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Sample records for blanket module iter

  1. ITER breeding blanket module design and analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kuroda, Toshimasa; Enoeda, Mikio; Kikuchi, Shigeto [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

    1998-11-01

    The ITER breeding blanket employs a ceramic breeder and Be neutron multiplier both in small spherical pebble form. Radial-poloidal cooling panels are arranged in the blanket box to remove the nuclear heating in these materials and to reinforce the blanket structure. At the first wall, Be armor is bonded onto the stainless steel (SS) structure to provide a low Z plasma-compatible surface and to protect the first wall/blanket structure from the direct contact with the plasma during off-normal events. Thermo-mechanical analyses and investigation of fabrication procedure have been performed for this breeding blanket. To evaluate thermo-mechanical behavior of the pebble beds including the dependency of the effective thermal conductivity on stress, analysis methods have been preliminary established by the use of special calculation option of ABAQUS code, which are briefly summarized in this report. The structural response of the breeding blanket module under internal pressure of 4 MPa (in case of in-blanket LOCA) resulted in rather high stress in the blanket side (toroidal end) wall, thus addition of a stiffening rib or increase of the wall thickness will be needed. Two-dimensional elasto-plastic analyses have been performed for the Be/SS bonded interface at the first wall taking a fabrication process based on HIP bonding and thermal cycle due to pulsed plasma operation into account. The stress-strain hysteresis during these process and operation was clarified, and a procedure to assess and/or confirm the bonding integrity was also proposed. Fabrication sequence of the breeding blanket module was preliminarily developed based on the procedure to fabricate part by part and to assemble them one by one. (author)

  2. Detailed 3-D nuclear analysis of ITER outboard blanket modules

    Energy Technology Data Exchange (ETDEWEB)

    Bohm, Tim, E-mail: tdbohm@wisc.edu [Fusion Technology Institute, University of Wisconsin-Madison, Madison, WI (United States); Davis, Andrew; Sawan, Mohamed; Marriott, Edward; Wilson, Paul [Fusion Technology Institute, University of Wisconsin-Madison, Madison, WI (United States); Ulrickson, Michael; Bullock, James [Formerly, Fusion Technology, Sandia National Laboratories, Albuquerque, NM (United States)

    2015-10-15

    Highlights: • Nuclear analysis was performed on detailed CAD models placed in a 40 degree model of ITER. • The regions examined include BM09, the upper ELM coil region (BM11–13), the neutral beam (NB) region (BM13–16), and BM18. • The results show that VV nuclear heating exceeds limits in the NB and upper ELM coil regions. • The results also show that the level of He production in parts of BM18 exceeds limits. • These calculations are being used to modify the design of the ITER blanket modules. - Abstract: In the ITER design, the blanket modules (BM) provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40 degree partially homogenized ITER global model. The regions analyzed include BM09, BM16 near the heating neutral beam injection (HNB) region, BM11–13 near the upper ELM coil region, and BM18. For the BM16 HNB region, the VV nuclear heating behind the NB region exceeds the design limit by up to 80%. For the BM11–13 region, the nuclear heating of the VV exceeds the design limit by up to 45%. For BM18, the results show that He production does not meet the limit necessary for re-welding. The results presented in this work are being used by the ITER Organization Blanket and Tokamak Integration groups to modify the BM design in the cases where limits are exceeded.

  3. Nuclear analysis of ITER Test Blanket Module Port Plug

    Energy Technology Data Exchange (ETDEWEB)

    Villari, Rosaria, E-mail: rosaria.villari@enea.it [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Kim, Byoung Yoon; Barabash, Vladimir; Giancarli, Luciano; Levesy, Bruno; Loughlin, Michael; Merola, Mario [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 Saint Paul-lez-Durance Cedex (France); Moro, Fabio [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Pascal, Romain [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 Saint Paul-lez-Durance Cedex (France); Petrizzi, Luigino [European Commission, DG Research & Innovation G5, CDMA 00/030, B-1049 Brussels (Belgium); Polunovsky, Eduard; Van Der Laan, Jaap G. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 Saint Paul-lez-Durance Cedex (France)

    2015-10-15

    Highlights: • 3D nuclear analysis of the ITER TBM Port Plug (PP). • Calculations of neutron fluxes, nuclear heating, damage and He-production in TBM PP components. • Shutdown dose rate assessment with Advanced D1S method considering different configurations. • Potential design improvements to reduce the shutdown dose rate in the port interspace. - Abstract: Nuclear analyses have been performed for the ITER Test Blanket Module Port Plug (TBM PP) using the MCNP-5 Monte Carlo Code. A detailed 3D model of the TBM Port Plug with dummy TBM has been integrated into the ITER MCNP model (B-lite v.3). Neutron fluxes, nuclear heating, helium production and neutron damage have been calculated in all the TBM PP components. Global shutdown dose rate calculations have also been performed with Advanced D1S method for different configurations of the TBM PP system. This paper presents the results of these analyses and discusses potential design improvements aiming to further reduce the shutdown dose rate in the port interspace.

  4. Detailed 3-D nuclear analysis of ITER blanket modules

    Energy Technology Data Exchange (ETDEWEB)

    Bohm, T.D., E-mail: tdbohm@wisc.edu [University of Wisconsin-Madison, Madison, WI (United States); Sawan, M.E.; Marriott, E.P.; Wilson, P.P.H. [University of Wisconsin-Madison, Madison, WI (United States); Ulrickson, M.; Bullock, J. [Sandia National Laboratories, Albuquerque, NM (United States)

    2014-10-15

    In ITER, the blanket modules (BM) are arranged around the plasma to provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. As a part of the BM design process, nuclear analysis is required to determine the level of nuclear heating, helium production, and radiation damage in the BM. Additionally, nuclear heating in the VV is also important for assessing the BM design. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40-degree partially homogenized ITER global model. The regions analyzed include BM01, the neutral beam injection (NB) region, and the upper port region. For BM01, the results show that He production meets the limit necessary for re-welding, and the VV heating behind BM01 is acceptable. For the NBI region, the VV nuclear heating behind the NB region exceeds the design limit by a factor of two. For the upper port region, the nuclear heating of the VV exceeds the design limit by up to 20%. The results presented in this work are being used to modify the BM design in the cases where limits are exceeded.

  5. RF test blanket sub-module with ceramic breeder and helium cooling for test in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Kovalenko, V. [N.A. Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation)]. E-mail: koval@nikiet.ru; Kapyshev, V. [N.A. Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation); Leshukov, A. [N.A. Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation); Poliksha, V. [N.A. Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation); Shatalov, G. [Russian Research Center ' Kurchatov Institute' , Kurchatov Square 1, 123182 Moscow (Russian Federation); Strebkov, Yu. [N.A. Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation); Strizhov, A. [N.A. Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation); Sviridenko, M. [N.A. Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, Moscow 101000 (Russian Federation)

    2006-02-15

    International thermonuclear experimental reactor (ITER) is anticipated as the only one step to DEMO fusion reactor. One of its main objectives is to demonstrate the availability and integration of technologies essential for a fusion reactor by testing of components for a future reactor including the test blanket modules (TBM) with different types of breeding materials. RF proposed to divide the TBM on two parts and to use two independent test blanket sub-modules (TBSM) which fixed on the frame in ITER horizontal experimental port for testing. CHC TBSM design description, its mechanical attachment on the frame, and principle schemes of helium cooling system and tritium cycle system are presented in this paper.

  6. Safety analysis of the US dual coolant liquid lead lithium ITER test blanket module

    Science.gov (United States)

    Merrill, Brad; Reyes, Susana; Sawan, Mohamed; Wong, Clement

    2007-07-01

    The US is proposing a prototype of a dual coolant liquid lead-lithium DEMO blanket concept for testing in the International Thermonuclear Experimental Reactor (ITER) as an ITER test blanket module (TBM). Because safety considerations are an integral part of the design process to ensure that this TBM does not adversely impact the safety of ITER, a safety assessment has been conducted for this TBM and its ancillary systems as requested by the ITER project. Four events were selected by the ITER international team (IT) to address specific reactor safety concerns, such as vaccum vessel (VV) pressurization, confinement building pressure build-up, TBM decay heat removal capability, tritium and activation products release from the TBM system and hydrogen and heat production from chemical reactions. This paper summarizes the results of this safety assessment conducted with the MELCOR computer code.

  7. ITER test blanket module error field simulation experiments at DIII-D

    NARCIS (Netherlands)

    Schaffer, M. J.; Snipes, J. A.; Gohil, P.; P. de Vries,; Evans, T. E.; Fenstermacher, M.E.; Gao, X.; Garofalo, A. M.; Gates, D. A.; Greenfield, C.M.; Heidbrink, W. W.; Kramer, G. J.; La Haye, R. J.; Liu, S.; Loarte, A.; Nave, M. F. F.; Osborne, T. H.; Oyama, N.; Park, J. K.; Ramasubramanian, N.; Reimerdes, H.; Saibene, G.; Salmi, A.; Shinohara, K.; Spong, D. A.; Solomon, W. M.; Tala, T.; Zhu, Y. B.; Boedo, J. A.; Chuyanov, V.; Doyle, E. J.; Jakubowski, M.; Jhang, H.; Nazikian, R. M.; Pustovitov, V. D.; Schmitz, O.; Srinivasan, R.; Taylor, T. S.; Wade, M. R.; You, K. I.; Zeng, L.

    2011-01-01

    Experiments at DIII-D investigated the effects of magnetic error fields similar to those expected from proposed ITER test blanket modules (TBMs) containing ferromagnetic material. Studied were effects on: plasma rotation and locking, confinement, L-H transition, the H-mode pedestal, edge localized m

  8. Fusion Reactor and Fusion Reactor Materials:Concept Design of the ITER Test Blanket Modules

    Institute of Scientific and Technical Information of China (English)

    HUANGJinhua; LIZaixing; ZHUYukun; HUGang

    2003-01-01

    Performances required: prospect to be adopted in DEMO. Shielding for V.V. and TFC in ITER. Design principles: the peak temperature and stress should not exceed technical limits. The structure of test blanket modules (TBM) should be simple for easy fabrication, and TBM should be robust for reliability.

  9. Technical issues of reduced activation ferritic/martensitic steels for fabrication of ITER test blanket modules

    Energy Technology Data Exchange (ETDEWEB)

    Tanigawa, H. [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan)], E-mail: tanigawa.hiroyasu@jaea.go.jp; Hirose, T.; Shiba, K. [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Kasada, R. [Institute of Advanced Energy, Kyoto University, Uji, Kyoto 611-0011 (Japan); Wakai, E. [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Serizawa, H.; Kawahito, Y. [Joining and Welding Research Institute, Osaka University, Ibaraki, Osaka 567-0047 (Japan); Jitsukawa, S. [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Kimura, A. [Institute of Advanced Energy, Kyoto University, Uji, Kyoto 611-0011 (Japan); Kohno, Y. [Department of Materials Science and Engineering, Muroran Institute of Technology, Muroran, Hokkaido 050-8585 (Japan); Kohyama, A. [Institute of Advanced Energy, Kyoto University, Uji, Kyoto 611-0011 (Japan); Katayama, S. [Joining and Welding Research Institute, Osaka University, Ibaraki, Osaka 567-0047 (Japan); Mori, H.; Nishimoto, K. [Division of Materials and Manufacturing Science, Osaka University, Ibaraki, Osaka 565-0871 (Japan); Klueh, R.L.; Sokolov, M.A.; Stoller, R.E.; Zinkle, S.J. [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831-6132 (United States)

    2008-12-15

    Reduced activation ferritic/martensitic steels (RAFMs) are recognized as the primary candidate structural materials for fusion blanket systems. The RAFM F82H was developed in Japan with emphasis on high-temperature properties and weldability. Extensive irradiation studies have conducted on F82H, and it has the most extensive available database of irradiated and unirradiated properties of all RAFMs. The objective of this paper is to review the R and D status of F82H and to identify the key technical issues for the fabrication of an ITER test blanket module (TBM) suggested from the recent research achievements in Japan. This work clarified that the primary issues with F82H involve welding techniques and the mechanical properties of weld joints. This is the result of the distinctive nature of the joint caused by the phase transformation that occurs in the weld joint during cooling, and its impact on the design of a TBM will be discussed.

  10. Influence of the blanket shield modules geometry on the operation of the ITER ICRF antenna

    Energy Technology Data Exchange (ETDEWEB)

    Louche, F., E-mail: fabrice.louche@rma.ac.be [Laboratory for Plasma Physics, Royal Military Academy, 30, Avenue de la Renaissance, 1000 Brussels (Belgium); Dumortier, P.; Durodié, F.; Messiaen, A. [Laboratory for Plasma Physics, Royal Military Academy, 30, Avenue de la Renaissance, 1000 Brussels (Belgium)

    2013-10-15

    Highlights: ► The ITER ICRF antenna and its surrounding blanket shield modules have been modeled with the 3D electromagnetic software Microwave Studio. ► Unexpected resonances are detected in the ITER relevant range of frequencies. ► These resonances are caused by the geometry of the modules, and in particular by the cavity at the back, between the module rear face and the port plug outer face, present in the considered model. ► Simplified modeling and transmission line computations validate this interpretation. ► The resonance is strongly dependent on the detailed geometry of the modules, but large voids should be avoided. -- Abstract: Three-dimensional electromagnetic simulations of the ITER ICRF antenna have been recently performed with the commercial code CST Microwave Studio{sup ®} (MWS) [1]. A detailed model imported from the CATIA{sup ®} file has been considered: it includes the 24 straps array (CY3.1 geometry [2]) and the surrounding blanket shield modules. The transient solver in MWS has detected the presence of a very localized peak in the input impedance matrix at a frequency of approximately 51 MHz in vacuum conditions. The presence of such a resonance in the ITER operating range of frequency is of concern and should be understood as previous analysis reported in [3] concluded that TEM and non-TEM modes are not expected in this frequency band as long as the antenna is grounded to the port at 1 m back from the antenna front face. By using a simplified model of the geometry we demonstrate that the resonance is a consequence of the considered geometry of the blanket shield modules and in particular of the cavity at the back of the modules made of the module attachment and the port plug outer face. We show that the presence of such a cavity locally increases the coaxial line impedance and allows for a TEM mode in the band. This physical analysis is supported by a transmission line model where the system made of the antenna and its surrounding

  11. Preliminary piping layout and integration of European test blanket modules subsystems in ITER CVCS area

    Energy Technology Data Exchange (ETDEWEB)

    Tarallo, Andrea, E-mail: andrea.tarallo@unina.it [CREATE, University of Naples Federico II, DII, P.le Tecchio, 80, 80125 Naples (Italy); Mozzillo, Rocco; Di Gironimo, Giuseppe [CREATE, University of Naples Federico II, DII, P.le Tecchio, 80, 80125 Naples (Italy); Aiello, Antonio; Utili, Marco [ENEA UTIS, C.R. Brasimone, Bacino del Brasimone, I-40032 Camugnano, BO (Italy); Ricapito, Italo [TBM& MD Project, Fusion for Energy, EU Commission, Carrer J. Pla, 2, Building B3, 08019 Barcelona (Spain)

    2015-04-15

    Highlights: • The use of human modeling tools for piping design in view of maintenance is discussed. • A possible preliminary layout for TBM subsystems in CVCS area has been designed with CATIA. • A DHM-based method to quickly check for maintainability of piping systems is suggested. - Abstract: This paper explores a possible integration of some ancillary systems of helium-cooled lithium lead (HCLL) and helium-cooled pebble-bed (HCPB) test blanket modules in ITER CVCS area. Computer-aided design and ergonomics simulation tools have been fundamental not only to define suitable routes for pipes, but also to quickly check for maintainability of equipment and in-line components. In particular, accessibility of equipment and systems has been investigated from the very first stages of the design using digital human models. In some cases, the digital simulations have resulted in changes in the initial space reservations.

  12. Experimental results and validation of a method to reconstruct forces on the ITER test blanket modules

    Energy Technology Data Exchange (ETDEWEB)

    Zeile, Christian, E-mail: christian.zeile@kit.edu; Maione, Ivan A.

    2015-10-15

    Highlights: • An in operation force measurement system for the ITER EU HCPB TBM has been developed. • The force reconstruction methods are based on strain measurements on the attachment system. • An experimental setup and a corresponding mock-up have been built. • A set of test cases representing ITER relevant excitations has been used for validation. • The influence of modeling errors on the force reconstruction has been investigated. - Abstract: In order to reconstruct forces on the test blanket modules in ITER, two force reconstruction methods, the augmented Kalman filter and a model predictive controller, have been selected and developed to estimate the forces based on strain measurements on the attachment system. A dedicated experimental setup with a corresponding mock-up has been designed and built to validate these methods. A set of test cases has been defined to represent possible excitation of the system. It has been shown that the errors in the estimated forces mainly depend on the accuracy of the identified model used by the algorithms. Furthermore, it has been found that a minimum of 10 strain gauges is necessary to allow for a low error in the reconstructed forces.

  13. A path to stable low-torque plasma operation in ITER with test blanket modules

    Science.gov (United States)

    Lanctot, M. J.; Snipes, J. A.; Reimerdes, H.; Paz-Soldan, C.; Logan, N.; Hanson, J. M.; Buttery, R. J.; deGrassie, J. S.; Garofalo, A. M.; Gray, T. K.; Grierson, B. A.; King, J. D.; Kramer, G. J.; La Haye, R. J.; Pace, D. C.; Park, J.-K.; Salmi, A.; Shiraki, D.; Strait, E. J.; Solomon, W. M.; Tala, T.; Van Zeeland, M. A.

    2017-03-01

    New experiments in the low-torque ITER Q  =  10 scenario on DIII-D demonstrate that n  =  1 magnetic fields from a single row of ex-vessel control coils enable operation at ITER performance metrics in the presence of applied non-axisymmetric magnetic fields from a test blanket module (TBM) mock-up coil. With n  =  1 compensation, operation below the ITER-equivalent injected torque is successful at three times the ITER equivalent toroidal magnetic field ripple for a pair of TBMs in one equatorial port, whereas the uncompensated TBM field leads to rotation collapse, loss of H-mode and plasma current disruption. In companion experiments at high plasma beta, where the n  =  1 plasma response is enhanced, uncorrected TBM fields degrade energy confinement and the plasma angular momentum while increasing fast ion losses; however, disruptions are not routinely encountered owing to increased levels of injected neutral beam torque. In this regime, n  =  1 field compensation leads to recovery of a dominant fraction of the TBM-induced plasma pressure and rotation degradation, and an 80% reduction in the heat load to the first wall. These results show that the n  =  1 plasma response plays a dominant role in determining plasma stability, and that n  =  1 field compensation alone not only recovers most of the impact on plasma performance of the TBM, but also protects the first wall from potentially damaging heat flux. Despite these benefits, plasma rotation braking from the TBM fields cannot be fully recovered using standard error field control. Given the uncertainty in extrapolation of these results to the ITER configuration, it is prudent to design the TBMs with as low a ferromagnetic mass as possible without jeopardizing the TBM mission.

  14. Effect of thick blanket modules on neoclassical tearing mode locking in ITER

    Science.gov (United States)

    La Haye, R. J.; Paz-Soldan, C.; Liu, Y. Q.

    2017-01-01

    The rotation of m/n  =  2/1 tearing modes can be slowed and stopped (i.e. locked) by eddy currents induced in resistive walls in conjunction with residual error fields that provide a final ‘notch’ point. This is a particular issue in ITER with large inertia and low applied torque (m and n are poloidal and toroidal mode numbers respectively). Previous estimates of tolerable 2/1 island widths in ITER found that the ITER electron cyclotron current drive (ECCD) system could catch and subdue such islands before they persisted long enough and grew large enough to lock. These estimates were based on a forecast of initial island rotation using the n  =  1 resistive penetration time of the inner vacuum vessel wall and benchmarked to DIII-D high-rotation plasmas, However, rotating tearing modes in ITER will also induce eddy currents in the blanket as the effective first wall that can shield the inner vessel. The closer fitting blanket wall has a much shorter time constant and should allow several times smaller islands to lock several times faster in ITER than previously considered; this challenges the ECCD stabilization. Recent DIII-D ITER baseline scenario (IBS) plasmas with low rotation through small applied torque allow better modeling and scaling to ITER with the blanket as the first resistive wall.

  15. ITER test blanket module error field simulation experiments at DIII-D

    Science.gov (United States)

    Schaffer, M. J.; Snipes, J. A.; Gohil, P.; de Vries, P.; Evans, T. E.; Fenstermacher, M. E.; Gao, X.; Garofalo, A. M.; Gates, D. A.; Greenfield, C. M.; Heidbrink, W. W.; Kramer, G. J.; La Haye, R. J.; Liu, S.; Loarte, A.; Nave, M. F. F.; Osborne, T. H.; Oyama, N.; Park, J.-K.; Ramasubramanian, N.; Reimerdes, H.; Saibene, G.; Salmi, A.; Shinohara, K.; Spong, D. A.; Solomon, W. M.; Tala, T.; Zhu, Y. B.; Boedo, J. A.; Chuyanov, V.; Doyle, E. J.; Jakubowski, M.; Jhang, H.; Nazikian, R. M.; Pustovitov, V. D.; Schmitz, O.; Srinivasan, R.; Taylor, T. S.; Wade, M. R.; You, K.-I.; Zeng, L.; DIII-D Team

    2011-10-01

    Experiments at DIII-D investigated the effects of magnetic error fields similar to those expected from proposed ITER test blanket modules (TBMs) containing ferromagnetic material. Studied were effects on: plasma rotation and locking, confinement, L-H transition, the H-mode pedestal, edge localized modes (ELMs) and ELM suppression by resonant magnetic perturbations, energetic particle losses, and more. The experiments used a purpose-built three-coil mock-up of two magnetized ITER TBMs in one ITER equatorial port. The largest effect was a reduction in plasma toroidal rotation velocity v across the entire radial profile by as much as Δv/v ~ 60% via non-resonant braking. Changes to global Δn/n, Δβ/β and ΔH98/H98 were ~3 times smaller. These effects are stronger at higher β. Other effects were smaller. The TBM field increased sensitivity to locking by an applied known n = 1 test field in both L- and H-mode plasmas. Locked mode tolerance was completely restored in L-mode by re-adjusting the DIII-D n = 1 error field compensation system. Numerical modelling by IPEC reproduces the rotation braking and locking semi-quantitatively, and identifies plasma amplification of a few n = 1 Fourier harmonics as the main cause of braking. IPEC predicts that TBM braking in H-mode may be reduced by n = 1 control. Although extrapolation from DIII-D to ITER is still an open issue, these experiments suggest that a TBM-like error field will produce only a few potentially troublesome problems, and that they might be made acceptably small.

  16. ITER Test Blanket Module Error Field Simulation Experiments at DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Schaffer, M. J. [General Atomics, San Diego; Testa, D. [CRPP, Switzerland; Snipes, J. A. [ITER Organization, Cadarache, France; Gohil, P. [General Atomics; De Vries, P. [Culham Centre for Fusion Energy, Culham, UK; Evans, T. E. [General Atomics, San Diego; Fenstermacher, M. E. [Lawrence Livermore National Laboratory (LLNL); Gao, X. [Academia Sinica, Institute of Plasma Physics, Hefei, China; Garofalo, A. [General Atomics, San Diego; Gates, D.A. [Princeton Plasma Physics Laboratory (PPPL); Greenfield, C. M. [General Atomics; Heidbrink, W. [University of California, Irvine; La Haye, R. [General Atomics, San Diego; Liu, S. [ASIPP, Hefei, China; Loarte, A. [ITER Organization, Cadarache, France; Nave, M. F. F. [Association EURATOM/IST, Lisbon, Portugal; Osborne, T.H. [General Atomics, San Diego; Oyama, N. [Japan Atomic Energy Agency (JAEA); Osakabe, M. [National Institute for Fusion Science, Toki, Japan; Park, J. K. [Princeton Plasma Physics Laboratory (PPPL); Ramasubramanian, N. [Institute for Plasma Research, Gandhinagar, India; Reimerdes, H. [Columbia University; Saibene, G. [Fusion for Energy (F4E), Barcelona, Spain; Saimi, A. [Aalto University, Finland; Shinohara, K. [Japan Atomic Energy Agency (JAEA), Naka; Spong, Donald A [ORNL; Solomon, W. M. [Princeton Plasma Physics Laboratory (PPPL); Tala, T. [Association Euratom-Tekes, Finland; Zhu, Y. B. [University of California, Irvine; Zhai, K. [University of Wisconsin, Madison; Boedo, J. [University of California, San Diego; Chuyanov, V. [ITER Organization, Cadarache, France; Doyle, E. J. [University of California, Los Angeles; Jakubowski, M. W. [Max-Planck-Institute for Plasmaphysik, EURATOM-Association, Greifswald, Germany; Jhang, H. [National Fusion Research Institute, Daejon, South Korea; Nazikian, Raffi [Princeton Plasma Physics Laboratory (PPPL); Pustovitov, V. D. [Russian Research Center, Kurchatov Institute, Moscow, Russia; Schmitz, O. [Forschungszentrum Julich, Julich, Germany; Sanchez, Raul [ORNL; Srinivasan, R. [Institute for Plasma Research, Gandhinagar, India; Taylor, T. S. [General Atomics, San Diego; Wade, M. [General Atomics, San Diego; You, K. I. [National Fusion Research Institute, Daejon, South Korea; Zeng, L. [University of California, Los Angeles

    2011-01-01

    Experiments at DIII-D investigated the effects of magnetic error fields similar to those expected from proposed ITER test blanket modules (TBMs) containing ferromagnetic material. Studied were effects on: plasma rotation and locking, confinement, L-H transition, the H-mode pedestal, edge localized modes (ELMs) and ELM suppression by resonant magnetic perturbations, energetic particle losses, and more. The experiments used a purpose-built three-coil mock-up of two magnetized ITER TBMs in one ITER equatorial port. The largest effect was a reduction in plasma toroidal rotation velocity v across the entire radial profile by as much as Delta upsilon/upsilon similar to 60% via non-resonant braking. Changes to global Delta n/n, Delta beta/beta and Delta H(98)/H(98) were similar to 3 times smaller. These effects are stronger at higher beta. Other effects were smaller. The TBM field increased sensitivity to locking by an applied known n = 1 test field in both L-and H-mode plasmas. Locked mode tolerance was completely restored in L-mode by re-adjusting the DIII-D n = 1 error field compensation system. Numerical modelling by IPEC reproduces the rotation braking and locking semi-quantitatively, and identifies plasma amplification of a few n = 1 Fourier harmonics as the main cause of braking. IPEC predicts that TBM braking in H-mode may be reduced by n = 1 control. Although extrapolation from DIII-D to ITER is still an open issue, these experiments suggest that a TBM-like error field will produce only a few potentially troublesome problems, and that they might be made acceptably small.

  17. Technical issues of RAFMs for the fabrication of ITER Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Tanigawa, Hiroyasu; Hirose, Takanori; Shiba, Kiyoyuki [Japan Atomic Energy Agency (JP)] (and others)

    2007-07-01

    Reduced activation ferritic/martensitic steels (RAFMs) are recognized as the primary candidate structural materials for fusion blanket systems, as it has they have been developed based on massive industrial experience of ferritic/martensitic steel replacing Mo and Nb of high chromium heat resistant martensitic steels (such as modified 9Cr-1Mo) with W and Ta, respectively. F82H and JLF-1 are RAFMs, which have been developed and studied in Japan and the various effects of irradiation were reported. F82H is designed with emphasis on high temperature property and weldability, and was provided and evaluated in various countries as a part of the IEA fusion materials development collaboration. The JAEA/US collaboration program also has been conducted with the emphasis on irradiation effects of F82H. Now, among the existing database for RAFMs the most extensive one is that for F82H. The objective of this paper is to review the R and D status of F82H and to identify the key technical issues for the fabrication of ITER Test Blanket Module (TBM) suggested from the recent achievements in Japan. It is desirable to make the status of RAFMs equivalent to commercial steels to use RAFMs as the ITER-TBM structural material. This would require demonstrating the reproducibility and weldability as well as providing the database. The excellent reproducibility of F82H has been demonstrated with four 5-ton-heats, and two of them were provided as F82H-IEA heats. It has been also proved that F82H could be provided as plates (thickness of 1.5 to 55 mm), pipes and rectangular tubes. It is also important to have the excellent weldability as the TBM has about 300m length of weld line, and it was proved through TIG, EB and YAG weld test performed in air atmosphere. Various mechanical and microstructural data have been accumulated including long-term tests such as creep rupture tests and aging tests. Although F82H is a well-perceived RAFM as the ITER-TBM structural material, some issues are

  18. Thermo-Mechanical Analyses of the High Heat Flux Component for ITER Dual Functional Lithium Lead Test Blanket Module

    Institute of Scientific and Technical Information of China (English)

    CHEN Hongli; BAI Yunqing

    2009-01-01

    The finite element code ANSYS is used to calculate the temperature and stress distributions for the first wall of DFLL-TBM (dual functional lithium lead-test blanket module),for testing in ITER. Preliminary analyses indicate that not only the low temperature design rules,the well-known 3Sm rules, are satisfied for the first wall, but the additional high temperature structural design criteria for the creep damage limits and creep-ratcheting limits are met as well.

  19. Modelling of 3D fields due to ferritic inserts and test blanket modules in toroidal geometry at ITER

    Science.gov (United States)

    Liu, Yueqiang; Äkäslompolo, Simppa; Cavinato, Mario; Koechl, Florian; Kurki-Suonio, Taina; Li, Li; Parail, Vassili; Saibene, Gabriella; Särkimäki, Konsta; Sipilä, Seppo; Varje, Jari

    2016-06-01

    Computations in toroidal geometry are systematically performed for the plasma response to 3D magnetic perturbations produced by ferritic inserts (FIs) and test blanket modules (TBMs) for four ITER plasma scenarios: the 15 MA baseline, the 12.5 MA hybrid, the 9 MA steady state, and the 7.5 MA half-field helium plasma. Due to the broad toroidal spectrum of the FI and TBM fields, the plasma response for all the n  =  1-6 field components are computed and compared. The plasma response is found to be weak for the high-n (n  >  4) components. The response is not globally sensitive to the toroidal plasma flow speed, as long as the latter is not reduced by an order of magnitude. This is essentially due to the strong screening effect occurring at a finite flow, as predicted for ITER plasmas. The ITER error field correction coils (EFCC) are used to compensate the n  =  1 field errors produced by FIs and TBMs for the baseline scenario for the purpose of avoiding mode locking. It is found that the middle row of the EFCC, with a suitable toroidal phase for the coil current, can provide the best correction of these field errors, according to various optimisation criteria. On the other hand, even without correction, it is predicted that these n  =  1 field errors will not cause substantial flow damping for the 15 MA baseline scenario.

  20. Irradiation behavior of Ti 4Al 2V (ΠT-3B) alloy for ITER blanket modules flexible attachment

    Science.gov (United States)

    Rodchenkov, B. S.; Kozlov, A. V.; Kuznetsov, Yu. G.; Kalinin, G. M.; Strebkov, Yu. S.

    2007-08-01

    Titanium alloys are recommended as a material to manufacture flexible attachments of the shield blanket modules in the ITER reactor owing to their advantageous combination of properties, i.e., high resistance to impact loading, strength, density and low thermal expansion coefficient. An additional factor for selecting Ti alloys is their fast induced radioactivity decay. The (α + β)-Ti alloys have higher strength than (α)-Ti alloys but are less developed. The data base on the irradiation behavior of these materials is limited. Neutron irradiation of (α)-Ti-4Al-2V (ΠT-3B) alloy has been performed in the framework of the ITER R&D programme. Specimens from a forging of Ti-4Al-2V alloy were irradiated in the IVV-2M reactor to doses of (0.32-0.43) dpa at temperatures of (240-260) °C. This paper describes the results of tensile, low cycle fatigue and fracture toughness tests of alloy in the unirradiated and neutron irradiated conditions. The results obtained are compared with those of the (α + β)-Ti-6Al-4V alloy.

  1. DEMO blanket testing in ITER. Influence on reaching DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Shatalov, G. E-mail: geshat@nfi.kiae.ru

    2001-10-01

    ITER goal was specified as one step between now and the DEMO fusion reactor. One of the major issues is the tritium breeding blankets test relevant to future reactors. The major objectives of blanket modules (TBM) experiments in ITER are reduced in comparison with proposed test objectives in ITER-FDR. Thus, results of DEMO blanket designs testing in ITER will provide limited (but still useful) information that will need strong support from non-fusion facilities testing. The role of non-fusion tests is increased now to provide additional data required for DEMO blanket construction and qualification. A strategy of testing steps to DEMO blanket qualifications has to include parallel testing in ITER and in non-fusion devices. Experiments in fission reactors are able to provide essential data on materials radiation properties; tritium release, inventory and permeation; and thermomechanical behavior of the blanket breeder/multiplier. However, the volume in fission reactors is rather small and neutron spectra differ from the fusion reactor one. Nonetheless in the near future one depends primarily on fission reactor irradiation. The powerful accelerator based neutron source IFMIF could also provide useful information on radiation material properties. Plasma based neutron sources of different fusion devices could be the best choice for testing DEMO materials and blanket mock-ups. Timetable and costs of these devices are not clear now.

  2. Current status of technology development for fabrication of Indian Test Blanket Module (TBM) of ITER

    Energy Technology Data Exchange (ETDEWEB)

    Jayakumar, T., E-mail: tjk@igcar.gov.in [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam 603102 (India); Rajendra Kumar, E. [TBM Division, Institute for Plasma Research (IPR), Bhat, Gandhinagar 382428 (India)

    2014-10-15

    Highlights: • Status of technology developments for Indian TBM to be installed in ITER is presented. • Procedure development for EB, laser and laser-hybrid welding of RAFM steel presented. • Filler wires for RAFM steel for TIG, NG-TIG and laser-hybrid welding have been developed. • Feasibility of production of channel plate by HIP technology has been demonstrated. - Abstract: Ever since India decided to install its Lead-Lithium Ceramic Breeder (LLCB) TBM in ITER, various technologies for fabrication of Indian TBM are being pursued by IPR and IGCAR, in collaboration with various research laboratories in India. Welding consumables for joining India specific RAFM steels (IN-RAFMS), procedures for hot isostatic pressing, electron beam welding, laser and laser-hybrid welding have been developed. Considering the complex nature and limited access available for inspection, innovative inspection procedures that involved use of phased array ultrasonic and C-scan imaging are also being pursued. This paper presents the current status of these developments and provides a roadmap for the future activities planned in realizing Indian TBM for testing in ITER.

  3. Progress in design and study of ITER test blanket modules%ITER氚增殖实验包层设计研究进展

    Institute of Scientific and Technical Information of China (English)

    刘松林; 柏云清; 陈红丽; 李春京; 黄群英; 吴宜灿; FDS团队

    2009-01-01

    The International Thermonuclear Experimental Reactor (ITER) will be the first experimental D-T fusion reactor to provide an exclusive test platform of physics and engineering technology for research and development of fusion, where the technology of Test Blanket Module (TBM) in ITER is one of the most critical kernels to achieve fusion power in the future. According to defined concepts of DEMO blanket, the parties had proposed DEMOrelevant TBM, respectively, which would be to be tested during ITER operation. Design of proposed TBM concepts, R&D status, and recommended port allocation in ITER are introduced in this contribution.%国际热核实验反应堆(ITER)为人类开发聚变能提供重要的物理和工程技术实验平台,ITER氚增殖实验包层模块(TBM)技术是必须掌握的关键技术.参与ITER计划的成员国根据本国商用演示堆包层发展策略,分别提出了各自的实验包层概念,以便在ITER运行期间进行实验.本文对ITER-TBM目前已经开展和正在进行的主要设计研究工作进展进行总结,介绍了各方提出的设计方案、支撑设计的相关技术研究进展,以及合作实验窗口的分配现状.

  4. The ITER EC H&CD upper launcher: Design, analysis and testing of a bolted joint for the Blanket Shield Module

    NARCIS (Netherlands)

    Gessner, R.; Aiello, G.; Grossetti, G.; Meier, A.; Ronden, D.; Spaeh, P.; Scherer, T.; Schreck, S.; Strauss, D.; Vaccaro, A.

    2013-01-01

    The final design of the structural system for the ITER EC H&CD upper launcher is in progress. Many design features of the preliminary design are under revision with the aim to achieve the built-to-print-status. This paper deals with design and analysis of a bolted joint for the Blanket Shield Mo

  5. Breeding blanket design for ITER and prototype (DEMO) fusion reactors and breeding materials issues

    Energy Technology Data Exchange (ETDEWEB)

    Takatsu, H.; Enoeda, M. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1998-03-01

    Current status of the designs of the ITER breeding blanket and DEMO blankets is introduced placing emphasis on the breeding materials selection and related issues. The former design is based on the up-to-date design activities, as of October 1997, being performed jointly by Joint Central Team (JCT) and Home Teams (HT`s), while the latter is based on the DEMO blanket test module designs being proposed by each Party at the TBWG (Test Blanket Working Group) meetings. (J.P.N.)

  6. Construction of a test platform for Test Blanket Module (TBM) systems integration and maintenance in ITER Port Cell #16

    Energy Technology Data Exchange (ETDEWEB)

    Vála, Ladislav, E-mail: ladislav.vala@cvrez.cz [Centrum výzkumu Řež, Hlavní 130, 250 68 Husinec-Řež (Czech Republic); Reungoat, Mathieu, E-mail: mathieu.reungoat@cvrez.cz [Centrum výzkumu Řež, Hlavní 130, 250 68 Husinec-Řež (Czech Republic); Vician, Martin [Centrum výzkumu Řež, Hlavní 130, 250 68 Husinec-Řež (Czech Republic); Poitevin, Yves; Ricapito, Italo; Zmitko, Milan; Panayotov, Dobromir [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain)

    2015-10-15

    Highlights: • A non-nuclear, full size facility – TBM platform – is under construction in CVR. • It is designed for tests, optimization and validation of TBS maintenance operations. • It will allow testing and validation of specific maintenance tools and RH equipment. • It reproduces ITER Port Cell #16, as well as the TBS interfaces and main equipment. • The TBM platform will be available for full operation in the first half of 2016. - Abstract: This paper describes a project of a non-nuclear, 1:1 scale testing platform dedicated to tests, optimization and validation of integration and maintenance operations for the European TBM systems in the ITER Port Cell #16. This TBM platform is currently under construction in Centrum výzkumu Řež, Czech Republic. The facility is realized within the scope of the SUSEN project and its full operation is foreseen in the first half of 2016.

  7. Materials development for ITER shielding and test blanket in China

    Energy Technology Data Exchange (ETDEWEB)

    Chen, J.M., E-mail: Chenjm@swip.ac.cn [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China); Wu, J.H.; Liu, X.; Wang, P.H. [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China); Wang, Z.H.; Li, Z.N. [Ningxia Orient Non-ferrous Metals Group Co. Ltd., P.O. Box 105, Shizuishan (China); Wang, X.S.; Zhang, P.C. [China Academy of Engineering Physics, P.O. Box 919-71, Mianyang 621900 (China); Zhang, N.M.; Fu, H.Y.; Liu, D.H. [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China)

    2011-10-01

    China is a member of the ITER program and is developing her own materials for its shielding and test blanket modules. The materials include vacuum-hot-pressing (VHP) Be, CuCrZr alloy, 316L(N) and China low activation ferritic/martensitic (CLF-1) steels. Joining technologies including Be/Cu hot isostatic pressing (HIP) and electron beam (EB) weldability of 316L(N) were investigated. Chinese VHP-Be showed good properties, with BeO content and ductility that satisfy the ITER requirements. Be/Cu mock-ups were fabricated for Be qualification tests at simulated ITER vertical displacement event (VDE) and heat flux cycling conditions. Fine microstructure and good mechanical strength of the CuCrZr alloy were achieved by a pre-forging treatment, while the weldability of 316L(N) by EB was demonstrated for welding depths varying from 5 to 80 mm. Fine microstructure, high strength, and good ductility were achieved in CLF-1 steel by an optimized normalizing, tempering and aging procedure.

  8. The ITER EC H and CD upper launcher: Design, analysis and testing of a bolted joint for the Blanket Shield Module

    Energy Technology Data Exchange (ETDEWEB)

    Gessner, Robby, E-mail: robby.gessner@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany); Aiello, Gaetano; Grossetti, Giovanni; Meier, Andreas [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany); Ronden, Dennis [DIFFER – Dutch Institute for Fundamental Energy Physics, P.O. Box 1207, NL-3430 BE Nieuwegein (Netherlands); Spaeh, Peter; Scherer, Theo; Schreck, Sabine; Strauss, Dirk; Vaccaro, Alessandro [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany)

    2013-10-15

    Highlights: ► The BSM of the ECH Launcher is attached to the Launcher Main Frame by a bolted joint. ► The bolts were designed as “captive” in order to avoid their accidental removal from the joint. ► The bolted flange connection using two sets of 15 captive bolts (M22 × 2) placed along the sides. ► The captive bolt design is based on a concept that uses a dedicated spring ring, a standard spiral spring and a tensioning screw with two threads to secure the bolts in a form-locking stop. -- Abstract: The final design of the structural system for the ITER EC H and CD upper launcher is in progress. Many design features of the preliminary design are under revision with the aim to achieve the built-to-print-status. This paper deals with design and analysis of a bolted joint for the Blanket Shield Module with special perspective on Remote Handling capability. The BSM of the ECH Launcher is attached to the Launcher Main Frame by a bolted joint conceived so that in the Hot Cell Facility, RH maintenance can be performed on internal components. The joint must be capable to resist very high Electro-Magnetic loads from disruptions, while it has to sustain substantial thermal cycling during operation. Thus the need for a rigid and reliable design is essential. Beside the set of pre-stressed bolts the flanges were therefore equipped with additional shear keys to divert radial moments away from the bolts. Main focus of the work performed was the mechanical design of the joint and the assessment of the structural integrity with respect to the loads applied and its capability for maintenance by RH procedures. To fulfill a major aspect of the RH requirements, the bolts were designed as “captive” in order to avoid their accidental removal from the joint. The captive bolt design is based on a concept that uses a dedicated spring ring, a standard spiral spring and a tensioning screw with two threads to secure the bolts in a form-locking stop. The final approval phase of

  9. An assessment of the base blanket for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Raffray, A.R.; Abdou, M.A.; Ying, A.

    1991-12-31

    Ideally, the ITER base blanket would provide the necessary tritium for the reactor to be self-sufficient during operation, while having minimal impact on the overall reactor cost, reliability and safety. A solid breeder blanket has been developed in CDA phase in an attempt to achieve such objectives. The reference solid breeder base blanket configurations at the end of the CDA phase has many attractive features such as a tritium breeding ratio (TBR) of 0.8--0.9 and a reasonably low tritium inventory. However, some concerns regarding the risk, cost and benefit of the base blanket have been raised. These include uncertainties associated with the solid breeder thermal control and the potentially high cost of the amount of Be used to achieve high TBR and to provide the necessary thermal barrier between the high temperature solid breeder and low temperature coolant. This work addresses these concerns. The basis for the selection of a breeding blanket is first discussed in light of the incremental risk, cost and benefits relative to a non-breeding blanket. Key issues associated with the CDA breeding blanket configurations are then analyzed. Finally, alternative schemes that could enhance the attractiveness and flexibility of a breeding blanket are explored.

  10. An assessment of the base blanket for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Raffray, A.R.; Abdou, M.A.; Ying, A.

    1991-01-01

    Ideally, the ITER base blanket would provide the necessary tritium for the reactor to be self-sufficient during operation, while having minimal impact on the overall reactor cost, reliability and safety. A solid breeder blanket has been developed in CDA phase in an attempt to achieve such objectives. The reference solid breeder base blanket configurations at the end of the CDA phase has many attractive features such as a tritium breeding ratio (TBR) of 0.8--0.9 and a reasonably low tritium inventory. However, some concerns regarding the risk, cost and benefit of the base blanket have been raised. These include uncertainties associated with the solid breeder thermal control and the potentially high cost of the amount of Be used to achieve high TBR and to provide the necessary thermal barrier between the high temperature solid breeder and low temperature coolant. This work addresses these concerns. The basis for the selection of a breeding blanket is first discussed in light of the incremental risk, cost and benefits relative to a non-breeding blanket. Key issues associated with the CDA breeding blanket configurations are then analyzed. Finally, alternative schemes that could enhance the attractiveness and flexibility of a breeding blanket are explored.

  11. Reduced activation martensitic steels as a structural material for ITER test blanket

    Energy Technology Data Exchange (ETDEWEB)

    Shiba, K. E-mail: shiba@realab01.tokai.jaeri.go.jp; Enoeda, M.; Jitsukawa, S

    2004-08-01

    A Japanese ITER test blanket module (TBM) is planed to use reduced-activation martensitic steel F82H. Feasibility of F82H for ITER test blanket module is discussed in this paper. Several kinds of property data, including physical properties, magnetic properties, mechanical properties and neutron-irradiation data on F82H have been obtained, and these data are complied into a database to be used for the designing of the ITER TBM. Currently obtained data suggests F82H will not have serious problems for ITER TBM. Optimization of F82H improves the induced activity, toughness and HIP resistance. Furthermore, modified F82H is resistant to temperature instability during material production.

  12. Analysis of the steady state hydraulic behaviour of the ITER blanket cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, P.A., E-mail: pietroalessandro.dimaio@unipa.it [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Dell’Orco, G.; Furmanek, A. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Garitta, S. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Merola, M.; Mitteau, R.; Raffray, R. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Spagnuolo, G.A.; Vallone, E. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy)

    2015-10-15

    Highlights: • Nominal steady state hydraulic behaviour of ITER blanket standard sector cooling system has been investigated. • Numerical simulations have been run adopting a qualified thermal-hydraulic system code. • Hydraulic characteristic functions and coolant mass flow rates, velocities and pressure drops have been assessed. • Most of the considered circuits are able to effectively cool blanket modules, meeting ITER requirements. - Abstract: The blanket system is the ITER reactor component devoted to providing a physical boundary for plasma transients and contributing to thermal and nuclear shielding of vacuum vessel, magnets and external components. It is expected to be subjected to significant heat loads under nominal conditions and its cooling system has to ensure an adequate cooling, preventing any risk of critical heat flux occurrence while complying with pressure drop limits. At the University of Palermo a study has been performed, in cooperation with the ITER Organization, to investigate the steady state hydraulic behaviour of the ITER blanket standard sector cooling system. A theoretical–computational approach based on the finite volume method has been followed, adopting the RELAP5 system code. Finite volume models of the most critical blanket cooling circuits have been set-up, realistically simulating the coolant flow domain. The steady state hydraulic behaviour of each cooling circuit has been investigated, determining its hydraulic characteristic function and assessing the spatial distribution of coolant mass flow rates, velocities and pressure drops under reference nominal conditions. Results obtained have indicated that the investigated cooling circuits are able to provide an effective cooling to blanket modules, generally meeting ITER requirements in term of pressure drop and velocity distribution, except for a couple of circuits that are being revised.

  13. Development of ITER shielding blanket prototype mockup by HIP bonding

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Satoshi; Furuya, Kazuyuki; Hatano, Toshihisa; Kuroda, Toshimasa; Enoeda, Mikio; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Takatsu, Hideyuki [Japan Atomic Energy Research Inst., Office of ITER Project Promotion, Tokyo (Japan)

    2000-07-01

    A prototype ({approx}900{sup H} x 1700{sup W} x 350{sup T} mm) of the ITER shielding blanket module has been fabricated following the previous successful fabrication of a small-scale ({approx}500{sup H} x 400{sup W} x 150{sup T} mm) and mid-scale ({approx}800{sup H} x 500{sup W} x 350{sup T} mm) mock-ups. This prototype incorporates most of key design features essential to the fabrication of the ITER shielding blanket module such as 1) the first wall heat sink made of Al{sub 2}O{sub 3} dispersion strengthened Cu (DSCu) with built-in SS316L coolant tubes bonded to a massive SS316LN shield block, 2) toroidally curved first wall with a radius of 5106 mm while straight in poloidal direction, 3) coolant channels oriented in poloidal direction in the first wall and in toroidal direction in the shield block, 4) the first wall coolant channel routing to avoid the interference with the front access holes, 5) coolant channels drilled through the forged SS316LN-IG shield block, and 6) four front access holes of 30 mm in diameter penetrated through the first wall and the shield block. For the joining method, especially for the first wall/side wall parts and the shield block, the solid HIP (Hot Isostatic Pressing) process was applied. It is difficult to apply conventional joining methods such as field welding, brazing, explosion bonding and mechanical one-axial diffusion bonding to a wide area bonding because sufficient mechanical strengths can not be obtained and excessive deformations occurs. In order to solve these fabrication issues, HIP bonding was applied. The first wall stainless steel (SS) coolant tubes of 10 mm in inner diameter and l mm in thickness were sandwiched by semi-circular grooved DSCu plates at the first wall and the front region of the side wall, and by semi-circular grooved SS plates at the back region of the side wall. After assembling of these first wall/side wall parts with the shield block, they were simultaneously bonded by single step HIP in order to

  14. ITER solid breeder blanket materials database

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.C. [Argonne National Lab., IL (United States); Dienst, W. [Kernforschungszentrum Karlsruhe GmbH (Germany). Inst. fuer Material- und Festkoerperforschung; Flament, T. [CEA Centre d`Etudes de Fontenay-aux-Roses (France). Commissariat A L`Energie Atomique; Lorenzetto, P. [NET Team, Garching (Germany); Noda, K. [Japan Atomic Energy Research Inst., Takai, Ibaraki, (Japan); Roux, N. [CEA Centre d`Etudes et de Recherches Les Materiaux (France). Commissariat a L`Energie Atomique

    1993-11-01

    The databases for solid breeder ceramics (Li{sub 2},O, Li{sub 4}SiO{sub 4}, Li{sub 2}ZrO{sub 3} and LiAlO{sub 2}) and beryllium multiplier material are critically reviewed and evaluated. Emphasis is placed on physical, thermal, mechanical, chemical stability/compatibility, tritium, and radiation stability properties which are needed to assess the performance of these materials in a fusion reactor environment. Correlations are selected for design analysis and compared to the database. Areas for future research and development in blanket materials technology are highlighted and prioritized.

  15. Ceramic helium-cooled blanket test module

    Energy Technology Data Exchange (ETDEWEB)

    Leshukov, A. E-mail: leshu@entek.ru; Kovalenko, V.; Shatalov, G.; Goroshkin, G.; Obukhov, A

    2000-11-01

    The design of RF DEMO-relevant ceramic helium cooled blanket test module (CHC BTM) for testing in international thermonuclear experimental reactor (ITER) is under consideration. The RF concept of DEMO BTM is based upon the breeder inside tube (BIT)-concept. This concept suggests the use of solid breeding ceramic material, helium as coolant and tritium purge-gas, ferrite-martensite steel as structural material, and beryllium as neutron multiplier. The parameters of the primary circuit coolant are the following, pressure -8 MPa, inlet/outlet temperature -300/550 deg. C, respectively. Helium (0.1 MPa pressure) is used for tritium removal from ceramic breeder. The ITER water coolant is the secondary circuit coolant of DEMO BTM cooling system. Lithium orthosilicate (Li{sub 4}SiO{sub 4}) is used as tritium breeding material (pebbles-bed of diameter 0.5-1 mm spheres). It is planned to use the beryllium as neutron multiplier (spheres diameter 1 mm pebbles-bed or the porous beryllium). The 3-D neutronic calculations on Monte Carlo method, in accordance with FENDL-1 library of the nuclear data, have been performed for CHC BTM. To validate the CHC BTM concept, the thermal hydraulic analysis has been performed for the design elements and cooling system equipment. The preliminary stress analysis for BTM design elements has been carried out on the ASME-code and RF strength regulations. The four types of LOFA and LOCA accidents have been investigated. The parameters of cooling, coolant purification and tritium extraction systems have been determined.

  16. Assessment of alkali metal coolants for the ITER blanket

    Science.gov (United States)

    Natesan, K.; Reed, C. B.; Mattas, R. F.

    1994-06-01

    The blanket system is one of the most important components of a fusion reactor because it has a major impact on both the economics and safety of fusion energy. The primary functions of the blanket in a deuterium/tritium-fueled fusion reactor are to convert the fusion energy into sensible heat and to breed tritium for the fuel cycle. The blanket comparison and selection study, conducted earlier, described the overall comparative performance of different blanket concepts, including liquid metal, molten salt, water, and helium. This paper will discuss the ITER requirements for a self-cooled blanket concept with liquid lithium and for indirectly cooled concepts that use other alkali metals such as NaK. The paper addresses the thermodynamics of interactions between the liquid metals (e.g., lithium and NaK) and structural materials (e.g., V-base alloys), together with associated corrosion/compatibility issues. Available experimental data are used to assess the long-term performance of the first wall in a liquid metal environment. Other key issues include development of electrical insulator coatings on the first-wall structural material to MHD pressure drop, and tritium permeation/inventory in self-cooled and indirectly cooled concepts. Acceptable types of coatings (based on their chemical compatibility and physical properties) are identified, and surface-modification avenues to achieve these coatings on the first wall are discussed. The assessment examines the extent of our knowledge on structural materials performance in liquid metals and identifies needed research and development in several of the areas in order to establish performance envelopes for the first wall in a liquid-metal environment.

  17. The State of the Art Report on the Development and Manufacturing Technology of Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. S.; Jeong, Y. H.; Park, S. Y.; Lee, M. H.; Choi, B. K.; Baek, J. H.; Park, J. Y.; Kim, J. H.; Kim, H. G.; Kim, K. H

    2006-07-15

    The main objective of the present R and D on breeder blanket is the development of test blanket modules (TBMs) to be installed and tested in International Thermonuclear Experimental Reactor (ITER). In the program of the blanket development, a blanket module test in the ITER is scheduled from the beginning of the ITER operation, and the performance test of TBM in ITER is the most important milestone for the development of the DEMO blanket. The fabrication of TBMs has been required to test the basic performance of the DEMO blanket, i.e., tritium production/recovery, high-grade heat generation and radiation shielding. Therefore, the integration of the TBM systems into ITER has been investigated with the aim to check the safety, reliability and compatibility under nuclear fusion state. For this reason, in the Test Blanket Working Group (TBWG) as an activity of the International Energy Association (IEA), a variety of ITER TBMs have been proposed and investigated by each party: helium-cooled ceramic (WSG-1), helium-cooled LiPb (WSG-2), water-cooled ceramic (WSG-3), self-cooled lithium (WSG-4) and self-cooled molten salt (WSG-5) blanket systems. Because we are still deficient in investigation of TBM development, the need of development became pressing. In this report, for the development of TBM sub-module and mock-up, it is necessary to analyze and examine the state of the art on the development of manufacturing technology of TBM in other countries. And we will be applied as basic data to establish a manufacturing technology.

  18. U.S. technical report for the ITER blanket/shield: A. blanket: Topical report, July 1990--November 1990

    Energy Technology Data Exchange (ETDEWEB)

    1995-01-01

    Three solid-breeder water-cooled blanket concepts have been developed for ITER based on a multilayer configuration. The primary difference among the concepts is in the fabricated form of breeder and multiplier. All the concepts have beryllium for neutron multiplication and solid-breeder temperature control. The blanket design does not use helium gaps or insulator material to control the solid breeder temperature. Lithium oxide (Li{sub 2}O) and lithium zirconate (Li{sub 2}ZrO{sub 3}) are the primary and the backup breeder materials, respectively. The lithium-6 enrichment is 95%. The use of high lithium-6 enrichment reduces the solid breeder volume required in the blanket and consequently the total tritium inventory in the solid breeder material. Also, it increases the blanket capability to accommodate power variation. The multilayer blanket configuration can accommodate up to a factor of two change in the neutron wall loading without violating the different design guidelines. The blanket material forms are sintered products and packed bed of small pebbles. The first concept has a sintered product material (blocks) for both the beryllium multiplier and the solid breeder. The second concept, the common ITER blanket, uses a packed bed breeder and beryllium blocks. The last concept is similar to the first except for the first and the last beryllium zones. Two small layers of beryllium pebbles are located behind the first wall and the back of the last beryllium zone to reduce the total inventory of the beryllium material and to improve the blanket performance. The design philosophy adopted for the blanket is to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. Also, the reliability and the safety aspects of the blanket are enhanced by using low-pressure water coolant and the separation of the tritium purge flow from the coolant system by several barriers.

  19. Overview on ITER and DEMO blanket fabrication activities of the KIT INR and related frameworks

    Energy Technology Data Exchange (ETDEWEB)

    Neuberger, Heiko, E-mail: heiko.neuberger@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology, Karlsruhe (Germany); Rey, Joerg; Weth, Axel von der; Hernandez, Francisco [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology, Karlsruhe (Germany); Martin, Tatiana [Karlsruhe Institute of Technology (KIT), Institute for Applied materials, Karlsruhe (Germany); Zmitko, Milan [Fusion for energy, ITER Department, Test Blanket Modules and Materials Development Project Team, Barcelona (Spain); Felde, Alexander [Institut für Umformtechnik (IFU), Universität Stuttgart (Germany); Niewöhner, Reinhard [Forschungsgesellschaft Umformtechnik (FGU), Stuttgart (Germany); Krüger, Friedhelm [Krüger Erodiertechnik, Biedenkopf (Germany)

    2015-10-15

    Highlights: • Recent achievements in fabricaition within different frameworks. • First Wall mockup with erosion technology. • Manufacturing of a HCPB TBM Cooling Plate Mockup (F4E) - Abstract: Fabrication experiments have been carried out in the KIT with the goal to qualify manufacturing technologies for the realization of fusion reactor components. The main focus of the activities managed by the fabrication team in the Institute of Neutron Physics and reactor technologies (INR) has been on the Test Blanket Module for ITER. Sets of fabrication and welding procedure specifications have been demonstrated and qualified in relevant scale for TBM structural and functional components. This paper presents interactions in between the different frameworks on domestic and European level to underline backgrounds of developments. It also summarizes results of development and their relevancy for DEMO and gives an outlook on the future development strategy for the DEMO blanket fabrication.

  20. Thermal-Hydraulic System Study of the Helium Cooled Pebble Bed (HCPB) Test Blanket Module (TBM) for ITER Using System Code RELAP5

    Institute of Scientific and Technical Information of China (English)

    Jin Xuezhou; R. Meyder

    2005-01-01

    The HCPB concept has been a European DEMO reference concept for nearly one decade. Detailed thermal-hydraulic study on the control behavior of the whole system is one of the important parts of this development. The thermal-hydraulic effect of the TBM-combined cooling circuit during a cyclic operation in ITER has been studied using the system code RELAP5. The RELAP5 is based on an one-dimensional, transient two-fluid model for the flow of a two-phase steam-water mixture that can contain noncondensable components like Helium. The RELAP5-models are modified to take the cyclic operation of the circulator, heat exchanger, bypass, valves etc in to account. A sequence of operational phases is investigated, starting from the cold state through the heating phase that brings the system to a stand-by condition, followed by typical power cycles applied in ITER. The results show that the implemented control mechanisms keep the inlet temperature to the TBM and the total mass flow rate at the required values through all phases.

  1. Thermal-Hydraulic System Study of the Helium Cooled Pebble Bed (HCPB) Test Blanket Module (TBM) for ITER Using System Code RELAP5

    Science.gov (United States)

    Jin, Xuezhou; R, Meyder

    2005-04-01

    The HCPB concept has been a European DEMO reference concept for nearly one decade. Detailed thermal-hydraulic study on the control behavior of the whole system is one of the important parts of this development. The thermal-hydraulic effect of the TBM-combined cooling circuit during a cyclic operation in ITER has been studied using the system code RELAP5. The RELAP5 is based on an one-dimensional, transient two-fluid model for the flow of a two-phase steam-water mixture that can contain noncondensable components like Helium. The RELAP5-models are modified to take the cyclic operation of the circulator, heat exchanger, bypass, valves etc in to account. A sequence of operational phases is investigated, starting from the cold state through the heating phase that brings the system to a stand-by condition, followed by typical power cycles applied in ITER. The results show that the implemented control mechanisms keep the inlet temperature to the TBM and the total mass flow rate at the required values through all phases.

  2. Development of Joining Technologies for the ITER Blanket First Wall

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Byoung Kwon; Jung, Yang Il; Park, Dong Jun; Kim, Hyun Gil; Park, Sang Yoon; Park, Jeong Yong; Jeong, Yong Hwan; Lee, Dong Won; Kim, Suk Kwon [KAERI, Daejeon (Korea, Republic of)

    2011-01-15

    The design of the ITER blanket first wall includes the Beryllium amour tiles joined to CuCrZr heat sink with stainless steel cooling tubes. For the ITER application, the Be/CuCrZr/SS joint was proposed as a first wall material. The joining of Be/CuCrZr as well as CuCrZr/SS was generally carried out by using a hot isostatic pressing (CuC) in many countries. The joining strength for Be/CuCrZr is relatively lower than that for CuCrZr/SS, since we usually forms surface oxides (BeO) and brittle a metallics with Cu. Therefore, the joining technology for the Be/CuCrZr joint has been investigated. Be is apt to adsorb oxygen in an air atmosphere, so we should be etched to eliminate the surface pre-oxide using a chemical solution and Ar ions in a vacuum chamber. Then we is coated with a first was to prevent further oxidation. The kinds of a first we are chosen to be able to enhance the joining strength as inhibiting excessive be diffusion. The performance of the Be/CuCrZr/SS joint used for the ITER first wall is primarily dependent on the joining strength of the Be/CuCrZr interface. The Cr/Cu and Ti/Cr/Cu interlayers enabled the successful joining of be tile to CuCrZr plate. Moreover, ion-beam assisted deposition (IBAD) increased joining strength of the Be/CuCrZr joint mock-ups. IBAD induced the increased packing of depositing atoms, which resulted in denser and more adhesive interlayers. The interlayers formed by IBAD process revealed about 40% improved resistance to the scratch test. It is suggested that the improved adhesion of coating interlayers enabled tight joining of Be and CuCrZr blocks. As compared to without IBAD coating, the shear strength as well as the 4-point bend strength were increased more than 20% depending on interlayer types and coating conditions

  3. ITER Blanket First Wall (WBS 1.6{sub 1}A)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Bong Guen; Kim, H. G.; Kim, J. H. (and others)

    2008-03-15

    -up fabrication was started; Cu/SS joints were fabricated and purchase of Be tiles was prepared. Fabrication manual and test manual such as mechanical tests and NDE were documented in the form of the TSD. Based on the design by the ITER-O, 3D modeling of the module no. 4 for ITER blanket FW was produced, thermal-hydraulic and thermo-mechanical analysis were performed. The developed NDE methods were applied to all fabricated mock-ups before HHF test and the UT results were compared with the IR images, which were generated when screening test during HHF test. ECT probes were prepared according to the previous simulation results and they were evaluated experimentally with the NDT mock-up, which has artificial defects. The developed NDE methods and their application were documented as an inspection manual and a QC document, and they were included in the TS000.

  4. ITER Central Solenoid Module Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Smith, John [General Atomics, San Diego, CA (United States)

    2016-09-23

    The fabrication of the modules for the ITER Central Solenoid (CS) has started in a dedicated production facility located in Poway, California, USA. The necessary tools have been designed, built, installed, and tested in the facility to enable the start of production. The current schedule has first module fabrication completed in 2017, followed by testing and subsequent shipment to ITER. The Central Solenoid is a key component of the ITER tokamak providing the inductive voltage to initiate and sustain the plasma current and to position and shape the plasma. The design of the CS has been a collaborative effort between the US ITER Project Office (US ITER), the international ITER Organization (IO) and General Atomics (GA). GA’s responsibility includes: completing the fabrication design, developing and qualifying the fabrication processes and tools, and then completing the fabrication of the seven 110 tonne CS modules. The modules will be shipped separately to the ITER site, and then stacked and aligned in the Assembly Hall prior to insertion in the core of the ITER tokamak. A dedicated facility in Poway, California, USA has been established by GA to complete the fabrication of the seven modules. Infrastructure improvements included thick reinforced concrete floors, a diesel generator for backup power, along with, cranes for moving the tooling within the facility. The fabrication process for a single module requires approximately 22 months followed by five months of testing, which includes preliminary electrical testing followed by high current (48.5 kA) tests at 4.7K. The production of the seven modules is completed in a parallel fashion through ten process stations. The process stations have been designed and built with most stations having completed testing and qualification for carrying out the required fabrication processes. The final qualification step for each process station is achieved by the successful production of a prototype coil. Fabrication of the first

  5. Numerical simulation of the transient thermal-hydraulic behaviour of the ITER blanket cooling system under the draining operational procedure

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, P.A. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128 Palermo (Italy); Dell’Orco, G.; Furmanek, A. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Garitta, S. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128 Palermo (Italy); Merola, M.; Mitteau, R.; Raffray, R. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Spagnuolo, G.A. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128 Palermo (Italy); Vallone, E., E-mail: eug.vallone@gmail.com [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128 Palermo (Italy)

    2015-10-15

    Highlights: • ITER blanket cooling system hydraulic behaviour is studied under draining transient. • A computational approach based on the finite volume method has been followed. • Draining efficiency has been assessed in term of transient duration and residual water. • Transient duration ranges from ∼40 to 50 s, under the reference draining scenario. • Residual water is predicted to range from few tens of gram up to few kilograms. - Abstract: Within the framework of the research and development activities supported by the ITER Organization on the blanket system issues, an intense analysis campaign has been performed at the University of Palermo with the aim to investigate the thermal-hydraulic behaviour of the cooling system of a standard 20° sector of ITER blanket during the draining transient operational procedure. The analysis has been carried out following a theoretical-computational approach based on the finite volume method and adopting the RELAP5 system code. In a first phase, attention has been focused on the development and validation of the finite volume models of the cooling circuits of the most demanding modules belonging to the standard blanket sector. In later phase, attention has been put to the numerical simulation of the thermal-hydraulic transient behaviour of each cooling circuit during the draining operational procedure. The draining procedure efficiency has been assessed in terms of both transient duration and residual amount of coolant inside the circuit, observing that the former ranges typically between 40 and 120 s and the latter reaches at most ∼8 kg, in the case of the cooling circuit of twinned modules #6–7. Potential variations to operational parameters and/or to circuit lay-out have been proposed and investigated to optimize the circuit draining performances. In this paper, the set-up of the finite volume models is briefly described and the key results are summarized and critically discussed.

  6. Development of the Water Cooled Ceramic Breeder Test Blanket Module in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Enoeda, Mikio, E-mail: enoeda.mikio@jaea.go.jp [Japan Atomic Energy Agency, Naka-shi, Ibaraki-ken 311-0193 (Japan); Tanigawa, Hisashi; Hirose, Takanori; Suzuki, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Yamanishi, Toshihiko; Hoshino, Tsuyoshi; Nakamichi, Masaru; Tanigawa, Hiroyasu; Ezato, Koichiro; Seki, Yohji; Yoshikawa, Akira; Tsuru, Daigo; Akiba, Masato [Japan Atomic Energy Agency, Naka-shi, Ibaraki-ken 311-0193 (Japan)

    2012-08-15

    The development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. For the TBM testing and evaluation toward DEMO blanket, the module fabrication technology development by a candidate structural material, reduced activation martensitic/ferritic steel, F82H, is one of the most critical items from the viewpoint of realization of TBM testing in ITER. In Japan, fabrication of a real scale first wall, side walls, a breeder pebble bed box and assembling of the first wall and side walls have succeeded. Recently, the real scale partial mockup of the back wall was fabricated. The fabrication procedure of the back wall, whose thickness is up to 90 mm, was confirmed toward the fabrication of the real scale back wall by F82H. Important key technologies are almost clarified for the fabrication of the real scale TBM module mockup. From the view point of testing and evaluation, development of the technology of the blanket tritium recovery, development of advanced breeder and multiplier pebbles and the development of the blanket neutronics measurement technology are also performed. Also, tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been started as the verification test of tritium production performance. This paper overviews the recent achievements of the development of the WCCB TBM in Japan.

  7. ITER (International Thermonuclear Experimental Reactor) shield and blanket work package report

    Energy Technology Data Exchange (ETDEWEB)

    1988-06-01

    This report summarizes nuclear-related work in support of the US effort for the International Thermonuclear Experimental Reactor (ITER) Study. The purpose of this work was to prepare for the first international ITER workshop devoted to defining a basic ITER concept that will serve as a basis for an indepth conceptual design activity over the next 2-1/2 years. Primary tasks carried out during the past year included: design improvements of the inboard shield developed for the TIBER concept, scoping studies of a variety of tritium breeding blanket options, development of necessary design guidelines and evaluation criteria for the blanket options, further safety considerations related to nuclear components and issues regarding structural materials for an ITER device. 44 refs., 31 figs., 29 tabs.

