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Sample records for compact demo slimcs

  1. Conceptual design of the SlimCS fusion DEMO reactor

    Tobita, Kenji; Nishio, Satoshi; Enoeda, Mikio; Nakamura, Hirofumi; Hayashi, Takumi; Asakura, Nobuyuki; Utoh, Hiroyasu; Tanigawa, Hiroyasu; Nishitani, Takeo; Isono, Takaaki; Sakurai, Shinji; Kurita, Genichi; Hayashi, Takao; Oyama, Naoyuki; Liu Changle; Hamamatsu, Kiyotaka; Inoue, Takashi; Ozeki, Takahisa; Sato, Masayasu; Suzuki, Satoshi; Kawashima, Hisato; Ezato, Koichiro; Tsuru, Daigo; Koizumi, Norikiyo; Sakamoto, Keiji; Ando, Masami; Sakamoto, Yoshiteru; Shibama, Yusuke; Suzuki, Takahiro; Takechi, Manabu; Takahashi, Koji; Hirose, Takanori; Sato, Satoru; Nozawa, Takashi; Tanigawa, Hisashi; Kakudate, Satoshi; Kawamura, Yoshinori; Yamanishi, Toshihiko; Hoshino, Tsuyoshi; Ochiai, Kentaro; Ide, Shunsuke; Aiba, Nobuyuki; Shimizu, Katsuhiro; Honda, Mitsuru; Nakamichi, Masaru; Nishi, Hiroshi; Seki, Yoji; Nakamura, Yukiharu; Tsuchiya, Kunihiko; Yoshida, Tohru; Song Yuntao

    2010-08-01

    This report describes the results of the conceptual design study of the SlimCS fusion DEMO reactor aiming at demonstrating fusion power production in a plant scale and allowing to assess the economic prospects of a fusion power plant. The design study has focused on a compact and low aspect ratio tokamak reactor concept with a reduced-sized central solenoid, which is novel compared with previous tokamak reactor concept such as SSTR (Steady State Tokamak Reactor). Owing to low aspect ratio, the reactor will be capable of having comparatively high beta limit and high elongation (which can elevate the Greenwald density limit), having potential for high power density. The reactor has the main parameters of a major radius of 5.5 m, aspect ratio of 2.6, elongation of 2.0, normalized beta of 4.3, fusion out put of 2.95 GW and average neutron wall load of 3 MW/m 2 . This report covers various aspects of design study including systematic design, physics design, torus configuration, blanket, superconducting magnet, maintenance and building, which were carried out increase the engineering feasibility of the concept. (author)

  2. Divertor Design and Physics Issues of Huge Power Handling for SlimCS Demo Reactor

    Asakura, N.; Hoshino, K.; Tobita, K.; Someya, Y.; Utoh, H.; Nakamura, M., E-mail: asakura.nobuyuki@jaea.go.jp [Japan Atomic Energy Agency, Rokkasho (Japan); Shimizu, K. [Japan Atomic Energy Agency, Naka (Japan); Takizuka, T. [Osaka University, Osaka (Japan)

    2012-09-15

    Full text: Power exhaust scenario for a 3 GW class fusion reactor with the ITER-size plasma has been developed with enhancing the radiation loss from seeding impurity. Transport of plasma, impurity and neutrals was simulated self-consistently, for the first time, under the Demo divertor condition using an integrated divertor simulation code SONIC. The total heat load, q{sub target}, was evaluated including radiation power load and neutral load, in addition to the plasma heat load. It was found that heat and particle diffusion coefficients significantly affect the plasma detachment. For the case of increasing the coefficients by the factor of two, peak q{sub target} is reduced from 18 MW/m{sup 2} to below the engineering design level of 10 MW/m{sup 2}, while the characteristic width of the heat flux at the midplane SOL increases slightly from 2.2 to 2.7 mm. It was also found that that enhancement of the local {chi} and D at the outer SOL affects a reduction in the peak q{sub target} near the separatrix. Effects of the divertor geometry such as the divertor leg were investigated. Outer divertor leg length was extended to 2.7 m, while the magnetic flux expansion at the target is reduced to a half compared to the reference case of 1.8 m. Large radiation volume is shifted further upstream from the target due to a reduction in T{sub e}. The peak q{sub target} decreases to 10 MW/m{sup 2} due to reduction in both the plasma heat load and the radiation power load. (author)

  3. Safety research on fusion DEMO in Japan: Toward development of safety strategy of a water-cooled DEMO

    Nakamura, Makoto, E-mail: nakamura.makoto@jaea.go.jp [Japan Atomic Energy Agency, Rokkasho 039-3212, Aomori (Japan); Tobita, Kenji; Someya, Youji; Utoh, Hiroyasu; Sakamoto, Yoshiteru [Japan Atomic Energy Agency, Rokkasho 039-3212, Aomori (Japan); Gulden, Werner [Fusion for Energy, Garching D-85748 (Germany)

    2016-11-01

    Highlights: • This paper reports the current status of a safety research on water-cooled fusion DEMO in Japan. • We report analyses of two transients: (i) complete loss of decay heat removal and (ii) major ex-VV LOCA. • The MELCOR analysis has clarified the temperature histories of the DEMO components in complete loss of decay heat removal. • A strategy to reduce the pressure load to the final barrier confining radioactive materials is proposed against the major ex-VV LOCA. - Abstract: This paper reports the current status of a safety research on water-cooled fusion DEMO in Japan. A basic strategy of development of the safety guidelines is described for DEMO based on a water-cooled solid pebble bed blanket. Clarification of safety features of the DEMO in accident situations is a key issue to develop the guidelines. Recent achievements in understanding of the safety features of the water-cooled DEMO are reported. The MELCOR analysis has clarified the temperature histories of the DEMO components in a complete loss of decay heat removal event. The transient behavior of the first wall temperature is found to be essentially different from that of ITER. The pressure load to the tokamak cooling water system vault (TCWSV) is analyzed based on a simple model equation of the energy conservation. If the amount of the primary coolant is the same as that of Slim-CS, the previous small Japanese DEMO, the discharged water does not damage the TCWSV with the volume and pressure-tightness similar to those of pressurized light water reactors. It is shown that implementation of a pressure suppression system to the small TCWSV is effective to suppress the pressure load to the second confinement barrier.

  4. A Feasible DEMO Blanket Concept Based on Water Cooled Solid Breeder

    Someya, Y.; Tobita, K.; Utoh, H.; Hoshino, K.; Asakura, N.; Nakamura, M.; Tanigawa, H.; Mikio, E.; Tanigawa, H.; Nakamichi, M.; Hoshino, T., E-mail: someya.yoji@jaea.go.jp [Japan Atomic Energy Agency, Rokkasho (Japan)

    2012-09-15

    Full text: JAEA has conducted the conceptual design study of blanket for a fusion DEMO reactor SlimCS. Considering DEMO specific requirements, we place emphasis on a blanket concept with durability to severe irradiation, ease of fabrication for mass production, operation temperature of blanket materials, and maintainability using remote handling equipment. This paper present a promising concept satisfying these requirements, which is characterized by minimized welding lines near the front, a simplified blanket interior consisting of cooling tubes and a mixed pebble bed of breeder and neutron multiplier, and approximately the same outlet temperature for all blanket modules. Neutronics calculation indicated that the blanket satisfies a self-sufficient production of tritium. An important finding is that little decrease is seen in tritium breeding ratio even when the gap between neighboring blanket modules is as wide as 0.03 m. This means that blanket modules can be arranged with such a significant clearance gap without sacrifice of tritium production, which will facilitate the access of remote handling equipment for replacement of the blanket modules and improve the access of diagnostics. (author)

  5. Breeding blanket for Demo

    Proust, E.; Giancarli, L.

    1992-01-01

    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 investigated within the framework of the European Test-Blanket Development Programme

  6. What must Demo do?

    Waganer, L.M.; Najmabadi, F.; Tillack, M.S.

    1995-01-01

    The US fusion demonstration plant (Demo) must satisfy certain top level requirements so that energy producers will confidently invest in a commercial fusion version for their next generation power plant. To instill that level of confidence to both the investor and the public, Demo must achieve high standards in safety, low environmental impact, reliability, and economics. This is a most difficult set of goals to meet. The public is demanding ever more strict environmental rules and regulations. The hazards of radioactive and toxic waste and emissions are becoming better understood. The difficulties of establishing and maintaining long-lived repositories are enormous. Neighborhood action groups have an aversion to large power plants in their back yards. Utilities and independent power producers are reluctant to commit to a long-term financial arrangement for a new technology. To achieve these stringent goals, the competition is continuing to improve to meet these challenges. Only the best can adapt and survive. The Demo plant is not expected to achieve all requirements demanded of the commercial power plant, but it must demonstrate values close enough to the commercial machine so that extrapolation to the commercial carries minimal risk in all key areas. Specifically, Demo must demonstrate all the major performance parameters in an integrated system similar to that of the commercial plant. It should be large enough so that all aspects of the Demo can be confidently scaled to that of the commercial plant, including the economics, reliability, availability, and operability

  7. Critical Design Factors for Sector Transport Maintenance in DEMO

    Utoh, H.; Someya, Y.; Tobita, K.; Asakura, N.; Hoshino, K.; Nakamura, M., E-mail: uto.hiroyasu@jaea.go.jp [Japan Atomic Energy Agency, Rokkasho (Japan)

    2012-09-15

    Full text: Maintenance is a critical issue for fusion DEMO reactor because the design conditions and requirements of DEMO maintenance scheme are different from that of ITER remote handling. The sector transport maintenance scheme has advantages to maintain blankets and divertors without the use of sophisticated remote handling devices including sensitive devices to radiation in the reactor. SlimCS designed in JAEA adopts the sector transport maintenance scheme in which every sector is pulled out horizontally through a port between TF coils. A critical design issue for the horizontal sector transport maintenance scheme is how to support an enormous turnover force of the toroidal field (TF) coils. We propose following two options; first option is the horizontal transport maintenance scheme in which every sector is pulled out through four horizontal ports connected with the corridor. Second option is the vertical sector transport maintenance scheme with small vertical maintenance ports (total: 6 ports). The new horizontal sector transport limited in the number of maintenance ports is a more realistic maintenance scheme, and the key engineering issue is the transferring mechanism of sector in the vacuum vessel. In the maintenance scenario, the key design factors are the cool down time in reactor and the cooling method in maintenance scheme for keeping components under operation temperature. By one-dimensional heat conduction analysis, the sector should be transported to hot cell within 40 hours in the case the cool down time is one month. In the horizontal sector transport maintenance, the maintenance time including removal of cooling piping, drain of cooling water and sector transport to hot cell is about 32 hours. Furthermore, the tritium release in the sector transport can be suppressed because the components temperature drops by forced-air cooling system. This paper mainly focuses on a sector transport maintenance scheme from the aspects of high plant availability

  8. Tritium extraction technologies and DEMO requirements

    Demange, D., E-mail: david.demange@kit.edu [Karlsruhe Institute of Technology, Institute for Technical Physics, Tritium Laboratory Karlsruhe, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Antunes, R.; Borisevich, O.; Frances, L. [Karlsruhe Institute of Technology, Institute for Technical Physics, Tritium Laboratory Karlsruhe, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Rapisarda, D. [Laboratorio Nacional de Fusión, EURATOM-CIEMAT, 28040 Madrid (Spain); Santucci, A. [ENEA for EUROfusion, Via E. Fermi 45, 00044 Frascati, Roma (Italy); Utili, M. [ENEA CR Brasimone, 40032 Camugnano, BO (Italy)

    2016-11-01

    Highlights: • We detail the R&D plan for tritium technology of the European DEMO breeding blanket. • We study advanced and efficient extraction techniques to improve tritium management. • We consider inorganic membranes and catalytic membrane reactor for solid blankets. • We consider permeator against vacuum and vacuum sieve tray for liquid blankets. - Abstract: The conceptual design of the tritium extraction system (TES) for the European DEMO reactor is worked out in parallel for four different breeding blankets (BB) retained by EUROfusion. The TES design has to be tackled in an integrated manner optimizing the synergy with the directly interfacing inner fuel cycle, while minimizing the tritium permeation into the coolant. Considering DEMO requirements, it is most likely that only advanced technologies will be suitable for the tritium extraction systems of the BB. This paper overviews the European work programme for R&D on tritium technology for the DEMO BB, summaries the general first outcomes, and details the specific and comprehensive R&D program to study experimentally immature but promising technologies such as vacuum sieve tray or permeator against vacuum for tritium extraction from PbLi, and advanced inorganic membranes and catalytic membrane reactor for tritium extraction from He. These techniques are simple, fully continuous, likely compact with contained energy consumption. Several European Laboratories are joining their efforts to deploy several new experimental setups to accommodate the tests campaigns that will cover small scale experiments with tritium and inactive medium scale tests so as to improve the technology readiness level of these advanced processes.

  9. DEMO diagnostics and burn control

    Biel, W.; De Baar, M.; Dinklage, A.; Felici, F.; König, R.; Meister, H.; Treutterer, W.; Wenninger, R.

    2015-01-01

    The development of the control system for a tokamak demonstration fusion reactor (DEMO) faces unprecedented challenges. First, the requirements for control reliability and accuracy are more stringent than on existing fusion devices: any loss of plasma control on DEMO may result in a disruption which

  10. Introduction: Undoing the demos

    Dean, Mitchell

    2017-01-01

    and usefulness of Michel Foucault’s notion of governmentality and Karl Marx’s analysis of capitalism for analysing neoliberalism; the way that neoliberalism ‘economises’ everything including politics and democracy; the nature of the state and of sovereignty, and how the left should relate to these......; and the nature of critique in its different forms (Kantian, Foucauldian, Marxist and others). These are issues that are important not only for the specific argument of Undoing the Demos, but more generally for social and political theory today....

