, ?, that depends only on the exponents of the power law relations, dictates the order in which roots of different diameters break. If ? 1, large roots break first. When ? = 1, all fibers break simultaneously, and the maximum tensile force is simply the roots' mean force times the number of roots in the bundle. Based on measurements of root geometry and mechanical properties, the value of ? is less than 1, usually ranging between 0 and 0.7. Thus, small roots always fail first. The model shows how geometrical and mechanical characteristics of roots and root diameter distribution affect the pullout force, its maximum and corresponding displacement. Comparing bundles of roots that have similar mean diameters, a bundle with a narrow variance in root diameter will result in a larger maximum force and a smaller displacement at maximum force than a bundle with a wide diameter distribution. Increasing the mean root diameter of a bundle without changing the distribution's shape increases

  8. Modelling packing interactions in parallel helix bundles: pentameric bundles of nicotinic receptor M2 helices.

    Science.gov (United States)

    Sankararamakrishnan, R; Sansom, M S

    1995-11-01

    The transbilayer pore of the nicotinic acetylcholine receptor (nAChR) is formed by a pentameric bundle of M2 helices. Models of pentameric bundles of M2 helices have been generated using simulated annealing via restrained molecular dynamics. The influence of: (a) the initial C alpha template; and (b) screening of sidechain electrostatic interactions on the geometry of the resultant M2 helix bundles is explored. Parallel M2 helices, in the absence of sidechain electrostatic interactions, pack in accordance with simple ridges-in-grooves considerations. This results in a helix crossing angle of ca. +12 degrees, corresponding to a left-handed coiled coil structure for the bundle as a whole. Tilting of M2 helices away from the central pore axis at their C-termini and/or inclusion of sidechain electrostatic interactions may perturb such ridges-in-grooves packing. In the most extreme cases right-handed coiled coils are formed. An interplay between inter-helix H-bonding and helix bundle geometry is revealed. The effects of changes in electrostatic screening on the dimensions of the pore mouth are described and the significance of these changes in the context of models for the nAChR pore domain is discussed.

  9. In-pile tests at Karlsruhe of LWR fuel-rod behavior during the heatup phase of a LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Karb, E.H.

    1980-01-01

    In order to investigate the influence of a nuclar environment on the mechanisms of fuel-rod failure, in-pile tests simulating the heatup phase of a loss-of-coolant accident in a pressurized-water reactor are being conducted with irradiated and unirradiated short-length single rods in the FR2 reactor at Kernforschungszentrum karlsruhe (Karlsruhe Nuclear Reasearch Center), Federal Republic of Germany, within the Project Nuclear Safety. With nearly 70% of the scheduled tests completed, no such influences have been found. The in-pile burst and deformation data are in good agreement with results from nonnuclear tests with electrically heated fuel-rod simulators. The phenomenon of pellet disintegration, which has been observed in all tests with previously irradiated rods, needs further investigation.

  10. Investigation of the radiation risk due to environmental contamination by 241Am from lightning rods disposed at uncontrolled garbage dumps.

    Science.gov (United States)

    Marumo, Júlio T; Isiki, Vera L K; Miyamoto, Hissae; Ferreira, Rafael V P; Bellini, Maria H; de Lima, Luis F C P

    2008-02-01

    Radioactive lightning rods were manufactured in Brazil until 1989, when the licenses for using radioactive sources in these products were lifted by the national nuclear authority. Since then, these rods have been replaced by the Franklin type and collected as radioactive waste. However, only 20% of the estimated total number of installed rods has been delivered to the Brazilian Nuclear Commission. This situation causes concern, since there is the possibility of the rods to be disposed as domestic waste. In Brazil, 64% of the municipal solid waste is disposed at garbage dumps without sufficient control. In addition, (241)Am, the radionuclide most commonly employed, is classified as a high-toxicity element, when incorporated. In the present study, (241)Am migration experiments were performed by means of a lysimeter system, in order to evaluate the risk of contamination caused by radioactive lightning rods disposed as common solid waste. (241)Am sources removed from lightning rods were placed inside lysimeters filled with organic waste that was collected at the restaurant of the Instituto de Pesquisas Energéticas e Nucleares. The generated leachate was periodically analyzed, and characteristics such as pH, redox potential, solid content and the concentration of the radioactive material were determined. The equivalent dose for members of the public was calculated considering ingestion of contaminated drinking water as the major path of exposure. Estimated doses were about 20-times below the effective dose limit of 1 mSv year(-1) for members of the public as recommended by the International Commission on Radiological Protection. This suggests the radiation risk caused by lightning rods disposed at uncontrolled garbage dumps to be low. It should be noted, however, that the number of investigated lightning rods was quite small. The results of this study might therefore not be entirely representative and should be interpreted with care. They provide, however, a very first

  11. Viscoelasticity of suspensions of long, rigid rods

    NARCIS (Netherlands)

    Dhont, Jan K.G.; Briels, W.J.

    2003-01-01

    A microscopic theory for the viscoelastic behaviour of suspensions of rigid rods with excluded volume interactions is presented, which is valid in the asymptotic limit of very long and thin rods. Stresses arising from translational and rotational Brownian motion and direct interactions are calculate

  12. Study of the rod style SFRFQ structure

    CERN Document Server

    Yan Xue Qing; Chen J

    2002-01-01

    There is a problem about upper limit of energy in the RFQ structure, although it is a wonderful low-energy-suited high current accelerating structure. After proposing an improved rod style SFRFQ structure without reversed field, the author studies its energy gain and transverse motion. The rod style SFRFQ structure is roughly compared with diaphragm SFRFQ structure

  13. An Alternative Bundle-to-Bundle Suturing Technique for Repairing Fresh Achilles Tendon Rupture.

    Science.gov (United States)

    Zhao, Jingjing; Yu, Bin; Xie, Ming; Huang, Ruokun; Xiao, Kai

    2016-01-01

    The main concern about conventional Achilles tendon repair surgical techniques is how to maintain the initial strength of the ruptured Achilles tendon through complicated suturing methods. The primary surgical problem lies in the properties of the soft tissue; the deterioration of the Achilles tendon, especially in its elasticity; and the surface lubricity of the local tissues. In the present study, we describe an innovative bundle-to-bundle suturing method that addresses these potential problems. Copyright © 2016 American College of Foot and Ankle Surgeons. Published by Elsevier Inc. All rights reserved.

  14. Assessment of ASSERT-PV for prediction of critical heat flux in CANDU bundles

    Energy Technology Data Exchange (ETDEWEB)

    Rao, Y.F., E-mail: raoy@aecl.ca; Cheng, Z., E-mail: chengz@aecl.ca; Waddington, G.M., E-mail: waddingg@aecl.ca

    2014-09-15

    Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for CHF prediction. • CANDU 28-, 37- and 43-element bundle CHF experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of CHF model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against five full-scale CANDU bundle experiments conducted in 1990s and in 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element (CANFLEX) bundles. A total of 15 CHF test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for CANDU reactors. Code predictions of channel dryout power and axial and radial CHF locations were compared against measurements from the SL CHF tests to quantify the code prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, the sensitivity studies evaluated the contribution of each CHF model change or enhancement to the improvement in CHF prediction. Overall, the assessment demonstrated significant improvement in prediction of channel dryout power and axial and radial CHF locations in horizontal fuel channels containing CANDU bundles.

  15. Assessment of ASSERT-PV for prediction of post-dryout heat transfer in CANDU bundles

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, Z., E-mail: chengz@aecl.ca; Rao, Y.F., E-mail: raoy@aecl.ca; Waddington, G.M., E-mail: waddingg@aecl.ca

    2014-10-15

    Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for PDO sheath temperature prediction. • CANDU 28-, 37- and 43-element bundle PDO experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of PDO model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of subchannel flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against PDO tests performed during five full-size CANDU bundle experiments conducted between 1992 and 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element bundles. A total of 10 PDO test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for existing CANDU reactors. Code predictions of maximum PDO fuel-sheath temperature were compared against measurements from the SL PDO tests to quantify the code's prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, separate-effects sensitivity studies quantified the contribution of each PDO model change or enhancement to the improvement in PDO heat transfer prediction. Overall, the assessment demonstrated significant improvement in prediction of PDO sheath temperature in horizontal fuel channels containing CANDU bundles.