  8. Establishment of design and fabrication technology and domestic qualification for ITER blanket system

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Bong Guen; In, S. R.; Bae, Y. D. (and others)

    2006-02-15

    To obtain and analyze the detailed design and manufacturing technology of the blanket system for each components, the related data are collected through the various sources. And also, design processes and results of the FWs, shield blocks, and TBMs are investigated. From these analysis of the blanket R and D status of each party, we develop the KO R and D plan and it is used in the selection of manufacturing method and the materials. For the ITA16-10 subtask1, we had the official agreement with ITER IT in December 2004 for the qualification of the FW panel fabrication methods and to establish the NDT methods for the FW panel. From the technical reports we published, we compare the manufacturing methods and the proposed material for each component according to the parties. Be is proposed as a plasma facing material and most parties have interest in S-65C. Cu alloy is proposed as a heat sink material and DSCu or CuCrZr are investigated now. For the structural material, stainless steel such as SS316L(N) is investigated internationally. HIP and brazing are proposed as the manufacturing methods. In order to establish the blanket system technology, design contents of shield block by ITER IT and other parties were investigated through participating the international workshop and meeting, dispatching the researcher to the ITER IT or other parties to collect the drafting and 3D modeling files. The modification items of blanket design were investigated and a researcher was dispatched in the ITER IT and participated in the analysis on cooling problem in shield block such as front header and drilled manifold. To investigate the development status of TBM, we participated the 14th TBWG meeting and proposed the KO HCSB and HCML as candidates. And also, we obtain the R and D results of other parties and make document about the R and D status of other parties for the TBM. Finally, we establish the KO TBM R and D plan and proposed it to ITER IT and other parties. In which, the

  9. Verification test results of a cutting technique for the ITER blanket cooling pipes

    Energy Technology Data Exchange (ETDEWEB)

    Shigematsu, Soichiro, E-mail: shigematsu.soichiro@jaea.go.jp [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki (Japan); Tanigawa, Hisashi; Aburadani, Atsushi; Takeda, Nobukazu; Kakudate, Satoshi [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki (Japan); Mori, Seiji; Nakahira, Masataka; Raffray, Rene; Merola, Mario [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France)

    2012-08-15

    For replacement of the first wall (FW) of the international thermonuclear experimental reactor (ITER), cutting and welding tools for the cooling pipes must be able to access a pipe from the surface side of the FW and cut/weld the pipe from the inside the cooling pipe (inner diameter: 42.72 mm, thickness: 2.77 mm). The cutting tool for the pipe end is required to cut a flat plate circularly from the surface side of the FW (cutting diameter: approximately 44 mm, plate thickness: 5 mm). To determine the specifications for both the tools and the blanket hydraulic connections, the ITER Organization (IO) and the Japan Domestic Agency (JADA) conducted research and development activities regarding the FW replacement. This paper describes the current status of the development of cutting tools for the cooling pipe connection.

  10. APT Blanket System Loss-of-Helium-Gas Accident Based on Initial Conceptual Design - Helium Supply Rupture into Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    The model results are used to determine if beam power shutdown is necessary (or not) as a result of the LOHGA accident to maintain the blanket system well below any of the thermal-hydraulic constraints imposed on the design. The results also provide boundary conditions to the detailed bin model to study the detailed temperature response of the hot blanket module structure. The results for these two cases are documented in the report.

  11. Summary report for ITER task - T68: MHD facility preparation for Li/V blanket option

    Energy Technology Data Exchange (ETDEWEB)

    Reed, C.B.; Haglund, R.C.; Miller, M.E. [and others

    1995-08-01

    A key feasibility issue for the ITER Vanadium/Lithium breeding blanket is the question of insulator coatings. Design calculations show that an electrically insulating layer is necessary to maintain an acceptably low MHD pressure drop. To enable experimental investigations of the MHD performance of candidate insulator materials and the technology for putting them in place, the room-temperature ALEX (Argonne`s Liquid Metal EXperiment) NaK facility was upgraded to a 300{degrees}C lithium system. The objective of this upgrade was to modify the existing facility to the minimum extent necessary, consistent with providing a safe, flexible, and easy to operate MHD test facility which uses lithium at ITER-relevant temperatures, Hartmann numbers, and interaction parameters. The facility was designed to produce MHD pressure drop data, test section voltage distributions, and heat transfer data for mid-scale test sections and blanket mockups. The system design description for this lithium upgrade of the ALEX facility is given in this document.

  12. Fabrication of ITER Semi-Prototype Blanket First Wall for the Final Qualification

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Byoung Kwon; Jung, Yang Il; Park, Jeong Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Suk Kwon; Lee, Don Won; Kim, Duck Hoi; Cho, Seung Yon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    The ITER semi-prototype was designed to qualify the manufacturing technology for the ITER blanket first wall. According to the design of the semi-prototype, its fabrication is expected to face great difficulty. The blanket first wall consists of three different materials, i.e., beryllium (Be), CuCrZr, and stainless-steel (SS), which are joined into one part. For fabrication of these multi-layered structures, hot isostatic pressing (HIP), which is one of the diffusion bonding methods, has been considered as a promising technology to realize sufficient mechanical integrity of a joint under the anticipated high neutron and stress fields. HIP provides high dimensional accuracy, low residual stress during the joining process, and the joining of three-dimensionally complex structures in comparison with other joining methods. Even though the joining technology for the different materials had been developed in the first stage of the qualification, the joining is still a key issue for the fabrication of the semi-prototype

  13. ITER屏蔽包层活化分析%Activation analysis for ITER shielding blanket

    Institute of Scientific and Technical Information of China (English)

    杨琪; 李斌; 郑剑; 何桃; 蒋洁琼; 吴宜灿

    2016-01-01

    作为国际热核聚变实验堆(ITER)的重要部件之一,屏蔽包层承受高强度聚变中子辐照,需要定期更换和维修。当活化的屏蔽包层从 ITER 托卡马克装置移到热室时,可能会给工作人员造成严重的辐射照射,是 ITER大厅和热室屏蔽设计的重要辐射源。文中基于 ITER最新中子学分析基准模型和“二步法”停堆剂量计算方法,使用超级蒙特卡罗核计算仿真软件系统 SuperMC针对15号屏蔽包层建立精细的中子学模型,并计算分析包层的活化情况及最严重情况下的周围辐射剂量率,并初步应用于 ITER赤道窗口室的屏蔽分析。计算结果显示,单个包层周围最大剂量率为350 Sv/hr,当传送小车停留在赤道窗口室内时,窗口室屏蔽门外剂量率高于10 mSv/hr,不足以满足设计要求。%As one of the key components of the International thermonuclear experiment reactor (ITER),blankets will sustain radiation from fusion neutrons with high intensity and may need to be replaced and maintained regularly. During the maintenance,the cask with activated blankets will be transferred to hot cell from Tokamak,which will cause high level of radiation in the building and radiation exposure for workers. Employing the Super Monte Carlo Simulation Program for Nuclear and Radiation Process (SuperMC),the activation of No.1 5 shielding blanket and the shutdown dose around was analyzed based on the latest ITER neutronics model named Blite-3. The results were applied in the shielding analysis for ITER equatorial port cell. From the results,the dose rate around one activated blanket should be as high as 350 Sv/hr. When the cask carrying four activated first walls was transferred to the equatorial port cell,the dose rate in the gallery outside the port cell could be more than 10 mSv/hr,not meeting with the design criteria.

  14. Fracture toughness of irradiated candidate materials for ITER first wall/blanket structures: Summary report

    Energy Technology Data Exchange (ETDEWEB)

    Alexander, D.J.; Pawel, J.E.; Grossbeck, M.L.; Rowcliffe, A.F. [Oak Ridge National Lab., TN (United States)] [and others

    1996-04-01

    Disk compact specimens of candidate materials for first wall/blanket structures in ITER have been irradiated to damage levels of about 3 dpa at nominal irradiation temperatures of either 90 250{degrees}C. These specimens have been tested over a temperature range from 20 to 250{degrees}C to determine J-integral values and tearing moduli. The results show that irradiation at these temperatures reduces the fracture toughness of austenic stainless steels, but the toughness remains quite high. The toughness decreases as the temperature increases. Irradiation at 250{degrees}C is more damaging that at 90{degrees}C, causing larger decreases in the fracture toughness. The ferritic-martensitic steels HT-9 and F82H show significantly greater reductions in fracture toughness that the austenitic stainless steels.

  15. Development of pipe welding, cutting and inspection tools for the ITER blanket

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Kiyoshi; Ito, Akira; Taguchi, Kou; Takiguchi, Yuji; Takahashi, Hiroyuki; Tada, Eisuke [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-07-01

    In D-T burning reactors such as International Thermonuclear Experimental Reactor (ITER), an internal access welding/cutting of blanket cooling pipe with bend sections is inevitably required because of spatial constraint due to nuclear shield and available port opening space. For this purpose, internal access pipe welding/cutting/inspection tools for manifolds and branch pipes are being developed according to the agreement of the ITER R and D task (T329). A design concept of welding/cutting processing head with a flexible optical fiber has been developed and the basic feasibility studies on welding, cutting and rewelding are performed using stainless steel plate (SS316L). In the same way, a design concept of inspection head with a non-destructive inspection probe (including a leak-testing probe) has been developed and the basic characteristic tests are performed using welded stainless steel pipes. In this report, the details of welding/cutting/inspection heads for manifolds and branch pipes are described, together with the basic experiment results relating to the welding/cutting and inspection. In addition, details of a composite type optical fiber, which can transmit both the high-power YAG laser and visible rays, is described. (author)

  16. Development of Reduced Activation Ferritic-Martensitic Steels and fabrication technologies for Indian test blanket module

    Energy Technology Data Exchange (ETDEWEB)

    Raj, Baldev [Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Jayakumar, T., E-mail: tjk@igcar.gov.in [Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India)

    2011-10-01

    For the development of Reduced Activation Ferritic-Martensitic Steel (RAFMS), for the Indian Test Blanket Module for ITER, a 3-phase programme has been adopted. The first phase consists of melting and detailed characterization of a laboratory scale heat conforming to Eurofer 97 composition, to demonstrate the capability of the Indian industry for producing fusion grade steel. In the second phase which is currently in progress, the chemical composition will be optimized with respect to tungsten and tantalum for better combination of mechanical properties. Characterization of the optimized commercial scale India-specific RAFM steel will be carried out in the third phase. The first phase of the programme has been successfully completed and the tensile, impact and creep properties are comparable with Eurofer 97. Laser and electron beam welding parameters have been optimized and welding consumables were developed for Narrow Gap - Gas Tungsten Arc welding and for laser-hybrid welding.

  17. First wall and blanket module safety enhancement by material selection and design decision

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, B.J.

    1980-01-01

    A thermal/mechanical study has been performed which illustrates the behavior of a fusion reactor first wall and blanket module during a loss of coolant flow event. The relative safety advantages of various material and design options were determined. A generalized first wall-blanket concept was developed to provide the flexibility to vary the structural material (stainless steel vs titanium), coolant (helium vs water), and breeder material (liquid lithium vs solid lithium aluminate). In addition, independent vs common first wall-blanket cooling and coupled adjacent module cooling design options were included in the study. The comparative analyses were performed using a modified thermal analysis code to handle phase change problems.

  18. Preliminary Failure Modes and Effects Analysis of the US DCLL Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Lee C. Cadwallader

    2010-06-01

    This report presents the results of a preliminary failure modes and effects analysis (FMEA) of a small tritium-breeding test blanket module design for the International Thermonuclear Experimental Reactor. The FMEA was quantified with “generic” component failure rate data, and the failure events are binned into postulated initiating event families and frequency categories for safety assessment. An appendix to this report contains repair time data to support an occupational radiation exposure assessment for test blanket module maintenance.

  19. Preliminary Failure Modes and Effects Analysis of the US DCLL Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Lee C. Cadwallader

    2007-08-01

    This report presents the results of a preliminary failure modes and effects analysis (FMEA) of a small tritium-breeding test blanket module design for the International Thermonuclear Experimental Reactor. The FMEA was quantified with “generic” component failure rate data, and the failure events are binned into postulated initiating event families and frequency categories for safety assessment. An appendix to this report contains repair time data to support an occupational radiation exposure assessment for test blanket module maintenance.

  20. Manufacturing and testing of full scale prototype for ITER blanket shield block

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sa-Woong, E-mail: swkim12@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); Kim, Duck-Hoi; Jung, Hun-Chea [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Sung-Ki [WONIL Co., Ltd., Haman (Korea, Republic of); Kang, Sung-Chan [POSCO Specialty Steel Co., Ltd., Changwon (Korea, Republic of); Zhang, Fu; Kim, Byoung-Yoon [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Ahn, Hee-Jae; Lee, Hyeon-Gon; Jung, Ki-Jung [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-04-15

    Highlights: • 316L(N)-IG forged steel was successfully fabricated and qualified. • Related R&D activities were implemented to resolve the fabrication issues. • SB #8 FSP was successfully manufactured with conventional fabrication techniques. • All of the validation tests were carried out and met the acceptance criteria. - Abstract: Based on the preliminary design of the ITER blanket shield block (SB) #8, the full scale prototype (FSP) has been manufactured and tested in accordance with pre-qualification program, and related R&D was performed to resolve the technical issues of fabrication. The objective of the SB pre-qualification program is to demonstrate the acceptable manufacturing quality by successfully passing the formal test program. 316L(N)-IG stainless steel forging blocks with 1.80L × 1.12W × 0.43t (m) were developed by using an electric arc furnace, and as a result, the material properties were satisfied with technical specification. In the course of applying conventional fabrication techniques such as cutting, milling, drilling and welding of the forged stainless steel block for the manufacturing of the SB #8 FSP, several technical problems have been addressed. And also, the hydraulic connector of cross-forged material re-melted by electro slag or vacuum arc requires the application of advanced joining techniques such as automatic bore TIG and friction welding. Many technical issues – drilling, welding, slitting, non-destructive test and so on – have been raised during manufacturing. Associated R&D including the computational simulation and coupon testing has been done in collaboration with relevant industries in order to resolve these engineering issues. This paper provides technical key issues and their possible resolutions addressed during the manufacture and formal test of the SB #8 FSP, and related R&D.

  1. Strategy for solving a coupled problem of the electromagnetic load analysis and design optimization for local conducting structures to support the ITER blanket development

    Energy Technology Data Exchange (ETDEWEB)

    Rozov, Vladimir, E-mail: vladimir.rozov@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul-lez-Durance (France); Belyakov, V.; Kukhtin, V.; Lamzin, E.; Mazul, I.; Sytchevsky, S. [D.V. Efremov Scientific Research Institute, 196641 St. Petersburg (Russian Federation)

    2014-11-15

    Highlights: • We present the way of modeling transient electro-magnetic loads on local conductive domains in the large magnetic system. • Simplification is achieved by decomposing of the problem, multi-scale integral-differential modeling and use of integral parameters. • The intrinsic scale of loads on a localized conductor with eddy is quantified through the load susceptibility tensor. • Solution is searched as response of a simple equivalent dynamic simulator, using control theory methods. • The concept is exemplified with multi-scenario assessment of EM eddy loads on ITER blanket modules. - Abstract: The complexity of the electromagnetic (EM) response of the tokamak structures is one of the key and design-driving issues for the ITER. We consider the specifics of the assessment of ponderomotive forces, acting on local components of a large electro-physical device during electromagnetic transients. A strategy and approach is proposed for the operative EM loads modeling and analysis that enables design optimization at early phases of development. The paper describes a method of principal simplification of the mathematical model, based on the analysis and exploiting specific features and peculiarities of the relevant technical problem, determined by the design and operation of the device and system under consideration. The application of the method for predictive EM loads analysis and corresponding numerical calculations are exemplified for the localized ITER blanket components — shield modules. The example demonstrates the efficiency of EM load analysis in complex electromagnetic systems via a set of simplified models with different scope, contents and level of detail.

  2. Current status on the detailed design and development of fabrication techniques for the ITER blanket shield block in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Duck-Hoi [National Fusion Research Institute, 52 Yeoeun-dong, Yuseong-gu, Daejeon 305-333 (Korea, Republic of)], E-mail: kdwh@nfri.re.kr; Cho, Seungyon; Ahn, Mu-Young; Lee, Eun-Seok; Jung, Ki Jung [National Fusion Research Institute, 52 Yeoeun-dong, Yuseong-gu, Daejeon 305-333 (Korea, Republic of); Kim, Do-Hyeong [ANST, Inc., 222-7 Guro3-dong, Guro-gu, Seoul 152-848 (Korea, Republic of)

    2008-12-15

    Recent activities and progress on the design and fabrication of the ITER blanket shield block in Korea are described in this paper. Hydraulic analyses, using a flow driver model for determining the gap between the radial cooling passages and flow drivers inside the shield block, were performed. The thermo-hydraulic analysis of half of a shield block was also conducted to investigate the uniformity of the flow stream in cooling passages and to evaluate the temperature distribution in the structure. The maximum temperature is below the allowable value, although hot spots occurred in the corner edge in the shield block. A manufacturing feasibility study for the development of the blanket shield block was performed in cooperation with KO industries. It was found that specific techniques would be required for the successful fabrication of an ITER blanket shield block, specifically electron-beam welding at a thickness up to 110 mm. The development of joining and drilling technologies for the thick shield block and lid joints is in progress. In addition, a full scale mock-up fabrication and the development of NDT techniques are planned in the near future.

  3. Activation Characteristics of Fuel Breeding Blanket Module in Fusion Driven Subcritical System

    Institute of Scientific and Technical Information of China (English)

    HUANG Qun-Ying; LI Jian-Gang; CHEN Yi-Xue

    2004-01-01

    @@ Shortage of energy resources and production of long-lived radioactivity wastes from fission reactors are among the main problems which will be faced in the world in the near future. The conceptual design of a fusion driven subcritical system (FDS) is underway in Institute of Plasma Physics, Chinese Academy of Sciences. There are alternative designs for multi-functional blanket modules of the FDS, such as fuel breeding blanket module (FBB)to produce fuels for fission reactors, tritium breeding blanket module to produce the fuel, i.e. tritium, for fusion reactor and waste transmutation blanket module to try to permanently dispose of long-lived radioactivity wastes from fission reactors, etc. Activation of the fuel breeding blanket of the fusion driven subcritical system (FDS-FBB) by D-T fusion neutrons from the plasma and fission neutrons from the hybrid blanket are calculated and analysed under the neutron wall loading 0.5 MW/m2 and neutron fluence 15 MW. yr/m2. The neutron spectrum is calculated with the worldwide-used transport code MCNP/4C and activation calculations are carried out with the well known European inventory code FISPACT/99 with the latest released IAEA Fusion Evaluated Nuclear Data Library FENDL-2.0 and the ENDF/B-V uranium evaluated data. Induced radioactivities, dose rates and afterheats, etc, for different components of the FDS-FBB are compared and analysed.

  4. Lithium-vanadium advanced blanket development. ITER final report on U.S. contribution: Task T219/T220

    Energy Technology Data Exchange (ETDEWEB)

    Smith, D.L.; Mattas, R.F. [comps.

    1997-07-01

    The objective of this task is to develop the required data base and demonstrate the performance of a liquid lithium-vanadium advanced blanket design. The task has two main activities related to vanadium structural material and liquid lithium system developments. The vanadium alloy development activity included four subtasks: (1.1) baseline mechanical properties of non irradiated base metal and weld metal joints; (1.2) compatibility with liquid lithium; (1.3) material irradiation tests; and (1.4) development of material manufacturing and joining methods. The lithium blanket technology activity included four subtasks: (2.1) electrical insulation development and testing for liquid metal systems; (2.2) MHD pressure drop and heat transfer study for self-cooled liquid metal systems; (2.3) chemistry of liquid lithium; and (2.4) design, fabrication and testing of ITER relevant size blanket mockups. A summary of the progress and results obtained during the period 1995 and 1996 in each of the subtask areas is presented in this report.

  5. Development of thermal-hydraulic analysis methodology for multiple modules of water-cooled breeder blanket in fusion DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Geon-Woo; Lee, Jeong-Hun [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Cho, Hyoung-Kyu, E-mail: chohk@snu.ac.kr [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Park, Goon-Cherl [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Im, Kihak [National Fusion Research Institute, 169-148, Yuseong-gu, Daejeon 305-806 (Korea, Republic of)

    2016-02-15

    Highlights: • A methodology to simulate the K-DEMO blanket system was proposed. • The results were compared with the CFD, to verify the prediction capability of MARS. • 46 Blankets in a single sector in K-DEMO were simulated using MARS-KS. • Supervisor program was devised to handle each blanket module individually. • The calculation results showed the flow rates, pressure drops, and temperatures. - Abstract: According to the conceptual design of the fusion DEMO reactor proposed by the National Fusion Research Institute of Korea, the water-cooled breeding blanket system incorporates a total of 736 blanket modules. The heat flux and neutron wall loading to each blanket module vary along their poloidal direction, and hence, thermal analysis for at least one blanket sector is required to confirm that the temperature limitations of the materials are satisfied in all the blanket modules. The present paper proposes a methodology of thermal analysis for multiple modules of the blanket system using a nuclear reactor thermal-hydraulic analysis code, MARS-KS. In order to overcome the limitations of the code, caused by the restriction on the number of computational nodes, a supervisor program was devised, which handles each blanket module separately at first, and then corrects the flow rate, considering pressure drops that occur in each module. For a feasibility test of the proposed methodology, 46 blankets in a single sector were simulated; the calculation results of the parameters, such as mass flow, pressure drops, and temperature distribution in the multiple blanket modules showed that the multi-module analysis method can be used for efficient thermal-hydraulic analysis of the fusion DEMO reactor.

  6. Fast Ion Effects During Test Blanket Module Simulation Experiments in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, G J; Ellis, R; Gorelenkova, M; Heidbrink, W W; Kurki-Suonio, T; Nazikian, R; Salmi, A; Schaffer, M J; Shinohara, K; Snipes, J A; Spong, D A; Koskela, T

    2011-06-03

    Fast beam-ion losses were studied in DIII-D in the presence of a scaled mockup of two Test Blanket Modules (TBM) for ITER. Heating of the protective tiles on the front of the TBM surface was found when neutral beams were injected and the TBM fields were engaged. The fast-ion core confinement was not significantly affected. Different orbit-following codes predict the formation of a hot spot on the TBM surface arising from beam-ions deposited near the edge of the plasma. The codes are in good agreement with each other on the total power deposited at the hot spot predicting an increase in power with decreasing separation between the plasma edge and the TBM surface. A thermal analysis of the heat flow through the tiles shows that the simulated power can account for the measured tile temperature rise. The thermal analysis, however, is very sensitive to the details of the localization of the hot spot which is predicted to be different among the various codes.

  7. Fast-ion effects during test blanket module simulation experiments in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, G. [Princeton Plasma Physics Laboratory (PPPL); Budny, R. V. [Princeton Plasma Physics Laboratory (PPPL); Ellis, R. [Princeton Plasma Physics Laboratory (PPPL); Gorelenkova, M. [Princeton Plasma Physics Laboratory (PPPL); Heidbrink, W. [University of California, Irvine; Kurki-Suonio, T. [Aalto University, Finland; Nazikian, Raffi [Princeton Plasma Physics Laboratory (PPPL); Saimi, A. [Aalto University, Finland; Schaffer, M. J. [General Atomics, San Diego; Shinohara, K. [Japan Atomic Energy Agency (JAEA), Naka; Snipes, J. A. [ITER Organization, Cadarache, France; Spong, Donald A [ORNL; Koskela, T. [Aalto University, Finland; Van Zeeland, Michael [General Atomics

    2011-01-01

    Fast beam-ion losses were studied in DIII-D in the presence of a scaled mock-up of two test blanket modules (TBM) for ITER. Heating of the protective tiles on the front of the TBM surface was found when neutral beams were injected and the TBM fields were engaged. The fast-ion core confinement was not significantly affected. Different orbit-following codes predict the formation of a hot spot on the TBM surface arising from beam ions deposited near the edge of the plasma. The codes are in good agreement with each other on the total power deposited at the hot spot, predicting an increase in power with decreasing separation between the plasma edge and the TBM surface. A thermal analysis of the heat flow through the tiles shows that the simulated power can account for the measured tile temperature rise. The thermal analysis, however, is very sensitive to the details of the localization of the hot spot, which is predicted to be different among the various codes.

  8. Development of the Helium Cooled Lithium Lead blanket for DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Aiello, G., E-mail: giacomo.aiello@cea.fr [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France); Aubert, J.; Jonquères, N. [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France); Li Puma, A. [CEA-Saclay, DEN/DANS/DM2S/SERMA/LPEC, 91191 Gif Sur Yvette Cedex (France); Morin, A.; Rampal, G. [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France)

    2014-10-15

    Highlights: • The HCLL blanket design has been modified to adapt it to the 2012 EFDA DEMO specifications. • The new design has been developed with the aim to capitalize on TBM experience in ITER. • A new attachment system for the modules has been proposed. - Abstract: The Helium Cooled Lithium Lead (HCLL) blanket is one of the candidate European blanket concepts selected for the DEMOnstration fusion power plant that should follow ITER. In a fusion power plant, the blanket is one of the key components because of its impact on the plant performance, availability, safety and economics. In 2012, the European Fusion Development Agreement (EFDA) agency issued new specifications for DEMO: this paper describes the work performed to adapt the previous 2007 HCLL-DEMO blanket design to those specifications. A new segmentation has been defined assuming straight surfaces for all blanket modules. Following the Multi Module Segment (MMS) option, all modules are attached to a common back supporting structure which also serves as manifold for Helium and PbLi distribution. A detailed CAD design of the central outboard module has been defined. Thermo-hydraulic and thermo-mechanical analyses on of the First Wall and Breeder Zone have been carried out. For the attachment of the modules to the common backplate, a new solution based on the use of Tie Rods, derived from the design of the corresponding HCLL Test Blanket Module for ITER, has been proposed. This paper also identifies the priorities for further development of the HCLL blanket design.

  9. Development of Thermal-hydraulic Analysis Methodology for Multi-module Breeding Blankets in K-DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Geon-Woo; Lee, Jeong-Hun; Park, Goon-Cherl; Cho, Hyoung-Kyu [Seoul National University, Seoul (Korea, Republic of); Im, Kihak [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    In this paper, the purpose of the analyses is to extend the capability of MARS-KS to the entire blanket system which includes a few hundreds of single blanket modules. Afterwards, the plan for the whole blanket system analysis using MARS-KS is introduced and the result of the multiple blanket module analysis is summarized. A thermal-hydraulic analysis code for a nuclear reactor safety, MARS-KS, was applied for the conceptual design of the K-DEMO breeding blanket thermal analysis. Then, a methodology to simulate multiple blanket modules was proposed, which uses a supervisor program to handle each blanket module individually at first and then distribute the flow rate considering pressure drops arises in each module. For a feasibility test of the proposed methodology, 10 outboard blankets in a toroidal field sector were simulated, which are connected with each other through the inlet and outlet common headers. The calculation results of flow rates, pressure drops, and temperatures showed the validity of the calculation and thanks to the parallelization using MPI, almost linear speed-up could be obtained.

  10. Shield fabrication development of ITER primary wall modules by powder HIP. ITER task T216-Subtask 3E1

    Energy Technology Data Exchange (ETDEWEB)

    Lind, A

    1997-12-01

    A research and development program for the blanket shield in the International Thermonuclear Experimental Reactor (ITER) has been implemented to provide input for the design and manufacture of full scale production components. It comprises fabrication and testing of mock-ups and prototype modules. The design, materials, manufacture, examination, testing and inspection of the mock-ups representing future full scale production modules. This work applies to the development of a shield block fabrication method by Hot Isostatic Pressing (HIP) starting from a gas atomised powder and pre-fabricated cooling tube galleries. The size of the block is 1250 x 650 x 250 mm and the weight is about 1400 kg. Examination and testing of the block was performed to determine properties, achieved fabrication tolerances, and quality of bonding. It is concluded that the today`s powder HIP route gives a 316 LN IG material with mechanical properties which fulfills the ITER material specification requirements and a fully dense block which is easy to examine with ultrasonic methods. The joints between tubes and matrix are excellent. In order to achieve and maintain accuracy in positioning of the tubes during fabrication improvements of the standard fabrication route have been identified, such as the positioning of tubes during welding, the powder particle distribution and the powder filling procedure. Modification of the actual HIP cycle may also be required

  11. Progress on DCLL Blanket Concept

    Energy Technology Data Exchange (ETDEWEB)

    Wong, Clement; Abdou, M.; Katoh, Yutai; Kurtz, Richard J.; Lumsdaine, A.; Marriott, Edward P.; Merrill, Brad; Morley, Neil; Pint, Bruce A.; Sawan, M.; Smolentsev, S.; Williams, Brian; Willms, Scott; Youssef, M.

    2013-09-01

    Under the US Fusion Nuclear Science and Technology Development program, we have selected the Dual Coolant Lead Lithium concept (DCLL) as a reference blanket, which has the potential to be a high performance DEMO blanket design with a projected thermal efficiency of >40%. Reduced activation ferritic/martensitic (RAF/M) steel is used as the structural material. The self-cooled breeder PbLi is circulated for power conversion and for tritium breeding. A SiC-based flow channel insert (FCI) is used as a means for magnetohydrodynamic pressure drop reduction from the circulating liquid PbLi and as a thermal insulator to separate the high-temperature PbLi (~700°C) from the helium-cooled RAF/M steel structure. We are making progress on related R&D needs to address critical Fusion Nuclear Science and Facility (FNSF) and DEMO blanket development issues. When performing the function as the Interface Coordinator for the DCLL blanket concept, we had been developing the mechanical design and performing neutronics, structural and thermal hydraulics analyses of the DCLL TBM module. We had estimated the necessary ancillary equipment that will be needed at the ITER site and a detailed safety impact report has been prepared. This provided additional understanding of the DCLL blanket concept in preparation for the FNSF and DEMO. This paper will be a summary report on the progress of the DCLL TBM design and R&Ds for the DCLL blanket concept.

  12. Synthesis and fabrication of lithium-titanate pebbles for ITER breeding blanket by solid state reaction and spherodization

    Energy Technology Data Exchange (ETDEWEB)

    Mandal, D., E-mail: dmandal@barc.gov.i [Chemical Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Shenoi, M.R.K.; Ghosh, S.K. [Chemical Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India)

    2010-08-15

    {sup 6}Li produces tritium by (n, {alpha}) nuclear reaction, {sup 6}Li + {sup 1}n {yields} {sup 4}He + {sup 3}H. Lithium titanate (Li{sub 2}TiO{sub 3}) enriched with {sup 6}Li, is the most promising candidate for solid test blanket module (TBM) material for fusion reactors. Various processes are reported in literature for the fabrication of Li{sub 2}TiO{sub 3} pebbles for its use as TBM material. A process has been developed based on the solid state reaction of lithium-carbonate and titanium-dioxide for the synthesis of lithium titanate and pebble fabrication by extrusion, spherodization and sintering. This paper discusses the sequence of steps followed in this process and the properties obtained. Analytical grade titanium-dioxide and lithium-carbonate were taken in stoichiometric ratio and were milled to ensure thorough intimate mixing and obtain fine particles less than 45 {mu}m before its calcination at 900 {sup o}C. Following calcination, the agglomerated product was again milled to fine particles of size less than 45 {mu}m. Aqueous solution of ploy-vinyl-alcohol was added as binder prior to its feeding in the extruder. The extruded strips were spherodized and spherical pebbles were dried and sintered at 900 {sup o}C for different duration. Pebbles of desired density and porosity were obtained by suitable combination of sintering temperature and duration of sintering. Properties of the prepared pebbles were also characterized for sphericity, pore size distribution, grain size, crushing load strength, etc. The values were found to be conforming to the desired properties for use as solid breeder. The attractive feature of this process is almost no waste generation.

  13. Thermal-mechanical and thermal-hydraulic integrated study of the Helium-Cooled Lithium Lead Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Chiovaro, P., E-mail: pg.chiovaro@din.unipa.it [Dipartimento di Ingegneria Nucleare, Universita di Palermo, Palermo (Italy); Di Maio, P.A.; Giammusso, R.; Lupo, Q.; Vella, G. [Dipartimento di Ingegneria Nucleare, Universita di Palermo, Palermo (Italy)

    2010-12-15

    The Helium-Cooled Lithium Lead Test Blanket Module (HCLL-TBM) is one of the two TBM to be installed in an ITER equatorial port since day 1 of operation, with the specific aim to investigate the main concept functionalities and issues such as high efficiency helium cooling, resistance to thermo-mechanical stresses, manufacturing techniques, as well as tritium transport, magneto-hydrodynamics effects and corrosion. In particular, in order to show a DEMO-relevant thermo-mechanical and thermal-hydraulic behavior, the HCLL-TBM has to meet several requirements especially as far as its coolant thermofluid-dynamic conditions and its thermal-mechanical field are concerned. The present paper is focused on the assessment of the HCLL-TBM thermal-mechanical performances under both nominal and accidental load conditions, by adopting a computational approach based on the Finite Element Method. A realistic 3D finite element model of the whole HCLL-TBM, in the horizontal first wall design has been set up, consisting of about 597,000 elements and 767,000 nodes. In particular, since the thermal fields of both the module and the coolant are strictly coupled, the helium flow domain has been modeled too and a thermal contact model has been set up to properly simulate the convective heat transfer between the structure wall and the coolant. Pure conductive heat transfer has been assumed within the Pb-Li eutectic alloy of the breeder units. The volumetric density of the nuclear deposited power, recently calculated at Department of Nuclear Engineering of the University of Palermo by the MCNP 4C code, has been applied as distributed thermal load in order to assess the potential influence on the module thermo-mechanical performances of the markedly non-uniform poloidal and toroidal distributions that have been predicted within the Segment Box. Different loading scenarios have been considered as to the heat flux onto the module First Wall. Steady state and transient thermal-mechanical analyses

  14. Conceptual design of a test facility for the remote handling operations of the ITER Test Blanker Modules; Diseno conceptual de la instalacion de demostracion de los dispositivos de manipulacion remota para los Modulos de Ensayo de la Envoltura Regeneradora de ITER

    Energy Technology Data Exchange (ETDEWEB)

    Marqueta, A.; Garcia, I.; Gomez, A.; Garcia, L.; Sedano, E.; Fernandez, I.

    2012-07-01

    Conceptual Design of a test facility for the remote handling operations of the ITER Test Blanket Modules. Conditions inside a fusion reactor are incompatible with conventional manual maintenance tasks. the same applies for ancillary equipment. As a consequence, it will become necessary to turn to remote visualization and remote handling techniques, which will have in consideration the extreme conditions, both physical and operating, of ITER. Main goal of the project has been the realization of the conceptual design for the test facility for the Test Blanket Modules of ITER and their associated systems, related to the Remote Handling operations regarding the Port Cell area. Besides the definition of the operations and the specification of the main components and ancillary systems of the TBM graphical simulation have been used for the design, verification and validation of the remote handling operations. (Author)

  15. Multiple Module Simulation of Water Cooled Breeding Blankets in K-DEMO Using Thermal-Hydraulic Analysis Code MARS-KS

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Geon-Woo; Lee, Jeong-Hun; Park, Goon-Cherl; Cho, Hyoung-Kyu [Seoul National University, Seoul (Korea, Republic of); Im, Kihak [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    A preliminary concept for the Korean fusion demonstration reactor (K-DEMO) has been studied by the National Fusion Research Institute (NFRI) based on the National Fusion Roadmap of Korea. The feasibility studies have been performed in order to establish the conceptual design guidelines of the breeding blanket. As a part of the NFRI research, Seoul National University (SNU) is conducting thermal design, evaluation and validation of the water-cooled breeding blanket for the K-DEMO reactor. The purpose of this study is to extend the capability of MARS-KS to the overall blanket system analysis which includes 736 blanket modules in total. The strategy for the multi-module blanket system analysis using MARS-KS is introduced and the analysis result of the 46 blanket modules of single sector was summarized. A thermal-hydraulic analysis code for a nuclear reactor safety, MARS-KS, was applied for thermal analysis of the conceptual design of the K-DEMO breeding blanket. Then, a methodology to simulate multiple blanket modules was proposed, which uses a supervisor program to handle each blanket module individually at first and then distribute the flow rate considering the pressure drop that occurs in each module. For a feasibility test of the proposed methodology, 46 blankets in a sector, which are connected with each other through the common headers for the sector inlet and outlet, were simulated. The calculation results of flow rates, pressure drops, and temperatures showed the validity of the calculation. Because of parallelization using the MPI system, the computational time could be reduced significantly. In future, this methodology will be extended to an efficient simulation of multiple sectors, and further validation for transient simulation will be carried out for more practical applications.

  16. Report of a technical evaluation panel on the use of beryllium for ITER plasma facing material and blanket breeder material

    Energy Technology Data Exchange (ETDEWEB)

    Ulrickson, M.A. [ed.] [Sandia National Labs., Albuquerque, NM (United States); Manly, W.D. [Oak Ridge National Lab., TN (United States); Dombrowski, D.E. [Brush Wellman, Inc., Cleveland, OH (United States)] [and others

    1995-08-01

    Beryllium because of its low atomic number and high thermal conductivity, is a candidate for both ITER first wall and divertor surfaces. This study addresses the following: why beryllium; design requirements for the ITER divertor; beryllium supply and unirradiated physical/mechanical property database; effects of irradiation on beryllium properties; tritium issues; beryllium health and safety; beryllium-coolant interactions and safety; thermal and mechanical tests; plasma erosion of beryllium; recommended beryllium grades for ITER plasma facing components; proposed manufacturing methods to produce beryllium parts for ITER; emerging beryllium materials; proposed inspection and maintenance techniques for beryllium components and coatings; time table and costs; and the importance of integrating materials and manufacturing personnel with designers.

  17. Low activity aluminum blanket

    Energy Technology Data Exchange (ETDEWEB)

    Benenati, R.; Tichler, P.; Powell, J.R.

    1976-03-01

    The basic design of the breeding blanket consists of cylindrical aluminium canisters filled with a ceramic bed of moderating, shielding, and breeding materials all suitably cooled. A technical analysis of the blanket for an EPR design is given. Activation studies are presented. The effect of pulsed magnetic fields on module structure is investigated. (MOW)

  18. Fabrication of an ITER CS module cross-section

    Science.gov (United States)

    Reed, R. P.; Irick, D.; Biermann, P.; Roundy, F.; Martovetsky, N. N.

    2014-01-01

    An array of 14 × 40 turn-insulated steel conduits, encompassed by the vertical and horizontal ground insulation designed for use in the ITER CS Modules, was fabricated using a specified vacuum-pressure, epoxy-resin, impregnation process. Detailed observations were conducted to assess resin impregnation and conduit realignment under the gravitational pressure of the weight of the stacked conduit. Electrical measurements were conducted to assess the voltage breakdown of turn and ground insulation. Observations were conducted of samples extracted from the resin-impregnated array to assess resin penetration and porosity and insulation quality. Dimensions were measured to obtain reliable estimates of changes in insulation build prior to, and following, resin impregnation and cure. Details of the resin transfer process are provided.

  19. The European ITER Test Blanket Modules: Current status of fabrication technologies development and a way forward

    Energy Technology Data Exchange (ETDEWEB)

    Zmitko, Milan, E-mail: milan.zmitko@f4e.europa.eu [Fusion for Energy (F4E), Josep Pla 2, Barcelona (Spain); Galabert, Jose [Fusion for Energy (F4E), Josep Pla 2, Barcelona (Spain); Thomas, Noël [ATMOSTAT, F-94815 Villejuif (France); Forest, Laurent [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Bucci, Philippe; Cogneau, Laurence [CEA-DRT, 38000 Grenoble (France); Rey, Jörg; Neuberger, Heiko [Karlsruhe Institute of Technology (KIT), Postfach 3640, Karlsruhe (Germany); Poitevin, Yves [Fusion for Energy (F4E), Josep Pla 2, Barcelona (Spain)

    2015-10-15

    Highlights: • Significant progress on development of welding procedures for European TBM achieved. • Fabrication processes feasibility based on diffusion and fusion welding demonstrated. • TBM box assembly welding scenarios investigated and welding scenarios identified. • Future qualification of pF/WPS proposed through realization of a number of QMUs. - Abstract: The paper reviews fabrication technologies and procedures applied for manufacturing of the TBM sub-components, like, HCLL and HCPB cooling plates, HCLL/HCPB stiffening plates, and HCLL/HCPB first wall and side caps. The used technologies are based on fusion and diffusion welding techniques taking into account specificities of the EUROFER-97 steel. Development of a standardized procedure complying with professional codes and standards (RCC-MRx), a preliminary fabrication/welding procedure specification (pF/WPS), is described as well as a fabrication and characterization of feasibility mock-ups (FMU) aimed at assessing the suitability of a fabrication process for fulfilling the design and fabrication specifications. Also, fabrication procedures for the TBM box assembly are presently under development through collaboration between European Fusion Laboratories and Industry for the establishment of an optimized assembly sequence/scenario and development of standardized welding procedure specifications. Selection of optimized assembly scenario takes into accounts not only the design requirements and fabrication possibilities/constraints but also maximum accessibility to the welds for sound non-destructive examination in compliance with welds classification. A future approach towards qualification of the developed fabrication technologies and procedures, through a number of medium to full-size qualification mock-ups according to European standards, is outlined before construction of the first TBMs.

  20. Three-dimensional nuclear analysis for the US dual coolant lead lithium ITER test blanket module

    Energy Technology Data Exchange (ETDEWEB)

    Sawan, M.E., E-mail: sawan@engr.wisc.edu [Fusion Technology Institute, University of Wisconsin, 1500 Engineering Dr., Madison, WI 53706 (United States); Smith, B.; Marriott, E.P.; Wilson, P.P.H. [Fusion Technology Institute, University of Wisconsin, 1500 Engineering Dr., Madison, WI 53706 (United States)

    2010-12-15

    Detailed 3-D neutronics calculations have been performed for the US DCLL TBM. The neutronics calculations were performed directly in the CAD model using the DAG-MCNP code that allows preserving the geometrical details. Detailed high-resolution, high-fidelity profiles of the nuclear parameters were generated using fine mesh tallies. These included tritium production, nuclear heating, and radiation damage. The TBM heterogeneity, exact source profile, and inclusion of the surrounding frame and other in-vessel components result in lower TBM nuclear parameters compared to the previous 1-D predictions. This work clearly demonstrates the importance of preserving geometrical details in nuclear analyses of geometrically complex components in fusion systems.

  1. Low activation steels welding with PWHT and coating for ITER test blanket modules and DEMO

    Science.gov (United States)

    Aubert, P.; Tavassoli, F.; Rieth, M.; Diegele, E.; Poitevin, Y.

    2011-02-01

    EUROFER weldability is investigated in support of the European material properties database and TBM manufacturing. Electron Beam, Hybrid, laser and narrow gap TIG processes have been carried out on the EUROFER-97 steel (thickness up to 40 mm), a reduced activation ferritic-martensitic steel developed in Europe. These welding processes produce similar welding results with high joint coefficients and are well adapted for minimizing residual distortions. The fusion zones are typically composed of martensite laths, with small grain sizes. In the heat-affected zones, martensite grains contain carbide precipitates. High hardness values are measured in all these zones that if not tempered would degrade toughness and creep resistance. PWHT developments have driven to a one-step PWHT (750 °C/3 h), successfully applied to joints restoring good material performances. It will produce less distortion levels than a full austenitization PWHT process, not really applicable to a complex welded structure such as the TBM. Different tungsten coatings have been successfully processed on EUROFER material. It has shown no really effect on the EUROFER base material microstructure.

  2. HHF test with 80x80x1 Be/Cu/SS Mock-ups for verifying the joining technology of the ITER blanket First Wall

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Bae, Young Dug; Kim, Suk Kwon; Hong, Bong Guen; Jeong, Yong Hwan; Park, Jeong Yong; Choi, Byung Kwon; Jung, Hyun Kyu

    2008-11-15

    Through the fabrication of the Cu/SS and Be/Cu joint specimens, fabrication procedure such as material preparation, canning, degassing, HIP (Hot Isostatic Pressing), PHHT (Post HIP heat treatment) was established. The HIP conditions (1050 .deg. C, 100 MPa 2 hr for Cu/SS, 580 .deg. C 100 MPa 2 hr for Be/Cu) were developed through the investigation on joint specimen fabricated with the various HIP conditions; the destructive tests of joint include the microstructure observation of the interface with the examination of the elemental distribution, tension test, bend test, Charpy impact test and fracture toughness test. However, since the joint should be tested under the High Heat Flux (HHF) conditions like the ITER operation for verifying its joint integrity, several HHF tests were performed like the previous HHF test with the Cu/SS, Be/Cu, Be/Cu/SS Mock-ups. In the present study, the HHF test with Be/Cu/SS Mock-ups, which have 80 mm x 80 mm single Be tile and each material depths were kept to be the same as the ITER blanket FW. The Mock-ups fabricated with three kinds of interlayers such as Cr/Ti/Cu, Ti/Cr/Cu, Ti/Cu, which were different from the developed interlayer (Cr/Cu), total 6 Mock-ups were fabricated. Preliminary analysis were performed to decide the test conditions; they were tested with up to 2.5 MW/m2 of heat fluxes and 20 cycles for each Mock-up in a given heat flux. They were tested with JUDITH-1 at FZJ in Germany. During tests, all Mock-ups showed delamination or full detachment of Be tile and it can be concluded that the joints with these interlayers have a bad joining but it can be used as a good data for developing the Be/Cu joint with HIP.

  3. Use of Nuclear Data Sensitivity and Uncertainty Analysis for the Design Preparation of the HCLL Breeder Blanket Mockup Experiment for ITER

    Directory of Open Access Journals (Sweden)

    I. Kodeli

    2008-01-01

    Full Text Available An experiment on a mockup of the test blanket module based on helium-cooled lithium lead (HCLL concept will be performed in 2008 in the Frascati Neutron Generator (FNG in order to study neutronics characteristics of the module and the accuracy of the computational tools. With the objective to prepare and optimise the design of the mockup in the sense to provide maximum information on the state-of-the-art of the cross-section data the mockup was pre-analysed using the deterministic codes for the sensitivity/uncertainty analysis. The neutron fluxes and tritium production rate (TPR, their sensitivity to the underlying basic cross-sections, as well as the corresponding uncertainties were calculated using the deterministic transport codes (DOORS package, the sensitivity/uncertainty code package SUSD3D, and the VITAMINJ/ COVA covariance matrix libraries. The cross-section reactions with largest contribution to the uncertainty of the calculated TPR were identified to be (n,2n and (n,3n reactions on lead. The conclusions of this work support the main benchmark design and suggest some modifications and improvements. In particular this study recommends the use, as far as possible, of both natural and enriched lithium pellets for the TRP measurements. The combined use is expected to provide additional and complementary information on the sensitive cross-sections.

  4. Experimental investigations of flow distribution in coolant system of Helium-Cooled-Pebble-Bed Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Ilić, M.; Schlindwein, G., E-mail: georg.schlindwein@kit.edu; Meyder, R.; Kuhn, T.; Albrecht, O.; Zinn, K.

    2016-02-15

    Highlights: • Experimental investigations of flow distribution in HCPB TBM are presented. • Flow rates in channels close to the first wall are lower than nominal ones. • Flow distribution in central chambers of manifold 2 is close to the nominal one. • Flow distribution in the whole manifold 3 agrees well with the nominal one. - Abstract: This paper deals with investigations of flow distribution in the coolant system of the Helium-Cooled-Pebble-Bed Test Blanket Module (HCPB TBM) for ITER. The investigations have been performed by manufacturing and testing of an experimental facility named GRICAMAN. The facility involves the upper poloidal half of HCPB TBM bounded at outlets of the first wall channels, at outlet of by-pass pipe and at outlets of cooling channels in breeding units. In this way, the focus is placed on the flow distribution in two mid manifolds of the 4-manifold system: (i) manifold 2 to which outlets of the first wall channels and inlet of by-pass pipe are attached and (ii) manifold 3 which supplies channels in breeding units with helium coolant. These two manifolds are connected with cooling channels in vertical/horizontal grids and caps. The experimental facility has been built keeping the internal structure of manifold 2 and manifold 3 exactly as designed in HCPB TBM. The cooling channels in stiffening grids, caps and breeding units are substituted by so-called equivalent channels which provide the same hydraulic resistance and inlet/outlet conditions, but have significantly simpler geometry than the real channels. Using the conditions of flow similarity, the air pressurized at 0.3 MPa and at ambient temperature has been used as working fluid instead of HCPB TBM helium coolant at 8 MPa and an average temperature of 370 °C. The flow distribution has been determined by flow rate measurements at each of 28 equivalent channels, while the pressure distribution has been obtained measuring differential pressure at more than 250 positions. The

  5. Iter

    Science.gov (United States)

    Iotti, Robert

    2015-04-01

    ITER is an international experimental facility being built by seven Parties to demonstrate the long term potential of fusion energy. The ITER Joint Implementation Agreement (JIA) defines the structure and governance model of such cooperation. There are a number of necessary conditions for such international projects to be successful: a complete design, strong systems engineering working with an agreed set of requirements, an experienced organization with systems and plans in place to manage the project, a cost estimate backed by industry, and someone in charge. Unfortunately for ITER many of these conditions were not present. The paper discusses the priorities in the JIA which led to setting up the project with a Central Integrating Organization (IO) in Cadarache, France as the ITER HQ, and seven Domestic Agencies (DAs) located in the countries of the Parties, responsible for delivering 90%+ of the project hardware as Contributions-in-Kind and also financial contributions to the IO, as ``Contributions-in-Cash.'' Theoretically the Director General (DG) is responsible for everything. In practice the DG does not have the power to control the work of the DAs, and there is not an effective management structure enabling the IO and the DAs to arbitrate disputes, so the project is not really managed, but is a loose collaboration of competing interests. Any DA can effectively block a decision reached by the DG. Inefficiencies in completing design while setting up a competent organization from scratch contributed to the delays and cost increases during the initial few years. So did the fact that the original estimate was not developed from industry input. Unforeseen inflation and market demand on certain commodities/materials further exacerbated the cost increases. Since then, improvements are debatable. Does this mean that the governance model of ITER is a wrong model for international scientific cooperation? I do not believe so. Had the necessary conditions for success

  6. Analysis of the thermo-mechanical behaviour of the DEMO Water-Cooled Lithium Lead breeding blanket module under normal operation steady state conditions

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, P.A.; Arena, P. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Aubert, J. [CEA Saclay, DEN/DANS/DM2S/SEMT, 91191 Gif sur Yvette Cedex (France); Bongiovì, G. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Chiovaro, P., E-mail: pierluigi.chiovaro@unipa.it [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Giammusso, R. [ENEA – C.R. Brasimone, 40032 Camugnano (Italy); Li Puma, A. [CEA Saclay, DEN/DANS/DM2S/SEMT, 91191 Gif sur Yvette Cedex (France); Tincani, A. [ENEA – C.R. Brasimone, 40032 Camugnano (Italy)

    2015-10-15

    Highlights: • A DEMO WCLL blanket module thermo-mechanical behaviour has been investigated. • Two models of the WCLL blanket module have been set-up adopting a code based on FEM. • The water flow domain in the module has been considered. • A set of uncoupled steady state thermo-mechanical analyses has been carried out. • Critical temperature is not overcome. Safety verifications are generally satisfied. - Abstract: Within the framework of DEMO R&D activities, a research cooperation has been launched between ENEA, the University of Palermo and CEA to investigate the thermo-mechanical behaviour of the outboard equatorial module of the DEMO1 Water-Cooled Lithium Lead (WCLL) blanket under normal operation steady state scenario. The research campaign has been carried out following a theoretical–computational approach based on the Finite Element Method (FEM) and adopting a qualified commercial FEM code. In particular, two different 3D FEM models (Model 1 and Model 2), reproducing respectively the central and the lateral poloidal–radial slices of the WCLL blanket module, have been set up. A particular attention has been paid to the modelling of water flow domain, within both the segment box channels and the breeder zone tubes, to simulate realistically the coolant-box thermal coupling. Results obtained are herewith reported and critically discussed.

  7. Residual stress in a laser welded EUROFER blanket module assembly using non-destructive neutron diffraction techniques

    Energy Technology Data Exchange (ETDEWEB)

    Hughes, D.J., E-mail: d.hughes@warwick.ac.uk [WMG, University of Warwick, Coventry CV4 7AL (United Kingdom); Koukovini-Platia, E. [CERN, CH-1211 Geneva 23 (Switzerland); Heeley, E.L. [Department of Physical Sciences, Open University, Walton Hall, Milton Keynes MK7 6AA (United Kingdom)

    2014-02-15

    Highlights: • Residual stresses were determined in a welded EUROFER blanket assembly with integrated cooling channels. • Good agreement was seen between experimentally determined and predicted stresses. • We show that microstructure changes that occur in EUROFER steels during welding must be considered for residual stress determination. • An experimental route is proposed for validation of predicted stresses in reactor components using non-destructive diffraction techniques. - Abstract: Whilst the structural integrity and lifetime considerations in welded joints for blanket modules can be predicted using finite element software, it is essential to prove the validity of these simulations. This paper provides detailed analysis for the first time, of the residual stress state in a laser-welded sample with integral cooling channels. State-of-the-art non-destructive neutron diffraction was employed to determine the triaxial stress state and to understand microstructural changes around the heat affected zone. Synchrotron X-ray diffraction was used to probe the variation of strain-free lattice reference parameter around the weld zone allowing correction of the neutron measurements. This paper details an important experimental route to validation of predicted stresses in complex safety-critical reactor components for future applications.

  8. Residual stress in a laser welded EUROFER blanket module assembly using non-destructive neutron diffraction techniques

    CERN Document Server

    Hughes, D J; Heeley, E L

    2014-01-01

    Whilst the structural integrity and lifetime considerations in welded joints for blanket modules can be predicted using finite element software, it is essential to prove the validity of these simulations. This paper provides detailed analysis for the first time, of the residual stress state in a laser-welded sample with integral cooling channels. State-of-the-art non-destructive neutron diffraction was employed to determine the triaxial stress state and to understand microstructural changes around the heat affected zone. Synchrotron X-ray diffraction was used to probe the variation of strain-free lattice reference parameter around the weld zone allowing correction of the neutron measurements. This paper details an important experimental route to validation of predicted stresses in complex safety-critical reactor components for future applications.

  9. An Analysis of Ripple and Error Fields Induced by a Blanket in the CFETR

    Science.gov (United States)

    Yu, Guanying; Liu, Xufeng; Liu, Songlin

    2016-10-01

    The Chinese Fusion Engineering Tokamak Reactor (CFETR) is an important intermediate device between ITER and DEMO. The Water Cooled Ceramic Breeder (WCCB) blanket whose structural material is mainly made of Reduced Activation Ferritic/Martensitic (RAFM) steel, is one of the candidate conceptual blanket design. An analysis of ripple and error field induced by RAFM steel in WCCB is evaluated with the method of static magnetic analysis in the ANSYS code. Significant additional magnetic field is produced by blanket and it leads to an increased ripple field. Maximum ripple along the separatrix line reaches 0.53% which is higher than 0.5% of the acceptable design value. Simultaneously, one blanket module is taken out for heating purpose and the resulting error field is calculated to be seriously against the requirement. supported by National Natural Science Foundation of China (No. 11175207) and the National Magnetic Confinement Fusion Program of China (No. 2013GB108004)

  10. Tritium processing for the European test blanket systems: current status of the design and development strategy

    Energy Technology Data Exchange (ETDEWEB)

    Ricapito, I.; Calderoni, P.; Poitevin, Y. [Fusion for Energy, Barcelona (Spain); Aiello, A.; Utili, M. [ENEA, Camugnano (Italy); Demange, D. [Karlsruhe Institute of Technology - KIT, Karlsruhe (Germany)

    2015-03-15

    Tritium processing technologies of the two European Test Blanket Systems (TBS), HCLL (Helium Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed), play an essential role in meeting the main objectives of the TBS experimental campaign in ITER. The compliancy with the ITER interface requirements, in terms of space availability, service fluids, limits on tritium release, constraints on maintenance, is driving the design of the TBS tritium processing systems. Other requirements come from the characteristics of the relevant test blanket module and the scientific programme that has to be developed and implemented. This paper identifies the main requirements for the design of the TBS tritium systems and equipment and, at the same time, provides an updated overview on the current design status, mainly focusing onto the tritium extractor from Pb-16Li and TBS tritium accountancy. Considerations are also given on the possible extrapolation to DEMO breeding blanket. (authors)

  11. Measurement of Velocity Profiles in a scaled Transparent Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Han; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of); Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Korea has developed two kinds of TBM for ITER; a Helium cooled solid breeder (HCSB) TBM and a Helium cooled molten lithium (HCML) TBM, respectively. Under the HCML TBM project, a 1/6 scaled mock-up of the TBM FW has been fabricated in Korea Atomic Energy Research Institute (KAERI). The size of the scaled mock-up is 260 mm height and 444 mm width. As coolant channels in the mock-up, there are rectangular shape of 10 channels with 10 mm height and 20 mm width. The scaled mock-up was manufactured by hot isostatic pressing bonding method using SS316L. Three components of the scaled mock-up were prepared; a front part of cooling channel 10 mm height with 20 mm width, a front cover plate, and a back plate. The front plate and the cover were bonded by welding, and the welded part and the back plate are attached by HIP process. A pair of manifolds, to distribute the coolant uniformly into 10 channels of the scaled mock-up, were designed and fabricated. The designed manifolds were then welded in inlet and outlet positions of the mock-up. To measure the flow distribution in each channel, the ultrasonic flowmeter (UFM) was used and the values were compared to a conventional flowmeter. Before the flow distribution test of the scaled mock-up, a calibration procedure was conducted with a single channel mock-up using the UFM and the flowmeter. The result showed a good agreement between the UFM and the flowmeter values in the single channel. The same test procedure conducted on the scaled mock-up; the velocity of each channel was measured by the UFM and total mass flow rate was measured with the flowmeter. The estimated velocities distributed from the manifold were simulated by ANSYS-CFX. However, there was a discrepancy between the measured and the simulated values. The current manifold could not provide uniform flow rate to the each channel or there would be a measurement error using the UFM in the specified mock-up. This means that the UFM measurement method should be validated

  12. Power system transient stability simulation based on module bi-directional iteration

    Institute of Scientific and Technical Information of China (English)

    FANG; Dazhong; YANG; Xiaodong

    2005-01-01

    A new simultaneous solution method using module bi-directional iteration is proposed for power system transient stability simulation. In this method, power network is partitioned into a tree hierarchy; computation modules are established for decomposed power networks and various power system components respectively. Through representing every computation module by a computation node, a computation tree is constructed by connecting the nodes together according to their electrical relations in power systems. A tree-traversing procedure called forward reduction and backward evaluation is performed to calculate correction factors of the variables in Newton iterations. This high-efficiency simulation method is feasible to be applied in parallel computation for large interconnected systems. Simulation tests are conducted on the New England 10-generator test power system and the North China-Northeast interconnected system, and the results are compared with those of the commercial software BPA to validate the effectiveness and correctness of this method.

  13. Design of the helium cooled lithium lead breeding blanket in CEA: from TBM to DEMO

    Science.gov (United States)

    Aiello, G.; Aubert, J.; Forest, L.; Jaboulay, J.-C.; Li Puma, A.; Boccaccini, L. V.

    2017-04-01

    The helium cooled lithium lead (HCLL) blanket concept was originally developed in CEA at the beginning of 2000: it is one of the two European blanket concepts to be tested in ITER in the form of a test blanket module (TBM) and one of the four blanket concepts currently being considered for the DEMOnstration reactor that will follow ITER. The TBM is a highly optimized component for the ITER environment that will provide crucial information for the development of the DEMO blanket, but its design needs to be adapted to the DEMO reactor. With respect to the TBM design, reduction of the steel content in the breeding zone (BZ) is sought in order to maximize tritium breeding reactions. Different options are being studied, with the potential of reaching tritium breeding ratio (TBR) values up to 1.21. At the same time, the design of the back supporting structure (BSS), which is a DEMO specific component that has to support the blanket modules inside the vacuum vessel (VV), is ongoing with the aim of maximizing the shielding power and minimizing pumping power. This implies a re-engineering of the modules’ attachment system. Design changes however, will have an impact on the manufacturing and assembly sequences that are being developed for the HCLL-TBM. Due to the differences in joint configurations, thicknesses to be welded, heat dissipation and the various technical constraints related to the accessibility of the welding tools and implementation of non-destructive examination (NDE), the manufacturing procedure should be adapted and optimized for DEMO design. Laser welding instead of TIG could be an option to reduce distortions. The time-of-flight diffraction (TOFD) technique is being investigated for NDE. Finally, essential information expected from the HCLL-TBM program that will be needed to finalize the DEMO design is discussed.

  14. RF TBMs for ITER tests

    Energy Technology Data Exchange (ETDEWEB)

    Kirillov, I.R. [D.V. Efremov Scientific Research Institute of Electrophysical Apparatus, 196641 St. Petersburg (Russian Federation)]. E-mail: kirillir@sintez.niiefa.spb.su; Shatalov, G.E. [Russian Research Center, Kurchatov Institute, Kurchatov Square 1, 123182 Moscow (Russian Federation); Strebkov, YU.S. [Dollezhal Research and Development Institute of Power Engineering, P.O. Box 788, 101000 Moscow (Russian Federation)

    2006-02-15

    Following the DEMO design analysis, two test blanket modules (TBM) were chosen in the RF for the development and testing in ITER: ceramic helium-cooled TBM and lithium self-cooled TBM. In the first one, lithium containing ceramics is used for tritium breeding, helium is used as a coolant and purge gas for tritium extraction, beryllium-as a multiplier. Ferritic steel is a structure material. In the second one lithium is used as tritium breeder and a coolant, and vanadium alloy of V-Cr-Ti system as a structure material. Conceptual designs of both TBMs and ancillary systems for their tests in ITER, strategy of tests, key R and D issues for both concepts are summarized. An international collaboration in R and D, development and testing of TBMs is of great importance due to shortage of testing space in ITER and due to high cost of the program.

  15. A NEW LABELING SEARCH ALGORITHM FOR BIT-INTERLEAVED CODED MODULATION WITH ITERATIVE DECODING

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    Bit-Interleaved Coded Modulation with Iterative Decoding (BICM-ID) is a bandwidth efficient transmission, where the bit error rate is reduced through the iterative information exchange between the inner demapper and the outer decoder. The choice of the symbol mapping is the crucial design parameter. This paper indicates that the Harmonic Mean of the Minimum Squared Euclidean (HMMSE) distance is the best criterion for the mapping design. Based on the design criterion of the HMMSE distance, a new search algorithm to find the optimized labeling maps for BICM-ID system is proposed. Numerical results and performance comparison show that the new labeling search method has a low complexity and outperforms other labeling schemes using other design criterion in BICM-ID system, therefore it is an optimized labeling method.

  16. An Iterative Time Windowed Signature Algorithm for Time Dependent Transcription Module Discovery.

    Science.gov (United States)

    Meng, Jia; Gao, Shou-Jiang; Huang, Yufei

    2008-01-01

    An algorithm for the discovery of time varying modules using genome-wide expression data is present here. When applied to large-scale time serious data, our method is designed to discover not only the transcription modules but also their timing information, which is rarely annotated by the existing approaches. Rather than assuming commonly defined time constant transcription modules, a module is depicted as a set of genes that are co-regulated during a specific period of time, i.e., a time dependent transcription module (TDTM). A rigorous mathematical definition of TDTM is provided, which is serve as an objective function for retrieving modules. Based on the definition, an effective signature algorithm is proposed that iteratively searches the transcription modules from the time series data. The proposed method was tested on the simulated systems and applied to the human time series microarray data during Kaposi's sarcoma-associated herpesvirus (KSHV) infection. The result has been verified by Expression Analysis Systematic Explorer.

  17. Design, fabrication, and testing of a helium-cooled module for the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Baxi, C.B.; Smith, J.P.; Youchison, D.

    1994-08-01

    The International Thermonuclear Reactor (ITER) will have a single-null divertor with total power flow of 200 MW and a peak heat flux of about 5 MW/m{sup 2}. The reference coolant for the divertor is water. However, helium is a viable alternative and offers advantages from safety considerations, such as excellent radiation stability and chemical inertness. In order to prove the feasibility of helium cooling at ITER relevant heat flux conditions, General Atomics designed, fabricated, and tested a helium-cooled divertor module. The module was made from dispersion strengthened copper, with a heat flux surface 25 mm wide and 80 mm long, designed for twice the ITER divertor heat flux. Different techniques were examined to enhance the heat transfer, which in turn reduced the flow and pumping power required to cool the module. It was concluded that an extended surface was the most practical solution. An optimization study was performed to find the best extended surface parameters. The optimum extended surface geometry consisted of fins: 10 mm high, 0.4 mm thick with a 1 mm pitch. It was estimated to require a pumping power of 150 W to remove 20 kW of power. This is more than an order of magnitude reduction in pumping power requirement, compared to smooth surface. The module was fabricated by electric discharge machining (EDM) process. The testing was carried out at SNLA during August 1993. The testing confirmed the design calculations. The peak heat flux during the test was 10 MW/m{sup 2} applied over a surface area of 20 cm{sup 2}. The pumping power calculated from flow rate and pressure drop measurement was about 160 W, which was less than 1% of the power removed. It is planned to test the module to higher temperature limits and higher heat fluxes during coming months. As a result of this effort we conclude that helium cooling of the ITER divertor is feasible without requiring a very large helium pressure or a large pumping power.

  18. RAMI analysis for DEMO HCPB blanket concept cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Dongiovanni, Danilo N., E-mail: danilo.dongiovanni@enea.it [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati (Italy); Pinna, Tonio [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati (Italy); Carloni, Dario [KIT, Institute of Neutron Physics and Reactor Technology (INR) – KIT (Germany)

    2015-10-15

    Highlights: • RAMI (reliability, availability, maintainability and inspectability) preliminary assessment for HCPB blanket concept cooling system. • Reliability block diagram (RBD) modeling and analysis for HCPB primary heat transfer system (PHTS), coolant purification system (CPS), pressure control system (PCS), and secondary cooling system. • Sensitivity analysis on system availability performance. • Failure models and repair models estimated on the base of data from the ENEA fusion component failure rate database (FCFRDB). - Abstract: A preliminary RAMI (reliability, availability, maintainability and inspectability) assessment for the HCPB (helium cooled pebble bed) blanket cooling system based on currently available design for DEMO fusion power plant is presented. The following sub-systems were considered in the analysis: blanket modules, primary cooling loop including pipework and steam generators lines, pressure control system (PCS), coolant purification system (CPS) and secondary cooling system. For PCS and CPS systems an extrapolation from ITER Test Blanket Module corresponding systems was used as reference design in the analysis. Helium cooled pebble bed (HCPB) system reliability block diagrams (RBD) models were implemented taking into account: system reliability-wise configuration, operating schedule currently foreseen for DEMO, maintenance schedule and plant evolution schedule as well as failure and corrective maintenance models. A simulation of plant activity was then performed on implemented RBDs to estimate plant availability performance on a mission time of 30 calendar years. The resulting availability performance was finally compared to availability goals previously proposed for DEMO plant by a panel of experts. The study suggests that inherent availability goals proposed for DEMO PHTS system and Tokamak auxiliaries are potentially achievable for the primary loop of the HCPB concept cooling system, but not for the secondary loop. A

  19. Safety Analysis on Dual-functional Lithium Lead Test Blanket Module With RELAP5%基于 RELAP5的双功能液态锂铅实验包层模块安全分析

    Institute of Scientific and Technical Information of China (English)

    李伟; 田文喜; 秋穗正; 苏光辉

    2013-01-01

    利用嵌入了液态锂铅(LiPb)的热工水力子模块的系统程序RELAP5/MOD3,对双功能液态锂铅(DFLL)实验包层模块(TBM)的安全特性进行评价。对DFLL-TBM 及其辅助冷却系统的稳态运行工况、预期运行事件和相关事故工况进行了建模、计算和分析。计算结果表明,稳态运行时第一壁(FW )结构材料表面最高温度低于允许值550℃。事故工况下氦气泄漏引起的ITER真空室(VV)、窗口设备室(port cell)以及托卡马克冷却水系统大厅拱顶(TCWS vault)的增压均低于ITER要求的限值0.2 MPa。实验包层钢结构不会熔化且可通过辐射换热有效地导出衰变余热。DFLL-TBM 的设计可满足ITER对其热工水力安全方面的要求。%Safety assessment on the dual-functional lithium lead test blanket module (DFLL-TBM) was performed with a modified version of RELAP5/MOD3 code in which the LiPb eutectic thermal-hydraulic sub-module was inserted .The DFLL-TBM and its ancillary cooling systems were modeled to conduct the computation and analysis for steady-state operation ,anticipated operational incidents and relevant accidents .Compu-tational results indicate that the maximum surface temperature of the first wall (FW) structural material is lower than the allowable value of 550 ℃ .For the accident analy-ses ,none of the pressure increases in ITER vacuum vessel (VV) ,port cell and TCWS vault induced by helium leaking is beyond the ITER safety limit of 0.2 MPa .No melting of the TBM box is found and the decay heat can be removed efficiently by the radiation heat transfer .With the current design ,DFLL-TBM can meet the thermal-hydraulic safety requirements from IT ER .