  11. The DEMO Quasisymmetric Stellarator

    Geoffrey B. McFadden

    2010-02-01

    Full Text Available The NSTAB nonlinear stability code solves differential equations in conservation form, and the TRAN Monte Carlo test particle code tracks guiding center orbits in a fixed background, to provide simulations of equilibrium, stability, and transport in tokamaks and stellarators. These codes are well correlated with experimental observations and have been validated by convergence studies. Bifurcated 3D solutions of the 2D tokamak problem have been calculated that model persistent disruptions, neoclassical tearing modes (NTMs and edge localized modes (ELMs occurring in the International Thermonuclear Experimental Reactor (ITER, which does not pass the NSTAB simulation test for nonlinear stability. So we have designed a quasiaxially symmetric (QAS stellarator with similar proportions as a candidate for the demonstration (DEMO fusion reactor that does pass the test [1]. The configuration has two field periods and an exceptionally accurate 2D symmetry that furnishes excellent thermal confinement and good control of the prompt loss of alpha particles. Robust coils are found from a filtered form of the Biot-Savart law based on a distribution of current over a control surface for the coils and the current in the plasma defined by the equilibrium calculation. Computational science has addressed the issues of equilibrium, stability, and transport, so it remains to develop an effective plan to construct the coils and build a diverter.

  12. DEMO diagnostics and burn control

    Biel, Wolfgang, E-mail: w.biel@fz-juelich.de [Institute of Energy and Climate Research, Forschungszentrum Jülich GmbH, Jülich (Germany); Department of Applied Physics, Ghent University (Belgium); Baar, Marco de [FOM-Institute DIFFER, Nieuwegein (Netherlands); Eindhoven University of Technology (Netherlands); Dinklage, Andreas [Max-Planck-Institut für Plasmaphysik, Greifswald (Germany); Felici, Federico [Eindhoven University of Technology (Netherlands); König, Ralf [Max-Planck-Institut für Plasmaphysik, Greifswald (Germany); Meister, Hans; Treutterer, Wolfgang [Max-Planck-Institut für Plasmaphysik, Garching (Germany); Wenninger, Ronald [Max-Planck-Institut für Plasmaphysik, Garching (Germany); EFDA Power Plant Physics and Technology, Garching (Germany)

    2015-10-15

    Highlights: • An initial concept for the DEMO diagnostic and control system is presented. • A preliminary list of control functions and candidate diagnostics is developed. • Challenges regarding disruptions, power exhaust and radiation control are highlighted. • The need for introducing realistic control margins is emphasized. • On outline of the future R&D plan is presented. - Abstract: The development of the control system for a tokamak demonstration fusion reactor (DEMO) faces unprecedented challenges. First, the requirements for control reliability and accuracy are more stringent than on existing fusion devices: any loss of plasma control on DEMO may result in a disruption which could damage the inner wall of the machine, while operating the device with larger margins against the operational limits would lead to a reduction of the electrical output power. Second, the performance of DEMO control is limited by space restrictions for the implementation of components (optimization of the tritium breeding rate), by lifetime issues for the front-end parts (neutron and gamma radiation, erosion and deposition acting on all components) and by slow, weak and indirect action of the available actuators (plasma shaping, heating and fuelling). The European DEMO conceptual design studies include the development of a reliable control system, since the details of the achievable plasma scenario and the machine design may depend on the actual performance of the control system. In the first phase of development, an initial understanding of the prime choices of diagnostic methods applicable to DEMO, implementation and performance issues, the interrelation with the plasma scenario definition, and the planning of necessary future R&D have been obtained.

  13. Review of fusion DEMO reactor study

    Seki, Yasushi

    1996-01-01

    Fusion DEMO Reactor is defined and the Steady State Tokamak Reactor (SSTR) concept is introduced as a typical example of a DEMO reactor. Recent DEMO reactor studies in Japan and abroad are introduced. The DREAM Reactor concept is introduced as an ultimate target of fusion research. (author)

  14. Demo of Gaze Controlled Flying

    Alapetite, Alexandre; Hansen, John Paulin; Scott MacKenzie, I.

    2012-01-01

    Development of a control paradigm for unmanned aerial vehicles (UAV) is a new challenge to HCI. The demo explores how to use gaze as input for locomotion in 3D. A low-cost drone will be controlled by tracking user’s point of regard (gaze) on a live video stream from the UAV.......Development of a control paradigm for unmanned aerial vehicles (UAV) is a new challenge to HCI. The demo explores how to use gaze as input for locomotion in 3D. A low-cost drone will be controlled by tracking user’s point of regard (gaze) on a live video stream from the UAV....

  15. PSO-Ensemble Demo Application

    2004-01-01

    Within the framework of the PSO-Ensemble project (FU2101) a demo application has been created. The application use ECMWF ensemble forecasts. Two instances of the application are running; one for Nysted Offshore and one for the total production (except Horns Rev) in the Eltra area. The output...

  16. Steady State versus Pulsed Tokamak DEMO

    Orsitto, F.P., E-mail: francesco.orsitto@enea.it [Associazione EURATOM-ENEA Unita Tecnica Fusione, Frascati (Italy); Todd, T. [CCFE/Fusion Association, Culham Science Centre, Abingdon (United Kingdom)

    2012-09-15

    Full text: The present report deals with a Review of problems for a Steady state(SS) DEMO, related argument is treated about the models and the present status of comparison between the characteristics of DEMO pulsed versus a Steady state device.The studied SS DEMO Models (SLIM CS, PPCS model C EU-DEMO, ARIES-RS) are analyzed from the point of view of the similarity scaling laws and critical issues for a steady state DEMO. A comparison between steady state and pulsed DEMO is therefore carried out: in this context a new set of parameters for a pulsed (6 - 8 hours pulse) DEMO is determined working below the density limit, peak temperature of 20 keV, and requiring a modest improvement in the confinement factor(H{sub IPBy2} = 1.1) with respect to the H-mode. Both parameters density and confinement parameter are lower than the DEMO models presently considered. The concept of partially non-inductive pulsed DEMO is introduced since a pulsed DEMO needs heating and current drive tools for plasma stability and burn control. The change of the main parameter design for a DEMO working at high plasma peak temperatures T{sub e} {approx} 35 keV is analyzed: in this range the reactivity increases linearly with temperature, and a device with smaller major radius (R = 7.5 m) is compatible with high temperature. Increasing temperature is beneficial for current drive efficiency and heat load on divertor, being the synchrotron radiation one of the relevant components of the plasma emission at high temperatures and current drive efficiency increases with temperature. Technology and engineering problems are examined including efficiency and availability R&D issues for a high temperature DEMO. Fatigue and creep-fatigue effects of pulsed operations on pulsed DEMO components are considered in outline to define the R&D needed for DEMO development. (author)

  17. The DEMO wall load challenge

    Wenninger, R.; Albanese, R.; Ambrosino, R.; Arbeiter, F.; Aubert, J.; Bachmann, C.; Barbato, L.; Barrett, T.; Beckers, M.; Biel, W.; Boccaccini, L.; Carralero, D.; Coster, D.; Eich, T.; Fasoli, A.; Federici, G.; Firdaouss, M.; Graves, J.; Horáček, Jan; Kovari, M.; Lanthaler, S.; Loschiavo, V.; Lowry, C.; Lux, H.; Maddaluno, G.; Maviglia, F.; Mitteau, R.; Neu, R.; Pfefferle, D.; Schmid, K.; Siccinio, M.; Sieglin, B.; Silva, C.; Snicker, A.; Subba, F.; Varje, J.; Zohm, H.

    2017-01-01

    Roč. 57, č. 4 (2017), č. článku 046002. ISSN 0029-5515 EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : DEMO * power loads * first wall Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 3.307, year: 2016 http://iopscience.iop.org/article/10.1088/1741-4326/aa4fb4

  18. High current superconductors for DEMO

    Bruzzone, Pierluigi, E-mail: pierluigi.bruzzone@psi.ch [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas (CRPP), Association Euratom – Confédération Suisse, CH-5232 Villigen PSI (Switzerland); Sedlak, Kamil; Stepanov, Boris [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas (CRPP), Association Euratom – Confédération Suisse, CH-5232 Villigen PSI (Switzerland)

    2013-10-15

    Highlights: ► Definition of requirement for TF coil based on the input of system code. ► A TF coil and conductor design for the European DEMO project. ► Use of React and Wind method opposite to Wind and React with related advantages. ► Hybridization of winding pack, Nb/Nb{sub 3}Sn, by graded layer winding. -- Abstract: In the assumption that DEMO will be an inductively driven tokamak, the number of load cycles will be in the range of several hundred thousands. The requirements for a new generation of Nb{sub 3}Sn based high current conductors for DEMO are drafted starting from the output of system code PROCESS. The key objectives include the stability of the DC performance over the lifetime of the machine and the effective use of the Nb{sub 3}Sn strand properties, for cost and reliability reasons. A preliminary layout of the winding pack and conductors for the toroidal field magnets is presented. To suppress the mechanism of reversible and irreversible degradation, i.e. to preserve in the cabled conductor the high critical current density of the strand, the thermal strain must be insignificant and no space for micro-bending under transverse load must be left in the strand bundle. The “react-and-wind” method is preferred here, with a graded, layer wound magnet, containing both Nb{sub 3}Sn and NbTi layers. The implications of the conductor choice on the coil design and technology are highlighted. A roadmap is sketched for the development of a full size prototype conductor sample and demonstration of the key technologies.

  19. Parameters of DEMO DN and JET DN

    Anon.

    1987-01-01

    The paper is an appendix to a study of the reactor relevance of the NET design concept. The latter study examines whether the technologies and design principles proposed for NET can be directly extrapolated to a demonstration (DEMO) reactor. The appendix presents the parameters of the DEMO and NET under the topic headings: power, geometry, plasma, toroidal and poloidal magnetic field coils, first wall engineering, divertor physics, divertor engineering, and blanket. (U.K.)

  20. Considerations on the DEMO pellet fuelling system

    Lang, P.T., E-mail: peter.lang@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Day, Ch. [Karlsruhe Institute of Technology, 76021 Karlsruhe (Germany); Fable, E. [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Igitkhanov, Y. [Karlsruhe Institute of Technology, 76021 Karlsruhe (Germany); Köchl, F. [Association EURATOM-Ö AW/ATI, Atominstitut, TU Wien, 1020 Vienna (Austria); Mooney, R. [Culham Centre for Fusion Energy, Culham Science Centre, Oxfordshire OX14 3DB (United Kingdom); Pegourie, B. [CEA, IRFM, 13108 Saint-Paul-lez-Durance (France); Ploeckl, B. [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Wenninger, R. [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); EFDA, Garching (Germany); Zohm, H. [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany)

    2015-10-15

    Graphical abstract: - Highlights: • Considerations are made for a core particle fuelling system covering all DEMO requirements. • Particle deposition beyond the pedestal top is needed to achieve efficient fuelling. • Conventional pellet technology enabling launching from the torus inboard side can be used. • Efforts have been taken for integrating a suitable pellet guiding system into the EU DEMO model. • In addition, further techniques bearing potential for advanced fuelling performance are considered. - Abstract: The Demonstration Fusion Power Reactor DEMO is the step foreseen to bridge the gap between ITER and the first commercial fusion power plant. One key element in the European work plan for DEMO is the elaboration of a conceptual design for a suitable core particle fuelling system. First considerations for such a system are presented in this contribution. Following the well-considered ITER solution, most analysis performed in this study assumes conventional pellet technology will be used for the fuelling system. However, taking advantage of the less compressed time frame for the DEMO project, several other techniques thought to bear potential for advanced fuelling performance are considered as well. In a first, basic analysis all actuation parameters at hand and their implications on the fuelling performance were considered. Tentative transport modeling of a reference scenario strongly indicates only particles deposited inside the plasma pedestal allow for efficient fuelling. Shallow edge fuelling results in an unbearable burden on the fuel cycle. Sufficiently deep particle deposition seems technically achievable, provided pellets are launched from the torus inboard at sufficient speed. All components required for a DEMO pellet system capable for high speed inboard pellet launch are already available or can be developed in due time with reasonable efforts. Furthermore, steps to integrate this solution into the EU DEMO model are taken.

  1. Considerations on the DEMO pellet fuelling system

    Lang, P.T.; Day, Ch.; Fable, E.; Igitkhanov, Y.; Köchl, F.; Mooney, R.; Pegourie, B.; Ploeckl, B.; Wenninger, R.; Zohm, H.

    2015-01-01

    Graphical abstract: - Highlights: • Considerations are made for a core particle fuelling system covering all DEMO requirements. • Particle deposition beyond the pedestal top is needed to achieve efficient fuelling. • Conventional pellet technology enabling launching from the torus inboard side can be used. • Efforts have been taken for integrating a suitable pellet guiding system into the EU DEMO model. • In addition, further techniques bearing potential for advanced fuelling performance are considered. - Abstract: The Demonstration Fusion Power Reactor DEMO is the step foreseen to bridge the gap between ITER and the first commercial fusion power plant. One key element in the European work plan for DEMO is the elaboration of a conceptual design for a suitable core particle fuelling system. First considerations for such a system are presented in this contribution. Following the well-considered ITER solution, most analysis performed in this study assumes conventional pellet technology will be used for the fuelling system. However, taking advantage of the less compressed time frame for the DEMO project, several other techniques thought to bear potential for advanced fuelling performance are considered as well. In a first, basic analysis all actuation parameters at hand and their implications on the fuelling performance were considered. Tentative transport modeling of a reference scenario strongly indicates only particles deposited inside the plasma pedestal allow for efficient fuelling. Shallow edge fuelling results in an unbearable burden on the fuel cycle. Sufficiently deep particle deposition seems technically achievable, provided pellets are launched from the torus inboard at sufficient speed. All components required for a DEMO pellet system capable for high speed inboard pellet launch are already available or can be developed in due time with reasonable efforts. Furthermore, steps to integrate this solution into the EU DEMO model are taken.