  16. Phase behavior and structure formation of hairy-rod supramolecules

    NARCIS (Netherlands)

    Subbotin, A; Stepanyan, R; Knaapila, M; Ikkala, O; ten Brinke, G

    2003-01-01

    Phase behavior and microstructure formation of rod and coil molecules, which can associate to form hairy-rod polymeric supramolecules, are addressed theoretically. Association induces considerable compatibility enhancement between the rod and coil molecules and various microscopically ordered struct

  17. Eulerian Formulation of Spatially Constrained Elastic Rods

    Science.gov (United States)

    Huynen, Alexandre

    Slender elastic rods are ubiquitous in nature and technology. For a vast majority of applications, the rod deflection is restricted by an external constraint and a significant part of the elastic body is in contact with a stiff constraining surface. The research work presented in this doctoral dissertation formulates a computational model for the solution of elastic rods constrained inside or around frictionless tube-like surfaces. The segmentation strategy adopted to cope with this complex class of problems consists in sequencing the global problem into, comparatively simpler, elementary problems either in continuous contact with the constraint or contact-free between their extremities. Within the conventional Lagrangian formulation of elastic rods, this approach is however associated with two major drawbacks. First, the boundary conditions specifying the locations of the rod centerline at both extremities of each elementary problem lead to the establishment of isoperimetric constraints, i.e., integral constraints on the unknown length of the rod. Second, the assessment of the unilateral contact condition requires, in principle, the comparison of two curves parametrized by distinct curvilinear coordinates, viz. the rod centerline and the constraint axis. Both conspire to burden the computations associated with the method. To streamline the solution along the elementary problems and rationalize the assessment of the unilateral contact condition, the rod governing equations are reformulated within the Eulerian framework of the constraint. The methodical exploration of both types of elementary problems leads to specific formulations of the rod governing equations that stress the profound connection between the mechanics of the rod and the geometry of the constraint surface. The proposed Eulerian reformulation, which restates the rod local equilibrium in terms of the curvilinear coordinate associated with the constraint axis, describes the rod deformed configuration

  18. Investigation of Minimum Film boiling Phenomena on Fuel Rods Under Blowdown Cooling Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Stephen M. Bajorek; Michael Gawron; Timothy Etzel; Lucas Peterson

    2003-06-30

    Blowdon cooling heat transfer is an important process that occurs early in a hypothetical large break loss-of-coolant accident (LOCA) in a pressurized water reactor. During blowdown, the flow through the hot assembly is a post-critical heat flux dispersed droplet flow. The heat transfer mechanisms that occur in blowdown cooling are complex and depend on droplet and heated surface interaction. In a safety analysis, it is of considerable importance to determine the thermal-hydraulic conditions leading to the minimum film boiling temperature, Tmin. A flow boiling rig for measurement of blowdown cooling heat transfer and quench phenomena on a nuclear fuel rod simulator was designed and constructed for operation at up to 12.4 MPa. The test section consisted of a concentric annulus, with a 9.5 mm OD nuclear fuel rod simulator at the center. The rod was contained within a 0.85 mm thick, 19 mm OD 316 stainless steel tube, forming the flow channel. Two types of rods were tested; one type was sheathed with Inconel 600 while the other was clad with Zircaloy-2. Water was injected into the test section at the top of the heated length through an injection header. This header was an annular sign that fit around the fuel rod simulator and within the stainless steel tube. Small spacers aligned the injection header and prevented contract with either the heater rod or the tube. A series of small diameter holes at the bottom of the header caused the formation of droplets that became entrained with the steam flow. The test section design was such that quench would take place on the rod, and not along the channel outer annulus.

  19. Anomalous water expulsion from carbon-based rods at high humidity

    Energy Technology Data Exchange (ETDEWEB)

    Nune, Satish K.; Lao, David B.; Heldebrant, David J.; Liu, Jian; Olszta, Matthew J.; Kukkadapu, Ravi K.; Gordon, Lyle M.; Nandasiri, Manjula I.; Whyatt, Greg; Clayton, Chris; Gotthold, David W.; Engelhard, Mark H.; Schaef, Herbert T.

    2016-06-13

    Managing water is critical for industrial applications including CO2 capture, catalysis, bio-oil separations and energy storage. Various classes of materials have been designed for these applications, achieving specific water adsorption capacities at a given relative humidity (RH). Three water adsorption-desorption mechanisms are common to inorganic materials: (1) chemisorption, which may lead to the modification of the first coordination sphere; (2) simple adsorption, which is reversible in nature; or (3) capillary condensation, which is irreversible in nature. Regardless of sorption mechanism, all materials known today increase water adsorption capacity with increasing RH; none exhibit repeated adsorption of water at low humidity and release at high humidity. We present here a material that breaks from this convention: a new class of nitrogen containing carbon rods along with nonstoichiometric FeXSY that adsorb water at low humidity, and spontaneously expel half of the adsorbed water when the RH exceeds a 50–80% threshold. Monolayers of water form on the surfaces of the carbon rods, with the amount of water adsorbed directly linked to the aspect ratio of the rods and the available surface area. This unprecedented water expulsion is a reversible physical process. Once a complete monolayer is formed, adjacent rods in the bundles begin to adhere together via formation of a bridging monolayer, reducing the surface area available for water to adhere to. We believe the unique surface chemistry of these carbon rods can be used on other functionalized materials. Such behaviour offers a paradigm shift in water purification and separation: water could be repeatedly adsorbed from a low humidity vapour stream and then expelled into a pure water vapour stream, or humidity-responsive membranes could change their water permeance or selectivity as a function of RH.

  20. Amplitude death of coupled hair bundles with stochastic channel noise

    CERN Document Server

    Kim, Kyung-Joong

    2014-01-01

    Hair cells conduct auditory transduction in vertebrates. In lower vertebrates such as frogs and turtles, due to the active mechanism in hair cells, hair bundles(stereocilia) can be spontaneously oscillating or quiescent. Recently, the amplitude death phenomenon has been proposed [K.-H. Ahn, J. R. Soc. Interface, {\\bf 10}, 20130525 (2013)] as a mechanism for auditory transduction in frog hair-cell bundles, where sudden cessation of the oscillations arises due to the coupling between non-identical hair bundles. The gating of the ion channel is intrinsically stochastic due to the stochastic nature of the configuration change of the channel. The strength of the noise due to the channel gating can be comparable to the thermal Brownian noise of hair bundles. Thus, we perform stochastic simulations of the elastically coupled hair bundles. In spite of stray noisy fluctuations due to its stochastic dynamics, our simulation shows the transition from collective oscillation to amplitude death as inter-bundle coupling str...

  1. Monopoles and Modifications of Bundles over Elliptic Curves

    Directory of Open Access Journals (Sweden)

    Andrey M. Levin

    2009-06-01

    Full Text Available Modifications of bundles over complex curves is an operation that allows one to construct a new bundle from a given one. Modifications can change a topological type of bundle. We describe the topological type in terms of the characteristic classes of the bundle. Being applied to the Higgs bundles modifications establish an equivalence between different classical integrable systems. Following Kapustin and Witten we define the modifications in terms of monopole solutions of the Bogomolny equation. We find the Dirac monopole solution in the case R × (elliptic curve. This solution is a three-dimensional generalization of the Kronecker series. We give two representations for this solution and derive a functional equation for it generalizing the Kronecker results. We use it to define Abelian modifications for bundles of arbitrary rank. We also describe non-Abelian modifications in terms of theta-functions with characteristic.

  2. A Tannakian approach to dimensional reduction of principal bundles

    CERN Document Server

    Álvarez-Cónsul, Luis; García-Prada, Oscar

    2016-01-01

    Let $P$ be a parabolic subgroup of a connected simply connected complex semisimple Lie group $G$. Given a compact K\\"ahler manifold $X$, the dimensional reduction of $G$-equivariant holomorphic vector bundles over $X\\times G/P$ was carried out by the first and third authors. This raises the question of dimensional reduction of holomorphic principal bundles over $X\\times G/P$. The method used for equivariant vector bundles does not generalize to principal bundles. In this paper, we adapt to equivariant principal bundles the Tannakian approach of Nori, to describe the dimensional reduction of $G$-equivariant principal bundles over $X\\times G/P$, and to establish a Hitchin--Kobayashi type correspondence. In order to be able to apply the Tannakian theory, we need to assume that $X$ is a complex projective manifold.

  3. The avalanche process of the fiber bundle model with defect

    Science.gov (United States)

    Hao, Da-Peng; Tang, Gang; Xia, Hui; Xun, Zhi-Peng; Han, Kui

    2017-04-01

    In order to explore the impacts of defect on the tensile fracture process of materials, the fiber bundle model with defect is constructed based on the classical fiber bundle model. In the fiber bundle model with defect, the two key parameters are the mean size and the density of defects. In both uniform and Weibull threshold distributions, the mean size and density all bring impacts on the threshold distribution of fibers. By means of analytical approximation and numerical simulation, we show that the two key parameters of the model have substantial effects on the failure process of the bundle. From macroscopic view, the defect described by the altering of threshold distribution of fibers will has a significant impact on the mechanical properties of the bundle. While in microscopic scale, the statistical properties of the model are still harmonious with the classical fiber bundle model.