  20. A fast iterative-clique percolation method for identifying functional modules in protein interaction networks

    Institute of Scientific and Technical Information of China (English)

    Penggang SUN; Lin GAO

    2009-01-01

    Accumulating evidence suggests that biological systems are composed of interacting, separable, functional modules-groups of vertices within which connections are dense but between which they are sparse. Identifying these modules is likely through capturing the biologically mean-ingful interactions. In recent years, many algorithms have been developed for detecting such structures. These al-gorithms, however, are computationally demanding, which limits their applications. In this paper, we propose a fast iterative-clique percolation method (ICPM) for identifying overlapping functional modules in protein-protein interac-tion (PPI) networks. Our method is based on clique percola-tion method (CPM), and it not only considers the degree of nodes to minimize the search space (the vertices in k-cliques must have the degree of k - 1 at least), but also converts k-cliques to (k - 1)-cliques. It finds k-cliques by append-ing one node to (k - 1)-cliques. By testing our method on PPI networks, our analysis of the yeast PPI network suggeststhat most of these modules have well-supported biological significance.

  1. The ITER project construction status

    Science.gov (United States)

    Motojima, O.

    2015-10-01

    The pace of the ITER project in St Paul-lez-Durance, France is accelerating rapidly into its peak construction phase. With the completion of the B2 slab in August 2014, which will support about 400 000 metric tons of the tokamak complex structures and components, the construction is advancing on a daily basis. Magnet, vacuum vessel, cryostat, thermal shield, first wall and divertor structures are under construction or in prototype phase in the ITER member states of China, Europe, India, Japan, Korea, Russia, and the United States. Each of these member states has its own domestic agency (DA) to manage their procurements of components for ITER. Plant systems engineering is being transformed to fully integrate the tokamak and its auxiliary systems in preparation for the assembly and operations phase. CODAC, diagnostics, and the three main heating and current drive systems are also progressing, including the construction of the neutral beam test facility building in Padua, Italy. The conceptual design of the Chinese test blanket module system for ITER has been completed and those of the EU are well under way. Significant progress has been made addressing several outstanding physics issues including disruption load characterization, prediction, avoidance, and mitigation, first wall and divertor shaping, edge pedestal and SOL plasma stability, fuelling and plasma behaviour during confinement transients and W impurity transport. Further development of the ITER Research Plan has included a definition of the required plant configuration for 1st plasma and subsequent phases of ITER operation as well as the major plasma commissioning activities and the needs of the accompanying R&D program to ITER construction by the ITER parties.

  2. An Overview of the US DCLL ITER TBM Program

    Energy Technology Data Exchange (ETDEWEB)

    Wong, C. P. C. [General Atomics, San Diego; Abdou, M [University of California, Los Angeles; Dagher, M [University of California, Los Angeles; Katoh, Yutai [ORNL; Kurtz, Richard [Pacific Northwest National Laboratory (PNNL); Malang, Siegfried [Consultant; Marriott, Edward [University of Wisconsin, Madison; Merrill, Brad [Idaho National Laboratory (INL); Messadek, K [University of California, Los Angeles; Morley, N.B. [University of California, Los Angeles; Pint, Bruce A [ORNL; Sawan, M. [University of Wisconsin; Sharafat, S [University of California, Los Angeles; Smolentsev, Sergey [University of California, Los Angeles; Sze, Dai-Kai [University of California, San Diego; Willms, Scott [Los Alamos National Laboratory (LANL); Ying, Alice [University of California, Los Angeles; Youssef, M Z [University of California, Los Angeles

    2010-01-01

    Under the US Fusion Nuclear Science and Technology program, we selected the Dual Coolant Lead Lithium (DCLL) concept as our primary Test Blanket Module (TBM) for testing in ITER. The DCLL blanket concept has the potential to be a high-performance DEMO blanket design with a projected thermal efficiency of >40%. Reduced activation ferritic/martensitic (RAF/M) steel is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled Pb-17Li breeder is circulated for power conversion and for tritium extraction. A SiC-based flow channel insert (FCI) is used as an electrical insulator for magnetohydrodynamic pressure drop reduction from the circulating Pb-17Li and as a thermal insulator to separate the high-temperature Pb-17Li ({approx}650-700 C) from the RAF/M structure, which has a corrosion temperature limit of {approx}480 C. The RAF/M material must also operate at temperatures above 350 C but less than 550 C. We are continuing the development of the mechanical design and performing neutronics, structural and thermal hydraulics analyses of the DCLL TBM module. Prototypical FCI structures were fabricated and further attention was paid to MHD effects and the design of the inboard blanket for DEMO. We are also making progress on related R&D needs to address key areas. This paper is a summary report on the progress and results of recent DCLL TBM development activities.

  3. An overview of the US DCLL ITER-TBM program

    Energy Technology Data Exchange (ETDEWEB)

    Wong, C.P.C., E-mail: wongc@fusion.gat.com [General Atomics, P.O. Box, 85608, San Diego, CA, 92186-5608 (United States); Abdou, M.; Dagher, M. [University of California-Los Angeles, CA (United States); Katoh, Y. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Kurtz, R.J. [Pacific Northwest National Laboratory, Richland, WA (United States); Malang, S. [Fusion Nuclear Technology Consulting, Linkenheim (Germany); Marriott, E.P. [University of Wisconsin, Madison, WI (United States); Merrill, B.J. [Idaho National Laboratory, Idaho Falls, ID (United States); Messadek, K.; Morley, N.B. [University of California-Los Angeles, CA (United States); Sawan, M.E. [University of Wisconsin, Madison, WI (United States); Sharafat, S.; Smolentsev, S. [University of California-Los Angeles, CA (United States); Sze, D.K. [University of California-San Diego, La Jolla, CA (United States); Willms, S. [Los Alamos National Laboratory, Los Alamos, NM (United States); Ying, A.; Youssef, M.Z. [University of California-Los Angeles, CA (United States)

    2010-12-15

    Under the US Fusion Nuclear Science and Technology program, we selected the Dual Coolant Lead Lithium (DCLL) concept as our primary Test Blanket Module (TBM) for testing in ITER. The DCLL blanket concept has the potential to be a high-performance DEMO blanket design with a projected thermal efficiency of >40%. Reduced activation ferritic/martensitic (RAF/M) steel is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled Pb-17Li breeder is circulated for power conversion and for tritium extraction. A SiC-based flow channel insert (FCI) is used as an electrical insulator for magnetohydrodynamic pressure drop reduction from the circulating Pb-17Li and as a thermal insulator to separate the high-temperature Pb-17Li ({approx}650-700 deg. C) from the RAF/M structure, which has a corrosion temperature limit of {approx}480 deg. C. The RAF/M material must also operate at temperatures above 350 deg. C but less than 550 deg. C. We are continuing the development of the mechanical design and performing neutronics, structural and thermal hydraulics analyses of the DCLL TBM module. Prototypical FCI structures were fabricated and further attention was paid to MHD effects and the design of the inboard blanket for DEMO. We are also making progress on related R and D needs to address key areas. This paper is a summary report on the progress and results of recent DCLL TBM development activities.

  4. Breeding blanket for DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Proust, E. (Commissariat a l' Energie Atomique (CEA), DRN/DMT/SERMA, CE, Saclay (France)); Anzidei, L. (ENEA/FUS, C.R.E., Frascati (Italy)); Casini, G. (Commission of the European Communities, Joint Research Center, Ispara (Italy)); Dalle Donne, M. (Kernforschungszentrum Karlsruhe GmbH (Germany)); Giancarli, L. (Commissariat a l' Energie Atomique (CEA), DRN/DMT/SERMA, CE, Saclay (France)); Malang, S. (Kernforschungszentrum Karlsruhe GmbH (Germany))

    1993-03-01

    This paper presents the main design features, their rationale, and the main critical issues for the development, of the four DEMO-relevant blanket concepts presently being investigated within the framework of the European Test-Blanket Development Programme. (orig.)

  5. Breeding blanket for Demo

    Energy Technology Data Exchange (ETDEWEB)

    Proust, E.; Giancarli, L. [CEA Centre d`Etudes de Saclay, 91 - Gif-sur-Yvette (France). Dept. de Mecanique et de Technologie; Anzidei, L. [ENEA, Frascati (Italy). Centro Ricerche Energia; Casini, G. [Commission of the European Communities, Ispra (Italy). Joint Research Centre; Dalle Donne, M.; Malang, S. [Kernforschungszentrum Karlsruhe GmbH (Germany)

    1992-12-31

    This paper presents the main design features, their rationale, and the main critical issues for the development, of the four DEMO-relevant blanket concepts presently investigated within the framework of the European Test-Blanket Development Programme.

  6. Energetic ions in ITER plasmas

    Science.gov (United States)

    Pinches, S. D.; Chapman, I. T.; Lauber, Ph. W.; Oliver, H. J. C.; Sharapov, S. E.; Shinohara, K.; Tani, K.

    2015-02-01

    This paper discusses the behaviour and consequences of the expected populations of energetic ions in ITER plasmas. It begins with a careful analytic and numerical consideration of the stability of Alfvén Eigenmodes in the ITER 15 MA baseline scenario. The stability threshold is determined by balancing the energetic ion drive against the dominant damping mechanisms and it is found that only in the outer half of the plasma ( r / a > 0.5 ) can the fast ions overcome the thermal ion Landau damping. This is in spite of the reduced numbers of alpha-particles and beam ions in this region but means that any Alfvén Eigenmode-induced redistribution is not expected to influence the fusion burn process. The influence of energetic ions upon the main global MHD phenomena expected in ITER's primary operating scenarios, including sawteeth, neoclassical tearing modes and Resistive Wall Modes, is also reviewed. Fast ion losses due to the non-axisymmetric fields arising from the finite number of toroidal field coils, the inclusion of ferromagnetic inserts, the presence of test blanket modules containing ferromagnetic material, and the fields created by the Edge Localised Mode (ELM) control coils in ITER are discussed. The greatest losses and associated heat loads onto the plasma facing components arise due to the use of the ELM control coils and come from neutral beam ions that are ionised in the plasma edge.

  7. Energetic ions in ITER plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Pinches, S. D. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul-lez-Durance Cedex (France); Chapman, I. T.; Sharapov, S. E. [CCFE, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Lauber, Ph. W. [Max-Planck-Institut für Plasmaphysik, EURATOM-Association, Boltzmanstraße 2, D-85748 Garching (Germany); Oliver, H. J. C. [H H Wills Physics Laboratory, University of Bristol, Royal Fort, Tyndall Avenue, Bristol BS8 1TL (United Kingdom); CCFE, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Shinohara, K. [Japan Atomic Energy Agency, Naka, Ibaraki 311-0193 (Japan); Tani, K. [Nippon Advanced Technology Co., Ltd, Naka, Ibaraki 311-0102 (Japan)

    2015-02-15

    This paper discusses the behaviour and consequences of the expected populations of energetic ions in ITER plasmas. It begins with a careful analytic and numerical consideration of the stability of Alfvén Eigenmodes in the ITER 15 MA baseline scenario. The stability threshold is determined by balancing the energetic ion drive against the dominant damping mechanisms and it is found that only in the outer half of the plasma (r/a>0.5) can the fast ions overcome the thermal ion Landau damping. This is in spite of the reduced numbers of alpha-particles and beam ions in this region but means that any Alfvén Eigenmode-induced redistribution is not expected to influence the fusion burn process. The influence of energetic ions upon the main global MHD phenomena expected in ITER's primary operating scenarios, including sawteeth, neoclassical tearing modes and Resistive Wall Modes, is also reviewed. Fast ion losses due to the non-axisymmetric fields arising from the finite number of toroidal field coils, the inclusion of ferromagnetic inserts, the presence of test blanket modules containing ferromagnetic material, and the fields created by the Edge Localised Mode (ELM) control coils in ITER are discussed. The greatest losses and associated heat loads onto the plasma facing components arise due to the use of the ELM control coils and come from neutral beam ions that are ionised in the plasma edge.

  8. Neutronics Comparison Analysis of the Water Cooled Ceramics Breeding Blanket for CFETR

    Science.gov (United States)

    Li, Jia; Zhang, Xiaokang; Gao, Fangfang; Pu, Yong

    2016-02-01

    China Fusion Engineering Test Reactor (CFETR) is an ITER-like fusion engineering test reactor that is intended to fill the scientific and technical gaps between ITER and DEMO. One of the main missions of CFETR is to achieve a tritium breeding ratio that is no less than 1.2 to ensure tritium self-sufficiency. A concept design for a water cooled ceramics breeding blanket (WCCB) is presented based on a scheme with the breeder and the multiplier located in separate panels for CFETR. Based on this concept, a one-dimensional (1D) radial built breeding blanket was first designed, and then several three-dimensional models were developed with various neutron source definitions and breeding blanket module arrangements based on the 1D radial build. A set of nuclear analyses have been carried out to compare the differences in neutronics characteristics given by different calculation models, addressing neutron wall loading (NWL), tritium breeding ratio (TBR), fast neutron flux on inboard side and nuclear heating deposition on main in-vessel components. The impact of differences in modeling on the nuclear performance has been analyzed and summarized regarding the WCCB concept design. supported by the National Special Project for Magnetic Confined Nuclear Fusion Energy (Nos. 2013GB108004, 2014GB122000, and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

  9. Fabrication study on the cooling module of the ITER neutral beam duct liner

    Energy Technology Data Exchange (ETDEWEB)

    Sa, J.W. [National Fusion Research Institute, Yuseong-gu, Daejeon (Korea, Republic of); Kim, H.S., E-mail: hskim@nfri.re.k [National Fusion Research Institute, Yuseong-gu, Daejeon (Korea, Republic of); Kim, B.Y.; Kim, B.C.; Ahn, H.J.; Bak, J.S. [National Fusion Research Institute, Yuseong-gu, Daejeon (Korea, Republic of); Jung, H.J. [Korean Intellectual Property Office, Seo-gu, Daejeon (Korea, Republic of); Han, M.H.; Hong, C.D.; Lee, J.S.; Kim, Y.K. [Hyundai Heavy Industries Co. Ltd., Dong-gu, Ulsan (Korea, Republic of); Urbani, M. [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Geli, F. [UKAEA Culham Division, Oxfordshire OX14 3DB, Abingdon (United Kingdom)

    2010-12-15

    Recently the new concept of the ITER neutral beam (NB) duct liner has been developed to improve thermo-mechanical performance and satisfy the requirements for remote handling and maintenance. The design concept of cooling module located inside neutron shield structure is to use deep-drilled panels instead of the original design concept of the casting-modularized component with tubes. In this study, the manufacturing feasibility has been investigated through the fabrication of small size coupons and full scale mock-up. Firstly, the small size coupons are for developing the electron beam welding processes. Secondly, the full scale mock-up which has 6 holes for cooling passage has been fabricated in order to develop the main fabrication processes such as deep drilling, bending and machining. In addition, the pressure and the leak tests have been carried out to check the required performance for completed cooling panel. Although some improvement is required, but the Electron Beam Welding (EBW) has been successfully achieved and generally the deep drilling and bending process also shown good results in dimensional control.

  10. Progress on the ITER TBM port plug design

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung-Yoon, E-mail: byoungyoon.kim@iter.org [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Marconi, Matteo [LTCalcoli, Piazza Prinetti 26/B, 23807 Merate (Italy); Neviere, Jean-Christophe [Comex Nucléaire, 13115 Saint Paul Lez Durance (France); Merola, Mario; Giancarli, Luciano M. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Lucca, Flavio [LTCalcoli, Piazza Prinetti 26/B, 23807 Merate (Italy); Gicquel, Stefan [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

    2015-10-15

    Highlights: • ITER TBM PP design progress with two Dummy TBMs was outlined. • Attachment of Dummy TBMs in TBM Frame was assessed to investigate mechanical and sealing performance. • Implementation of cover plates was described and the preliminary maintenance sequence was developed. • Clearance between TBM Frame and Dummy TBMs were reduced to have better shielding performance. • Cooling layout and castellation design were updated to minimize thermal induced stress. - Abstract: In three of the ITER equatorial ports, tritium breeding blanket concepts will be validated and tested using mock-up breeding blankets called test blanket modules (TBM). In these ports, two TBM-Sets are mechanically attached in a TBM Frame to form a TBM port plug (TBM PP). The ITER Organization is responsible for the design and manufacture of both this TBM Frame and the Dummy TBMs which will fill them. As a part of this development, in 2013, a conceptual design review (CDR) of TBM PP with two dummy TBMs revealed the need for improvement of design performance, interfaces and maintainability. This paper presents the main design improvements after the CDR as well as associated feasibility analysis of the improved design focusing on the attachment. Finally the work plan for the Preliminary Design phase is summarized.

  11. Conceptual design of Blanket Remote Handling System for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Wei, Jianghua, E-mail: weijh@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Song, Yuntao, E-mail: songyt@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); University of Science and Technology of China, Hefei (China); Pei, Kun; Zhao, Wenlong; Zhang, Yu; Cheng, Yong [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China)

    2015-11-15

    Highlights: • The concept for the blanket maintenance is carried out, including three sub-systems. • The basic maintenance procedure for blanket between VV and hot cell is carried out. • The primary kinematics study is used to verify the feasibility of BRHS. • Virtual reality is adopted as another approach to verify the concept design. - Abstract: The China Fusion Engineering Testing Reactor (CFETR), which is a new superconducting tokamak device being designed by China, has a mission to achieve a high duty time (0.3–0.5). To accomplish this great mission, the big modular blanket option has been adopted to achieve the high efficiency of the blanket maintenance. Considering this mission and the large and heavy blanket module, a novel conceptual blanket maintenance system for CFETR has been carried out by us over the past year. This paper presents the conceptual design of the Blanket Remote Handling System (BRHS), which mainly comprises the In-Vessel-Maintenance-System (IVMS), Lifting System and Blanket-Tool-Manipulator System (BTMS). The BRHS implements the extraction and replacement between in-vessel (the blanket module operation configuration location) and ex-vessel (inside of the vertical maintenance cask) by the collaboration of these three sub systems. What is more, this paper represents the blanket maintenance procedure between the docking station (between hot cell building and tokamak building) and inside the vacuum vessel, in tokamak building. Virtual reality technology is also used to verify and optimize our concept design.

  12. Preliminary structural design and thermo-mechanical analysis of helium cooled solid breeder blanket for Chinese Fusion Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Li, Min; Chen, Hongli, E-mail: hlchen1@ustc.edu.cn; Zhou, Guangming; Liu, Qianwen; Wang, Shuai; Lv, Zhongliang; Ye, Minyou

    2015-02-15

    Highlights: • A helium cooled solid breeder blanket module was designed for CFETR. • Multilayer U-shaped pebble beds were adopted in the blanket module. • Thermal and thermo-mechanical analyses were carried out under normal operating conditions. • The analysis results were found to be acceptable. - Abstract: With the aim to bridge the R&D gap between ITER and fusion power plant, the Chinese Fusion Engineering Test Reactor (CFETR) was proposed to be built in China. The mission of CFETR is to address the essential R&D issues for achieving practical fusion energy. Its blanket is required to be tritium self-sufficient. In this paper, a helium cooled solid breeder blanket adopting multilayer U-shaped pebble beds was designed and analyzed. Thermo-mechanical analysis of the first wall and side wall combined with breeder unit was carried out for normal operating steady state conditions. The results showed that the maximum temperatures of the structural material, neutron multiplier and tritium breeder pebble beds are 523 °C, 558 °C and 787 °C, respectively, which are below the corresponding limits of 550 °C, 650 °C and 920 °C. The maximum equivalent stress of the structure is under the allowable value with a margin about 14.5%.

  13. Demonstration Tokamak Hybrid Reactor (DTHR) blanket design study, December 1978

    Energy Technology Data Exchange (ETDEWEB)

    1978-01-01

    This work represents only the second iteration of the conceptual design of a DTHR blanket; consequently, a number of issues important to a detailed blanket design have not yet been evaluated. The most critical issues identified are those of two-phase flow maldistribution, flow instabilities, flow stratification for horizontal radial inflow of boiling water, fuel rod vibrations, corrosion of clad and structural materials by high quality steam, fretting and cyclic loads. Approaches to minimizing these problems are discussed and experimental testing with flow mock-ups is recommended. These implications on a commercial blanket design are discussed and critical data needs are identified.

  14. Multiphysics Engineering Analysis for an Integrated Design of ITER Diagnostic First Wall and Diagnostic Shield Module Design

    Energy Technology Data Exchange (ETDEWEB)

    Zhai, Y. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Loesser, G. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Smith, M. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Udintsev, V. [ITER Org, F-13115 St Paul Les Durance, France.; Giacomin, T., T. [ITER Org, F-13115 St Paul Les Durance, France.; Khodak, A. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Johnson, D, [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Feder, R, [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)

    2015-07-01

    ITER diagnostic first walls (DFWs) and diagnostic shield modules (DSMs) inside the port plugs (PPs) are designed to protect diagnostic instrument and components from a harsh plasma environment and provide structural support while allowing for diagnostic access to the plasma. The design of DFWs and DSMs are driven by 1) plasma radiation and nuclear heating during normal operation 2) electromagnetic loads during plasma events and associate component structural responses. A multi-physics engineering analysis protocol for the design has been established at Princeton Plasma Physics Laboratory and it was used for the design of ITER DFWs and DSMs. The analyses were performed to address challenging design issues based on resultant stresses and deflections of the DFW-DSM-PP assembly for the main load cases. ITER Structural Design Criteria for In-Vessel Components (SDC-IC) required for design by analysis and three major issues driving the mechanical design of ITER DFWs are discussed. The general guidelines for the DSM design have been established as a result of design parametric studies.

  15. Integration of remote refurbishment performed on ITER components

    Energy Technology Data Exchange (ETDEWEB)

    Dammann, A., E-mail: alexis.dammann@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Antola, L. [AMEC, 31 Parc du Golf, CS 90519, 13596 Aix en Provence (France); Beaudoin, V. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Dremel, C. [Westinghouse, Electrique France/Astare, 122 Avenue de Hambourg, 13008 Marseille (France); Evrard, D. [SOGETI High Tech, 180 Rue René Descartes, 13851 Aix en Provence (France); Friconneau, J.P. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Lemée, A. [SOGETI High Tech, 180 Rue René Descartes, 13851 Aix en Provence (France); Levesy, B.; Pitcher, C.S. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2015-10-15

    Highlights: • System engineering approach to consolidate requirements to modify the layout of the Hot Cell. • Illustration of the loop between requirement and design. • Verification process. - Abstract: Internal components of the ITER Tokamak are replaced and transferred to the Hot Cell by remote handling equipment. These components include port plugs, cryopumps, divertor cassettes, blanket modules, etc. They are brought to the refurbishment area of the ITER Hot Cell Building for cleaning and maintenance, using remote handling techniques. The ITER refurbishment area will be unique in the world, when considering combination of size, quantity of complex component to refurbish in presence of radiation, activated dust and tritium. The refurbishment process to integrate covers a number of workstations to perform specific remote operations fully covered by a mast on crane system. This paper describes the integration of the Refurbishment Area, explaining the functions, the methodology followed, some illustrations of trade-off and safety improvements.

  16. A photoelastic-modulator-based motional Stark effect polarimeter for ITER that is insensitive to polarized broadband background reflections

    Science.gov (United States)

    Thorman, A.; Michael, C.; De Bock, M.; Howard, J.

    2016-07-01

    A motional Stark effect polarimeter insensitive to polarized broadband light is proposed. Partially polarized background light is anticipated to be a significant source of systematic error for the ITER polarimeter. The proposed polarimeter is based on the standard dual photoelastic modulator approach, but with the introduction of a birefringent delay plate, it generates a sinusoidal spectral filter instead of the usual narrowband filter. The period of the filter is chosen to match the spacing of the orthogonally polarized Stark effect components, thereby increasing the effective signal level, but resulting in the destructive interference of the broadband polarized light. The theoretical response of the system to an ITER like spectrum is calculated and the broadband polarization tolerance is verified experimentally.

  17. 扩散连接技术在核聚变反应堆包层模块制造中的应用%Application of Diffusion Bonding Technique in Fabrication of Blanket Module Components of Nuclear Fusion Reactor

    Institute of Scientific and Technical Information of China (English)

    刘晨曦; 刘永长; 周晓胜; 马宗青; 王颖; 李会军; 杨建国

    2015-01-01

    国际受控热核聚变实验堆计划是全球规模最大、影响最深远的国际科研合作项目之一,有望彻底解决能源危机。核聚变反应堆关键部件———包层模块的结构复杂、体积庞大,且服役环境恶劣,焊接接头成为影响反应堆安全运行的薄弱环节。以扩散连接为代表的固相焊接技术对接头性能及组织影响较小,已逐渐取代熔化焊应用于包层模块复杂构件制造。在简要介绍扩散连接及其原理的基础上,对包层模块构件扩散连接的研究进展进行了阐述,包括低活化铁素体/马氏体钢及氧化物弥散强化钢构件的扩散连接,Be,W,SiC等其他先进高温材料的扩散连接等。%International Thermonuclear Experimental Reactor is one of the world′s largest and the most far-reaching in-ternational scientific collaborative projects, which is expected to solve the energy crisis.As a key built-up part, blanket module has complex structure and large size, and serves under harsh service conditions.The welding joints of blanket mod-ule have become the weak links affecting the operation of the nuclear fusion reactor.Solid-phase welding technology, repre-sented by diffusion bonding, have relatively low effect on the mechanical properties and microstructure of the joints, and has gradually taken the place of the fusion welding technology used for fabrication of the blanket module complex compo-nents.Based on the brief presentation of diffusion bondingand its bonding mechanism, the research progress in diffusion bonding of blanket module components was discussed in this paper, including the diffusion bonding of reduced activation ferritic/martensitic steels and oxide dispersion strengthened steels, and the diffusion bonding of Be, W, SiC and/or other advanced high-temperature materials.

  18. Performance assessment of the antenna setup for the ITER plasma position reflectometry in-vessel systems

    Science.gov (United States)

    Varela, P.; Belo, J. H.; Quental, P. B.

    2016-11-01

    The design of the in-vessel antennas for the ITER plasma position reflectometry diagnostic is very challenging due to the need to cope both with the space restrictions inside the vacuum vessel and with the high mechanical and thermal loads during ITER operation. Here, we present the work carried out to assess and optimise the design of the antenna. We show that the blanket modules surrounding the antenna strongly modify its characteristics and need to be considered from the early phases of the design. We also show that it is possible to optimise the antenna performance, within the design restrictions.

  19. Development of in-vessel components of the microfission chamber for ITER.

    Science.gov (United States)

    Ishikawa, M; Kondoh, T; Ookawa, K; Fujita, K; Yamauchi, M; Hayakawa, A; Nishitani, T; Kusama, Y

    2010-10-01

    Microfission chambers (MFCs) will measure the total neutron source strength in ITER. The MFCs will be installed behind blanket modules in the vacuum vessel (VV). Triaxial mineral insulated (MI) cables will carry signals from the MFCs. The joint connecting triaxial MI cables in the VV must be considered because the MFCs and the MI cables will be installed separately at different times. Vacuum tight triaxial connector of the MI cable has been designed and a prototype has been constructed. Performance tests indicate that the connector can be applied to the ITER environment. A small bending-radius test of the MI cable indicates no observed damage at a curvature radius of 100 mm.

  20. Performance assessment of the antenna setup for the ITER plasma position reflectometry in-vessel systems.

    Science.gov (United States)

    Varela, P; Belo, J H; Quental, P B

    2016-11-01

    The design of the in-vessel antennas for the ITER plasma position reflectometry diagnostic is very challenging due to the need to cope both with the space restrictions inside the vacuum vessel and with the high mechanical and thermal loads during ITER operation. Here, we present the work carried out to assess and optimise the design of the antenna. We show that the blanket modules surrounding the antenna strongly modify its characteristics and need to be considered from the early phases of the design. We also show that it is possible to optimise the antenna performance, within the design restrictions.

  1. Preliminary design of the ITER ECH Upper Launcher

    Energy Technology Data Exchange (ETDEWEB)

    Strauss, D., E-mail: dirk.strauss@kit.edu [Karlsruhe Institute of Technology, Assoc. KIT-EURATOM, D-76021 Karlsruhe (Germany); Aiello, G. [Karlsruhe Institute of Technology, Assoc. KIT-EURATOM, D-76021 Karlsruhe (Germany); Chavan, R. [Centre de Recherches en Physique des Plasmas, CRPP–EPFL, CH-1015 Lausanne (Switzerland); Cirant, S. [Istituto di Fisica del Plasma CNR, Euratom Association, 20125 Milano (Italy); Baar, M. de [FOM, Van Vollenhovenlaan 659, 3527 JP, Utrecht (Netherlands); Farina, D. [Istituto di Fisica del Plasma CNR, Euratom Association, 20125 Milano (Italy); Gantenbein, G. [Karlsruhe Institute of Technology, Assoc. KIT-EURATOM, D-76021 Karlsruhe (Germany); Goodman, T.P. [Centre de Recherches en Physique des Plasmas, CRPP–EPFL, CH-1015 Lausanne (Switzerland); Henderson, M.A. [ITER Organization, 13108 Saint-Paul-lez-Durance (France); Kasparek, W. [Institut für Plasmaforschung, IPF, D-70569 Stuttgart (Germany); Kleefeldt, K. [Karlsruhe Institute of Technology, Assoc. KIT-EURATOM, D-76021 Karlsruhe (Germany); Landis, J.-D. [Centre de Recherches en Physique des Plasmas, CRPP–EPFL, CH-1015 Lausanne (Switzerland); Meier, A. [Karlsruhe Institute of Technology, Assoc. KIT-EURATOM, D-76021 Karlsruhe (Germany); Moro, A.; Platania, P. [Istituto di Fisica del Plasma CNR, Euratom Association, 20125 Milano (Italy); Plaum, B. [Institut für Plasmaforschung, IPF, D-70569 Stuttgart (Germany); Poli, E. [Max-Planck-IPP, Euratom Association, D-85748 Garching (Germany); Ramponi, G. [Istituto di Fisica del Plasma CNR, Euratom Association, 20125 Milano (Italy); Ronden, D. [FOM, Van Vollenhovenlaan 659, 3527 JP, Utrecht (Netherlands); Saibene, G. [Fusion for Energy, Barcelona (Spain); and others

    2013-11-15

    Highlights: • Front steering mirror design. • Plasma facing blanket shield module/first wall panel design. • Fixed frequency torus CVD diamond window serving as first tritium barrier. • Prototypes and tests of the above key components in the Launcher Handling and Testing Facility. -- Abstract: The design of the ITER electron cyclotron launchers recently reached the preliminary design level - the last major milestone before design finalization. The ITER ECH system contains 24 installed gyrotrons providing a maximum ECH injected power of 20 MW through transmission lines towards the tokamak. There are two EC launcher types both using a front steering mirror; one equatorial launcher (EL) for plasma heating and four upper launchers (UL) for plasma mode stabilization (neoclassical tearing modes and the sawtooth instability). A wide steering angle range of the ULs allows focusing of the beam on magnetic islands which are expected on the rational magnetic flux surfaces q = 1 (sawtooth instability), q = 3/2 and q = 2 (NTMs). In this paper the preliminary design of the ITER ECH UL is presented, including the optical system and the structural components. Highlights of the design include the torus CVD-diamond windows, the frictionless, front steering mechanism and the plasma facing blanket shield module (BSM). Numerical simulations as well as prototype tests are used to verify the design.

  2. Neutronics Analysis of the ITER In-Vessel Viewing System

    CERN Document Server

    Turner, Andrew; Puiu, Adrian

    2013-01-01

    The In-Vessel Viewing System (IVVS) in ITER consists of six identical units which are deployed between pulses or during shutdown, to perform visual examination and metrology of plasma facing components. The system is housed in dedicated ports at B1 level, with deployment at the level between the divertor cassettes and the lowermost outboard blanket modules. Boron carbide shielding blocks are envisaged to protect the sensitive components of the IVVS from damage during operations, and personnel from radiation fields. In order to progress the design of the IVVS beyond the pre-conceptual stage, analyses were conducted using MCNP to determine the acceptability of a series of different shielding configurations.

  3. An overview of the US DCLL ITER-TBM program

    Energy Technology Data Exchange (ETDEWEB)

    Wong, C. P. C.; Abdou, M.; Dagher, M.; Katoh, Y.; Kurtz, R. J.; Malang, S.; Marriott, E. P.; Merrill, B. J.; Messadek, K.; Morley, N. B.; Sawan, M. E.; Sharafat, S.; Smolentsev, S.; Sze, D. K.; Willms, S.; Ying, A.; Youssef, M. Z.

    2010-12-01

    Under the US Fusion Nuclear Science and Technology program, we selected the Dual Coolant Lead Lithium (DCLL) concept as our primary Test Blanket Module (TBM) for testing in ITER. The DCLL blanket concept has the potential to be a high-performance DEMO blanket design with a projected thermal efficiency of >40%. Reduced activation ferritic/martensitic (RAF/M) steel is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled Pb-17Li breeder is circulated for power conversion and for tritium extraction. A SiC-based flow channel insert (FCI) is used as an electrical insulator for magnetohydrodynamic pressure drop reduction from the circulating Pb-17Li and as a thermal insulator to separate the high-temperature Pb-17Li (~650–700 °C) from the RAF/M structure, which has a corrosion temperature limit of ~480 °C. The RAF/M material must also operate at temperatures above 350 °C but less than 550 °C. We are continuing the development of the mechanical design and performing neutronics, structural and thermal hydraulics analyses of the DCLL TBM module. Prototypical FCI structures were fabricated and further attention was paid to MHD effects and the design of the inboard blanket for DEMO. We are also making progress on related R&D needs to address key areas. Finally, this paper is a summary report on the progress and results of recent DCLL TBM development activities.

  4. Flexible armored blanket development

    Energy Technology Data Exchange (ETDEWEB)

    Roth, E.S.

    1978-05-01

    An exploratory development contract was undertaken on December 23, 1977 which had as its purpose the development and demonstration of a flexible armored blanket design suitable for providing ballistic protection to nuclear weapons during shipment. Objectives were to design and fabricate a prototype blanket which will conform to the weapon shape, is troop-handleable in the field, and which, singly or in multiple layers, can defeat a range of kinetic energy armor piercing (AP) ammunition potentially capable of damaging the critical portion of the nuclear weapon. Following empirical testing, including the firing of threat ammunition under controlled laboratory and field test conditions, materials were selected and assembled into two blanket designs, each weighing approximately 54 kg/m{sup 2} (11 lbs/ft{sup 2}) and estimated to cost from $111 to $180 per ft{sup 2} in production. A firing demonstration to evidence blanket performance against terrorist/light infantry weapons, heavy infantry weapons, and aircraft cannon was conducted for representatives of the DOD and interested Sandia employees on April 12, 1978. The blankets performed better than anticipated defeating bullets up to 7.62 mm x 51 mm AP with one layer and projectiles up to 23 mm HEI with two layers. Based on these preliminary tests it is recommended that development work be continued with the following objectives: (1) the selection by the DOD of priority applications, (2) the specific design and fabrication of sufficient quantities of armored blankets for field testing, (3) the evaluation of the blankets by DOD operational units, with reports to Sandia Laboratories to enable final design.

  5. Neutronic analysis for bolometers in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Suarez, A., E-mail: alejandro.suarez@iter.org [CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain); Reichle, R.; Loughlin, M.; Polunovskiy, E.; Walsh, M. [ITER Organization, Route de Vinon sur Verdon, 13115, St. Paul lez Durance (France)

    2013-10-15

    Highlights: ► Radiation damage calculations for the bolometers in ITER. ► Redesign of the bolometric diagnostic in EPP01. ► New bolometer radiation damage values in EPP01 in the safe zone. -- Abstract: Neutronic considerations in ITER have such importance that they drive the design of many diagnostics and components of the machine, and bolometers are not an exception. Bolometer cameras will be installed on the vacuum vessel, viewing the plasma through the gaps between blanket modules, divertor, equatorial and upper port plugs. The ITER reference bolometer sensors are of a resistive type. For this study it is assumed that they are composed of a thin silicon nitride carrier film and platinum resistors disposed in a Wheatstone bridge configuration. Their assumed radiation hardness is 0.1 dpa. Neutronic calculations were performed with the Monte Carlo program MCNP5, the FENDL 2.1 nuclear data library and the latest B-lite ITER neutronic model with the appropriate modifications using the CAD to MCNP converter MCAM. A complete characterization of the neutron fluxes in all the bolometer locations and the calculation of neutron damage were performed. Values above the failure threshold damage were obtained for some of the bolometers, leading to a complete redesign of some parts of the bolometric system in order to extend its lifetime.

  6. Development of liquid metal type TBM technology for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Bong Guen; Kwak, J. G.; Kim, Y. (and others)

    2008-03-15

    The objectives of the ITER project for the construction and operation are to perform the test related to the neutronics, blanket module, tritium treatment technology, advanced plasma technology, and to test the heat extraction and tritium breeding in the test blanket for the fusion reactor. Other parties have been developing the Test Blanket Module (TBM) for testing in the ITER for these purposes. Through this project, we can secure the TBM design and related technology, which will be used as the core technology for the DEMO construction, our own fusion reactor development. In 1st year, the optimized design procedure was established with the existing tools, which have been used in nuclear reactor design, and the optimized HCML TBM design was obtained through iteration method according to the developed design procedure. He cooling system as a TBM auxiliary system was designed considering the final design of the KO HCML TBM such as coolant capacity and operation pressure. Layout for this system was prepared to be installed in the ITER TCWS vault. MHD effect of liquid Li breeder by magnetic flux in ITER such as much higher pressure drop was evaluated with CFD-ACE and it was concluded that the Li breeder should have a slow velocity to reduce this effect. Most results were arranged in the form of DDD including preliminary safety analysis report. In 2nd year, the optimized design procedure was complemented and updated. In performance analysis on thermal-hydraulic and thermo-mechanical one, full 3D meshes were generated and used in this analysis in order to obtain the more exact temperature, deformation, and stress solution. For liquid Li breeder system, design parameters were induced before the detailed design of the system and were used in the design of the liquid Li test loop. LOCA analysis, activation analysis in LOCA, EM analysis were performed as a preliminary safety analysis. In order to develop the manufacturing technology, Be+FMS and FMS to FMS joining conditions

  7. Tokamak blanket design study, final report

    Energy Technology Data Exchange (ETDEWEB)

    1980-08-01

    A cylindrical module concept was developed, analyzed, and incorporated in a tokamak blanket system that includes piping systems, vacuum boundary sealing, and support structures. The design is based on the use of state-of-the-art structural materials (20% cold-worked type 316 stainless steel), lithium as the breeding material, and pressurized helium as the coolant. The module design consists of nested concentric cylinders (with an outer diameter of 10 cm) and features direct wall cooling by helium flowing between the outer (first-wall) cylinder and the inner (lithium-containing) cylinder. Each cylinder can withstand full coolant pressure, thus enhancing reliability. Results show that stainless steel is a viable material for a first wall subjected to a neutron wall loading of 4 MW/m/sup 2/ and a particle heat flux of 1 MW/m/sup 2/. Lifetime analysis shows that the first-wall design meets the goal of operating at 20-min cycles with 95% duty for 100,000 cycles. To reduce system complexity, a larger 20-cm-diam module also was analyzed for incorporation in the blanket assembly. Reliability assessment indicates that it may be possible to double the module in size from 10 to 20 cm in diameter. With a modest increase in coolant pumping power, a blanket assembly comprising 20-cm-diam modules can still achieve 100,000 operating cycles - equivalent to a 3.6-year design lifetime - with only one or two helium coolant leaks into the plasma.

  8. The Haida Button Blanket.

    Science.gov (United States)

    Johnson, Vesta

    In the Haida nation, there are two phratries, Eagle and Raven, divided into a number of clans sharing one or more emblems. These emblems, inherited from the mother's line, adorn the button blankets which are the traditional ceremonial robes that serve to identify the family of the wearer. Written instructions and diagrams guide students in…

  9. TIG and HIP joining of Reduced Activation Ferrite/Martensitic steel for the Korean ITER-TBM

    Energy Technology Data Exchange (ETDEWEB)

    Ku, Duck Young [ITER Korea, National Fusion Research Institute, Gwahangno 113, Yuseong-gu, Daejeon 305-333 (Korea, Republic of); Oh, Seungjin, E-mail: osj0423@gmail.com [Nuclear Engineering and Technology Institute, KHNP, 508 Keumbyeong-ro, Yuseong-gu, Daejeon 305-343 (Korea, Republic of); Ahn, Mu-Young; Yu, In-Keun; Kim, Duck-Hoi; Cho, Seungyon [ITER Korea, National Fusion Research Institute, Gwahangno 113, Yuseong-gu, Daejeon 305-333 (Korea, Republic of); Choi, Im-Sub; Kwon, Ki-Bum [Soles Co. LTD., 9-1 Horim-dong, Dalseo-gu, Daegu 704-249 (Korea, Republic of)

    2011-10-01

    Korea is developing a Helium Cooled Solid Breeder Test Blanket Module for ITER. The primary candidate structural material is a Reduced Activation Ferritic/Martensitic steel. The complex TBM structure requires developing joining technologies for successful fabrication. The characteristics of Tungsten Inert Gas welding and Hot Isostatic Pressing joining of RAFM steel were investigated. Metallurgical examinations showed that the steel grain size was increased after HIP joining and recovered by post joining heat treatment. Both TIG welding and HIP joining are found to be acceptable for ITER TBM based on mechanical tests and microstructure examination.

  10. Engineering an achromatic Bessel beam using a phase-only spatial light modulator and an iterative Fourier transformation algorithm

    Science.gov (United States)

    Walde, Marie; Jost, Aurélie; Wicker, Kai; Heintzmann, Rainer

    2017-01-01

    Bessel illumination is an established method in optical imaging and manipulation to achieve an extended depth of field without compromising the lateral resolution. When broadband or multicolour imaging is required, wavelength-dependent changes in the radial profile of the Bessel illumination can complicate further image processing and analysis. We present a solution for engineering a multicolour Bessel beam that is easy to implement and promises to be particularly useful for broadband imaging applications. A phase-only spatial light modulator (SLM) in the image plane and an iterative Fourier Transformation algorithm (IFTA) are used to create an annular light distribution in the back focal plane of a lens. The 2D Fourier transformation of such a light ring yields a Bessel beam with a constant radial profile for different wavelength.

  11. Automatic Frequency Identification under Sample Loss in Sinusoidal Pulse Width Modulation Signals Using an Iterative Autocorrelation Algorithm

    Directory of Open Access Journals (Sweden)

    Alejandro Said

    2016-08-01

    Full Text Available In this work, we present a simple algorithm to calculate automatically the Fourier spectrum of a Sinusoidal Pulse Width Modulation Signal (SPWM. Modulated voltage signals of this kind are used in industry by speed drives to vary the speed of alternating current motors while maintaining a smooth torque. Nevertheless, the SPWM technique produces undesired harmonics, which yield stator heating and power losses. By monitoring these signals without human interaction, it is possible to identify the harmonic content of SPWM signals in a fast and continuous manner. The algorithm is based in the autocorrelation function, commonly used in radar and voice signal processing. Taking advantage of the symmetry properties of the autocorrelation, the algorithm is capable of estimating half of the period of the fundamental frequency; thus, allowing one to estimate the necessary number of samples to produce an accurate Fourier spectrum. To deal with the loss of samples, i.e., the scan backlog, the algorithm iteratively acquires and trims the discrete sequence of samples until the required number of samples reaches a stable value. The simulation shows that the algorithm is not affected by either the magnitude of the switching pulses or the acquisition noise.

  12. ITER test programme

    Science.gov (United States)

    Abdou, M.; Baker, C.; Casini, G.

    1991-07-01

    The International Thermonuclear Experimental Reactor (ITER) was designed to operate in two phases. The first phase, which lasts for 6 years, is devoted to machine checkout and physics testing. The second phase lasts for 8 years and is devoted primarily to technology testing. This report describes the technology test program development for ITER, the ancillary equipment outside the torus necessary to support the test modules, the international collaboration aspects of conducting the test program on ITER, the requirements on the machine major parameters and the R and D program required to develop the test modules for testing in ITER.

  13. High precision three-dimensional iterative indoor localization algorithm using code division multiple access modulation based on visible light communication

    Science.gov (United States)

    Guan, Weipeng; Wu, Yvxiang; Wen, Shangsheng; Yang, Chen; Chen, Hao; Zhang, Zhaoze; Chen, Yingcong

    2016-10-01

    To solve the problem of positioning accuracy affected by mutual interference among multiple reference points in the traditional visible light communication positioning system, an iterative algorithm of received signal strength (RSS) based on code division multiple access (CDMA) coding is proposed. Every light-emitting diode (LED) source broadcasts a unique CDMA modulation identification (ID) code, which is associated with geographic position. The mobile terminal receives a mixed light signal from each LED reference point. By using the orthogonality of the spreading codes, the corresponding ID position information and the intensity attenuation factor of each LED reference point source can be available. According to the ID information and signal attenuation intensity, the location area of each LED and the distance between the receiver end and each LED can be determined. The three-dimensional (3-D) position of the receiver can be obtained by using the iterative algorithm of RSS triangulation. The experimental results show that the proposed algorithm can achieve a positioning accuracy of 5.25 cm in a two-dimensional (2-D) positioning system. And in the 3-D positioning system, the maximum positioning error is 10.27 cm, the minimum positioning error is 0.45 cm, the average positioning error is 3.97 cm, and the proportion of the positioning error exceeding 5 cm is <25%. With a very good positioning accuracy, this system is simple and does not require synchronization processing. What is more, it can be applied to both the 2-D and 3-D localization systems, which has a broad application prospect.

  14. SERIALLY CONCATENATED CONTINUOUS PHASE MODULATION WITH REDUCED ITERATIVE DEMODULATION AND DETECTION

    Institute of Scientific and Technical Information of China (English)

    Sun Jinhua; Li Jiandong; Jin Lijun

    2007-01-01

    A reduced state Soft Input Soft Output (SISO) a posteriori probability algorithm for Serially Concatenated Continuous Phase Modulation (SCCPM) is proposed in this paper. Based on the Reduced State Sequence Detection (RSSD), it has more general form compared with other reduced state SISO algorithms. The proposed algorithm can greatly reduce the state number, thus leads to the computation complexity reduction. It also minimizes the degradation in Euclidean distance with decision feedback in the reduced state trellis. Analysis and simulation results show that the performance degradation is little with proper reduction scheme.

  15. A reduced state SISO iterative decoding algorithm for serially concatenated continuous phase modulation

    Institute of Scientific and Technical Information of China (English)

    SUN JinHua; LI JianDong; JIN LiJun

    2007-01-01

    In this paper, serially concatenated continuous phase modulation (SCCPM) system is analyzed and a reduced state soft input soft output (SISO) a posteriori probability algorithm is proposed. Based on the reduced state sequence detection (RSSD), it has the more general form compared with other reduced state SISO algorithms. The proposed algorithm can greatly reduce the state number, thus leads to the computation complexity reduction. It also minimizes the degradation in Euclidean distance with decision feedback in the reduced state trellis. Analysis and simulation results show that the performance degradation is little with proper reduction scheme.

  16. Development of the breeding blanket and shield model for the fusion power reactors system SYCOMORE

    Energy Technology Data Exchange (ETDEWEB)

    Li-Puma, Antonella, E-mail: antonella.lipuma@cea.fr [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Jaboulay, Jean-Charles, E-mail: Jean-Charles.jaboulay@cea.fr [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Martin, Brunella, E-mail: brunella.martin@gmail.com [Incka, 19-21 Rue du 8 mai 1945, F-94110 Arcueil (France)

    2014-10-15

    SYCOMORE, a fusion reactor system code based on a modular approach is under development at CEA. Within this framework, this paper describes the relevant sub-modules which have been implemented to model the main outputs of the breeding blanket and shield block of the system code: tritium breeding ratio, peak energy deposition in toroidal field coils, reactor layout and power deposition, blanket pressure drops and materials inventory. Blanket and shield requirements are calculated by several sub-modules: the blanket assembly and layout sub-module, the neutronic sub-module, the blanket design sub-module (thermal hydraulic and thermo-mechanic pre-design tool). A power flow module has also been developed which is directly linked to the blanket thermo-dynamic performances, which is not described in this paper. For the blanket assembly and layout and the blanket module design sub-modules, explicit analytic models have been developed and implemented; for the neutronic sub-module neural networks that replicate the results of appropriate simplified 1D and 2D neutronic simulations have been built. Presently, relevant model for the Helium Cooled Lithium Lead is available. Sub-modules have been built in a way that they can run separately or coupled into the breeding blanket and shield module in order to be integrated in SYCOMORE. In the paper, the objective and main input/output parameters of each sub-module are reported and relevant models discussed. The application to previous studied reactor models (PPCS model AB, DEMO-HCLL 2006–2007 studies) is also presented.

  17. Development of the water cooled lithium lead blanket for DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Aubert, J., E-mail: julien.aubert@cea.fr [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France); Aiello, G.; Jonquères, N. [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France); Li Puma, A. [CEA-Saclay, DEN/DANS/DM2S/SERMA/LPEC, 91191 Gif Sur Yvette Cedex (France); Morin, A.; Rampal, G. [CEA-Saclay, DEN/DANS/DM2S/SEMT/BCCR, 91191 Gif Sur Yvette Cedex (France)

    2014-10-15

    Highlights: • The WCLL blanket design has been modified to adapt it to the 2012 EFDA DEMO specifications. • Preliminary CAD design of the equatorial outboard module of the WCLL blanket has been developed for DEMO. • Finite elements analyses have been carried out in order to assess the module thermal behavior in the straight part of the module. - Abstract: The water cooled lithium lead (WCLL) blanket, based on near-future technology requiring small extrapolation from present-day knowledge both on physical and technological aspect, is one of the breeding blanket concepts considered as possible candidates for the EU DEMOnstration power plant. In 2012, the EFDA agency issued new specifications for DEMO: this paper describes the work performed to adapt the WCLL blanket design to those specifications. Relatively small modules with straight surfaces are attached to a common Back Supporting Structure housing feeding pipes. Each module features reduced activation ferritic-martensitic steel as structural material, liquid Lithium-Lead as breeder, neutron multiplier and carrier. Water at typical Pressurized Water Reactors (PWR) conditions is chosen as coolant. A preliminary design of the equatorial outboard module has been achieved. Finite elements analyses have been carried out in order to assess the module thermal behavior. Two First Wall (FW) concepts have been proposed, one favoring the thermal efficiency, the other favoring the manufacturability. The Breeding Zone has been designed with C-shaped Double-Walled Tubes in order to minimize the Water/Pb-15.7Li interaction likelihood. The priorities for further development of the WCLL blanket concept are identified in the paper.

  18. Conceptual design and analysis of the helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Hongli, E-mail: hlchen1@ustc.edu.cn; Li, Min; Lv, Zhongliang; Zhou, Guangming; Liu, Qianwen; Wang, Shuai; Wang, Xiaoliang; Zheng, Jie; Ye, Minyou

    2015-10-15

    Highlights: • A helium cooled solid blanket was proposed as a candidate blanket concept for CFETR. • Material selection, basic structure and gas flow scheme of the blanket were introduced. • A series of performance analyses for the blanket were summarized. - Abstract: To bridge the gap between ITER and DEMO and to realize the fusion energy in China, a fusion device Chinese Fusion Engineering Test Reactor (CFETR) was proposed and is being designed mainly to demonstrate 50–200 MW fusion power, 30–50% duty time factor, tritium self-sustained. Because of the high demand of tritium production and the realistic engineering consideration, the design of tritium breeding blanket for CFETR is a challenging work and getting special attention. As a blanket candidate, a helium cooled solid breeder blanket has been designed with the emphasis on conservative design and realistic blanket technology. This paper introduces the basic blanket scheme, including the material selection, structural design, cooling scheme and purge gas flow path. In addition, some results of neutronics, thermal-hydraulic and stress analysis are presented.

  19. Iterating skeletons

    DEFF Research Database (Denmark)

    Dieterle, Mischa; Horstmeyer, Thomas; Berthold, Jost;

    2012-01-01

    block inside a bigger structure. In this work, we present a general framework for skeleton iteration and discuss requirements and variations of iteration control and iteration body. Skeleton iteration is expressed by synchronising a parallel iteration body skeleton with a (likewise parallel) state...

  20. Tailorable Advanced Blanket Insulation (TABI)

    Science.gov (United States)

    Sawko, Paul M.; Goldstein, Howard E.

    1987-01-01

    Single layer and multilayer insulating blankets for high-temperature service fabricated without sewing. TABI woven fabric made of aluminoborosilicate. Triangular-cross-section flutes of core filled with silica batting. Flexible blanket formed into curved shapes, providing high-temperature and high-heat-flux insulation.

  1. Structural Design and Force Analysis for Equatorial Diagnostic Port Plug of ITER

    Institute of Scientific and Technical Information of China (English)

    YU Dongman; YAO Damao; XIE Han; BAO Liman; XU Tiejun; LIU Sumei

    2009-01-01

    The structural design of equatorial diagnostic port plug for international thermonuclear experimental reactor(ITER)was discussed in this paper.The induced electromagnetic (EM) loads of port plug and blanket shielding module(BSM)were presented.The finite element analysis (FEA) and numerical calculation of the system were performed under the maximum loads.According to the American Society of Mechanical Engineers(ASME)criteria,the EM stress is far below the admissible failure boundary,which indicates that the system can resist the cooperative action of EM loads and device deadweight.The analysis could provide technical data and references for further analysis and optimized design of the system.

  2. Progress of the ECRH Upper Launcher design for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Strauss, D., E-mail: dirk.strauss@kit.edu [Karlsruhe Institute of Technology, Assoc. KIT-EURATOM, D-76021 Karlsruhe (Germany); Aiello, G. [Karlsruhe Institute of Technology, Assoc. KIT-EURATOM, D-76021 Karlsruhe (Germany); Bruschi, A. [Istituto di Fisica del Plasma CNR, Euratom Association, 20125 Milano (Italy); Chavan, R. [Centre de Recherches en Physique des Plasmas, CRPP–EPFL, CH-1015 Lausanne (Switzerland); Farina, D.; Figini, L. [Istituto di Fisica del Plasma CNR, Euratom Association, 20125 Milano (Italy); Gagliardi, M.; Garcia, V. [Fusion for Energy, Barcelona (Spain); Goodman, T.P. [Centre de Recherches en Physique des Plasmas, CRPP–EPFL, CH-1015 Lausanne (Switzerland); Grossetti, G. [Karlsruhe Institute of Technology, Assoc. KIT-EURATOM, D-76021 Karlsruhe (Germany); Heemskerk, C. [Heemskerk Innovative Technology, Merelhof 2, 2172 HZ Sassenheim (Netherlands); Henderson, M.A. [ITER Organization, 13108 Saint-Paul-lez-Durance (France); Kasparek, W. [Institut für Plasmaforschung, IPF, D-70569 Stuttgart (Germany); Krause, A.; Landis, J.-D. [Centre de Recherches en Physique des Plasmas, CRPP–EPFL, CH-1015 Lausanne (Switzerland); Meier, A. [Karlsruhe Institute of Technology, Assoc. KIT-EURATOM, D-76021 Karlsruhe (Germany); Moro, A.; Platania, P. [Istituto di Fisica del Plasma CNR, Euratom Association, 20125 Milano (Italy); Plaum, B. [Institut für Plasmaforschung, IPF, D-70569 Stuttgart (Germany); Poli, E. [Max-Planck-IPP, Euratom Association, D-85748 Garching (Germany); and others

    2014-10-15

    The design of the ITER ECRH system provides 20 MW millimeter wave power for central plasma heating and MHD stabilization. The system consists of an array of 24 gyrotrons with power supplies coupled to a set of transmission lines guiding the beams to the four upper and the equatorial launcher. The front steering upper launcher design described herein has passed successfully the preliminary design review, and it is presently in the final design stage. The launcher consists of a millimeter wave system and steering mechanism with neutron shielding integrated into an upper port plug with the plasma facing blanket shield module (in-vessel) and a set of ex-vessel waveguides connecting the launcher to the transmission lines. Part of the transmission lines are the ultra-low loss CVD torus diamond windows and a shutter valve, a miter bend section and the feedthroughs integrated in the plug closure plate. These components are connected by corrugated waveguides and form together the first confinement system (FCS). In-vessel, the millimeter-wave system includes a quasi-optical beam propagation system including four mirror sets and a front steering mirror. The millimeter wave system is integrated into a specifically optimized upper port plug providing structural stability to withstand plasma disruption forces and the high heat load from the plasma side with a dedicated blanket shield module. A recent update in the ITER interface definition has resulted in the recession of the upper port plug first wall panels, which is now integrated into the design. Apart from the millimeter wave system the upper port plug houses also a set of shield blocks which provide neutron shielding. An overview of the actual ITER ECRH Upper Launcher is given together with some highlights of the design.

  3. A fail–safe and cost effective fabrication route for blanket First Walls

    Energy Technology Data Exchange (ETDEWEB)

    Commin, L., E-mail: lorelei.commin@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials (IAM-AWP), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Rieth, M.; Dafferner, B.; Zimmermann, H.; Bolich, D.; Baumgärtner, S.; Ziegler, R. [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials (IAM-AWP), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Dichiser, S.; Fabry, T.; Fischer, S.; Hildebrand, W.; Palussek, O.; Ritz, H.; Sponda, A. [Karlsruhe Institute of Technology (KIT), Technische Infrastruktur und Dienste (TID), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2013-11-15

    Helium Cooled Lithium Lead and Helium Cooled Pebble Bed concepts have been selected as European Test Blanket Modules (TBM) for ITER. The TBM fabrication will need the assembly of six Reduced Activation Ferritic Martensitic steel sub-components, namely First Wall, Caps, Stiffening Grid, Breeding Units, Back Plates/Manifolds, and Attachment system. The fabrication of the First Wall requires the production of cooling channels inside 30 mm thick bended plates. For this specific component, the main issues consist of the lack of accessibility of some areas to join, the process tolerances, the dimensional stability and the resulting assembly mechanical properties. Several fabrication routes have been already investigated, which involve diffusion welding and fusion welding (electron beam, laser beam, hybrid MIG/laser). In this study, an alternative processing method was developed, based on Hot Isostatic Pressing of inner pipes within two half-shells. This method presents some major advantages over the existing ones, in particular its inherent fail–safe design due to the application of the double containment principle, the solely use of cost effective standard fabrication processes and the resulting component dimensional stability. A four channel mock-up was fabricated and analyzed to validate the fabrication procedure. The joint quality was assessed using microstructural characterization and Charpy tests. The results confirm the predicted perfect weld lines as well as the preservation of the mechanical properties. Therefore, the presented fabrication procedure is very appropriate for the fabrication of First Walls for fusion reactor blankets.

  4. Radiation dose reduction sinogram affirmed iterative reconstruction and automatic tube voltage modulation(CARE kV) in abdominal CT

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Hyun Joo; Chung, Yong Eun; Lee, Young Han; Choi, Jin Young; Park, Mi Suk; Kim, Myeong Jin; Kim, Ki Whang [Dept. of Radiology, Research Institute of Radiological Science, Severance Hospital, Yonsei University College of Medicine, Seoul (Korea, Republic of)

    2013-12-15

    To evaluate the feasibility of sinogram-affirmed iterative reconstruction (SAFIRE) and automated kV modulation (CARE kV) in reducing radiation dose without increasing image noise for abdominal CT examination. This retrospective study included 77 patients who received CT imaging with an application of CARE kV with or without SAFIRE and who had comparable previous CT images obtained without CARE kV or SAFIRE, using the standard dose (i.e., reference mAs of 240) on an identical CT scanner and reconstructed with filtered back projection (FBP) within 1 year. Patients were divided into two groups: group A (33 patients, CT scanned with CARE kV); and group B (44 patients, scanned after reducing the reference mAs from 240 to 170 and applying both CARE kV and SAFIRE). CT number, image noise for four organs and radiation dose were compared among the two groups. Image noise increased after CARE kV application (p < 0.001) and significantly decreased as SAFIRE strength increased (p < 0.001). Image noise with reduced-mAs scan (170 mAs) in group B became similar to that of standard-dose FBP images after applying CARE kV and SAFIRE strengths of 3 or 4 when measured in the aorta, liver or muscle (p ≥ 0.108). Effective doses decreased by 19.4% and 41.3% for groups A and B, respectively (all, p < 0.001) after application of CARE kV with or without SAFIRE. Combining CARE kV, reduction of mAs from 240 to 170 mAs and noise reduction by applying SAFIRE strength 3 or 4 reduced the radiation dose by 41.3% without increasing image noise compared with the standard-dose FBP images.

  5. Blanket comparison and selection study. Volume II

    Energy Technology Data Exchange (ETDEWEB)

    1983-10-01

    This volume contains extensive data for the following chapters: (1) solid breeder tritium recovery, (2) solid breeder blanket designs, (3) alternate blanket concept screening, and (4) safety analysis. The following appendices are also included: (1) blanket design guidelines, (2) power conversion systems, (3) helium-cooled, vanadium alloy structure blanket design, (4) high wall loading study, and (5) molten salt safety studies. (MOW)

  6. Preliminary failure modes and effects analysis on Korean HCCR TBS to be tested in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Mu-Young, E-mail: myahn74@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of); Jin, Hyung Gon; Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Yi-Hyun; Lee, Youngmin [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Highlights: • Postulated initiating events are identified through failure modes and effects analysis on the current HCCR TBS design. • A set of postulated initiating events are selected for consideration of deterministic analysis. • Accident evolutions on the selected postualted initiating events are qualitatively described for deterministic analysis. - Abstract: Korean Helium cooled ceramic reflector (HCCR) Test blanket system (TBS), which comprises Test blanket module (TBM) and ancillary systems in various locations of ITER building, is operated at high temperature and pressure with decay heat. Therefore, safety is utmost concern in design process and it is required to demonstrate that the HCCR TBS is designed to comply with the safety requirements and guidelines of ITER. Due to complexity of the system with many interfaces with ITER, a systematic approach is necessary for safety analysis. This paper presents preliminary failure modes and effects analysis (FMEA) study performed for the HCCR TBS. FMEA is a systematic methodology in which failure modes for components in the system and their consequences are studied from the bottom-up. Over eighty failure modes have been investigated on the HCCR TBS. The failure modes that have similar consequences are grouped as postulated initiating events (PIEs) and total seven reference accident scenarios are derived from FMEA study for deterministic accident analysis. Failure modes not covered here due to evolving design of the HCCR TBS and uncertainty in maintenance procedures will be studied further in near future.

  7. Iterating skeletons

    DEFF Research Database (Denmark)

    Dieterle, Mischa; Horstmeyer, Thomas; Berthold, Jost;

    2012-01-01

    Skeleton-based programming is an area of increasing relevance with upcoming highly parallel hardware, since it substantially facilitates parallel programming and separates concerns. When parallel algorithms expressed by skeletons involve iterations – applying the same algorithm repeatedly...... block inside a bigger structure. In this work, we present a general framework for skeleton iteration and discuss requirements and variations of iteration control and iteration body. Skeleton iteration is expressed by synchronising a parallel iteration body skeleton with a (likewise parallel) state......-based iteration control, where both skeletons offer supportive type safety by dedicated types geared towards stream communication for the iteration. The skeleton iteration framework is implemented in the parallel Haskell dialect Eden. We use example applications to assess performance and overhead....

  8. DESIGN OF THE ITER IN-VESSEL COILS

    Energy Technology Data Exchange (ETDEWEB)

    Neumeyer, C; Bryant, L; Chrzanowski, J; Feder, R; Gomez, M; Heitzenroeder, P; Kalish, M; Lipski, A; Mardenfeld, M; Simmons, R; Titus, P; Zatz, I; Daly, E; Martin, A; Nakahira, M; Pillsbury, R; Feng, J; Bohm, T; Sawan, M; Stone, H; Griffiths, I

    2010-11-27

    The ITER project is considering the inclusion of two sets of in-vessel coils, one to mitigate the effect of Edge Localized Modes (ELMs) and another to provide vertical stabilization (VS). The in-vessel location (behind the blanket shield modules, mounted to the vacuum vessel inner wall) presents special challenges in terms of nuclear radiation (~3000 MGy) and temperature (100oC vessel during operations, 200oC during bakeout). Mineral insulated conductors are well suited to this environment but are not commercially available in the large cross section required. An R&D program is underway to demonstrate the production of mineral insulated (MgO or Spinel) hollow copper conductor with stainless steel jacketing needed for these coils. A preliminary design based on this conductor technology has been developed and is presented herein.

  9. In-Vessel Coil Material Failure Rate Estimates for ITER Design Use

    Energy Technology Data Exchange (ETDEWEB)

    L. C. Cadwallader

    2013-01-01

    The ITER international project design teams are working to produce an engineering design for construction of this large tokamak fusion experiment. One of the design issues is ensuring proper control of the fusion plasma. In-vessel magnet coils may be needed for plasma control, especially the control of edge localized modes (ELMs) and plasma vertical stabilization (VS). These coils will be lifetime components that reside inside the ITER vacuum vessel behind the blanket modules. As such, their reliability is an important design issue since access will be time consuming if any type of repair were necessary. The following chapters give the research results and estimates of failure rates for the coil conductor and jacket materials to be used for the in-vessel coils. Copper and CuCrZr conductors, and stainless steel and Inconel jackets are examined.

  10. Calculating the 3D magnetic field of ITER for European TBM studies

    CERN Document Server

    Äkäslompolo, Simppa; Bergmans, Thijs; Gagliardi, Mario; Galabert, Jose; Hirvijoki, Eero; Kurki-Suonio, Taina; Sipilä, Seppo; Snicker, Antti

    2015-01-01

    The magnetic perturbation due to the ferromagnetic test blanket modules (TBMs) may deteriorate fast ion confinement in ITER. This effect must be quantified by numerical studies in 3D. We have implemented a combined finite element method (FEM) -- Biot-Savart law integrator method (BSLIM) to calculate the ITER 3D magnetic field and vector potential in detail. Unavoidable geometry simplifications changed the mass of the TBMs and ferritic inserts (FIs) up to 26%. This has been compensated for by modifying the nonlinear ferromagnetic material properties accordingly. Despite the simplifications, the computation geometry and the calculated fields are highly detailed. The combination of careful FEM mesh design and using BSLIM enables the use of the fields unsmoothed for particle orbit-following simulations. The magnetic field was found to agree with earlier calculations and revealed finer details. The vector potential is intended to serve as input for plasma shielding calculations.

  11. Attachment system for helium-cooled blanket of RF DEMO fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Leshukov, A. E-mail: leshu@entek.ru; Blinov, Y.; Kovalenko, V.; Shatalov, G.; Strebkov, Y.; Strizhov, A

    2002-11-01

    The development of DEMO thermonuclear reactor is a part of Russian national program on the fusion process mastering. The DEMO-S (stationary thermonuclear reactor) should be the logic continuation of the ITER-type projects (pulse thermonuclear reactors) and the prototype for commercial power plants. DEMO reactor layout suggests to use the segmented blanket with mounting/dismounting procedure through the vacuum vessel vertical ports. Taking into account this layout the blanket attachment system has been developed and the present paper is devoted to this subject. The considered attachment system includes the lower and upper toroidal support assemblies which connect all the blanket segments in the enclosed ring. In it's turn the lower support assemblies attached to the vacuum vessel through the system of hinged support pillars. The heights of support pillars for inboard and outboard blankets are selected so that to indemnify the blanket massif thermal expansions in vertical and radial directions. The support pillars have been calculated on strength taking into account the electromagnetic loads from the plasma disruptions and blanket mass. The selection of high-strength chromium steel as a structural material for the support pillars could be considered as the results of strength analysis. The conclusions on the possibility to apply this attachment system for fusion reactor blanket and the critical issues are contained in this paper too.