  2. Initial DEMO tokamak design configuration studies

    Bachmann, Christian, E-mail: christian.bachmann@efda.org [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Aiello, G. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Albanese, R.; Ambrosino, R. [ENEA/CREATE, Universita di Napoli Federico II, Naples (Italy); Arbeiter, F. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Aubert, J. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Boccaccini, L.; Carloni, D. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Federici, G. [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Fischer, U. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Kovari, M. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Li Puma, A. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Loving, A. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Maione, I. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Mattei, M. [ENEA/CREATE, Universita di Napoli Federico II, Naples (Italy); Mazzone, G. [ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati, Roma (Italy); Meszaros, B. [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Palermo, I. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Pereslavtsev, P. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Riccardo, V. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); and others

    2015-10-15

    Highlights: • A definition of main DEMO requirements. • A description of the DEMO tokamak design configuration. • A description of issues yet to be solved. - Abstract: To prepare the DEMO conceptual design phase a number of physics and engineering assessments were carried out in recent years in the frame of EFDA concluding in an initial design configuration of a DEMO tokamak. This paper gives an insight into the identified engineering requirements and constraints and describes their impact on the selection of the technologies and design principles of the main tokamak components. The EU DEMO program aims at making best use of the technologies developed for ITER (e.g., magnets, vessel, cryostat, and to some degree also the divertor). However, other systems in particular the breeding blanket require design solutions and advanced technologies that will only partially be tested in ITER. The main differences from ITER include the requirement to breed, to extract, to process and to recycle the tritium needed for plasma operation, the two orders of magnitude larger lifetime neutron fluence, the consequent radiation dose levels, which limit remote maintenance options, and the requirement to use low-activation steel for in-vessel components that also must operate at high temperature for efficient energy conversion.

  3. On the definition of a DEMO (demonstration) reactor

    Cole, H.C.; Challender, R.S.

    1987-01-01

    The paper is an appendix to a study of the reactor relevance of the NET design concept. The latter study examines whether the technologies and design principles proposed for NET can be directly extrapolated to a demonstration (DEMO) reactor. The authors have suggested a definition of a DEMO, and listed what they considered to be the most important implications of this definition. A table of parameters is included comparing published DEMO's with typical commercial reactor and 'pre-DEMO' studies. (U.K.)

  4. Versatile Desktop Experiment Module (DEMo) on Heat Transfer

    Minerick, Adrienne R.

    2010-01-01

    This paper outlines a new Desktop Experiment Module (DEMo) engineered for a chemical engineering junior-level Heat Transfer course. This new DEMo learning tool is versatile, fairly inexpensive, and portable such that it can be positioned on student desks throughout a classroom. The DEMo system can illustrate conduction of various materials,…

  5. Objective Provision Tree for K-DEMO

    Oh, Kyemin; Kang, Myung-suk; Heo, Gyunyoung; Kim, Hyoung-chan

    2014-01-01

    In current nuclear field based on technology-neutral approach, safety principles and design have been considered for Generation IV (Gen-IV) nuclear power plants in parallel. This strategy can save resource, time, and manpower while keeping achievable safety. For this reason, the studies related with safety affecting significant design parameters for planned construction or fusion plants was needed and required even though K-DEMO is staying in pre-conceptual design phase. Objective Provision Tree (OPT) is one of the tools of Integrated Safety Assessment Methodology (ISAM) developed by Risk and Safety Working Group (RSWG) for design and assessment of Gen-IV. This is suitable tool to recognize and investigate safety issues from previous engineering experience. The purpose of this paper is to suggest multiple barriers/critical safety function and to describe the current status of the OPT for the conceptual design of K-DEMO. In this paper, critical safety functions were defined and OPT for K-DEMO was described and performed. We have carried out researches related to safety for fusion power plant in collaboration with the academies funded by NFRI during the past 4 years. As part of this research, Integrated Safety Assessment Methodology (ISAM), which was used to develop GEN-IV nuclear systems, was used to determine the technical safety issues and regulatory requirements for K-DEMO. OPT is one of ISAM tools

  6. Constructing a Tibetan Demos in Exile

    Brox, Trine

    2012-01-01

    homeland. Two specific instances of the construction of a transnational exile demos are investigated: citizenship and political representation. The Tibetan Government-in-Exile's formalized idea of citizenship builds upon ideals of equal and loyal members who form a single unit bounded by a common cause...

  7. Efficiency of water coolant for DEMO divertor

    Fetzer, Renate; Igitkhanov, Yuri; Bazylev, Boris

    2015-01-01

    Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

  8. Efficiency of water coolant for DEMO divertor

    Fetzer, Renate, E-mail: renate.fetzer@kit.edu; Igitkhanov, Yuri; Bazylev, Boris

    2015-10-15

    Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

  9. European blanket development for a demo reactor

    Giancarli, L.; Proust, E.; Anzidei, L.

    1994-01-01

    There are four breeding blanket concepts for a fusion DEMO reactor under development within the framework of the fusion technology programme of the European Union (EU). This paper describes the design of these concepts, the accompanying R + D programme and the status of the development. (authors). 8 figs., 1 tab

  10. Simulations with COREDIV Code of DEMO Discharges

    Zagorski, R.; Stankiewicz, R.; Ivanova-Stanik, I., E-mail: roman.zagorski@ipplm.pl [Institute of Plasma Physics and Laser Microfusion, Warsaw (Poland)

    2012-09-15

    Full text: The reduction of divertor target power load due to radiation of sputtered and externally seeded impurities in fusion reactor is investigated in this paper. The approach is based on integrated numerical modelling of DEMO discharges using the COREDIV code, which self-consistently solves 1D radial transport equations of plasma and impurities in the core region and 2D multifluid transport in the SOL. The model is fully self-consistent with respect to both the effects of impurities on the alpha-power level and the interaction between seeded and intrinsic impurities. The code has been already successfully benchmarked with the data from present day experiments (JET, ADEX). Calculations have been performed for inductive DEMO scenario and DEMO Steady-State configuration with tungsten walls and Ar seeding. In case of DEMO Steady-State scenario strong increase of Z{sub eff} and significant reduction of the alpha power are observed with the increase of Ar influx which is caused by the decrease of fuel ions density due to the dilution effect. It leads to the reduction of the target plate heat loads but surprisingly the radiation level remains almost constant with the increased seeding which is the result of the interplay between the energy losses and tungsten source due to sputtering processes. It has been found that the W radiation is the dominant energy loss mechanism and it accounts for 90% of all radiation losses. In case of pulsed DEMO scenario, it appears that the helium accumulation might be a serious problem. Even without seeding the resulting Z{sub eff} is very large (> 2.6) and consequently only relatively weak seeding can be applied for pulsed scenario. It is found that helium accumulation depends strongly on the transport model used for helium, if the helium diffusion is increased than the accumulation effect is mitigated. (author)

  11. Energy Storage System for a Pulsed DEMO

    Lucas, J.; Cortes, M.; Mendez, P.; Maisonnier, D.; Hayward, J.

    2006-01-01

    Several designs have been proposed for DEMO, some of which will operate in pulsed mode. Since a fusion power plant will be required to deliver continuous output, this challenge must be solved. For the reference DEMO, energy storage is required at a level of 250 MWhe with a capability of delivering a power of 1 GWe. Although DEMO is scheduled to be built in about 30 years, the design of the energy storage system must be based on current technology, focusing on commercially available products and on their expected future trends. From a thorough review of the different technologies available, thermal energy storage, compressed air energy storage, water pumping, fuel cells, batteries, flywheels and ultracapacitors are the most promising solutions to energy storage for a pulsed DEMO. An outline of each of these technologies is described in the paper, showing its basis, features, advantages and disadvantages for this application. Following this review, the most suitable methods capable of storing the required energy are examined. Fuel cells are not suitable due to the power requirement. Compressed air energy storage has a lower efficiency than the required one. Thermal energy storage, based on molten salts, so more energy can be stored with a better efficiency, and water pumping are shown as the main solutions, based on existing technology. However, those are not the only solutions capable of solving our challenge. Hydrogen production, using water electrolysis, hydrogen storage and combustion in a combined cycle can achieve our energy and power requirements with an acceptable efficiency. All these solutions are studied in detail and described, evaluating their current cost and efficiency in order to compare them all. (author)

  12. DEMO port plug design and integration studies

    Grossetti, G.; Boccaccini, L. V.; Cismondi, F.; Del Nevo, A.; Fischer, U.; Franke, T.; Granucci, G.; Hernández, F.; Mozzillo, R.; Strauß, D.; Tran, M. Q.; Vaccaro, A.; Villari, R.

    2017-11-01

    The EUROfusion Consortium established in 2014 and composed by European Fusion Laboratories, and in particular the Power Plant Physics and Technology department aims to develop a conceptual design for the Fusion DEMOnstration Power Plant, DEMO. With respect to present experimental machines and ITER, the main goals of DEMO are to produce electricity continuously for a period of about 2 h, with a net electrical power output of a few hundreds of MW, and to allow tritium self-sufficient breeding with an adequately high margin in order to guarantee its planned operational schedule, including all planned maintenance intervals. This will eliminate the need to import tritium fuel from external sources during operations. In order to achieve these goals, extensive engineering efforts as well as physics studies are required to develop a design that can ensure a high level of plant reliability and availability. In particular, interfaces between systems must be addressed at a very early phase of the project, in order to proceed consistently. In this paper we present a preliminary design and integration study, based on physics assessments for the EU DEMO1 Baseline 2015 with an aspect ratio of 3.1 and 18 toroidal field coils, for the DEMO port plugs. These aim to host systems like electron cyclotron heating launchers currently developed within the Work Package Heating and Current Drive that need an external radial access to the plasma and through in-vessel systems like the breeder blanket. A similar approach shown here could be in principle followed by other systems, e.g. other heating and current drive systems or diagnostics. The work addresses the interfaces between the port plug and the blanket considering the helium-cooled pebble bed and the water cooled lithium lead which are two of four breeding blanket concepts under investigation in Europe within the Power Plant Physics and Technology Programme: the required openings will be evaluated in terms of their impact onto the

  13. Diagnostics for plasma control on DEMO: challenges of implementation

    Donne, A. J. H.; Costley, A. E.; Morris, A. W.

    2012-01-01

    As a test fusion power plant, DEMO will have to demonstrate reliability and very long pulse/steady-state operation, which calls for unprecedented robustness and reliability of all diagnostic systems (also requiring adequate redundancy). But DEMO will have higher levels of neutron and gamma fluxes,

  14. Energy storage system for a pulsed DEMO

    Lucas, J.; Cortes, M.; Mendez, P.; Hayward, J.; Maisonnier, D.

    2007-01-01

    Several designs have been proposed for the DEMO fusion reactor. Some of them are working in a non-steady state mode. Since a power plant should be able to deliver to the grid a constant power, this challenge must be solved. Energy storage is required at a level of 250 MWh e with the capability of delivering a power of 1 GWe. A review of different technologies for energy storage is made. Thermal energy storage (TES), fuel cells and other hydrogen storage, compressed air storage, water pumping, batteries, flywheels and supercapacitors are the most promising solutions to energy storage. Each one is briefly described in the paper, showing its basis, features, advantages and disadvantages for this application. The conclusion of the review is that, based on existing technology, thermal energy storage using molten salts and a system based on hydrogen storage are the most promising candidates to meet the requirements of a pulsed DEMO. These systems are investigated in more detail together with an economic assessment of each

  15. A Fast-Track Path to DEMO Enabled by ITER and FNSF-AT

    Garofalo, A. M.; Choi, M.; Humphreys, D. A.; Kinsey, J. E.; Lao, L. L.; Snyder, P. B.; John, H. E.St.; Turnbull, A. D.; Taylor, T.S., E-mail: garofalo@fusion.gat.com [General Atomics, San Diego (United States); Chan, V. S.; Canik, J. M. [Oak Ridge National Laboratory, Oak Ridge (United States); Sawan, M. E. [University of Wisconsin, Madison (United States); Stangeby, P. C. [University of Toronto Institute for Aerospace Studies, Toronto (Canada)

    2012-09-15

    Full text: A Fusion Nuclear Science Facility based on the Advanced Tokamak concept (FNSF-AT) [1] is a key element of a fast track plan to a commercially attractive fusion DEMO. The next step forward on the path towards fusion commercialization must be a device that complements ITER in addressing the community identified science and technology gaps to DEMO, and that enables a DEMO construction decision triggered by the achievement of Q = 10 in ITER, presently scheduled for the year 2030. This paper elucidates the logic flow leading to the FNSF-AT approach for such a next step forward, and presents the results of recent analysis resolving key physics and engineering issues. A FNSF-AT will show fusion can make its own fuel, provide a materials irradiation facility, show fusion can produce high-grade process heat and electricity. In order to accomplish these goals, the FNSF has to operate steady-state with significant duty cycle and significant neutron fluence. In FNSF-AT, advanced tokamak physics enables steady-state burning plasmas with the high fluence required for FNSF's nuclear science development objective, in the compact size required to demonstrate Tritium fuel self-sufficiency using only a moderate quantity of the limited supply of Tritium. Physics based integrated modeling has found a steady-state baseline equilibrium with good stability and controllability properties. 2-D analysis assuming ITER heat and particle diffusion coefficients in the SOL predicts peak heat flux < 10 MW/m{sup 2} at the outer divertor targets. High fidelity and high-resolution 3D neutronics calculations have also been carried out, showing acceptable cumulative end-of-life organic insulator dose levels in all the device coils, and TBR > 1 for two blanket concepts considered. This FNSF-AT baseline plasma scenario has significant margin to meet the FNSF nuclear science mission. Moreover, the facility allows the development of more advanced scenarios to close the physics gaps to DEMO

  16. Neutronics requirements for a DEMO fusion power plant

    Fischer, U., E-mail: ulrich.fischer@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bachmann, C. [EUROfusion Consortium , Boltzmannstraße 2, 85748 Garching (Germany); Palermo, I. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Pereslavtsev, P. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Villari, R. [ENEA UT-FUS C.R. Frascati, Via E. Fermi 45, 00044 Frascati (Italy)

    2015-10-15

    Highlights: • Discussion and specification of neutronic requirements for a DEMO power plant. • TBR uncertainties are reviewed/discussed and design margins are elaborated. • Limits are given for radiation loads to super-conducting magnets and steel structural components. • Available DEMO results are compared to recommended limits and TBR design target. - Abstract: This paper addresses the neutronic requirements a DEMO fusion power plant needs to fulfil for a reliable and safe operation. The major requirement is to ensure Tritium self-sufficiency taking into account the various uncertainties and plant-internal losses that occur during DEMO operation. A further major requirement is to ensure sufficient protection of the superconducting magnets against the radiation penetrating in-vessel components and vessel. Reliable criteria for the radiation loads need to be defined and verified to ensure the reliable operation of the magnets over the lifetime of DEMO. Other issues include radiation induced effects on structural materials such as the accumulated displacement damage, the generation of gases such as helium which may deteriorate the material performance. The paper discusses these issues and their impact on design options for DEMO taking into account results obtained in the frame of European Power Plant Physics and Technology (PPPT) 2013 programme activities with DEMO models employing the helium cooled pebble bed (HCPB), the helium cooled lithium lead (HCLL), and the water-cooled (WCLL) blanket concepts.