  4. Comparison study of the thermal mechanical performance of fuel rods during BWR fuel preconditioning operations using the computer codes FUELSIM and FEMAXI-V

    Energy Technology Data Exchange (ETDEWEB)

    Pantoja C, R. [IPN, Escuela Superior de Fisica y Matematicas, Departamento de Ingenieria Nuclear, Av. Instituto Politecnico Nacional s/n, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Ortiz V, J.; Castillo D, R., E-mail: rafael.pantoja10@yahoo.com.m [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, Ocoyoacac 52750, Estado de Mexico (Mexico)

    2010-10-15

    The safety of nuclear power plants requires monitoring those parameters having some direct or indirect effect on safety. The thermal limits are values set for those parameters considered having most impact on the safe operation of a nuclear power reactor. Some thermal limits monitoring requires the thermal-mechanical analysis of the rods containing the nuclear fuel. The fuel rod thermal-mechanical behaviour under irradiation is a complex process in which there exists a great deal of interrelated physical and chemical phenomena, so that the fuel rod performance analysis in the core of a nuclear power reactor is generally accomplished by using computer codes, which integrate several of the phenomena that are expected to occur during the lifetime of the fuel rod in the core. In the operation of a nuclear power reactor, pre-conditioning simulations are necessary to determine in advance limit values for the power that can be generated in a fuel rod during any power ramp, and mainly during reactor startup, and thus avoiding any rod damage. In this work, a first analysis of the thermal-mechanical performance of typical fuel rods used in nuclear reactors of the type BWR is performed. This study includes two types of fuel rods: one from a fuel assembly design with array 8 x 8, and the other one from a 10 x 10 fuel assembly design, and a comparison of the thermal-mechanical performance between the two different rod designs is performed. The performance simulations were performed by the code FUELSIM, and compared against results previously obtained from similar simulation with the code FEMAXI-V. (Author)

  5. Preliminary Content Evaluation of the North Anna High Burn-Up Sister Fuel Rod Segments for Transportation in the 10-160B and NAC-LWT

    Energy Technology Data Exchange (ETDEWEB)

    Ketusky, E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-08-09

    The U.S. Department of Energy’s (DOE’s) Used Fuel Disposition Campaign (UFDC) Program has transported high-burnup nuclear sister fuel rods from a commercial nuclear power plant for purposes of evaluation and testing. The evaluation and testing of high-burnup used nuclear fuel is integral to DOE initiatives to collect information useful in determining the integrity of fuel cladding for future safe transportation of the fuel, and for determining the effects of aging, on the integrity of UNF subjected to extended storage and subsequent transportation. The UFDC Program, in collaboration with the U.S. Nuclear Regulatory Commission and the commercial nuclear industry, has obtained individual used nuclear fuel rods for testing. The rods have been received at Oak Ridge National Laboratory (ORNL) for both separate effects testing (SET) and small-scale testing (SST). To meet the research objectives, testing on multiple 6 inch fuel rod pins cut from the rods at ORNL will be performed at Pacific Northwest National Laboratory (PNNL). Up to 10 rod equivalents will be shipped. Options were evaluated for multiple shipments using the 10-160B (based on 4.5 rod equivalents) and a single shipment using the NAC-LWT. Based on the original INL/Virginia Power transfer agreement, the rods are assumed to 152 inches in length with a 0.374-inch diameter. This report provides a preliminary content evaluation for use of the 10-160B and NAC-LWT for transporting those fuel rod pins from ORNL to PNNL. This report documents the acceptability of using these packagings to transport the fuel segments from ORNL to PNNL based on the following evaluations: enrichment, A2 evaluation, Pu-239 FGE evaluation, heat load, shielding (both gamma and neutron), and content weight/structural evaluation.

  6. Cosmic multimuon bundles detected by DELPHI

    CERN Document Server

    Rídky, J

    2004-01-01

    The DELPHI detector located at LEP accelerator has been used also to measure multimuon bundles originated from cosmic ray interactions. Two subdetectors-hadron calorimeter and time projection chamber, are used for this purpose. The 1999 and 2000 data are analyzed over wide range of multiplicities. The multiplicity distribution is compared with prediction of Monte Carlo simulation based on CORSIKA/QGSJET. The Monte-Carlo does not describe the large multiplicity part of data. Even the extreme assumption on the cosmic ray composition (pure iron nuclei) hardly predicts comparable number of high-multiplicity events.

  7. Differential geometry of complex vector bundles

    CERN Document Server

    Kobayashi, Shoshichi

    2014-01-01

    Holomorphic vector bundles have become objects of interest not only to algebraic and differential geometers and complex analysts but also to low dimensional topologists and mathematical physicists working on gauge theory. This book, which grew out of the author's lectures and seminars in Berkeley and Japan, is written for researchers and graduate students in these various fields of mathematics. Originally published in 1987. The Princeton Legacy Library uses the latest print-on-demand technology to again make available previously out-of-print books from the distinguished backlist of Princeto

  8. Higher order mechanics on graded bundles

    Science.gov (United States)

    Bruce, Andrew James; Grabowska, Katarzyna; Grabowski, Janusz

    2015-05-01

    In this paper we develop a geometric approach to higher order mechanics on graded bundles in both, the Lagrangian and Hamiltonian formalism, via the recently discovered weighted algebroids. We present the corresponding Tulczyjew triple for this higher order situation and derive in this framework the phase equations from an arbitrary (also singular) Lagrangian or Hamiltonian, as well as the Euler-Lagrange equations. As important examples, we geometrically derive the classical higher order Euler-Lagrange equations and analogous reduced equations for invariant higher order Lagrangians on Lie groupoids.

  9. Bundling Products and Services Through Modularization Strategies

    DEFF Research Database (Denmark)

    Bask, Anu; Hsuan, Juliana; Rajahonka, Mervi;

    2012-01-01

    Modularity has been recognized as a powerful tool in improving the efficiency and management of product design and manufacturing. However, the integrated view on covering both, product and service modularity for product-service systems (PSS), is under researched. Therefore, in this paper our...... objective is to contribute to the PSS modularity. Thus, we describe configurations of PSSs and the bundling of products and services through modularization strategies. So far there have not been tools to analyze and determine the correct combinations of degrees of product and service modularities....

  10. Compression of a bundle of light rays.

    Science.gov (United States)

    Marcuse, D

    1971-03-01

    The performance of ray compression devices is discussed on the basis of a phase space treatment using Liouville's theorem. It is concluded that the area in phase space of the input bundle of rays is determined solely by the required compression ratio and possible limitations on the maximum ray angle at the output of the device. The efficiency of tapers and lenses as ray compressors is approximately equal. For linear tapers and lenses the input angle of the useful rays must not exceed the compression ratio. The performance of linear tapers and lenses is compared to a particular ray compressor using a graded refractive index distribution.

  11. Vector bundles on complex projective spaces

    CERN Document Server

    Okonek, Christian; Spindler, Heinz

    1980-01-01

    This expository treatment is based on a survey given by one of the authors at the Séminaire Bourbaki in November 1978 and on a subsequent course held at the University of Göttingen. It is intended to serve as an introduction to the topical question of classification of holomorphic vector bundles on complex projective spaces, and can easily be read by students with a basic knowledge of analytic or algebraic geometry. Short supplementary sections describe more advanced topics, further results, and unsolved problems.

  12. Morphoelastic rods Part II: Growing birods

    Science.gov (United States)

    Lessinnes, Thomas; Moulton, Derek E.; Goriely, Alain

    2017-03-01

    The general problem of determining the shape and response of two attached growing elastic Kirchhoff rods is considered. A description of the kinematics of the individual interacting rods is introduced. Each rod has a given intrinsic shape and constitutive laws, and a map associating points on the two rods is defined. The resulting filamentary structure, a growing birod, can be seen as a new filamentary structure. This kinematic description is used to derive the general equilibrium equations for the shape of the rods under loads, or equivalently, for the new birod. It is shown that, in general, the birod is not simply a Kirchhoff rod but rather, due to the internal constraints, new effects can appear. The two-dimensional restriction is then considered explicitly and the limit for small deformation is shown to be equivalent to the classic Timsohenko bi-metallic strip problem. A number of examples and applications are presented. In particular, the problem of two attached rods with intrinsic helical shape and uniform growth is computed in detail and a host of new interesting solutions and bifurcations are observed.

  13. Granular materials interacting with thin flexible rods

    Science.gov (United States)

    Neto, Alfredo Gay; Campello, Eduardo M. B.

    2017-04-01

    In this work, we develop a computational model for the simulation of problems wherein granular materials interact with thin flexible rods. We treat granular materials as a collection of spherical particles following a discrete element method (DEM) approach, while flexible rods are described by a large deformation finite element (FEM) rod formulation. Grain-to-grain, grain-to-rod, and rod-to-rod contacts are fully permitted and resolved. A simple and efficient strategy is proposed for coupling the motion of the two types (discrete and continuum) of materials within an iterative time-stepping solution scheme. Implementation details are shown and discussed. Validity and applicability of the model are assessed by means of a few numerical examples. We believe that robust, efficiently coupled DEM-FEM schemes can be a useful tool to the simulation of problems wherein granular materials interact with thin flexible rods, such as (but not limited to) bombardment of grains on beam structures, flow of granular materials over surfaces covered by threads of hair in many biological processes, flow of grains through filters and strainers in various industrial segregation processes, and many others.

  14. Magnetically controlled growing rods for scoliosis surgery.

    Science.gov (United States)

    Metkar, Umesh; Kurra, Swamy; Quinzi, David; Albanese, Stephen; Lavelle, William F

    2017-02-01

    Early onset scoliosis can be both a disfiguring as well as a life threatening condition. When more conservative treatments fail, pediatric spinal surgeons are forced to consider operative interventions. Traditionally, these interventions have involved the insertion of a variety of implants into the patient with a limited number of anchor points controlling the spine. In the past, these pediatric patients have had multiple surgeries for elective lengthening of these devices to facilitate their growth while attempting to control the scoliosis. These patients often experience a physical and emotional toll from their multiple repeated surgeries. Growing spine techniques have also had a noted high complication rate due to implant dislodgement and infections. Recently, the development of non-invasively, self-lengthening growing rods has occurred. These devices have the potential to allow for the devices to be lengthened magnetically in a conscious patient in the surgeon's office. Areas covered: This review summarized previously published articles in the English literature using a key word search in PubMed for: 'magnetically controlled growing rods', 'Magec rods', 'magnetic growing rods' and 'growing rods'. Expert commentary: Magnetically controlled growing rods have an advantage over growing rods in lengthening the growing spine in the absence of repetitive surgeries.