  12. Characterization methodology for Difficult To Measure nuclides in the Type B rad waste from the ITER

    Energy Technology Data Exchange (ETDEWEB)

    Ji, Youngyong; Hong, Kwonpyo; Oh, Wanho; Kang, Munja; Na, Byungchan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    In general, it is not possible to directly detect beta rays from the rad waste in the field measurement due to their extremely low penetration through the materials. Only lab-scale measurements with proper shield and detecting system are available for the nondestructive assay. However, the disposal sites in many countries require the determination of inventories of the difficult to-measure (DTM) nuclides in the waste before their acceptance for disposal. Many sites that generate rad wastes thus are adapting the indirect method to characterize the DTM nuclides in the rad waste to be disposed. The rad waste from the operation of an international thermonuclear experimental reactor (ITER) will be sent to the hot cell building (HCB) after packing it to the basket and they are then treated into the disposal form as well as characterized through the nondestructive assay. The rad waste properties from the ITER are that high density material such as a steel, a copper, and a tungsten accounts for the main substance and many nuclides due to the neutron irradiation including the DTM nuclides exists in that waste. Therefore, the ITER is also facing with the problem for the characterization of DTM nuclides. The scaling factor for the radiological relationship between the gamma and the beta nuclides is one of the indirect measurements to characterize the DTM nuclides in the waste. The methodology of the scaling factor to apply this method to the characterization the Type B rad waste from the ITER are presented in this paper. There are several types of the in-vessel components (IVCs) in a Tokamak which will be activated by neutron and they will be divided into different types of the rad waste such as the divertor cassette, blanket module, and port plugs. In this paper, the characterization of DTM nuclides will be focused on the rad waste from a blanket module out of IVCs.

  13. Predictive capabilities, analysis and experiments for Fusion Nuclear Technology, and ITER R D

    Energy Technology Data Exchange (ETDEWEB)

    1991-01-01

    This report discusses the following topics on ITER research and development: trituim modeling; liquid metal blanket modeling; free surface liquid metal studies; and thermal conductance and thermal control experiments and modeling. (LIP)

  14. 迭代系统中连续相位调制技术的研究%Research of Continuous Phase Modulation in Iterative System

    Institute of Scientific and Technical Information of China (English)

    侯长波; 赵旦峰; 张霖; 薛睿

    2011-01-01

    Continuous Phase Modulation(CPM) itself has the characteristics of high power and spectral efficiency in wireless communication makes it especially widely used in military communications. According to Rimoldi decomposition model of CPM, which can be decomposed into a continuous phase encoder (CPE) and memoryless modulator (MM) in combination, based on the CPE’s memory and recursive features, combined with error-correcting code and the random interleaver, designing iterative detection receiver based on soft input soft output (SISO) algorithm can effectively improve the bit error rate performance and bandwidth efficiency. This paper gives continuous phase modulation algorithm model in iterative system, and simulates the BER performance of iterative detection receiver which is composed of CPM and LDPC codes. This paper can provide a reference for the design of iterative detection receiver.%连续相位调制(CPM)本身所具有的高功率和频谱利用率特性使得它在无线通信领域尤其是军事通信领域中应用广泛。根据CPM的Rimoldi分解模型,其可分解为连续相位编码(CPE)和无记忆调制(MM)的组合,基于CPE的记忆与递归特性,结合纠错码及随机交织器,设计基于软输入软输出(SISO)算法的迭代检测接收机,可有效的提高系统的误比特率性能和频带利用率。本文给出了迭代系统中连续相位调制的算法模型,并仿真了LDPC码与CPM迭代系统的误码率性能,为其他迭代检测接收机的设计提供了参考。

  15. APT Blanket Safety Analysis: Preliminary Analyses of Downflow Through a Lateral Row 1 Blanket Model Under Near RHR Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    To address a concern about a potential maldistribution of coolant flow through an APT blanket module under low flow near RHR conditions, a scoping study of downflow mixed convection in parallel channels was conducted. Buoyancy will adversely effect the flow distribution in module bins with downflow and non-uniform power distributions. This study consists of two parts: a simple analytical model of flow in a two channel network, and a lumped eleven channel FLOWTRAN-TF model of a front lateral Row-1 blanket module bin. Results from both models indicate that the concern about coolant flow in a vertical model being diverted away from high power regions by buoyancy is warranted. The FLOWTRAN-TF model predicted upflow (i.e., a flow reversal) through several of the high power channels, under some low flow conditions. The transition from the regime with downflow in all channels to a regime with upflow in some channels was abrupt.

  16. Preliminary Design of a Helium-Cooled Ceramic Breeder Blanket for CFETR Based on the BIT Concept

    Science.gov (United States)

    Ma, Xuebin; Liu, Songlin; Li, Jia; Pu, Yong; Chen, Xiangcun

    2014-04-01

    CFETR is the “ITER-like” China fusion engineering test reactor. The design of the breeding blanket is one of the key issues in achieving the required tritium breeding radio for the self-sufficiency of tritium as a fuel. As one option, a BIT (breeder insider tube) type helium cooled ceramic breeder blanket (HCCB) was designed. This paper presents the design of the BIT—HCCB blanket configuration inside a reactor and its structure, along with neutronics, thermo-hydraulics and thermal stress analyses. Such preliminary performance analyses indicate that the design satisfies the requirements and the material allowable limits.

  17. Development and trial manufacturing of 1/2-scale partial mock-up of blanket box structure for fusion experimental reactor

    Science.gov (United States)

    Hashimoto, Toshiyuki; Takatsu, Hideyuki; Sato, Satoshi

    1994-07-01

    Conceptual design of breeding blanket has been discussed during the CDA (Conceptual Design Activities) of ITER (International Thermonuclear Experimental Reactor). Structural concept of breeding blanket is based on box structure integrated with first wall and shield, which consists of three coolant manifolds for first wall, breeding and shield regions. The first wall must have cooling channels to remove surface heat flux and nuclear heating. The box structure includes plates to form the manifolds and stiffening ribs to withstand enormous electromagnetic load, coolant pressure and blanket internal (purge gas) pressure. A 1/2-scale partial model of the blanket box structure for the outboard side module near midplane is manufactured to estimate the fabrication technology, i.e. diffusion bonding by HIP (Hot Isostatic Pressing) and EBW (Electron Beam Welding) procedure. Fabrication accuracy is a key issue to manufacture first wall panel because bending deformation during HIP may not be small for a large size structure. Data on bending deformation during HIP was obtained by preliminary manufacturing of HIP elements. For the shield structure, it is necessary to reduce the welding strain and residual stress of the weldment to establish the fabrication procedure. Optimal shape of the parts forming the manifolds, welding locations and welding sequence have been investigated. In addition, preliminary EBW tests have been performed in order to select the EBW conditions, and fundamental data on built-up shield have been obtained. Especially, welding deformation by joining the first wall panel to the shield has been measured, and total deformation to build-up shield by EBW has been found to be smaller than 2 mm. Consequently, the feasibility of fabrication technologies has been successfully demonstrated for a 1m-scaled box structure including the first wall with cooling channels by means of HIP, EBW and TIG (Tungsten Inert Gas arc)-welding.

  18. Activation calculation analysis for the China 2×6 solid breeder test blanket module%中国2×6固态实验包层模块活化计算分析

    Institute of Scientific and Technical Information of China (English)

    韩静茹; 陈义学; 张国书; 曹启祥

    2011-01-01

    Based on the new design of China 2x6 helium cooled solid breeder (CH-HCSB) test blanket module (TBM), three-dimensional activation calculation analysis was performed by using the Monte Carlo particle transport code MCNP and European activation code FISPACT. The results show that, at the beginning moment after shutdown, the total activity is 1.78×1016Bq, and the total afterheat is 3.01kW. They are both dominated by the structural material CLF-1. Meanwhile, the dominant radioactivity nuclides and reaction pathways have been identified. The results will provide useful indications for materials selection and further optimization design of the TBM. On basis of the calculated contact dose rate, the activated materials can be reused with the remote handling recycling options. That is effective for preventing from the radiation environmental hazard.%基于中国氦冷固态实验包层模块(CH-HCSB-TBM)的新设计方案,采用蒙特卡罗粒子输运程序MCNP和欧洲活化计算程序FISPACT,对CH-HCSB-TBM进行了三维活化计算分析.计算结果表明,停堆初期TBM总的放射性活度、衰变余热分别为1.78× 1016Bq和3.01kW,主要受结构材料CLF-1影响.同时给出了影响TBM材料活化特性的主要核素及其生成途径,为TBM设计的材料选取和优化提供依据.根据计算的停堆剂量率可知,TBM中的活化材料都能采取远程操作实现再循环利用,可有效防止放射性环境危害问题.

  19. Crucial issues on liquid metal blanket design

    Energy Technology Data Exchange (ETDEWEB)

    Malang, S. (Kernforschungszentrum Karlsruhe (Germany)); Leroy, P. (CEA, CEN Saclay, 91 - Gif-sur-Yvette (France)); Casini, G.P. (CEC, Joint Research Centre (JRC), Ispra (Italy)); Mattas, R.F. (Argonne National Lab., IL (United States)); Strebkov, Yu. (Research and Development Inst. of Power Engineering, Moscow (USSR))

    1991-12-01

    Typical design concepts of liquid metal breeder blankets for power reactors are explained and characterized. The major problems of these concepts are described for both water-cooled blankets and self-cooled blankets. Three crucial issues of liquid metal breeder blankets are investigated. They are in the fields of magnetohydrodynamics, tritium control and safety. The influence of the magnetic field on liquid metal flow is of special interest for self-cooled blankets. The main problems in this field and the status of the related R and D work are described. Tritium permeation losses to the cooling water is a crucial issue for water-cooled blankets. Methods for its reduction are discussed. An inherent problem of all liquid breeder blankets is the potential release of activated products in the case of chemical reactions between the breeder material and water or reactive gases. The most important issues in this field are described. (orig.).

  20. APT {sup 3}He target/blanket. Topical report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-03-01

    The {sup 3}He target/blanket (T/B) preconceptual design for the 3/8-Goal facility is based on a 1000-MeV, 200-mA accelerator to produce a high-intensity proton beam that is expanded and then strikes one of two T/B modules. Each module consists of a centralized neutron source made of tungsten and lead, a proton beam backstop region made of zirconium and lead, and a moderator made of D{sub 2}O. Helium-3 gas is circulated through the neutron source region and the blanket to create tritium through neutron capture. The gas is continually processed to extract the tritium with an online separation process.

  1. Conceptual design of a water cooled breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Songlin, E-mail: slliu@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Pu, Yong; Cheng, Xiaoman [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Li, Jia; Peng, ChangHong [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China); Ma, Xuebing [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Chen, Lei [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China)

    2014-10-15

    Highlights: • We proposed a water cooled ceramic breeder blanket with superheated steam. • Superheated steam is generated at the first wall and the front part of breeder zone. • Superheated steam has negligible impact on neutron absorption by coolant in FW and improves TBR. • The superheated steam at higher temperature can improve thermal efficiency. - Abstract: China Fusion Engineering Test Reactor (CFETR) is an ITER-like superconducting tokamak reactor. Its major radius is 5.7 m, minor radius is 1.6 m and elongation ratio is 1.8. Its mission is to achieve 50–200 MW of fusion power, 30–50% of duty time factor, and tritium breeding ratio not less than 1.2 to ensure the self-sufficiency. As one of the breeding blanket candidates for CFETR, a water cooled breeder blanket with superheated steam is proposed and its conceptual design is being carried out. In this design, sub-cooling water at 265 °C under the pressure of 7 MPa is fed into cooling plates in breeding zone and is heated up to 285 °C with saturated steam generated, and then this steam is pre-superheated up to 310 °C in first wall (FW), final, the pre-superheated steam coming from several blankets is fed into the other one blanket to superheat again up to 517 °C. Due to low density of superheated steam, it has negligible impact on neutron absorption by coolant in FW so that the high energy neutrons entering into breeder zone moderated by water in cooling plate help enhance tritium breeding by {sup 6}Li(n,α)T reaction. Li{sub 2}TiO{sub 3} pebbles and Be{sub 12}Ti pebbles are chosen as tritium breeder and neutron multiplier respectively, because Li{sub 2}TiO{sub 3} and Be{sub 12}Ti are expected to have better chemical stability and compatibility with water in high temperature. However, Be{sub 12}Ti may lead to a reduction in tritium breeding ratio (TBR). Furthermore, a spot of sintered Be plate is used to improve neutron multiplying capacity in a multi-layer structure. As one alternative option

  2. Three-dimensional dual-flow fields analysis of the DFLL TBM for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Wang Weihua, E-mail: whwang@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); New Star Institute of Applied Technology, Hefei, Anhui 230031 (China); Li Jinling [New Star Institute of Applied Technology, Hefei, Anhui 230031 (China); Liu Songling [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Pei Xi; Huang Qunying [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China)

    2012-08-15

    This paper concerns the design calculations and performance evaluation of the Dual Function Lithium Lead Test Blanket Module (DFLL TBM) for ITER. Detailed three-dimensional dual-flow field calculations of helium gas and lithium lead (LiPb) have been performed for the DFLL TBM. The commercial Computational Fluid Dynamics (CFD) code FLUENT based finite volume method Navier-Stokes solver capable of solving conjugate flow and heat transfer between dual-flow field and structure is used. The CFD calculations are conducted directly in the CAD model using the CATIA code that allows preserving the geometrical details. The computational results show that the current TBM design is reasonable under the ITER normal condition. The detailed dual-flow fields, which include temperature, velocity, pressure and heat transfer of liquid LiPb and helium gas, are presented to optimize and improve the design of DFLL TBM system for ITER, and to supply more robust database and make a significant joint contribution to the future TBM testing in EAST and ITER.

  3. Thermo-mechanical study of high heat flux component mock-ups for ITER TBM

    Energy Technology Data Exchange (ETDEWEB)

    Bonelli, Flavia [Institute for Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (Germany); Dipartimento Energia, Politecnico di Torino (Italy); Boccaccini, Lorenzo Virgilio, E-mail: lorenzo.boccaccini@kit.edu [Institute for Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (Germany); Kunze, André; Maione, Ivan Alessio [Institute for Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (Germany); Savoldi, Laura; Zanino, Roberto [Dipartimento Energia, Politecnico di Torino (Italy)

    2015-10-15

    Highlights: • Infrared radiation heaters for test of plasma facing component available at KIT. • Numerical model developed and validated to check uniformity of heat flux. • Thermo-mechanical calculations performed on a mock-up of the HCPB TBM FW. • Assessment done of representativity of stress conditions for the ITER TBMs. - Abstract: Commercial infrared heaters have been proposed to be used in the HELOKA facility under construction at Karlsruhe Institute of Technology (KIT) to test a mock-up of the first wall (FW), called thermo-cycle mock-up (TCM) plate, under stress loading comparable to those experienced by the test blanket modules (TBMs) in ITER. Two related issues are analyzed in this paper, in relation to the ongoing European project aimed at the design of the two EU TBMs: (1) the possibility to reproduce, by means of those heaters, high heat flux loading conditions on the TCM plate similar to those expected on the ITER TBMs, and (2) the thermo-mechanical analysis of the TCM itself, in order to define a suitable choice of experimental parameters and mechanical constraints leading to a relevant stress condition. A suitable heater model is developed and validated against experimental data from an ad-hoc test campaign. A thermo-mechanical study of the TCM plate is presented, showing that the structure is able to withstand high thermal loads, even in the most constrained case, reaching stress levels comparable to the ITER TBM.

  4. Structural materials for ITER in-vessel component design

    Science.gov (United States)

    Kalinin, G.; Gauster, W.; Matera, R.; Tavassoli, A.-A. F.; Rowcliffe, A.; Fabritsiev, S.; Kawamura, H.

    1996-10-01

    The materials proposed for ITER in-vessel components have to exhibit adequate performance for the operating lifetime of the reactor or for specified replacement intervals. Estimates show that maximum irradiation dose to be up to 5-7 dpa (for 1 MWa/m 2 in the basic performance phase (BPP)) within a temperature range from 20 to 300°C. Austenitic SS 316LN-ITER Grade was defined as a reference option for the vacuum vessel, blanket, primary wall, pipe lines and divertor body. Conventional technologies and mill products are proposed for blanket, back plate and manifold manufacturing. HIPing is proposed as a reference manufacturing method for the primary wall and blanket and as an option for the divertor body. The existing data show that mechanical properties of HIPed SS are no worse than those of forged 316LN SS. Irradiation will result in property changes. Minimum ductility has been observed after irradiation in an approximate temperature range between 250 and 350°C, for doses of 5-10 dpa. In spite of radiation-induced changes in tensile deformation behavior, the fracture remains ductile. Irradiation assisted corrosion cracking is a concern for high doses of irradiation and at high temperatures. Re-welding is one of the critical issues because of the need to replace failed components. It is also being considered for the replacement of shielding blanket modules by breeding modules after the BPP. Estimates of radiation damage at the locations for re-welding show that the dose will not exceed 0.05 dpa (with He generation of 1 appm) for the manifold and 0.01 dpa (with He generation 0.1 appm) for the back plate for the BPP of ITER operation. Existing experimental data show that these levels will not result in property changes for SS; however, neutron irradiation and He generation promote crack formation in the heat affected zone during welding. Cu based alloys, DS-Cu (Glidcop A125) and PHCu CuCrZr bronze) are proposed as a structural materials for high heat flux

  5. Geometric feasibility of flexible cask transportation system for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Lima, P.; Ribeiro, M.I.; Aparicio, P. [Instituto Superior Tecnico-Instituto de Sistemas e Robotica, Lisboa (Portugal)

    1998-07-01

    One of the remote operations that has to be carried out in the International Thermonuclear Experimental Reactor (ITER) is the transportation of sealed casks between the various ports of the Tokamak Building (TB) and the Hot Cell Building (HCB). The casks may contain different in-vessel components (e.g. blanket modules, divertors) and are designed for a maximum load of about 80 ton. To improve the safety and flexibility of ITER Remote Handling (RH) transport vehicles, the cask is not motorized by itself, but instead, a motorized platform carrying the cask was proposed. This paper addresses the geometric feasibility of the flexible cask transportation system, taking into account the vehicle kinematics. The feasibility issues studied include planning smooth paths to increase safety, the discussion of building constraints by the evaluation of the vehicle spanned areas when following a planned path, and the analysis of the clearance required to remove the platform from underneath the cask at different possible failure locations. Simulation results are presented for the recommended trajectory, the spanned area and the rescue manoeuvres at critical locations along the path. (authors)

  6. Remote handling assessment of attachment concepts for DEMO blanket segments

    Energy Technology Data Exchange (ETDEWEB)

    Iglesias, Daniel, E-mail: daniel.iglesias@ccfe.ac.uk [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Bastow, Roger; Cooper, Dave; Crowe, Robert; Middleton-Gear, Dave [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Sibois, Romain [VTT, Technical Research Centre of Finland, Industrial Systems, ROViR, Tampere (Finland); Carloni, Dario [Institute of Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (KIT) (Germany); Vizvary, Zsolt; Crofts, Oliver [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Harman, Jon [EFDA Close Support Unit Garching, Boltzmannstaße 2, D-85748 Garching bei München (Germany); Loving, Antony [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)

    2015-10-15

    Highlights: • Challenges are identified for the remote handling of blanket segments’ attachments. • Two attachment design approaches are assessed for remote handling (RH) feasibility. • An alternative is proposed, which potentially simplifies and speeds-up RH operations. • Up to three different assemblies are proposed for the remote handling of the attachments. • Proposed integrated design of upper port is compatible with the attachment systems. - Abstract: The replacement strategy of the massive Multi-Module Blanket Segments (MMS) is a key driver in the design of several DEMO systems. These include the blankets themselves, the vacuum vessel (VV) and its ports and the Remote Maintenance System (RMS). Common challenges to any blanket attachment system have been identified, such as the need for applying a preload to the MMS manifold, the effects of the decay heat and several uncertainties related to permanent deformations when removing the blanket segments after service. The WP12 kinematics of the MMS in-vessel transportation was adapted to the requirements of each of the supports during 2013 and 2014 design activities. The RM equipment envisaged for handling attachments and earth connections may be composed of up to three different assemblies. An In-Vessel Mover at the divertor level handles the lower support and earth bonding, and could stabilize the MMS during transportation. A Shield Plug crane with a 6 DoF manipulator operates the upper attachment and earth straps. And a Vertical Maintenance Crane is responsible for the in-vessel MMS transportation and can handle the removable upper support pins. A final proposal is presented which can potentially reduce the number of required systems, at the same time that speeds-up the RMS global operations.

  7. Fusion reactor blanket/shield design study

    Energy Technology Data Exchange (ETDEWEB)

    Smith, D.L.; Clemmer, R.G.; Harkness, S.D.

    1979-07-01

    A joint study of tokamak reactor first-wall/blanket/shield technology was conducted by Argonne National Laboratory (ANL) and McDonnell Douglas Astronautics Company (MDAC). The objectives of this program were the identification of key technological limitations for various tritium-breeding-blanket design concepts, establishment of a basis for assessment and comparison of the design features of each concept, and development of optimized blanket designs. The approach used involved a review of previously proposed blanket designs, analysis of critical technological problems and design features associated with each of the blanket concepts, and a detailed evaluation of the most tractable design concepts. Tritium-breeding-blanket concepts were evaluated according to the proposed coolant. The ANL effort concentrated on evaluation of lithium- and water-cooled blanket designs while the MDAC effort focused on helium- and molten salt-cooled designs. A joint effort was undertaken to provide a consistent set of materials property data used for analysis of all blanket concepts. Generalized nuclear analysis of the tritium breeding performance, an analysis of tritium breeding requirements, and a first-wall stress analysis were conducted as part of the study. The impact of coolant selection on the mechanical design of a tokamak reactor was evaluated. Reference blanket designs utilizing the four candidate coolants are presented.

  8. Packed fluidized bed blanket for fusion reactor

    Science.gov (United States)

    Chi, John W. H.

    1984-01-01

    A packed fluidized bed blanket for a fusion reactor providing for efficient radiation absorption for energy recovery, efficient neutron absorption for nuclear transformations, ease of blanket removal, processing and replacement, and on-line fueling/refueling. The blanket of the reactor contains a bed of stationary particles during reactor operation, cooled by a radial flow of coolant. During fueling/refueling, an axial flow is introduced into the bed in stages at various axial locations to fluidize the bed. When desired, the fluidization flow can be used to remove particles from the blanket.

  9. Lithium as a blanket coolant

    Energy Technology Data Exchange (ETDEWEB)

    Wells, W.M.

    1977-01-01

    Recent re-assessment of tokamak reactors which move towards smaller size and lower required field strength (higher beta)/sup 2/ change the picture as regards the magnitude of MHD effects on flow resistance for lithium coolant. Perhaps the most important consequence of this as regards use of this coolant is that of clear acceptability of such effects when the flow is predominantly transverse to the magnetic field. This permits defining a blanket that consists entirely of round tubes containing the circulated lithium with voids between the tubes. Required thermal-hydraulic calculations are then on bases which are well established, especially in view of recent results dealing with perturbations of ducts and magnetic fields. Mitigation of MHD effects is feasible through tapering of tube wall thickness or use of insulated layers, but their use was not mandatory for the assumed conditions. Blanket configurations utilizing flowing lithium in round tubes immersed in static lithium may be suitable, but calculational methods do not now exist for this situation. Use of boiling potassium or cesium appears to be prohibitive in terms of vapor flow area when temperature levels are consistent with stainless steel. Liquid sodium, in addition to not being a breeding material, requires higher velocity than lithium for the same heat removal.

  10. Classification Using Markov Blanket for Feature Selection

    DEFF Research Database (Denmark)

    Zeng, Yifeng; Luo, Jian

    2009-01-01

    Selecting relevant features is in demand when a large data set is of interest in a classification task. It produces a tractable number of features that are sufficient and possibly improve the classification performance. This paper studies a statistical method of Markov blanket induction algorithm...... induction as a feature selection method. In addition, we point out an important assumption behind the Markov blanket induction algorithm and show its effect on the classification performance....... for filtering features and then applies a classifier using the Markov blanket predictors. The Markov blanket contains a minimal subset of relevant features that yields optimal classification performance. We experimentally demonstrate the improved performance of several classifiers using a Markov blanket...

  11. Active beam spectroscopy for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Hellermann, M.G. von, E-mail: mgvh@jet.u [FOM Institute Rijnhuizen, Euratom Association, 3430BE Nieuwegein (Netherlands); Barnsley, R. [ITER Organization, 13108 St.-Paul-Lez-Durance, Cadarache (France); Biel, W. [Institut fuer Energieforschung, Plasmaphysik, Forschungszentrum Juelich, Euratom Association, 52425 Juelich (Germany); Delabie, E. [FOM Institute Rijnhuizen, Euratom Association, 3430BE Nieuwegein (Netherlands); Hawkes, N. [Culham Centre for Fusion Energy, Euratom Association, Culham OX14 3DB (United Kingdom); Jaspers, R. [FOM Institute Rijnhuizen, Euratom Association, 3430BE Nieuwegein (Netherlands); Johnson, D. [Princeton Plasma Physics Laboratory, Princeton, NJ-08548 (United States); Klinkhamer, F. [TNO Science and Industry, Stieltjesweg 1, 2628CK Delft (Netherlands); Lischtschenko, O. [FOM Institute Rijnhuizen, Euratom Association, 3430BE Nieuwegein (Netherlands); Marchuk, O. [Institut fuer Energieforschung, Plasmaphysik, Forschungszentrum Juelich, Euratom Association, 52425 Juelich (Germany); Schunke, B. [ITER Organization, 13108 St.-Paul-Lez-Durance, Cadarache (France); Singh, M.J. [Institute for Plasma Research, Bhat, Gandhinagar, Gurajat 384828 (India); Snijders, B. [TNO Science and Industry, Stieltjesweg 1, 2628CK Delft (Netherlands); Summers, H.P. [Culham Centre for Fusion Energy, Euratom Association, Culham OX14 3DB (United Kingdom); Thomas, D. [ITER Organization, 13108 St.-Paul-Lez-Durance, Cadarache (France); Tugarinov, S. [TRINITI Troitsk, Moscow Region 142092 (Russian Federation); Vasu, P. [Institute for Plasma Research, Bhat, Gandhinagar, Gurajat 384828 (India)

    2010-11-11

    Since the first feasibility studies of active beam spectroscopy on ITER in 1995 the proposed diagnostic has developed into a well advanced and mature system. Substantial progress has been achieved on the physics side including comprehensive performance studies based on an advanced predictive code, which simulates active and passive features of the expected spectral ranges. The simulation has enabled detailed specifications for an optimized instrumentation and has helped to specify suitable diagnostic neutral beam parameters. Four ITER partners share presently the task of developing a suite of ITER active beam diagnostics, which make use of the two 0.5 MeV/amu 18 MW heating neutral beams and a dedicated 0.1 MeV/amu, 3.6 MW diagnostic neutral beam. The IN ITER team is responsible for the DNB development and also for beam physics related aspects of the diagnostic. The RF will be responsible for edge CXRS system covering the outer region of the plasma (1>r/a>0.4) using an equatorial observation port, and the EU will develop the core CXRS system for the very core (0ITER environment. Additionally, an essential change of the orientation of the DNB injection angle and specification of suitable blanket aperture has been made to avoid trapped particle damage to the first wall.

  12. The ITER in-vessel system

    Energy Technology Data Exchange (ETDEWEB)

    Lousteau, D.C.

    1994-09-01

    The overall programmatic objective, as defined in the ITER Engineering Design Activities (EDA) Agreement, is to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes. The ITER EDA Phase, due to last until July 1998, will encompass the design of the device and its auxiliary systems and facilities, including the preparation of engineering drawings. The EDA also incorporates validating research and development (R&D) work, including the development and testing of key components. The purpose of this paper is to review the status of the design, as it has been developed so far, emphasizing the design and integration of those components contained within the vacuum vessel of the ITER device. The components included in the in-vessel systems are divertor and first wall; blanket and shield; plasma heating, fueling, and vacuum pumping equipment; and remote handling equipment.

  13. Development of Fabrication Procedure and Welding Performance Test for ITER HCCR TBM

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, J. S.; Lee, E. H.; Kim, S. K.; Lee, D. W.; Shin, K. I.; Jin, H. G. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, S. [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Korea has developed and plans to test a Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in the ITER. The HCCR TBM is composed of four sub-modules and a back manipulator (BM). Each sub-module is composed of a first wall (FW), a breeding box with a seven-layer breeding zone (BZ), and side walls (SW) with a cooling path. The front surface of the sub-module is 231 mm in width and 835 mm in height. The test specimens of the ARAA were prepared to test the weldability for tungsten inert gas (TIG) welding and electron beam (EB) welding. To establish and optimize the welding procedure in an EB weld from ARRA material, the variation in the bead width and penetration depth according to the welding current and welding speed were investigated. To verify the weldability and fabrication procedure for a complex structure such as the breeding zone, a small box with a cooling channel is being fabricated using the ARAA steel under development. A fabrication procedure of a HCCR TBM submodule was designed for a first wall, breeding zone including the side wall, and assembled sub-module by welding the first wall and breeding zone box. A plate for the test specimen to evaluate the welding characteristics was fabricated based on the optimized E-beam welding results.

  14. Multifractal Framework Based on Blanket Method

    Science.gov (United States)

    Paskaš, Milorad P.; Reljin, Irini S.; Reljin, Branimir D.

    2014-01-01

    This paper proposes two local multifractal measures motivated by blanket method for calculation of fractal dimension. They cover both fractal approaches familiar in image processing. The first two measures (proposed Methods 1 and 3) support model of image with embedded dimension three, while the other supports model of image embedded in space of dimension three (proposed Method 2). While the classical blanket method provides only one value for an image (fractal dimension) multifractal spectrum obtained by any of the proposed measures gives a whole range of dimensional values. This means that proposed multifractal blanket model generalizes classical (monofractal) blanket method and other versions of this monofractal approach implemented locally. Proposed measures are validated on Brodatz image database through texture classification. All proposed methods give similar classification results, while average computation time of Method 3 is substantially longer. PMID:24578664

  15. Exploratory Study of Blanket Liquid Curtain

    Institute of Scientific and Technical Information of China (English)

    HUGang; HUANGJinhua; FENGKaiming

    2003-01-01

    Blankets and other in-vessel components are easily damaged owing to their circumstance of high radiation and high heat. To protect them, first wall design should be considered. Owing to its high heat removal nd self-refreshing capability, liquid metal first wall has been seen as a potential first wall for a fusion reactor in the future. Blanketliquid curtain is actually a special liquid metal wall to protect blanket.

  16. Study on fission blanket fuel cycling of a fusion-fission hybrid energy generation system

    Science.gov (United States)

    Zhou, Z.; Yang, Y.; Xu, H.

    2011-10-01

    This paper presents a preliminary study on neutron physics characteristics of a light water cooled fission blanket for a new type subcritical fusion-fission hybrid reactor aiming at electric power generation with low technical limits of fission fuel. The major objective is to study the fission fuel cycling performance in the blanket, which may possess significant impacts on the feasibility of the new concept of fusion-fission hybrid reactor with a high energy gain (M) and tritium breeding ratio (TBR). The COUPLE2 code developed by the Institute of Nuclear and New Energy Technology of Tsinghua University is employed to simulate the neutronic behaviour in the blanket. COUPLE2 combines the particle transport code MCNPX with the fuel depletion code ORIGEN2. The code calculation results show that soft neutron spectrum can yield M > 20 while maintaining TBR >1.15 and the conversion ratio of fissile materials CR > 1 in a reasonably long refuelling cycle (>five years). The preliminary results also indicate that it is rather promising to design a high-performance light water cooled fission blanket of fusion-fission hybrid reactor for electric power generation by directly loading natural or depleted uranium if an ITER-scale tokamak fusion neutron source is achievable.

  17. Multiaxial fatigue analysis for IMIC of ITER upper ELM coil

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, S.W., E-mail: zhangsw@ipp.cn; Song, Y.T.; Wang, Z.W.; Lu, S.; Ji, X.; Du, S.S.; Liu, X.F.; Feng, C.L.; Yang, H.; Wang, S.K.; Luo, Z.R.

    2014-04-15

    Highlights: • The structural analysis provides the initial stresses. • Constant and variation of principal stress direction provide the design codes. • Two methods can be recommended for the ELM coils. • IMIC meets the fatigue criteria. - Abstract: Inconel Jacketed Mineral Insulated Conductor (IMIC) is a very important component of International Thermonuclear Experimental Reactor (ITER) Edge Localized Modes (ELM) coils, which are located between the vacuum vessel (VV) and blanket shield modules and subject to high radiation levels, high temperature and high magnetic field. These coils will experience thermal pulsed, cyclic electromagnetic (EM) load during operation. They are designed to sustain at 1.5e8 total stress cycles and shall have sufficient strength and excellent fatigue to transport and bear the high cyclic load. For IMIC, multiaxial fatigue analysis is used to evaluate failure. Two methods based on the alternating stress and mean stress in American Society of Mechanical Engineers (ASME) code provide the design codes for multiaxial fatigue evaluation: constant principal stress direction and variation of principal stress direction. Results show that using the two methods obtains basically the same equivalent alternating stress. Both of them can be recommended for the ELM coils and IMIC can meet the fatigue criteria.

  18. Multi-purpose deployer for ITER in-vessel maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chang-Hwan, E-mail: Chang-Hwan.CHOI@iter.org [ITER Organization, Route de Vinon-sur-Verdon, 13115 St Paul lez Durance (France); Tesini, Alessandro; Subramanian, Rajendran [ITER Organization, Route de Vinon-sur-Verdon, 13115 St Paul lez Durance (France); Rolfe, Alan; Mills, Simon; Scott, Robin; Froud, Tim; Haist, Bernhard; McCarron, Eddie [Oxford Technologies Ltd., 7 Nuffield Way, Abingdon, OXON (United Kingdom)

    2015-10-15

    Highlights: • ITER RH system called as the multi-purpose deployer (MPD) is introduced. • The MPD performs dust and tritium inventory control, in-service inspection. • The MPD performs leak localization, in-vessel diagnostics maintenance. • The MPD has nine degrees of freedom with a payload capacity up to 2 tons. - Abstract: The multi-purpose deployer (MPD) is a general purpose in-vessel remote handling (RH) system in the ITER RH system. The MPD provides the means for deployment and handling of in-vessel tools or components inside the vacuum vessel (VV) for dust and tritium inventory control, in-service inspection, leak localization, and in-vessel diagnostics. It also supports the operation of blanket first wall maintenance and neutral beam duct liner module maintenance operations. This paper describes the concept design of the MPD. The MPD is a cask based system, i.e. it stays in the hot cell building during the machine operation, and is deployed to the VV using the cask system for the in-vessel operations. The main part of the MPD is the articulated transporter which provides transportation and positioning of the in-vessel tools or components. The articulated transporter has nine degrees of freedom with a payload capacity up to 2 tons. The articulated transporter can cover the whole internal surface of the VV by switching between the four equatorial RH ports. Additionally it can use two non-RH equatorial ports to transfer large tools or components. A concept for in-cask tool exchange is developed which minimizes the cask transportation by allowing the MPD to stay in the VV during the tool exchange.

  19. The ITER EC H and CD Upper Launcher: Analysis of vertical Remote Handling applied to the BSM maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Grossetti, Giovanni, E-mail: giovanni.grossetti@kit.edu [Karlsruhe Institute of Technology, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Aiello, Gaetano [Karlsruhe Institute of Technology, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Heemskerk, Cock [Heemskerk Innovative Technology, Merelhof 2, 2172 HZ Sassenheim (Netherlands); Elzendoorn, Ben [FOM Institute DIFFER, P.O. Box 1207, 3430 BE Nieuwegein (Netherlands); Geßner, Robby [Karlsruhe Institute of Technology, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Koning, Jarich [Heemskerk Innovative Technology, Merelhof 2, 2172 HZ Sassenheim (Netherlands); Meier, Andreas [Karlsruhe Institute of Technology, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Ronden, Dennis [FOM Institute DIFFER, P.O. Box 1207, 3430 BE Nieuwegein (Netherlands); Späh, Peter; Scherer, Theo; Schreck, Sabine; Strauß, Dirk; Vaccaro, Alessandro [Karlsruhe Institute of Technology, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany)

    2013-10-15

    This paper deals with Remote Handling activities foreseen on the Blanket Shield Module, the plasma facing component of the ITER Electron Cyclotron Heating and Current Drive Upper Launcher. The maintenance configuration considered here is the Vertical Remote Handling, meaning gravity acting along the launcher radial axis. The plant, where the maintenance under consideration is occurring, is the Hot Cell Facility Work Cell. The study here reported has been carried out within the presently ongoing EFDA Goal Oriented Training program on Remote Handling (GOT-RH), which aims to support ITER activities. This document and its contents have to be considered as part of a more vast RAMI analysis to be developed within the GOT-RH, which aims to maximize the Electron Cyclotron Heating and Current Drive system availability. The Baseline CAD model of the Electron Cyclotron Heating and Current Drive Upper Launcher is currently in its preliminary design phase and does not provide enough details for developing a fully detailed maintenance strategy. Therefore, through a System Engineering approach, a set of assumptions was conceived on the launcher structure, as a basis for development of a Remote Handling strategy. Moreover, to compare different design solutions related to the possibility of integrating a quasi-optical component into the Blanket Shield Module, a Trade-Off was made, and its contents are shown here. The outcome of this System Engineering approach has been formalized into Task Definition Forms whose contents are reported here. The Remote Handling strategy presented in this work will be tested in the near future both through Virtual Reality simulations and through prototype experiments.

  20. Towards the development of technical specifications for the production of lithium-lead alloys for the ITER HCLL TMB

    Energy Technology Data Exchange (ETDEWEB)

    Barrado, Ana Isabel, E-mail: anaisabel.barrado@ciemat.es [CIEMAT. Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas (Spain); Conde, Estefania; Fernandez, Marta [CIEMAT. Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas (Spain); Gomez-Salazar, Jose Maria [UCM. Dpto. De Ciencia de Materiales e Ingenieria Metalurgica (Spain); Quejido, Alberto; Quinones, Javier [CIEMAT. Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas (Spain)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer Pb-Li alloy plays a key role in the new commercial fusion reactors functionality. Black-Right-Pointing-Pointer It is important to have a complete characterization to define their physicochemical properties. Black-Right-Pointing-Pointer Methodology developed is a key tool that allows performing quality control procedures. Black-Right-Pointing-Pointer Determine concentrations of major and trace elements, which can be found in Pb-Li alloy. - Abstract: The ITER and DEMO projects are developing new Test Blanket Modules (TBM), where the Pb-Li alloy plays a key role in the new commercial fusion reactors functionality. The Breeding Blanket (BB) has to perform several functions which are essential for the reactor operation. The HCLL TBM is one of the Breeding Blanket concepts to be tested in ITER. It is cooled by He and uses the eutectic liquid metal LLE (Lithium-Lead Eutectic) as breeder material (enriched at 90% in {sup 6}Li). Pb-Li eutectic alloy has no known uses outside of fusion technology, so the available databases of this material are currently incomplete. It is very important, within the material specifications, to have a complete characterization in order to define their chemical and physical properties, because any variation in the alloy composition has significant consequences in their behaviour, and therefore in their regenerative function inside the blanket. The chemical characterization methodology developed and presented in this paper (useful for both Pb-Li alloys as any Pb alloy) is a key tool that allows performing standard quality control procedures for base material and/or monitoring the alloy during the reactor operation. This report provides a procedure to perform a wide material chemical characterization, assessing the concentrations of major elements, as well as a review of trace level elements that can be found both in the eutectic alloy and in starting materials. In this determination plays an important

  1. The requirements for processing tritium recovered from liquid lithium blankets: The blanket interface

    Energy Technology Data Exchange (ETDEWEB)

    Clemmer, R.G.; Finn, P.A.; Greenwood, L.R.; Grimm, T.L.; Sze, D.K.; Bartlit, J.R.; Anderson, J.L.; Yoshida, H.; Naruse

    1988-03-01

    We have initiated a study to define a blanket processing mockup for Tritium Systems Test Assembly. Initial evaluation of the requirements of the blanket processing system have been started. The first step of the work is to define the condition of the gaseous tritium stream from the blanket tritium recovery system. This report summarizes this part of the work for one particular blanket concept, i.e., a self-cooled lithium blanket. The total gas throughput, the hydrogen to tritium ratio, the corrosive chemicals, and the radionuclides are defined. The key discoveries are: the throughput of the blanket gas stream (including the helium carrier gas) is about two orders of magnitude higher than the plasma exhaust stream;the protium to tritium ratio is about 1, the deuterium to tritium ratio is about 0.003;the corrosion chemicals are dominated by halides;the radionuclides are dominated by C-14, P-32, and S-35;their is high level of nitrogen contamination in the blanket stream. 77 refs., 6 figs., 13 tabs.

  2. Proceedings of the sixth international workshop on ceramic breeder blanket interactions

    Energy Technology Data Exchange (ETDEWEB)

    Noda, Kenji [ed.

    1998-03-01

    This report is the Proceedings of `the Sixth International Workshop on Ceramic Breeder Blanket Interactions` which was held as a workshop on ceramic breeders under Annex II of IEA Implementing Agreement on a Programme of Research and Development on Fusion Materials, and Japan-US Workshop 97FT4-01. This workshop was held in Mito city, Japan on October 22-24, 1997. About forty experts from EU, Japan, USA, and Chile attended the workshop. The scope of the workshop included the following: (1) fabrication and characterization of ceramic breeders, (2) properties data for ceramic breeders, (3) tritium release characteristics, (4) modeling of tritium behavior, (5) irradiation effects on performance behavior, (6) blanket design and R and D requirements, (7) hydrogen behavior in materials, and (8) blanket system technology and structural materials. In the workshop, information exchange was performed for fabrication technology of ceramic breeder pebbles in EU and Japan, data of various properties of Li{sub 2}TiO{sub 3}, tritium release behavior of Li{sub 2}TiO{sub 3} and Li{sub 2}ZrO{sub 3} including tritium diffusion, modeling of tritium release from Li{sub 2}ZrO{sub 3} in ITER condition, helium release behavior from Li{sub 2}O, results of tritium release irradiation tests of Li{sub 4}SiO{sub 4} pebbles in EXOTIC-7, R and D issues for ceramic breeders for ITER and DEMO blankets, etc. The 23 of the papers are indexed individually. (J.P.N.)

  3. Preliminary Analysis of Remote Handling Feasibility of the ITER Type B Radwaste Management System

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Kwon Pyo; Oh, Wan Ho; Hong, Dae Seok; Ji, Young Yong; Kim, Tae Kook; Ahn, Sang Bok; Ryu, Wu Seog; Cheon, Je Keun; Na, Byung Chan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Jung, Ki Jung [National Fusion Research Institute, Daejeon (Korea, Republic of); John Blight [ITER Organization, Paul lez Durance (France)

    2010-10-15

    ITER Type B radwaste, which is defined as intermediate level and long lived radioactive waste, is mainly resulted from the replacement of in-vessel components of Tokamak such as divertor targets and domes and first walls of blankets, etc. They are mostly metallic components activated by neutrons and contaminated by tritium. The Type B radwaste is to be stored in ITER site for the life time of ITER. For this purpose the Type B radwaste treatment and storage system (RWTS) is located in the level 2 basement of the ITER Hot Cell Building (HCB)

  4. Lightweight IMM PV Flexible Blanket Assembly

    Science.gov (United States)

    Spence, Brian

    2015-01-01

    Deployable Space Systems (DSS) has developed an inverted metamorphic multijunction (IMM) photovoltaic (PV) integrated modular blanket assembly (IMBA) that can be rolled or z-folded. This IMM PV IMBA technology enables a revolutionary flexible PV blanket assembly that provides high specific power, exceptional stowed packaging efficiency, and high-voltage operation capability. DSS's technology also accommodates standard third-generation triple junction (ZTJ) PV device technologies to provide significantly improved performance over the current state of the art. This SBIR project demonstrated prototype, flight-like IMM PV IMBA panel assemblies specifically developed, designed, and optimized for NASA's high-voltage solar array missions.

  5. 48 CFR 613.303 - Blanket purchase agreements (BPAs).

    Science.gov (United States)

    2010-10-01

    ... 48 Federal Acquisition Regulations System 4 2010-10-01 2010-10-01 false Blanket purchase agreements (BPAs). 613.303 Section 613.303 Federal Acquisition Regulations System DEPARTMENT OF STATE....303 Blanket purchase agreements (BPAs)....

  6. 48 CFR 1313.303 - Blanket Purchase Agreements (BPAs).

    Science.gov (United States)

    2010-10-01

    ... 48 Federal Acquisition Regulations System 5 2010-10-01 2010-10-01 false Blanket Purchase Agreements (BPAs). 1313.303 Section 1313.303 Federal Acquisition Regulations System DEPARTMENT OF COMMERCE....303 Blanket Purchase Agreements (BPAs)....

  7. 48 CFR 13.303 - Blanket purchase agreements (BPAs).

    Science.gov (United States)

    2010-10-01

    ... 48 Federal Acquisition Regulations System 1 2010-10-01 2010-10-01 false Blanket purchase agreements (BPAs). 13.303 Section 13.303 Federal Acquisition Regulations System FEDERAL ACQUISITION... Methods 13.303 Blanket purchase agreements (BPAs)....

  8. 48 CFR 313.303 - Blanket purchase agreements.

    Science.gov (United States)

    2010-10-01

    ... 48 Federal Acquisition Regulations System 4 2010-10-01 2010-10-01 false Blanket purchase agreements. 313.303 Section 313.303 Federal Acquisition Regulations System HEALTH AND HUMAN SERVICES....303 Blanket purchase agreements....

  9. A recursion identity for formal iterated logarithms and iterated exponentials

    CERN Document Server

    Robinson, Thomas J

    2010-01-01

    We prove a recursive identity involving formal iterated logarithms and formal iterated exponentials. These iterated logarithms and exponentials appear in a natural extension of the logarithmic formal calculus used in the study of logarithmic intertwining operators and logarithmic tensor category theory for modules for a vertex operator algebra. This extension has a variety of interesting arithmetic properties. We develop one such result here, the aforementioned recursive identity. We have applied this identity elsewhere to certain formal series expansions related to a general formal Taylor theorem and these series expansions in turn yield a sequence of combinatorial identities which have as special cases certain classical combinatorial identities involving (separately) the Stirling numbers of the first and second kinds.

  10. Review: BNL graphite blanket design concepts

    Energy Technology Data Exchange (ETDEWEB)

    Fillo, J.A.; Powell, J.R.

    1976-03-01

    A review of the Brookhaven National Laboratory (BNL) minimum activity graphite blanket designs is made. Three designs are identified and discussed in the context of an experimental power reactor (EPR) and commercial power reactor. Basically, the three designs employ a thick graphite screen (typically 30 cm or greater, depending on type as well as application-experimental power reactor or commercial reactor). Bremsstrahlung energy is deposited on the graphite surface and re-radiated away as thermal radiation. Fast neutrons are slowed down in the graphite, depositing most of their energy. This energy is then either radiated to a secondary blanket with coolant tubes, as in types A and B, or is removed by intermittent direct gas cooling (type C). In types A and B, radiation damage to the structural material of the coolant tubes in the secondary blanket is reduced by one or two orders of magnitude by the graphite screen, while in type C, the blanket is only cooled when the reactor is shut down, so that coolant cannot quench the plasma, whatever the degree of radiation damage.

  11. The climatic impact of supervolcanic ash blankets

    Energy Technology Data Exchange (ETDEWEB)

    Jones, Morgan T.; Sparks, R.S.J. [University of Bristol, Department of Earth Sciences, Bristol (United Kingdom); Valdes, Paul J. [University of Bristol, School of Geographical Sciences, Bristol (United Kingdom)

    2007-11-15

    Supervolcanoes are large caldera systems that can expel vast quantities of ash, volcanic gases in a single eruption, far larger than any recorded in recent history. These super-eruptions have been suggested as possible catalysts for long-term climate change and may be responsible for bottlenecks in human and animal populations. Here, we consider the previously neglected climatic effects of a continent-sized ash deposit with a high albedo and show that a decadal climate forcing is expected. We use a coupled atmosphere-ocean General Circulation Model (GCM) to simulate the effect of an ash blanket from Yellowstone volcano, USA, covering much of North America. Reflectivity measurements of dry volcanic ash show albedo values as high as snow, implying that the effects of an ash blanket would be severe. The modeling results indicate major disturbances to the climate, particularly to oscillatory patterns such as the El Nino Southern Oscillation (ENSO). Atmospheric disruptions would continue for decades after the eruption due to extended ash blanket longevity. The climatic response to an ash blanket is not significant enough to investigate a change to stadial periods at present day boundary conditions, though this is one of several impacts associated with a super-eruption which may induce long-term climatic change. (orig.)

  12. Fidget Blankets: A Sensory Stimulation Outreach Program.

    Science.gov (United States)

    Kroustos, Kelly Reilly; Trautwein, Heidi; Kerns, Rachel; Sobota, Kristen Finley

    2016-01-01

    Behavioral and Psychological Symptoms of Dementia (BPSD) include behaviors such as aberrant motor behavior, agitation, anxiety, apathy, delusions, depression, disinhibition, elation, hallucinations, irritability, and sleep or appetite changes. A student-led project to provide sensory stimulation in the form of "fidget blankets" developed into a community outreach program. The goal was to decrease the use of antipsychotics used for BPSD.

  13. A blanket design, apparatus, and fabrication techniques for the mass production of multilayer insulation blankets for the Superconducting Super Collider

    Energy Technology Data Exchange (ETDEWEB)

    Gonczy, J.D.; Boroski, W.N.; Niemann, R.C.; Otavka, J.G.; Ruschman, M.K.; Schoo, C.J.

    1989-09-01

    The multilayer insulation (MLI) system for the Superconducting Super Collider (SSC) consists of full cryostat length assemblies of aluminized polyester film fabricated in the form of blankets and installed as blankets to the 4.5K cold mass and the 20K and 80K thermal radiation shields. Approximately 40,000 MLI blankets will be required in the 10,000 cryogenic devices comprising the SSC accelerator. Each blanket is nearly 17 meters long and 1.8 meters wide. This paper reports the blanket design, an apparatus, and the fabrication method used to mass produce pre-fabricated MLI blankets. Incorporated in the blanket design are techniques which automate quality control during installation of the MLI blankets in the SSC cryostat. The apparatus and blanket fabrication method insure consistency in the mass produced blankets by providing positive control of the dimensional parameters which contribute to the thermal performance of the MLI blanket. By virtue of the fabrication process, the MLI blankets have inherent features of dimensional stability three-dimensional uniformity, controlled layer density, layer-to-layer registration, interlayer cleanliness, and interlayer material to accommodate thermal contraction differences. 11 refs., 6 figs., 1 tab.

  14. Molecule-surface interaction processes of relevance to gas blanket type fusion device divertor design

    Energy Technology Data Exchange (ETDEWEB)

    Snowdon, K.J. [Newcastle Univ. (United Kingdom). Dept. of Physics; Tawara, H.

    1997-01-01

    The mechanisms which may lead to the departure of molecular species from surfaces exposed to low energy (0.1-100 eV) particle or photon and electron irradiation are reviewed. Where possible, the charge and electronic state, angular, translational and internal energy distributions of the departing molecules are described and the physical origin of the nature of those distributions identified. The consequences, for the departing molecules, of certain material choices become apparent from such an analysis. Such information may help guide the choice of appropriate materials for plasma facing components of gas-blanket type divertors such as that recently proposed for the International Thermonuclear Experimental Reactor (ITER). (author). 71 refs.

  15. Embeddings of Iteration Trees

    OpenAIRE

    Mitchell, William

    1992-01-01

    This paper, dating from May 1991, contains preliminary (and unpublishable) notes on investigations about iteration trees. They will be of interest only to the specialist. In the first two sections I define notions of support and embeddings for tree iterations, proving for example that every tree iteration is a direct limit of finite tree iterations. This is a generalization to models with extenders of basic ideas of iterated ultrapowers using only ultrapowers. In the final section (which is m...

  16. Multi-Sensor Data Fusion Technologies for Blanket Jamming Localization

    Institute of Scientific and Technical Information of China (English)

    WANG Ju; WU Si-liang; ZENG Tao

    2005-01-01

    The localization of the blanket jamming is studied and a new method of solving the localization ambiguity is proposed. Radars only can acquire angle information without range information when encountering the blanket jamming. Netted radars could get position information of the blanket jamming by make use of radars' relative position and the angle information, when there is one blanket jamming. In the presence of error, the localization method and the accuracy analysis of one blanket jamming are given. However, if there are more than one blanket jamming, and the two blanket jamming and two radars are coplanar, the localization of jamming could be error due to localization ambiguity. To solve this confusion, the Kalman filter model is established for all intersections, and through the initiation and association algorithm of multi-target, the false intersection can be eliminated. Simulations show that the presented method is valid.

  17. A water cooled, lithium lead breeding blanket for a DEMO fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G.; Rieger, M.; Biggio, M.; Farfaletti-Casali, F.; Tominetti, S.; Wu, J.; Zucchetti, M. (Commission of the European Communities, Ispra (Italy). Joint Research Centre); Labbe, P.; Baraer, L.; Gervaise, G.; Giancarli, L.; Roze, M.; Severi, Y.; Quintric-Bossy, J. (CEA Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France))

    1991-04-01

    The main features of a tritium breeding blanket for a Demonstration Power Reactor involving the eutectic Pb-17Li as liquid breeder and water as coolant are presented. The configuration of the blanket segments and breeder modules as well as their arrangement inside the reactor vacuum vessel are outlined. The main design aspects and the corresponding design limits are reviewed, namely those related to thermomechanics, neutronics, magneto-hydrodynamics, tritium permeation and recovery. First results of safety analysis, in particular those connected with the rupture of a coolant tube in the breeder module are presented and discussed. As a conclusion, the feasibility of the concept look attractive. A problem which requires further investigation is that of the tritium self-sufficiency. It is shown that a net tritium production near to one can be obtained if berylium tiles are placed in front of the plasma, provided that they are cooled by heavy water. (orig.).

  18. Finalization of the conceptual design of the auxiliary circuits for the European test blanket systems

    Energy Technology Data Exchange (ETDEWEB)

    Aiello, A., E-mail: antonio.aiello@enea.it [ENEA UTIS – C.R. Brasimone, Bacino del Brasimone, I-40032 Camugnano, BO (Italy); Ghidersa, B.E. [Karlsruher Institut für Technologie (KIT) – Institut für Neutronenphysik und Reaktortechnik (INR), D-76021 Karlsruhe (Germany); Utili, M. [ENEA UTIS – C.R. Brasimone, Bacino del Brasimone, I-40032 Camugnano, BO (Italy); Vala, L. [Sustainable Energy (SUSEN), Technological Experimental Circuits, Centrum vyzkumu Rez s.r.o. (CV Rez), Hlavni c.p. 130, CZ-250 68 Husinec-Rez (Czech Republic); Ilkei, T. [Institute for Particle and Nuclear Physics, Wigner Research Centre for Physics, Hungarian Academy of Sciences, Budapest H-1525 (Hungary); Di Gironimo, G.; Mozzillo, R.; Tarallo, A. [CREATE/University of Naples Federico II, Department of Industrial Engineering, P.le Tecchio 80, 80125 Naples (Italy); Ricapito, I.; Calderoni, P. [TBM& MD Project, Fusion for Energy, EU Commission, Carrer J. Pla, 2, Building B3, 08019 Barcelona (Spain)

    2015-10-15

    In view of the ITER conceptual design review, the design of the ancillary systems of the European test blanket systems presented in [1] has been updated and made consistent with the ITER requirements for the present design phase. Europe is developing two concepts of TBM, the helium cooled lithium lead (HCLL) and the helium cooled pebble bed (HCPB) one, having in common the cooling media, pressurized helium at 8 MPa [2]. TBS, namely helium cooling system (HCS), coolant purification system (CPS), lead lithium loop and tritium extraction/removal system (TES–TRS) have the purpose to cool down the TBM and to remove tritium to be driven to TEP from breeder and coolant. These systems are placed in port cell 16 (PC#16), chemical and volume control system (CVCS) area and tritium building. Starting from the pre-conceptual design developed in the past, more mature technical interfaces with the ITER facility have been consolidated and iterative design activities were performed to comply with design requirements/specifications requested by IO to conclude the conceptual design phase. In this paper the present status of design of the TBS is presented together with the preliminary integration in ITER areas.

  19. A Precambrian proximal ejecta blanket from Scotland

    Science.gov (United States)

    Amor, Kenneth; Hesselbo, Stephen P.; Porcelli, Don; Thackrey, Scott; Parnell, John

    2008-04-01

    Ejecta blankets around impact craters are rarely preserved onEarth. Although impact craters are ubiquitous on solid bodiesthroughout the solar system, on Earth they are rapidly effaced,and few records exist of the processes that occur during emplacementof ejecta. The Stac Fada Member of the Precambrian Stoer Groupin Scotland has previously been described as volcanic in origin.However, shocked quartz and biotite provide evidence for high-pressureshock metamorphism, while chromium isotope values and elevatedabundances of platinum group metals and siderophile elementsindicate addition of meteoritic material. Thus, the unit isreinterpreted here as having an impact origin. The ejecta blanketreaches >20 m in thickness and contains abundant dark green,vesicular, devitrified glass fragments. Field observations suggestthat the deposit was emplaced as a single fluidized flow thatformed as a result of an impact into water-saturated sedimentarystrata. The continental geological setting and presence of groundwatermake this deposit an analogue for Martian fluidized ejecta blankets.

  20. Stellar model atmospheres with magnetic line blanketing

    CERN Document Server

    Kochukhov, O; Shulyak, D

    2004-01-01

    Model atmospheres of A and B stars are computed taking into account magnetic line blanketing. These calculations are based on the new stellar model atmosphere code LLModels which implements direct treatment of the opacities due to the bound-bound transitions and ensures an accurate and detailed description of the line absorption. The anomalous Zeeman effect was calculated for the field strengths between 1 and 40 kG and a field vector perpendicular to the line of sight. The model structure, high-resolution energy distribution, photometric colors, metallic line spectra and the hydrogen Balmer line profiles are computed for magnetic stars with different metallicities and are discussed with respect to those of non-magnetic reference models. The magnetically enhanced line blanketing changes the atmospheric structure and leads to a redistribution of energy in the stellar spectrum. The most noticeable feature in the optical region is the appearance of the 5200 A depression. However, this effect is prominent only in ...

  1. Chicxulub Ejecta Blanket Deposits From Belize

    Science.gov (United States)

    Ocampo, A.

    1995-01-01

    The Chicxulub impact into a thick sequence of carbonates and sulfates released over a trillion tons of volatiles. The importance of the explosive release of such a large mass of volatiles has been greatly underestimated in studies of ejecta depositional processes. Proximal Chicxulub ejecta blanket deposits recent discovered on Albion Island in Belize provide a key to understanding the role of volatile-rich target material during large impact events.

  2. Analysis of Consistency of Printing Blankets using Correlation Technique

    Directory of Open Access Journals (Sweden)

    Lalitha Jayaraman

    2010-01-01

    Full Text Available This paper presents the application of an analytical tool to quantify material consistency of offset printing blankets. Printing blankets are essentially viscoelastic rubber composites of several laminas. High levels of material consistency are expected from rubber blankets for quality print and for quick recovery from smash encountered during the printing process. The present study aims at determining objectively the consistency of printing blankets at three specific torque levels of tension under two distinct stages; 1. under normal printing conditions and 2. on recovery after smash. The experiment devised exhibits a variation in tone reproduction properties of each blanket signifying the levels of inconsistency also in thicknessdirection. Correlation technique was employed on ink density variations obtained from the blanket on paper. Both blankets exhibited good consistency over three torque levels under normal printing conditions. However on smash the recovery of blanket and its consistency was a function of manufacturing and torque levels. This study attempts to provide a new metrics for failure analysis of offset printing blankets. It also underscores the need for optimizing the torque for blankets from different manufacturers.

  3. Analysis of Consistency of Printing Blankets using Correlation Technique

    Directory of Open Access Journals (Sweden)

    Balaraman Kumar

    2010-06-01

    Full Text Available This paper presents the application of an analytical tool to quantify material consistency of offset printing blankets. Printing blankets are essentially viscoelastic rubber composites of several laminas. High levels of material consistency are expected from rubber blankets for quality print and for quick recovery from smash encountered during the printing process. The present study aims at determining objectively the consistency of printing blankets at three specific torque levels of tension under two distinct stages; 1. under normal printing conditions and 2. on recovery after smash. The experiment devised exhibits a variation in tone reproduction properties of each blanket signifying the levels of inconsistency also in thickness direction. Correlation technique was employed on ink density variations obtained from the blanket on paper. Both blankets exhibited good consistency over three torque levels under normal printing conditions. However on smash the recovery of blanket and its consistency was a function of manufacturing and torque levels. This study attempts to provide a new metrics for failure analysis of offset printing blankets. It also underscores the need for optimising the torque for blankets from different manufacturers.

  4. Detection of Breeding Blankets Using Antineutrinos

    Science.gov (United States)

    Cogswell, Bernadette; Huber, Patrick

    2016-03-01

    The Plutonium Management and Disposition Agreement between the United States and Russia makes arrangements for the disposal of 34 metric tons of excess weapon-grade plutonium. Under this agreement Russia plans to dispose of its excess stocks by processing the plutonium into fuel for fast breeder reactors. To meet the disposition requirements this fuel would be burned while the fast reactors are run as burners, i.e., without a natural uranium blanket that can be used to breed plutonium surrounding the core. This talk discusses the potential application of antineutrino monitoring to the verification of the presence or absence of a breeding blanket. It is found that a 36 kg antineutrino detector, exploiting coherent elastic neutrino-nucleus scattering and made of silicon, could determine the presence of a breeding blanket at a liquid sodium cooled fast reactor at the 95% confidence level within 90 days. Such a detector would be a novel non-intrusive verification tool and could present a first application of coherent elastic neutrino-nucleus scattering to a real-world challenge.

  5. Design of ITER vacuum vessel in-wall shielding

    Energy Technology Data Exchange (ETDEWEB)

    Wang, X., E-mail: xiaoyu.wang@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Ioki, K. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Morimoto, M. [Mitsubishi Heavy Industries, 1-1, Wadasaki-cho 1-chome, Hyogo-ku, Kobe (Japan); Choi, C.H.; Utin, Y.; Sborchia, C. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); TaiLhardat, O. [Assystem EOS, ZAC SAINT MARTIN, 23 rue Benjamin Franklin, 84120 Pertuis (France); Mille, B.; Terasawa, A.; Gribov, Y.; Barabash, V.; Polunovskiy, E.; Dani, S. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Pathak, H.; Raval, J. [ITER-India, Institute for Plasma Research, Gandhinagar 382025 (India); Liu, S.; Lu, M.; Du, S. [Institute of Plasma Physics, China Academy of Sciences, Shushanhu Road 350, Hefei (China)

    2014-10-15

    The ITER vacuum vessel is a torus-shaped, double wall structure. The space between the double walls of the VV is filled with in-wall shielding (IWS) and cooling water. The main purpose of the in-wall shielding is to provide neutron shielding together with the blanket and VV shells and water during ITER plasma operation and to reduce the ripple of the Toroidal magnetic field. Based on ITER vacuum vessel structure and related requirements, in-wall shielding are designed as about 8900 individual blocks with different sizes and several different materials distributed over nine vessel sectors and nine field joints of vessel sectors. This paper presents the design of the IWS, considering loads, structural stresses and assembly method, and also shows neutron shielding effect and TF ripple reduced by the IWS.

  6. High power density self-cooled lithium-vanadium blanket.

    Energy Technology Data Exchange (ETDEWEB)

    Gohar, Y.; Majumdar, S.; Smith, D.

    1999-07-01

    A self-cooled lithium-vanadium blanket concept capable of operating with 2 MW/m{sup 2} surface heat flux and 10 MW/m{sup 2} neutron wall loading has been developed. The blanket has liquid lithium as the tritium breeder and the coolant to alleviate issues of coolant breeder compatibility and reactivity. Vanadium alloy (V-4Cr-4Ti) is used as the structural material because it can accommodate high heat loads. Also, it has good mechanical properties at high temperatures, high neutron fluence capability, low degradation under neutron irradiation, good compatibility with the blanket materials, low decay heat, low waste disposal rating, and adequate strength to accommodate the electromagnetic loads during plasma disruption events. Self-healing electrical insulator (CaO) is utilized to reduce the MHD pressure drop. A poloidal coolant flow with high velocity at the first wall is used to reduce the peak temperature of the vanadium structure and to accommodate high surface heat flux. The blanket has a simple blanket configuration and low coolant pressure to reduce the fabrication cost, to improve the blanket reliability, and to increase confidence in the blanket performance. Spectral shifter, moderator, and reflector are utilized to improve the blanket shielding capability and energy multiplication, and to reduce the radial blanket thickness. Natural lithium is used to avoid extra cost related to the lithium enrichment process.

  7. Monte Carlo radiation shielding and activation analyses for the Diagnostic Equatorial Port Plug in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Serikov, A., E-mail: arkady.serikov@kit.edu [Karlsruhe Institute of Technology KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Fischer, U.; Leichtle, D. [Karlsruhe Institute of Technology KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Pitcher, C.S. [ITER Organization, Route de Vinon sur Verdon, 13115, St. Paul lez Durance (France)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer Systematic neutronics analyses were conducted to assess the ITER Equatorial Port Plug radiation shielding performance. Black-Right-Pointing-Pointer Shielding optimization was achieved by parametric analyses of several design variants using the MCNP5, FISPACT-2007, and R2Smesh codes. Black-Right-Pointing-Pointer Dominant effect of radiation streaming along the port plug gaps was recognized. Black-Right-Pointing-Pointer Combination of the gap labyrinths and streaming stoppers or rails reduces shutdown doses by 2 orders of magnitude. Black-Right-Pointing-Pointer Using the proposed shielding, the shutdown dose in the ITER port interspace is less than the personnel access limit of 100 {mu}Sv/h. - Abstract: This paper addresses neutronics aspects of the design development of the Diagnostic Generic Equatorial Port Plug (EPP) in ITER. To secure the personnel access at the EPP back-end interspace, parametric neutronics analyses of the EPP radiation environment have been performed and practical shielding solutions have been found. Radiation transport was performed with the Monte Carlo MCNP5 code. Activation calculations were conducted with the FISPACT-2007 inventory code. The R2Smesh approach was applied to couple transport and activation calculations. Newly created EPP local MCNP5 model was devised by extracting the EPP and adjacent blanket modules from the ITER Alite-4.1 model with proper modification of the EPP geometry in accordance with recent 3D CAD CATIA model. The EPP local model reproduces the EPP neutronically important features and allows investigation of the EPP neutronics effects in isolation from all other ITER components. Thorough EPP parametric analyses revealed dominant effect of gaps around EPP and several EPP design improvements were implemented as the outcomes of the analyses. Gap labyrinths and streaming stoppers inserted into the gaps were shown are capable to reduce the shutdown dose rate which is below the 100

  8. Approximate Modified Policy Iteration

    CERN Document Server

    Scherrer, Bruno; Ghavamzadeh, Mohammad; Geist, Matthieu

    2012-01-01

    Modified policy iteration (MPI) is a dynamic programming (DP) algorithm that contains the two celebrated policy and value iteration methods. Despite its generality, MPI has not been thoroughly studied, especially its approximation form which is used when the state and/or action spaces are large or infinite. In this paper, we propose three approximate MPI (AMPI) algorithms that are extensions of the well-known approximate DP algorithms: fitted-value iteration, fitted-Q iteration, and classification-based policy iteration. We provide an error propagation analysis for AMPI that unifies those for approximate policy and value iteration. We also provide a finite-sample analysis for the classification-based implementation of AMPI (CBMPI), which is more general (and somehow contains) than the analysis of the other presented AMPI algorithms. An interesting observation is that the MPI's parameter allows us to control the balance of errors (in value function approximation and in estimating the greedy policy) in the fina...

  9. 75 FR 51482 - Woven Electric Blankets From China

    Science.gov (United States)

    2010-08-20

    ... publishing the notice in the Federal Register of March 11, 2010 (75 FR 11557). The hearing was held in... COMMISSION Woven Electric Blankets From China Determination On the basis of the record \\1\\ developed in the... United States is materially injured by reason of imports from China of woven electric blankets,...

  10. 48 CFR 213.303 - Blanket purchase agreements (BPAs).

    Science.gov (United States)

    2010-10-01

    ... 48 Federal Acquisition Regulations System 3 2010-10-01 2010-10-01 false Blanket purchase agreements (BPAs). 213.303 Section 213.303 Federal Acquisition Regulations System DEFENSE ACQUISITION... PROCEDURES Simplified Acquisition Methods 213.303 Blanket purchase agreements (BPAs)....

  11. 48 CFR 8.405-3 - Blanket purchase agreements (BPAs).

    Science.gov (United States)

    2010-10-01

    ... 48 Federal Acquisition Regulations System 1 2010-10-01 2010-10-01 false Blanket purchase... Blanket purchase agreements (BPAs). (a)(1) Establishment. Ordering activities may establish BPAs under any..., before placing an order exceeding the micro-purchase threshold, the ordering activity shall— (i)...