  17. Comparative evaluation of remote maintenance schemes for fusion DEMO reactor

    Utoh, Hiroyasu, E-mail: uto.hiroyasu@jaea.go.jp; Tobita, Kenji; Someya, Youji; Asakura, Nobuyuki; Sakamoto, Yoshiteru; Hoshino, Kazuo; Nakamura, Makoto

    2015-10-15

    Highlights: • Various remote maintenance schemes for DEMO were comparatively assessed based on requirements for DEMO remote maintenance. • The banana shape segment transport using all vertical maintenance ports would be more probable DEMO reactor maintenance scheme. • The key engineering issues are in-vessel transferring mechanism of segment, pipe connection and conducting shell design for plasma vertical stability. - Abstract: Maintenance schemes are one of the critical issues in DEMO design, significantly affecting the configuration of in-vessel components, the size of toroidal field (TF) coil, the arrangement of poloidal field (PF) coils, reactor building, hot cell and so forth. Therefore, the maintenance schemes should satisfy many design requirements and criteria to assure reliable and safe plant operation and to attain reasonable plant availability. The plant availability depends on reliability of remote maintenance scheme, inspection of pipe connection and plasma operation. In this paper, various remote maintenance schemes for DEMO were comparatively assessed based on requirements for DEMO remote maintenance. From the view points of the reliability of inspection on hot cell, TF coil size, stored energy of PF coil and portability of segment, the banana shape segment transport using all vertical maintenance ports would be more probable DEMO reactor maintenance scheme, and it has key engineering issues such as in-vessel transferring mechanism of segment, pipe connection and conducting shell design for plasma vertical stability.

  18. Comparative evaluation of remote maintenance schemes for fusion DEMO reactor

    Utoh, Hiroyasu; Tobita, Kenji; Someya, Youji; Asakura, Nobuyuki; Sakamoto, Yoshiteru; Hoshino, Kazuo; Nakamura, Makoto

    2015-01-01

    Highlights: • Various remote maintenance schemes for DEMO were comparatively assessed based on requirements for DEMO remote maintenance. • The banana shape segment transport using all vertical maintenance ports would be more probable DEMO reactor maintenance scheme. • The key engineering issues are in-vessel transferring mechanism of segment, pipe connection and conducting shell design for plasma vertical stability. - Abstract: Maintenance schemes are one of the critical issues in DEMO design, significantly affecting the configuration of in-vessel components, the size of toroidal field (TF) coil, the arrangement of poloidal field (PF) coils, reactor building, hot cell and so forth. Therefore, the maintenance schemes should satisfy many design requirements and criteria to assure reliable and safe plant operation and to attain reasonable plant availability. The plant availability depends on reliability of remote maintenance scheme, inspection of pipe connection and plasma operation. In this paper, various remote maintenance schemes for DEMO were comparatively assessed based on requirements for DEMO remote maintenance. From the view points of the reliability of inspection on hot cell, TF coil size, stored energy of PF coil and portability of segment, the banana shape segment transport using all vertical maintenance ports would be more probable DEMO reactor maintenance scheme, and it has key engineering issues such as in-vessel transferring mechanism of segment, pipe connection and conducting shell design for plasma vertical stability.

  19. Phenomena Identification and Ranking Table for K-DEMO

    Oh, Kye Min; Kang, Myung Suk; Heo, Gyun Young; Kim, Hyoung Chan

    2013-01-01

    The purpose of this paper is to describe the current status of the Phenomena Identification Ranking Table (PIRT) for the conceptual design of K-DEMO, Korean Fusion DEMO Plant. K-DEMO is to be planned as the first fusion power plant constructed in South Korea. However, several key technologies such as plasma, materials, and cooling still have large uncertainties. There are also no relevant references to facilitate the design process of K-DEMO due to its different size, commercializing purpose, and regulatory framework. It was proposed to define the phenomena of systems, components, and processes in an accident condition. In this paper, PIRT for K-DEMO was described and analysis based on this tool was performed. We have carried out researches related to safety for fusion power plant in collaboration with the academies funded by NFRI during the past 3 years. As part of this research, Integrated Safety Assessment Methodology (ISAM), which was used to develop GEN-IV nuclear systems, was used to determine the technical safety issues and regulatory requirements for K-DEMO. PIRT is one of ISAM tools. It can recognize vulnerabilities of systems and identify the gaps in technical areas requiring additional research. The results through this tool are expected to contribute on detailed design for K-DEMO as guidance for regulatory requirements and safety systems in the future

  20. Phenomena Identification and Ranking Table for K-DEMO

    Oh, Kye Min; Kang, Myung Suk; Heo, Gyun Young [Kyung Hee Univ., Yongin (Korea, Republic of); Kim, Hyoung Chan [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The purpose of this paper is to describe the current status of the Phenomena Identification Ranking Table (PIRT) for the conceptual design of K-DEMO, Korean Fusion DEMO Plant. K-DEMO is to be planned as the first fusion power plant constructed in South Korea. However, several key technologies such as plasma, materials, and cooling still have large uncertainties. There are also no relevant references to facilitate the design process of K-DEMO due to its different size, commercializing purpose, and regulatory framework. It was proposed to define the phenomena of systems, components, and processes in an accident condition. In this paper, PIRT for K-DEMO was described and analysis based on this tool was performed. We have carried out researches related to safety for fusion power plant in collaboration with the academies funded by NFRI during the past 3 years. As part of this research, Integrated Safety Assessment Methodology (ISAM), which was used to develop GEN-IV nuclear systems, was used to determine the technical safety issues and regulatory requirements for K-DEMO. PIRT is one of ISAM tools. It can recognize vulnerabilities of systems and identify the gaps in technical areas requiring additional research. The results through this tool are expected to contribute on detailed design for K-DEMO as guidance for regulatory requirements and safety systems in the future.

  1. Assessment of DEMO challenges in technology and physics

    Zohm, Hartmut, E-mail: zohm@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, EURATOM Association, D-85748 Garching (Germany)

    2013-10-15

    Highlights: ► It is very important to respect the interlinks between physics and technology when developing designs for DEMO. ► Pulsed operation of a tokamak DEMO should seriously be considered in conservative DEMO designs. ► Optimization of both plasma CD efficiency as well as wall plug efficiency of the CD system is important. ► Exhaust requirements lead to an unprecedented high level of core radiation loss by impurity seeding in DEMO. -- Abstract: The challenges that DEMO designs encounter in both technology and physics are reviewed. It is shown that it is very important to respect the interlinks between these fields when developing designs for DEMO. Examples for areas where such interlinks put very strict requirements are the development of a steady state tokamak operation scenario and the question of power exhaust taking into account the boundary conditions set by materials questions. Concerning steady state operation, we find that demands on the physics scenario are so high that pulsed operation of a tokamak DEMO should seriously be considered in conservative DEMO designs. Alternatively, the device could foresee a large fraction of externally driven current which calls for optimization of both plasma CD efficiency as well as wall plug efficiency of the CD system. In the exhaust area, a realistic estimate of the admissable time averaged peak heat flux at the target is of the order of 5 MW/m{sup 2}, leading to strict requirements for the operational scenario, which has to rely on an unprecedented high level of radiation loss by impurity seeding and the facilitation of partial detachment. Thus, exhaust scenarios along these lines have to be developed which are compatible with the confinement needs and the H-L back transition power for DEMO. In both areas, we discuss possible risk mitigation strategies based on conceptually different approaches.

  2. Japanese endeavors to establish technological bases for DEMO

    Yamada, Hiroshi, E-mail: yamada.hiroshi@nifs.ac.jp [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Kasada, Ryuta [Institute of Advanced Energy, Kyoto University, Uji, Kyoto 611-0011 (Japan); Ozaki, Akira [Japan Atomic Industrial Forum, Inc., Minato-ku, Tokyo 105-8605 (Japan); Sakamoto, Ryuichi [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Sakamoto, Yoshiteru [Rokkasho Fusion Institute, Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212 (Japan); Takenaga, Hidenobu [Naka Fusion Institute, Japan Atomic Energy Agency, Naka, Ibaraki 311-0193 (Japan); Tanaka, Teruya [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Tanigawa, Hisashi [Rokkasho Fusion Institute, Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212 (Japan); Okano, Kunihiko [Keio University, Yokohama, Kanagawa 223-0061 (Japan); Tobita, Kenji [Rokkasho Fusion Institute, Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212 (Japan); Kaneko, Osamu [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Ushigusa, Kenkichi [Rokkasho Fusion Institute, Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212 (Japan)

    2016-11-01

    Highlights: • The strategy for DEMO has been discussed by a joint effort in Japan. • DEMO should be aimed at steady power generation beyond several hundred MW. • DEMO should be aimed at availability extendable to commercialization. • DEMO should be aimed at tritium breeding to fulfill self-sufficiency of fuels. • Related actions are emerging to deliberate the Japanese fusion roadmap. - Abstract: The establishment of technology bases required for the development of a fusion demonstration reactor (DEMO) has been discussed by a joint effort throughout the Japanese fusion community. The basic concept of DEMO premised for investigation has been identified and the structure of technological issues to ensure the feasibility of this DEMO concept has been examined. The Joint-Core Team, which was launched along with the request by the ministerial council, has compiled analyses in two reports to clarify technology which should be secured, maintained, and developed in Japan, to share the common targets among industry, government, and academia, and to activate actions under a framework for implementation throughout Japan. The reports have pointed out that DEMO should be aimed at steady power generation beyond several hundred thousand kilowatts, availability which must be extended to commercialization, and overall tritium breeding to fulfill self-sufficiency of fuels. The necessary technological activities, such as superconducting coils, blanket, divertor, and others, have been sorted out and arranged in the chart with the time line toward the decision on DEMO. Based upon these Joint-Core Team reports, related actions are emerging to deliberate the Japanese fusion roadmap.

  3. He-cooled divertor development for DEMO

    Norajitra, P.; Giniyatulin, R.; Ihli, T.; Janeschitz, G.; Krauss, W.; Kruessmann, R.; Kuznetsov, V.; Mazul, I.; Widak, V.; Ovchinnikov, I.; Ruprecht, R.; Zeep, B.

    2007-01-01

    Goal of the He-cooled divertor development for future fusion power plants is to resist a high heat flux of at least 10 MW/m 2 . The development includes the fields of design, analyses, and experiments. A helium-cooled modular jet concept (HEMJ) has been defined as reference solution, which is based on jet impingement cooling. In cooperation with the Efremov Institute, work was aimed at construction and high heat flux tests of prototypical tungsten mockups to demonstrate their manufacturability and their performances. A helium loop was built for this purpose to simulate the realistic thermo-hydraulics conditions close to those of DEMO (10 MPa He, 600 deg. C). The first high heat flux test results confirm the feasibility and the performance of the divertor design

  4. European DEMO BOT solid breeder blanket

    Dalle Donne, M.

    1994-11-01

    The BOT (Breeder Outside Tube) Solid Breeder Blanket for a fusion DEMO reactor is presented. This is one of the four blanket concepts under development in the frame of the European fusion technology program with the aim to select in 1995 the two most promising ones for further development. In the paper the reference blanket design and external loops are described as well as the results of the theoretical and experimental work in the fields of neutronics, thermohydraulics, mechanical stresses, tritium control and extraction, development and irradiation of the ceramic breeder material, beryllium development, ferromagnetic forces caused by disruptions, safety and reliability. An outlook is given on the remaining open questions and on the required R and D program. (orig.) [de

  5. DEMO development strategy based on China FPP program

    Pan Chuanhong; Feng, K.M.; Wu, W.C.; Liu, S.L.

    2007-01-01

    The DEMO in China is to demonstrate the safety, reliability and environment feasibility of the fusion power plants, while to demonstrate the prospective economic feasibility of the commercial fusion power plants. Considering that there is still a long way to go towards an economically competitive commercial power plant, DEMO in China should be an indispensable step prior to the commercial one. Two options of breeding blanket with ceramic and lead lithium breeders might be chosen as DEMO concepts under the conditions of meeting the requirement of the neutronics, thermal-hydraulics and mechanics aspects. The DEMO development strategy, related R and D activities, based on China fusion power plant (FPP) program are presented. (orig.)

  6. Advances in the physics basis for the European DEMO design

    Wenninger, R.; Arbeiter, F.; Aubert, J.; Aho-Mantila, L.; Albanese, R.; Ambrosino, R.; Angioni, C.; Artaud, J.-F.; Bernert, M.; Fable, E.; Fasoli, A.; Federici, G.; Garcia, J.; Giruzzi, G.; Jenko, F.; Maget, P.; Mattei, M.; Maviglia, F.; Poli, E.; Ramogida, G.; Reux, C.; Schneider, M.; Sieglin, B.; Villone, F.; Wischmeier, M.; Zohm, H.

    2015-06-01

    In the European fusion roadmap, ITER is followed by a demonstration fusion power reactor (DEMO), for which a conceptual design is under development. This paper reports the first results of a coherent effort to develop the relevant physics knowledge for that (DEMO Physics Basis), carried out by European experts. The program currently includes investigations in the areas of scenario modeling, transport, MHD, heating & current drive, fast particles, plasma wall interaction and disruptions.