  15. Control Rod Malfunction at the NRAD Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Thomas L. Maddock

    2010-05-01

    The neutron Radiography Reactor (NRAD) is a training, research, and isotope (TRIGA) reactor located at the INL. The reactor is normally shut down by the insertion of three control rods that drop into the core when power is removed from electromagnets. During a routine shutdown, indicator lights on the console showed that one of the control rods was not inserted. It was initially thought that the indicator lights were in error because of a limit switch that was out of adjustment. Through further testing, it was determined that the control rod did not drop when the scram switch was initially pressed. The control rod anomaly led to a six month shutdown of the reactor and an in depth investigation of the reactor protective system. The investigation looked into: scram switch operation, console modifications, and control rod drive mechanisms. A number of latent issues were discovered and corrected during the investigation. The cause of the control rod malfunction was found to be a buildup of corrosion in the control rod drive mechanism. The investigation resulted in modifications to equipment, changes to both operation and maintenance procedures, and additional training. No reoccurrences of the problem have been observed since corrective actions were implemented.

  16. Estimation of irradiated control rod worth

    Energy Technology Data Exchange (ETDEWEB)

    Varvayanni, M., E-mail: melina@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PO Box 60228, 15310 Aghia Paraskevi (Greece); Catsaros, N. [NCSR ' DEMOKRITOS' , PO Box 60228, 15310 Aghia Paraskevi (Greece); Antonopoulos-Domis, M. [School of Electrical and Computer Engineering, Aristotle University of Thessaloniki, Thessaloniki (Greece)

    2009-11-15

    When depleted control rods are planned to be used in new core configurations, their worth has to be accurately predicted in order to deduce key design and safety parameters such as the available shutdown margin. In this work a methodology is suggested for the derivation of the distributed absorbing capacity of a depleted rod, useful in the case that the level of detail that is known about the irradiation history of the control rod does not allow an accurate calculation of the absorber's burnup. The suggested methodology is based on measurements of the rod's worth carried out in the former core configuration and on corresponding calculations based on the original (before first irradiation) absorber concentration. The methodology is formulated for the general case of the multi-group theory; it is successfully tested for the one-group approximation, for a depleted control rod of the Greek Research Reactor, containing five neutron absorbers. The computations reproduce satisfactorily the irradiated rod worth measurements, practically eliminating the discrepancy of the total rod worth, compared to the computations based on the nominal absorber densities.

  17. Granular materials interacting with thin flexible rods

    Science.gov (United States)

    Neto, Alfredo Gay; Campello, Eduardo M. B.

    2016-01-01

    In this work, we develop a computational model for the simulation of problems wherein granular materials interact with thin flexible rods. We treat granular materials as a collection of spherical particles following a discrete element method (DEM) approach, while flexible rods are described by a large deformation finite element (FEM) rod formulation. Grain-to-grain, grain-to-rod, and rod-to-rod contacts are fully permitted and resolved. A simple and efficient strategy is proposed for coupling the motion of the two types (discrete and continuum) of materials within an iterative time-stepping solution scheme. Implementation details are shown and discussed. Validity and applicability of the model are assessed by means of a few numerical examples. We believe that robust, efficiently coupled DEM-FEM schemes can be a useful tool to the simulation of problems wherein granular materials interact with thin flexible rods, such as (but not limited to) bombardment of grains on beam structures, flow of granular materials over surfaces covered by threads of hair in many biological processes, flow of grains through filters and strainers in various industrial segregation processes, and many others.

  18. Heat transfer in bundles of finned tubes in crossflow

    Energy Technology Data Exchange (ETDEWEB)

    Stasiulevicius, J.; Skrinska, A.; Zukauskas, A.; Hewitt, G.F.

    1986-01-01

    This book provides correlations of heat transfer and hydraulic data for bundles of finned tubes in crossflow at high Reynolds numbers. Results of studies of the effectiveness of the fin, local, and mean heat transfer coefficients are presented. The effect of geometric parameters of the fins and of the location of tubes in the bundle on heat transfer and hydraulic drag are described. The resistance of the finned tube bundles under study and other factors are examined.

  19. Heat Transfer Analysis in Wire Bundles for Aerospace Vehicles

    Science.gov (United States)

    Rickman, S. L.; Iamello, C. J.

    2016-01-01

    Design of wiring for aerospace vehicles relies on an understanding of "ampacity" which refers to the current carrying capacity of wires, either, individually or in wire bundles. Designers rely on standards to derate allowable current flow to prevent exceedance of wire temperature limits due to resistive heat dissipation within the wires or wire bundles. These standards often add considerable margin and are based on empirical data. Commercial providers are taking an aggressive approach to wire sizing which challenges the conventional wisdom of the established standards. Thermal modelling of wire bundles may offer significant mass reduction in a system if the technique can be generalized to produce reliable temperature predictions for arbitrary bundle configurations. Thermal analysis has been applied to the problem of wire bundles wherein any or all of the wires within the bundle may carry current. Wire bundles present analytical challenges because the heat transfer path from conductors internal to the bundle is tortuous, relying on internal radiation and thermal interface conductance to move the heat from within the bundle to the external jacket where it can be carried away by convective and radiative heat transfer. The problem is further complicated by the dependence of wire electrical resistivity on temperature. Reduced heat transfer out of the bundle leads to higher conductor temperatures and, hence, increased resistive heat dissipation. Development of a generalized wire bundle thermal model is presented and compared with test data. The steady state heat balance for a single wire is derived and extended to the bundle configuration. The generalized model includes the effects of temperature varying resistance, internal radiation and thermal interface conductance, external radiation and temperature varying convective relief from the free surface. The sensitivity of the response to uncertainties in key model parameters is explored using Monte Carlo analysis.

  20. Isothermal microcalorimetry, a tool for probing SWNT bundles.

    Science.gov (United States)

    Marquis, Renaud; Greco, Carla; Schultz, Patrick; Meunier, Stéphane; Mioskowski, Charles

    2009-11-01

    The bundling state of several dry single-walled carbon nanotube (SWNT) samples is compared using isothermal microcalorimetry (IMC). So as to get different dry samples with various bundling states, the pristine SWNTs were pretreated with a solution of an aromatic amphiphile with or without sonication, washed and dried before being studied by IMC. The bundling state of the different SWNT samples, which was first analyzed by TEM, was then correlated to the obtained IMC data thanks to the interpretation of the observed energy transfer phenomena. From our results, IMC appears to be an interesting technique for the surface probing of dry SWNT samples, and herein for the evaluation of the bundling state.

  1. Restriction Theorem for Principal bundles in Arbitrary Characteristic

    DEFF Research Database (Denmark)

    Gurjar, Sudarshan

    2015-01-01

    The aim of this paper is to prove two basic restriction theorem for principal bundles on smooth projective varieties in arbitrary characteristic generalizing the analogues theorems of Mehta-Ramanathan for vector bundles. More precisely, let G be a reductive algebraic group over an algebraically...... closed field k and let X be a smooth, projective variety over k together with a very ample line bundle O(1). The main result of the paper is that if E is a semistable (resp. stable) principal G-bundle on X w.r.t O(1), then the restriction of E to a general, high multi-degree, complete-intersection curve...

  2. Enthalpy and void distributions in subchannels of PHWR fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. W.; Choi, H.; Rhee, B. W. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    Two different types of the CANDU fuel bundles have been modeled for the ASSERT-IV code subchannel analysis. From calculated values of mixture enthalpy and void fraction distribution in the fuel bundles, it is found that net buoyancy effect is pronounced in the central region of the DUPIC fuel bundle when compared with the standard CANDU fuel bundle. It is also found that the central region of the DUPIC fuel bundle can be cooled more efficiently than that of the standard fuel bundle. From the calculated mixture enthalpy distribution at the exit of the fuel channel, it is found that the mixture enthalpy and void fraction can be highest in the peripheral region of the DUPIC fuel bundle. On the other hand, the enthalpy and the void fraction were found to be highest in the central region of the standard CANDU fuel bundle at the exit of the fuel channel. This study shows that the subchannel analysis is very useful in assessing thermal behavior of the fuel bundle that could be used in CANDU reactors. 10 refs., 4 figs., 2 tabs. (Author)

  3. Tipping time of a quantum rod

    Energy Technology Data Exchange (ETDEWEB)

    Parrikar, Onkar [Birla Institute of Technology and Science-Pilani, Goa campus, Zuarinagar, Goa 4032726 (India)], E-mail: onkarsp@gmail.com

    2010-03-15

    The behaviour of a quantum rod, pivoted at its lower end on an impenetrable floor and restricted to moving in the vertical plane under the gravitational potential, is studied analytically under the approximation that the rod is initially localized to a 'small-enough' neighbourhood around the point of classical unstable equilibrium. It is shown that the rod evolves out of this neighbourhood. The time required for this to happen, i.e. the tipping time, is calculated using the semi-classical path integral. It is shown that equilibrium is recovered in the classical limit, and that our calculations are consistent with the uncertainty principle.