  12. Applied iterative methods

    CERN Document Server

    Hageman, Louis A

    2004-01-01

    This graduate-level text examines the practical use of iterative methods in solving large, sparse systems of linear algebraic equations and in resolving multidimensional boundary-value problems. Assuming minimal mathematical background, it profiles the relative merits of several general iterative procedures. Topics include polynomial acceleration of basic iterative methods, Chebyshev and conjugate gradient acceleration procedures applicable to partitioning the linear system into a "red/black" block form, adaptive computational algorithms for the successive overrelaxation (SOR) method, and comp

  13. Properties of unirradiated and irradiated Ti-6Al-4V alloy for ITER flexible connectors

    Energy Technology Data Exchange (ETDEWEB)

    Rodchenkov, B.S., E-mail: rodchen@nikiet.ru [Research and Development Institute of Power Engineering (RDIPE), P.O. Box 788, Moscow 101000 (Russian Federation); Evseev, M.V. [Institute of Reactor Materials, Zarechnyi, Sverdlovsk Region 624051 (Russian Federation); Strebkov, Yu.S. [Research and Development Institute of Power Engineering (RDIPE), P.O. Box 788, Moscow 101000 (Russian Federation); Sinelnikov, L.P.; Shushlebin, V.V. [Institute of Reactor Materials, Zarechnyi, Sverdlovsk Region 624051 (Russian Federation)

    2011-10-01

    The high strength ({alpha} + {beta}) Ti-6Al-4V alloy was selected as the material for flexible attachments of the shield blanket modules in the ITER reactor. The different technologies used for manufacturing this alloy are: forging, stamping or pressing. The microstructures resulting from these processing methods can vary significantly and as a consequence the properties, including irradiation behavior, also vary. There are limited data available on the irradiation behavior of these materials. Specimens cut in the longitudinal and transversal directions of forged and stamped material were studied, with some of the specimens hydrogen charged to {approx}400 ppm H{sub 2}. In the unirradiated condition the forged alloy had slightly more ductility than the stamped alloy. The strength properties of both were practically the same. Neutron irradiation of these materials in the IVV-2M reactor reached doses of {approx}0.2 and 0.3 dpa at temperatures 240-260 deg. C. Irradiation resulted in substantial hardening and significant decrease of the fracture toughness of specimens from both materials.

  14. Iteration, Not Induction

    Science.gov (United States)

    Dobbs, David E.

    2009-01-01

    The main purpose of this note is to present and justify proof via iteration as an intuitive, creative and empowering method that is often available and preferable as an alternative to proofs via either mathematical induction or the well-ordering principle. The method of iteration depends only on the fact that any strictly decreasing sequence of…

  15. Pre-conceptual design study on K-DEMO ceramic breeder blanket

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Sung, E-mail: jspark@nfri.re.kr [National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Kwon, Sungjin; Im, Kihak; Kim, Keeman [National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Brown, Thomas; Neilson, George [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States)

    2015-11-15

    A pre-conceptual design study has been carried out for the Korean fusion demonstration reactor (K-DEMO) tokamak featured by high magnetic field (B{sub T0} = 7.4 T), R = 6.8 m, a = 2.1 m, and a steady-state operation. The design concepts of the K-DEMO blanket system considering the cooling in-vessel components with pressurized water and a solid pebble breeder are described herein. The structure of the K-DEMO blanket is toroidally subdivided into 16 inboard and 32 outboard sectors, in order to allow the vertical maintenance. Each blanket module is composed of plasma-facing first wall, layers of breeding parts, shielding and manifolds. A ceramic breeder using Li{sub 4}SiO{sub 4} pebbles with Be{sub 12}Ti as neuron multiplier is employed for study. MCNP neutronic simulations and thermo-hydraulic analyses are interactively performed in order to satisfy two key aspects: achieving a global Tritium Breeding Ratio (TBR) >1.05 and operating within the maximum allowable temperature ranges of materials.

  16. ITER at Cadarache; ITER a Cadarache

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-06-15

    This public information document presents the ITER project (International Thermonuclear Experimental Reactor), the definition of the fusion, the international cooperation and the advantages of the project. It presents also the site of Cadarache, an appropriate scientifical and economical environment. The last part of the documentation recalls the historical aspect of the project and the today mobilization of all partners. (A.L.B.)

  17. Simulation of sludge blanket height in clarifiers

    Institute of Scientific and Technical Information of China (English)

    ZHOU Zhen; WU Zhi-chao; WANG Zhi-wei; GU Guo-wei

    2009-01-01

    Sludge blanket height (SBH) is an important parameter in the clarifier design,operation and control.Based on an overview and classification of SBH algorithms,a modifed SBH algorithm is proposed by incorporating a threshold concentration limit into a relative concentration sharp change algorithm to eliminate the disturbance of compression interfaces on the correct simulation of SBH.Pilot-scale test data are adopted to compare reliability of three SBH algorithms reported in literature and the modified SBH algorithm developed in this paper.Calculated results demonstrate that the three SBH algorithms give results with large deviation (>50%) from measured SBH,especially under low solid flux conditions.The modified algorithm is computationally efficient and reliable in matching the measured data.It is incorporated into a onedimensional clarifier model for stable simulation of pilot-scale experimental clarifier data and into dynamic simulation of a full-scale wastewater treatment plant (WWTP) clarifier data.

  18. Assessment of the measurement performance of the in-vessel system of gap 6 of the ITER plasma position reflectometer using a finite-difference time-domain Maxwell full-wave code

    Science.gov (United States)

    da Silva, F.; Heuraux, S.; Ricardo, E.; Quental, P.; Ferreira, J.

    2016-11-01

    We conducted a first assessment of the measurement performance of the in-vessel components at gap 6 of the ITER plasma position reflectometry with the aid of a synthetic Ordinary Mode (O-mode) broadband frequency-modulated continuous-wave reflectometer implemented with REFMUL, a 2D finite-difference time-domain full-wave Maxwell code. These simulations take into account the system location within the vacuum vessel as well as its access to the plasma. The plasma case considered is a baseline scenario from Fusion for Energy. We concluded that for the analyzed scenario, (i) the plasma curvature and non-equatorial position of the antenna have neglectable impact on the measurements; (ii) the cavity-like space surrounding the antenna can cause deflection and splitting of the probing beam; and (iii) multi-reflections on the blanket wall cause a substantial error preventing the system from operating within the required error margin.

  19. Effects of Complex Symmetry-Breakings on Alpha Particle Power Loads on First Wall Structures and Equilibrium in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Shinohara, K. [Japan Atomic Energy Agency (JAEA), Naka; Kurki-Suonio, T. [Aalto University, Finland; Spong, Donald A [ORNL; Asunta, O. [Aalto University, Finland; Tani, K. [Japan Atomic Energy Agency (JAEA), Naka; Strumberger, E. [Max Planck Institute for Plasma Physics, Garching, Germany; Briguglio, S. [EURATOM / ENEA, Italy; Koskela, T. [Aalto University, Finland; Vlad, G. [EURATOM / ENEA, Italy; Günter, S. [Max-Planck Institute, Garching, Germany; Kramer, G. [Princeton Plasma Physics Laboratory (PPPL); Putvinski, S. [ITER Organization, Cadarache, France; Hamamatsu, K. [Japan Atomic Energy Agency (JAEA), Naka

    2011-01-01

    Within the ITPA Topical Group on Energetic Particles, we have investigated the impact that various mechanisms breaking the tokamak axisymmetry can have on the fusion alpha particle confinement in ITER as well as on the wall power loads due to these alphas. In addition to the well-known TF ripple, the 3D effect due to ferromagnetic materials (in ferritic inserts and test blanket modules) and ELM mitigation coils are included in these mechanisms. ITER scenario 4 was chosen since, due to its lower plasma current, it is more vulnerable for various off-normal features. First, the validity of using a 2D equilibrium was investigated: a 3D equilibrium was reconstructed using the VMEC code, and it was verified that no 3D equilibrium reconstruction is needed but it is sufficient to add the vacuum field perturbations onto an axisymmetric equilibrium. Then the alpha particle confinement was studied using three independent codes, ASCOT, DELTA5D and F3D OFMC, all of which assume MHD quiescent background plasma and no anomalous diffusion. All the codes gave a loss power fraction of about 0.2%. The distribution of the peak power load was found to depend on the first wall shape. We also made the first attempt to accommodate the effect of fast-ion-related MHD on the wall loads in ITER using the HMGC and ASCOT codes. The power flux to the wall was found to increase due to the redistribution of fast ions by the MHD activity. Furthermore, the effect of the ELM mitigation field on the fast-ion confinement was addressed by simulating NBI ions with the F3D OFMC code. The loss power fraction of NBI ions was found to increase from 0.3% without the ELM mitigation field to 4-5% with the ELM mitigation field.

  20. The development of ferritic steels for DEMO blanket

    Energy Technology Data Exchange (ETDEWEB)

    Kohyama, A. [Kyoto Univ. (Japan). Inst. of Advanced Energy; Hishinuma, A.; Shiba, K. [Tokai Establishment, JAERI, Tokai, Ibaraki (Japan); Kohno, Y. [Department of Materials Science, University of Tokyo, Hongo, Tokyo 113 (Japan); Sagara, A. [National Institute for Fusion Science, Toki, Gifu (Japan)

    1998-09-01

    The development of low-activation ferritic/martensitic steels is a key to the achievement of nuclear fusion as a safe, environmentally attractive and economically competitive energy source. The Japanese and the European Fusion Materials programs have put low-activation ferritic and martensitic steels R and D at the highest priority for a demonstration reactor (DEMO) and the beyond. An international collaborative test program on low-activation ferritic/martensitic steels for fusion is in progress as an activity of the International Energy Agency (IEA) fusion materials working group to verify the feasibility of using ferritic/martensitic steels for fusion by an extensive test program covering the most relevant technical issues for the qualification of a material for a nuclear application. The development of a comprehensive data base on the representative industrially processed reduced-activation steels of type 8-9Cr-2WVTa is underway for providing designers a preliminary set of material data for the mechanical design of components, e.g. for DEMO relevant blanket modules. The current design status of FFHR and SSTR utilizing low-activation ferritic steels is reviewed and future prospects are defined. (orig.) 12 refs.

  1. Design analyses of self-cooled liquid metal blankets

    Energy Technology Data Exchange (ETDEWEB)

    Gohar, Y.

    1986-12-01

    A trade-off study of liquid metal self-cooled blankets was carried out to define the performance of these blankets and to determine the potential to operate at the maximum possible values of the performance parameters. The main parameters considered during the course of the study were the tritium breeding ratio (TBR), the blanket energy multiplication factor, the energy fraction lost to the shield, the lithium-6 enrichment in the breeder material, the total blanket thickness, the reflector material selection, and the compositions of the different blanket zones. Also, a study was carried out to assess the impact of different reactor design choices on the reactor performance parameters. The design choices include the impurity control system (limiter or divertor), the material choice for the limiter, the elimination of tritium breeding from the inboard section of tokamak reactors, and the coolant choice for the nonbreeding inboard blanket. In addition, tritium breeding benchmark calculations were performed using different transport codes and nuclear data libraries. The importance of the TBR in the blanket design motivated the benchmark calculations.

  2. Heat transfer problems in gas-cooled solid blankets

    Energy Technology Data Exchange (ETDEWEB)

    Fillo, J.A.; Powell, J.R.

    1976-01-01

    In all fusion reactors using the deuterium-tritium fuel cycle, a large fraction approximately 80 percent of the fusion energy will be released as approximately 14 MeV neutrons which must be slowed down in a relatively thick blanket surrounding the plasma, thereby, converting their kinetic energy to high temperature heat which can be continuously removed by a coolant stream and converted in part to electricity in a conventional power turbine. Because of the primary goal of achieving minimum radioactivity, to date Brookhaven blanket concepts have been restricted to the use of some form of solid lithium, with inert gas-cooling and in some design cases, water-cooling of the shell structure. Aluminum and graphite have been identified as very promising structural materials for fusion blankets, and conceptual designs based on these materials have been made. Depending on the thermal loading on the ''first'' wall which surrounds the plasma as well as blanket design, heat transfer problems may be noticeably different in gas-cooled solid blankets. Approaches to solution of heat removal problems as well as explanation of: (a) the after-heat problems in blankets; (b) tritium breeding in solids; and (c) materials selection for radiation shields relative to the minimum activity blanket efforts at Brookhaven are discussed.

  3. ETR/ITER systems code

    Energy Technology Data Exchange (ETDEWEB)

    Barr, W.L.; Bathke, C.G.; Brooks, J.N.; Bulmer, R.H.; Busigin, A.; DuBois, P.F.; Fenstermacher, M.E.; Fink, J.; Finn, P.A.; Galambos, J.D.; Gohar, Y.; Gorker, G.E.; Haines, J.R.; Hassanein, A.M.; Hicks, D.R.; Ho, S.K.; Kalsi, S.S.; Kalyanam, K.M.; Kerns, J.A.; Lee, J.D.; Miller, J.R.; Miller, R.L.; Myall, J.O.; Peng, Y-K.M.; Perkins, L.J.; Spampinato, P.T.; Strickler, D.J.; Thomson, S.L.; Wagner, C.E.; Willms, R.S.; Reid, R.L. (ed.)

    1988-04-01

    A tokamak systems code capable of modeling experimental test reactors has been developed and is described in this document. The code, named TETRA (for Tokamak Engineering Test Reactor Analysis), consists of a series of modules, each describing a tokamak system or component, controlled by an optimizer/driver. This code development was a national effort in that the modules were contributed by members of the fusion community and integrated into a code by the Fusion Engineering Design Center. The code has been checked out on the Cray computers at the National Magnetic Fusion Energy Computing Center and has satisfactorily simulated the Tokamak Ignition/Burn Experimental Reactor II (TIBER) design. A feature of this code is the ability to perform optimization studies through the use of a numerical software package, which iterates prescribed variables to satisfy a set of prescribed equations or constraints. This code will be used to perform sensitivity studies for the proposed International Thermonuclear Experimental Reactor (ITER). 22 figs., 29 tabs.

  4. Review of past experiments at the FELIX facility and future plans for ITER applications

    Energy Technology Data Exchange (ETDEWEB)

    Hua, T.Q.; Turner, L.R.

    1993-10-01

    FELIX is an experimental test facility at Argonne National Laboratory (ANL) for the study of electromagnetic effects in first wall, blanket, shield systems of fusion reactors. From 1983 to 1986 five major test series, including static and dynamic tests, were conducted and are reviewed in this paper. The dynamic tests demonstrated an important coupling effect between eddy currents and motion in a conducting structure. Recently the US has proposed to the ITER Joint Central Team to use FELIX for testing mock-up components to study electromagnetic effects encountered during plasma disruptions and other off-normal events. The near and long term plans for ITER applications are discussed.

  5. Advanced Acoustic Blankets for Improved Aircraft Interior Noise Reduction Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The objective of the proposed Phase II research effort is to develop heterogeneous (HG) blankets for improved sound reduction in aircraft structures. Phase I...

  6. Performance of uncoated AFRSI blankets during multiple Space Shuttle flights

    Science.gov (United States)

    Sawko, Paul M.; Goldstein, Howard E.

    1992-01-01

    Uncoated Advanced Flexible Reusable Surface Insulation (AFRSI) blankets were successfully flown on seven consecutive flights of the Space Shuttle Orbiter OV-099 (Challenger). In six of the eight locations monitored (forward windshield, forward canopy, mid-fuselage, upper wing, rudder/speed brake, and vertical tail) the AFRSI blankets performed well during the ascent and reentry exposure to the thermal and aeroacoustic environments. Several of the uncoated AFRSI blankets that sustained minor damage, such as fraying or broken threads, could be repaired by sewing or by patching with a surface coating called C-9. The chief reasons for replacing or completely coating a blanket were fabric embrittlement and fabric abrasion caused by wind erosion. This occurred in the orbiter maneuvering system (OMS) pod sidewall and the forward mid-fuselage locations.

  7. Advanced Acoustic Blankets for Improved Aircraft Interior Noise Reduction Project

    Data.gov (United States)

    National Aeronautics and Space Administration — In this project advanced acoustic blankets for improved low frequency interior noise control in aircraft will be developed and demonstrated. The improved performance...

  8. 18 CFR 284.402 - Blanket marketing certificates.

    Science.gov (United States)

    2010-04-01

    ... effective for an affiliated marketer with respect to transactions involving affiliated pipelines when an affiliated pipeline receives its blanket certificate pursuant to § 284.284. (2) Should a marketer...

  9. Lightweight IMM Multi-Junction Photovoltaic Flexible Blanket Assembly Project

    Data.gov (United States)

    National Aeronautics and Space Administration — DSS's recently completed successful NASA SBIR Phase 1 program has established a TRL 3/4 classification for an innovative IMM PV Integrated Modular Blanket Assembly...

  10. Approximate iterative algorithms

    CERN Document Server

    Almudevar, Anthony Louis

    2014-01-01

    Iterative algorithms often rely on approximate evaluation techniques, which may include statistical estimation, computer simulation or functional approximation. This volume presents methods for the study of approximate iterative algorithms, providing tools for the derivation of error bounds and convergence rates, and for the optimal design of such algorithms. Techniques of functional analysis are used to derive analytical relationships between approximation methods and convergence properties for general classes of algorithms. This work provides the necessary background in functional analysis a

  11. Axial blanket for 16NGF Angra 1 fuel type

    Energy Technology Data Exchange (ETDEWEB)

    Sadde, Luciano Martins; Faria, Eduardo Fernandes [Industrias Nucleares do Brasil (INB), Resende, RJ (Brazil)]. E-mails: sadde@inb.gov.br; faria@inb.gov.br; Sang-Keun You [Korea Nuclear Fuel Co. Ltd. (KNFC), Taejon (Korea, Republic of)]. E-mail: skyou@knfc.co.kr

    2007-07-01

    Angra-1, Kori-2 and Krsko are nuclear power plants with the same design. However, the fuel assemblies have some differences in design due to the countries strategies and the differences in the fabrication process. The 16NGF (16x16 Next Generation Fuel) was developed by INB, KNFC and Westinghouse in order to be used in these three nuclear power plants and the 'Axial Blanket' is one of the new features for the 16NGF design. The main purpose of the Axial Blanket Optimization study is to determine which axial blanket enrichment and length would provide the better fuel cycle cost benefit. All of the calculations were performed using Gadolinium as Burnable Absorber and solid pellets type for Axial Blanket. The results indicate 1.8 w/o U235 enrichment and 8 inches length as the best option of Axial Blanket from the fuel cycle cost benefit standpoint. The economy is about 1.8%. The difference in the reload cost in the range between 1.5 and 2.6 w/o U235 enrichment and for the 6 and 8 inches length is not so significant. Due that, from the Fq limit standpoint and also for longer cycle length requirements, a higher axial blanket enrichment (2.6 w/o) and shorter length (6 inches) is recommended. (author)

  12. Neutron Measurement Instrumentation Development at KIT for the European ITER TBM

    Energy Technology Data Exchange (ETDEWEB)

    Klix, A.; Fischer, U.; Raj, P.; Reimann, Th.; Szalkai, D.; Tian, K. [Association KIT-EURATOM, Karlsruhe Institute of Technology, D-76344 Eggenstein-Leopoldshafen (Germany); Angelone, M. [Associazione ENEA-EURATOM sulla Fusione, ENEA C.R., I-00044 Frascati (Italy); Gehre, D. [Technical University of Dresden, D-01069 Dresden (Germany); Lyoussi, A. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France)

    2015-07-01

    Fusion power reactors will rely on the internal production of the fuel tritium from lithium in the tritium breeding blanket. Test Blanket Modules (TBM) will be installed in ITER with the aim to investigate the nuclear performance of different breeding blanket designs. Currently there is no fully qualified nuclear instrumentation available for the measurement of neutron fluxes and tritium production rates which would be able to withstand the harsh environment conditions in the TBM such as high temperature (>400 deg. C) and, depending on the operation scenario, intense radiation levels. As partner of the European Consortium on Nuclear Data and Measurement Techniques in the framework of several F4E specific grants and contracts, KIT and ENEA have jointly studied the possibility to develop and test detectors suitable to operate in ITER-TBMs. Here we present an overview of ongoing work on three types of neutron flux monitors under development for the TBMs with focus on the KIT activities. A neutron activation system (NAS) with pneumatic sample transport could provide absolute neutron flux measurements in selected positions. A test system for investigating activation materials with short half-lives was constructed at the DT neutron generator laboratory of Technical University of Dresden to investigate the neutronics aspects. Several irradiations have been performed with focus on the simultaneous measurement of the extracted activated probes. An engineering assessment of a TBM NAS in the conceptual design phase has been done which considered issues of design requirements and integration. Last but not least, a mechanical test bench is under construction at KIT which will address issues of driving the activation probes, solutions for loading the system etc. experimentally. Self-powered neutron detectors (SPND) are widely applied in fission reactor monitoring, and the commercially available SPNDs are sensitive to thermal neutrons. We are investigating novel materials for

  13. LDPC-coded unitary space-time modulation with low-complexity iterative demodulate-decoding scheme%具有低复杂度的LDPC已编码非相干酉空时调制系统设计

    Institute of Scientific and Technical Information of China (English)

    王利娇; 张林; 杨玲玲; 彭立; 符达伟

    2014-01-01

    提出一种克服无线信道瑞利衰落和高斯白噪声干扰的非相干编码调制MIMO系统方案。纠错码采用IEEE 802.16e 中的非规则 QC-LDPC 码,非相干调制采用基于三角函数的酉空时调制(SC-USTM)。在接收端,推导出SC-USTM的最大后验概率(MAP)解调算法;为了降低复杂度,构造了SC-USTM 的双解调器方案;为了改善双解调的性能,在置信传播(BP)迭代解码器和MAP解调器之间引入了迭代反馈机制。仿真实验表明LDPC已编码SC-USTM的MIMO系统比未编码USTM的MIMO系统在510-误码率时,性能改善15~17 dB,并且整个系统具有较低的计算复杂度。%A practical low-density parity-check(LDPC) coded unitary space-time modulation (USTM) noncoherent MI-MO communication system over Rayleigh flat fading and additive white Gaussian noise (AWGN) channel was con-structed, where the LDPC code is selected from IEEE 802.16e Standard and USTM is constructed by the unitary matrix based on the sinc-cosinc function, SC-USTM for short. At the receiver, the maximum a posteriori probability (MAP) de-modulating algorithm of the SC-USTM was first designed;for reducing the complexity of MAP demodulator, the dual demodulator was then conceived;for improving the performance, the iterative feedback between the belief propagation (BP) decoder and the MPA demodulator was finally introduced. The simulation test shows that the system of the 510-LDPC coded SC-USTM is superior to the uncoded SC-USTM about 15~17 dB at BER , and this system has very low complexity.

  14. Robust iterative methods

    Energy Technology Data Exchange (ETDEWEB)

    Saadd, Y.

    1994-12-31

    In spite of the tremendous progress achieved in recent years in the general area of iterative solution techniques, there are still a few obstacles to the acceptance of iterative methods in a number of applications. These applications give rise to very indefinite or highly ill-conditioned non Hermitian matrices. Trying to solve these systems with the simple-minded standard preconditioned Krylov subspace methods can be a frustrating experience. With the mathematical and physical models becoming more sophisticated, the typical linear systems which we encounter today are far more difficult to solve than those of just a few years ago. This trend is likely to accentuate. This workshop will discuss (1) these applications and the types of problems that they give rise to; and (2) recent progress in solving these problems with iterative methods. The workshop will end with a hopefully stimulating panel discussion with the speakers.

  15. Quantum Iterated Function Systems

    CERN Document Server

    Lozinski, A; Slomczynski, W; Lozinski, Artur; Zyczkowski, Karol; Slomczynski, Wojciech

    2003-01-01

    Iterated functions system (IFS) is defined by specifying a set of functions in a classical phase space, which act randomly on the initial point. In an analogous way, we define quantum iterated functions system (QIFS), where functions act randomly with prescribed probabilities in the Hilbert space. In a more general setting a QIFS consists of completely positive maps acting in the space of density operators. We present exemplary classical IFSs, the invariant measure of which exhibits fractal structure, and study properties of the corresponding QIFSs and their invariant state.

  16. Iterative List Decoding

    DEFF Research Database (Denmark)

    Justesen, Jørn; Høholdt, Tom; Hjaltason, Johan

    2005-01-01

    We analyze the relation between iterative decoding and the extended parity check matrix. By considering a modified version of bit flipping, which produces a list of decoded words, we derive several relations between decodable error patterns and the parameters of the code. By developing a tree...

  17. Iterative software kernels

    Energy Technology Data Exchange (ETDEWEB)

    Duff, I.

    1994-12-31

    This workshop focuses on kernels for iterative software packages. Specifically, the three speakers discuss various aspects of sparse BLAS kernels. Their topics are: `Current status of user lever sparse BLAS`; Current status of the sparse BLAS toolkit`; and `Adding matrix-matrix and matrix-matrix-matrix multiply to the sparse BLAS toolkit`.

  18. 75 FR 38459 - Certain Woven Electric Blankets From the People's Republic of China: Final Determination of Sales...

    Science.gov (United States)

    2010-07-02

    ... Antidumping Investigations involving Non-Market Economy Countries,'' which states: \\23\\ See Certain Woven... International Trade Administration Certain Woven Electric Blankets From the People's Republic of China: Final... Department'') has determined that certain woven electric blankets (``woven electric blankets'') from...

  19. Direct LiT Electrolysis in a Metallic Fusion Blanket

    Energy Technology Data Exchange (ETDEWEB)

    Olson, Luke [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-09-30

    A process that simplifies the extraction of tritium from molten lithium-based breeding blankets was developed. The process is based on the direct electrolysis of lithium tritide using a ceramic Li ion conductor that replaces the molten salt extraction step. Extraction of tritium in the form of lithium tritide in the blankets/targets of fusion/fission reactors is critical in order to maintain low concentrations. This is needed to decrease the potential tritium permeation to the surroundings and large releases from unforeseen accident scenarios. Extraction is complicated due to required low tritium concentration limits and because of the high affinity of tritium for the blanket. This work identified, developed and tested the use of ceramic lithium ion conductors capable of recovering hydrogen and deuterium through an electrolysis step at high temperatures.

  20. MHD pressure drop in ferritic pipes of fusion blankets

    Energy Technology Data Exchange (ETDEWEB)

    Reimann, J.; Buehler, Leo E-mail: leo.buehler@iket.fzk.de; Messadek, K.; Stieglitz, R

    2003-09-01

    Magnetohydrodynamic flows in pipes of ferromagnetic material is an important issue for liquid metal blanket concepts using MANET as wall material. Fusion relevant magnetic fields of 4-8 T cause high pressure drop in the blanket header where a massive structure of ferromagnetic material exists. It is briefly outlined that in the blanket the reduction of pressure drop due to magnetic shielding is limited to about 10%. Remarkable reduction of pressure drop is possible by means of electrical insulation that prevents currents from short-circuiting through the very thick walls of the headers. Direct contact of the insulating material with the liquid metal is excluded by using metallic liners. Results are reported on the fabrication of such a test section and corresponding pressure drop measurements confirm the effective contribution of the electrical decoupling.

  1. Direct Lit Electrolysis In A Metallic Lithium Fusion Blanket

    Energy Technology Data Exchange (ETDEWEB)

    Colon-Mercado, H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Babineau, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Elvington, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Garcia-Diaz, B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Teprovich, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Vaquer, A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-10-13

    A process that simplifies the extraction of tritium from molten lithium based breeding blankets was developed.  The process is based on the direct electrolysis of lithium tritide using a ceramic Li ion conductor that replaces the molten salt extraction step. Extraction of tritium in the form of lithium tritide in the blankets/targets of fission/fusion reactors is critical in order to maintained low concentrations.  This is needed to decrease the potential tritium permeation to the surroundings and large releases from unforeseen accident scenarios. Because of the high affinity of tritium for the blanket, extraction is complicated at the required low levels. This work identified, developed and tested the use of ceramic lithium ion conductors capable of recovering the hydrogen and deuterium thru an electrolysis step at high temperatures. 

  2. Direct LiT Electrolysis in a Metallic Fusion Blanket

    Energy Technology Data Exchange (ETDEWEB)

    Olson, Luke [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-09-30

    A process that simplifies the extraction of tritium from molten lithium based breeding blankets was developed. The process is based on the direct electrolysis of lithium tritide using a ceramic Li ion conductor that replaces the molten salt extraction step. Extraction of tritium in the form of lithium tritide in the blankets/targets of fusion/fission reactors is critical in order to maintain low concentrations. This is needed to decrease the potential tritium permeation to the surroundings and large releases from unforeseen accident scenarios. Extraction is complicated due to required low tritium concentration limits and because of the high affinity of tritium for the blanket. This work identified, developed and tested the use of ceramic lithium ion conductors capable of recovering hydrogen and deuterium through an electrolysis step at high temperatures.

  3. First wall and blanket concepts for experimental fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G.; Biggio, M.; Cardella, A.; Daenner, W.; Farfaletti-Casali, F.; Ponti, C.; Rieger, M.; Vieider, G.

    1985-07-01

    The paper describes the progress of the studies on first wall and liquid breeder blankets for tritium production in the Next European Torus (NET). Two concepts of first wall/blanket segments are described, using 17Li83Pb as breeder and water as coolant. In both concepts the first wall is integrated in a steel box enveloping the breeder units which are cylindrical vessels with an inside heat transfer system. The thermomechanical and neutronics features of the two concepts are evaluated. Finally, the questions related to tritium permeation into coolant and tritium recovery from breeder are discussed on the basis of the analysis in progress in Europe.

  4. Iterative Algorithms for Nonexpansive Mappings

    Directory of Open Access Journals (Sweden)

    Yao Yonghong

    2008-01-01

    Full Text Available Abstract We suggest and analyze two new iterative algorithms for a nonexpansive mapping in Banach spaces. We prove that the proposed iterative algorithms converge strongly to some fixed point of .

  5. Iterative supervirtual refraction interferometry

    KAUST Repository

    Al-Hagan, Ola

    2014-05-02

    In refraction tomography, the low signal-to-noise ratio (S/N) can be a major obstacle in picking the first-break arrivals at the far-offset receivers. To increase the S/N, we evaluated iterative supervirtual refraction interferometry (ISVI), which is an extension of the supervirtual refraction interferometry method. In this method, supervirtual traces are computed and then iteratively reused to generate supervirtual traces with a higher S/N. Our empirical results with both synthetic and field data revealed that ISVI can significantly boost up the S/N of far-offset traces. The drawback is that using refraction events from more than one refractor can introduce unacceptable artifacts into the final traveltime versus offset curve. This problem can be avoided by careful windowing of refraction events.

  6. Iterative participatory design

    DEFF Research Database (Denmark)

    Simonsen, Jesper; Hertzum, Morten

    2010-01-01

    iterative process of mutual learning by designers and domain experts (users), who aim to change the users’ work practices through the introduction of information systems. We provide an illustrative case example with an ethnographic study of clinicians experimenting with a new electronic patient record......The theoretical background in this chapter is information systems development in an organizational context. This includes theories from participatory design, human-computer interaction, and ethnographically inspired studies of work practices. The concept of design is defined as an experimental...... system, focussing on emergent and opportunity-based change enabled by appropriating the system into real work. The contribution to a general core of design research is a reconstruction of the iterative prototyping approach into a general model for sustained participatory design....

  7. SUMMARY REPORT-FY2006 ITER WORK ACCOMPLISHED

    Energy Technology Data Exchange (ETDEWEB)

    Martovetsky, N N

    2006-04-11

    Six parties (EU, Japan, Russia, US, Korea, China) will build ITER. The US proposed to deliver at least 4 out of 7 modules of the Central Solenoid. Phillip Michael (MIT) and I were tasked by DoE to assist ITER in development of the ITER CS and other magnet systems. We work to help Magnets and Structure division headed by Neil Mitchell. During this visit I worked on the selected items of the CS design and carried out other small tasks, like PF temperature margin assessment.

  8. New Iterated Decoding Algorithm Based on Differential Frequency Hopping System

    Institute of Scientific and Technical Information of China (English)

    LIANG Fu-lin; LUO Wei-xiong

    2005-01-01

    A new iterated decoding algorithm is proposed for differential frequency hopping (DFH) encoder concatenated with multi-frequency shift-key (MFSK) modulator. According to the character of the frequency hopping (FH) pattern trellis produced by DFH function, maximum a posteriori (MAP) probability theory is applied to realize the iterate decoding of it. Further, the initial conditions for the new iterate algorithm based on MAP algorithm are modified for better performance. Finally, the simulation result compared with that from traditional algorithms shows good anti-interference performance.

  9. Status of experimental data related to Be in ITER materials R and D data bank

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, Shigeru [ITER Joint Central Team, Muenchen (Germany)

    1998-01-01

    To keep traceability of many valuable raw data that were experimentally obtained in the ITER Technology R and D Tasks related to materials for In-Vessel components (divertor, first wall, blanket, vacuum vessel, etc.) and to easily make the best use of these data in the ITER design activities, the `ITER Materials R and D Data Bank` has been built up, with the use of Excel{sup TM} spread sheets. The paper describes status of experimental data collected in this data bank on thermo-mechanical properties of unirradiated and neutron irradiated Be, on plasma-material interactions of Be, on mechanical properties of various kinds of Be/Cu joints (including plasma sprayed Be), and on thermal fatigue tests of Be/Cu mock-ups. (author)

  10. The Iterate Manual

    Science.gov (United States)

    1990-10-01

    is probably a bad idea. A better versica would use a temporary: (defmacro sum-of-squares (expr) (let ((temp ( gensym ))) ’(lot (,temp ,expr)) (sum...val ( gensym )) (tempi ( gensym )) (temp2 ( gensym )) (winner (or var iterate::*result-var*))) ’(progn (with ,max-val - nil) (with ,winner = nil) (cond ((null...the elements of a vector (disregards fill-pointer)" (let ((vect ( gensym )) (end ( gensym )) (index ( gensym ))) ’(progn (with ,vect - v) (with ,end = (array

  11. ITER Safety Analyses with ISAS

    Science.gov (United States)

    Gulden, W.; Nisan, S.; Porfiri, M.-T.; Toumi, I.; de Gramont, T. Boubée

    1997-06-01

    Detailed analyses of accident sequences for the International Thermonuclear Experimental Reactor (ITER), from an initiating event to the environmental release of activity, have involved in the past the use of different types of computer codes in a sequential manner. Since these codes were developed at different time scales in different countries, there is no common computing structure to enable automatic data transfer from one code to the other, and no possibility exists to model or to quantify the effect of coupled physical phenomena. To solve this problem, the Integrated Safety Analysis System of codes (ISAS) is being developed, which allows users to integrate existing computer codes in a coherent manner. This approach is based on the utilization of a command language (GIBIANE) acting as a “glue” to integrate the various codes as modules of a common environment. The present version of ISAS allows comprehensive (coupled) calculations of a chain of codes such as ATHENA (thermal-hydraulic analysis of transients and accidents), INTRA (analysis of in-vessel chemical reactions, pressure built-up, and distribution of reaction products inside the vacuum vessel and adjacent rooms), and NAUA (transport of radiological species within buildings and to the environment). In the near future, the integration of S AFALY (simultaneous analysis of plasma dynamics and thermal behavior of in-vessel components) is also foreseen. The paper briefly describes the essential features of ISAS development and the associated software architecture. It gives first results of a typical ITER accident sequence, a loss of coolant accident (LOCA) in the divertor cooling loop inside the vacuum vessel, amply demonstrating ISAS capabilities.

  12. Strength Evaluation of Heat Affected Zone in Electron Beam Welded ARAA for HCCR TBM in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, J. S.; Kim, S. K.; Jin, H. G.; Lee, E. H.; Lee, D. W. [KAERI, Daejeon (Korea, Republic of); Cho, S. [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The Korean helium cooled ceramic reflector (HCCR) test blanket module (TBM) has been developed for ITER, and Korean reduced activation ferritic martensitic (RAFM) steel, called advanced reduced activation alloy (ARAA), has also been developed for a structural material of the HCCR TBM. One case of limited optimized electron beam (EB) welding conditions was selected based on previous work, and the weldability of an EB weld was evaluated for TBM fabrication. The micro-hardness was measured from the base to the weld region, and the microstructures were also observed. A small punch (SP) test considering the HAZ was carried out at room and high (550 .deg. C) temperatures. The empirical mechanical properties of HAZ in the EB weld were evaluated, and the fracture behavior was investigated after the SP test. The SP results show that the estimated yield and tensile strength of the HAZ were higher than the base metal at both temperatures. Korean RAFM steel, ARAA, was developed as a TBM structural material. Using one of the program alloys in ARAA (F206), one case of a limited optimized EB welding condition was selected based on previous works, and the weldability of an EB weld using the SP test was evaluated for TBM fabrication at room and high (550 .deg. C) temperatures. From a micro-Vickers hardness evaluation, the HAZ gave the highest values compared with the other regions. The irregular grain boundaries in the HAZ were observed, but its width was narrower than the TIG weld from the previous results. The optimized welding methods such as the TIG, EB, and laser weld, and the welding procedure considering the PWHT are being established, and the weldability evaluation is also progressing according to the development of the ARAA for the fusion material application in Korea.

  13. 75 FR 11557 - Woven Electric Blankets From China

    Science.gov (United States)

    2010-03-11

    ... permitted by section 201.8 of the Commission's rules, as amended, 67 FR 68036 (November 8, 2002). Even where... specified in II (C) of the Commission's Handbook on Electronic Filing Procedures, 67 FR 68168, 68173... COMMISSION Woven Electric Blankets From China AGENCY: United States International Trade Commission....

  14. First-wall/blanket materials selection for STARFIRE tokamak reactor

    Energy Technology Data Exchange (ETDEWEB)

    Smith, D.L.; Mattas, R.F.; Clemmer, R.G.; Davis, J.W.

    1980-01-01

    The development of the reference STARFIRE first-wall/blanket design involved numerous trade-offs in the materials selection process for the breeding material, coolant structure, neutron multiplier, and reflector. The major parameters and properties that impact materials selection and design criteria are reviewed.

  15. Technical issues for beryllium use in fusion blanket applications

    Energy Technology Data Exchange (ETDEWEB)

    McCarville, T.J.; Berwald, D.H.; Wolfer, W.; Fulton, F.J.; Lee, J.D.; Maninger, R.C.; Moir, R.W.; Beeston, J.M.; Miller, L.G.

    1985-01-01

    Beryllium is an excellent non-fissioning neutron multiplier for fusion breeder and fusion electric blanket applications. This report is a compilation of information related to the use of beryllium with primary emphasis on the fusion breeder application. Beryllium resources, production, fabrication, properties, radiation damage and activation are discussed. A new theoretical model for beryllium swelling is presented.

  16. ERGODIC THEOREM FOR INFINITE ITERATED FUNCTION SYSTEMS

    Institute of Scientific and Technical Information of China (English)

    O Hyong-chol; Ro Yong-hwa; Kil Won-gun

    2005-01-01

    A set of contraction maps of a metric space is called an iterated function systems.Iterated function systems with condensation can be considered infinite iterated function systems. Infinite iterated function systems on compact metric spaces were studied. Using the properties of Banach limit and uniform contractiveness, it was proved that the random iterating algorithms for infinite iterated function systems on compact metric spaces satisfy ergodicity. So the random iterating algorithms for iterated function systems with condensation satisfy ergodicity, too.

  17. Shut-Down Dose Rate analysis for ITER Diagnostic Equatorial and Upper Ports

    Energy Technology Data Exchange (ETDEWEB)

    Serikov, Arkady, E-mail: arkady.serikov@kit.edu [Karlsruhe Institute of Technology KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bertalot, Luciano [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Fischer, Ulrich [Karlsruhe Institute of Technology KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Pitcher, Charles Spencer; Suarez, Alejandro; Udintsev, Victor S. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Weinhorst, Bastian [Karlsruhe Institute of Technology KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2014-10-15

    Highlights: •Shut-Down Dose Rate (SDDR) analysis for ITER Diagnostic Equatorial and Upper Port Plugs (EPP and UPP). •ALARA principle and minimization of SDDR are used for the optimization of the port plugs shielding. •Activation and radiation shielding analyses with the MCNP5, FISPACT-2007, D1S and R2Smesh codes. •Significance of contribution of ELM/in-vessel coils and blanket manifolds into the port SDDR is shown. •Shielding improvements for EPP, UPP, and adjacent ITER components were proposed. -- Abstract: The Shut-Down Dose Rate (SDDR) is an important criterion of radiation safety for the personnel access for maintenance operations in ITER ports after the cessation of the D-T 14 MeV neutron fusion source. Therefore, the problem of the SDDR calculations attracts the attention of fusion neutronics community because SDDR in such a large and geometrically complicated fusion device as the ITER tokamak is challenging to compute. This challenge has been faced and overcome by applying dedicated methodological approaches explained in this paper. The results of the SDDR analysis allowed us to propose several design solutions for improvement of the radiation shielding of the ITER Generic Diagnostic Equatorial and Upper Port Plugs (EPP and UPP). The SDDR analysis was focused on the interspace area located between the ITER bioshield and port plugs where the personnel access is envisaged at ∼12 days after the ITER shut-down. By this analysis the radiation streaming pathways and dominant sources of decay radiation were revealed and the methods to mitigate the streaming and subsequent activation were found. The optimization of the port shielding was targeted on minimization of the SDDR in the interspace area following the ALARA principle and taking into account the feasibility to implement proposed shielding options with the actual hardware. Among them, wrapping the EPP walls with the B{sub 4}C tiles improves the EPP shielding performance. While void around the ELM

  18. Quantum iterated function systems.

    Science.gov (United States)

    Łoziński, Artur; Zyczkowski, Karol; Słomczyński, Wojciech

    2003-10-01

    An iterated function system (IFS) is defined by specifying a set of functions in a classical phase space, which act randomly on an initial point. In an analogous way, we define a quantum IFS (QIFS), where functions act randomly with prescribed probabilities in the Hilbert space. In a more general setting, a QIFS consists of completely positive maps acting in the space of density operators. This formalism is designed to describe certain problems of nonunitary quantum dynamics. We present exemplary classical IFSs, the invariant measure of which exhibits fractal structure, and study properties of the corresponding QIFSs and their invariant states.

  19. Iterative Magnetometer Calibration

    Science.gov (United States)

    Sedlak, Joseph

    2006-01-01

    This paper presents an iterative method for three-axis magnetometer (TAM) calibration that makes use of three existing utilities recently incorporated into the attitude ground support system used at NASA's Goddard Space Flight Center. The method combines attitude-independent and attitude-dependent calibration algorithms with a new spinning spacecraft Kalman filter to solve for biases, scale factors, nonorthogonal corrections to the alignment, and the orthogonal sensor alignment. The method is particularly well-suited to spin-stabilized spacecraft, but may also be useful for three-axis stabilized missions given sufficient data to provide observability.

  20. TOKOPS: Tokamak Reactor Operations Study: The influence of reactor operations on the design and performance of tokamaks with solid-breeder blankets: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Conn, R.W.; Ghoniem, N.M.; Firestone, M.A. (eds.)

    1986-09-01

    Reactor system operation and procedures have a profound impact on the conception and design of power plants. These issues are studied here using a model tokamak system employing a solid-breeder blanket. The model blanket is one which has evolved from the STARFIRE and BCSS studies. The reactor parameters are similar to those characterizing near-term fusion engineering reactors such as INTOR or NET (Next European Tokamak). Plasma startup, burn analysis, and methods for operation at various levels of output power are studied. A critical, and complicating, element is found to be the self-consistent electromagnetic response of the system, including the presence of the blanket and the resulting forces and loadings. Fractional power operation, and the strategy for burn control, is found to vary depending on the scaling law for energy confinement, and an extensive study is reported. Full-power reactor operation is at a neutron wall loading pf 5 MW/m/sup 2/ and a surface heat flux of 1 MW/m/sup 2/. The blanket is a pressurized steel module with bare beryllium rods and low-activation HT-9-(9-C-) clad LiAlO/sub 2/ rods. The helium coolant pressure is 5 MPa, entering the module at 297/sup 0/C and exiting at 550/sup 0/C. The system power output is rated at 1000 MW(e). In this report, we present our findings on various operational scenarios and their impact on system design. We first start with the salient aspects of operational physics. Time-dependent analyses of the blanket and balance of plant are then presented. Separate abstracts are included for each chapter.

  1. Design of In-vessel neutron monitor using micro fission chambers for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Nishitani, Takeo; Kasai, Satoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ebisawa, Katsuyuki [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Walker, Chris [ITER Joint Central Team, Garching (Germany)

    2001-10-01

    A neutron monitor using micro fission chambers to be installed inside the vacuum vessel has been designed for compact ITER (ITER-FEAT). We investigated the responses of the micro fission chambers to find the suitable position of micro fission chambers by a neutron Monte Carlo calculation using MCNP version 4b code. It was found that the averaged output of the micro fission chambers behind blankets at upper outboard and lower outboard is insensitive to the changes in the plasma position and the neutron source profile. A set of {sup 235}U micro fission chamber and ''blank'' detector which is a fissile material free detector to identify noise issues such as from {gamma}-rays are installed behind blankets. Employing both pulse counting mode and Campbelling mode in the electronics, the ITER requirement of 10{sup 7} dynamic range with 1 ms temporal resolution can be accomplished. The in-situ calibration has been simulated by MCNP calculation, where a point source of 14 MeV neutrons is moving on the plasma axis. It was found that the direct calibration is possible by using a neutron generator with an intensity of 10{sup 11} n/s. The micro fission chamber system can meet the required 10% accuracy for a fusion power monitor. (author)

  2. ITER中国液态锂铅实验包层模块热工水力学设计与分析%Thermal hydraulic design and analysis of the dual-function lithium-lead test blanket module for ITER

    Institute of Scientific and Technical Information of China (English)

    汪卫华; 刘松林; 吴宜灿

    2006-01-01

    依据ITER堆芯物理参数和中子学计算结果,给出了双功能锂铅实验包层模块热工水力学设计方案,并对氦气系统和液态金属锂铅系统压降和驱动功率进行了初步计算.相关中子学、结构、安全及MHD模拟结果证实了设计方案的可行性.

  3. Runaway electrons and ITER

    Science.gov (United States)

    Boozer, Allen

    2016-10-01

    ITER planning for avoiding runaway damage depends on magnetic surface breakup in fast relaxations. These arise in thermal quenches and in the spreading of impurities from massive gas injection or shattered pellets. Surface breakup would prevent a runaway to relativistic energies were it not for non-intercepting flux tubes, which contain magnetic field lines that do not intercept the walls. Such tubes persist near the magnetic axis and in the cores of islands but must dissipate before any confining surfaces re-form. Otherwise, a highly dangerous situation arises. Electrons that were trapped and accelerated in these flux tubes can fill a large volume of stochastic field lines and serve as a seed for the transfer of the full plasma current to runaways. If the outer confining surfaces are punctured, as by a drift into the wall, then the full runaway inventory will be lost in a short pulse along a narrow flux tube. Although not part of ITER planning, currents induced in the walls by the fast magnetic relaxation could be used to passively prevent outer surfaces re-forming. If magnetic surface breakup can be avoided during impurity injection, the plasma current could be terminated in tens of milliseconds by plasma cooling with no danger of runaway. Support by DoE Office of Fusion Energy Science Grant De-FG02-03ER54696.

  4. ITER helium ash accumulation

    Energy Technology Data Exchange (ETDEWEB)

    Hogan, J.T.; Hillis, D.L.; Galambos, J.; Uckan, N.A. (Oak Ridge National Lab., TN (USA)); Dippel, K.H.; Finken, K.H. (Forschungszentrum Juelich GmbH (Germany, F.R.). Inst. fuer Plasmaphysik); Hulse, R.A.; Budny, R.V. (Princeton Univ., NJ (USA). Plasma Physics Lab.)

    1990-01-01

    Many studies have shown the importance of the ratio {upsilon}{sub He}/{upsilon}{sub E} in determining the level of He ash accumulation in future reactor systems. Results of the first tokamak He removal experiments have been analysed, and a first estimate of the ratio {upsilon}{sub He}/{upsilon}{sub E} to be expected for future reactor systems has been made. The experiments were carried out for neutral beam heated plasmas in the TEXTOR tokamak, at KFA/Julich. Helium was injected both as a short puff and continuously, and subsequently extracted with the Advanced Limiter Test-II pump limiter. The rate at which the He density decays has been determined with absolutely calibrated charge exchange spectroscopy, and compared with theoretical models, using the Multiple Impurity Species Transport (MIST) code. An analysis of energy confinement has been made with PPPL TRANSP code, to distinguish beam from thermal confinement, especially for low density cases. The ALT-II pump limiter system is found to exhaust the He with maximum exhaust efficiency (8 pumps) of {approximately}8%. We find 1<{upsilon}{sub He}/{upsilon}{sub E}<3.3 for the database of cases analysed to date. Analysis with the ITER TETRA systems code shows that these values would be adequate to achieve the required He concentration with the present ITER divertor He extraction system.

  5. Iterative guided image fusion

    Directory of Open Access Journals (Sweden)

    Alexander Toet

    2016-08-01

    Full Text Available We propose a multi-scale image fusion scheme based on guided filtering. Guided filtering can effectively reduce noise while preserving detail boundaries. When applied in an iterative mode, guided filtering selectively eliminates small scale details while restoring larger scale edges. The proposed multi-scale image fusion scheme achieves spatial consistency by using guided filtering both at the decomposition and at the recombination stage of the multi-scale fusion process. First, size-selective iterative guided filtering is applied to decompose the source images into approximation and residual layers at multiple spatial scales. Then, frequency-tuned filtering is used to compute saliency maps at successive spatial scales. Next, at each spatial scale binary weighting maps are obtained as the pixelwise maximum of corresponding source saliency maps. Guided filtering of the binary weighting maps with their corresponding source images as guidance images serves to reduce noise and to restore spatial consistency. The final fused image is obtained as the weighted recombination of the individual residual layers and the mean of the approximation layers at the coarsest spatial scale. Application to multiband visual (intensified and thermal infrared imagery demonstrates that the proposed method obtains state-of-the-art performance for the fusion of multispectral nightvision images. The method has a simple implementation and is computationally efficient.

  6. ITER fast ion collective Thomson scattering, conceptual design of 60 GHz system

    Energy Technology Data Exchange (ETDEWEB)

    Meo, F.; Bindslev, H.; Korsholm, S.B.

    2007-08-15

    The collective Thomson scattering diagnostic for ITER at the 60 GHz range is capable of measuring the fast ion distribution parallel and perpendicular to the magnetic field at different radial locations simultaneously. The design is robust technologically with no moveable components near the plasma. The fast ion CTS diagnostic consists of two separate systems. Each system has its own RF launcher and separate set of detectors. The first system measures the perpendicular component of the fast ion velocity distribution. It consists of radially directed RF launcher and receiver, both located in the equatorial port on the low field side (LFS). This system will be referred to by the acronym LFS-BS system referring to the location of the receiver and the fact that it measures backscattered radiation. The second part of the CTS diagnostic measures the parallel component of the fast ion distribution. It consists of an RF launcher located in the mid-plane port on the LFS and a receiver mounted on the inner vacuum vessel wall that views the plasma from between two blanket modules. This system will be referred to as HFS-FS referring to the location of the receivers and that they measure forward scattered radiation. The design of both LFS-BS and HFS-FS receivers is aimed at measuring at different spatial locations simultaneously with no moveable components near the plasma. This report is a preliminary study of the hardware design and engineering constraints for this frequency range. Section 2 conceptually describes the two systems and their main components. Section 3 clarifies the impact of design parameters such as beam widths and scattering angle on the CTS measurements. With this in hand, the ITER measurement requirements are translated into constraints on the CTS system designs. An important result in this section is that systems can be designed inside these constraints. Section 4 outlines the technical feasibility and describes in more detail the design and the engineering

  7. On the Convergence of Iterative Receiver Algorithms Utilizing Hard Decisions

    Directory of Open Access Journals (Sweden)

    Jürgen F. Rößler

    2009-01-01

    Full Text Available The convergence of receivers performing iterative hard decision interference cancellation (IHDIC is analyzed in a general framework for ASK, PSK, and QAM constellations. We first give an overview of IHDIC algorithms known from the literature applied to linear modulation and DS-CDMA-based transmission systems and show the relation to Hopfield neural network theory. It is proven analytically that IHDIC with serial update scheme always converges to a stable state in the estimated values in course of iterations and that IHDIC with parallel update scheme converges to cycles of length 2. Additionally, we visualize the convergence behavior with the aid of convergence charts. Doing so, we give insight into possible errors occurring in IHDIC which turn out to be caused by locked error situations. The derived results can directly be applied to those iterative soft decision interference cancellation (ISDIC receivers whose soft decision functions approach hard decision functions in course of the iterations.

  8. Iterated crowdsourcing dilemma game

    Science.gov (United States)

    Oishi, Koji; Cebrian, Manuel; Abeliuk, Andres; Masuda, Naoki

    2014-02-01

    The Internet has enabled the emergence of collective problem solving, also known as crowdsourcing, as a viable option for solving complex tasks. However, the openness of crowdsourcing presents a challenge because solutions obtained by it can be sabotaged, stolen, and manipulated at a low cost for the attacker. We extend a previously proposed crowdsourcing dilemma game to an iterated game to address this question. We enumerate pure evolutionarily stable strategies within the class of so-called reactive strategies, i.e., those depending on the last action of the opponent. Among the 4096 possible reactive strategies, we find 16 strategies each of which is stable in some parameter regions. Repeated encounters of the players can improve social welfare when the damage inflicted by an attack and the cost of attack are both small. Under the current framework, repeated interactions do not really ameliorate the crowdsourcing dilemma in a majority of the parameter space.

  9. Laboratory experiments on drought and runoff in blanket peat

    OpenAIRE

    Holden, J; Burt, T. P.

    2002-01-01

    Global warming might change the hydrology of upland blanket peats in Britain. We have therefore studied in laboratory experiments the impact of drought on peat from the North Pennines of the UK. Runoff was dominated by surface and near-surface flow; flow decreased rapidly with depth and differed from one type of cover to another. Infiltration depended on the intensity of rain, and runoff responded rapidly to rain, with around 50% of rainwater emerging as overland flow. Drought changed the str...

  10. Development of insulating coatings for liquid metal blankets

    Energy Technology Data Exchange (ETDEWEB)

    Malang, S.; Borgstedt, H.U. [Kernforschungszentrum Karlsruhe GmbH (Germany); Farnum, E.H. [Los Alamos National Lab., NM (United States); Natesan, K. [Argonne National Lab., IL (United States); Vitkovski, I.V. [Efremov Inst., St. Petersburg (Russian Federation). MHD-Machines Lab.

    1994-07-01

    It is shown that self-cooled liquid metal blankets are feasible only with electrically insulating coatings at the duct walls. The requirements on the insulation properties are estimated by simple analytical models. Candidate insulator materials are selected based on insulating properties and thermodynamic consideration. Different fabrication technologies for insulating coatings are described. The status of the knowledge on the most crucial feasibility issue, the degradation of the resisivity under irradiation, is reviewed.

  11. MFTF-B Upgrade for blanket-technology testing

    Energy Technology Data Exchange (ETDEWEB)

    Thomassen, K.I.; Doggett, J.N.; Logan, B.G.

    1982-10-22

    Based on preliminary studies at Lawrence Livermore National Laboratory (LLNL), we believe the Mirror Fusion Test Facility (MFTF-B) could be upgraded for operation in a hot-ion Kelley mode in a portion of the central cell to provide fusion nuclear engineering data, particularly blanket technology information, by the end of the decade. Cost of this mode of operation would be modest compared with that of the other fusion devices considered in the last few years for such purposes.

  12. Blanket comparison and selection study. Final report. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1984-09-01

    The study focused on: (1) Development of reference design guidelines, evaluation criteria, and a methodology for evaluating and ranking candidate blanket concepts. (2) Compilation of the required data base and development of a uniform systems analysis for comparison. (3) Development of conceptual designs for the comparative evaluation. (4) Evaluation of leading concepts for engineering feasibility, economic performance, and safety. (5) Identification and prioritization of R and D requirements for the leading blanket concepts. Sixteen concepts (nine TMR and seven tokamak) which were identified as leading candidates in the early phases of the study, were evaluated in detail. The overall evaluation concluded that the following concepts should provide the focus for the blanket R and D program: (Breeder/Coolant/Structure), Lithium/Lithium/Vanadium Alloy, Li/sub 2/O/Helium/Ferritic Steel, LiPb Alloy/LiPb Alloy/Vanadium Alloy, and Lithium/Helium/Ferritic Steel. The primary R and D issues for the Li/Li/V concept are the development of an advanced structural alloy, resolution of MHD and corrosion problems, provision for an inert atmosphere (e.g., N/sub 2/) in the reactor building, and the development of non-water cooled near-plasma components, particularly for the tokamak. The main issues for the LiPb/LiPb/V concept are similar to the Li/Li/V blanket with the addition of resolving the tritium recovery issue. The R and D issues for Li/sub 2/O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concept are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue.

  13. Blanket comparison and selection study. Final report. Volume 3

    Energy Technology Data Exchange (ETDEWEB)

    1984-09-01

    The study focused on: (1) Development of reference design guidelines, evaluation criteria, and a methodology for evaluating and ranking candidate blanket concepts. (2) Compilation of the required data base and development of a uniform systems analysis for comparison. (3) Development of conceptual designs for the comparative evaluation. (4) Evaluation of leading concepts for engineering feasibility, economic performance, and safety. (5) Identification and prioritization of R and D requirements for the leading blanket concepts. Sixteen concepts (nine TMR and seven tokamak) which were identified as leading candidates in the early phases of the study, were evaluated in detail. The overall evaluation concluded that the following concepts should provide the focus for the blanket R and D program: (Breeder/Coolant/Structure), Lithium/Lithium/Vanadium Alloy, Li/sub 2/O/Helium/Ferritic Steel, LiPb Alloy/LiPb Alloy/Vanadium Alloy, and Lithium/Helium/Ferritic Steel. The primary R and D issues for the Li/Li/V concept are the development of an advanced structural alloy, resolution of MHD and corrosion problems, provision for an inert atmosphere (e.g., N/sub 2/) in the reactor building, and the development of non-water cooled near-plasma components, particularly for the tokamak. The main issues for the LiPb/LiPb/V concept are similar to the Li/Li/V blanket with the addition of resolving the tritium recovery issue. The R and D issues for Li/sub 2/O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concepts are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue.

  14. Blanket comparison and selection study. Final report. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    1984-09-01

    The study focused on: (1) Development of reference design guidelines, evaluation criteria, and a methodology for evaluating and ranking candidate blanket concepts. (2) Compilation of the required data base and development of a uniform systems analysis for comparison. (3) Development of conceptual designs for the comparative evaluation. (4) Evaluation of leading concepts for engineering feasibility, economic performance, and safety. (5) Identification and prioritization of R and D requirements for the leading blanket concepts. Sixteen concepts (nine TMR and seven tokamak) which were identified as leading candidates in the early phases of the study, were evaluated in detail. The overall evaluation concluded that the following concepts should provide the focus for the blanket R and D program: (Breeder/Coolant/Structure), Lithium/Lithium/Vanadium Alloy, Li/sub 2/O/Helium/Ferritic Steel, LiPb Alloy/LiPb Alloy/Vanadium Alloy, and Lithium/Helium/Ferritic Steel. The primary R and D issues for the Li/Li/V concept are the development of an advanced structural alloy, resolution of MHD and corrosion problems, provision for an inert atmosphere (e.g., N/sub 2/) in the reactor building, and the development of non-water cooled near-plasma components, particularly for the tokamak. The main issues for the LiPb/LiPb/V concepts are similar to the Li/Li/V blanket with the addition of resolving the tritium recovery issue. The R and D issues for Li/sub 2/O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concept are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue.

  15. Model problem of MHD flow in a lithium blanket

    Energy Technology Data Exchange (ETDEWEB)

    Cherepanov, V.Y.

    1978-01-01

    A model problem is considered for a feasibility study concerning controlled MHD flow in the blanket of a Tokamak nuclear reactor. The fundamental equations for the steady flow of an incompressible viscous fluid in a uniform transverse magnetic field are solved in rectangular coordinates, in the zero-induction approximation and with negligible induced currents. A numerical solution obtained for a set of appropriate boundary constraints establishes the conditions under which no stagnation zones will be formed.

  16. IVVS actuating system compatibility test to ITER gamma radiation conditions

    Energy Technology Data Exchange (ETDEWEB)

    Rossi, Paolo, E-mail: paolo.rossi@enea.it [Associazione EURATOM-ENEA sulla Fusione, 45 Via Enrico Fermi, 00044 Frascati, Rome (Italy); Collibus, M. Ferri de; Florean, M.; Monti, C.; Mugnaini, G.; Neri, C.; Pillon, M.; Pollastrone, F. [Associazione EURATOM-ENEA sulla Fusione, 45 Via Enrico Fermi, 00044 Frascati, Rome (Italy); Baccaro, S.; Piegari, A. [ENEA CR Casaccia, 301 Via Anguillarese, 00123 Santa Maria di Galeria, Rome (Italy); Damiani, C.; Dubus, G. [Fusion For Energy c/Josep Pla, n° 2 Torres Diagonal Litoral, 08019 Barcelona (Spain)

    2013-10-15

    Highlights: • ENEA developed and tested a prototype of a laser In Vessel Viewing and ranging System (IVVS) for ITER. • One piezo-motor prototype has been tested on the ENEA Calliope gamma irradiation facility to verify its compatibility to ITER gamma radiation conditions. • After a total dose of more than 4 MGy the piezo-motor maintained almost the same working parameters monitored before test without any evident and significant degradation of functionality. • After the full gamma irradiation test, the same piezo-motor assembly will be tested with 14 MeV neutrons irradiation using ENEA FNG facility. -- Abstract: The In Vessel Viewing System (IVVS) is a fundamental remote handling equipment, which will be used to make a survey of the status of the blanket first wall and divertor plasma facing components. A design and testing activity is ongoing, in the framework of a Fusion for Energy (F4E) grant agreement, to make the IVVS probe design compatible with ITER operating conditions and in particular, but not only, with attention to neutrons and gammas fluxes and both space constraints and interfaces. The paper describes the testing activity performed on the customized piezoelectric motors and the main components of the actuating system of the IVVS probe with reference to ITER gamma radiation conditions. In particular the test is performed on the piezoelectric motor, optical encoder and small scale optical samples .The test is carried out on the ENEA Calliope gamma irradiation facility at ITER relevant gamma fields at rate of about 2.5 kGy/h and doses of 4 MGy. The paper reports in detail the setup arrangement of the test campaign in order to verify significant working capability of the IVVS actuating components and the results are shown in terms of functional performances and parameters. The overall test campaign on IVVS actuating system will be completed on other ENEA testing facilities in order to verify compatibility to Magnetic field, neutrons and thermal

  17. Continuous fine pattern formation by screen-offset printing using a silicone blanket

    Science.gov (United States)

    Nomura, Ken-ichi; Kusaka, Yasuyuki; Ushijima, Hirobumi; Nagase, Kazuro; Ikedo, Hiroaki; Mitsui, Ryosuke; Takahashi, Seiya; Nakajima, Shin-ichiro; Iwata, Shiro

    2014-09-01

    Screen-offset printing combines screen-printing on a silicone blanket with transference of the print from the blanket to a substrate. The blanket absorbs organic solvents in the ink, and therefore, the ink does not disperse through the material. This prevents blurring and allows fine patterns with widths of a few tens of micrometres to be produced. However, continuous printing deteriorates the pattern’s shape, which may be a result of decay in the absorption abilities of the blanket. Thus, we have developed a new technique for refreshing the blanket by substituting high-boiling-point solvents present on the blanket surface with low-boiling-point solvents. We analyse the efficacy of this technique, and demonstrate continuous fine pattern formation for 100 screen-offset printing processes.

  18. Design and safety analysis of the helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Shuai; Zhou, Guangming; Lv, Zhongliang; Jin, Cheng; Chen, Hongli [University of Science and Technology of China, Anhui (China). School of Nuclear Science and Technology

    2016-05-15

    This paper reports the design and safety analysis results of the helium cooled solid breeder blanket of the Chinese Fusion Engineering Test Reactor (CFETR). Materials selection and basic structure of the blanket have been presented. Performance analysis including neutronics analysis and thermo-mechanical analysis has shown good results. And the safety analysis of the blanket under Loss Of Coolant Accident (LOCA) conditions has been described. Results showed the current design can deal well with the selected accident scenarios.

  19. Prevalence of enterobiasis and its incidence after blanket chemotherapy in a male orphanage.

    Science.gov (United States)

    Sirivichayakul, C; Pojjaroen-anant, C; Wisetsing, P; Lalitphiphat, A; Chanthavanich, P; Kabkaew, K

    2000-03-01

    A prospective observational study was conducted in a male orphanage to find out the prevalence of enterobiasis and its incidence after blanket chemotherapy using mebendazole. We found that the prevalence of enterobiasis was 28.9%. The incidence density of enterobiasis after blanket chemotherapy was 379.82 per 1,000 person-years which was quite high. We suggest that blanket chemotherapy should be repeated at every 6 months interval to control enterobiasis in orphanages.

  20. 76 FR 44903 - Kinder Morgan Interstate Gas Transmission, LLC; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2011-07-27

    ...-000] Kinder Morgan Interstate Gas Transmission, LLC; Notice of Request Under Blanket Authorization Take notice that on June 30, 2011 Kinder Morgan Interstate Gas Transmission, LLC (KMIGT), Post...

  1. Some geometrical iteration methods for nonlinear equations

    Institute of Scientific and Technical Information of China (English)

    LU Xing-jiang; QIAN Chun

    2008-01-01

    This paper describes geometrical essentials of some iteration methods (e.g. Newton iteration,secant line method,etc.) for solving nonlinear equations and advances some geomet-rical methods of iteration that are flexible and efficient.

  2. The first fusion reactor: ITER

    Science.gov (United States)

    Campbell, D. J.

    2016-11-01

    Established by the signature of the ITER Agreement in November 2006 and currently under construction at St Paul-lez-Durance in southern France, the ITER project [1,2] involves the European Union (including Switzerland), China, India, Japan, the Russian Federation, South Korea and the United States. ITER (`the way' in Latin) is a critical step in the development of fusion energy. Its role is to provide an integrated demonstration of the physics and technology required for a fusion power plant based on magnetic confinement.

  3. PICARD ITERATION FOR NONSMOOTH EQUATIONS

    Institute of Scientific and Technical Information of China (English)

    Song-bai Sheng; Hui-fu Xu

    2001-01-01

    This paper presents an analysis of the generalized Newton method, approximate Newton methods, and splitting methods for solving nonsmooth equations from Picard iteration viewpoint. It is proved that the radius of the weak Jacobian (RGJ) of Picard iteration function is equal to its least Lipschitz constant. Linear convergence or superlinear convergence results can be obtained provided that RGJ of the Picard iteration function at a solution point is less than one or equal to zero. As for applications, it is pointed out that the approximate Newton methods, the generalized Newton method for piecewise C1problems and splitting methods can be explained uniformly with the same viewpoint.

  4. Study for reducing radioactive solid waste at ITER decommissioning period

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Shinichi; Araki, Masanori; Ohmori, Junji; Ohno, Isamu; Sato, Satoshi; Yamauchi, Michinori; Nishitani, Takeo [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    2002-11-01

    It is one of the foremost goals for ITER to demonstrate the attractiveness with regard to safety and environmental potential. This implies that the radioactive materials and waste at decommissioning phase should carefully be treated with prescribed regulations. As possible activities during the Coordinated Technical Activity (CTA), the authors have performed a feasibility study for searching the possibility of effective reduction in the activated level as reasonably achievable as possible by taking account of minimum material changes while keeping original design concept and structure. Major induced activation in ITER comes from activated nickel and cobalt so that it is effective for the major structural components to minimize their material contents. Employing less Ni and Co steel in place of high-Ni austenitic stainless steel for blanket shield block, vacuum vessel shield material and TF coil casing has been considered as one of the effective plans to reduce the activated materials at the decommissioning phase. In this study, two less-Ni austenitic stainless steels are evaluated; one is high-Mn austenitic stainless steel JK2 which is developing for jacket material of ITER CS coil and the other is SS204L/ASTM-XM-11 which is also high-Mn steel specified in the popular standards such as American Society of Testing and Material (ASTM). Based on the material changes, activation analyses have been performed to investigate the possibility of reducing radioactive wastes. As a most impressive result, at 40 years after the termination some of main components such as a TF coil casing will reach to the clearance level which is specified by IAEA, and most components will be categorized into extremely low level waste except for limited components. These results will give the appropriate short decommissioning period that is assumed to start at 100 years after the termination in the original design. (author)

  5. Existence test for asynchronous interval iterations

    DEFF Research Database (Denmark)

    Madsen, Kaj; Caprani, O.; Stauning, Ole

    1997-01-01

    as a synchronous (sequential) interval iteration:In each iteration step all components of the iterate are calculatedbased on the previous iterate. In this case it is straight forward to base simple interval existence and non-existencetests on the calculations done in each step of the iteration. The search can also...... be performed as an asynchronous (parallel) iteration: Only a few components are changed in each stepand this calculation is in general based on components from differentprevious iterates. For the asynchronous iteration it turns out thatsimple tests of existence and non-existence can be based...