  7. Technology and Plasma Physics Developments needed for DEMO

    Lackner, K.

    2006-01-01

    Although no universally agreed definition of the next step after ITER exists at present it is commonly accepted that significant progress beyond the ITER base-line operating physics modes and the technologies employed in it are needed. We first review the role of DEMO in the different proposed fusion road maps and derive from them the corresponding performance requirements. A fast track to commercial fusion implies that DEMO is already close to a first of a kind power plant in all aspects except average availability. Existing power plant studies give therefore also a good approximation to the needs of DEMO. We outline the options for achieving the needed physics progress in the different characteristic parameters, and the implications for the experimental programme of ITER and accompanying satellite devices. On the time scale of the operation of ITER and of the planning DEMO, ab-initio modelling of fusion plasmas is also expected to assume a qualitatively new role. Besides the mapping of the reactor regime of plasma physics and the integration of a burning plasma with the principal reactor technologies on ITER, the development of functional and structural materials capable of handling the high power fluxes and neutron fluences, respectively is also on the critical path to DEMO. Finally we discuss the potential contributions of other confinement concepts (stellarators and spherical tokamaks) to the design of DEMO. (author)

  8. Neutronic performance issues of the breeding blanket options for the European DEMO fusion power plant

    Fischer, U., E-mail: ulrich.fischer@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bachmann, C. [EUROfusion—Programme Management Unit, Boltzmannstr. 2, 85748 Garching (Germany); Jaboulay, J.-C. [CEA-Saclay, DEN, DM2S, SERMA, LPEC, 91191 Gif-sur-Yvette (France); Moro, F. [ENEA, Dipartimento Fusione e tecnologie per la Sicurezza Nucleare, Via E. Fermi 45, 00044 Frascati, Rome (Italy); Palermo, I. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Pereslavtsev, P. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Villari, R. [ENEA, Dipartimento Fusione e tecnologie per la Sicurezza Nucleare, Via E. Fermi 45, 00044 Frascati, Rome (Italy)

    2016-11-01

    Highlights: • Breeder blanket concepts for DEMO—design features. • Neutronic characteristics of breeder blankets. • Evaluation of Tritium breeding potential. • Evaluation of shielding performance. - Abstract: This paper presents nuclear performance issues of the HCPB, HCLL, DCLL and WCLL breeder blankets, which are under development within the PPPT (Power Plant Physics and Technology) programme of EUROfusion, with the objective to assess the potential and suitability of the blankets for the application to DEMO. The assessment is based on the initial design versions of the blankets developed in 2014. The Tritium breeding potential is considered sufficient for all breeder blankets although the initial design versions of the HCPB, HCLL and DCLL blankets were shown to require further design improvements. Suitable measures have been proposed and proven to be sufficient to achieve the required Tritium Breeding Ratio (TBR) ≥ 1.10. The shielding performance was shown to be sufficient to protect the super-conducting toroidal field coil provided that efficient shielding material mixtures including WC or borated water are utilized. The WCLL blanket does not require the use of such shielding materials due to a very compact blanket support structure/manifold configuration which yet requires design verification. The vacuum vessel can be safely operated over the full anticipated DEMO lifetime of 6 full power years for all blanket concepts considered.

  9. Demonstration tokamak-power-plant study (DEMO)

    1982-09-01

    A study of a Demonstration Tokamak Power Plant (DEMO) has been completed. The study's objective was to develop a conceptual design of a prototype reactor which would precede commercial units. Emphasis has been placed on defining and analyzing key design issues and R and D needs in five areas: noninductive current drivers, impurity control systems, tritium breeding blankets, radiation shielding, and reactor configuration and maintenance features. The noninductive current drive analysis surveyed a wide range of candidates and selected relativistic electron beams for the reference reactor. The impurity control analysis considered both a single-null poloidal divertor and a pumped limiter. A pumped limiter located at the outer midplane was selected for the reference design because of greater engineering simplicity. The blanket design activity focused on two concepts: a Li 2 O solid breeder with high pressure water cooling and a lead-rich Li-Pb eutectic liquid metal breeder (17Li-83Pb). The reference blanket concept is the Li 2 O option with a PCA structural material. The first wall concept is a beryllium-clad corrugated panel design. The radiation shielding effort concentrated on reducing the cost of bulk and penetration shielding; the relatively low-cost outborad shield is composed of concrete, B 4 C, lead, and FE 1422 structural material

  10. Development of a master model concept for DEMO vacuum vessel

    Mozzillo, Rocco; Marzullo, Domenico; Tarallo, Andrea [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy); Bachmann, Christian [EUROfusion PMU, Boltzmannstraße 2, 85748 Garching (Germany); Di Gironimo, Giuseppe, E-mail: peppe.digironimo@gmail.com [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy)

    2016-11-15

    Highlights: • The present work concerns the development of a first master concept model for DEMO vacuum vessel. • A parametric-associative CAD master model concept of a DEMO VV sector has been developed in accordance with DEMO design guidelines. • A proper CAD design methodology has been implemented in view of the later FEM analyses based on “shell elements”. - Abstract: This paper describes the development of a master model concept of the DEMO vacuum vessel (VV) conducted within the framework of the EUROfusion Consortium. Starting from the VV space envelope defined in the DEMO baseline design 2014, the layout of the VV structure was preliminarily defined according to the design criteria provided in RCC-MRx. A surface modelling technique was adopted and efficiently linked to the finite element (FE) code to simplify future FE analyses. In view of possible changes to shape and structure during the conceptual design activities, a parametric design approach allows incorporating modifications to the model efficiently.

  11. Development of a master model concept for DEMO vacuum vessel

    Mozzillo, Rocco; Marzullo, Domenico; Tarallo, Andrea; Bachmann, Christian; Di Gironimo, Giuseppe

    2016-01-01

    Highlights: • The present work concerns the development of a first master concept model for DEMO vacuum vessel. • A parametric-associative CAD master model concept of a DEMO VV sector has been developed in accordance with DEMO design guidelines. • A proper CAD design methodology has been implemented in view of the later FEM analyses based on “shell elements”. - Abstract: This paper describes the development of a master model concept of the DEMO vacuum vessel (VV) conducted within the framework of the EUROfusion Consortium. Starting from the VV space envelope defined in the DEMO baseline design 2014, the layout of the VV structure was preliminarily defined according to the design criteria provided in RCC-MRx. A surface modelling technique was adopted and efficiently linked to the finite element (FE) code to simplify future FE analyses. In view of possible changes to shape and structure during the conceptual design activities, a parametric design approach allows incorporating modifications to the model efficiently.

  12. Issues and strategies for DEMO in-vessel component integration

    Bachmann, C.; Arbeiter, F.; Boccaccini, L.V.; Coleman, M.; Federici, G.; Fischer, U.; Kemp, R.; Maviglia, F.; Mazzone, G.; Pereslavtsev, P.; Roccella, R.; Taylor, N.; Villari, R.; Villone, F.; Wenninger, R.; You, J.-H.

    2016-01-01

    In the frame of the EUROfusion Consortium activities were launched in 2014 to develop a concept of a DEMO reactor including a large R&D program and the integrated design of the tokamak systems. The integration of the in-vessel components (IVCs) must accommodate numerous constraints imposed by their operating environment, the requirements for precise alignment, high performance, reliability, and remote maintainability. This makes the development of any feasible design a major challenge. Although DEMO is defined to be a one-of-a-kind device there needs to be in addition to the development of the IVC design solutions a remarkable emphasis on the optimization of these solutions already at the conceptual level. Their design has a significant impact on the machine layout, complexity, and performance. This paper identifies design and technology limitations of IVCs, their consequences on the integration principles, and introduces strategies currently considered in the DEMO tokamak design approach.

  13. Comparative study of cost models for tokamak DEMO fusion reactors

    Oishi, Tetsutarou; Yamazaki, Kozo; Arimoto, Hideki; Ban, Kanae; Kondo, Takuya; Tobita, Kenji; Goto, Takuya

    2012-01-01

    Cost evaluation analysis of the tokamak-type demonstration reactor DEMO using the PEC (physics-engineering-cost) system code is underway to establish a cost evaluation model for the DEMO reactor design. As a reference case, a DEMO reactor with reference to the SSTR (steady state tokamak reactor) was designed using PEC code. The calculated total capital cost was in the same order of that proposed previously in cost evaluation studies for the SSTR. Design parameter scanning analysis and multi regression analysis illustrated the effect of parameters on the total capital cost. The capital cost was predicted to be inside the range of several thousands of M$s in this study. (author)

  14. Issues and strategies for DEMO in-vessel component integration

    Bachmann, C., E-mail: christian.bachmann@euro-fusion.org [EUROfusion PMU, Garching (Germany); Arbeiter, F.; Boccaccini, L.V. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Coleman, M.; Federici, G. [EUROfusion PMU, Garching (Germany); Fischer, U. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Kemp, R. [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Maviglia, F. [EUROfusion PMU, Garching (Germany); Mazzone, G. [ENEA Dipartimento Fusione e Sicurezza Nucleare C. R. Frascati – via E. Fermi 45, 00044 Frascati, Roma (Italy); Pereslavtsev, P. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Roccella, R. [ITER Organization, St. Paul Lez Durance (France); Taylor, N. [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Villari, R. [ENEA Dipartimento Fusione e Sicurezza Nucleare C. R. Frascati – via E. Fermi 45, 00044 Frascati, Roma (Italy); Villone, F. [ENEA-CREATE Association, DIEI, Università di Cassino e del Lazio Meridiona (Italy); Wenninger, R. [EUROfusion PMU, Garching (Germany); You, J.-H. [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Garching (Germany)

    2016-11-15

    In the frame of the EUROfusion Consortium activities were launched in 2014 to develop a concept of a DEMO reactor including a large R&D program and the integrated design of the tokamak systems. The integration of the in-vessel components (IVCs) must accommodate numerous constraints imposed by their operating environment, the requirements for precise alignment, high performance, reliability, and remote maintainability. This makes the development of any feasible design a major challenge. Although DEMO is defined to be a one-of-a-kind device there needs to be in addition to the development of the IVC design solutions a remarkable emphasis on the optimization of these solutions already at the conceptual level. Their design has a significant impact on the machine layout, complexity, and performance. This paper identifies design and technology limitations of IVCs, their consequences on the integration principles, and introduces strategies currently considered in the DEMO tokamak design approach.

  15. A preliminary conceptual design study for Korean fusion DEMO reactor

    Kim, Keeman, E-mail: kkeeman@nfri.re.kr [National Fusion Research Institute, 169-148 Gwahak-ro, Daejeon 305-806 (Korea, Republic of); Kim, Hyoung Chan; Oh, Sangjun; Lee, Young Seok; Yeom, Jun Ho; Im, Kihak; Lee, Gyung-Su [National Fusion Research Institute, 169-148 Gwahak-ro, Daejeon 305-806 (Korea, Republic of); Neilson, George; Kessel, Charles; Brown, Thomas; Titus, Peter [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543 (United States)

    2013-10-15

    Highlights: ► Perform a preliminary conceptual study for a steady-state Korean DEMO reactor. ► Present design guidelines and requirements of Korean DEMO reactor. ► Present a preliminary design of TF (toroidal field) and CS (central solenoid) magnet. ► Present a preliminary result of the radial build scheme of Korean DEMO reactor. -- Abstract: As the ITER is being constructed, there is a growing anticipation for an earlier realization of fusion energy, so called fast-track approach. Korean strategy for fusion energy can be regarded as a fast-track approach and one special concept discussed in this paper is a two-stage development plan. At first, a steady-state Korean DEMO Reactor (K-DEMO) is designed not only to demonstrate a net electricity generation and a self-sustained tritium cycle, but also to be used as a component test facility. Then, at its second stage, a major upgrade is carried out by replacing in-vessel components in order to show a net electric generation on the order of 300 MWe and the competitiveness in cost of electricity (COE). The major radius is designed to be just below 6.5 m, considering practical engineering feasibilities. By using high performance Nb{sub 3}Sn-based superconducting cable currently available, high magnetic field at the plasma center above 8 T can be achieved. A design concept for TF magnets and radial builds for the K-DEMO considering a vertical maintenance scheme, are presented together with preliminary design parameters.

  16. Relevance of NET first wall concept for DEMO DN

    Kiltie, J.S.

    1987-01-01

    Design studies for the Next European Torus (NET) have produced a design concept for the first wall. This concept features poloidal water cooling, double contained in a welded steel structure which is protected by radiatively cooled tiles. In this appendix the relevance of this concept to a DEMO is examined with particular emphasis given to the ability of the cooling tube arrangement to remove the heat. A suggested modification to the arrangement of coolant tubes is suggested so that the design can operate at the higher loadings of a DEMO. (author)

  17. A preliminary systems assessment of the Starlite Demo candidates

    Bathke, C.G.

    1995-01-01

    The Starlite project has evaluated the following five tokamaks as candidates for the US Demo Power Plant: (1) steady state, first stability regime; (2) pulsed, first stability regime; (3) steady state, second stability regime; (4) steady state, reversed shear; and (5) steady state, low aspect ratio. Systems analysis of these candidates has played an important role in the selection of a reversed-shear tokamak for further conceptual design as a US Demo Power Plant. The cost-based systems analysis that led to the selection of a reversed-shear tokamak is described herein

  18. On tungsten technologies and qualification for DEMO

    Laan, J. van der; Hegeman, H.; Wouters, O.; Luzginova, N.; Jonker, B.; Van der Marck, S.; Opschoor, J.; Wang, J.; Dowling, G.; Stuivenga, M.; Carton, E.