  4. Heat flux and temperature determination on the control rod outer surface

    Energy Technology Data Exchange (ETDEWEB)

    Taler, J.; Cebula, A. [Cracow Univ. of Tech., Cracow (Poland); Marcinkiewicz, J.; Tinoco, H. [Forsmarks Kraftgrupp AB, Osthammar (Sweden)

    2011-07-01

    The paper presents heat transfer calculation results concerning a control rod of Unit 3 of Forsmark Nuclear Power Plant (NPP). The part of the control rod, which is the object of interest, is surrounded by a mixing region of hot and cold flows and, as a consequence, is subjected to thermal fluctuations. The paper describes a numerical test which validates the method based on the solution of the inverse heat conduction problem (IHCP). The comparison of the results achieved by two methods, CFD and IHCP, including a description of the IHCP method used in the calculation process, shows a very good agreement between the methods. (author)

  5. Verification of heat flux and temperature calculation on the control rod outer surface

    Science.gov (United States)

    Taler, Jan; Cebula, Artur

    2011-12-01

    The paper presents heat transfer calculation results concerning a control rod of Forsmark Nuclear Power Plant (NPP). The part of the control rod, which is the object of interest, is surrounded by a mixing region of hot and cold flows and, as a consequence, is subjected to thermal fluctuations. The paper describes a numerical test which validates the method based on the solution of the inverse heat conduction problem (IHCP). The comparison of the results achieved by two methods, computational fluid dynamics (CFD) simulations and IHCP, including a description of the IHCP method used in the calculation process, shows a very good agreement between the methods.

  6. Bundled capillary electrophoresis using microstructured fibres.

    Science.gov (United States)

    Rogers, Benjamin; Gibson, Graham T T; Oleschuk, Richard D

    2011-01-01

    Joule heating, arising from the electric current passing through the capillary, causes many undesired effects in CE that ultimately result in band broadening. The use of narrow-bore capillaries helps to solve this problem as smaller cross-sectional area results in decreased Joule heating and the rate of heat dissipation is increased by the larger surface-to-volume ratio. Issues arising from such small capillaries, such as poor detection sensitivity, low loading capacity and high flow-induced backpressure (complicating capillary loading) can be avoided by using a bundle of small capillaries operating simultaneously that share buffer reservoirs. Microstructured fibres, originally designed as waveguides in the telecommunication industry, are essentially a bundle of parallel ∼5 μm id channels that extend the length of a fibre having otherwise similar dimensions to conventional CE capillaries. This work presents the use of microstructured fibres for CZE, taking advantage of their relatively high surface-to-volume ratio and the small individual size of each channel to effect highly efficient separations, particularly for dye-labelled peptides.

  7. Development boiling to sprinkled tube bundle

    Directory of Open Access Journals (Sweden)

    Kracík Petr

    2016-01-01

    Full Text Available This paper presents results of a studied heat transfer coefficient at the surface of a sprinkled tube bundle where boiling occurs. Research in the area of sprinkled exchangers can be divided into two major parts. The first part is research on heat transfer and determination of the heat transfer coefficient at sprinkled tube bundles for various liquids, whether boiling or not. The second part is testing of sprinkle modes for various tube diameters, tube pitches and tube materials and determination of individual modes’ interface. All results published so far for water as the falling film liquid apply to one to three tubes for which the mentioned relations studied are determined in rigid laboratory conditions defined strictly in advance. The sprinkled tubes were not viewed from the operational perspective where there are more tubes and various modes may occur in different parts with various heat transfer values. The article focuses on these processes. The tube is located in a low-pressure chamber where vacuum is generated using an exhauster via ejector. The tube consists of smooth copper tubes of 12 mm diameter placed horizontally one above another.

  8. Development boiling to sprinkled tube bundle

    Science.gov (United States)

    Kracík, Petr; Pospíšil, Jiří

    2016-03-01

    This paper presents results of a studied heat transfer coefficient at the surface of a sprinkled tube bundle where boiling occurs. Research in the area of sprinkled exchangers can be divided into two major parts. The first part is research on heat transfer and determination of the heat transfer coefficient at sprinkled tube bundles for various liquids, whether boiling or not. The second part is testing of sprinkle modes for various tube diameters, tube pitches and tube materials and determination of individual modes' interface. All results published so far for water as the falling film liquid apply to one to three tubes for which the mentioned relations studied are determined in rigid laboratory conditions defined strictly in advance. The sprinkled tubes were not viewed from the operational perspective where there are more tubes and various modes may occur in different parts with various heat transfer values. The article focuses on these processes. The tube is located in a low-pressure chamber where vacuum is generated using an exhauster via ejector. The tube consists of smooth copper tubes of 12 mm diameter placed horizontally one above another.

  9. Fascin regulates nuclear actin during Drosophila oogenesis.

    Science.gov (United States)

    Kelpsch, Daniel J; Groen, Christopher M; Fagan, Tiffany N; Sudhir, Sweta; Tootle, Tina L

    2016-10-01

    Drosophila oogenesis provides a developmental system with which to study nuclear actin. During Stages 5-9, nuclear actin levels are high in the oocyte and exhibit variation within the nurse cells. Cofilin and Profilin, which regulate the nuclear import and export of actin, also localize to the nuclei. Expression of GFP-tagged Actin results in nuclear actin rod formation. These findings indicate that nuclear actin must be tightly regulated during oogenesis. One factor mediating this regulation is Fascin. Overexpression of Fascin enhances nuclear GFP-Actin rod formation, and Fascin colocalizes with the rods. Loss of Fascin reduces, whereas overexpression of Fascin increases, the frequency of nurse cells with high levels of nuclear actin, but neither alters the overall nuclear level of actin within the ovary. These data suggest that Fascin regulates the ability of specific cells to accumulate nuclear actin. Evidence indicates that Fascin positively regulates nuclear actin through Cofilin. Loss of Fascin results in decreased nuclear Cofilin. In addition, Fascin and Cofilin genetically interact, as double heterozygotes exhibit a reduction in the number of nurse cells with high nuclear actin levels. These findings are likely applicable beyond Drosophila follicle development, as the localization and functions of Fascin and the mechanisms regulating nuclear actin are widely conserved.

  10. Nuclear power demonstrating

    Energy Technology Data Exchange (ETDEWEB)

    Basmajian, V. V.; Haldeman, C. W.

    1980-08-12

    Apparatus for demonstrating the operation of a closed loop nuclear steam electric generating plant includes a transparent boiler assembly having immersion heating elements, which may be quartz lamps or stainless steel encased resistive immersion heating units with a quartz iodide lamp providing a source of visible radiation when using the encased immersion heating units. A variable voltage autotransformer is geared to a support rod for simulated reactor control rods for controlling the energy delivered to the heating elements and arranged so that when the voltage is high, the rods are withdrawn from the boiler to produce increased heating and illumination proportional to rod position, thereby simulating nuclear reaction. A relief valve, steam outlet pipe and water inlet pipe are connected to the boiler with a small stainless steel resistive heating element in the steam outlet pipe providing superheat. This heater is connected in series with a rheostat mounted on the front panel to provide superheat adjustments and an interlock switch that prevents the superheater from being energized when the steam valve is off with with no flow through the superheater. A heavy blue plastic radiation shield surrounds the boiler inside a bell jar.

  11. Nuclear Energy in Southeast Asia: Pull Rods or Scram

    Science.gov (United States)

    2009-06-01

    Statement of the Director General at Nobel Laureate Public Lecture, Kuala Lumpur , July 2007, http://www.iaea.org/NewsCenter/Statements/2007... Kuala Lumpur July 2007. http://www.iaea.org/NewsCenter/Statements/2007/ ebsp2007n011.html (accessed April 3, 2009). Gourevitch, Peter A. Politics in...Thailand’s political turmoil has crossed over into the economic sphere and eroded investor confidence. Even the tourism industry, one of Thailand main

  12. Evaluation of Single-Bundle versus Double-Bundle PCL Reconstructions with More Than 10-Year Follow-Up

    Directory of Open Access Journals (Sweden)

    Masataka Deie

    2015-01-01

    Full Text Available Background. Posterior cruciate ligament (PCL injuries are not rare in acute knee injuries, and several recent anatomical studies of the PCL and reconstructive surgical techniques have generated improved patient results. Now, we have evaluated PCL reconstructions performed by either the single-bundle or double-bundle technique in a patient group followed up retrospectively for more than 10 years. Methods. PCL reconstructions were conducted using the single-bundle (27 cases or double-bundle (13 cases method from 1999 to 2002. The mean age at surgery was 34 years in the single-bundle group and 32 years in the double-bundle group. The mean follow-up period was 12.5 years. Patients were evaluated by Lysholm scoring, the gravity sag view, and knee arthrometry. Results. The Lysholm score after surgery was 89.1±5.6 points for the single-bundle group and 91.9±4.5 points for the double-bundle group. There was no significant difference between the methods in the side-to-side differences by gravity sag view or knee arthrometer evaluation, although several cases in both groups showed a side-to-side difference exceeding 5 mm by the latter evaluation method. Conclusions. We found no significant difference between single- and double-bundle PCL reconstructions during more than 10 years of follow-up.