  6. Updated safety analysis of ITER

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, Neill, E-mail: neill.taylor@iter.org [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Baker, Dennis; Ciattaglia, Sergio; Cortes, Pierre; Elbez-Uzan, Joelle; Iseli, Markus; Reyes, Susana; Rodriguez-Rodrigo, Lina; Rosanvallon, Sandrine; Topilski, Leonid [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France)

    2011-10-15

    An updated version of the ITER Preliminary Safety Report has been produced and submitted to the licensing authorities. It is revised and expanded in response to requests from the authorities after their review of an earlier version in 2008, to reflect enhancements in ITER safety provisions through design changes, to incorporate new and improved safety analyses and to take into account other ITER design evolution. The updated analyses show that changes to the Tokamak cooling water system design have enhanced confinement and reduced potential radiological releases as well as removing decay heat with very high reliability. New and updated accident scenario analyses, together with fire and explosion risk analyses, have shown that design provisions are sufficient to minimize the likelihood of accidents and reduce potential consequences to a very low level. Taken together, the improvements provided a stronger demonstration of the very good safety performance of the ITER design.

  7. Iterative solution of linear systems

    Science.gov (United States)

    Freund, Roland W.; Golub, Gene H.; Nachtigal, Noel M.

    1992-01-01

    Recent advances in the field of iterative methods for solving large linear systems are reviewed. The main focus is on developments in the area of conjugate gradient-type algorithms and Krylov subspace methods for nonHermitian matrices.

  8. Cooperation between CERN and ITER

    CERN Multimedia

    2008-01-01

    CERN and the International Fusion Organisation ITER have just signed a first cooperation agreeement. Kaname Ikeda, the Director-General of the International Fusion Energy Organisation (ITER) (on the right) and Robert Aymar, Director-General of CERN, signing the agreement.The Director-General of the International Fusion Energy Organization, Mr Kaname Ikeda, and CERN Director-General, Robert Aymar, signed a cooperation agreement at a meeting on the Meyrin site on Thursday 6 March. One of the main purposes of this agreement is for CERN to give ITER the benefit of its experience in the field of technology as well as in administrative domains such as finance, procurement, human resources and informatics through the provision of consultancy services. Currently in its start-up phase at its Cadarache site, 70 km from Marseilles (France), ITER will focus its research on the scientific and technical feasibility of using fusion energy as a fu...

  9. Iterative Reconstruction Methods for Hybrid Inverse Problems in Impedance Tomography

    DEFF Research Database (Denmark)

    Hoffmann, Kristoffer; Knudsen, Kim

    2014-01-01

    For a general formulation of hybrid inverse problems in impedance tomography the Picard and Newton iterative schemes are adapted and four iterative reconstruction algorithms are developed. The general problem formulation includes several existing hybrid imaging modalities such as current density...... impedance imaging, magnetic resonance electrical impedance tomography, and ultrasound modulated electrical impedance tomography, and the unified approach to the reconstruction problem encompasses several algorithms suggested in the literature. The four proposed algorithms are implemented numerically in two...... be based on a theoretical analysis of the underlying inverse problem....

  10. ITER leader to head CERN

    CERN Multimedia

    Feder, Toni

    2003-01-01

    After successfully chairing an external review committee for CERN last year, Robert Aymar will leave ITER to become director general of the European particle physics laboratory rom 2004. Before ITER he also successfully managed the startup or Tore Supra. He will attempt to ensure that the LHC begins operating in 2007 - two years late - and is paid for by 2010 and will also start the planning for life after the LHC (1 page)

  11. ITER neutral beam system US conceptual design

    Energy Technology Data Exchange (ETDEWEB)

    Purgalis, P.

    1990-09-01

    In this document we present the US conceptual design of a neutral beam system for International Thermonuclear Experimental Reactor (ITER). The design incorporates a barium surface conversion D{sup {minus}} source feeding a linear array of accelerator channels. The system uses a dc accelerator with electrostatic quadrupoles for strong focusing. A high voltage power supply that is integrated with the accelerator is presented as an attractive option. A gas neutralizer is used and residual ions exiting the neutralizer are deflected to water-cooled dumps. Cryopanels are located at the accelerator exit to pump excess gas from the source and the neutralizer, and in the ion dump cavity to pump re-neutralized ions and neutralizer gas. All the above components are packaged in compact identical, independent modules which can be removed for remote maintenance. The neutral beam system delivers 75 MW of DO at 1.3 MeV, into three ports with a total of 9 modules arranged in stacks of three modules per port . To increase reliability each module is designed to deliver up to 10 MW; this allows eight modules operating at partial capacity to deliver the required power in the event one module is out of service, and provides 20% excess capacity to improve availability. Radiation protection is provided by shielding and by locating critical components in the source and accelerator 46.5 m from the torus centerline. Neutron shielding in the drift duct and neutralizer provides the added feature of limiting conductance and thus reducing gas flow to and from the torus.

  12. Inhibition of Frying Oil Oxidation by Carbon Dioxide Blanketing.

    Science.gov (United States)

    Totani, Nagao; Inoue, Ryota; Yawata, Miho

    2016-06-01

    The oxidation of oil starts, in general, from the penetration of atmospheric oxygen into oil. Inhibition of the vigorous oxidation of oil at deep-frying temperature under carbon dioxide flow, by disrupting the contact between oil and air, was first demonstrated using oil in a round bottom flask. Next, the minimum carbon dioxide flow rate necessary to blanket 4 L of frying oil in an electric fryer (surface area 690 cm(2)) installed with nonwoven fabric cover, was found to be 40 L/h. Then deep-frying of potato was done accordingly; immediately after deep-frying, an aluminum cover was placed on top of the nonwoven fabric cover to prevent the loss of carbon dioxide and the carbon dioxide flow was shut off. In conclusion, the oxidation of oil both at deep-frying temperature and during standing was remarkably inhibited by carbon dioxide blanketing at a practical flow rate and volume. Under the deep-frying conditions employed in this study, the increase in polar compound content was reduced to half of that of the control.

  13. Uranium self-shielding in fast reactor blankets

    Energy Technology Data Exchange (ETDEWEB)

    Kadiroglu, O.K.; Driscoll, M.J.

    1976-03-01

    The effects of heterogeneity on resonance self-shielding are examined with particular emphasis on the blanket region of the fast breeder reactor and on its dominant reaction--capture in /sup 238/U. The results, however, apply equally well to scattering resonances, to other isotopes (fertile, fissile and structural species) and to other environments, so long as the underlying assumptions of narrow resonance theory apply. The heterogeneous resonance integral is first cast into a modified homogeneous form involving the ratio of coolant-to-fuel fluxes. A generalized correlation (useful in its own right in many other applications) is developed for this ratio, using both integral transport and collision probability theory to infer the form of correlation, and then relying upon Monte Carlo calculations to establish absolute values of the correlation coefficients. It is shown that a simple linear prescription can be developed for the flux ratio as a function of only fuel optical thickness and the fraction of the slowing-down source generated by the coolant. This in turn permitted derivation of a new equivalence theorem relating the heterogeneous self-shielding factor to the homogeneous self-shielding factor at a modified value of the background scattering cross section per absorber nucleus. A simple version of this relation is developed and used to show that heterogeneity has a negligible effect on the calculated blanket breeding ratio in fast reactors.

  14. Elevator mode convection in liquid metal blankets for fusion reactors

    Science.gov (United States)

    Zikanov, Oleg; Liu, Li

    2015-11-01

    The work is motivated by the design of liquid-metal blankets for nuclear fusion reactors. Mixed convection in a downward flow in a vertical duct with strong contant-rate heating of one wall (the Grashof number up to 1012) and strong transverse magnetic field (the Hartmann number up to 104) is considered. It is found that in an infinitely long duct the flow is dominated by exponentially growing elevator modes having the form of a combination of ascending and descending jets. An analytical solution approximating the growth rate of the modes is derived. Analogous flows in finite-length pipes and ducts are analyzed using the high-resolution numerical simulations. The results of the recent experiments are reproduced and explained. It is found that the flow evolves in cycles consisting of periods of exponential growth and breakdowns of the jets. The resulting high-amplitude fluctuations of temperature is a feature potentially dangerous for operation of a reactor blanket. Financial support was provided by the US NSF (Grant CBET 1232851).

  15. Effect of reactor size on the breeding economics of LMFBR blankets

    Energy Technology Data Exchange (ETDEWEB)

    Tagishi, A.; Driscoll, M.J.

    1975-02-01

    The effect of reactor size on the neutronic and economic performance of LMFBR blankets driven by radially-power-flattened cores has been investigated using both simple models and state-of-the-art computer methods. Reactor power ratings in the range 250 to 3000 MW(e) were considered. Correlations for economic breakeven and optimum irradiation times and blanket thicknesses have been developed for batch-irradiated blankets. It is shown that a given distance from the core-blanket interface the fissile buildup rate per unit volume remains very nearly constant in the radial blanket as (radially-power-flattened, constant-height) core size increases. As a consequence, annual revenue per blanket assembly, and breakeven and optimum irradiation times and optimum blanket dimensions, are the same for all reactor sizes. It is also shown that the peripheral core fissile enrichment, hence neutron leakage spectra, of the (radially-power-flattened, constant-height) cores remains essentially constant as core size increases. Coupled with the preceding observations, this insures that radial blanket breeding performance in demonstration-size LMFBR units will be a good measure of that in much larger commercial LMFBR's.

  16. 77 FR 31004 - Southern Natural Gas Company; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2012-05-24

    ...] Southern Natural Gas Company; Notice of Request Under Blanket Authorization Take notice that on May 9, 2012, Southern Natural Gas Company (Southern), 569 Brookwood Village, Suite 501, Birmingham, Alabama 35209, filed... Commission's regulations under the Natural Gas Act (NGA), and Southern's blanket certificate issued in...

  17. 48 CFR 313.303-5 - Purchases under blanket purchase agreements.

    Science.gov (United States)

    2010-10-01

    ... 48 Federal Acquisition Regulations System 4 2010-10-01 2010-10-01 false Purchases under blanket purchase agreements. 313.303-5 Section 313.303-5 Federal Acquisition Regulations System HEALTH AND HUMAN... Methods 313.303-5 Purchases under blanket purchase agreements. (e)(5) HHS personnel that sign...

  18. 76 FR 58488 - Dominion Cove Point LNG, LP; Application for Blanket Authorization to Export Previously Imported...

    Science.gov (United States)

    2011-09-21

    ... Dominion Cove Point LNG, LP; Application for Blanket Authorization to Export Previously Imported Liquefied... (Application), filed on August 8, 2011, by Dominion Cove Point LNG, LP (DCP), requesting blanket authorization to export liquefied natural gas (LNG) that previously had been imported into the United States...

  19. 75 FR 60095 - Sempra LNG Marketing, LLC; Application for Blanket Authorization To Export Liquefied Natural Gas

    Science.gov (United States)

    2010-09-29

    ... LNG Marketing, LLC; Application for Blanket Authorization To Export Liquefied Natural Gas AGENCY..., by Sempra LNG Marketing, LLC (Sempra), requesting blanket authorization to export up to a total of 250 billion cubic feet (Bcf) of foreign sourced liquefied natural gas (LNG) for a two-year...

  20. 78 FR 35263 - Freeport LNG Development, L.P.; Application for Blanket Authorization To Export Previously...

    Science.gov (United States)

    2013-06-12

    ... Freeport LNG Development, L.P.; Application for Blanket Authorization To Export Previously Imported... receipt of an application (Application), filed on April 19, 2013, by Freeport LNG Development, L.P. (Freeport LNG), requesting blanket authorization to export liquefied natural gas (LNG) that previously...

  1. 77 FR 76013 - Sempra LNG Marketing, LLC; Application for Blanket Authorization To Export Previously Imported...

    Science.gov (United States)

    2012-12-26

    ... LNG Marketing, LLC; Application for Blanket Authorization To Export Previously Imported Liquefied... application (Application), filed on October 26, 2012, by Sempra LNG Marketing, LLC (Sempra LNG Marketing), requesting blanket authorization to export liquefied natural gas (LNG) that previously had been imported...

  2. 75 FR 38092 - The Dow Chemical Company; Application for Blanket Authorization To Export Liquefied Natural Gas

    Science.gov (United States)

    2010-07-01

    ... Chemical Company; Application for Blanket Authorization To Export Liquefied Natural Gas AGENCY: Office of... The Dow Chemical Company (Dow), requesting blanket authorization to export liquefied natural gas (LNG... equivalent of 390 billion cubic feet (Bcf) of natural gas on a short-term or spot market basis. The LNG...

  3. 75 FR 19954 - Cheniere Marketing, LLC; Application for Blanket Authorization To Export Liquefied Natural Gas

    Science.gov (United States)

    2010-04-16

    ... Cheniere Marketing, LLC; Application for Blanket Authorization To Export Liquefied Natural Gas AGENCY... Cheniere Marketing, LLC (CMI), requesting blanket authorization to export liquefied natural gas (LNG) that... 500 Billion cubic feet (Bcf) of natural gas on a short-term or spot market basis. The LNG would...

  4. 78 FR 4400 - Eni USA Gas Marketing LLC; Application for Blanket Authorization To Export Previously Imported...

    Science.gov (United States)

    2013-01-22

    ... USA Gas Marketing LLC; Application for Blanket Authorization To Export Previously Imported Liquefied... and order (Order No. 2923) that granted Eni USA Gas Marketing authority to export a cumulative total... Application, Eni USA Gas Marketing requests blanket authorization to export LNG from the Cameron Terminal...

  5. Needs and gaps in the development of aluminum-based corrosion and T-permeation barriers for DEMO blankets

    Energy Technology Data Exchange (ETDEWEB)

    Wulf, Sven-Erik, E-mail: sven-erik.wulf@kit.edu; Krauss, Wolfgang; Konys, Jürgen

    2015-10-15

    Highlights: • New processes for barriers based on electroplating introduced in the last years. • New processes ECA and ECX able to overcome former fabrication problems. • Scales by ECA showed long-term compatibility in flowing Pb–Li (>12,000 h). • Further fusion relevant characterization and optimization of scales is required. • Qualification of T-permeation properties is urgently needed. - Abstract: Low-activation-ferritic–martensitic (RAFM) steels are candidates for structural materials in different blanket designs foreseen for DEMO and partly for TBM's tested in ITER. In all designs the liquid breeder Pb–15.7Li is in direct contact with the structural material, and thus two major topics – corrosion and T-permeation – influence the reliable, safe and economical application of such combination of breeder and structural material. As bare RAFM steels exhibit high corrosion rates of up to 400 μm/h in flowing Pb–15.7Li, Al-based coatings made by different coating processes were developed during the last 15 years and showed promising results in protecting RAFM steels from corrosion and T-permeation reduction. Especially barriers made by HDA, and electroplating (ECA, ECX), proved their ability to protect Eurofer against corrosion in flowing Pb–15.7Li. However, available T-permeation data for coated RAFM steels are rare and partly ambiguous for these coatings. This paper summarizes the state-of-the-art of aluminum-based barrier development and points out gaps and needs in future scale characterization and T-permeation barrier development. Additionally, necessary qualification steps on the path toward a reliable fabrication route are presented that is required to produce aluminum-based corrosion and T-permeation barriers on RAFM steels for blanket applications in future fusion reactors like DEMO.

  6. Development of a full scale remote steerable ECRH mm-wave launching system test set-up for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Elzendoorn, B.S.Q. [FOM Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, Partner in the Trilateral Euregio Cluster, Nieuwegein, Edisonbaan 14, 3439 MN Nieuwegein (Netherlands)], E-mail: ben@rijnh.nl; Bongers, W.A.; Graswinckel, M.F. [FOM Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, Partner in the Trilateral Euregio Cluster, Nieuwegein, Edisonbaan 14, 3439 MN Nieuwegein (Netherlands); Jamar, J. [TNO Science and Industry, Business Unit Design and Manufacturing, Eindhoven, P.O. Box 6235, 5600 HE Eindhoven (Netherlands); Kruijt, O.G.; Lamers, B.; Ronden, D.M.S. [FOM Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, Partner in the Trilateral Euregio Cluster, Nieuwegein, Edisonbaan 14, 3439 MN Nieuwegein (Netherlands); Verhoeven, A.G.A

    2007-10-15

    FOM is developing a full-scale mock-up of the mm-wave upper-port launching system for electron cyclotron resonance heating in ITER. The test set-up foresees in most of the typical ITER operation conditions, e.g. primary vacuum and the secondary vacuum, blanket cooling and component cooling conditions. All relevant operational ITER conditions except neutron radiation can be simulated, including baking conditions. The fabrication of the launching system mock-up will be the last step in the development phase, which will be characterized by the fabrication of test samples and mm-wave tests and calculations. The parameterized modeling of mm-wave beam propagation in CATIA [D.M.S. Ronden, et al., Parameterized modeling of mm-wave beam propagation of the ITER ECRH remote steering upper port launcher in CATIA, this conference] (8th figure in this article) the test and calculation results will be used for the built to print design of the remote steering ECRH launcher for the ITER upper ports [A.G.A. Verhoeven, et al., Design of the Remote Steering ITER ECRH Upper Port Launcher, this conference].

  7. Vacuum hot-pressed beryllium and TiC dispersion strengthened tungsten alloy developments for ITER and future fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Xiang, E-mail: xliu@swip.ac.cn [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041, Sichuan (China); Chen, Jiming; Lian, Youyun; Wu, Jihong; Xu, Zengyu; Zhang, Nianman; Wang, Quanming; Duan, Xuro [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041, Sichuan (China); Wang, Zhanhong; Zhong, Jinming [Northwest Rare Metal Material Research Institute, CNMC, Ningxia Orient Group Co. Ltd.,No.119 Yejin Road, Shizuishan City, Ningxia,753000 (China)

    2013-11-15

    Beryllium and tungsten have been selected as the plasma facing materials of the ITER first wall (FW) and divertor chamber, respectively. China, as a participant in ITER, will share the manufacturing tasks of ITER first-wall mockups with the European Union and Russia. Therefore ITER-grade beryllium has been developed in China and a kind of vacuum hot-pressed (VHP) beryllium, CN-G01, was characterized for both physical, and thermo-mechanical properties and high heat flux performance, which indicated an equivalent performance to U.S. grade S-65C beryllium, a reference grade beryllium of ITER. Consequently CN-G01 beryllium has been accepted as the armor material of ITER-FW blankets. In addition, a modification of tungsten by TiC dispersion strengthening was investigated and a W–TiC alloy with TiC content of 0.1 wt.% has been developed. Both surface hardness and recrystallization measurements indicate its re-crystallization temperature approximately at 1773 K. Deuterium retention and thermal desorption behaviors of pure tungsten and the TiC alloy were also measured by deuterium ion irradiation of 1.7 keV energy to the fluence of 0.5–5 × 10{sup 18} D/cm{sup 2}; a main desorption peak at around 573 K was found and no significant difference was observed between pure tungsten and the tungsten alloy. Further characterization of the tungsten alloy is in progress.

  8. Chinese Tritium Technology of the Liquid Lithium-lead Alloy Experimental Loop for ITER%ITER中国液态锂铅实验回路中的氚技术

    Institute of Scientific and Technical Information of China (English)

    谢波; 吴宜灿; 翁葵平; 杨通在; 刘云怒; 宋勇; FDS团队

    2011-01-01

    The multinational cooperation in the International Thermonuclear Experimental Reactor (ITER) plan aims to show that fusion reactors are a new and viable way to address global energy concerns.The Chinese Dual Function Lithium Lead Testing Blanket Module (DFLL-TBM) is one of the major research programs and uses liquid lithium-lead as both breeder and cooler, and a helium-hydrogen gas bubbling method is used to extract tritium.So, tritium technology is a key issue in the liquid metal blanket.Based on the development strategy for Chinese liquid lithium-lead experimental blanket technology, development of Chinese tritium technology for liquid lithiumlead loops between 2004 and 2010 was elaberated in three fields, namely, theoretical analysis and calculation, experimental research, and engineering design.Some important information were introduced, such as the simulation-design-develop of liquid lithium-lead bubbler,tritium analysis and permeation barriers in the loops, tritium release from lithium-lead after irradiation, design of tritium extraction system for the blanket, etc.These works indicate that it is possible to completely overcome the difficulties involving very small solubility of tritium in the lithium-lead, accumulation of Magneto-Hydro-Dynamics (MHD) after a long period of continuous operation, materials corrosion together with the pressure drop caused by wall stress, and many technical problems, such as tritium retention, penetration,recovery and environment pollution, can be thoroughly solved.%多方合作的国际热核实验堆(ITER)计划是全球能源问题关注的一个重要进展标志,中国参与提出的双功能液态锂铅包层模块(DFLL-TBM)是一重要组成部分,采用液态锂铅合金作为氚增殖剂和冷却剂,氢-氦混合气鼓泡方式提取氚.因此,氚技术成为关系液态金属包层成败的关键问题之一.结合中国液态锂铅实验包层技术的发展战略,从理论分析与计算

  9. Wash resistance and repellent properties of Africa University mosquito blankets against mosquitoes

    Directory of Open Access Journals (Sweden)

    N. Lukwa

    2013-04-01

    Full Text Available The effect of permethrin-treated Africa University (AU mosquito blankets on susceptible female Anopheles gambiae sensu lato mosquitoes was studied under laboratory conditions at Africa University Campus in Mutare, Zimbabwe. Wash resistance (ability to retain an effective dose that kills ≥80% of mosquitoes after a number of washes and repellence (ability to prevent ≥80% of mosquito bites properties were studied. The AU blankets were wash resistant when 100% mortality was recorded up to 20 washes, declining to 90% after 25 washes. Untreated AU blankets did not cause any mortality on mosquitoes. However, mosquito repellence was 96%, 94%, 97.9%, 87%, 85% and 80.7% for treated AU blankets washed 0, 5, 10, 15, 20 and 25 times, respectively. Mosquito repellence was consistently above 80% from 0-25 washes. In conclusion, AU blankets washed 25 times were effective in repelling and killing An. gambiae sl mosquitoes under laboratory conditions.

  10. Evaluation of ITER MSE Viewing Optics

    Energy Technology Data Exchange (ETDEWEB)

    Allen, S; Lerner, S; Morris, K; Jayakumar, J; Holcomb, C; Makowski, M; Latkowski, J; Chipman, R

    2007-03-26

    formed an intermediate image that then was relayed out of the port plug with more ideal (dielectric) mirrors. Engineering models of the optics, port plug, and neutral beam geometry were also created, using the CATIA ITER models. Two video conference calls with the USIPO provided valuable design guidelines, such as the minimum distance of the first optic from the plasma. A second focus of the project was the calibration of the system. Several different techniques are proposed, both before and during plasma operation. Fixed and rotatable polarizers would be used to characterize the system in the no-plasma case. Obtaining the full modulation spectrum from the polarization analyzer allows measurement of polarization effects and also MHD plasma phenomena. Light from neutral beam interaction with deuterium gas (no plasma) has been found useful to determine the wavelength of each spatial channel. The status of the optical design for the edge (upper) and core (lower) systems is included in the following figure. Several issues should be addressed by a follow-on study, including whether the optical labyrinth has sufficient neutron shielding and a detailed polarization characterization of actual mirrors.

  11. Evaluating ITER remote handling middleware concepts

    Energy Technology Data Exchange (ETDEWEB)

    Koning, J.F., E-mail: j.f.koning@differ.nl [FOM Institute DIFFER, Association EURATOM-FOM, Partner in the Trilateral Euregio Cluster and ITER-NL, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Heemskerk, C.J.M.; Schoen, P.; Smedinga, D. [Heemskerk Innovative Technology, Noordwijk (Netherlands); Boode, A.H. [University of Applied Sciences InHolland, Alkmaar (Netherlands); Hamilton, D.T. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    Highlights: ► Remote Handling Study Centre: middleware system setup and modules built. ► Aligning to ITER RH Control System Layout: prototype of database, VR and simulator. ► OpenSplice DDS, ZeroC ICE messaging and object oriented middlewares reviewed. ► Windows network latency found problematic for semi-realtime control over the network. -- Abstract: Remote maintenance activities in ITER will be performed by a unique set of hardware systems, supported by an extensive software kit. A layer of middleware will manage and control a complex set of interconnections between teams of operators, hardware devices in various operating theatres, and databases managing tool and task logistics. The middleware is driven by constraints on amounts and timing of data like real-time control loops, camera images, and database access. The Remote Handling Study Centre (RHSC), located at FOM institute DIFFER, has a 4-operator work cell in an ITER relevant RH Control Room setup which connects to a virtual hot cell back-end. The centre is developing and testing flexible integration of the Control Room components, resulting in proof-of-concept tests of this middleware layer. SW components studied include generic human-machine interface software, a prototype of a RH operations management system, and a distributed virtual reality system supporting multi-screen, multi-actor, and multiple independent views. Real-time rigid body dynamics and contact interaction simulation software supports simulation of structural deformation, “augmented reality” operations and operator training. The paper presents generic requirements and conceptual design of middleware components and Operations Management System in the context of a RH Control Room work cell. The simulation software is analyzed for real-time performance and it is argued that it is critical for middleware to have complete control over the physical network to be able to guarantee bandwidth and latency to the components.

  12. Summary report for ITER Task-T19: MHD pressure drop and heat transfer study for liquid metal systems

    Science.gov (United States)

    Reed, Claude B.; Hua, Thanh Q.; Natesan, Ken; Kirillov, Igor R.; Vitkovski, Ivan V.; Anisimov, Aleksandr M.

    1995-03-01

    A key feasibility issue for the ITER Vanadium/Lithium breeding blanket is the question of insulator coatings. Design calculations show that an electrically insulating layer is necessary to maintain an acceptably low MHD pressure drop. To begin experimental investigations of the MHD performance of candidate insulator materials and the technology for putting them in place, a new test section was prepared. Aluminum oxide was chosen as the first candidate insulating material because it may be used in combination with NaK in the ITER vacuum vessel and/or the divertor. Details on the methods used to produce the aluminum oxide layer as well as the microstructures of the coating and the aluminide sublayer are presented and discussed. The overall MHD pressure drop, local MHD pressure gradient, local transverse MHD pressure difference, and surface voltage distributions in both the circumferential and the axial directions are reported and discussed. The positive results obtained here for high-temperature NaK have two beneficial implications for ITER. First, since NaK may be used in the vacuum vessel and/or the divertor, these results support the design approach of using electrically insulating coatings to substantially reduce MHD pressure drop. Secondly, while Al2O3/SS is not the same coating/base material combination which would be used in the advanced blanket, this work nonetheless shows that it is possible to produce a viable insulating coating which is stable in contact with a high temperature alkali metal coolant.

  13. Relaxation Criteria for Iterated Traffic Simulations

    Science.gov (United States)

    Kelly, Terence; Nagel, Kai

    Iterative transportation microsimulations adjust traveler route plans by iterating between a microsimulation and a route planner. At each iteration, the route planner adjusts individuals' route choices based on the preceding microsimulations. Empirically, this process yields good results, but it is usually unclear when to stop the iterative process when modeling real-world traffic. This paper investigates several criteria to judge relaxation of the iterative process, emphasizing criteria related to traveler decision-making.

  14. Thermal and structural design issues of breeding blankets for testing in the Next European Torus (NET)

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G.

    1988-05-01

    A review of the breeding blankets under study in Europe for testing in the Next European Torus is presented. In many concepts, the breeder modules are enclosed in boxes whose side walls in front of the plasma act as the first wall of the machine. Various types of breeder modules are investigated, involving both liquid and solid breeders, namely: - Pb-17Li liquid breeder concepts, the coolant being either water or Pb-17Li itself; - solid (ceramic) breeder concepts, the coolant being in all cases helium. The various ceramic concepts differ in the breeder/coolant arrangement (breeder-out-of-tube and breeder-in-tube), the orientation of the coolant tubes (poloidal or toroidal) and the breeder geometry (rods, plates or pebble bed). For each of these concepts the main design features are shown and the thermomechanical problems are discussed. The problems related to a coolant tube rupture are in many cases the most severe from the structural design point of view. The first wall box enclosing the breeder modules appears to be a weak secondary containment barrier. The liquid breeder-water cooled concept looks manageable from the thermal and structural design of point view. In the case of the self-cooled liquid breeder concept, the main problems are related to the magnetohydrodynamic effects. Solutions are envisaged to overcome these difficulties. In the case of ceramic breeders, the use of plates implies small dimensions in order to limit the thermal stresses and a poor exploitation of the permitted temperature operation window. Solutions involving rods associated with a multipass cooling scheme or pebble bed enable achievement of better thermomechanical conditions and, therefore, are preferred in the current investigations. However, they lead to design complications and require experimental verification which is in progress at the European laboratories.

  15. A Linear Iterative Unfolding Method

    CERN Document Server

    Laszlo, Andras

    2011-01-01

    A frequently faced task in experimental physics is to measure the probability distribution of some quantity. Often this quantity to be measured is smeared by a non-ideal detector response or by some physical process. The procedure of removing this smearing effect from the measured distribution is called unfolding, and is a delicate problem in signal processing. Due to the numerical ill-posedness of this task, various methods were invented which, given some assumptions on the initial probability distribution, try to regularize the problem. Most of these methods definitely introduce bias on the estimate of the initial probability distribution. We propose a linear iterative method (motivated by the Neumann series / Landweber iteration known in functional analysis), which has the advantage that no assumptions on the initial probability distribution is needed, and the only regularization parameter is the stopping order of the iteration. Convergence is proved under certain quite general conditions, which hold for p...

  16. Rollout Sampling Approximate Policy Iteration

    CERN Document Server

    Dimitrakakis, Christos

    2008-01-01

    Several researchers have recently investigated the connection between reinforcement learning and classification. We are motivated by proposals of approximate policy iteration schemes without value functions which focus on policy representation using classifiers and address policy learning as a supervised learning problem. This paper proposes variants of an improved policy iteration scheme which addresses the core sampling problem in evaluating a policy through simulation as a multi-armed bandit machine. The resulting algorithm offers comparable performance to the previous algorithm achieved, however, with significantly less computational effort. An order of magnitude improvement is demonstrated experimentally in two standard reinforcement learning domains: inverted pendulum and mountain-car.

  17. Heating performances of a IC in-blanket ring array

    Science.gov (United States)

    Bosia, G.; Ragona, R.

    2015-12-01

    An important limiting factor to the use of ICRF as candidate heating method in a commercial reactor is due to the evanescence of the fast wave in vacuum and in most of the SOL layer, imposing proximity of the launching structure to the plasma boundary and causing, at the highest power level, high RF standing and DC rectified voltages at the plasma periphery, with frequent voltage breakdowns and enhanced local wall loading. In a previous work [1] the concept for an Ion Cyclotron Heating & Current Drive array (and using a different wave guide technology, a Lower Hybrid array) based on the use of periodic ring structure, integrated in the reactor blanket first wall and operating at high input power and low power density, was introduced. Based on the above concept, the heating performance of such array operating on a commercial fusion reactor is estimated.

  18. Heating performances of a IC in-blanket ring array

    Energy Technology Data Exchange (ETDEWEB)

    Bosia, G., E-mail: gbosia@to.infn.it [Department of Physics, University of Turin (Italy); Ragona, R. [Laboratory for Plasma Physics-LPP-ERM/KMS, Brussels (Belgium)

    2015-12-10

    An important limiting factor to the use of ICRF as candidate heating method in a commercial reactor is due to the evanescence of the fast wave in vacuum and in most of the SOL layer, imposing proximity of the launching structure to the plasma boundary and causing, at the highest power level, high RF standing and DC rectified voltages at the plasma periphery, with frequent voltage breakdowns and enhanced local wall loading. In a previous work [1] the concept for an Ion Cyclotron Heating & Current Drive array (and using a different wave guide technology, a Lower Hybrid array) based on the use of periodic ring structure, integrated in the reactor blanket first wall and operating at high input power and low power density, was introduced. Based on the above concept, the heating performance of such array operating on a commercial fusion reactor is estimated.

  19. Experimental Investigation of Ternary Alloys for Fusion Breeding Blankets

    Energy Technology Data Exchange (ETDEWEB)

    Choi, B. William [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Chiu, Ing L. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-10-26

    Future fusion power plants based on the deuterium-tritium (DT) fuel cycle will be required to breed the T fuel via neutron reactions with lithium, which will be incorporated in a breeding blanket that surrounds the fusion source. Recent work by LLNL proposed the used of liquid Li as the breeder in an inertial fusion energy (IFE) power plant. Subsequently, an LDRD was initiated to develop alternatives ternary alloy liquid metal breeders that have reduced chemical reactivity with water and air compared to pure Li. Part of the work plan was to experimentally investigate the phase diagrams of ternary alloys. Of particular interest was measurement of the melt temperature, which must be low enough to be compatible with the temperature limits of the steel used in the construction of the chamber and heat transfer system.

  20. Fusion Blanket Coolant Section Criteria, Methodology, and Results

    Energy Technology Data Exchange (ETDEWEB)

    DeMuth, J. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Meier, W. R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Jolodosky, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Frantoni, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Reyes, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-10-02

    The focus of this LDRD was to explore potential Li alloys that would meet the tritium breeding and blanket cooling requirements but with reduced chemical reactivity, while maintaining the other attractive features of pure Li breeder/coolant. In other fusion approaches (magnetic fusion energy or MFE), 17Li- 83Pb alloy is used leveraging Pb’s ability to maintain high TBR while lowering the levels of lithium in the system. Unfortunately this alloy has a number of potential draw-backs. Due to the high Pb content, this alloy suffers from very high average density, low tritium solubility, low system energy, and produces undesirable activation products in particular polonium. The criteria considered in the selection of a tritium breeding alloy are described in the following section.

  1. Cosmetic wastewater treatment by upflow anaerobic sludge blanket reactor

    Energy Technology Data Exchange (ETDEWEB)

    Puyol, D.; Monsalvo, V.M.; Mohedano, A.F. [Seccion de Ingenieria Quimica, Facultad de Ciencias, Universidad Autonoma de Madrid, C/ Francisco Tomas y Valiente 7, 28049, Madrid (Spain); Sanz, J.L. [Departamento de Biologia Molecular, Facultad de Ciencias, Universidad Autonoma de Madrid, C/ Francisco Tomas y Valiente 7, 28049, Madrid (Spain); Rodriguez, J.J., E-mail: juanjo.rodriguez@uam.es [Seccion de Ingenieria Quimica, Facultad de Ciencias, Universidad Autonoma de Madrid, C/ Francisco Tomas y Valiente 7, 28049, Madrid (Spain)

    2011-01-30

    Anaerobic treatment of pre-settled cosmetic wastewater in batch and continuous experiments has been investigated. Biodegradability tests showed high COD and solid removal efficiencies (about 70%), being the hydrolysis of solids the limiting step of the process. Continuous treatment was carried out in an upflow anaerobic sludge blanket reactor. High COD and TSS removal efficiencies (up to 95% and 85%, respectively) were achieved over a wide range of organic load rate (from 1.8 to 9.2 g TCOD L{sup -1} day{sup -1}). Methanogenesis inhibition was observed in batch assays, which can be predicted by means of a Haldane-based inhibition model. Both COD and solid removal were modelled by Monod and pseudo-first order models, respectively.

  2. Physical Model Development and Benchmarking for MHD Flows in Blanket Design

    Energy Technology Data Exchange (ETDEWEB)

    Ramakanth Munipalli; P.-Y.Huang; C.Chandler; C.Rowell; M.-J.Ni; N.Morley; S.Smolentsev; M.Abdou

    2008-06-05

    An advanced simulation environment to model incompressible MHD flows relevant to blanket conditions in fusion reactors has been developed at HyPerComp in research collaboration with TEXCEL. The goals of this phase-II project are two-fold: The first is the incorporation of crucial physical phenomena such as induced magnetic field modeling, and extending the capabilities beyond fluid flow prediction to model heat transfer with natural convection and mass transfer including tritium transport and permeation. The second is the design of a sequence of benchmark tests to establish code competence for several classes of physical phenomena in isolation as well as in select (termed here as “canonical”,) combinations. No previous attempts to develop such a comprehensive MHD modeling capability exist in the literature, and this study represents essentially uncharted territory. During the course of this Phase-II project, a significant breakthrough was achieved in modeling liquid metal flows at high Hartmann numbers. We developed a unique mathematical technique to accurately compute the fluid flow in complex geometries at extremely high Hartmann numbers (10,000 and greater), thus extending the state of the art of liquid metal MHD modeling relevant to fusion reactors at the present time. These developments have been published in noted international journals. A sequence of theoretical and experimental results was used to verify and validate the results obtained. The code was applied to a complete DCLL module simulation study with promising results.

  3. Synthesis and Characterization of Fibre Reinforced Silica Aerogel Blankets for Thermal Protection

    Directory of Open Access Journals (Sweden)

    S. Chakraborty

    2016-01-01

    Full Text Available Using tetraethoxysilane (TEOS as the source of silica, fibre reinforced silica aerogels were synthesized via fast ambient pressure drying using methanol (MeOH, trimethylchlorosilane (TMCS, ammonium fluoride (NH4F, and hexane. The molar ratio of TEOS/MeOH/(COOH2/NH4F was kept constant at 1 : 38 : 3.73 × 10−5 : 0.023 and the gel was allowed to form inside the highly porous meta-aramid fibrous batting. The wet gel surface was chemically modified (silylation process using various concentrations of TMCS in hexane in the range of 1 to 20% by volume. The fibre reinforced silica aerogel blanket was obtained subsequently through atmospheric pressure drying. The aerogel blanket samples were characterized by density, thermal conductivity, hydrophobicity (contact angle, and Scanning Electron Microscopy. The radiant heat resistance of the aerogel blankets was examined and compared with nonaerogel blankets. It has been observed that, compared to the ordinary nonaerogel blankets, the aerogel blankets showed a 58% increase in the estimated burn injury time and thus ensure a much better protection from heat and fire hazards. The effect of varying the concentration of TMCS on the estimated protection time has been examined. The improved thermal stability and the superior thermal insulation of the flexible aerogel blankets lead to applications being used for occupations that involve exposure to hazards of thermal radiation.

  4. Cooperation between CERN and ITER

    CERN Multimedia

    CERN Audiovisual Service

    2008-01-01

    CERN and the International Fusion Organisation ITER have just signed a first cooperation agreeement. The Director-General of the International Fusion Energy Organization, Mr Kaname Ikeda, and CERN Director-General, Robert Aymar, signed a cooperation agreement at a meeting on the Meyrin site on Thursday 6 March.

  5. Bounded Fixed-Point Iteration

    DEFF Research Database (Denmark)

    Nielson, Hanne Riis; Nielson, Flemming

    1992-01-01

    they obtain a quadratic bound. These bounds are shown to be tight. Specializing the case of strict and additive functions to functionals of a form that would correspond to iterative programs they show that a linear bound is tight. This is related to several analyses studied in the literature (including...

  6. Iterative Specialisation of Horn Clauses

    DEFF Research Database (Denmark)

    Nielsen, Christoffer Rosenkilde; Nielson, Flemming; Nielson, Hanne Riis

    2008-01-01

    We present a generic algorithm for solving Horn clauses through iterative specialisation. The algorithm is generic in the sense that it can be instantiated with any decidable fragment of Horn clauses, resulting in a solution scheme for general Horn clauses that guarantees soundness and terminatio...

  7. Iterative method for interferogram processing

    Science.gov (United States)

    Kotlyar, Victor V.; Seraphimovich, P. G.; Zalyalov, Oleg K.

    1994-12-01

    We have developed and numerically evaluated an iterative algorithm for interferogram processing including the Fourier-transform method, the Gerchberg-Papoulis algorithm and Wiener's filter-based regularization used in combination. Using a signal-to-noise ratio not less than 1, it has been possible to reconstruct the phase of an object field with accuracy better than 5%.

  8. Preliminary Analysis of Liquid Metal MHD Pressure Drop in the Blanket for the FDS

    Institute of Scientific and Technical Information of China (English)

    王红艳; 吴宜灿; 何晓雄

    2002-01-01

    Preliminary analysis and calculation of liquid metal Li17Pb83 magnetohydrodynamic (MHD) pressure drop in the blanket for the FDS have been presented to evaluate the significance of MHD effects on the thermal-hydraulic design of the blanket. To decrease the liquid metal MHD pressure drop, Al2O3 is applied as an electronically insulated coating onto the inner surface of the ducts. The requirement for the insulated coating to reduce the additional leakage pressure drop caused by coating imperfections has been analyzed. Finally, the total liquid metal MHD pressure drop and magnetic pump power in the FDS blanket have been given.

  9. Qualification of the Joints for the ITER Central Solenoid

    Energy Technology Data Exchange (ETDEWEB)

    Martovetsky, N; Berryhill, A; Kenney, S

    2011-09-01

    The ITER Central Solenoid has 36 interpancake joints, 12 bus joints, and 12 feeder joints in the magnet. The joints are required to have resistance below 4 nOhm at 45 kA at 4.5 K. The US ITER Project Office developed two different types of interpancake joints with some variations in details in order to find a better design, qualify the joints, and establish a fabrication process. We built and tested four samples of the sintered joints and two samples with butt-bonded joints (a total of eight joints). Both designs met the specifications. Results of the joint development, test results, and selection of the baseline design are presented and discussed in the paper. The ITER Central Solenoid (CS) consists of six modules. Each module is composed of six wound hexapancakes and one quadrapancake. The multipancakes are connected electrically and hydraulically by in-line interpancake joints. The joints are located at the outside diameter (OD) of the module. Cable in conduit conductor (CICC) high-current joints are critical elements in the CICC magnets. In addition to low resistivity, the CS joints must fit a space envelope equivalent to the regular conductor cross section and must have low hydraulic impedance and enough structural strength to withstand the hoop and compressive forces during operation, including cycling. This paper is the continuation of the work reported on the intermodule joints.

  10. In plain sight: the Chesapeake Bay crater ejecta blanket

    Directory of Open Access Journals (Sweden)

    D. L. Griscom

    2012-02-01

    Full Text Available The discovery nearly two decades ago of a 90 km-diameter impact crater below the lower Chesapeake Bay has gone unnoted by the general public because to date all published literature on the subject has described it as "buried". To the contrary, evidence is presented here that the so-called "upland deposits" that blanket ∼5000 km2 of the U.S. Middle-Atlantic Coastal Plain (M-ACP display morphologic, lithologic, and stratigraphic features consistent with their being ejecta from the 35.4 Ma Chesapeake Bay Impact Structure (CBIS and absolutely inconsistent with the prevailing belief that they are of fluvial origin. Specifically supporting impact origin are the facts that (i a 95 %-pure iron ore endemic to the upland deposits of southern Maryland, eastern Virginia, and the District of Columbia has previously been proven to be impactoclastic in origin, (ii this iron ore welds together a small percentage of well-rounded quartzite pebbles and cobbles of the upland deposits into brittle sheets interpretable as "spall plates" created in the interference-zone of the CBIS impact, (iii the predominantly non-welded upland gravels have long ago been shown to be size sorted with an extreme crater-centric gradient far too large to have been the work of rivers, but well explained as atmospheric size-sorted interference-zone ejecta, (iv new evidence is provided here that ~60 % of the non-welded quartzite pebbles and cobbles of the (lower lying gravel member of the upland deposits display planar fractures attributable to interference-zone tensile waves, (v the (overlying loam member of the upland deposits is attributable to base-surge-type deposition, (vi several exotic clasts found in a debris flow topographically below the upland deposits can only be explained as jetting-phase crater ejecta, and (vii an allogenic granite boulder found among the upland deposits is deduced to have been launched into space and sculpted by hypervelocity air friction

  11. ITERATIVE ALGORITHMS FOR DATA ASSIMILATION PROBLEMS

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    Iterative algorithms for solving the data assimilation problems are considered, based on the main and adjoint equations. Spectral properties of the control operators of the problem are studied, the iterative algorithms are justified.

  12. Facilities, testing program and modeling needs for studying liquid metal magnetohydrodynamic flows in fusion blankets

    Energy Technology Data Exchange (ETDEWEB)

    Bühler, L., E-mail: leo.buehler@kit.edu [Karlsruhe Institute of Technology (KIT), Postfach 3640, 76021 Karlsruhe (Germany); Mistrangelo, C.; Konys, J. [Karlsruhe Institute of Technology (KIT), Postfach 3640, 76021 Karlsruhe (Germany); Bhattacharyay, R. [Institute for Plasma Research, Gandhinagar, Gujarat 382428 (India); Huang, Q. [Institute of Nuclear Energy Safety Technology (INEST), Chinese Academy of Sciences (CAS) (China); Obukhov, D. [D.V. Efremov Scientific Research Institute of Electrophysical Apparatus (NIIEFA) (Russian Federation); Smolentsev, S. [University of California Los Angeles (UCLA) (United States); Utili, M. [ENEA C.R. Brasimone, Camugnano 40032 (Italy)

    2015-11-15

    Since many years, liquid metal flows for applications in fusion blankets have been investigated worldwide. A review is given about modeling requirements and existing experimental facilities for investigations of liquid metal related issues in blankets with the focus on magnetohydrodynamics (MHD). Most of the performed theoretical and experimental works were dedicated to fundamental aspects of MHD flows under very strong magnetic fields as they may occur in generic elements of fusion blankets like pipes, ducts, bends, expansions and contractions. Those experiments are required to progressively validate numerical tools with the purpose of obtaining codes capable to predict MHD flows at fusion relevant parameters in complex blanket geometries, taking into account electrical and thermal coupling between fluid and structural materials. Scaled mock-up experiments support the theoretical activities and help deriving engineering correlations for cases which cannot be calculated with required accuracy up to now.

  13. Normal Operation (NO) of APT Blanket System and its Components Based on Initial Conceptual Design

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports documenting accident scenario simulations for the Accelerator Production of Tritium (APT) blanket heat removal systems. The simulations were performed in support of the Preliminary Safety Analysis Report (PSAR) for the APT.

  14. Properties of Ejecta Blanket Deposits Surrounding Morasko Meteorite Impact Craters (Poland)

    Science.gov (United States)

    Szokaluk, M.; Muszyński, A.; Jagodziński, R.; Szczuciński, W.

    2016-08-01

    Morasko impact craters are a record of the fall of a meteorite into the soft sediments. The presented results illustrate the geological structure of the area around the crater as well as providing evidence of the occurrence of ejecta blanket.

  15. Acoustic contributions of a sound absorbing blanket placed in a double panel structure: Absorption Versus Transmission

    CERN Document Server

    Doutres, Olivier; 10.1121/1.3458845

    2010-01-01

    The objective of this paper is to propose a simple tool to estimate the absorption vs. transmission loss contributions of a multilayered blanket unbounded in a double panel structure and thus guide its optimization. The normal incidence airborne sound transmission loss of the double panel structure, without structure-borne connections, is written in terms of three main contributions; (i) sound transmission loss of the panels, (ii) sound transmission loss of the blanket and (iii) sound absorption due to multiple reflections inside the cavity. The method is applied to four different blankets frequently used in automotive and aeronautic applications: a non-symmetric multilayer made of a screen in sandwich between two porous layers and three symmetric porous layers having different pore geometries. It is shown that the absorption behavior of the blanket controls the acoustic behavior of the treatment at low and medium frequencies and its transmission loss at high frequencies. Acoustic treatment having poor sound ...

  16. Axial Neutron Flux Evaluation in a Tokamak System: a Possible Transmutation Blanket Position for a Fusion-Fission Transmutation System

    Science.gov (United States)

    Velasquez, Carlos E.; de P. Barros, Graiciany; Pereira, Claubia; Fortini Veloso, Maria A.; Costa, Antonella L.

    2012-08-01

    A sub-critical advanced reactor based on Tokamak technology with a D-T fusion neutron source is an innovative type of nuclear system. Due to the large number of neutrons produced by fusion reactions, such a system could be useful in the transmutation process of transuranic elements (Pu and minor actinides (MAs)). However, to enhance the MA transmutation efficiency, it is necessary to have a large neutron wall loading (high neutron fluence) with a broad energy spectrum in the fast neutron energy region. Therefore, it is necessary to know and define the neutron fluence along the radial axis and its characteristics. In this work, the neutron flux and the interaction frequency along the radial axis are evaluated for various materials used to build the first wall. W alloy, beryllium, and the combination of both were studied, and the regions more suitable to transmutation were determined. The results demonstrated that the best zone in which to place a transmutation blanket is limited by the heat sink and the shield block. Material arrangements of W alloy/W alloy and W alloy/beryllium would be able to meet the requirements of the high fluence and hard spectrum that are needed for transuranic transmutation. The system was simulated using the MCNP code, data from the ITER Final Design Report, 2001, and the Fusion Evaluated Nuclear Data Library/MC-2.1 nuclear data library.

  17. Climate-driven expansion of blanket bogs in Britain during the Holocene

    Directory of Open Access Journals (Sweden)

    A. V. Gallego-Sala

    2015-10-01

    Full Text Available Blanket bog occupies approximately 6 % of the area of the UK today. The Holocene expansion of this hyperoceanic biome has previously been explained as a consequence of Neolithic forest clearance. However, the present distribution of blanket bog in Great Britain can be predicted accurately with a simple model (PeatStash based on summer temperature and moisture index thresholds, and the same model correctly predicts the highly disjunct distribution of blanket bog worldwide. This finding suggests that climate, rather than land-use history, controls blanket-bog distribution in the UK and everywhere else. We set out to test this hypothesis for blanket bogs in the UK using bioclimate envelope modelling compared with a database of peat initiation age estimates. We used both pollen-based reconstructions and climate model simulations of climate changes between the mid-Holocene (6000 yr BP, 6 ka and modern climate to drive PeatStash and predict areas of blanket bog. We compiled data on the timing of blanket-bog initiation, based on 228 age determinations at sites where peat directly overlies mineral soil. The model predicts large areas of northern Britain would have had blanket bog by 6000 yr BP, and the area suitable for peat growth extended to the south after this time. A similar pattern is shown by the basal peat ages and new blanket bog appeared over a larger area during the late Holocene, the greatest expansion being in Ireland, Wales and southwest England, as the model predicts. The expansion was driven by a summer cooling of about 2 °C, shown by both pollen-based reconstructions and climate models. The data show early Holocene (pre-Neolithic blanket-bog initiation at over half of the sites in the core areas of Scotland, and northern England. The temporal patterns and concurrence of the bioclimate model predictions and initiation data suggest that climate change provides a parsimonious explanation for the early Holocene distribution and later

  18. Resolution of proliferation issues for a SFR blanket with a specific application

    Energy Technology Data Exchange (ETDEWEB)

    Stauff, N.E. [31 rue baudelaire, voisins le bretonneux, 78960 (France); Massachusetts Institute of Technology, 77 Massachusetts Ave, Cambridge, MA 02139 (United States); Forget, B.; Driscoll, M.J. [Massachusetts Institute of Technology, 77 Massachusetts Ave, Cambridge, MA 02139 (United States)

    2009-06-15

    The Sodium Fast Reactor is seen as the most realistic Gen-IV reactor to be built in the near future. France and the US are still developing their designs; these will require improved safety, competitive economics, and also proliferation resistance. To meet this last requirement, both French and American designers show some concerns with the use of breeding blankets. France and the USA won't need breeding blankets to produce plutonium because they already have large amounts of plutonium bred from their LWR fleet to start a new SFR fleet, thus breeding blankets are mainly of interest for minor actinide burning. On the contrary, India and China express great interest in blankets for their SFR designs, to reach a positive breeding gain. For example, the Indian PFBR, a 500 MWe oxide-fueled SFR has a breeding ratio of 1.05. Blankets are used in a Fast Reactor to increase the breeding ratio of the core, by breeding a significant amount of plutonium. The Plutonium bred within these blankets, if these are loaded with Uranium only, is generally of a very high quality, which makes it easily used in a nuclear explosive device. Our research has shown that the plutonium in breeding blankets can be made less attractive to make a nuclear explosive device than LWR-bred plutonium with a burnup of 50 MWd/Kg. Minor actinide doping and moderator addition were the two options studied, as they increase Pu{sup 238} and Pu{sup 240} production. In the work reported here, the methodology developed for securing a breeding blanket was successfully applied to the Indian PFBR. The full paper will describe a design of the PFBR breeding proliferation resistant plutonium within its blankets. The blankets were rendered secure by adding a zirconium hydride moderator and a small volume of MAs. It was demonstrated that reducing the attractiveness of the blanket plutonium would require no external MA dependency by choosing an adequate fuel cycle. The characteristics and performance of this design

  19. Iterative Channel-Tracking Techniques for 5.9 GHz DSRC Applications

    Directory of Open Access Journals (Sweden)

    Harb Abdulhamid

    2008-01-01

    Full Text Available This letter proposes a novel channel-tracking scheme to improve the performance of the dedicated short-range communication (DSRC systems affected by rapid fluctuations in channel envelopes. The proposed technique is called “iterative (turbo compensation.” It utilizes additional information extracted from the receivers output to further improve the accuracy of the channel estimation. Simulation results show that the iterative scheme performs better than noniterative techniques in higher constellation modulations at high vehicle speeds.

  20. Non-iterative phase hologram computation for low speckle holographic image projection

    OpenAIRE

    Ürey, Hakan; Ulusoy, Erdem; Mengü, Deniz

    2016-01-01

    Phase-only spatial light modulators (SLMs) are widely used in holographic display applications, including holographic image projection (HIP). Most phase computer generated hologram (CGH) calculation algorithms have an iterative structure with a high computational load, and also are prone to speckle noise, as a result of the random phase terms applied on the desired images to mitigate the encoding noise. In this paper, we present a non-iterative algorithm, where simple Discrete Fourier Transfo...

  1. On One-Point Iterations and DIIS

    DEFF Research Database (Denmark)

    Østerby, Ole; Sørensen, Hans Henrik Brandenborg

    2009-01-01

    We analyze various iteration procedures in many dimensions inspired by the SCF iteration used in first principles electronic structure calculations. We show that the simple mixing of densities can turn a divergent (or slowly convergent) iteration into a (faster) convergent process provided all th...

  2. REMARK ON STABILITY OF ISHIKAWA ITERATIVE PROCEDURES

    Institute of Scientific and Technical Information of China (English)

    薛志群; 田虹

    2002-01-01

    The stability of the Ishikawa iteration procedures was studied for one class ofcontinuity strong pseudocontraction and continuity strongly accretive operators with boundedrange in real uniformly smooth Banach space. Under parameters satisfying certainconditions, the convergence of iterative sequences was proved. The results improve andextend the recent corresponding results, and supply the basis of theory for further discussingconvergence of iteration procedures with errors.

  3. Wash resistance and repellent properties of Africa University mosquito blankets against mosquitoes

    OpenAIRE

    N. Lukwa; A. Makuwaza; T. Chiwade; Mutambu, S L; M. Zimba; P. Munosiyei

    2013-01-01

    The effect of permethrin-treated Africa University (AU) mosquito blankets on susceptible female Anopheles gambiae sensu lato mosquitoes was studied under laboratory conditions at Africa University Campus in Mutare, Zimbabwe. Wash resistance (ability to retain an effective dose that kills ≥80% of mosquitoes after a number of washes) and repellence (ability to prevent ≥80% of mosquito bites) properties were studied. The AU blankets were wash resistant when 100% mortality was recorded up t...

  4. ITER LHCD plans and design

    Energy Technology Data Exchange (ETDEWEB)

    Bibet, Ph.; Beaumont, B.; Delpech, L.; Ekedahl, A.; Kazarian, F.; Litaudon, X.; Prou, M. [CEA Cadarache, Dept. de Recherches sur la Fusion Controlee (DRFC), 13 - Saint-Paul-lez-Durance (France); Belo, J.H.; Bizarro, J.P.S. [Centro de Fusao Nuclear, Associacao Euratom-IST, Instituto Superior Tecnico, Lisboa (Portugal); Granucci, G. [Associazione EURATOM-ENEA sulla Fusione, Milano (Italy); Kuzikov, S. [Institute of Applied Physics, Nizhny Novgorod (Russian Federation); Mailloux, J. [Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon (United Kingdom); Mirizzi, F.; Pericoli, V.; Tuccillo, A.A. [Association Euratom-ENEA sulla Fusione, Centro Ricerche Energia Frascati (Italy); Rantamaki, K. [Association Euratom-Tekes, VTT (Finland)

    2005-07-01

    LH waves experimentally exhibit the highest Current Drive efficiency at low plasma temperature, therefore they are the most suitable candidates for controlling the current profile in the off axis part of ITER Steady State plasmas. For this purpose, a 5 GHz, 20 MW CW LH system has been designed, that relies on a generator made of 24 klystrons, 1 MW each, 60 metres long circular oversized transmission lines, one antenna, based on the Passive Active Multi-function (PAM) concept. High reliability of the launcher is achieved, by limiting the power density to 33 MW/m{sup 2}. Together with the overall system description, the present results achieved toward ITER are presented. The different ongoing project are listed. The remaining outstanding problems are depicted. (authors)

  5. An Iterative Rejection Sampling Method

    CERN Document Server

    Sherstnev, A

    2008-01-01

    In the note we consider an iterative generalisation of the rejection sampling method. In high energy physics, this sampling is frequently used for event generation, i.e. preparation of phase space points distributed according to a matrix element squared $|M|^2$ for a scattering process. In many realistic cases $|M|^2$ is a complicated multi-dimensional function, so, the standard von Neumann procedure has quite low efficiency, even if an error reducing technique, like VEGAS, is applied. As a result of that, many of the $|M|^2$ calculations go to ``waste''. The considered iterative modification of the procedure can extract more ``unweighted'' events, i.e. distributed according to $|M|^2$. In several simple examples we show practical benefits of the technique and obtain more events than the standard von Neumann method, without any extra calculations of $|M|^2$.

  6. Truncated States Obtained by Iteration

    Institute of Scientific and Technical Information of China (English)

    W.B.Cardoso; N.G.de Almeida

    2008-01-01

    We introduce the concept of truncated states obtained via iterative processes(TSI)and study its statistical features,making an analogy with dynamical systems theory(DST).As a specific example,we have studied TSI for the doubring and the logistic functions,which are standard functions in studying chaos.TSI for both the doubling and logistic functions exhibit certain similar patterns when their statistical features are compared from the point of view of DST.

  7. Iterative Goal Refinement for Robotics

    Science.gov (United States)

    2014-06-01

    Iterative Goal Refinement for Robotics Mark Roberts1, Swaroop Vattam1, Ronald Alford2, Bryan Auslander3, Justin Karneeb3, Matthew Molineaux3... robotics researchers and practitioners. We present a goal lifecycle and define a formal model for GR that (1) relates distinct disciplines concerning...researchers to collaborate in exploring this exciting frontier. 1. Introduction Robotic systems often act using incomplete models in environments

  8. Design of the Remote Steerable ECRH launching system for the ITER upper ports

    Energy Technology Data Exchange (ETDEWEB)

    Verhoeven, A G A [FOM-Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, Trilateral Euregio Cluster, Nieuwegein (Netherlands); Elzendoorn, B S Q [FOM-Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, Trilateral Euregio Cluster, Nieuwegein (Netherlands); Bongers, W A [FOM-Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, Trilateral Euregio Cluster, Nieuwegein (Netherlands)] (and others)

    2005-01-01

    An ECRH (electron-cyclotron resonance heating) launching system for the ITER upper ports is being designed. The aim of the system is to inject Electron Cyclotron Waves (ECW) in the ITER plasma in order to stabilize neoclassical tearing modes (NTM). Each of the four upper-port launchers consists of six mm-wave lines capable of transmitting high power up to 2 MW per line at 170 GHz. In order to exploit the capability of ECW for localized heating and current drive over a range of plasma radii in ITER, the ECH and CD upper port launcher must have a beam steering capability. The Remote Steering (RS) principle has great advantages, because it enables to avoid steerable mirrors with flexible cooling lines at the plasma-facing end of the launcher. The principle consists of a long, corrugated, square waveguide having the steerable optics placed outside of the first confinement boundary of the vacuum vessel. All vulnerable components are far away from the hostile plasma environment. Furthermore, the RS launching system enables to do maintenance on the system during shutdown, without affecting the torus vacuum and the blanket cooling circuits.

  9. Liquid immersion blanket design for use in a compact modular fusion reactor

    Science.gov (United States)

    Sorbom, Brandon; Ball, Justin; Barnard, Harold; Haakonsen, Christian; Hartwig, Zachary; Olynyk, Geoffrey; Sierchio, Jennifer; Whyte, Dennis

    2012-10-01

    Traditional tritium breeding blankets in fusion reactor designs include a large amount of structural material. This results in complex engineering requirements, complicated sector maintenance, and marginal tritium breeding ratios (TBR). We present a conceptual design of a fully liquid blanket. To maximize tritium breeding volume, the vacuum vessel is completely immersed in a continuously recycled FLiBe blanket, with the exception of small support posts. FLiBe has a wide liquid temperature window (459 C to 1430 C), low electrical conductivity to minimize MHD effects, similar thermal/fluid characteristics to water, and is chemically inert. While tritium breeding with FLiBe in traditional blankets is poor, we use MCNP neutronics analysis to show that the immersion blanket design coupled with a beryllium neutron multiplier results in TBR > 1. FLiBe is shown to be a sufficient radiation shield for the toroidal field magnets and can be used as a coolant for the vacuum vessel and divertor, allowing for a simplified single-phase, low-pressure, single-fluid cooling scheme. When coupled with a high-field compact reactor design, the immersion blanket eliminates the need for complex sector maintenance, allows the vacuum vessel to be a replaceable component, and reduces financial cost.

  10. Multiplier, moderator, and reflector materials for lithium-vanadium fusion blankets.

    Energy Technology Data Exchange (ETDEWEB)

    Gohar, Y.; Smith, D. L.

    1999-10-07

    The self-cooled lithium-vanadium fusion blanket concept has several attractive operational and environmental features. In this concept, liquid lithium works as the tritium breeder and coolant to alleviate issues of coolant breeder compatibility and reactivity. Vanadium alloy (V-4Cr-4Ti) is used as the structural material because of its superior performance relative to other alloys for this application. However, this concept has poor attenuation characteristics and energy multiplication for the DT neutrons. An advanced self-cooled lithium-vanadium fusion blanket concept has been developed to eliminate these drawbacks while maintaining all the attractive features of the conventional concept. An electrical insulator coating for the coolant channels, spectral shifter (multiplier, and moderator) and reflector were utilized in the blanket design to enhance the blanket performance. In addition, the blanket was designed to have the capability to operate at high loading conditions of 2 MW/m{sup 2} surface heat flux and 10 MW/m{sup 2} neutron wall loading. This paper assesses the spectral shifter and the reflector materials and it defines the technological requirements of this advanced blanket concept.

  11. The ITER EC H and CD upper launcher: Structural system

    Energy Technology Data Exchange (ETDEWEB)

    Spaeh, Peter, E-mail: peter.spaeh@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Aiello, Gaetano; Gessner, Robby; Grossetti, Giovanni; Meier, Andreas; Scherer, Theo; Schreck, Sabine [Karlsruhe Institute of Technology, Institute for Applied Materials, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Serikov, Arkady [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Strauss, Dirk; Vaccaro, Alessandro [Karlsruhe Institute of Technology, Institute for Applied Materials, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Weinhorst, Bastian [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany)

    2013-10-15

    Highlights: ► ITER EC H and CD Launcher goes towards final design. ► European consortium founded for bringing the EC H and CD launcher towards final design. ► New design updates for dedicated components of the ITER EC H and CD launcher. -- Abstract: After approval of the preliminary design of the ITER EC H and CD Upper Launcher, ECHUL-CA, a consortium of several European research institutes, was founded to pool resources for approaching the final design. At the end of 2011 the consortium has signed a 2 years contract with F4E to go ahead with the work on the launcher. The contract deals with design work on both the port plug, forming the structural system, and the mm-wave-system, which injects the RF-power into the plasma. Within the period of this contract all components being part of the Tritium confinement, of which the closure plate, the support flange, the diamond windows and the waveguide feed-throughs are the most outstanding ones, will get the status of the final design. Important steps to be done for the structural system are the optimization of the mechanical behavior of the launcher, leading to minimum deflections of the port plug during plasma disruptions and optimum seismic resistance. To reduce the effect of halo currents it was decided to recess the first wall of 100 mm compared to the regular blanket tangent. This recess requires substantial changes of the cooling system and the thermo-hydraulic design of the launcher. Also the layout of the shielding arrangement and the integration of the mm-wave system need significant revision. Moreover manufacturing aspects and enhanced remote handling capability are taken into account. For the final design also quality aspects must be considered; thus the design is elaborated with respect to applicable codes and standards, material specifications, risk analyses and the RAMI (reliability, availability, maintainability and inspectability) analysis to guarantee maximum performance of the device. This paper

  12. Vacuum Permeator Analysis for Extraction of Tritium from DCLL Blankets

    Energy Technology Data Exchange (ETDEWEB)

    Humrickhouse, Paul Weston [Idaho National Laboratory; Merrill, Brad Johnson [Idaho National Laboratory

    2014-11-01

    It is envisioned that tritium will be extracted from DCLL blankets using a vacuum permeator. We derive here an analytical solution for the extraction efficiency of a permeator tube, which is a function of only two dimensionless numbers: one that indicates whether radial transport is limited in the PbLi or in the solid membrane, and another that is the ratio of axial and radial transport times in the PbLi. The permeator efficiency is maximized by decreasing the velocity and tube diameter, and increasing the tube length. This is true regardless of the mass transport correlation used; we review several here and find that they differ little, and the choice of correlation is not a source of significant uncertainty here. The PbLi solubility, on the other hand, is a large source of uncertainty, and we identify upper and lower bounds from the literature data. Under the most optimistic assumptions, we find that a ferritic steel permeator operating at 550 °C will need to be at least an order of magnitude larger in volume than previous conceptual designs using niobium and operating at higher temperatures.

  13. Iterative methods for mixed finite element equations

    Science.gov (United States)

    Nakazawa, S.; Nagtegaal, J. C.; Zienkiewicz, O. C.

    1985-01-01

    Iterative strategies for the solution of indefinite system of equations arising from the mixed finite element method are investigated in this paper with application to linear and nonlinear problems in solid and structural mechanics. The augmented Hu-Washizu form is derived, which is then utilized to construct a family of iterative algorithms using the displacement method as the preconditioner. Two types of iterative algorithms are implemented. Those are: constant metric iterations which does not involve the update of preconditioner; variable metric iterations, in which the inverse of the preconditioning matrix is updated. A series of numerical experiments is conducted to evaluate the numerical performance with application to linear and nonlinear model problems.

  14. ITER fast plant system controller prototype based on ATCA platform

    Energy Technology Data Exchange (ETDEWEB)

    Goncalves, B., E-mail: bruno@ipfn.ist.utl.pt [Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, Universidade Tecnica de Lisboa, P-1049-001 Lisboa (Portugal); Sousa, J.; Carvalho, B.B.; Batista, A.; Neto, A.; Santos, B.; Duarte, A.; Valcarcel, D.; Alves, D.; Correia, M.; Rodrigues, A.P.; Carvalho, P.F. [Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, Universidade Tecnica de Lisboa, P-1049-001 Lisboa (Portugal); Ruiz, M. [Grupo de Investigacion en Instrumentacion y Acustica Aplicada, Universidad Politecnica de Madrid (Spain); Vega, J.; Castro, R. [Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas, CIEMAT, Av. Complutense, Madrid (Spain); Lopez, J.M. [Grupo de Investigacion en Instrumentacion y Acustica Aplicada, Universidad Politecnica de Madrid (Spain); Utzel, N.; Makijarvi, P. [ITER Organization, CS 90 046, 13067 St. Paul lez Durance Cedex (France)

    2012-12-15

    The ITER fast plan system controllers (FPSC) are based on embedded technologies. The FPSCs [1] will be devoted to data acquisition tasks (sampling rates >1 kSPS) and control purposes in closed-control loops whose cycle times are below 1 ms. Fast controllers will be dedicated industrial controllers with the ability to supervise other fast and/or slow controllers and interface to actuators, sensors and high performance networks. This contribution presents an FPSC prototype, specialized for data acquisition, based on the ATCA (Advanced Telecommunications Computing Architecture) standard. This prototyping activity contributes to the ITER Plant Control Design Handbook (PCDH) effort of standardization, specifically regarding fast controller characteristics. For the prototype, IPFN has developed a new family of ATCA modules targeting ITER requirements. This family of modules comprises an AMC (Advanced Mezzanine Card) carrier/data hub/timing hub, compliant with the upcoming ATCA extensions for Physics, and a multi-channel galvanically isolated PnP digitizer, designed for serviceability. The design and test of a peer-to-peer communications layer for the implementation of a reflective memory over PCI Express and the design and test of an IEEE-1588 transport layer over an high performance serial link were also performed. In this contribution, a complete description of the solution is presented as well as the integration of the controller into the standard CODAC environment. The most relevant test results will be addressed, focusing in the benefits and limitations of the applied technologies.