    2009-01-01

    Tungsten alloys are considered prime candidates for the in-vessel components directly facing the plasma. For example, in the HEMJ helium cooled divertor design tiles may be operated at temperatures up to 1700 deg. C, supported by a structure partially consisting of tungsten at temperatures from 600 to 1000 deg. C, and connected to a HT steel structure. The tungsten armoured primary wall is operated at 500-900 deg. C. Irradiation doses will be few tens dpa at minimum, but FPR requirements for plants availability will stretch these targets. Recently injection moulding technology was developed for pure tungsten and representative parts were manufactured for ITER monobloc divertors and DEMO HEMJ thimbles. The major advantages for this technology are the efficient use of material feedstock/resources and the intrinsic possibility to produce near-finished product, avoiding machining processes that are costly and may introduce surface defects deteriorating the component in service performance. It is well suited for mass-manufacturing of components as well known in e.g. lighting industries. To further qualify this material technology various specimen types were produced with processing parameters identical to the components, and tested successfully, showing the high potential for implementation in (fusion) devices. Furthermore, the engineering approach can clearly be tailored away from conventional design and manufacturing technologies based on bulk materials. The technology is suitable for shaping of new W-alloys and W-ODS variants as well. Basically this technology allows a particular qualification trajectory. There is no need to produce large batches of material during the material development and optimization stage. For the verification of irradiation behaviour in the specific neutron spectra, there is a further attractive feature to use e.g. isotope tailored powders to adjust to available irradiation facilities like MTR's. In addition the ingrowth of transmutation

  19. Global shutdown dose rate maps for a DEMO conceptual design

    Leichtle, D.; Pereslavtsev, P.; Sanz, J.; Catalan, J.P.; Juarez, R.

    2015-01-01

    Highlights: • Application of R2S-method on high-resolution full torus sector mesh for DEMO. • Absorbed dose rates after shutdown for a variely of RH equipment at typical locations. • Idenification of radiation levels at several port based locations. - Abstract: For the calculations of highly reliable shutdown dose rate (SDR) maps in fusion devices like a DEMO plant, the Rigorous-2-step (R2S) method is nowadays routinely applied using high-resolution decay gamma sources from initial high-resolution neutron flux meshes activating all materials in the system. This approach has been utilized in the present paper with the objective to provide SDR results relevant for RH systems of a conceptual DEMO design developed in the EU. The primary objective was to assess specific locations of interest for RH equipment inside the vessel and along the extension of maintenance ports. To this end, a provisional DEMO MCNP model has been used, featuring HCLL-type blankets, tungsten/copper divertor, manifolds, vacuum vessel with ports and toroidal field coils. The operational scenario assumed 2.1 GW fusion power and a life-time of 20 years with plant availability of 30%, where removable parts will be extracted after 5.2 years. Results of absorbed dose rate distributions for several relevant materials are presented and discussed in terms of the different contributions from the various activated components.

  20. Scoping the parameter space for demo and the engineering test

    Meier, W R.

    1999-01-01

    In our IFE development plan, we have set a goal of building an Engineering Test Facility (ETF) for a total cost of $2B and a Demo for $3B. In Mike Campbell s presentation at Madison, we included a viewgraph with an example Demo that had 80 to 250 MWe of net power and showed a plausible argument that it could cost less than $3B. In this memo, I examine the design space for the Demo and then briefly for the ETF. Instead of attempting to estimate the costs of the drivers, I pose the question in a way to define R ampersand D goals: As a function of key design and performance parameters, how much can the driver cost if the total facility cost is limited to the specified goal? The design parameters examined for the Demo included target gain, driver energy, driver efficiency, and net power output. For the ETF; the design parameters are target gain, driver energy, and target yield. The resulting graphs of allowable driver cost determine the goals that the driver R ampersand D programs must seek to meet

  1. Configuration management of the EU DEMO conceptual design data

    Meszaros, Botond; Shannon, Mark; Marzullo, Domenico; Woodley, Colin; Rowe, Steve; Di Gironimo, Giuseppe

    2016-01-01

    Highlights: • Description of the selection of the DEMO Product Data Management tool. • Introduction of the DEMO configuration management philosophy for the CAD design data. • Description of the enabling tools and systems of the configuration management. - Abstract: The EUROfusion Consortium is setting up – as part of the EU Fusion Roadmap – the framework for the implementation of the (pre)conceptual design phase of the DEMO reactor. Configuration management needs have been identified as one of the key elements of this framework and is the topic of this paper, in particular the configuration of the CAD design data. The desire is to keep the definition and layout of the corresponding systems “light weight” and relatively easy to manage, whilst simultaneously providing a level of detail in the definition of the design configuration that is fit for the purpose of a conceptual design. This paper aims to describe the steps followed during the definition of the configuration management system of the DEMO design data in terms of (i) the identification of the appropriate product data management system, (ii) the description of the philosophy of the configuration management of the design data, and (iii) the introduction of the most important enabling processes.

  2. Design and development of ceramic breeder demo blanket

    Enoeda, M.; Sato, S.; Hatano, T.

    2001-01-01

    Ceramic breeder blanket development has been widely conducted in Japan from fundamental researches to project-oriented engineering scaled development. A long term R and D program has been launched in JAERI since 1996 as a course of DEMO blanket development. The objectives of this program are to provide engineering data base and fabrication technologies of the DEMO blanket, aiming at module testing in ITER currently scheduled to start from the beginning of the ITER operation as a near-term target. Two types of DEMO blanket systems, water cooled blanket and helium cooled blanket, have been designed to be consistent with the SSTR (Steady State Tokamak Reactor) which is the reference DEMO reactor design in JAERI. Both of them utilize packed small pebbles of breeder Li 2 O or Li 2 TiO 3 as a candidate) and neutron multiplier (Be) and rely on the development of advanced structural materials (a reduced activation ferritic steel F82H) compatible with high temperature operation. (author)

  3. Configuration management of the EU DEMO conceptual design data

    Meszaros, Botond; Shannon, Mark [EUROfusion Consortium, PPPT Department, Garching, Boltzmannstr. 2 (Germany); Marzullo, Domenico [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Woodley, Colin; Rowe, Steve [CCFE, Culham Science Centre, Oxfordshire OX14 3DB, Abingdon (United Kingdom); Di Gironimo, Giuseppe [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy)

    2016-11-01

    Highlights: • Description of the selection of the DEMO Product Data Management tool. • Introduction of the DEMO configuration management philosophy for the CAD design data. • Description of the enabling tools and systems of the configuration management. - Abstract: The EUROfusion Consortium is setting up – as part of the EU Fusion Roadmap – the framework for the implementation of the (pre)conceptual design phase of the DEMO reactor. Configuration management needs have been identified as one of the key elements of this framework and is the topic of this paper, in particular the configuration of the CAD design data. The desire is to keep the definition and layout of the corresponding systems “light weight” and relatively easy to manage, whilst simultaneously providing a level of detail in the definition of the design configuration that is fit for the purpose of a conceptual design. This paper aims to describe the steps followed during the definition of the configuration management system of the DEMO design data in terms of (i) the identification of the appropriate product data management system, (ii) the description of the philosophy of the configuration management of the design data, and (iii) the introduction of the most important enabling processes.

  4. Global shutdown dose rate maps for a DEMO conceptual design

    Leichtle, D., E-mail: dieter.leichtle@f4e.europa.eu [Karlsruhe Institute of Technology KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Pereslavtsev, P. [Karlsruhe Institute of Technology KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Sanz, J.; Catalan, J.P.; Juarez, R. [Universidad Nacional de Educación a Distancia(UNED), E.T.S. Ingenieros Industriales, C/ Juan del Rosal 12, 28040 Madrid (Spain)

    2015-10-15

    Highlights: • Application of R2S-method on high-resolution full torus sector mesh for DEMO. • Absorbed dose rates after shutdown for a variely of RH equipment at typical locations. • Idenification of radiation levels at several port based locations. - Abstract: For the calculations of highly reliable shutdown dose rate (SDR) maps in fusion devices like a DEMO plant, the Rigorous-2-step (R2S) method is nowadays routinely applied using high-resolution decay gamma sources from initial high-resolution neutron flux meshes activating all materials in the system. This approach has been utilized in the present paper with the objective to provide SDR results relevant for RH systems of a conceptual DEMO design developed in the EU. The primary objective was to assess specific locations of interest for RH equipment inside the vessel and along the extension of maintenance ports. To this end, a provisional DEMO MCNP model has been used, featuring HCLL-type blankets, tungsten/copper divertor, manifolds, vacuum vessel with ports and toroidal field coils. The operational scenario assumed 2.1 GW fusion power and a life-time of 20 years with plant availability of 30%, where removable parts will be extracted after 5.2 years. Results of absorbed dose rate distributions for several relevant materials are presented and discussed in terms of the different contributions from the various activated components.

  5. Japanese perspective of fusion nuclear technology from ITER to DEMO

    Tanaka, Satoru; Takatsu, Hideyuki

    2007-01-01

    The world fusion community is now launching construction of ITER, the first nuclear-grade fusion machine in the world. In parallel to the ITER program, Broader Approach (BA) activities are to be initiated in this year by EU and Japan, mainly at Rokkasho BA site in Japan, as complementary activities to ITER toward DEMO. The BA activities include IFMIFEVEDA (International Fusion Materials Irradiation Facility-Engineering Validation and Engineering Design Activities) and DEMO design activities with generic technology R and Ds, both of which are critical to the rapid development of DEMO and commercial fusion power plants. The Atomic Energy Commission of Japan reviewed on-going third phase fusion program and issued the results of the review, 'On the policy of Nuclear Fusion Research and Development' in November 2005. In this report, it is anticipated that the ITER will be made operational in a decade and the programmatic objective can be met in the succeeding seven or eight years. Under this condition, the report presents a roadmap toward the DEMO and beyond and R and D items on fusion nuclear technology, indispensable for fusion energy utilization, are re-aligned. In the present paper, Japanese view and policy on ITER and beyond is summarized mainly from the viewpoints of nuclear fusion technology, and a minimum set of R and D elements on fusion nuclear technology, essential for fusion energy utilization, is presented. (orig.)

  6. Massively Clustered CubeSats NCPS Demo Mission

    Robertson, Glen A.; Young, David; Kim, Tony; Houts, Mike

    2013-01-01

    Technologies under development for the proposed Nuclear Cryogenic Propulsion Stage (NCPS) will require an un-crewed demonstration mission before they can be flight qualified over distances and time frames representative of a crewed Mars mission. In this paper, we describe a Massively Clustered CubeSats platform, possibly comprising hundreds of CubeSats, as the main payload of the NCPS demo mission. This platform would enable a mechanism for cost savings for the demo mission through shared support between NASA and other government agencies as well as leveraged commercial aerospace and academic community involvement. We believe a Massively Clustered CubeSats platform should be an obvious first choice for the NCPS demo mission when one considers that cost and risk of the payload can be spread across many CubeSat customers and that the NCPS demo mission can capitalize on using CubeSats developed by others for its own instrumentation needs. Moreover, a demo mission of the NCPS offers an unprecedented opportunity to invigorate the public on a global scale through direct individual participation coordinated through a web-based collaboration engine. The platform we describe would be capable of delivering CubeSats at various locations along a trajectory toward the primary mission destination, in this case Mars, permitting a variety of potential CubeSat-specific missions. Cameras on various CubeSats can also be used to provide multiple views of the space environment and the NCPS vehicle for video monitoring as well as allow the public to "ride along" as virtual passengers on the mission. This collaborative approach could even initiate a brand new Science, Technology, Engineering and Math (STEM) program for launching student developed CubeSat payloads beyond Low Earth Orbit (LEO) on future deep space technology qualification missions. Keywords: Nuclear Propulsion, NCPS, SLS, Mars, CubeSat.

  7. Scoping studies for NBI launch geometries on DEMO

    Jenkins, I., E-mail: ian.jenkins@ukaea.uk; Challis, C.D.; Keeling, D.L.; Surrey, E.

    2016-05-15

    Highlights: • NBCD scans are done for beam energies of 1.5 MeV and 1.0 MeV in two DEMO scenarios. • NBCD scan profiles are fed into genetic algorithm to fit a target current profile. • The result gives location and power of sources to give best fit to target profile. • This method can help provide requirements for DEMO beamline geometry. - Abstract: Engineering and technical constraints on Neutral Beam Injection (NBI) in DEMO may determine the available beam energy and may also strongly impact the Neutral Beam Current Drive (NBCD) efficiency by restricting available beam tangential radii. These latter are determined by factors such as the inter-TF coil spacing, as well as the degree of required shielding. In order to illustrate how these factors may affect the contribution of NBCD on DEMO operating scenarios, scans of NBI tangency radii and elevation on two possible DEMO scenarios have been performed with two beam energies, 1.5 MeV and 1.0 MeV, in order to determine the most favourable options for NBCD efficiency. In addition, a method using a genetic algorithm has been used to seek optimised solutions of NBI source locations and powers to attempt to synthesize a target total plasma driven-current profile. It is found that certain beam trajectories may be proscribed by limitations on shinethrough onto the vessel wall. This may affect the ability of NBCD to extend the duration of a pulse in a scenario where it must complement the induced plasma current. Operating at the lower beam energy reduces the restrictions due to shinethrough and is attractive for technical reasons as it will required less development, but in the scenarios examined here this results in a spatial broadening of the NBCD profile, which may make it more challenging to achieve desired total driven-current profiles.