  13. Safety assessment for the CANFLEX-NU fuel bundles with respect to the 37-element fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Suk, H. C.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-11-01

    The KAERI and AECL have jointly developed an advanced CANDU fuel, called CANFLEX-NU fuel bundle. CANFLEX 43-element bundle has some improved features of increased operating margin and enhanced safety compared to the existing 37-element bundle. Since CANFLEX fuel bundle is designed to be compatible with the CANDU-6 reactor design, the behaviour in the thermalhydraulic system will be nearly identical with 37-element bundle. But due to different element design and linear element power distribution between the two bundles, it is expected that CANFLEX fuel behaviour would be different from the behaviour of the 37-element fuel. Therefore, safety assessments on the design basis accidents which result if fuel failures are performed. For all accidents selected, it is observed that the loading of CANFLEX bundle in an existing CANDU-6 reactor would not worsen the reactor safety. It is also predicted that fission product release for CANFLEX fuel bundle generally is lower than that for 37-element bundle. 3 refs., 2 figs., 2 tabs. (Author)

  14. Application of CFD Codes in Nuclear Reactor Safety Analysis

    Directory of Open Access Journals (Sweden)

    T. Höhne

    2010-01-01

    Full Text Available Computational Fluid Dynamics (CFD is increasingly being used in nuclear reactor safety (NRS analyses as a tool that enables safety relevant phenomena occurring in the reactor coolant system to be described in more detail. Numerical investigations on single phase coolant mixing in Pressurised Water Reactors (PWR have been performed at the FZD for almost a decade. The work is aimed at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity. For the experimental investigation of horizontal two phase flows, different non pressurized channels and the TOPFLOW Hot Leg model in a pressure chamber was build and simulated with ANSYS CFX. In a common project between the University of Applied Sciences Zittau/Görlitz and FZD the behaviour of insulation material released by a LOCA released into the containment and might compromise the long term emergency cooling systems is investigated. Moreover, the actual capability of CFD is shown to contribute to fuel rod bundle design with a good CHF performance.

  15. Renewability and sustainability aspects of nuclear energy

    Energy Technology Data Exchange (ETDEWEB)

    Şahin, Sümer, E-mail: ssahin@atilim.edit.tr [Department of Mechanical Engineering, Faculty of Engineering, ATILIM University, 06836 İncek, Gölbaşı, Ankara (Turkey)

    2014-09-30

    Renewability and sustainability aspects of nuclear energy have been presented on the basis of two different technologies: (1) Conventional nuclear technology; CANDU reactors. (2) Emerging nuclear technology; fusion/fission (hybrid) reactors. Reactor grade (RG) plutonium, {sup 233}U fuels and heavy water moderator have given a good combination with respect to neutron economy so that mixed fuel made of (ThO{sub 2}/RG‐PuO{sub 2}) or (ThC/RG-PuC) has lead to very high burn up grades. Five different mixed fuel have been selected for CANDU reactors composed of 4 % RG‐PuO{sub 2} + 96 % ThO{sub 2}; 6 % RG‐PuO{sub 2} + 94 % ThO{sub 2}; 10 % RG‐PuO{sub 2} + 90 % ThO{sub 2}; 20 % RG‐PuO{sub 2} + 80 % ThO{sub 2}; 30 % RG‐PuO{sub 2} + 70 % ThO{sub 2}, uniformly taken in each fuel rod in a fuel channel. Corresponding operation lifetimes have been found as ∼ 0.65, 1.1, 1.9, 3.5, and 4.8 years and with burn ups of ∼ 30 000, 60 000, 100 000, 200 000 and 290 000 MW.d/ton, respectively. Increase of RG‐PuO{sub 2} fraction in radial direction for the purpose of power flattening in the CANDU fuel bundle has driven the burn up grade to 580 000 MW.d/ton level. A laser fusion driver power of 500 MW{sub th} has been investigated to burn the minor actinides (MA) out of the nuclear waste of LWRs. MA have been homogenously dispersed as carbide fuel in form of TRISO particles with volume fractions of 0, 2, 3, 4 and 5 % in the Flibe coolant zone in the blanket surrounding the fusion chamber. Tritium breeding for a continuous operation of the fusion reactor is calculated as TBR = 1.134, 1.286, 1.387, 1.52 and 1.67, respectively. Fission reactions in the MA fuel under high energetic fusion neutrons have lead to the multiplication of the fusion energy by a factor of M = 3.3, 4.6, 6.15 and 8.1 with 2, 3, 4 and 5 % TRISO volume fraction at start up, respectively. Alternatively with thorium, the same fusion driver would produce ∼160 kg {sup 233}U per year in addition to fission

  16. Renewability and sustainability aspects of nuclear energy

    Science.gov (United States)

    Şahin, Sümer

    2014-09-01

    Renewability and sustainability aspects of nuclear energy have been presented on the basis of two different technologies: (1) Conventional nuclear technology; CANDU reactors. (2) Emerging nuclear technology; fusion/fission (hybrid) reactors. Reactor grade (RG) plutonium, 233U fuels and heavy water moderator have given a good combination with respect to neutron economy so that mixed fuel made of (ThO2/RG-PuO2) or (ThC/RG-PuC) has lead to very high burn up grades. Five different mixed fuel have been selected for CANDU reactors composed of 4 % RG-PuO2 + 96 % ThO2; 6 % RG-PuO2 + 94 % ThO2; 10 % RG-PuO2 + 90 % ThO2; 20 % RG-PuO2 + 80 % ThO2; 30 % RG-PuO2 + 70 % ThO2, uniformly taken in each fuel rod in a fuel channel. Corresponding operation lifetimes have been found as ˜ 0.65, 1.1, 1.9, 3.5, and 4.8 years and with burn ups of ˜ 30 000, 60 000, 100 000, 200 000 and 290 000 MW.d/ton, respectively. Increase of RG-PuO2 fraction in radial direction for the purpose of power flattening in the CANDU fuel bundle has driven the burn up grade to 580 000 MW.d/ton level. A laser fusion driver power of 500 MWth has been investigated to burn the minor actinides (MA) out of the nuclear waste of LWRs. MA have been homogenously dispersed as carbide fuel in form of TRISO particles with volume fractions of 0, 2, 3, 4 and 5 % in the Flibe coolant zone in the blanket surrounding the fusion chamber. Tritium breeding for a continuous operation of the fusion reactor is calculated as TBR = 1.134, 1.286, 1.387, 1.52 and 1.67, respectively. Fission reactions in the MA fuel under high energetic fusion neutrons have lead to the multiplication of the fusion energy by a factor of M = 3.3, 4.6, 6.15 and 8.1 with 2, 3, 4 and 5 % TRISO volume fraction at start up, respectively. Alternatively with thorium, the same fusion driver would produce ˜160 kg 233U per year in addition to fission energy production in situ, multiplying the fusion energy by a factor of ˜1.3.

  17. Impact of AD995 alumina rods

    Energy Technology Data Exchange (ETDEWEB)

    Chhabildas, L.C.; Furnish, M.D.; Reinhart, W.D. [Sandia National Labs., Albuquerque, NM (United States); Grady, D.E. [Applied Research Associates, Inc., Albuquerque, NM (United States)

    1997-10-01

    Gas guns and velocity interferometric techniques have been used to determine the loading behavior of an AD995 alumina rod 19 mm in diameter by 75 mm and 150 mm long, respectively. Graded-density materials were used to impact both bare and sleeved alumina rods while the velocity interferometer was used to monitor the axial-velocity of the free end of the rods. Results of these experiments demonstrate that (1) a time-dependent stress pulse generated during impact allows an efficient transition from the initial uniaxial strain loading to a uniaxial stress state as the stress pulse propagates through the rod, and (2) the intermediate loading rates obtained in this configuration lie between split Hopkinson bar and shock-loading techniques.

  18. Computer simulation of rod-sphere mixtures

    CERN Document Server

    Antypov, D

    2003-01-01

    Results are presented from a series of simulations undertaken to investigate the effect of adding small spherical particles to a fluid of rods which would otherwise represent a liquid crystalline (LC) substance. Firstly, a bulk mixture of Hard Gaussian Overlap particles with an aspect ratio of 3:1 and hard spheres with diameters equal to the breadth of the rods is simulated at various sphere concentrations. Both mixing-demixing and isotropic-nematic transition are studied using Monte Carlo techniques. Secondly, the effect of adding Lennard-Jones particles to an LC system modelled using the well established Gay-Berne potential is investigated. These rod-sphere mixtures are simulated using both the original set of interaction parameters and a modified version of the rod-sphere potential proposed in this work. The subject of interest is the internal structure of the binary mixture and its dependence on density, temperature, concentration and various parameters characterising the intermolecular interactions. Both...