  15. Meromorphic iterative roots of linear fractional functions

    Institute of Scientific and Technical Information of China (English)

    SHI YongGuo; CHEN Li

    2009-01-01

    Iterative root problem can be regarded as a weak version of the problem of embedding a homeomorphism into a flow. There are many results on iterative roots of monotone functions. However, this problem gets more difficult in non-monotone cases. Therefore, it is interesting to find iterative roots of linear fractional functions (abbreviated as LFFs), a class of non-monotone functions on R. In this paper, iterative roots of LFFs are studied on C. An equivalence between the iterative functional equation for non-constant LFFs and the matrix equation is given. By means of a method of finding matrix roots, general formulae of all meromorphic iterative roots of LFFs are obtained and the precise number of roots is also determined in various cases. As applications, we present all meromorphic iterative roots for functions z and 1/z.

  16. Performance bounds for Lambda Policy Iteration

    CERN Document Server

    Scherrer, Bruno

    2007-01-01

    We consider the discrete-time infinite-horizon discounted stationary optimal control problem formalized by Markov Decision Processes. We study Lambda Policy Iteration, a family of algorithms parameterized by lambda, originally introduced by Ioffe and Bertsekas. Lambda Policy Iteration generalizes the standard algorithms Value Iteration and Policy Iteration, and has some connections with TD(Lambda) introduced by Sutton & Barto. We deepen the original theory developped by Ioffe and Bertsekas by providing convergence rate bounds which generalize standard bounds for Value Iteration described for instance by Puterman. We also develop the theory of this algorithm when it is used in an approximate form. Doing so, we extend and unify the separate analyses developped by Munos for Approximate Value Iteration and Approximate Policy Iteration.

  17. Engineering design of the ITER Collective Thomson Scattering diagnostic. Contract EFDA 06-1478

    Energy Technology Data Exchange (ETDEWEB)

    Michelsen, P.K.; Furtula, V.; Korsholm, S.B.; Leipold, F.; Meo, F.; Salewski, M.; Bindslev, H.; Lauritzen, B.; Lucas, M.; Nonboel, E.

    2009-12-15

    This report describes the work done under EFDA contract 06-1478 (EFDA Ref.: TW6-TPDS-DIASUP10). The main part of the work has been focused on: 1) An outline plan for the full development of the CTS diagnostic for ITER, including specifications for future design tasks on the system and R and D tasks on critical components. 2) An engineering design and test in a blanket mock-up of the frontend quasi-optical High Field Side (HFS) antenna system,. 3) Some considerations on the waveguide mounting. 4) Neutronics and thermo-elastic calculations on nuclear and radiative heating of the first mirror required to provide input to the engineering design. 5) An engineering design of the front-end quasi-optical components for the Low Field Side (LFS) system in the port plug. 6) A discussion on possible calibration methods. (author)

  18. Evaluation of induced activity, decay heat and dose rate distribution after shutdown in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Maki, Koichi [Hitachi Ltd., Ibaraki (Japan). Hitachi Research Lab.; Satoh, Satoshi; Hayashi, Katsumi; Yamada, Koubun; Takatsu, Hideyuki; Iida, Hiromasa

    1997-03-01

    Induced activity, decay heat and dose rate distributions after shutdown were estimated for 1MWa/m{sup 2} operation in ITER. The activity in the inboard blanket one day after shutdown is 1.5x10{sup 11}Bq/cm{sup 3}, and the average decay heating rate 0.01w/cm{sup 3}. The dose rate outside the 120cm thick concrete biological shield is two order higher than the design criterion of 5{mu}Sv/h. This indicates that the biological shield thickness should be enhanced by 50cm in concrete, that is, total thickness 170cm for workers to enter the reactor room and to perform maintenance. (author)

  19. TRIPOLI-4{sup ®} Monte Carlo code ITER A-lite neutronic model validation

    Energy Technology Data Exchange (ETDEWEB)

    Jaboulay, Jean-Charles, E-mail: jean-charles.jaboulay@cea.fr [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Cayla, Pierre-Yves; Fausser, Clement [MILLENNIUM, 16 Av du Québec Silic 628, F-91945 Villebon sur Yvette (France); Damian, Frederic; Lee, Yi-Kang; Puma, Antonella Li; Trama, Jean-Christophe [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France)

    2014-10-15

    3D Monte Carlo transport codes are extensively used in neutronic analysis, especially in radiation protection and shielding analyses for fission and fusion reactors. TRIPOLI-4{sup ®} is a Monte Carlo code developed by CEA. The aim of this paper is to show its capability to model a large-scale fusion reactor with complex neutron source and geometry. A benchmark between MCNP5 and TRIPOLI-4{sup ®}, on the ITER A-lite model was carried out; neutron flux, nuclear heating in the blankets and tritium production rate in the European TBMs were evaluated and compared. The methodology to build the TRIPOLI-4{sup ®} A-lite model is based on MCAM and the MCNP A-lite model. Simplified TBMs, from KIT, were integrated in the equatorial-port. A good agreement between MCNP and TRIPOLI-4{sup ®} is shown; discrepancies are mainly included in the statistical error.

  20. Neutronics Evaluation of Lithium-Based Ternary Alloys in IFE Blankets

    Energy Technology Data Exchange (ETDEWEB)

    Jolodosky, A. [Univ. of California, Berkeley, CA (United States); Fratoni, M. [Univ. of California, Berkeley, CA (United States)

    2015-09-22

    Lithium is often the preferred choice as breeder and coolant in fusion blankets as it offers excellent heat transfer and corrosion properties, and most importantly, it has a very high tritium solubility and results in very low levels of tritium permeation throughout the facility infrastructure. However, lithium metal vigorously reacts with air and water and exacerbates plant safety concerns. For this reason, over the years numerous blanket concepts have been proposed with the scope of reducing concerns associated with lithium. The European helium cooled pebble bed breeding blanket (HCPB) physically confines lithium within ceramic pebbles. The pebbles reside within a low activation martensitic ferritic steel structure and are cooled by helium. The blanket is composed of the tritium breeding lithium ceramic pebbles and neutron multiplying beryllium pebbles. Other blanket designs utilize lead to lower chemical reactivity; LiPb alone can serve as a breeder, coolant, neutron multiplier, and tritium carrier. Blankets employing LiPb coolants alongside silicon carbide structural components can achieve high plant efficiency, low afterheat, and low operation pressures. This alloy can also be used alongside of helium such as in the dual-coolant lead-lithium concept (DCLL); helium is utilized to cool the first wall and structural components made up of low-activation ferritic steel, whereas lithium-lead (LiPb) acts as a self-cooled breeder in the inner channels of the blanket. The helium-cooled steel and lead-lithium alloy are separated by flow channel inserts (usually made out of silicon carbide) which thermally insulate the self-cooled breeder region from the helium cooled steel walls. This creates a LiPb breeder with a much higher exit temperature than the steel which increases the power cycle efficiency and also lowers the magnetohydrodynamic (MHD) pressure drop [6]. Molten salt blankets with a mixture of lithium, beryllium, and fluorides (FLiBe) offer good tritium breeding

  1. P-SPARSLIB: A parallel sparse iterative solution package

    Energy Technology Data Exchange (ETDEWEB)

    Saad, Y. [Univ. of Minnesota, Minneapolis, MN (United States)

    1994-12-31

    Iterative methods are gaining popularity in engineering and sciences at a time where the computational environment is changing rapidly. P-SPARSLIB is a project to build a software library for sparse matrix computations on parallel computers. The emphasis is on iterative methods and the use of distributed sparse matrices, an extension of the domain decomposition approach to general sparse matrices. One of the goals of this project is to develop a software package geared towards specific applications. For example, the author will test the performance and usefulness of P-SPARSLIB modules on linear systems arising from CFD applications. Equally important is the goal of portability. In the long run, the author wishes to ensure that this package is portable on a variety of platforms, including SIMD environments and shared memory environments.

  2. Cryogenic instrumentation for ITER magnets

    Science.gov (United States)

    Poncet, J.-M.; Manzagol, J.; Attard, A.; André, J.; Bizel-Bizellot, L.; Bonnay, P.; Ercolani, E.; Luchier, N.; Girard, A.; Clayton, N.; Devred, A.; Huygen, S.; Journeaux, J.-Y.

    2017-02-01

    Accurate measurements of the helium flowrate and of the temperature of the ITER magnets is of fundamental importance to make sure that the magnets operate under well controlled and reliable conditions, and to allow suitable helium flow distribution in the magnets through the helium piping. Therefore, the temperature and flow rate measurements shall be reliable and accurate. In this paper, we present the thermometric chains as well as the venturi flow meters installed in the ITER magnets and their helium piping. The presented thermometric block design is based on the design developed by CERN for the LHC, which has been further optimized via thermal simulations carried out by CEA. The electronic part of the thermometric chain was entirely developed by the CEA and will be presented in detail: it is based on a lock-in measurement and small signal amplification, and also provides a web interface and software to an industrial PLC. This measuring device provides a reliable, accurate, electromagnetically immune, and fast (up to 100 Hz bandwidth) system for resistive temperature sensors between a few ohms to 100 kΩ. The flowmeters (venturi type) which make up part of the helium mass flow measurement chain have been completely designed, and manufacturing is on-going. The behaviour of the helium gas has been studied in detailed thanks to ANSYS CFX software in order to obtain the same differential pressure for all types of flowmeters. Measurement uncertainties have been estimated and the influence of input parameters has been studied. Mechanical calculations have been performed to guarantee the mechanical strength of the venturis required for pressure equipment operating in nuclear environment. In order to complete the helium mass flow measurement chain, different technologies of absolute and differential pressure sensors have been tested in an applied magnetic field to identify equipment compatible with the ITER environment.

  3. A passively-safe fusion reactor blanket with helium coolant and steel structure

    Energy Technology Data Exchange (ETDEWEB)

    Crosswait, K.M.

    1994-04-01

    Helium is attractive for use as a fusion blanket coolant for a number of reasons. It is neutronically and chemically inert, nonmagnetic, and will not change phase during any off-normal or accident condition. A significant disadvantage of helium, however, is its low density and volumetric heat capacity. This disadvantage manifests itself most clearly during undercooling accident conditions such as a loss of coolant accident (LOCA) or a loss of flow accident (LOFA). This thesis describes a new helium-cooled tritium breeding blanket concept which performs significantly better during such accidents than current designs. The proposed blanket uses reduced-activation ferritic steel as a structural material and is designed for neutron wall loads exceeding 4 MW/m{sup 2}. The proposed geometry is based on the nested-shell concept developed by Wong, but some novel features are used to reduce the severity of the first wall temperature excursion. These features include the following: (1) A ``beryllium-joint`` concept is introduced, which allows solid beryllium slabs to be used as a thermal conduction path from the first wall to the cooler portions of the blanket. The joint concept allows for significant swelling of the beryllium (10 percent or more) without developing large stresses in the blanket structure. (2) Natural circulation of the coolant in the water-cooled shield is used to maintain shield temperatures below 100 degrees C, thus maintaining a heat sink close to the blanket during the accident. This ensures the long-term passive safety of the blanket.

  4. Electrical behaviour of ceramic breeder blankets in pebble form after γ-radiation

    Directory of Open Access Journals (Sweden)

    E. Carella

    2015-07-01

    Full Text Available Lithium orthosilicate (Li4SiO4 ceramics in from of pebble bed is the European candidate for ITER testing HCPB (Helium Cooled Pebble Bed breeding modules. The breeder function and the shielding role of this material, represent the areas upon which attention is focused. Electrical measurements are proposed for monitoring the modification created by ionizing radiation and at the same time provide information on lithium movement in this ceramic structure. The electrical tests are performed on pebbles fabricated by Spray-dryer method before and after gamma-irradiation through a 60Co source to a fluence of 4.8 Gy/s till a total dose of 5 ∗ 105 Gy. The introduction of thermal annealing treatments during the electrical impedance spectroscopy (EIS measurements points out the recombination effect of the temperature on the γ-induced defects.

  5. IHadoop: Asynchronous iterations for MapReduce

    KAUST Repository

    Elnikety, Eslam Mohamed Ibrahim

    2011-11-01

    MapReduce is a distributed programming frame-work designed to ease the development of scalable data-intensive applications for large clusters of commodity machines. Most machine learning and data mining applications involve iterative computations over large datasets, such as the Web hyperlink structures and social network graphs. Yet, the MapReduce model does not efficiently support this important class of applications. The architecture of MapReduce, most critically its dataflow techniques and task scheduling, is completely unaware of the nature of iterative applications; tasks are scheduled according to a policy that optimizes the execution for a single iteration which wastes bandwidth, I/O, and CPU cycles when compared with an optimal execution for a consecutive set of iterations. This work presents iHadoop, a modified MapReduce model, and an associated implementation, optimized for iterative computations. The iHadoop model schedules iterations asynchronously. It connects the output of one iteration to the next, allowing both to process their data concurrently. iHadoop\\'s task scheduler exploits inter-iteration data locality by scheduling tasks that exhibit a producer/consumer relation on the same physical machine allowing a fast local data transfer. For those iterative applications that require satisfying certain criteria before termination, iHadoop runs the check concurrently during the execution of the subsequent iteration to further reduce the application\\'s latency. This paper also describes our implementation of the iHadoop model, and evaluates its performance against Hadoop, the widely used open source implementation of MapReduce. Experiments using different data analysis applications over real-world and synthetic datasets show that iHadoop performs better than Hadoop for iterative algorithms, reducing execution time of iterative applications by 25% on average. Furthermore, integrating iHadoop with HaLoop, a variant Hadoop implementation that caches

  6. Iterated Stretching of Viscoelastic Jets

    Science.gov (United States)

    Chang, Hsueh-Chia; Demekhin, Evgeny A.; Kalaidin, Evgeny

    1999-01-01

    We examine, with asymptotic analysis and numerical simulation, the iterated stretching dynamics of FENE and Oldroyd-B jets of initial radius r(sub 0), shear viscosity nu, Weissenberg number We, retardation number S, and capillary number Ca. The usual Rayleigh instability stretches the local uniaxial extensional flow region near a minimum in jet radius into a primary filament of radius [Ca(1 - S)/ We](sup 1/2)r(sub 0) between two beads. The strain-rate within the filament remains constant while its radius (elastic stress) decreases (increases) exponentially in time with a long elastic relaxation time 3We(r(sup 2, sub 0)/nu). Instabilities convected from the bead relieve the tension at the necks during this slow elastic drainage and trigger a filament recoil. Secondary filaments then form at the necks from the resulting stretching. This iterated stretching is predicted to occur successively to generate high-generation filaments of radius r(sub n), (r(sub n)/r(sub 0)) = square root of 2[r(sub n-1)/r(sub 0)](sup 3/2) until finite-extensibility effects set in.

  7. ITER Port Interspace Pressure Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, Juan J [ORNL; Van Hove, Walter A [ORNL

    2016-01-01

    The ITER Vacuum Vessel (VV) is equipped with 54 access ports. Each of these ports has an opening in the bioshield that communicates with a dedicated port cell. During Tokamak operation, the bioshield opening must be closed with a concrete plug to shield the radiation coming from the plasma. This port plug separates the port cell into a Port Interspace (between VV closure lid and Port Plug) on the inner side and the Port Cell on the outer side. This paper presents calculations of pressures and temperatures in the ITER (Ref. 1) Port Interspace after a double-ended guillotine break (DEGB) of a pipe of the Tokamak Cooling Water System (TCWS) with high temperature water. It is assumed that this DEGB occurs during the worst possible conditions, which are during water baking operation, with water at a temperature of 523 K (250 C) and at a pressure of 4.4 MPa. These conditions are more severe than during normal Tokamak operation, with the water at 398 K (125 C) and 2 MPa. Two computer codes are employed in these calculations: RELAP5-3D Version 4.2.1 (Ref. 2) to calculate the blowdown releases from the pipe break, and MELCOR, Version 1.8.6 (Ref. 3) to calculate the pressures and temperatures in the Port Interspace. A sensitivity study has been performed to optimize some flow areas.

  8. Status of US ITER Diagnostics

    Science.gov (United States)

    Stratton, B.; Delgado-Aparicio, L.; Hill, K.; Johnson, D.; Pablant, N.; Barnsley, R.; Bertschinger, G.; de Bock, M. F. M.; Reichle, R.; Udintsev, V. S.; Watts, C.; Austin, M.; Phillips, P.; Beiersdorfer, P.; Biewer, T. M.; Hanson, G.; Klepper, C. C.; Carlstrom, T.; van Zeeland, M. A.; Brower, D.; Doyle, E.; Peebles, A.; Ellis, R.; Levinton, F.; Yuh, H.

    2013-10-01

    The US is providing 7 diagnostics to ITER: the Upper Visible/IR cameras, the Low Field Side Reflectometer, the Motional Stark Effect diagnostic, the Electron Cyclotron Emission diagnostic, the Toroidal Interferometer/Polarimeter, the Core Imaging X-Ray Spectrometer, and the Diagnostic Residual Gas Analyzer. The front-end components of these systems must operate with high reliability in conditions of long pulse operation, high neutron and gamma fluxes, very high neutron fluence, significant neutron heating (up to 7 MW/m3) , large radiant and charge exchange heat flux (0.35 MW/m2) , and high electromagnetic loads. Opportunities for repair and maintenance of these components will be limited. These conditions lead to significant challenges for the design of the diagnostics. Space constraints, provision of adequate radiation shielding, and development of repair and maintenance strategies are challenges for diagnostic integration into the port plugs that also affect diagnostic design. The current status of design of the US ITER diagnostics is presented and R&D needs are identified. Supported by DOE contracts DE-AC02-09CH11466 (PPPL) and DE-AC05-00OR22725 (UT-Battelle, LLC).

  9. Experimental developments towards an ITER thermography diagnostic

    Energy Technology Data Exchange (ETDEWEB)

    Reichle, R. [Association EURATOM-CEA, DSM/DRFC, CEA Cadarache, F-13108 St. Paul-lez-Durance (France)]. E-mail: roger.reichle@cea.fr; Brichard, B. [SCK.CEN, EURATOM Ass. Belgium, Boeretang 200, B-2400 Mol (Belgium); Escourbiac, F. [Association EURATOM-CEA, DSM/DRFC, CEA Cadarache, F-13108 St. Paul-lez-Durance (France); Gardarein, J.L. [Association EURATOM-CEA, DSM/DRFC, CEA Cadarache, F-13108 St. Paul-lez-Durance (France); Hernandez, D. [PROMES CNRS 8521, Centre Felix Trombe, BP5, Odeillo, F-66125 Font Romeu (France); Le Niliot, C. [IUSTI UMR CNRS 65 95, Uni. de Provence, 5 rue Enrico Fermi, F-13453 Marseille (France); Rigollet, F. [IUSTI UMR CNRS 65 95, Uni. de Provence, 5 rue Enrico Fermi, F-13453 Marseille (France); Serra, J.J. [DGA/DET/CEP/LOT/EHF, 10 rue des fours solaires, BP 6, F-66125 Font Romeu (France); Badie, J.M. [PROMES CNRS 8521, Centre Felix Trombe, BP5, Odeillo, F-66125 Font Romeu (France); van Ierschot, S. [SCK.CEN, EURATOM Ass. Belgium, Boeretang 200, B-2400 Mol (Belgium); Jouve, M. [Association EURATOM-CEA, DSM/DRFC, CEA Cadarache, F-13108 St. Paul-lez-Durance (France); Martinez, S. [IUSTI UMR CNRS 65 95, Uni. de Provence, 5 rue Enrico Fermi, F-13453 Marseille (France); Ooms, H. [SCK.CEN, EURATOM Ass. Belgium, Boeretang 200, B-2400 Mol (Belgium); Pocheau, C. [Association EURATOM-CEA, DSM/DRFC, CEA Cadarache, F-13108 St. Paul-lez-Durance (France); Rauber, X. [ENSAM, 2, cours des Arts et Metiers, F-13617 Aix-en-Provence (France); Sans, J.L. [PROMES CNRS 8521, Centre Felix Trombe, BP5, Odeillo, F-66125 Font Romeu (France); Scheer, E. [DGA/DET/CEP/LOT/EHF, 10 rue des fours solaires, BP 6, F-66125 Font Romeu (France); Berghmans, F. [SCK.CEN, EURATOM Ass. Belgium, Boeretang 200, B-2400 Mol (Belgium); Decreton, M. [SCK.CEN, EURATOM Ass. Belgium, Boeretang 200, B-2400 Mol (Belgium)

    2007-06-15

    In the course of the development of a concept for a spectrally resolving thermography diagnostic for the ITER divertor using optical fibres experimental development work has been carried out in three different areas. Firstly ZrF{sub 4} fibres and hollow fibres (silica capillaries with internal AG/AgJ coating) were tested in a Co{sup 60} irradiation facility under {gamma} irradiation up to doses of 5 kGy and 27 kGy, respectively. The ZrF{sub 4} fibres suffered more radiation induced degradation (>1 db/m) then the hollow fibres (0-0.4 db/m). Secondly multi-colour pyroreflectometry is being developed towards tokamak applicability. The emissivity and temperature of tungsten samples were measured in the range of 700-1500 {sup o}C. The angular working range for off normal observation of the method was 20-30{sup o}. The working distance of the method has been be increased from cm to the m range. Finally, encouraging preliminary results have been obtained concerning the application of pulsed and modulated active thermography.

  10. Design of ITER NBI power supply system

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Kazuhiro; Ohara, Yoshihiro; Okumura, Yoshikazu [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Higa, Osamu; Kawashima, Syuichi; Ono, Youichi; Tanaka, Masanobu; Yasutomi, Sei

    1997-07-01

    Power supply system for the ITER neutral beam injector (NBI) whose total injection power is 1 MeV, 50 MW from three modules, has been designed. The power supply system consists of a source power supply for negative ion production/extraction and a DC 1 MV, 45 A power supply for negative ion acceleration. An inverter controlled multi-transformer/rectifier system has been adopted to the acceleration power supply. An inverter frequency of 150 Hz was selected to satisfy required specifications which are rise time of <100 ms, voltage ripple of <10% peak to peak and cut off speed of <200{mu}s. It was confirmed that the rise time, the ripple and the cut off speed is about 50 ms, 7% and <200{mu}s respectively by computation. It was also confirmed that a surge current and an energy input to the ion source at the breakdown can be suppressed lower than 3 kA and 10 J, which are considered to be lower than allowable values. A 1 MV transmission line has been designed from a view point of electric field on the inner conductors and grounded conductor. The results from the design study indicate that all the required specification to the power supply system can be satisfied and that R and D on the transmission line is one of the most important subjects. (author)

  11. Improved structure and long-life blanket concepts for heliotron reactors

    Science.gov (United States)

    Sagara, A.; Imagawa, S.; Mitarai, O.; Dolan, T.; Tanaka, T.; Kubota, Y.; Yamazaki, K.; Watanabe, K. Y.; Mizuguchi, N.; Muroga, T.; Noda, N.; Kaneko, O.; Yamada, H.; Ohyabu, N.; Uda, T.; Komori, A.; Sudo, S.; Motojima, O.

    2005-04-01

    New design approaches are proposed for the LHD-type heliotron D-T demo-reactor FFHR2 to solve the key engineering issues of blanket space limitation and replacement difficulty. A major radius of over 14 m is selected to permit a blanket-shield thickness of about 1 m and to reduce the neutron wall loading and toroidal field, while achieving an acceptable cost of electricity. Two sets of optimization are successfully carried out. One is to reduce the magnetic hoop force on the helical coil support structures by adjustment of the helical winding coil pitch parameter and the poloidal coils design, which facilitates expansion of the maintenance ports. The other is a long-life blanket concept using carbon armour tiles that soften the neutron energy spectrum incident on the self-cooled flibe-reduced activation ferritic steel blanket. In this adaptation of the spectral-shifter and tritium breeder blanket (STB) concept a local tritium breeding ratio over 1.2 is feasible by optimized arrangement of the neutron multiplier Be in the carbon tiles, and the radiation shielding of the superconducting magnet coils is also significantly improved. Using constant cross sections of a helically winding shape, the 'screw coaster' concept is proposed to replace in-vessel components such as the STB armour tiles. The key R&D issues for developing the STB concept, such as radiation effects on carbon and enhanced heat transfer of Flibe, are elucidated.

  12. Preliminary Neutronics Analysis of the ITER Toroidal Interferometer and Polarimeter Diagnostic Corner Cube Retroreflectors

    Energy Technology Data Exchange (ETDEWEB)

    Tresemer, K. R.

    2015-07-01

    ITER is an international project under construction in France that will demonstrate nuclear fusion at a power plant-relevant scale. The Toroidal Interferometer and Polarimeter (TIP) Diagnostic will be used to measure the plasma electron line density along 5 laser-beam chords. This line-averaged density measurement will be input to the ITER feedback-control system. The TIP is considered the primary diagnostic for these measurements, which are needed for basic ITER machine control. Therefore, system reliability & accuracy is a critical element in TIP’s design. There are two major challenges to the reliability of the TIP system. First is the survivability and performance of in-vessel optics and second is maintaining optical alignment over long optical paths and large vessel movements. Both of these issues greatly depend on minimizing the overall distortion due to neutron & gamma heating of the Corner Cube Retroreflectors (CCRs). These are small optical mirrors embedded in five first wall locations around the vacuum vessel, corresponding to certain plasma tangency radii. During the development of the design and location of these CCRs, several iterations of neutronics analyses were performed to determine and minimize the total distortion due to nuclear heating of the CCRs. The CCR corresponding to TIP Channel 2 was chosen for analysis as a good middle-road case, being an average distance from the plasma (of the five channels) and having moderate neutron shielding from its blanket shield housing. Results show that Channel 2 meets the requirements of the TIP Diagnostic, but barely. These results suggest other CCRs might be at risk of exceeding thermal deformation due to nuclear heating.

  13. Activation analyses for the Korea helium cooled ceramic reflector test blanket module

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Cheol Woo, E-mail: cwl@kaeri.re.kr [Korea Atomic Energy Research Institute, 989 Daeduck-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Lee, Young-Ouk [Korea Atomic Energy Research Institute, 989 Daeduck-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Ahn, Mu-Young; Cho, Seungyon [National Fusion Research Institute, Gwahangno, Yuseong-gu, Daejeon 305-333 (Korea, Republic of); Lee, Dong Won [Korea Atomic Energy Research Institute, 989 Daeduck-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2013-10-15

    The activation analyses were performed to obtain activities, inventories and dose rates after shutdown for the Korea HCCR TBM. MonteBurns code coupled with MCNP and CINDER codes was used in the activation calculation and the irradiation history based on the SA2 scenario was applied. The total activity in TBM after shutdown was evaluated as a value of 4.12 × 10{sup 16} Bq. Dose rates after shutdown from the activated HCCR TBM was estimated. The decay gamma-ray spectra in each region inside TBM were estimated based on the calculated activities. The dose rate at 0 cm–3 m from the FW surface of HCCR TBM was evaluated according to the cooling time of 1 day, 1 week, 1 month, 6 months and 1 year. The dose rate at 0 cm from the FW surface was evaluated to be 575.84 Sv/h at 1 day after shutdown.

  14. Preconditioned iterations to calculate extreme eigenvalues

    Energy Technology Data Exchange (ETDEWEB)

    Brand, C.W.; Petrova, S. [Institut fuer Angewandte Mathematik, Leoben (Austria)

    1994-12-31

    Common iterative algorithms to calculate a few extreme eigenvalues of a large, sparse matrix are Lanczos methods or power iterations. They converge at a rate proportional to the separation of the extreme eigenvalues from the rest of the spectrum. Appropriate preconditioning improves the separation of the eigenvalues. Davidson`s method and its generalizations exploit this fact. The authors examine a preconditioned iteration that resembles a truncated version of Davidson`s method with a different preconditioning strategy.

  15. A FAST CONVERGENT METHOD OF ITERATED REGULARIZATION

    Institute of Scientific and Technical Information of China (English)

    Huang Xiaowei; Wu Chuansheng; Wu Di

    2009-01-01

    This article presents a fast convergent method of iterated regularization based on the idea of Landweber iterated regularization, and a method for a-posteriori choice by the Morozov discrepancy principle and the optimum asymptotic convergence order of the regularized solution is obtained. Numerical test shows that the method of iterated regu-larization can quicken the convergence speed and reduce the calculation burden efficiently.

  16. Iterative solution of the reduced eigenvalue problem

    Energy Technology Data Exchange (ETDEWEB)

    Sauer, G. (Technischer Ueberwachungs-Verein Bayern e.V., Muenchen (Germany, F.R.))

    1991-04-01

    The Guyan method of reducing the stiffness and mass matrices of large linear structures introduces errors in the reduced mass matrix. These errors cannot be completely avoided even if the analysis coordinates are chosen optimally. However, they can be elimiated by iterating on the eigenvectors found from the Guyan reduced matrices. The necessary iteration steps follow directly from the eigenvalue problem. The resulting iteration procedures are presented and applied to two test problems showing that the iterations enable the exact eigensolutions to be extracted. All errors from the Guyan reduced matrices are removed or substantially decreased. (orig.).

  17. Research at ITER towards DEMO: Specific reactor diagnostic studies to be carried out on ITER

    Science.gov (United States)

    Krasilnikov, A. V.; Kaschuck, Y. A.; Vershkov, V. A.; Petrov, A. A.; Petrov, V. G.; Tugarinov, S. N.

    2014-08-01

    In ITER diagnostics will operate in the very hard radiation environment of fusion reactor. Extensive technology studies are carried out during development of the ITER diagnostics and procedures of their calibration and remote handling. Results of these studies and practical application of the developed diagnostics on ITER will provide the direct input to DEMO diagnostic development. The list of DEMO measurement requirements and diagnostics will be determined during ITER experiments on the bases of ITER plasma physics results and success of particular diagnostic application in reactor-like ITER plasma. Majority of ITER diagnostic already passed the conceptual design phase and represent the state of the art in fusion plasma diagnostic development. The number of related to DEMO results of ITER diagnostic studies such as design and prototype manufacture of: neutron and γ-ray diagnostics, neutral particle analyzers, optical spectroscopy including first mirror protection and cleaning technics, reflectometry, refractometry, tritium retention measurements etc. are discussed.

  18. Research at ITER towards DEMO: Specific reactor diagnostic studies to be carried out on ITER

    Energy Technology Data Exchange (ETDEWEB)

    Krasilnikov, A. V.; Kaschuck, Y. A.; Vershkov, V. A.; Petrov, A. A.; Petrov, V. G.; Tugarinov, S. N. [Institution Project center ITER, Moscow (Russian Federation)

    2014-08-21

    In ITER diagnostics will operate in the very hard radiation environment of fusion reactor. Extensive technology studies are carried out during development of the ITER diagnostics and procedures of their calibration and remote handling. Results of these studies and practical application of the developed diagnostics on ITER will provide the direct input to DEMO diagnostic development. The list of DEMO measurement requirements and diagnostics will be determined during ITER experiments on the bases of ITER plasma physics results and success of particular diagnostic application in reactor-like ITER plasma. Majority of ITER diagnostic already passed the conceptual design phase and represent the state of the art in fusion plasma diagnostic development. The number of related to DEMO results of ITER diagnostic studies such as design and prototype manufacture of: neutron and γ–ray diagnostics, neutral particle analyzers, optical spectroscopy including first mirror protection and cleaning technics, reflectometry, refractometry, tritium retention measurements etc. are discussed.

  19. Numerical design of the Seed-Blanket Unit for the thorium nuclear fuel cycle

    Directory of Open Access Journals (Sweden)

    Oettingen Mikołaj

    2016-01-01

    Full Text Available In the paper we present the Monte Carlo modelling by the means of the Monte Carlo Continuous Energy Burn-up Code of the 17x17 Pressurized Water Reactor fuel assembly designed according to the Radkowsky Thorium Fuel concept. The design incorporates the UO2 seed fuel located in the centre and (Th,UO2 blanket fuel located in the peripheries of fuel assembly. The high power seed region supplies neutrons for the low power blanket region and thus induces breeding of fissile 233U from fertile 232Th. The both regions are physically separated and thus this approach is also known as either the heterogonous approach or the Seed-Blanket Unit. In the numerical analysis we consider the time evolutions of infinite neutron multiplication factor, axial/radial power density profile, 233U, 235U and 232Th.

  20. Electrically insulating coatings for V-Li self-cooled blanket in a fusion system

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Reed, C. B.; Uz, M.; Park, J. H.; Smith, D. L.

    2000-05-17

    The blanket system is one of the most important components in a fusion reactor because it has a major impact on both the economics and safety of fusion energy. The primary functions of the blanket in a deuterium/tritium-fueled fusion reactor are to convert the fusion energy into sensible heat and to breed tritium for the fuel cycle. The liquid-metal blanket concept requires an electrically insulating coating on the first-wall structural material to minimize the magnetohydrodynamic pressure drop that occurs during the flow of liquid metal in a magnetic field. Based on the thermodynamics of interactions between the coating and the liquid lithium on one side and the structural V-base alloy on the other side, several coating candidates are being examined to perform the insulating function over a wide range of temperatures and lithium chemistries.

  1. Neutronics Analysis of Water-Cooled Ceramic Breeder Blanket for CFETR

    Science.gov (United States)

    Zhu, Qingjun; Li, Jia; Liu, Songlin

    2016-07-01

    In order to investigate the nuclear response to the water-cooled ceramic breeder blanket models for CFETR, a detailed 3D neutronics model with 22.5° torus sector was developed based on the integrated geometry of CFETR, including heterogeneous WCCB blanket models, shield, divertor, vacuum vessel, toroidal and poloidal magnets, and ports. Using the Monte Carlo N-Particle Transport Code MCNP5 and IAEA Fusion Evaluated Nuclear Data Library FENDL2.1, the neutronics analyses were performed. The neutron wall loading, tritium breeding ratio, the nuclear heating, neutron-induced atomic displacement damage, and gas production were determined. The results indicate that the global TBR of no less than 1.2 will be a big challenge for the water-cooled ceramic breeder blanket for CFETR. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB108004, 2014GB122000, and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

  2. Tritium recovery in Pb17Li-water cooled blanket systems

    Energy Technology Data Exchange (ETDEWEB)

    Malara, C. [Safety Technology Inst., Ispra (Italy); Casini, G. [Systems Engineering & Information Inst., Ispra (Italy); Viola, A. [Univ. of Cagliari (Italy)

    1994-12-31

    The question of tritium recovery in Pb17Li, water cooled blankets is under investigation since several years at JRC Ispra. The method which has been more extensively analyzed is that of slowly circulating the breeder out from the blanket units and of extracting the tritium from it outside the plasma vacuum vessel by helium gas purging in a suited process apparatus. The design features of the process systems are related to: (1) the very low tritium solubility in Pb17Li which implies high permeation rates through the containment structures; (2) the need of keeping as low as possible the tritium concentration in the cooling water both for safety and economical reasons. A computerized model of the tritium behavior in the blanket units and in the extraction system has been developed.

  3. Iterative Demodulation and Decoding Scheme with 16QAM

    Institute of Scientific and Technical Information of China (English)

    LIU Lin-nan; KUANG Jing-ming; LI Ming; FEI Ze-song

    2006-01-01

    Iterative demodulation and decoding scheme is analyzed and modulation labeling is considered to be one of the crucial factors to this scheme. By analyzing the existent mapping design criterion, four aspects are found as the key techniques for choosing a label mapping. Based on this discovery, a novel mapping design criterion is proposed and two label mappings are searched according to it. Simulation results show that the performance of BICM-ID using the novel mappings is better than the former ones. The extrinsic information transfer(EXIT) chart is introduced and it is used to evaluate the proposed mapping design criteria.

  4. ITER CS Model Coil and CS Insert Test Results

    Energy Technology Data Exchange (ETDEWEB)

    Martovetsky, N; Michael, P; Minervina, J; Radovinsky, A; Takayasu, M; Thome, R; Ando, T; Isono, T; Kato, T; Nakajima, H; Nishijima, G; Nunoya, Y; Sugimoto, M; Takahashi, Y; Tsuji, H; Bessette, D; Okuno, K; Ricci, M

    2000-09-07

    The Inner and Outer modules of the Central Solenoid Model Coil (CSMC) were built by US and Japanese home teams in collaboration with European and Russian teams to demonstrate the feasibility of a superconducting Central Solenoid for ITER and other large tokamak reactors. The CSMC mass is about 120 t, OD is about 3.6 m and the stored energy is 640 MJ at 46 kA and peak field of 13 T. Testing of the CSMC and the CS Insert took place at Japan Atomic Energy Research Institute (JAERI) from mid March until mid August 2000. This paper presents the main results of the tests performed.

  5. Truncated states obtained by iteration

    CERN Document Server

    Cardoso, W B

    2007-01-01

    Quantum states of the electromagnetic field are of considerable importance, finding potential application in various areas of physics, as diverse as solid state physics, quantum communication and cosmology. In this paper we introduce the concept of truncated states obtained via iterative processes (TSI) and study its statistical features, making an analogy with dynamical systems theory (DST). As a specific example, we have studied TSI for the doubling and the logistic functions, which are standard functions in studying chaos. TSI for both the doubling and logistic functions exhibit certain similar patterns when their statistical features are compared from the point of view of DST. A general method to engineer TSI in the running-wave domain is employed, which includes the errors due to the nonidealities of detectors and photocounts.

  6. Planning as an Iterative Process

    Science.gov (United States)

    Smith, David E.

    2012-01-01

    Activity planning for missions such as the Mars Exploration Rover mission presents many technical challenges, including oversubscription, consideration of time, concurrency, resources, preferences, and uncertainty. These challenges have all been addressed by the research community to varying degrees, but significant technical hurdles still remain. In addition, the integration of these capabilities into a single planning engine remains largely unaddressed. However, I argue that there is a deeper set of issues that needs to be considered namely the integration of planning into an iterative process that begins before the goals, objectives, and preferences are fully defined. This introduces a number of technical challenges for planning, including the ability to more naturally specify and utilize constraints on the planning process, the ability to generate multiple qualitatively different plans, and the ability to provide deep explanation of plans.

  7. Benchmarking ICRF simulations for ITER

    Energy Technology Data Exchange (ETDEWEB)

    R. V. Budny, L. Berry, R. Bilato, P. Bonoli, M. Brambilla, R.J. Dumont, A. Fukuyama, R. Harvey, E.F. Jaeger, E. Lerche, C.K. Phillips, V. Vdovin, J. Wright, and members of the ITPA-IOS

    2010-09-28

    Abstract Benchmarking of full-wave solvers for ICRF simulations is performed using plasma profiles and equilibria obtained from integrated self-consistent modeling predictions of four ITER plasmas. One is for a high performance baseline (5.3 T, 15 MA) DT H-mode plasma. The others are for half-field, half-current plasmas of interest for the pre-activation phase with bulk plasma ion species being either hydrogen or He4. The predicted profiles are used by seven groups to predict the ICRF electromagnetic fields and heating profiles. Approximate agreement is achieved for the predicted heating power partitions for the DT and He4 cases. Profiles of the heating powers and electromagnetic fields are compared.

  8. Aggregation-iterative analogues and generalizations of projection-iterative methods

    Directory of Open Access Journals (Sweden)

    Shuvar B.F.

    2013-06-01

    Full Text Available Aggregation-iterative algorithms for linear operator equations are constructed and investigated. These algorithms cover methods of iterative aggregation and projection-iterative methods. In convergence conditions there is neither requirement for the corresponding operator of fixed sign no restriction to the spectral radius to be less than one.

  9. Installation of the ITER committee industry. Participants guide; Installation du Comite industrie ITER. Dossier des participants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-07-01

    ITER is an international project to design and build an experimental fusion reactor based on the tokamak concept. This guide presents the ITER project and objectives and the associated organizations in France, the recommendations and actions for ITER, the industrial mobilization, the industrial committee and its members, technological sheets for the enterprises and the statistical document of the SESSI. (A.L.B.)

  10. Neutronics optimization study for D-D fusion reactor blanket/shield

    Energy Technology Data Exchange (ETDEWEB)

    Shiba, T.; Kanda, Y.; Nakashima, H.

    1985-12-01

    Position-dependent optimization calculations have been carried out on a D-D fusion reactor blanket/shield to maximize the energy gain in the blanket and to minimize the atomic displacement rate of the copper stabilizer in the superconducting magnet. The results obtained by using the optimization code SWAN indicate the advantage of D/sub 2/O coolant over H/sub 2/O coolant with respect to increasing the energy gain, and the difference in the optimal shield distributions between D-T and D-D neutron sources. The possibility of improving both the energy gain and radiation shielding characteristics is also discussed.

  11. Advanced Burner Reactor with Breed-and-Burn Thorium Blankets for Improved Economics and Resource Utilization

    OpenAIRE

    Zhang, Guanheng

    2015-01-01

    This study assesses the feasibility of designing Seed and Blanket (S&B) Sodium-cooled Fast Reactor (SFR) to generate a significant fraction of the core power from radial thorium fueled blankets that operate on the Breed-and-Burn (B&B) mode without exceeding the 200 Displacements per Atom (DPA) radiation damage constraint of presently verified cladding materials. The S&B core is designed to have an elongated seed (or “driver”) to maximize the fraction of neutrons that radially leak into the su...

  12. Study of thorium-uranium based molten salt blanket in a fusion-fission hybrid reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zhao Jing, E-mail: zhao_jing@mail.tsinghua.edu.cn [INET, Tsinghua University, Beijing 100084 (China); Yang Yongwei; Zhou Zhiwei [INET, Tsinghua University, Beijing 100084 (China)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer A molten salt blanket has been designed for the fusion-fission hybrid reactor. Black-Right-Pointing-Pointer The use of Thorium in the molten salt fuels has been studied. Black-Right-Pointing-Pointer The molten salt was consisted of F-Li-Be and with the thickness of 40 cm. Black-Right-Pointing-Pointer The concentration of {sup 6}Li was chosen to be the natural enrichment ratio. Black-Right-Pointing-Pointer The result shows that TBR is greater than 1, M is about 15-16. - Abstract: Not only solid fuels, but also liquid fuels can be used for the fusion-fission symbiotic reactor blanket. The operational record of the molten salt reactor with F-Li-Be was very successful, so the F-Li-Be blanket was chosen for research. The molten salt has several features which are suited for the fusion-fission applications. The fuel material uranium and thorium were dissolved in the F-Li-Be molten salt. A combined program, COUPLE, was used for neutronics analysis of the molten salt blanket. Several cases have been calculated and compared. Not only the influence of the different fuels have been studied, but also the thickness of the molten salt, and the concentration of the {sup 6}Li in the molten salt. Preliminary studies indicate that when thorium-uranium-plutonium fuels were added into a F-Li-Be molten salt blanket and with a component of 71% LiF-2% BeF{sub 2}-13.5% ThF{sub 4}-8.5% UF{sub 4}-5% PuF{sub 3}, and also with the molten salt thickness of 40 cm and natural concentration of {sup 6}Li, the appropriate blanket energy multiplication factor and TBR can be obtained. The result shows that thorium-uranium molten salt can be used in the blanket of a fusion-fission symbiotic reactor. The research on the molten salt blanket must be valuable for the design of fusion-fission symbiotic reactor.

  13. APT Blanket Detailed Bin Model Based on Initial Plate-Type Design -3D FLOWTRAN-TF Model

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    This report provides background information for a series of reports documenting accident scenario simulations for the Accelerator Production of Tritium (APT) blanket heat removal systems. The simulations were performed in support of the Preliminary Safety Analysis Report for the APT. This report gives a brief description of the FLOWTRAN-TF code which was used for detailed blanket bin modeling.

  14. 76 FR 2093 - Eni USA Gas Marketing LLC; Application for Blanket Authorization To Export Liquefied Natural Gas

    Science.gov (United States)

    2011-01-12

    ... Gas Marketing LLC; Application for Blanket Authorization To Export Liquefied Natural Gas AGENCY... November 30, 2010, by Eni USA Gas Marketing LLC (Eni USA), requesting blanket authorization to export... and Gas Global Security and Supply, Office of Fossil Energy, Forrestal Building, Room 3E-042,...

  15. 76 FR 33746 - Freeport LNG Development, L.P.; Application for Blanket Authorization To Export Liquefied Natural...

    Science.gov (United States)

    2011-06-09

    ... Freeport LNG Development, L.P.; Application for Blanket Authorization To Export Liquefied Natural Gas..., 2011, by Freeport LNG Development, L.P. (Freeport LNG), requesting blanket authorization to export liquefied natural gas (LNG) that previously had been imported into the United States from foreign sources...

  16. 75 FR 13755 - Freeport LNG Development, L.P.; Application To Amend Blanket Authorization To Export Liquefied...

    Science.gov (United States)

    2010-03-23

    ... Freeport LNG Development, L.P.; Application To Amend Blanket Authorization To Export Liquefied Natural Gas... application filed on March 4, 2010, by Freeport LNG Development, L.P. (Freeport LNG), requesting an amendment to its blanket authorization to export liquefied natural gas (LNG) granted by DOE/FE on May 28,...

  17. 75 FR 62510 - Chevron U.S.A. Inc.; Application for Blanket Authorization To Export Liquefied Natural Gas

    Science.gov (United States)

    2010-10-12

    ... U.S.A. Inc.; Application for Blanket Authorization To Export Liquefied Natural Gas AGENCY: Office of..., 2010, by Chevron U.S.A. Inc. (Chevron), requesting blanket authorization to export liquefied natural... up to the equivalent of 72 billion cubic feet (Bcf) of natural gas on a short-term or spot...

  18. New concurrent iterative methods with monotonic convergence

    Energy Technology Data Exchange (ETDEWEB)

    Yao, Qingchuan [Michigan State Univ., East Lansing, MI (United States)

    1996-12-31

    This paper proposes the new concurrent iterative methods without using any derivatives for finding all zeros of polynomials simultaneously. The new methods are of monotonic convergence for both simple and multiple real-zeros of polynomials and are quadratically convergent. The corresponding accelerated concurrent iterative methods are obtained too. The new methods are good candidates for the application in solving symmetric eigenproblems.

  19. Iterative Brinkman penalization for remeshed vortex methods

    DEFF Research Database (Denmark)

    Hejlesen, Mads Mølholm; Koumoutsakos, Petros; Leonard, Anthony;

    2015-01-01

    We introduce an iterative Brinkman penalization method for the enforcement of the no-slip boundary condition in remeshed vortex methods. In the proposed method, the Brinkman penalization is applied iteratively only in the neighborhood of the body. This allows for using significantly larger time s...

  20. Rater Variables Associated with ITER Ratings

    Science.gov (United States)

    Paget, Michael; Wu, Caren; McIlwrick, Joann; Woloschuk, Wayne; Wright, Bruce; McLaughlin, Kevin

    2013-01-01

    Advocates of holistic assessment consider the ITER a more authentic way to assess performance. But this assessment format is subjective and, therefore, susceptible to rater bias. Here our objective was to study the association between rater variables and ITER ratings. In this observational study our participants were clerks at the University of…

  1. An iterative method for spherical bounces

    CERN Document Server

    Buniy, Roman V

    2016-01-01

    We develop a new iterative method for finding approximate solutions for spherical bounces associated with the decay of the false vacuum in scalar field theories. The method works for any generic potential in any number of dimensions, contains Coleman's thin-wall approximation as its first iteration, and greatly improves its accuracy by including higher order terms.

  2. Iterative methods for weighted least-squares

    Energy Technology Data Exchange (ETDEWEB)

    Bobrovnikova, E.Y.; Vavasis, S.A. [Cornell Univ., Ithaca, NY (United States)

    1996-12-31

    A weighted least-squares problem with a very ill-conditioned weight matrix arises in many applications. Because of round-off errors, the standard conjugate gradient method for solving this system does not give the correct answer even after n iterations. In this paper we propose an iterative algorithm based on a new type of reorthogonalization that converges to the solution.

  3. ITER Fast Ion Collective Thomson Scattering

    DEFF Research Database (Denmark)

    Bindslev, Henrik; Meo, Fernando; Korsholm, Søren Bang

    for measurements of the confined fusion alpha particles in ITER set by the ITER team. Then we outline the considerations, which enter into the selection and evaluation of CTS systems. System definition includes choice of probe frequency, geometry of probe and receiver beam patterns and probe power, but ultimately...

  4. Experimental studies of ITER demonstration discharges

    NARCIS (Netherlands)

    Sips, A.C.C.; Casper, T. A.; Doyle, E. J.; Giruzzi, G.; Gribov, Y.; Hobirk, J.; Hogeweij, G. M. D.; Horton, L. D.; Hubbard, A. E.; Hutchinson, I.; Ide, S.; Isayama, A.; Imbeaux, F.; Jackson, G. L.; Kamada, Y.; Kessel, C.; Kochl, F.; Lomas, P.; Litaudon, X.; Luce, T. C.; Marmar, E.; Mattei, M.; Nunes, I.; Oyama, N.; Parail, V.; Portone, A.; Saibene, G.; Sartori, R.; Stober, J. K.; Suzuki, T.; Wolfe, S. M.

    2009-01-01

    Key parts of the ITER scenarios are determined by the capability of the proposed poloidal field (PF) coil set. They include the plasma breakdown at low loop voltage, the current rise phase, the performance during the flat top (FT) phase and a ramp down of the plasma. The ITER discharge evolution has

  5. APT Blanket System Loss-of-Coolant Accident (LOCA) Based on Initial Conceptual Design - Case 1: External HR Break Near Inlet Header

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.

    1998-10-07

    The APT blanket system has about 57 MW of thermal energy deposited within the blanket region under normal operating conditions from the release of neutrons and the interaction of the High energy particles with the blanket materials. This corresponds to about 48 percent of total thermal energy deposited in the APT target/blanket system. The deposited thermal energy under normal operation conditions is an important input parameter used in the thermal-hydraulic design and accident analysis.

  6. Complex light modulation for lensless image projection

    Institute of Scientific and Technical Information of China (English)

    M. Makowski; A. Kolodziejczyk; A. Siemion; I. Ducin; K. Kakarenko; M. Sypek; A. M. Siemion; J. Suszek; D. Wojnowski; Z. Jaroszewicz

    2011-01-01

    We present a lensless projection of color images based on computer-generated Fourier holograms. Amplitude and phase modulation of three primary-colored laser beams is performed by a matched pair of spatial light modulators. The main advantage of the complex light modulation is the lack of iterative phase retrieval techniques. The advantage is the lack of speckles in the projected images. Experimental results are given and compared with the outcome of classical phase-only modulation.%We present a lensless projection of color images based on computer-generated Fourier holograms.Amplitude and phase modulation of three primary-colored laser beams is performed by a matched pair of spatial light modulators.The main advantage of the complex light modulation is the lack of iterative phase retrieval techniques.The advantage is the lack of speckles in the projected images.Experimental results are given and compared with the outcome of classical phase-only modulation.

  7. Distribution of bog and heath in a Newfoundland blanket bog complex: topographic limits on the hydrological processes governing blanket bog development

    Directory of Open Access Journals (Sweden)

    P. A. Graniero

    1999-01-01

    Full Text Available This research quantified the role of topography and hydrological processes within and, hence, the development of, blanket bogs. Topographic characteristics were derived from digital elevation models (DEMs developed for the surface and underlying substrate at three blanket bog sites on the southeastern lobe of the Avalon Peninsula, Newfoundland. A multinomial logit (MNL model of the probability of bog occurrence was constructed in terms of relevant topographic characteristics. The resulting model was then used to investigate the probabilistic boundary conditions of bog occurrence within the landscape. Under average curvatures for the sites studied, substrate slopes up to 0.065 favoured blanket bog development. However, steeper slopes could, theoretically, be occupied by blanked bog where water is concentrated by convergent curvatures or large contributing areas. Near community boundaries, bog and heath communities both occupied similar topographic conditions. Since these boundary locations are capable of supporting the hydrological conditions necessary for bog development, the heath is likely to be encroached upon by bog.

  8. Landweber iterative regularization for nearfield acoustic holography

    Institute of Scientific and Technical Information of China (English)

    BI Chuanxing; CHEN Xinzhao; ZHOU Rong; CHEN Jian

    2006-01-01

    On the basis of the distributed source boundary point method (DSBPM)-based nearfield acoustic holography (NAH), Landweber iterative regularization method is proposed to stabilize the NAH reconstruction process, control the influence of measurement errors on the reconstructed results and ensure the validity of the reconstructed results. And a new method, the auxiliary surface method, is proposed to determine the optimal iterative number for optimizing the regularization effect. Here, the optimal number is determined by minimizing the relative error between the calculated pressure on the auxiliary surface corresponding to each iterative number and the measured pressure. An experiment on a speaker is investigated to demonstrate the high sensitivity of the reconstructed results to measurement errors and to validate the chosen method of the optimal iterative number and the Landweber iterative regularization method for controlling the influence of measurement errors on the reconstructed results.

  9. Turbo iterative equalization for HSDPA systems

    Institute of Scientific and Technical Information of China (English)

    WU QiHui; ZHAO ChunMing; WANG JinLong

    2007-01-01

    In this paper, a turbo iterative receiver structure with chip equalization is proposed for the 3G high-speed downlink packet access (HSDPA) systems. The receiver equalizes the channel prior to the dispreading and then performs two successive soft-output decisions, achieved by a soft-input soft-output (SISO) multi-code detector and a SISO turbo decoder through an iterative process. At each iteration, extrinsic information is extracted from detection and decoding stages and is then used as a priori information in the next iteration, just as in turbo decoding. Computer simulations show that the turbo iterative receiver structure with chip equalization offers significant performance gain over the traditional receiver structure.

  10. Techniques in Iterative Proton CT Image Reconstruction

    CERN Document Server

    Penfold, Scott

    2015-01-01

    This is a review paper on some of the physics, modeling, and iterative algorithms in proton computed tomography (pCT) image reconstruction. The primary challenge in pCT image reconstruction lies in the degraded spatial resolution resulting from multiple Coulomb scattering within the imaged object. Analytical models such as the most likely path (MLP) have been proposed to predict the scattered trajectory from measurements of individual proton location and direction before and after the object. Iterative algorithms provide a flexible tool with which to incorporate these models into image reconstruction. The modeling leads to a large and sparse linear system of equations that can efficiently be solved by projection methods-based iterative algorithms. Such algorithms perform projections of the iterates onto the hyperlanes that are represented by the linear equations of the system. They perform these projections in possibly various algorithmic structures, such as block-iterative projections (BIP), string-averaging...

  11. Development of laser beam injection system for the Edge Thomson Scattering (ETS) in ITER

    Science.gov (United States)

    Yatsuka, E.; Hatae, T.; Suitoh, S.; Ohara, M.; Hagita, K.; Inoue, K.; Bassan, M.; Walsh, M.; Itami, K.

    2016-01-01

    This paper focuses on the design and development of the laser injection system for the ITER Edge Thomson Scattering system (ETS). The ITER ETS achieves a temporal resolution of 100 Hz by firing two 50 Hz laser beams alternatively. The use of dual lasers enables us to perform the Thomson scattering measurements at a temporal resolution of 50 Hz in case that one of the laser systems stops functioning. A new type of beam combiner was developed to obtain a single beam that is collinear and fixed linearly polarized from two laser beams using a motor-driven rotating half-wave plate. The rotating half-wave plate method does not induce misalignment even if the rotating mechanism malfunctions. The combined beam is relayed from the diagnostic hall to the plasma using mirror optics and is absorbed at the beam dump integrated on the inner blanket. The beam alignment system was designed to direct the laser beam onto the center of the beam dump head. The beam position at the beam dump is monitored by four alignment laser beams which propagate parallel to the diagnostic Nd:YAG laser beam and imaging systems installed outside the diagnostic port.

  12. Characteristics of microstructure and tritium release properties of different kinds of beryllium pebbles for application in tritium breeding modules

    Energy Technology Data Exchange (ETDEWEB)

    Kurinskiy, P., E-mail: petr.kurinskiy@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials – Applied Materials Physics (IAM-AWP), P.O. Box 3640, Karlsruhe 76021 (Germany); Vladimirov, P.; Moeslang, A. [Karlsruhe Institute of Technology, Institute for Applied Materials – Applied Materials Physics (IAM-AWP), P.O. Box 3640, Karlsruhe 76021 (Germany); Rolli, R. [Karlsruhe Institute of Technology, Institute for Applied Materials – Materials and Biomechanics (IAM-WBM), P.O. Box 3640, Karlsruhe 76021 (Germany); Zmitko, M. [The European Joint Undertaking for ITER and the Development of Fusion Energy, c/Josep Pla, no. 2, Torres Diagonal Litoral, Edificio B3, Barcelona 08019 (Spain)

    2014-10-15

    Highlights: • Tritium release properties and characteristics of microstructure of beryllium pebbles having different sizes of grains were studied. • Fine-grained beryllium pebbles showed the best ability to release tritium compared to pebbles from another charges. • Be pebbles with the grain sizes exceeding 100 μm contain a great number of small pores and inclusions presumably referring to the history of material fabrication. • The sizes of grains are one of a key characteristic of microstructure which influences the parameters of tritium release. - Abstract: Beryllium pebbles with diameters of 1 mm are considered to be perspective material for the use as neutron multiplier in tritium breeding modules of fusion reactors. Up to now, the design of helium-cooled breeding blanket in ITER project foresees the use of 1 mm beryllium pebbles fabricated by NGK Insulators Ltd., Japan. It is notable that beryllium pebbles from Russian Federation and USA are also available and the possibility of their large-scale fabrication is under study. Presented work is dedicated to a study of characteristics of microstructure and parameters of tritium release of beryllium pebbles produced by Bochvar Institute, Russian Federation, and Materion Corporation, USA.

  13. On the interplay between inner and outer iterations for a class of iterative methods

    Energy Technology Data Exchange (ETDEWEB)

    Giladi, E. [Stanford Univ., CA (United States)

    1994-12-31

    Iterative algorithms for solving linear systems of equations often involve the solution of a subproblem at each step. This subproblem is usually another linear system of equations. For example, a preconditioned iteration involves the solution of a preconditioner at each step. In this paper, the author considers algorithms for which the subproblem is also solved iteratively. The subproblem is then said to be solved by {open_quotes}inner iterations{close_quotes} while the term {open_quotes}outer iteration{close_quotes} refers to a step of the basic algorithm. The cost of performing an outer iteration is dominated by the solution of the subproblem, and can be measured by the number of inner iterations. A good measure of the total amount of work needed to solve the original problem to some accuracy c is then, the total number of inner iterations. To lower the amount of work, one can consider solving the subproblems {open_quotes}inexactly{close_quotes} i.e. not to full accuracy. Although this diminishes the cost of solving each subproblem, it usually slows down the convergence of the outer iteration. It is therefore interesting to study the effect of solving each subproblem inexactly on the total amount of work. Specifically, the author considers strategies in which the accuracy to which the inner problem is solved, changes from one outer iteration to the other. The author seeks the `optimal strategy`, that is, the one that yields the lowest possible cost. Here, the author develops a methodology to find the optimal strategy, from the set of slowly varying strategies, for some iterative algorithms. This methodology is applied to the Chebychev iteration and it is shown that for Chebychev iteration, a strategy in which the inner-tolerance remains constant is optimal. The author also estimates this optimal constant. Then generalizations to other iterative procedures are discussed.

  14. Iterants, Fermions and Majorana Operators

    Science.gov (United States)

    Kauffman, Louis H.

    Beginning with an elementary, oscillatory discrete dynamical system associated with the square root of minus one, we study both the foundations of mathematics and physics. Position and momentum do not commute in our discrete physics. Their commutator is related to the diffusion constant for a Brownian process and to the Heisenberg commutator in quantum mechanics. We take John Wheeler's idea of It from Bit as an essential clue and we rework the structure of that bit to a logical particle that is its own anti-particle, a logical Marjorana particle. This is our key example of the amphibian nature of mathematics and the external world. We show how the dynamical system for the square root of minus one is essentially the dynamics of a distinction whose self-reference leads to both the fusion algebra and the operator algebra for the Majorana Fermion. In the course of this, we develop an iterant algebra that supports all of matrix algebra and we end the essay with a discussion of the Dirac equation based on these principles.

  15. Active beam spectroscopy for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Von Hellermann, M.; Giroud, C.; Jaspers, R. [Association Euratom-Fom, FOM Institute for Plasma Physics Rijnhuizen, Trilateral Euregio Cluster (Netherlands); Hawkes, N.C.; Mullane, M.O.; Zastrow, K.D. [Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon (United Kingdom); Krasilnikov, A.; Tugarinov, S. [SRC RF TRINITI, Troitsk, Moscow region (Russian Federation); Lotte, P. [Association Euratom-CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; McKee, G. [Wisconsin Univ., Madison, WI (United States); Malaquias, A. [Associacao EURATOM/IST, Instituto Superior Tecnico, Lisboa (Portugal); Rachlew, E. [Kungliga Tekniska Hoegskolan (KTH), Stockholm(Sweden)

    2003-07-01

    The latest status of 'Active Beam' related spectroscopy aspects as part of the ITER diagnostic scenario is presented. A key issue of the proposed scheme is based on the concept that in order to achieve the ultimate goal of global data consistency, all particles involved, that is, intrinsic and seeded impurity ions as well as helium ash ions and bulk plasma ions and also the plasma background data (e.g. magnetic and electric fields, electron density and temperature profiles) need to be addressed. A further sensible step in this direction is the decision of exploiting both a dedicated low-energy, low-power diagnostic beam (DNB, 2.2 MW 100 keV/amu) as well as the high-power, high-energy heating beams (HNB, 17 MW 500 keV/amu) for maximum diagnostic information. The authors report some new aspects referring to the use of DNB for motional Stark effect (MSE) where the main idea is to treat both beams (HNB and DNB) as potential diagnostic tools with complementary roles. The equatorial ports for the DNB promise excellent spatial resolution, however, the angles are less favourable for a polarimetric MSE exploitation. HNB can be used as probe beam for diagnosing slowing-down fusion alpha with a birth energy of 3,5 MeV.

  16. Non-LTE Line Blanketing in Stars With Extended Outflowing Atmospheres.

    Science.gov (United States)

    Hillier, D. J.; Miller, D. L.

    1995-05-01

    With continuing advances in radiative transfer techniques, increases in computing power, and the availability of at least some of the necessary atomic data, it is now possible to consider the computation of detailed non-LTE model atmospheres in which the full effects of non-LTE line blanketing are taken into account. We discuss our own implementation of non-LTE line blanketing in a spherical non-LTE code developed for the investigation of objects with extended outflows. A partial linearization technique is used to simultaneously solve the radiative transfer equation in conjunction with the equations of statistical equilibrium. Convergence properties are similar to that obtained with an ``Optimal'' Approximate-Lambda Operator. CNO line blanketing has been incorporated without major difficulty, while Fe blanketing is currently being installed. Comparisons of model spectra with recent HST observations of an LMC WC star will be presented. When completed we anticipate the code will be applicable to the study of a wide range of phenomena exhibiting outflows including Luminous-Blue variables, Supernovae, Wold-Rayet stars and Novae. Partial support for this work was provided by NASA through grant Nos GO-5460.01-93A and GO-4550.01-92A from the Space Science Institute which is operated under the Association of Universities for Research in Astronomy, Inc., under NASA contract NAS5-26555. Support from NASA award NAGW-3828 is also gratefully acknowledged.

  17. 18 CFR 284.284 - Blanket certificates for unbundled sales services.

    Science.gov (United States)

    2010-04-01

    ... granted a blanket certificate of public convenience and necessity pursuant to section 7 of the Natural Gas... the sales customer to arrange for any pipeline-provided service necessary to deliver gas to the customer. (e) Small customer cost-based rate. A pipeline that provided bundled sales service to a...

  18. 78 FR 44558 - Stingray Pipeline Company, L.L.C.; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2013-07-24

    ... Energy Regulatory Commission Stingray Pipeline Company, L.L.C.; Notice of Request Under Blanket Authorization Take notice that on July 3, 2013, Stingray Pipeline Company, L.L.C. (Stingray), 1100 Louisiana... in the federal waters offshore Louisiana. Specifically, Stingray proposes to abandon, by sale, its...

  19. 76 FR 18216 - Southern Natural Gas Company; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2011-04-01

    ... Federal Energy Regulatory Commission Southern Natural Gas Company; Notice of Request Under Blanket Authorization Take notice that on March 16, 2011, Southern Natural Gas Company (Southern), Post Office Box 2563... and 157.216 of the Commission's Regulations under the Natural Gas Act (NGA) as amended, to abandon...

  20. 75 FR 3232 - Northern Natural Gas Company; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2010-01-20

    ... Energy Regulatory Commission Northern Natural Gas Company; Notice of Request Under Blanket Authorization January 8, 2010. Take notice that on December 30, 2009, Northern Natural Gas Company (Northern), 1111... sections 157.205 and 157.214 of the Commission's regulations under the Natural Gas Act for authorization...

  1. 75 FR 13535 - Northern Natural Gas Company; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2010-03-22

    ... Energy Regulatory Commission Northern Natural Gas Company; Notice of Request Under Blanket Authorization March 16, 2010. Take notice that on March 12, 2010, Northern Natural Gas Company (Northern), 1111 South... External Affairs, Northern Natural Gas Company, 1111 South 103rd Street, Omaha, Nebraska 68124, at...

  2. Acoustic contributions of a sound absorbing blanket placed in a double panel structure: absorption versus transmission.

    Science.gov (United States)

    Doutres, Olivier; Atalla, Noureddine

    2010-08-01

    The objective of this paper is to propose a simple tool to estimate the absorption vs. transmission loss contributions of a multilayered blanket unbounded in a double panel structure and thus guide its optimization. The normal incidence airborne sound transmission loss of the double panel structure, without structure-borne connections, is written in terms of three main contributions; (i) sound transmission loss of the panels, (ii) sound transmission loss of the blanket and (iii) sound absorption due to multiple reflections inside the cavity. The method is applied to four different blankets frequently used in automotive and aeronautic applications: a non-symmetric multilayer made of a screen in sandwich between two porous layers and three symmetric porous layers having different pore geometries. It is shown that the absorption behavior of the blanket controls the acoustic behavior of the treatment at low and medium frequencies and its transmission loss at high frequencies. Acoustic treatment having poor sound absorption behavior can affect the performance of the double panel structure.

  3. Analysis of Time-Dependent Tritium Breeding Capability of Water Cooled Ceramic Breeder Blanket for CFETR

    Science.gov (United States)

    Gao, Fangfang; Zhang, Xiaokang; Pu, Yong; Zhu, Qingjun; Liu, Songlin

    2016-08-01

    Attaining tritium self-sufficiency is an important mission for the Chinese Fusion Engineering Testing Reactor (CFETR) operating on a Deuterium-Tritium (D-T) fuel cycle. It is necessary to study the tritium breeding ratio (TBR) and breeding tritium inventory variation with operation time so as to provide an accurate data for dynamic modeling and analysis of the tritium fuel cycle. A water cooled ceramic breeder (WCCB) blanket is one candidate of blanket concepts for the CFETR. Based on the detailed 3D neutronics model of CFETR with the WCCB blanket, the time-dependent TBR and tritium surplus were evaluated by a coupling calculation of the Monte Carlo N-Particle Transport Code (MCNP) and the fusion activation code FISPACT-2007. The results indicated that the TBR and tritium surplus of the WCCB blanket were a function of operation time and fusion power due to the Li consumption in breeder and material activation. In addition, by comparison with the results calculated by using the 3D neutronics model and employing the transfer factor constant from 1D to 3D, it is noted that 1D analysis leads to an over-estimation for the time-dependent tritium breeding capability when fusion power is larger than 1000 MW. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB108004, 2015GB108002, and 2014GB119000), and by National Natural Science Foundation of China (No. 11175207)

  4. 32 CFR Appendix C to Part 327 - DeCA Blanket Routine Uses

    Science.gov (United States)

    2010-07-01

    .... 76-07. (h) Routine Use—Disclosure to the Office of Personnel Management. A record from a system of... Personnel Management (OPM) concerning information on pay and leave, benefits, retirement deduction, and any... Blanket Routine Uses (a) Routine Use—Law Enforcement. If a system of records maintained by a DoD...

  5. 75 FR 33803 - Sabine Pipe Line LLC; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2010-06-15

    ... Energy Regulatory Commission Sabine Pipe Line LLC; Notice of Request Under Blanket Authorization June 8, 2010. Take notice that on June 1, 2010, Sabine Pipe Line LLC (Sabine), 4800 Fournace Place, Bellaire...-free, (866) 208-3676 or TTY, (202) 502-8659. Specifically, Sabine proposes to abandon, in place,...

  6. 77 FR 53874 - The Dow Chemical Company; Application for Blanket Authorization To Export Previously Imported...