  8. Development, simulation and testing of structural materials for DEMO

    Laesser, R.; Baluc, N.; Boutard, J.-L.; Diegele, E.; Gasparotto, M.; Riccardi, B.; Dudarev, S.; Moeslang, A.; Pippan, R.; Schaaf, B. van der

    2006-01-01

    In DEMO the structural and functional materials of the in-vessel components will be exposed to a very intense flux of fusion neutrons with energies up to 14 MeV creating displacement cascades and gaseous transmutation products. Point defects and transmutations will induce new microstructures leading to changes in mechanical and physical properties such as hardening, swelling, loss of fracture toughness and creep strength. The kinetics of microstructural evolution depends on time, temperature and defect production rates. The structural materials to be used in DEMO should have very special properties: high radiation resistance up to the dose of 100 dpa, low residual activation, high creep strength and good compatibility with the cooling media in as wide a temperature operational window as possible for the achievement of high thermal efficiency. The most promising materials are: Reduced Activation Ferritic Martensitic (RAFM) steels (Eurofer and F82H), Oxide Dispersion Strengthened (ODS) RAFM and RAF steels, SiC fibres reinforced SiC matrix composites (SiCf/SiC), tungsten (W) and W-alloys. Each of these materials has its advantages and drawbacks and will be best used under certain conditions. Presently the best studied group of materials are the RAFM steels. They require the smallest extrapolation for use in DEMO but also offer the lowest upper temperature limit of operation (550 o C) and thus the lowest thermal efficiency. The other materials foreseen for more advanced breeder blanket and divertor concepts require intense fundamental R(and)D and testing before their acceptance, whereas the so-called Test Blanket Modules (TBMs) will be constructed using RAFM steel and tested in ITER. Validation of the DEMO structural materials will be done in IFMIF, the International Fusion Materials Irradiation Facility, which will produce neutron damage and transmutation products very similar to those characterising a fusion device and will allow accelerated testing with damage rates

  9. Reel success creating demo reels and animation portfolios

    Cabrera, Cheryl

    2013-01-01

    Are you an animator looking to get your foot in the door to the top studios?It's tough if you don't have a demo reel and portfolio that reflects your unique style and incredible talents.  The reception of that reel will make or break you; so it's no wonder that creating a demo reel can be such a daunting task.  Reel Success by Cheryl Cabrera can help.  This book guides you into putting the right content into your portfolio, how to cater to the right audience, and how to harness the power of social media and network effectively.  Accompanied by case studies of actual students

  10. DEMO and fusion power plant conceptual studies in Europe

    Maisonnier, David; Cook, Iau; Pierre, Sardain; Lorenzo, Boccaccini; Luigi, Di Pace; Luciano, Giancarli; Prachai, Norajitra; Aldo, Pizzuto

    2006-01-01

    Within the European Power Plant Conceptual Study (PPCS) four fusion power plant 'models' have been developed. Two of these models were developed considering limited extrapolations both in physics and in technology. For the two other models, advanced physics scenarios have been identified and combined with advanced blanket concepts that allow higher thermodynamic efficiencies of the power conversion systems. For all the PPCS models, systems analyses were used to integrate the plasma physics and technology constraints to produce self-consistent plant parameter sets. The broad features of the conclusions of previous studies on safety, environmental impact and economics have been confirmed for the new models and demonstrated with increased confidence. The PPCS also helps in the definition of the objectives and in the identification of the design drivers of DEMO, i.e. the device between the next step (ITER) and a first-of-a-kind reactor. These will constitute the basis of the European DEMO Conceptual Study that has recently started

  11. RAMI analysis for DEMO HCPB blanket concept cooling system

    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

  12. LTS and HTS high current conductor development for DEMO

    Bruzzone, Pierluigi; Sedlak, Kamil; Uglietti, Davide; Bykovsky, Nikolay; Muzzi, Luigi; De Marzi, Gainluca; Celentano, Giuseppe; Della Corte, Antonio; Turtù, Simonetta; Seri, Massimo

    2015-01-01

    Highlights: • Design and R&D for DEMO TF conductors. • Wind&react vs. react&wind options for Nb_3Sn high grade TF conductors. • Progress in the manufacture of short length Nb_3Sn proptotypes. • Design and prototype manufacture for high current HTS cabled conductors. - Abstract: The large size of the magnets for DEMO calls for very large operating current in the forced flow conductor. A plain extrapolation from the superconductors in use for ITER is not adequate to fulfill the technical and cost requirements. The proposed DEMO TF magnets is a graded winding using both Nb_3Sn and NbTi conductors, with operating current of 82 kA @ 13.6 T peak field. Two Nb_3Sn prototypes are being built in 2014 reflecting the two approaches suggested by CRPP (react&wind method) and ENEA (wind&react method). The Nb_3Sn strand (overall 200 kg) has been procured at technical specification similar to ITER. Both the Nb_3Sn strand and the high RRR, Cr plated copper wire (400 kg) have been delivered. The cabling trials are carried out at TRATOS Cavi using equipment relevant for long length production. The completion of the manufacture of the two 20 m long prototypes is expected in the end of 2014 and their test is planned in 2015 at CRPP. In the scope of a long term technology development, high current HTS conductors are built at CRPP and ENEA. A DEMO-class prototype conductor is developed and assembled at CRPP: it is a flat cable composed of 20 twisted stacks of coated conductor tape soldered into copper shells. The 10 kA conductor developed at ENEA consists of stacks of coated conductor tape inserted into a slotted and twisted Al core, with a central cooling channel. Samples have been manufactured in industrial environment and the scalability of the process to long production lengths has been proven.

  13. Reliability and availability requirements analysis for DEMO: fuel cycle system

    Pinna, T.; Borgognoni, F.

    2015-01-01

    The Demonstration Power Plant (DEMO) will be a fusion reactor prototype designed to demonstrate the capability to produce electrical power in a commercially acceptable way. Two of the key elements of the engineering development of the DEMO reactor are the definitions of reliability and availability requirements (or targets). The availability target for a hypothesized Fuel Cycle has been analysed as a test case. The analysis has been done on the basis of the experience gained in operating existing tokamak fusion reactors and developing the ITER design. Plant Breakdown Structure (PBS) and Functional Breakdown Structure (FBS) related to the DEMO Fuel Cycle and correlations between PBS and FBS have been identified. At first, a set of availability targets has been allocated to the various systems on the basis of their operating, protection and safety functions. 75% and 85% of availability has been allocated to the operating functions of fuelling system and tritium plant respectively. 99% of availability has been allocated to the overall systems in executing their safety functions. The chances of the systems to achieve the allocated targets have then been investigated through a Failure Mode and Effect Analysis and Reliability Block Diagram analysis. The following results have been obtained: 1) the target of 75% for the operations of the fuelling system looks reasonable, while the target of 85% for the operations of the whole tritium plant should be reduced to 80%, even though all the tritium plant systems can individually reach quite high availability targets, over 90% - 95%; 2) all the DEMO Fuel Cycle systems can reach the target of 99% in accomplishing their safety functions. (authors)

  14. Conceptual design study of the K-DEMO magnet system

    Kim, Keeman, E-mail: kkeeman@nfri.re.kr [National Fusion Research Institute, 169-148 Gwahak-ro, Daejeon 305-806 (Korea, Republic of); Oh, Sangjun; Park, Jong Sung; Lee, Chulhee; Im, Kihak; Kim, Hyung Chan; Lee, Gyung-Su [National Fusion Research Institute, 169-148 Gwahak-ro, Daejeon 305-806 (Korea, Republic of); Neilson, George; Brown, Thomas; Kessel, Charles; Titus, Peter; Zhai, Yuhu [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States)

    2015-10-15

    Highlights: • Perform a preliminary conceptual study for a steady-state Korean DEMO reactor. • Present a preliminary design of TF (toroidal field) magnet. • Present a preliminary design of CS (central solenoid) magnet. • Present a preliminary design of PF (toroidal field) magnet. - Abstract: As the ITER is being constructed, there is a growing anticipation for an earlier realization of fusion energy. A major design philosophy for the initiated conceptual design study for a steady-state Korean fusion demonstration reactor (K-DEMO) is engineering feasibility. A two-staged development plan is envisaged. K-DEMO is designed not only to demonstrate a net electricity generation and a self-sustained tritium cycle, but also to be used, in its initial stage, as a component test facility. Then, in its second stage, a major upgrade is carried out by replacing in-vessel components in order to show a net electricity generation on the order of 500 MWe. After a thorough 0-D system analysis, the major radius and minor radius are chosen to be 6.8 m and 2.1 m, respectively. In order to minimize wave deflection, a top-launch high frequency (>200 GHz) electron cyclotron current drive (ECCD) system will be the key system for the current profile control. For matching the high frequency ECCD, a high toroidal field (TF) is required and can be achieved by using high current density Nb{sub 3}Sn superconducting conductor. The peak magnetic field reaches to 16 T with the magnetic field at the plasma center above 7 T. Key features of the K-DEMO magnet system include the use of two TF coil winding packs, each of a different conductor design, to reduce the construction cost and save the space for the magnet structure material.

  15. Conceptual design study of the K-DEMO magnet system

    Kim, Keeman; Oh, Sangjun; Park, Jong Sung; Lee, Chulhee; Im, Kihak; Kim, Hyung Chan; Lee, Gyung-Su; Neilson, George; Brown, Thomas; Kessel, Charles; Titus, Peter; Zhai, Yuhu

    2015-01-01

    Highlights: • Perform a preliminary conceptual study for a steady-state Korean DEMO reactor. • Present a preliminary design of TF (toroidal field) magnet. • Present a preliminary design of CS (central solenoid) magnet. • Present a preliminary design of PF (toroidal field) magnet. - Abstract: As the ITER is being constructed, there is a growing anticipation for an earlier realization of fusion energy. A major design philosophy for the initiated conceptual design study for a steady-state Korean fusion demonstration reactor (K-DEMO) is engineering feasibility. A two-staged development plan is envisaged. K-DEMO is designed not only to demonstrate a net electricity generation and a self-sustained tritium cycle, but also to be used, in its initial stage, as a component test facility. Then, in its second stage, a major upgrade is carried out by replacing in-vessel components in order to show a net electricity generation on the order of 500 MWe. After a thorough 0-D system analysis, the major radius and minor radius are chosen to be 6.8 m and 2.1 m, respectively. In order to minimize wave deflection, a top-launch high frequency (>200 GHz) electron cyclotron current drive (ECCD) system will be the key system for the current profile control. For matching the high frequency ECCD, a high toroidal field (TF) is required and can be achieved by using high current density Nb_3Sn superconducting conductor. The peak magnetic field reaches to 16 T with the magnetic field at the plasma center above 7 T. Key features of the K-DEMO magnet system include the use of two TF coil winding packs, each of a different conductor design, to reduce the construction cost and save the space for the magnet structure material.

  16. DEMO concepts and their roles within the fusion programme

    Tran, Minh Quang

    2007-01-01

    In the past years, the international fusion community has developed models of fusion power plants, which were extremely useful in showing the key advantages of fusion energy and pointing out he areas of development. The present view is that between ITER and such power plants (even of ''first of kind'' type), there is a need for one or two intermediate steps. The need to have a ''fast rack'' towards such a fusion reactor, suggested that the steps after ITER, which are usually considered to be a Demonstration power plant followed by a Prototypical one, could be combines into one known as a DEMO. DEMO would then be a device capable of producing electricity, paving the way towards fusion power plants which would be economically viable. This talk outlines the DEMO concepts as the necessary physics and technological extrapolation from the envisaged future steps (ITER, IFMIF) are discussed. It attempts to provide a coverage of the different concepts developed by various countries, The key issues, as foreseen today, and their implications for the programme are highlighted. (orig.)

  17. Objectives and status of EUROfusion DEMO blanket studies

    Boccaccini, L.V., E-mail: lorenzo.boccaccini@kit.edu [Karlsruhe Institute of Technology (KIT) (Germany); Aiello, G.; Aubert, J. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Bachmann, C. [EUROfusion, PPPT, Garching (Germany); Barrett, T. [CCFE, Abingdon OX14 3DB (United Kingdom); Del Nevo, A. [ENEA CR Brasimone, 40032 Camugnano, BO (Italy); Demange, D. [Karlsruhe Institute of Technology (KIT) (Germany); Forest, L. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Hernandez, F.; Norajitra, P. [Karlsruhe Institute of Technology (KIT) (Germany); Porempovic, G. [Fuziotech Engineering Ltd (Hungary); Rapisarda, D. [CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain); Sardain, P. [CEA/IRFM, 13115 Saint-Paul-lès-Durance (France); Utili, M. [ENEA CR Brasimone, 40032 Camugnano, BO (Italy); Vala, L. [Centrum výzkumu Řež, 250 68 Husinec-Řež (Czech Republic)

    2016-11-01

    Highlights: • Short description of the new Breeding Blanket Project in the EUROfusion consortium for the design of the EU PPPT DEMO: objectives. • Presentation of the design approach used in the development of the Breeding Blanket design: requirements. • Breeding Blanket design; in particular the four blanket concepts included in the study are presented, recent results highlighted and the status discussed. • Auxiliary systems and related R&D programme: in particular the work areas addressed in the Project (Tritium Technology, Pb-Li and Solid Breeders Technology, First Wall Design and R&D, Manufacturing) are presented, recent results highlighted and the status discussed. - Abstract: The design of a DEMO reactor requires the design of a blanket system suitable of reliable T production and heat extraction for electricity production. In the frame of the EUROfusion Consortium activities, the Breeding Blanket Project has been constituted in 2014 with the goal to develop concepts of Breeding Blankets for the EU PPPT DEMO; this includes an integrated design and R&D programme with the goal to select after 2020 concepts on fusion plants for the engineering phase. The design activities are presently focalized around a pool of solid and liquid breeder blanket with helium, water and PbLi cooling. Development of tritium extraction and control technology, as well manufacturing and development of solid and PbLi breeders are part of the programme.

  18. Model improvements for tritium transport in DEMO fuel cycle

    Santucci, Alessia, E-mail: alessia.santucci@enea.it [Unità Tecnica Fusione – ENEA C. R. Frascati, Via E. Fermi 45, 00044 Frascati (Roma) (Italy); Tosti, Silvano [Unità Tecnica Fusione – ENEA C. R. Frascati, Via E. Fermi 45, 00044 Frascati (Roma) (Italy); Franza, Fabrizio [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany)

    2015-10-15

    Highlights: • T inventory and permeation of DEMO blankets have been assessed under pulsed operation. • 1-D model for T transport has been developed for the HCLL DEMO blanket. • The 1-D model evaluated T partial pressure and T permeation rate radial profiles. - Abstract: DEMO operation requires a large amount of tritium, which is directly produced inside the reactor by means of Li-based breeders. During its production, recovering and purification, tritium comes in contact with large surfaces of hot metallic walls, therefore it can permeate through the blanket cooling structure, reach the steam generator and finally the environment. The development of dedicated simulation tools able to predict tritium losses and inventories is necessary to verify the accomplishment of the accepted tritium environmental releases as well as to guarantee a correct machine operation. In this work, the FUS-TPC code is improved by including the possibility to operate in pulsed regime: results in terms of tritium inventory and losses for three pulsed scenarios are shown. Moreover, the development of a 1-D model considering the radial profile of the tritium generation is described. By referring to the inboard segment on the equatorial axis of the helium-cooled lithium–lead (HCLL) blanket, preliminary results of the 1-D model are illustrated: tritium partial pressure in Li–Pb and tritium permeation in the cooling and stiffening plates by assuming several permeation reduction factor (PRF) values. Future improvements will consider the application of the model to all segments of different blanket concepts.