  19. Bouncing Balls and Hot Rod Races.

    Science.gov (United States)

    Tibbs, Peggy; Sherrill, Donna

    This paper presents the Bouncing Ball Experiment which models quadratic and exponential functions, and the Hot Rod Races activity that explores velocity and acceleration. Activities include directions for the use of TI-82 and TI-83 calculators. (YDS)

  20. Investigation of Thermal Hydraulics of a Nuclear Reactor Moderator

    Science.gov (United States)

    Sarchami, Araz

    A three-dimensional numerical modeling of the thermo hydraulics of Canadian Deuterium Uranium (CANDU) nuclear reactor is conducted. The moderator tank is a Pressurized heavy water reactor which uses heavy water as moderator in a cylindrical tank. The main use of the tank is to bring the fast neutrons to the thermal neutron energy levels. The moderator tank compromises of several bundled tubes containing nuclear rods immersed inside the heavy water. It is important to keep the water temperature in the moderator at sub-cooled conditions, to prevent potential failure due to overheating of the tubes. Because of difficulties in measuring flow characteristics and temperature conditions inside a real reactor moderator, tests are conducted using a scaled moderator in moderator test facility (MTF) by Chalk River Laboratories of Atomic Energy of Canada Limited (CRL, AECL). MTF tests are conducted using heating elements to heat tube surfaces. This is different than the real reactor where nuclear radiation is the source of heating which results in a volumetric heating of the heavy water. The data recorded inside the MTF tank have shown levels of fluctuations in the moderator temperatures and requires in depth investigation of causes and effects. The purpose of the current investigation is to determine the causes for, and the nature of the moderator temperature fluctuations using three-dimensional simulation of MTF with both (surface heating and volumetric heating) modes. In addition, three dimensional simulation of full scale actual moderator tank with volumetric heating is conducted to investigate the effects of scaling on the temperature distribution. The numerical simulations are performed on a 24-processor cluster using parallel version of the FLUENT 12. During the transient simulation, 55 points of interest inside the tank are monitored for their temperature and velocity fluctuations with time.

  1. Coherent hollow-core waveguide bundles for thermal imaging.

    Science.gov (United States)

    Gal, Udi; Harrington, James; Ben-David, Moshe; Bledt, Carlos; Syzonenko, Nicholas; Gannot, Israel

    2010-09-01

    There has been very little work done in the past to extend the wavelength range of fiber image bundles to the IR range. This is due, in part, to the lack of IR transmissive fibers with optical and mechanical properties analogous to the oxide glass fibers currently employed in the visible fiber bundles. Our research is aimed at developing high-resolution hollow-core coherent IR fiber bundles for transendoscopic infrared imaging. We employ the hollow glass waveguide (HGW) technology that was used successfully to make single-HGWs with Ag/AgI thin film coatings to form coherent bundles for IR imaging. We examine the possibility of developing endoscopic systems to capture thermal images using hollow waveguide fiber bundles adjusted to the 8-10?mum spectral range and investigate the applicability of such systems. We carried out a series of measurements in order to characterize the optical properties of the fiber bundles. These included the attenuation, resolution, and temperature response. We developed theoretical models and simulation tools that calculate the light propagation through HGW bundles, and which can be used to calculate the optical properties of the fiber bundles. Finally, the HGW fiber bundles were used to transmit thermal images of various heated objects; the results were compared with simulation results. The experimental results are encouraging, show an improvement in the resolution and thermal response of the HGW fiber bundles, and are consistent with the theoretical results. Nonetheless, additional improvements in the attenuation of the bundles are required in order to be able to use this technology for medical applications.

  2. Microelectrophoresis of Silica Rods Using Confocal Microscopy.

    Science.gov (United States)

    Bakker, Henriëtte E; Besseling, Thijs H; Wijnhoven, Judith E G J; Helfferich, Peter H; van Blaaderen, Alfons; Imhof, Arnout

    2017-01-31

    The electrophoretic mobility and the zeta potential (ζ) of fluorescently labeled colloidal silica rods, with an aspect ratio of 3.8 and 6.1, were determined with microelectrophoresis measurements using confocal microscopy. In the case where the colloidal particles all move at the same speed parallel to the direction of the electric field, we record a xyz-stack over the whole depth of the capillary. This method is faster and more robust compared to taking xyt-series at different depths inside the capillary to obtain the parabolic flow profile, as was done in previous work from our group. In some cases, rodlike particles do not move all at the same speed in the electric field, but exhibit a velocity that depends on the angle between the long axis of the rod and the electric field. We measured the orientation-dependent velocity of individual silica rods during electrophoresis as a function of κa, where κ(-1) is the double layer thickness and a is the radius of the rod associated with the diameter. Thus, we determined the anisotropic electrophoretic mobility of the silica rods with different sized double layers. The size of the double layer was tuned by suspending silica rods in different solvents at different electrolyte concentrations. We compared these results with theoretical predictions. We show that even at already relatively high κa when the Smoluchowski limiting law is assumed to be valid (κa > 10), an orientation dependent velocity was measured. Furthermore, we observed that at decreasing values of κa the anisotropy in the electrophoretic mobility of the rods increases. However, in low polar solvents with κa < 1, this trend was reversed: the anisotropy in the electrophoretic mobility of the rods decreased. We argue that this decrease is due to end effects, which was already predicted theoretically. When end effects are not taken into account, this will lead to strong underestimation of the experimentally determined zeta potential.

  3. High Power Performance of Rod Fiber Amplifiers

    DEFF Research Database (Denmark)

    Johansen, Mette Marie; Michieletto, Mattia; Kristensen, Torben

    2015-01-01

    An improved version of the DMF rod fiber is tested in a high power setup delivering 360W of stable signal power. Multiple testing degrades the fiber and transverse modal instability threshold from >360W to ~290W.......An improved version of the DMF rod fiber is tested in a high power setup delivering 360W of stable signal power. Multiple testing degrades the fiber and transverse modal instability threshold from >360W to ~290W....

  4. IMPACT CONICAL ROD ON HARD LIMITER

    Directory of Open Access Journals (Sweden)

    Ulitin G.

    2014-12-01

    Full Text Available The problem is considered of longitudinal impact conical rod in article. A recommendation on the use of the approximate method of calculation is based on an analysis of the influence of design parameters on the value of the main oscillation frequency. There was obtained an equation of the displacement and stress of the rod. Engineering dependence has been proposed to determine the maximum force in the impact section.

  5. Self-diagnosing braided composite rod

    OpenAIRE

    Fangueiro, Raúl; Zdraveva, E.; Pereira, Cristiana Gonilho; Ferreira, A; Lanceros-Méndez, S.

    2010-01-01

    This paper presents the development of a braided reinforced composite rod (BCR) able to both reinforce and monitor the stress state of concrete structures. Carbon fibers have been used as sensing and reinforcing materials along with glass fiber. Various composites rods have been produced using an author patented technique based on a modified conventional braiding machine. The materials investigated were prepared with different carbon fiber content as follows: BCR2 (77% glass/23...

  6. Measurements of control rod efficiency in RBMK critical assembly upon dropping of the rods

    Energy Technology Data Exchange (ETDEWEB)

    Zhitarev, V. E., E-mail: vejitarev@nnrd.kiae.su; Kachanov, V. M.; Sergevnin, A. Yu.; Lebedev, G. V., E-mail: lgv2004@mail.ru [National Research Center Kurchatov Institute (Russian Federation)

    2014-12-15

    The efficiency of control rods in the RBMK critical assembly was measured in the case where one manual-control rod (MCR) is dropped from a steady critical state, and several other MCRs were additionally dropped after 44 s. The measured number of neutrons in the assembly during and after dropping of the rods was used to calculate the efficiency values of the rods by solution of the system of point kinetics equations. A series of methods of the initial data treatment for determination of the desired values of reactivity without the calculated corrections were used.

  7. Measurements of control rod efficiency in RBMK critical assembly upon dropping of the rods

    Science.gov (United States)

    Zhitarev, V. E.; Kachanov, V. M.; Sergevnin, A. Yu.; Lebedev, G. V.

    2014-12-01

    The efficiency of control rods in the RBMK critical assembly was measured in the case where one manual-control rod (MCR) is dropped from a steady critical state, and several other MCRs were additionally dropped after 44 s. The measured number of neutrons in the assembly during and after dropping of the rods was used to calculate the efficiency values of the rods by solution of the system of point kinetics equations. A series of methods of the initial data treatment for determination of the desired values of reactivity without the calculated corrections were used.

  8. High-throughput rod-induced electrospinning

    Science.gov (United States)

    Wu, Dezhi; Xiao, Zhiming; Teh, Kwok Siong; Han, Zhibin; Luo, Guoxi; Shi, Chuan; Sun, Daoheng; Zhao, Jinbao; Lin, Liwei

    2016-09-01

    A high throughput electrospinning process, directly from flat polymer solution surfaces induced by a moving insulating rod, has been proposed and demonstrated. Different rods made of either phenolic resin or paper with a diameter of 1-3 cm and a resistance of about 100-500 MΩ, has been successfully utilized in the process. The rod is placed approximately 10 mm above the flat polymer solution surface with a moving speed of 0.005-0.4 m s-1 this causes the solution to generate multiple liquid jets under an applied voltage of 15-60 kV for the tip-less electrospinning process. The local electric field induced by the rod can boost electrohydrodynamic instability in order to generate Taylor cones and liquid jets. Experimentally, it is found that a large rod diameter and a small solution-to-rod distance can enhance the local electrical field to reduce the magnitude of the applied voltage. In the prototype setup with poly (ethylene oxide) polymer solution, an area of 5 cm  ×  10 cm and under an applied voltage of 60 kV, the maximum throughput of nanofibers is recorded to be approximately144 g m-2 h-1.