    Science.gov (United States)

    2012-09-04

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY The Dow... application (Application), filed on July 13, 2012, by The Dow Chemical Company (Dow), requesting blanket... on a short-term or spot market basis for a two-year period commencing on October 5, 2012.\\1\\...

  7. Salted lamb meat blanket of Petrolina-Pernambuco, Brazil: process and quality

    Directory of Open Access Journals (Sweden)

    Nely de Almeida Pedrosa

    2014-03-01

    Full Text Available Salted lamb meat blanket, originated from boning, salting, and drying of whole lamb carcass, was studied aiming at obtaining information that support the search for guarantees of origin for this typical regional product from the city of Petrolina-Pernambuco-Brazil. Data from three processing units were obtained, where it was observed the use of a traditional local technology that uses salting, an ancient preservation method; however, with a peculiar boning technique, resulting in a meat product with great potential for exploitation in the form of meat blanket. Based on the values of pH (6.22 ± 0.22, water activity (0.97 ± 0.02, and moisture (69.86 ± 2.26 lamb meat blanket is considered a perishable product, and consequently it requires the use of other preservation methods combined with salt, which along with the results of the microbiological analyses (absence of Salmonella sp, score <10 MPN/g of halophilic bacteria, total coliforms between 6.7 × 10³ and 5.2 × 10(6 FUC/g, and Staphylococcus from 8.1 × 10³ CFU/g at uncountable reinforce the need of hygienic practices to ensure product safety. These results, together with the product notoriety and the organization of the sector are important factors in achieving Geographical Indication of the Salted lamb Meat blanket of Petrolina.

  8. 78 FR 51182 - Sea Robin Pipeline Company, LLC; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2013-08-20

    ... Energy Regulatory Commission Sea Robin Pipeline Company, LLC; Notice of Request Under Blanket Authorization Take notice that on July 31, 2013, Sea Robin Pipeline Company, LLC (Sea Robin), P. O. Box 4967....205(b) and 157.216 of the Commission's Regulations under the Natural Gas Act (NGA), and Sea...

  9. 76 FR 31326 - Gulf LNG Pipeline, LLC; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2011-05-31

    ... Energy Regulatory Commission Gulf LNG Pipeline, LLC; Notice of Request Under Blanket Authorization Take notice that on May 18, 2011, Gulf LNG Pipeline, LLC (GLNG Pipeline), Colonial Brookwood Center, 569... to Margaret G. Coffman, Counsel, Gulf LNG Pipeline Company, LLC, Colonial Brookwood Center,...

  10. 75 FR 8327 - Golden Pass Pipeline LLC; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2010-02-24

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Golden Pass Pipeline LLC; Notice of Request Under Blanket Authorization February 17, 2010. Take notice that on October 29, 2009, Golden Pass Pipeline, LLC (GPPL), filed in...

  11. 77 FR 38622 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2012-06-28

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization Take notice that on June 4, 2012, Southern Star Central Gas Pipeline, Inc. (Southern Star), 4700 Highway 56, Owensboro,...

  12. 78 FR 68835 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2013-11-15

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization Take notice that on October 31, 2013, Southern Star Central Gas Pipeline, Inc. (Southern Star), 4700 Highway 56, Owensboro,...

  13. 78 FR 25264 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2013-04-30

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization Take notice that on April 16, 2013, Southern Star Central Gas Pipeline, Inc. (Southern Star), 4700 Highway 56, Owensboro,...

  14. 78 FR 53746 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2013-08-30

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization Take notice that on August 13, 2013, Southern Star Central Gas Pipeline, Inc. (Southern Star), 4700 Highway 56, Owensboro,...

  15. 77 FR 14517 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2012-03-12

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization Take notice that on February 21, 2012 Southern Star Central Gas Pipeline, Inc. (Southern Star), 4700 State Highway 56,...

  16. 78 FR 13663 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2013-02-28

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization Take notice that on February 11, 2013, Southern Star Central Gas Pipeline, Inc. (Southern Star), 4700 Highway 56, P.O. Box...

  17. 75 FR 8053 - Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2010-02-23

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Southern Star Central Gas Pipeline, Inc.; Notice of Request Under Blanket Authorization February 16, 2010. Take notice that on January 29, 2010, Southern Star Central Gas Pipeline, Inc. (Southern Star), 4700...

  18. 75 FR 17708 - Kinder Morgan Louisiana Pipeline LLC; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2010-04-07

    ... Energy Regulatory Commission Kinder Morgan Louisiana Pipeline LLC; Notice of Request Under Blanket Authorization March 30, 2010. Take notice that on March 25, 2010, Kinder Morgan Louisiana Pipeline LLC (KMLP... directed to Norman Watson, Director, Business Development, Kinder Morgan Louisiana Pipeline LLC, 500...

  19. 75 FR 35019 - Kinder Morgan Interstate Gas Transmission LLC; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2010-06-21

    ... Energy Regulatory Commission Kinder Morgan Interstate Gas Transmission LLC; Notice of Request Under Blanket Authorization June 11, 2010. Take notice that on June 3, 2009, Kinder Morgan Interstate Gas..., Kinder Morgan Interstate Gas Transmission LLC, P.O. Box 281304, Lakewood, Colorado 80228-8304, or...

  20. 75 FR 53966 - Kinder Morgan Interstate Gas Transmission, LLC; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2010-09-02

    ... Federal Energy Regulatory Commission Kinder Morgan Interstate Gas Transmission, LLC; Notice of Request Under Blanket Authorization August 27, 2010. Take notice that on August 25, 2010, Kinder Morgan Interstate Gas Transmission, LLC (Kinder Morgan), 370 Van Gordon Street, Lakewood, Colorado 80228-8304...

  1. 75 FR 45111 - Kinder Morgan Interstate Gas Transmission LLC; Notice of Request Under Blanket Authorization

    Science.gov (United States)

    2010-08-02

    ... Energy Regulatory Commission Kinder Morgan Interstate Gas Transmission LLC; Notice of Request Under Blanket Authorization July 26, 2010. Take notice that on July 20, 2010, Kinder Morgan Interstate Gas..., Vice President, Regulatory, Kinder Morgan Interstate Gas Transmission LLC, 370 Van Gordon...

  2. Stochastic modeling to determine the economic effects of blanket, selective, and no dry cow therapy

    NARCIS (Netherlands)

    Huijps, K.; Hogeveen, H.

    2007-01-01

    In many countries, blanket dry cow therapy (DCT) is the standard way to dry off cows. Because of concerns about antibiotic resistance, selective DCT is proposed as an alternative. The economic consequences of different types of DCT were studied previously, but variation between input traits and diff

  3. 77 FR 25711 - Cheniere Marketing, LLC; Application for Blanket Authorization To Export Previously Imported...

    Science.gov (United States)

    2012-05-01

    ... Cheniere Marketing, LLC; Application for Blanket Authorization To Export Previously Imported Liquefied... to export * * *'' \\9\\ \\6\\ Cheniere Marketing, LLC, DOE/FE Order No 2795 at 11. \\7\\ See Dominion Cove... application (Application), filed on March 30, 2012, by Cheniere Marketing, LLC (CMI), requesting...

  4. Newton—Like Iteration Method for Solving Algebraic Equations

    Institute of Scientific and Technical Information of China (English)

    JihuanHE

    1998-01-01

    In this paper,a Newton-like iteration method is proposed to solve an approximate solution of an algebraic equation.The iteration formula obtained by homotopy perturbation method contains the well-known Newton iteration formulain logic.

  5. Frozen Landweber Iteration for Nonlinear Ill-Posed Problems

    Institute of Scientific and Technical Information of China (English)

    J.Xu; B.Han; L.Li

    2007-01-01

    In this paper we propose a modification of the Landweber iteration termed frozen Landweber iteration for nonlinear ill-posed problems.A convergence analysis for this iteration is presented.The numerical performance of this frozen Landweber iteration for a nonlinear Hammerstein integral equation is compared with that of the Landweber iteration.We obtain a shorter running time of the frozen Landweber iteration based on the same convergence accuracy.

  6. Halpern Iteration in CAT(κ) Spaces

    Institute of Scientific and Technical Information of China (English)

    Bo(z)ena PI(A)TEK

    2011-01-01

    In this paper we show that an iterative sequence generated by the Halpern algorithm converges to a fixed point in the case of complete CAT(κ) spaces. Similar results for Hadamard manifolds were obtained in[Li,C.,López, G., Martín-Márquez, V.:Iterative algorithms for nonexpansive mappings on Hadamard manifolds. Taiwanese J. Math., 14, 541-559 (2010)], but we study a much more general case. Moreover, we discuss the Halpern iteration procedure for set-valued mappings.

  7. Iterative restoration algorithms for nonlinear constraint computing

    Science.gov (United States)

    Szu, Harold

    A general iterative-restoration principle is introduced to facilitate the implementation of nonlinear optical processors. The von Neumann convergence theorem is generalized to include nonorthogonal subspaces which can be reduced to a special orthogonal projection operator by applying an orthogonality condition. This principle is shown to permit derivation of the Jacobi algorithm, the recursive principle, the van Cittert (1931) deconvolution method, the iteration schemes of Gerchberg (1974) and Papoulis (1975), and iteration schemes using two Fourier conjugate domains (e.g., Fienup, 1981). Applications to restoring the image of a double star and division by hard and soft zeros are discussed, and sample results are presented graphically.

  8. Equalization characteristics of an upflow sludge blanket-aerated biofilter (USB-AF) system.

    Science.gov (United States)

    Jun, H B; Park, S M; Park, J K; Lee, S H

    2005-01-01

    Equalization characteristics of the upflow sludge blanket-aerated bio-filter (USB-AF) were investigated with the fluctuated raw domestic sewage. Recycle of nitrified effluent from AF to USB triggered the equalization characteristics of the sludge blanket on both soluble and particulate organic matter. Increment of EPS in sludge blanket by nitrate recycle was detected and removal of turbidity and particulates increased at higher recycle ratios by bio-flocculation. Increased TCOD removal in the USB was due to both denitrification of recycled nitrate and entrapment of the particulate organic matter in sludge blanket. Capture of both soluble and particulate organic matter increased sludge blanket layer in the USB, which improved the reactor performances and reduced the organic load on the subsequent AF. Overall TCOD and SS removal efficiencies were about 98% and 96%, respectively in the USB-AF system. Turbidity in the USB effluent was about 44, 20 and 5.5 NTU, at recycle ratios of 0, 100 and 200%, respectively. Particle counts in the range 2-4 microm in the USB effluent were higher than those in influent without nitrate recycle, while particle counts in the range of 0.5-15 microm in the USB effluent decreased 70% at recycle ratio of 200%. The major constituent of EPS extracted from anaerobic sludge was protein and total EPS increased from 109.1 to 165.7 mg/g-VSS with nitrate recycle of 100%. Removal efficiency and concentration of T-N in the UBS-AF effluent was over 70% and below 16 mg/L, respectively.

  9. FEPI-MB: identifying SNPs-disease association using a Markov Blanket-based approach

    Directory of Open Access Journals (Sweden)

    Han Bing

    2011-11-01

    Full Text Available Abstract Background The interactions among genetic factors related to diseases are called epistasis. With the availability of genotyped data from genome-wide association studies, it is now possible to computationally unravel epistasis related to the susceptibility to common complex human diseases such as asthma, diabetes, and hypertension. However, the difficulties of detecting epistatic interaction arose from the large number of genetic factors and the enormous size of possible combinations of genetic factors. Most computational methods to detect epistatic interactions are predictor-based methods and can not find true causal factor elements. Moreover, they are both time-consuming and sample-consuming. Results We propose a new and fast Markov Blanket-based method, FEPI-MB (Fast EPistatic Interactions detection using Markov Blanket, for epistatic interactions detection. The Markov Blanket is a minimal set of variables that can completely shield the target variable from all other variables. Learning of Markov blankets can be used to detect epistatic interactions by a heuristic search for a minimal set of SNPs, which may cause the disease. Experimental results on both simulated data sets and a real data set demonstrate that FEPI-MB significantly outperforms other existing methods and is capable of finding SNPs that have a strong association with common diseases. Conclusions FEPI-MB algorithm outperforms other computational methods for detection of epistatic interactions in terms of both the power and sample-efficiency. Moreover, compared to other Markov Blanket learning methods, FEPI-MB is more time-efficient and achieves a better performance.

  10. Structural analysis of the ITER Vacuum Vessel regarding 2012 ITER Project-Level Loads

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, J.-M., E-mail: jean-marc.martinez@live.fr [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul lez Durance (France); Jun, C.H.; Portafaix, C.; Choi, C.-H.; Ioki, K.; Sannazzaro, G.; Sborchia, C. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul lez Durance (France); Cambazar, M.; Corti, Ph.; Pinori, K.; Sfarni, S.; Tailhardat, O. [Assystem EOS, 117 rue Jacquard, L' Atrium, 84120 Pertuis (France); Borrelly, S. [Sogeti High Tech, RE2, 180 rue René Descartes, Le Millenium – Bat C, 13857 Aix en Provence (France); Albin, V.; Pelletier, N. [SOM Calcul – Groupe ORTEC, 121 ancien Chemin de Cassis – Immeuble Grand Pré, 13009 Marseille (France)

    2014-10-15

    Highlights: • ITER Vacuum Vessel is a part of the first barrier to confine the plasma. • ITER Vacuum Vessel as Nuclear Pressure Equipment (NPE) necessitates a third party organization authorized by the French nuclear regulator to assure design, fabrication, conformance testing and quality assurance, i.e. Agreed Notified Body (ANB). • A revision of the ITER Project-Level Load Specification was implemented in April 2012. • ITER Vacuum Vessel Loads (seismic, pressure, thermal and electromagnetic loads) were summarized. • ITER Vacuum Vessel Structural Margins with regards to RCC-MR code were summarized. - Abstract: A revision of the ITER Project-Level Load Specification (to be used for all systems of the ITER machine) was implemented in April 2012. This revision supports ITER's licensing by accommodating requests from the French regulator to maintain consistency with the plasma physics database and our present understanding of plasma transients and electro-magnetic (EM) loads, to investigate the possibility of removing unnecessary conservatism in the load requirements and to review the list and definition of incidental cases. The purpose of this paper is to present the impact of this 2012 revision of the ITER Project-Level Load Specification (LS) on the ITER Vacuum Vessel (VV) loads and the main structural margins required by the applicable French code, RCC-MR.

  11. Clinical applications of iterative reconstruction

    Energy Technology Data Exchange (ETDEWEB)

    Eberl, S. [Royal Prince Alfred Hospital, Camperdown, NSW (Australia). Department of PET and Nuclear Medicine

    1998-03-01

    Expectation maximisation (EM) reconstruction largely eliminates the hot and cold streaking artifacts characteristic of filtered-back projection (FBP) reconstruction around localised hot areas, such as the bladder. It also substantially reduces the problem of decreased inferior wall counts in MIBI myocardial perfusion studies due to ``streaking`` from high liver uptake. Non-uniform attenuation and scatter correction, resolution recovery, anatomical information, e.g. from MRI or CT tracer kinetic modelling, can all be built into the EM reconstruction imaging model. The properties of ordered subset EM (OSEM) have also been used to correct for known patient motion as part of the reconstruction process. These uses of EM are elaborated more fully in some of the other abstracts of this meeting. Currently we use OSEM routinely for: (i) studies where streaking is a problem, including all MIBI myocardial perfusion studies, to avoid hot liver inferior wall artifact, (ii) all whole body FDG PET, all lung V/Q SPECT (which have a short acquisition time) and all gated {sup 201}TI myocardial perfusion studies due to improved noise characteristics of OSEM in these studies; (iii) studies with measured, non-uniform attenuation correction. With the accelerated OSEM algorithm, iterative reconstruction is practical for routine clinical applications and we have found OSEM to provide clearly superior reconstructions for the areas listed above and are investigating its application to other studies. In clinical use, we have not found OSEM to introduce artifacts which would not also occur with FBP, e.g. uncorrected patient motion will cause artifacts with both OSEM and FBP.

  12. The Iterative Signature Algorithm for the analysis of large scale gene expression data

    CERN Document Server

    Bergmann, S R; Barkai, N; Bergmann, Sven; Ihmels, Jan; Barkai, Naama

    2003-01-01

    We present a new approach for the analysis of genome-wide expression data. Our method is designed to overcome the limitations of traditional techniques, when applied to large-scale data. Rather than alloting each gene to a single cluster, we assign both genes and conditions to context-dependent and potentially overlapping transcription modules. We provide a rigorous definition of a transcription module as the object to be retrieved from the expression data. An efficient algorithm, that searches for the modules encoded in the data by iteratively refining sets of genes and conditions until they match this definition, is established. Each iteration involves a linear map, induced by the normalized expression matrix, followed by the application of a threshold function. We argue that our method is in fact a generalization of Singular Value Decomposition, which corresponds to the special case where no threshold is applied. We show analytically that for noisy expression data our approach leads to better classificatio...

  13. Status and verification strategy for ITER neutronics

    Energy Technology Data Exchange (ETDEWEB)

    Loughlin, Michael, E-mail: michael.loughlin@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Angelone, Maurizio [Associazione EURATOM-ENEA Sulla Fusione, Via E. Fermi 45, I-00044 Frascati, Roma (Italy); Batistoni, Paola [Associazione EURATOM-ENEA Sulla Fusione, Via E. Fermi 45, I-00044 Frascati, Roma (Italy); JET-EFDA, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Bertalot, Luciano [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Eskhult, Jonas [Studsvik Nuclear AB, SE-611 Nyköping (Sweden); Konno, Chikara [Japan Atomic Energy Agency Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan); Pampin, Raul [F4E Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, Barcelona 08019 (Spain); Polevoi, Alexei; Polunovskiy, Eduard [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2014-10-15

    The paper summarizes the current status of neutronics at ITER and a first set of proposals for experimental programmes to be conducted in the early operational life-time of ITER are described for the more crucial areas. These include a TF coils heating benchmark, a streaming benchmark and streaming measurements by activation on ITER itself. Also on ITER the measurement of activated water from triton burn-up should be planned and performed. This will require the measurement of triton burn-up in DD phase. Measurements of neutron flux in the tokamak building during DD operations should also be carried out. The use of JET for verification of shut down dose rate estimates is desirable. Other facilities to examine the production and behaviour of activated corrosion products and the shielding properties of concretes to high energy (6 MeV) gamma-rays are recommended.

  14. Accelerating Iterative Big Data Computing Through MPI

    Institute of Scientific and Technical Information of China (English)

    梁帆; 鲁小亿

    2015-01-01

    Current popular systems, Hadoop and Spark, cannot achieve satisfied performance because of the inefficient overlapping of computation and communication when running iterative big data applications. The pipeline of computing, data movement, and data management plays a key role for current distributed data computing systems. In this paper, we first analyze the overhead of shuffle operation in Hadoop and Spark when running PageRank workload, and then propose an event-driven pipeline and in-memory shuffle design with better overlapping of computation and communication as DataMPI-Iteration, an MPI-based library, for iterative big data computing. Our performance evaluation shows DataMPI-Iteration can achieve 9X∼21X speedup over Apache Hadoop, and 2X∼3X speedup over Apache Spark for PageRank and K-means.

  15. Anderson Acceleration for Fixed-Point Iterations

    Energy Technology Data Exchange (ETDEWEB)

    Walker, Homer F. [Worcester Polytechnic Institute, MA (United States)

    2015-08-31

    The purpose of this grant was to support research on acceleration methods for fixed-point iterations, with applications to computational frameworks and simulation problems that are of interest to DOE.

  16. On the safety of ITER accelerators.

    Science.gov (United States)

    Li, Ge

    2013-01-01

    Three 1 MV/40A accelerators in heating neutral beams (HNB) are on track to be implemented in the International Thermonuclear Experimental Reactor (ITER). ITER may produce 500 MWt of power by 2026 and may serve as a green energy roadmap for the world. They will generate -1 MV 1 h long-pulse ion beams to be neutralised for plasma heating. Due to frequently occurring vacuum sparking in the accelerators, the snubbers are used to limit the fault arc current to improve ITER safety. However, recent analyses of its reference design have raised concerns. General nonlinear transformer theory is developed for the snubber to unify the former snubbers' different design models with a clear mechanism. Satisfactory agreement between theory and tests indicates that scaling up to a 1 MV voltage may be possible. These results confirm the nonlinear process behind transformer theory and map out a reliable snubber design for a safer ITER.

  17. Archimedes' Pi--An Introduction to Iteration.

    Science.gov (United States)

    Lotspeich, Richard

    1988-01-01

    One method (attributed to Archimedes) of approximating pi offers a simple yet interesting introduction to one of the basic ideas of numerical analysis, an iteration sequence. The method is described and elaborated. (PK)

  18. Stability of Jungck-type iterative procedures

    Directory of Open Access Journals (Sweden)

    S. L. Singh

    2005-01-01

    Full Text Available We introduce and discuss the stability of Jungck and Jungck-Mann iterative procedures for a pair of Jungck-Osilike-type maps on an arbitrary set with values in a metric or linear metric space.

  19. Progress in the Integrated Simulation of Thermal-Hydraulic Operation of the ITER Magnet System

    CERN Document Server

    Bagnasco, M; Bessette, D; Marinucci, C

    2010-01-01

    A new integrated computer code is being developed for the simulations of the overall behavior of the ITER magnet cryo-system. The existing THEA, FLOWER and POWER codes, assembled as modules of a computational environment (Super-Magnet) have been upgraded to perform global simulations of the cooling circuit for the ITER magnet system. The thermal coupling resulting from the generic geometric configurations has been implemented to realize quasi-three-dimensional simulations of the winding pack. In this paper we present details on the model.

  20. Efficient iterative adaptive pole placement algorithm

    Institute of Scientific and Technical Information of China (English)

    李俊民; 李靖; 杨磊

    2004-01-01

    An iterative adaptive pole placement algorithm is presented. The stability and the convergence of the algorithm are respectively established. Since one-step iterative formulation in computing controller's parameters is used, the on-line computation cost is greatly reduced with respected to the traditional algorithm. The algorithm with the feed-forward can follow arbitrarily bounded output. The algorithm is also extended to multivariate case. Simulation examples show the efficiency and robustness of the algorithm.

  1. Accelerated Schwarz iterations for Helmholtz equation

    Science.gov (United States)

    Nagid, Nabila; Belhadj, Hassan; Amattouch, Mohamed Ridouan

    2017-01-01

    In this paper, the Restricted additive Schwarz (RAS) method is applied to solve Helmholtz equation. To accelerate the RAS iterations, we propose to apply the vector ɛ-algorithm. Some convergence analysis of the proposed method is presented, and applied succeffully to Helmholtz problem. The obtained results show the efficiency of the proposed approach. Moreover, the algorithm yields much faster convergence than the classical Schwarz iterations.

  2. Iterative consolidation of unorganized point clouds.

    Science.gov (United States)

    Liu, Shengjun; Chan, Kwan-Chung; Wang, Charlie C L

    2012-01-01

    Unorganized point clouds obtained from 3D shape acquisition devices usually present noise, outliers, and nonuniformities. The proposed framework consolidates unorganized points through an iterative procedure of interlaced downsampling and upsampling. Selection operations remove outliers while preserving geometric details. The framework improves the uniformity of points by moving the downsampled particles and refining point samples. Surface extrapolation fills missed regions. Moreover, an adaptive sampling strategy speeds up the iterations. Experimental results demonstrate the framework's effectiveness.

  3. Novel aspects of plasma control in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Humphreys, D.; Jackson, G.; Walker, M.; Welander, A. [General Atomics P.O. Box 85608, San Diego, California 92186-5608 (United States); Ambrosino, G.; Pironti, A. [CREATE/University of Naples Federico II, Napoli (Italy); Vries, P. de; Kim, S. H.; Snipes, J.; Winter, A.; Zabeo, L. [ITER Organization, St. Paul Lez durance Cedex (France); Felici, F. [Eindhoven University of Technology, Eindhoven (Netherlands); Kallenbach, A.; Raupp, G.; Treutterer, W. [Max-Planck Institut für Plasmaphysik, Garching (Germany); Kolemen, E. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States); Lister, J.; Sauter, O. [Centre de Recherches en Physique des Plasmas, Ecole Polytechnique Federale de Lausanne, Lausanne (Switzerland); Moreau, D. [CEA, IRFM, 13108 St. Paul-lez Durance (France); Schuster, E. [Lehigh University, Bethlehem, Pennsylvania (United States)

    2015-02-15

    ITER plasma control design solutions and performance requirements are strongly driven by its nuclear mission, aggressive commissioning constraints, and limited number of operational discharges. In addition, high plasma energy content, heat fluxes, neutron fluxes, and very long pulse operation place novel demands on control performance in many areas ranging from plasma boundary and divertor regulation to plasma kinetics and stability control. Both commissioning and experimental operations schedules provide limited time for tuning of control algorithms relative to operating devices. Although many aspects of the control solutions required by ITER have been well-demonstrated in present devices and even designed satisfactorily for ITER application, many elements unique to ITER including various crucial integration issues are presently under development. We describe selected novel aspects of plasma control in ITER, identifying unique parts of the control problem and highlighting some key areas of research remaining. Novel control areas described include control physics understanding (e.g., current profile regulation, tearing mode (TM) suppression), control mathematics (e.g., algorithmic and simulation approaches to high confidence robust performance), and integration solutions (e.g., methods for management of highly subscribed control resources). We identify unique aspects of the ITER TM suppression scheme, which will pulse gyrotrons to drive current within a magnetic island, and turn the drive off following suppression in order to minimize use of auxiliary power and maximize fusion gain. The potential role of active current profile control and approaches to design in ITER are discussed. Issues and approaches to fault handling algorithms are described, along with novel aspects of actuator sharing in ITER.

  4. A steerable ECRF launcher for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Grunloh, H.; Prater, R.; Doane, J.L.; Moeller, C.P. [General Atomics, San Diego (United States); Makowski, M. [ITER Joint Work Site, Garching (Germany)

    1998-07-01

    A design is proposed to steer the electron cyclotron heating and current drive power for ITER using rotatable, water-cooled mirrors and long-pressure hydraulic actuators, and to accommodate changes in length of the waveguide when the temperatures of the vacuum vessel and the cryostat change using waveguide bellows. An alternative concept is also introduced that requires no moving parts within the ITER cryostat and that utilizes wave reconstruction within the waveguide to effect the steering. (author)

  5. Experiences with iterated traffic microsimulations in Dallas

    CERN Document Server

    Nagel, K

    1997-01-01

    This paper reports experiences with iterated traffic microsimulations in the context of a Dallas study. ``Iterated microsimulations'' here means that the information generated by a microsimulation is fed back into the route planner so that the simulated individuals can adjust their routes to circumvent congestion. This paper gives an overview over what has been done in the Dallas context to better understand the relaxation process, and how to judge the robustness of the results.

  6. Development and test of prototype components for ITER; Entwicklung und Test von Prototypkomponenten fuer ITER

    Energy Technology Data Exchange (ETDEWEB)

    Biel, Wolfgang; Behr, Wilfried; Castano-Bardawil, David; and others

    2015-08-15

    The scientific program of the project is divided into the following partial projects: (1.) ITER Diagnostic Port Plug for the charge-exchange spectroscopy (CXRS) with the subthemes: (a) Development of prototypes for critical mechanical components, (b) development of a roboter for the laser welding of vacuum seals and pipings at the Port Plug, (c) mirror studies, (d) CXRS prototype spectrometer, (2.) ITER tritium retention diagnostics (TR), (3.) ITER disruption mitigation ventile (DMV).

  7. Nitrogen segregation and blister formation of 316LN austenitic steel during electron beam welding tests for ITER gravity supports

    Science.gov (United States)

    Lee, P. Y.; Hou, B. L.; Wu, J. H.; Yang, D.; Zhang, G. R.; Zhang, C. P.

    2009-04-01

    316LN has been widely applied in the design of ITER components, such as shield blanket and gravity supports, due to its excellent corrosion resistance and high strength. The behavior of nitrogen in this steel during welding is important for the mechanical properties of the components. In this study, a focused 150 kv high voltage electron beam with 300 mA beam current has been used to weld 316LN steel under vacuum condition. The microstructure and composition of the welding area were observed and analyzed. The influence of welding on the shock resistance and tensile strength at both room temperature and low temperature were examined. It was found that the mechanical properties are strongly related to the defects formed in the welding process.

  8. ITER Remote Maintenance System (IRMS) lifecycle management

    Energy Technology Data Exchange (ETDEWEB)

    Tesini, Alessandro, E-mail: alessandro.tesini@iter.org [ITER Organization, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Otto' , Bede [Oxford Technologies Ltd, 7, Nuffield Way, Abingdon, Oxon OX14 1RJ (United Kingdom); Blight, John [FAAST 31c Allee de la Granette, 13600 Ceyreste (France); Choi, Chang-Hwan; Friconneau, Jean-Pierre; Gotewal, Krishan Kumar; Hamilton, David [ITER Organization, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Heckendorn, Frank [FD Technologies, PO Box 6686, Aiken, SC (United States); Martins, Jean-Pierre [ITER Organization, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Marty, Thomas [Westinghouse, 122, avenue de Hambourg, 13008 Marseille (France); Nakahira, Masataka; Palmer, Jim; Subramanian, Rajendran [ITER Organization, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

    2011-10-15

    The availability of the ITER machine to perform its scientific program is strongly dependent on the performance of the different Remote Handling (RH) systems constituting the ITER Remote Maintenance System (IRMS). The lifecycle of the IRMS will largely exceed 40 years from initial concept design and proof testing through to machine decommissioning. Such a long lifecycle requires that a rigorous approach is put in place to guarantee the technical capabilities of the highly innovative IRMS, its efficiency and its availability. For this purpose, an IRMS System Engineering and IRMS lifecycle management approach has been adopted by ITER. The approach aims at ensuring the IRMS full operability and availability at an acceptable cost of ownership over the full ITER machine assembly and operations period. The IRMS lifecycle management method described in this paper covers such subjects as specific requirements for IRMS design reviews, monitoring during manufacture, factory and site acceptance testing, integrated commissioning, decontamination, maintenance and re-qualification strategies, requirements for Integrated Logistical Support during operations. The updating and implementation of the IRMS lifecycle strategy and this procedure will be managed and monitored by the Remote Handling Integrated Product Team (RH-IPT). Although developed for the IRMS, the basic principles and procedures of lifecycle management could be applied to other ITER plant systems whose reliability and availability will be essential for the continued operation of the ITER machine.

  9. Exploring climatic controls on blanket bog litter decomposition across an altitudinal gradient

    Science.gov (United States)

    Bell, Michael; Ritson, Jonathan P.; Clark, Joanna M.; Verhoef, Anne; Brazier, Richard E.

    2016-04-01

    The hydrological and ecological functioning of blanket bogs is strongly coupled, involving multiple ecohydrological feedbacks which can affect carbon cycling. Cool and wet conditions inhibit decomposition, and favour the growth of Sphagnum mosses which produce highly recalcitrant litter. A small but persistent imbalance between production and decomposition has led to blanket bogs in the UK accumulating large amounts of carbon. Additionally, healthy bogs provide a suite of other ecosystems services including water regulation and drinking water provision. However, there is concern that climate change could increase rates of litter decomposition and disrupt this carbon sink. Furthermore, it has been argued that the response of these ecosystems in the warmer south west and west of the UK may provide an early analogue for later changes in the more extensive northern peatlands. In order to investigate the effects of climate change on blanket bog litter decomposition, we set-up a litter bag experiment across an altitudinal gradient spanning 200 m of elevation (including a transition from moorland to healthy blanket bog) on Dartmoor, an area of hitherto unstudied, climatically marginal blanket bog in the south west of the UK. At seven sites, water table depth and soil and surface temperature were recorded continuously. Litter bags filled with the litter of three vegetation species dominant on Dartmoor were incubated just below the bog surface and retrieved over a period of 12 months. We found significant differences in the rate of decomposition between species. At all sites, decomposition progressed in the order Calluna vulgaris (dwarf shrub) > Molinia caerulea (graminoid) > Sphagnum (bryophyte). However, while soil temperature did decrease along the altitudinal gradient, being warmer in the lower altitudes, a hypothesised accompanying decrease in decomposition rates did not occur. This could be explained by greater N deposition at the higher elevation sites (estimated

  10. ITER neutral beam system US conceptual design. Final vesion

    Energy Technology Data Exchange (ETDEWEB)

    Purgalis, P.

    1990-09-01

    In this document we present the US conceptual design of a neutral beam system for International Thermonuclear Experimental Reactor (ITER). The design incorporates a barium surface conversion D{sup {minus}} source feeding a linear array of accelerator channels. The system uses a dc accelerator with electrostatic quadrupoles for strong focusing. A high voltage power supply that is integrated with the accelerator is presented as an attractive option. A gas neutralizer is used and residual ions exiting the neutralizer are deflected to water-cooled dumps. Cryopanels are located at the accelerator exit to pump excess gas from the source and the neutralizer, and in the ion dump cavity to pump re-neutralized ions and neutralizer gas. All the above components are packaged in compact identical, independent modules which can be removed for remote maintenance. The neutral beam system delivers 75 MW of DO at 1.3 MeV, into three ports with a total of 9 modules arranged in stacks of three modules per port . To increase reliability each module is designed to deliver up to 10 MW; this allows eight modules operating at partial capacity to deliver the required power in the event one module is out of service, and provides 20% excess capacity to improve availability. Radiation protection is provided by shielding and by locating critical components in the source and accelerator 46.5 m from the torus centerline. Neutron shielding in the drift duct and neutralizer provides the added feature of limiting conductance and thus reducing gas flow to and from the torus.

  11. Upflow anaerobic sludge blanket reactor--a review.

    Science.gov (United States)

    Bal, A S; Dhagat, N N

    2001-04-01

    . Concentrated waste (usually sewage sludge) can be added continuously or periodically (semi-batch operation), where it is mixed with the contents of the reactor. Theoretically, the conventional digester is operated as a once-through, completely mixed, reactor. In this particular mode of operation the hydraulic retention time (HRT) is equal to the solids retention time (SRT). Basically, the required process efficiency is related to the sludge retention time (SRT), and hence longer SRT provided, results in satisfactory population (by reproduction) for further waste stabilization. By reducing the hydraulic retention time (HRT) in the conventional mode reactor, the quantity of biological solids within the reactor is also decreased as the solids escape with the effluent. The limiting HRT is reached when the bacteria are removed from the reactor faster than they can grow. Methanogenic bacteria are slow growers and are considered the rate-limiting component in the anaerobic digestion process. The first anaerobic process developed, which separated the SRT from the HRT was the anaerobic contact process. In 1963, Young and McCarty (1968) began work, which eventually led to the development of the anaerobic upflow filter (AF) process. The anaerobic filter represented a significant advance in anaerobic waste treatment, since the filter can trap and maintain a high concentration of biological solids. By trapping these solids, long SRT's could be obtained at large waste flows, necessary to anaerobically treat low strength wastes at nominal temperatures economically. Another anaerobic process which relies on the development of biomass on the surfaces of a media is an expanded bed upflow reactor. The primary concept of the process consists of passing wastewater up through a bed of inert sand sized particles at sufficient velocities to fluidize and partially expand the sand bed. One of the more interesting new processes is the upflow anaerobic sludge blanket process (UASB), which was developed

  12. Purchasing. They Got Out From Under Blanket Orders With a Stores-Based Buy-Out System

    Science.gov (United States)

    Deal, Ralph E.

    1974-01-01

    By having a university 'stores' set up with a blanket order with about 115 local vendors, a university has eliminated petty cash disbursements for small purchases and proliferating purchase orders. (Author/PG)

  13. A study on the enhancement of the reliability in gravure offset roll printing with blanket swelling control

    Science.gov (United States)

    Eul Kim, Ga; Woo, Kyoohee; Kang, Dongwoo; Jang, Yunseok; Choi, Young-Man; Lee, Moon G.; Lee, Taik-Min; Kwon, Sin

    2016-10-01

    In roll-offset printing (patterning) technology with a PDMS blanket as a transfer medium, one of the major reliability issues is the occurrence of swelling, which involves absorption of the ink solvent in the printing blanket with repeated printing. This study developed a method to resolve blanket swelling in gravure offset roll printing and performed experiments for performance verification. The physical phenomena of mass and heat transfer were applied to fabricate a device based on convection drying. The proposed device managed to effectively control blanket swelling through drying by blowing air and additional temperature control. The experiments verified that printing quality (in particular the variation of the width of printed patterns) was maintained over 500 continuous printing.

  14. CORSICA modelling of ITER hybrid operation scenarios

    Science.gov (United States)

    Kim, S. H.; Bulmer, R. H.; Campbell, D. J.; Casper, T. A.; LoDestro, L. L.; Meyer, W. H.; Pearlstein, L. D.; Snipes, J. A.

    2016-12-01

    The hybrid operating mode observed in several tokamaks is characterized by further enhancement over the high plasma confinement (H-mode) associated with reduced magneto-hydro-dynamic (MHD) instabilities linked to a stationary flat safety factor (q ) profile in the core region. The proposed ITER hybrid operation is currently aiming at operating for a long burn duration (>1000 s) with a moderate fusion power multiplication factor, Q , of at least 5. This paper presents candidate ITER hybrid operation scenarios developed using a free-boundary transport modelling code, CORSICA, taking all relevant physics and engineering constraints into account. The ITER hybrid operation scenarios have been developed by tailoring the 15 MA baseline ITER inductive H-mode scenario. Accessible operation conditions for ITER hybrid operation and achievable range of plasma parameters have been investigated considering uncertainties on the plasma confinement and transport. ITER operation capability for avoiding the poloidal field coil current, field and force limits has been examined by applying different current ramp rates, flat-top plasma currents and densities, and pre-magnetization of the poloidal field coils. Various combinations of heating and current drive (H&CD) schemes have been applied to study several physics issues, such as the plasma current density profile tailoring, enhancement of the plasma energy confinement and fusion power generation. A parameterized edge pedestal model based on EPED1 added to the CORSICA code has been applied to hybrid operation scenarios. Finally, fully self-consistent free-boundary transport simulations have been performed to provide information on the poloidal field coil voltage demands and to study the controllability with the ITER controllers. Extended from Proc. 24th Int. Conf. on Fusion Energy (San Diego, 2012) IT/P1-13.

  15. ITER Fast Plant System Controller prototype based on PXIe platform

    Energy Technology Data Exchange (ETDEWEB)

    Ruiz, M., E-mail: mariano.ruiz@upm.es [Grupo de Investigacion en Instrumentacion y Acustica Aplicada, CAEND CSIC-UPM Universidad Politecnica de Madrid, Crta. Valencia Km-7, Madrid 28031 (Spain); Vega, J.; Castro, R. [Asociacion EURATOM/CIEMAT para Fusion, Madrid (Spain); Sanz, D.; Lopez, J.M.; Arcas, G. de; Barrera, E.; Nieto, J. [Grupo de Investigacion en Instrumentacion y Acustica Aplicada, CAEND CSIC-UPM Universidad Politecnica de Madrid, Crta. Valencia Km-7, Madrid 28031 (Spain); Goncalves, B.; Sousa, J.; Carvalho, B. [Associacao EURATOM/IST, Lisbon (Portugal); Utzel, N.; Makijarvi, P. [ITER Organization, CS 90 046, 13067 St. Paul lez Durance Cedex (France)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer Implementation of Fast Plant System Controller (FPSC) for ITER CODAC. Black-Right-Pointing-Pointer Efficient data acquisition and data movement using EPICS. Black-Right-Pointing-Pointer Performance of PCIe technologies in the implementation of FPSC. - Abstract: The ITER Fast Plant System Controller (FPSC) is based on embedded technologies. The FPSC will be devoted to both data acquisition tasks (sampling rates higher than 1 kHz) and control purposes (feedback loop actuators). Some of the essential requirements of these systems are: (a) data acquisition and data preprocessing; (b) interfacing with different networks and high speed links (Plant Operation Network, timing network based on IEEE1588, synchronous data transference and streaming/archiving networks); and (c) system setup and operation using EPICS (Experimental Physics and Industrial Control System) process variables. CIEMAT and UPM have implemented a prototype of FPSC using a PXIe (PCI eXtension for Instrumentation) form factor in a R and D project developed in two phases. The paper presents the main features of the two prototypes developed that have been named alpha and beta. The former was implemented using LabVIEW development tools as it was focused on modeling the FPSC software modules, using the graphical features of LabVIEW applications, and measuring the basic performance in the system. The alpha version prototype implements data acquisition with time-stamping, EPICS monitoring using waveform process variables (PVs), and archiving. The beta version prototype is a complete IOC implemented using EPICS with different software functional blocks. These functional blocks are integrated and managed using an ASYN driver solution and provide the basic functionalities required by ITER FPSC such as data acquisition, data archiving, data pre-processing (using both CPU and GPU) and streaming.

  16. Retractable tube design issues in ITER CXRS UPP no. 3

    Energy Technology Data Exchange (ETDEWEB)

    Baross, T., E-mail: baross.teteny@wigner.mta.hu [WIGNER RCP, RMKI, PO Box 91, H-1521 Budapest (Hungary); Biel, W. [Institut für Energieforschung – Plasmaphysik, Forschungszentrum Jülich Gmbh, Association EURATOM-FZJ, Member of Trilateral Euregio Cluster, 52425 Jülich (Germany); Krejczinger, A. [WIGNER RCP, RMKI, PO Box 91, H-1521 Budapest (Hungary); Krasikov, Yu.; Panin, A. [Institut für Energieforschung – Plasmaphysik, Forschungszentrum Jülich Gmbh, Association EURATOM-FZJ, Member of Trilateral Euregio Cluster, 52425 Jülich (Germany)

    2013-10-15

    Highlights: ► The retractable tube is located in the upper port plug no. 3 of ITER core charge exchange recombination spectroscopy (cCXRS) that carries the cleaning system of the first mirror. ► Various manufacturing methods were investigated for the proper layout of the tube cooling structure. ► The tube support boundary conditions were examined. ► A flexible compensator element was developed between the floating and structural flange of the tube. -- Abstract: The retractable tube is located in the upper port plug no. 3 of ITER core charge exchange recombination spectroscopy (cCXRS), where it is surrounded by a set of mirrors, diagnostic shielding module (DSM), outer shell, cleaning device and shutter. Accordingly to the recent port plug concept the tube is an optional unit that carries the cleaning system of the first mirror. The cleaning device shall be a compact instrument at the tube head that influences the tube diameter primarily. Independently of the cleaning device the tube cooling channels and its proper support concept has to be developed. Various manufacturing methods are investigated for the proper layout of the tube cooling structure. They are the gun-drilled deep holes, the solid state bonded sheets forming cooling channels between the welded layers, ‘tube-in-tube’ approach. Most of them were analyzed in view of their thermal conditions. The tube support system design is a complex task, that has to take into account several boundary conditions and limitations. The tube must be compatible with the assumed ITER generic installation procedure, withstand the electromagnetic (EM) and thermal loads. Furthermore, the tube has to be exchanged several times during its lifetime. The paper summarizes the tube cooling layout possibilities and the assumptions on its support concepts.

  17. PREFACE: Progress in the ITER Physics Basis

    Science.gov (United States)

    Ikeda, K.

    2007-06-01

    I would firstly like to congratulate all who have contributed to the preparation of the `Progress in the ITER Physics Basis' (PIPB) on its publication and express my deep appreciation of the hard work and commitment of the many scientists involved. With the signing of the ITER Joint Implementing Agreement in November 2006, the ITER Members have now established the framework for construction of the project, and the ITER Organization has begun work at Cadarache. The review of recent progress in the physics basis for burning plasma experiments encompassed by the PIPB will be a valuable resource for the project and, in particular, for the current Design Review. The ITER design has been derived from a physics basis developed through experimental, modelling and theoretical work on the properties of tokamak plasmas and, in particular, on studies of burning plasma physics. The `ITER Physics Basis' (IPB), published in 1999, has been the reference for the projection methodologies for the design of ITER, but the IPB also highlighted several key issues which needed to be resolved to provide a robust basis for ITER operation. In the intervening period scientists of the ITER Participant Teams have addressed these issues intensively. The International Tokamak Physics Activity (ITPA) has provided an excellent forum for scientists involved in these studies, focusing their work on the high priority physics issues for ITER. Significant progress has been made in many of the issues identified in the IPB and this progress is discussed in depth in the PIPB. In this respect, the publication of the PIPB symbolizes the strong interest and enthusiasm of the plasma physics community for the success of the ITER project, which we all recognize as one of the great scientific challenges of the 21st century. I wish to emphasize my appreciation of the work of the ITPA Coordinating Committee members, who are listed below. Their support and encouragement for the preparation of the PIPB were

  18. An electro-hydraulic servo control system research for CFETR blanket RH

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Changqi [Hefei University of Technology, Hefei 230009, Anhui (China); Tang, Hongjun, E-mail: taurustang@126.com [Hefei University of Technology, Hefei 230009, Anhui (China); Qi, Songsong [Hefei University of Technology, Hefei 230009, Anhui (China); Cheng, Yong; Feng, Hansheng; Peng, Xuebing; Song, Yuntao [Institute of Plasma Physics Chinese Academy of Sciences, Hefei 230031, Anhui (China)

    2014-11-15

    Highlights: • We discussed the conceptual design of CFETR blanket RH maintenance system. • The mathematical model of electro-hydraulic servo system was calculated. • A fuzzy adaptive PD controller was designed based on control theory and experience. • The co-simulation models of the system were established with AMESim/Simulink. • The fuzzy adaptive PD algorithm was designed as the core strategy of the system. - Abstract: Based on the technical design requirements of China Fusion Engineering Test Reactor (CFETR) blanket remote handling (RH) maintenance, this paper focus on the control method of achieving high synchronization accuracy of electro-hydraulic servo system. Based on fuzzy control theory and practical experience, a fuzzy adaptive proportional-derivative (PD) controller was designed. Then a more precise co-simulation model was established with AMESim/Simulink. Through the analysis of simulation results, a fuzzy adaptive PD control algorithm was designed as the core strategy of electro-hydraulic servo control system.

  19. Integrated application of upflow anaerobic sludge blanket reactor for the treatment of wastewaters.

    Science.gov (United States)

    Latif, Muhammad Asif; Ghufran, Rumana; Wahid, Zularisam Abdul; Ahmad, Anwar

    2011-10-15

    The UASB process among other treatment methods has been recognized as a core method of an advanced technology for environmental protection. This paper highlights the treatment of seven types of wastewaters i.e. palm oil mill effluent (POME), distillery wastewater, slaughterhouse wastewater, piggery wastewater, dairy wastewater, fishery wastewater and municipal wastewater (black and gray) by UASB process. The purpose of this study is to explore the pollution load of these wastewaters and their treatment potential use in upflow anaerobic sludge blanket process. The general characterization of wastewater, treatment in UASB reactor with operational parameters and reactor performance in terms of COD removal and biogas production are thoroughly discussed in the paper. The concrete data illustrates the reactor configuration, thus giving maximum awareness about upflow anaerobic sludge blanket reactor for further research. The future aspects for research needs are also outlined.

  20. Preliminary lifetime predictions for 304 stainless steel as the LANL ABC blanket material

    Energy Technology Data Exchange (ETDEWEB)

    Park, J.J.; Buksa, J.J.; Houts, M.G.; Arthur, E.D.

    1997-11-01

    The prediction of materials lifetime in the preconceptual Los Alamos National Laboratory (LANL) Accelerator-Based Conversion of Plutonium (ABC) is of utmost interest. Because Hastelloy N showed good corrosion resistance to the Oak Ridge National Laboratory Molten Salt Reactor Experiment fuel salt that is similar to the LANL ABC fuel salt, Hastelloy N was originally proposed for the LANL ABC blanket material. In this paper, the possibility of using 304 stainless steel as a replacement for the Hastelloy N is investigated in terms of corrosion issues and fluence-limit considerations. An attempt is made, based on the previous Fast Flux Test Facility design data, to predict the preliminary lifetime estimate of the 304 stainless steel used in the blanket region of the LANL ABC.

  1. Use of Ball Blanket in attention-deficit/hyperactivity disorder sleeping problems

    DEFF Research Database (Denmark)

    Hvolby, Allan; Bilenberg, Niels

    2011-01-01

    Objectives: Based on actigraphic surveillance, attention-deficit/hyperactivity disorder (ADHD) symptom rating and sleep diary, this study will evaluate the effect of Ball Blanket on sleep for a sample of 8-13-year-old children with ADHD. Design: Case-control study. Setting: A child and adolescent...... psychiatric department of a teaching hospital. Participants: 21 children aged 8-13 years with a diagnosis of ADHD and 21 healthy control subjects. Intervention: Sleep was monitored by parent-completed sleep diaries and 28 nights of actigraphy. For 14 of those days, the child slept with a Ball Blanket. Main...... outcome measures: The sleep latency, number of awakenings and total length of sleep was measured, as was the possible influence on parent- and teacher-rated ADHD symptom load. Results: The results of this study will show that the time it takes for a child to fall asleep is shortened when using a Ball...

  2. Two-dimensional TBR calculations for conceptual compact reversed-field pinch reactor blanket

    Science.gov (United States)

    Davidson, J. W.; Battat, M. E.; Dudziak, D. J.

    A detailed two-dimensional nucleonic analysis was performed for a conceptual first wall, blanket, and shield design for the Compact Reversed-Field Pinch Reactor. The design includes significant two-dimensional aspects presented by the limiter, vacuum ducts, and coolant manifolds; these aspects seriously degrade the tritium-breeding reaction (TBR) predicted by one-dimensional calculations. A range of design change to increase the TBR were investigated within the two-dimensional analysis. The results of this investigation indicated that an adequate TBR could be achieved with a thinning copper first wall, a (6)Li enrichment near 90%, the proper selection of reflector, and a small addition to the blanket thickness, determined by the one-dimensional analysis.

  3. Preconceptual engineering design for the APT {sup 3}He target/blanket concept

    Energy Technology Data Exchange (ETDEWEB)

    Mensink, D.L. [Babcock & Wilcox Co., Naval Nuclear Fuel Division, P.O. Box 785, Mt. Athos Rd., Lynchburg, Virginia 24505-0785 (United States); Rose, S.C. Jr. [Reactor Design and Analysis, Los Alamos National Laboratory, Los Alamos, New Mexico 87544 (United States)

    1995-01-20

    A preconceptual engineering design has been developed for the {sup 3}He Target/Blanket (T/B) System for the Accelerator Production of Tritium Project. This concept uses an array of pressure tubes containing tungsten rods for the neutron spallation source and {sup 3}He gas contained in a metal tank and blanket tubes as the tritium production material. The engineering design is based on a physics model optimized for efficient tritium production. Principle engineering consideration were: provisions for cooling all materials including the {sup 3}He gas; containment of the gas and radionuclides; remote handling; material compatibility; minimization of {sup 3}He, D{sub 2}O, and activated waste; modularity; and manufacturability. The design provides a basis for estimating the cost to implement the system.

  4. Microstructure and hardness of HIP-bonded regions in F82H blanket structures

    Science.gov (United States)

    Furuya, K.; Wakai, E.; Ando, M.; Sawai, T.; Nakamura, K.; Takeuchi, H.; Iwabuchi, A.

    2002-12-01

    Metallurgical examinations and hardness measurements were performed at hot isostatic pressing (HIP)-bonded regions in blanket structures made from F82H alloy in order to investigate the HIP-bondability and the influence on the microstructure due to the HIP and heat treatments which would correspond to the fabrication of an actual blanket. The metallurgical examination showed that the HIP-bonded interfaces were sufficiently diffusion-bonded without significant defects, i.e. voids and/or exfoliations, although grain coarsening was observed at a part of the HIP interfaces. Hardness was nearly equal in the coarsening region and a region without coarsening, but about a 10 Hv increase was found in a boundary in between the regions with and without coarsening. Microcrystallized grains were observed in a region about ˜6 μm from HIP interfaces, and the hardness increased by about 0.2 GPa in the region.

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

    Science.gov (United States)

    Peterson, Per F.

    2013-05-14

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

  6. Neutron transport-burnup code MCORGS and its application in fusion fission hybrid blanket conceptual research

    Science.gov (United States)

    Shi, Xue-Ming; Peng, Xian-Jue

    2016-09-01

    Fusion science and technology has made progress in the last decades. However, commercialization of fusion reactors still faces challenges relating to higher fusion energy gain, irradiation-resistant material, and tritium self-sufficiency. Fusion Fission Hybrid Reactors (FFHR) can be introduced to accelerate the early application of fusion energy. Traditionally, FFHRs have been classified as either breeders or transmuters. Both need partition of plutonium from spent fuel, which will pose nuclear proliferation risks. A conceptual design of a Fusion Fission Hybrid Reactor for Energy (FFHR-E), which can make full use of natural uranium with lower nuclear proliferation risk, is presented. The fusion core parameters are similar to those of the International Thermonuclear Experimental Reactor. An alloy of natural uranium and zirconium is adopted in the fission blanket, which is cooled by light water. In order to model blanket burnup problems, a linkage code MCORGS, which couples MCNP4B and ORIGEN-S, is developed and validated through several typical benchmarks. The average blanket energy Multiplication and Tritium Breeding Ratio can be maintained at 10 and 1.15 respectively over tens of years of continuous irradiation. If simple reprocessing without separation of plutonium from uranium is adopted every few years, FFHR-E can achieve better neutronic performance. MCORGS has also been used to analyze the ultra-deep burnup model of Laser Inertial Confinement Fusion Fission Energy (LIFE) from LLNL, and a new blanket design that uses Pb instead of Be as the neutron multiplier is proposed. In addition, MCORGS has been used to simulate the fluid transmuter model of the In-Zinerater from Sandia. A brief comparison of LIFE, In-Zinerater, and FFHR-E will be given.

  7. Numerical analysis of heat transfer in the first wall of CFETR WCSB blanket

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Pinghui, E-mail: phzhao@mail.ustc.edu.cn; Deng, Weiping; Ge, Zhihao; Li, Yuanjie

    2016-04-15

    Highlights: • Detailed numerical analysis of heat transfer in a water-cooling first wall was carried out based on the conceptual design of CFETR WCSB blanket. • Investigation of the influences of buoyancy effect and surface roughness on heat transfer in the water-cooling first wall was presented. • Analysis of the effect of the front wall thickness on temperature was carried out for the water-cooling first wall design. • Simulation results of two 1D CFD methods were evaluated by the 3D CFD data. - Abstract: China Fusion Engineering Test Reactor (CFETR), the first fusion reactor experiment project planned in China, is now being investigated in detail. Recently, a conceptual structural design of the Water-Cooled-Solid-Breeder (WCSB) blanket was proposed as one of the breeding blanket candidates for CFETR. In this research, based on the present design of the CFETR WCSB blanket, the heat transfer performance in the first wall (FW) under the pressurized water cooling condition was analyzed. The 3D computational fluid dynamics (CFD) results show that the maximal temperature of the FW will not exceed the limited temperature under normal or even higher heat flux condition. In addition, the effect of buoyancy on heat transfer is negligible under both conditions. The influence of roughness becomes increasingly important when the roughness height lies in the fully turbulent regime. The maximal temperature increases approximately linearly as the thickness of the front wall increases. It is also found that the heat flux and the local heat transfer coefficient are extremely non-uniform in the circumferential direction. Two 1D CFD methods are also evaluated by 3D CFD data, with the conclusion that both 1D results have some differences with the 3D data. The improved 1D method is more accurate than the former one. However, we ascertain that 1D methods should be used with caution for the water-cooling FW design.

  8. Test program element II blanket and shield thermal-hydraulic and thermomechanical testing, experimental facility survey

    Energy Technology Data Exchange (ETDEWEB)

    Ware, A.G.; Longhurst, G.R.

    1981-12-01

    This report presents results of a survey conducted by EG and G Idaho to determine facilities available to conduct thermal-hydraulic and thermomechanical testing for the Department of Energy Office of Fusion Energy First Wall/Blanket/Shield Engineering Test Program. In response to EG and G queries, twelve organizations (in addition to EG and G and General Atomic) expressed interest in providing experimental facilities. A variety of methods of supplying heat is available.

  9. Analysis of tritium behaviour and recovery from a water-cooled Pb17Li blanket

    Energy Technology Data Exchange (ETDEWEB)

    Malara, C. [Institute Regional des Materiaux Avances, Ispra (Italy); Casini, G. [Systems Engineering and Informatics Institute, JRC Ispra, Ispra (Vatican City State, Holy See) (Italy); Viola, A. [Department of Chemical Engineering, University of Cagliari, Cagliari (Italy)

    1995-03-01

    The question of the tritium recovery in water-cooled Pb17Li blankets has been under investigation for several years at JRC Ispra. The method which has been more extensively analysed is that of slowly circulating the breeder out from the blanket units and of extracting the tritium from it outside the plasma vacuum vessel by helium gas purging or vacuum degassing in a suited process apparatus. A computerized model of the tritium behaviour in the blanket units and in the extraction system was developed. It includes four submodels: (1) tritium permeation process from the breeder to the cooling water as a function of the local operative conditions (tritium concentration in Pb17Li, breeder temperature and flow rate); (2) tritium mass balance in each breeding unit; (3) tritium desorption from the breeder material to the gas phase of the extraction system; (4) tritium extraction efficiency as a function of the design parameters of the recovery apparatus. In the present paper, on the basis of this model, a parametric study of the tritium permeation rate in the cooling water and of the tritium inventory in the blanket is carried out. Results are reported and discussed in terms of dimensionless groups which describe the relative effects of the overall resistance on tritium transfer to the cooling water (with and without permeation barriers), circulating Pb17Li flow rate and extraction efficiency of the tritium recovery unit. The parametric study is extended to the recovery unit in the case of tritium extraction by helium purge or vacuum degassing in a droplet spray unit. (orig.).

  10. Neutronics Optimization of Tritium Breeding Blan-ket for the FDS

    Institute of Scientific and Technical Information of China (English)

    郑善良; 吴宜灿; 黄群英

    2002-01-01

    Neutronics optimization calculations have been performed for thc tritium breeding blankets with solid ceramic breeder Li2O and liquid eutectie breeder Li17Pb83, respectively,based on a 2-D geometrical configuration using the Monte Carlo neutron-photon transport code MCNP/4B. The effects of beryllium, 6Li enrichment and various structural materials on Tritium Breeding Ratio have been systematically analyzed.

  11. A New Iterative Scheme of Modified Mann Iteration in Banach Space

    Directory of Open Access Journals (Sweden)

    Jinzuo Chen

    2014-01-01

    Full Text Available We introduce the modified iterations of Mann's type for nonexpansive mappings and asymptotically nonexpansive mappings to have the strong convergence in a uniformly convex Banach space. We study approximation of common fixed point of asymptotically nonexpansive mappings in Banach space by using a new iterative scheme. Applications to the accretive operators are also included.

  12. Picard iteration converges faster than Mann iteration for a class of quasi-contractive operators

    Directory of Open Access Journals (Sweden)

    Vasile Berinde

    2004-06-01

    Full Text Available In the class of quasi-contractive operators satisfying Zamfirescu's conditions, the most used fixed point iterative methods, that is, the Picard, Mann, and Ishikawa iterations, are all known to be convergent to the unique fixed point. In this paper, the comparison of the first two methods with respect to their convergence rate is obtained.

  13. Power Flattening and Rejuvenation of PWR Spent Fuel Blanket for Hybrid Fusion-Fission Reactor%功率展平的压水堆乏燃料发电包层中子学初步研究

    Institute of Scientific and Technical Information of China (English)

    马续波; 陈义学; 王继亮; 王悦; 韩静茹; 陆道纲

    2011-01-01

    The hybrid fusion-fission reactor has advantages of breeding of the nuclear fuel and transmutation of the long-life nuclear waste and having inherent safety. Meanwhile, the engineering and technological demand of hybrid reactor is significantly reduced comparing with that of pure fusion reactor. A generating electricity blanket concept using the PWR spent fuel directly was proposed, which was based on ITER parameter level achieved. Different volume fractions of the fuel in blanket enabled to realize a power flattening in the fissile zone. The results show that the peak-to-average power factor becomes less than no power flattening, and the output power of the fuel zone raises more than 21. 7%. At the end of the operation, the maximum fuel enrichment is 5. 23%. The blanket is feasible from the neutronics viewpoint.%聚变裂变混合堆在增殖核燃料、嬗变长寿命核废料及固有安全性等方面具有较大优势,同时,它比纯聚变堆在工程及技术方面要求低,因此较聚变堆更易实现.本工作基于目前国际聚变实验堆(ITER)所能达到的技术水平,提出一种直接利用乏燃料进行发电的聚变裂变混合堆包层概念,利用在不同位置放置不同乏燃料体积分数的方法对燃料增殖区实现了功率展平.计算结果表明:功率展平后的包层功率不均匀系数更小,且包层中燃料区的能量输出要比不展平情况下的能量输出高约21.7%.燃料富集度到运行末期最大可达5.23%.从中子学角度初步论证了该包层的可行性.

  14. Status of ITER neutron diagnostic development

    Science.gov (United States)

    Krasilnikov, A. V.; Sasao, M.; Kaschuck, Yu. A.; Nishitani, T.; Batistoni, P.; Zaveryaev, V. S.; Popovichev, S.; Iguchi, T.; Jarvis, O. N.; Källne, J.; Fiore, C. L.; Roquemore, A. L.; Heidbrink, W. W.; Fisher, R.; Gorini, G.; Prosvirin, D. V.; Tsutskikh, A. Yu.; Donné, A. J. H.; Costley, A. E.; Walker, C. I.

    2005-12-01

    Due to the high neutron yield and the large plasma size many ITER plasma parameters such as fusion power, power density, ion temperature, fast ion energy and their spatial distributions in the plasma core can be measured well by various neutron diagnostics. Neutron diagnostic systems under consideration and development for ITER include radial and vertical neutron cameras (RNC and VNC), internal and external neutron flux monitors (NFMs), neutron activation systems and neutron spectrometers. The two-dimensional neutron source strength and spectral measurements can be provided by the combined RNC and VNC. The NFMs need to meet the ITER requirement of time-resolved measurements of the neutron source strength and can provide the signals necessary for real-time control of the ITER fusion power. Compact and high throughput neutron spectrometers are under development. A concept for the absolute calibration of neutron diagnostic systems is proposed. The development, testing in existing experiments and the engineering integration of all neutron diagnostic systems into ITER are in progress and the main results are presented.

  15. Study of neutron spectrometers for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Kaellne, Jan

    2005-11-15

    A review is presented of the developments in the field of neutron emission spectrometry (NES) which is of relevance for identifying the role of NES diagnostics on ITER and selecting suitable instrumentation. Neutron spectrometers will be part of the ITER neutron diagnostic complement and this study makes a special effort to examine which performance characteristics the spectrometers should possess to provide the best burning plasma diagnostic information together with neutron cameras and neutron yield monitors. The performance of NES diagnostics is coupled to how much interface space can be provided which has lead to an interest to find compact instruments and their NES capabilities. This study assesses all known spectrometer types of potential interest for ITER and makes a ranking of their performance (as demonstrated or projected), which, in turn, are compared with ITER measurement requirements as a reference; the ratio of diagnostic performance to interface cost for different spectrometers is also discussed for different spectrometer types. The overall result of the study is an assessment of which diagnostic functions neutron measurements can provide in burning plasma fusion experiments on ITER and the role that NES can play depending on the category of instrument installed. Of special note is the result that much higher quality diagnostic information can be obtained from neutron measurements with total yield monitors, profile flux cameras and spectrometers when the synergy in the data is considered in the analysis and interpretation.

  16. [Air conditioning units and warm air blankets in the operating room].

    Science.gov (United States)

    Kerwat, Klaus; Piechowiak, Karolin; Wulf, Hinnerk

    2013-01-01

    Nowadays almost all operating rooms are equipped with air conditioning (AC units). Their main purpose is climatization, like ventilation, moisturizing, cooling and also the warming of the room in large buildings. In operating rooms they have an additional function in the prevention of infections, especially the avoidance of postoperative wound infections. This is achieved by special filtration systems and by the creation of specific air currents. Since hypothermia is known to be an unambiguous factor for the development of postoperative wound infections, patients are often actively warmed intraoperatively using warm air blankets (forced-air warming units). In such cases it is frequently discussed whether such warm air blankets affect the performance of AC units by changing the air currents or whether, in contrast, have exactly the opposite effect. However, it has been demonstrated in numerous studies that warm air blankets do not have any relevant effect on the functioning of AC units. Also there are no indications that their use increases the rate of postoperative wound infections. By preventing the patient from experiencing hypothermia, the rate of postoperative wound infections can even be decreased thereby.

  17. Preflow stresses in Martian rampart ejecta blankets - A means of estimating the water content

    Science.gov (United States)

    Woronow, A.

    1981-01-01

    Measurements of extents of rampart ejecta deposits as a function of the size of the parent craters support models which, for craters larger than about 6 km diameter, constrain ejecta blankets to all have a similar maximum thickness regardless of the crater size. These volatile-rich ejecta blankets may have failed under their own weights, then flowed radially outward. Assuming this to be so, some of the physicomechanical properties of the ejecta deposits at the time of their emplacement can then be determined. Finite-element studies of the stress magnitudes, distributions, and directions in hypothetical Martian rampart ejecta blankets reveal that the material most likely failed when the shear stresses were less than 500 kPa and the angle of internal friction was between 26 and 36 deg. These figures imply that the ejecta has a water content between 16 and 72%. Whether the upper limit or the lower limit is more appropriate depends on the mode of failure which one presumes: namely, viscous flow of plastic deformation.

  18. Parametric Weight Comparison of Advanced Metallic, Ceramic Tile, and Ceramic Blanket Thermal Protection Systems

    Science.gov (United States)

    Myers, David E.; Martin, Carl J.; Blosser, Max L.

    2000-01-01

    A parametric weight assessment of advanced metallic panel, ceramic blanket, and ceramic tile thermal protection systems (TPS) was conducted using an implicit, one-dimensional (I-D) finite element sizing code. This sizing code contained models to account for coatings fasteners, adhesives, and strain isolation pads. Atmospheric entry heating profiles for two vehicles, the Access to Space (ATS) vehicle and a proposed Reusable Launch Vehicle (RLV), were used to ensure that the trends were not unique to a certain trajectory. Ten TPS concepts were compared for a range of applied heat loads and substructural heat capacities to identify general trends. This study found the blanket TPS concepts have the lightest weights over the majority of their applicable ranges, and current technology ceramic tiles and metallic TPS concepts have similar weights. A proposed, state-of-the-art metallic system which uses a higher temperature alloy and efficient multilayer insulation was predicted to be significantly lighter than the ceramic tile stems and approaches blanket TPS weights for higher integrated heat loads.

  19. AB Blanket for Cities (for continual pleasant weather and protection from chemical, biological and radioactive weapons)

    CERN Document Server

    Bolonkin, Alexander

    2009-01-01

    In a series of previous articles (see references) the author offered to cover a city or other important large installations or subregions by a transparent thin film supported by a small additional air overpressure under the form of an AB Dome. The building of a gigantic inflatable AB Dome over an empty flat surface is not difficult. However, if we want to cover a city, garden, forest or other obstacle course we cannot easily deploy the thin film over building or trees. In this article is suggested a new method which solves this problem. The idea is to design a double film blanket filled by light gas (for example, methane, hydrogen, or helium). Sections of this AB Blanket are lighter then air and fly in atmosphere. They can be made on a flat area (serving as an assembly area) and delivered by dirigible or helicopter to station at altitude over the city. Here they connect to the already assembled AB Blanket subassemblies, cover the city in an AB Dome and protect it from bad weather, chemical, biological and rad...

  20. Blanketing effect of expansion foam on liquefied natural gas (LNG) spillage pool.

    Science.gov (United States)

    Zhang, Bin; Liu, Yi; Olewski, Tomasz; Vechot, Luc; Mannan, M Sam

    2014-09-15

    With increasing consumption of natural gas, the safety of liquefied natural gas (LNG) utilization has become an issue that requires a comprehensive study on the risk of LNG spillage in facilities with mitigation measures. The immediate hazard associated with an LNG spill is the vapor hazard, i.e., a flammable vapor cloud at the ground level, due to rapid vaporization and dense gas behavior. It was believed that high expansion foam mitigated LNG vapor hazard through warming effect (raising vapor buoyancy), but the boil-off effect increased vaporization rate due to the heat from water drainage of foam. This work reveals the existence of blocking effect (blocking convection and radiation to the pool) to reduce vaporization rate. The blanketing effect on source term (vaporization rate) is a combination of boil-off and blocking effect, which was quantitatively studied through seven tests conducted in a wind tunnel with liquid nitrogen. Since the blocking effect reduces more heat to the pool than the boil-off effect adds, the blanketing effect contributes to the net reduction of heat convection and radiation to the pool by 70%. Water drainage rate of high expansion foam is essential to determine the effectiveness of blanketing effect, since water provides the boil-off effect.