  9. Rigid rod anchored to infinite membrane.

    Science.gov (United States)

    Guo, Kunkun; Qiu, Feng; Zhang, Hongdong; Yang, Yuliang

    2005-08-15

    We investigate the shape deformation of an infinite membrane anchored by a rigid rod. The density profile of the rod is calculated by the self-consistent-field theory and the shape of the membrane is predicted by the Helfrich membrane elasticity theory [W. Helfrich, Z. Naturforsch. 28c, 693 (1973)]. It is found that the membrane bends away from the rigid rod when the interaction between the rod and the membrane is repulsive or weakly attractive (adsorption). However, the pulled height of the membrane at first increases and then decreases with the increase of the adsorption strength. Compared to a Gaussian chain with the same length, the rigid rod covers much larger area of the membrane, whereas exerts less local entropic pressure on the membrane. An evident gap is found between the membrane and the rigid rod because the membrane's curvature has to be continuous. These behaviors are compared with that of the flexible-polymer-anchored membranes studied by previous Monte Carlo simulations and theoretical analysis. It is straightforward to extend this method to more complicated and real biological systems, such as infinite membrane/multiple chains, protein inclusion, or systems with phase separation.

  10. Long-Rod Moving-Plate Interaction

    Science.gov (United States)

    Partom, Y.

    2002-07-01

    Understanding the mechanics of interaction of a long rod projectile with a forward moving plate at an angle is essential to understanding long rod interaction with an explosive reactive armor cassette. To investigate the mechanics of such an interaction we use AUTODIN2D/EULER in plane geometry, although the problem is 3D. We assume that this is a satisfactory approximation, as we're only interested in the main features, and are not comparing fine details to experimental results. From the simulations we learn that the interaction never reaches steady state. Initially each material splits into two streams, and the interaction plane is perpendicular to the rod. But with time the interaction plane rotates slowly, until it becomes parallel to the rod, which is then able to continue moving forward without interruption. During this process interacting rod material of length DeltaL is diverted at an angle and becomes ineffective for penetrating the main target. We made many such runs to determine the dependence of DeltaL on the parameters of the problem. This dependence makes it possible to predict DeltaL for a variety of rod-plate situations.

  11. Topological optimisation of rod-stirring devices

    CERN Document Server

    Finn, Matthew D

    2011-01-01

    There are many industrial situations where rods are used to stir a fluid, or where rods repeatedly stretch a material such as bread dough or taffy. The goal in these applications is to stretch either material lines (in a fluid) or the material itself (for dough or taffy) as rapidly as possible. The growth rate of material lines is conveniently given by the topological entropy of the rod motion. We discuss the problem of optimising such rod devices from a topological viewpoint. We express rod motions in terms of generators of the braid group, and assign a cost based on the minimum number of generators needed to write the braid. We show that for one cost function -- the topological entropy per generator -- the optimal growth rate is the logarithm of the golden ratio. For a more realistic cost function,involving the topological entropy per operation where rods are allowed to move together, the optimal growth rate is the logarithm of the silver ratio, $1+\\sqrt{2}$. We show how to construct devices that realise th...

  12. Degradation in steam of 60 cm-long B4C control rods

    Science.gov (United States)

    Dominguez, C.; Drouan, D.

    2014-08-01

    In the framework of nuclear reactor core meltdown accident studies, the degradation of boron carbide control rod segments exposed to argon/steam atmospheres was investigated up to about 2000 °C in IRSN laboratories. The sequence of the phenomena involved in the degradation has been found to take place as expected. Nevertheless, the ZrO2 oxide layer formed on the outer surface of the guide tube was very protective, significantly delaying and limiting the guide tube failure and therefore the boron carbide pellet oxidation. Contrary to what was expected, the presence of the control rod decreases the hydrogen release instead of increasing it by additional oxidation of boron compounds. Boron contents up to 20 wt.% were measured in metallic mixtures formed during degradation. It was observed that these metallic melts are able to attack the surrounding fuel rods, which could have consequences on fuel degradation and fission product release kinetics during severe accidents.

  13. Nuclear Pasta Formation

    CERN Document Server

    Schneider, A S; Hughto, J; Berry, D K

    2013-01-01

    The formation of complex nonuniform phases of nuclear matter, known as nuclear pasta, is studied with molecular dynamics simulations containing 51200 nucleons. A phenomenological nuclear interaction is used that reproduces the saturation binding energy and density of nuclear matter. Systems are prepared at an initial density of 0.10fm$^{-3}$ and then the density is decreased by expanding the simulation volume at different rates to densities of 0.01 fm$^{-3}$ or less. An originally uniform system of nuclear matter is observed to form spherical bubbles ("swiss cheese"), hollow tubes, flat plates ("lasagna"), thin rods ("spaghetti") and, finally, nearly spherical nuclei with decreasing density. We explicitly observe nucleation mechanisms, with decreasing density, for these different pasta phase transitions. Topological quantities known as Minkowski functionals are obtained to characterize the pasta shapes. Different pasta shapes are observed depending on the expansion rate. This indicates non equilibrium effects...

  14. International symposium on fuel rod simulators: development and application

    Energy Technology Data Exchange (ETDEWEB)

    McCulloch, R.W. (comp.)

    1981-05-01

    Separate abstracts are included for each of the papers presented concerning fuel rod simulator operation and performance; simulator design and evaluation; clad heated fuel rod simulators and fuel rod simulators for cladding investigations; fuel rod simulator components and inspection; and simulator analytical modeling. Ten papers have previously been input to the Energy Data Base.

  15. Non-abelian higher gauge theory and categorical bundle

    CERN Document Server

    Viennot, David

    2012-01-01

    A gauge theory is associated with a principal bundle endowed with a connection permitting to define horizontal lifts of paths. The horizontal lifts of surfaces cannot be defined into a principal bundle structure. An higher gauge theory is an attempt to generalize the bundle structure in order to describe horizontal lifts of surfaces. A such attempt is particularly difficult for the non-abelian case. Some structures have been proposed to realize this goal (twisted bundle, gerbes with connection, bundle gerbe, 2-bundle). Each of them uses a category in place of the total space manifold of the usual principal bundle structure. Some of them replace also the structure group by a category (more precisely a Lie crossed module viewed as a category). But the base space remains still a simple manifold (possibly viewed as a trivial category with only identity arrows). We propose a new principal categorical bundle structure, with a Lie crossed module as structure groupoid, but with a base space belonging to a bigger clas...

  16. Presenting Lexical Bundles for Explicit Noticing with Schematic Linguistic Representation

    Science.gov (United States)

    Thomson, Haidee Elizabeth

    2016-01-01

    Lexical bundles are essential for fluency, but their incompleteness is a stumbling block for learners. In this study, two presentation methods to increase awareness of lexical bundles through explicit noticing are explored and compared with incidental exposure. The three conditions in this study were as follows: noticing with schematic linguistic…

  17. Smooth Bundling of Large Streaming and Sequence Graphs

    NARCIS (Netherlands)

    Hurter, C.; Ersoy, O.; Telea, A.

    2013-01-01

    Dynamic graphs are increasingly pervasive in modern information systems. However, understanding how a graph changes in time is difficult. We present here two techniques for simplified visualization of dynamic graphs using edge bundles. The first technique uses a recent image-based graph bundling met

  18. Helical twist controls the thickness of F-actin bundles

    NARCIS (Netherlands)

    Claessens, M.M.A.E.; Semmrich, C.; Ramos, L.; Bausch, A.R.

    2008-01-01

    In the presence of condensing agents such as nonadsorbing polymer, multivalent counter ions, and specific bundling proteins, chiral biopolymers typically form bundles with a finite thickness, rather than phase-separating into a polymer-rich phase. Although short-range repulsive interactions or geome

  19. Lexical Bundles in L1 and L2 Academic Writing

    Science.gov (United States)

    Chen, Yu-Hua; Baker, Paul

    2010-01-01

    This paper adopts an automated frequency-driven approach to identify frequently-used word combinations (i.e., "lexical bundles") in academic writing. Lexical bundles retrieved from one corpus of published academic texts and two corpora of student academic writing (one L1, the other L2), were investigated both quantitatively and qualitatively.…

  20. Lexical Bundles: Facilitating University "Talk" in Group Discussions

    Science.gov (United States)

    Heng, Chan Swee; Kashiha, Hadi; Tan, Helen

    2014-01-01

    Group discussion forms an integral language experience for most language learners, providing them with an opportunity to express themselves in a naturalistic setting. Multi-word expressions are commonly used and one of them is lexical bundles. Lexical bundles are types of extended collocations that occur more commonly than we expect; they are…