  19. Design study of ITER-like divertor target for DEMO

    Crescenzi, Fabio; Bachmann, C.; Richou, M.; Roccella, S.; Visca, E.; You, J.-H.

    2015-01-01

    Highlights: • ‘DEMO’ is a near-term Power Plant Conceptual Study (PPCS). • The ITER-like design concept represents a promising solution also for DEMO plasma facing units. • The optimization of PFUs aims to enhance the thermo-mechanical behaviour of the component. • The optimized geometry was evaluated by ITER SDC-IC criteria and in terms of low cycle fatigue (LCF). - Abstract: A near-term water-cooled target solution has to be evaluated together with the required technologies and its power exhaust limit under ‘DEMO’ conditions. The ITER-like design concept based on the mono-block technology using W as armour material and the CuCrZr-IG as structural material with an interlayer of pure copper represents a promising solution also for DEMO. This work reports the design study of an “optimized” ITER-like Water Cooled Divertor able to withstand a heat flux of 10 MW m"−"2, as requested for DEMO operating conditions. The optimization of plasma facing unit (PFU) aims to enhance the thermo-mechanical behaviour of the component by varying some geometrical parameters (monoblock size, interlayer thickness and, tube diameter and thickness). The optimization was performed by means of the multi-variable optimization algorithms using the FEM code ANSYS. The coolant hydraulic conditions (inlet pressure, temperature and velocity) were fixed for simplicity. This study is based on elastic analysis and 3 dimensional modelling. The resulting optimized geometry was evaluated on the basis of the ITER SDC-IC criteria and in terms of low cycle fatigue (LCF). The margin to the critical heat flux (CHF) was also estimated. Further design study (taking into account the effect of neutron radiation on the material properties) together with mock-up fabrication and high-heat-flux (HHF) tests are foreseen in next work programmes.

  20. Design study of ITER-like divertor target for DEMO

    Crescenzi, Fabio, E-mail: fabio.crescenzi@enea.it [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); Bachmann, C. [EFDA, Power Plant Physics and Technology, Boltzmannstraße 2, 85748 Garching (Germany); Richou, M. [CEA, IRFM, F-13108 Saint Paul Lez Durance (France); Roccella, S.; Visca, E. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); You, J.-H. [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany)

    2015-10-15

    Highlights: • ‘DEMO’ is a near-term Power Plant Conceptual Study (PPCS). • The ITER-like design concept represents a promising solution also for DEMO plasma facing units. • The optimization of PFUs aims to enhance the thermo-mechanical behaviour of the component. • The optimized geometry was evaluated by ITER SDC-IC criteria and in terms of low cycle fatigue (LCF). - Abstract: A near-term water-cooled target solution has to be evaluated together with the required technologies and its power exhaust limit under ‘DEMO’ conditions. The ITER-like design concept based on the mono-block technology using W as armour material and the CuCrZr-IG as structural material with an interlayer of pure copper represents a promising solution also for DEMO. This work reports the design study of an “optimized” ITER-like Water Cooled Divertor able to withstand a heat flux of 10 MW m{sup −2}, as requested for DEMO operating conditions. The optimization of plasma facing unit (PFU) aims to enhance the thermo-mechanical behaviour of the component by varying some geometrical parameters (monoblock size, interlayer thickness and, tube diameter and thickness). The optimization was performed by means of the multi-variable optimization algorithms using the FEM code ANSYS. The coolant hydraulic conditions (inlet pressure, temperature and velocity) were fixed for simplicity. This study is based on elastic analysis and 3 dimensional modelling. The resulting optimized geometry was evaluated on the basis of the ITER SDC-IC criteria and in terms of low cycle fatigue (LCF). The margin to the critical heat flux (CHF) was also estimated. Further design study (taking into account the effect of neutron radiation on the material properties) together with mock-up fabrication and high-heat-flux (HHF) tests are foreseen in next work programmes.

  1. Progress on DEMO blanket attachment concept with keys and pins

    Vizvary, Zsolt; Iglesias, Daniel; Cooper, David; Crowe, Robert; Riccardo, Valeria

    2015-01-01

    Highlights: • DEMO blanket attachment system with keys and pins (without using bolts). • Blanket segments are preloaded by progressively designed springs. • Blanket back plate flexibility has a major impact on spring design. • Mechanical analysis of other components indicates no unresolvable issues. • Thermal analysis indicates acceptable temperatures for the support system. - Abstract: The blanket attachment has to cope with gravity, thermal and electromagnetic loads, also it has to be installed and serviced by remote handling. Pre-stressed components suffer from stress relaxation in irradiated environments such as DEMO. To circumvent this problem pre-stressed component should be either avoided or shielded, and where possible keys and pins should be used. This strategy has been proposed for the DEMO multi-module segments (MMS). The blanket segments are held by two tapered keys each, designed to allow thermal expansions while providing contact with the vacuum vessel and to resist the poloidal and radial moments the latter being dominant at 9.1 MNm inboard and 15 MNm outboard. On the top of the blanket segment there is a pin which provides vertical support. At the bottom another vertical support has to lock them in position after installation and manage the pre-load on the segments. The pre-load is required to deal with the electromagnetic loads during disruption. This is provided by a set of springs, which require shielding as they are preloaded. These are sized to cope with the force (3 MN inboard, 1.4 MN outboard) due to halo currents and the toroidal moment which can reverse. Calculations show that the flexibility of the blanket segment itself plays a significant role in defining the required support system. The blanket segment acts as a preloaded spring and it has to be part of the attachment design as well.

  2. An exploratory study on the gaps and pathways to the Korean fusion DEMO

    Kim, Hyuck Jong; Heo, Gyunyoung; Kim, Hyung Chan; Yeom, Jun Ho; Kim, Jong Kyung; Lee, Young-seok; Kwon, Myeun; Lee, Gyung-Su; Kim, Yong-soo; Kim, Eunbae; Lee, Chul-sik

    2012-01-01

    With the vision of being an early demonstrator of fusion energy, the strategic plans for the Fusion DEMO program of Korea (K-DEMO program) has been developed. A staged development of the K-DEMO plant was considered in the strategic plans as to verify technical feasibility in the first stage and economic feasibility in the second stage. The top-tier design requirements and assumptions of the first stage K-DEMO plant are defined and postulated. With these requirements and assumptions, the desired and current status of nuclear fusion technologies are compared to identify the gaps to be filled to design, fabricate, construct, and operate it. The pathways from KSTAR, ITER to K-DEMO plant have also been studied to identify R and D activities for K-DEMO program that are to go in parallel with KSTAR and ITER are extracted from the pathways. Cross-cutting with the fusion R and D activities of the other countries and utilizing the commonalities with the existing systems are discussed with the provision of open-innovation strategy that is one of the key strategies of K-DEMO program. The priority of the R and D activities of K-DEMO program is qualitatively determined in consideration of the gaps, cross-cutting, and risks associated with the R and D investments.

  3. The Role of Community Colleges in Advancing Upward Mobility: A Demos Perspective

    Huelsman, Mark

    2015-01-01

    This article provides a short background on Demos, a public policy organization that works on issues of political and economic inequality. Demos views community colleges as a linchpin in the American higher education system, and it has worked over several years to research ways to increase state support for higher education and direct support…

  4. Demos as an Explanatory Lens in Teacher Educators' Elusive Search for Social Justice

    Oikonomidoy, Eleni M.; Brock, Cynthia H.; Obenchain, Kathryn M.; Pennington, Julie L.

    2013-01-01

    Borrowing insights from the Ancient Greek ideal conceptions of a democratic civic space (demos), this article examines the applicability of this framework to four teacher educators' journey to implement social justice in their programs. It is proposed that the three constitutive dimensions of demos (freedom of speech, equality to vote and hold…

  5. Overview of EU activities on DEMO liquid metal breeder blanket

    Giancarli, L.; Proust, E.; Malang, S.; Reimann, J.; Perujo, A.

    1994-01-01

    The present paper gives an overview of both design and experimental activities within the European Union (EU) concerning the development of liquid metal breeder blankets for DEMO. After several years of studies on breeding blankets, two blanket concepts are presently considered, both using the eutectic Pb-17Li: the dual-coolant concept and the water-cooled concept. The analysis of such concepts has permitted to identify the experimental areas where further data are required. Tritium control and MHD-issues are, at present, the activities on which is devoted the greatest effort within the EU. (authors). 4 figs., 4 tabs., 39 refs

  6. Constitutional Crowdsourcing to Reconcile Demos with Aristos and Nomos

    Abat Ninet, Antoni

    2017-01-01

    it is framed, been liberal democracies or authoritarian states. Derrida stated there is a sort of “semantic indeterminacy” at the core of democracy and that constitutional crowdsourcing is a way to intervene in this indeterminacy. The Icelandic example enlightened that there is a way to mediate between....... The final segment of the paper aims to obtain different elements to improve the constitutional crowdsourcing to be considered in future constituent processes around the world. From a formal perspective the paper simulates a judgment between a Plaintiff Demos (representing “We the People” the entitled...

  7. Compact NMR

    Bluemich, Bernhard; Haber-Pohlmeier, Sabina; Zia, Wasif [RWTH Aachen Univ. (Germany). Inst. fuer Technische und Makromolekulare Chemie (ITMC)

    2014-06-01

    Nuclear Magnetic Resonance (NMR) spectroscopy is the most popular method for chemists to analyze molecular structures, while Magnetic Resonance Imaging (MRI) is a non-invasive diagnostic tool for medical doctors that provides high-contrast images of biological tissue. In both applications, the sample (or patient) is positioned inside a large, superconducting magnet to magnetize the atomic nuclei. Interrogating radio-frequency pulses result in frequency spectra that provide the chemist with molecular information, the medical doctor with anatomic images, and materials scientist with NMR relaxation parameters. Recent advances in magnet technology have led to a variety of small permanent magnets to allow compact and low-cost instruments. The goal of this book is to provide an introduction to the practical use of compact NMR at a level nearly as basic as the operation of a smart phone.

  8. Compact vortices

    Bazeia, D.; Losano, L.; Marques, M.A.; Zafalan, I. [Universidade Federal da Paraiba, Departamento de Fisica, Joao Pessoa, PB (Brazil); Menezes, R. [Universidade Federal da Paraiba, Departamento de Ciencias Exatas, Rio Tinto, PB (Brazil); Universidade Federal de Campina Grande, Departamento de Fisica, Campina Grande, PB (Brazil)

    2017-02-15

    We study a family of Maxwell-Higgs models, described by the inclusion of a function of the scalar field that represent generalized magnetic permeability. We search for vortex configurations which obey first-order differential equations that solve the equations of motion. We first deal with the asymptotic behavior of the field configurations, and then implement a numerical study of the solutions, the energy density and the magnetic field. We work with the generalized permeability having distinct profiles, giving rise to new models, and we investigate how the vortices behave, compared with the solutions of the corresponding standard models. In particular, we show how to build compact vortices, that is, vortex solutions with the energy density and magnetic field vanishing outside a compact region of the plane. (orig.)

  9. Časopis Demos (Internationale Ethnographische und Folkloristische Informationen) v novém tisíciletí

    Woitsch, Jiří

    2003-01-01

    Roč. 6, - (2003), s. 74-78 ISSN 1210-1109 Institutional research plan: CEZ:AV0Z9058907 Keywords : journal Demos * history of the journal Demos * new developments in publishing of the journal Demos Subject RIV: AC - Archeology, Anthropology, Ethnology

  10. Neutronic performance optimization study of Indian fusion demo reactor first wall and breeding blanket

    Swami, H.L.; Danani, C.

    2015-01-01

    In frame of design studies of Indian Nuclear Fusion DEMO Reactor, neutronic performance optimization of first wall and breeding blanket are carried out. The study mainly focuses on tritium breeding ratio (TBR) and power density responses estimation of breeding blanket. Apart from neutronic efficiency of existing breeding blanket concepts for Indian DEMO i.e. lead lithium ceramic breeder and helium cooled solid breeder concept other concepts like helium cooled lead lithium and helium-cooled Li_8PbO_6 with reflector are also explored. The aim of study is to establish a neutronically efficient breeding blanket concept for DEMO. Effect of first wall materials and thickness on breeding blanket neutronic performance is also evaluated. For this study 1 D cylindrical neutronic model of DEMO has been constructed according to the preliminary radial build up of Indian DEMO. The assessment is being done using Monte Carlo based radiation transport code and nuclear cross section data file ENDF/B- VII. (author)

  11. Compact stars

    Estevez-Delgado, Gabino; Estevez-Delgado, Joaquin

    2018-05-01

    An analysis and construction is presented for a stellar model characterized by two parameters (w, n) associated with the compactness ratio and anisotropy, respectively. The reliability range for the parameter w ≤ 1.97981225149 corresponds with a compactness ratio u ≤ 0.2644959374, the density and pressures are positive, regular and monotonic decrescent functions, the radial and tangential speed of sound are lower than the light speed, moreover, than the plausible stability. The behavior of the speeds of sound are determinate for the anisotropy parameter n, admitting a subinterval where the speeds are monotonic crescent functions and other where we have monotonic decrescent functions for the same speeds, both cases describing a compact object that is also potentially stable. In the bigger value for the observational mass M = 2.05 M⊙ and radii R = 12.957 Km for the star PSR J0348+0432, the model indicates that the maximum central density ρc = 1.283820319 × 1018 Kg/m3 corresponds to the maximum value of the anisotropy parameter and the radial and tangential speed of the sound are monotonic decrescent functions.

  12. ITER, the 'Broader Approach', a DEMO fusion reactor

    Janeschitz, G.; Bahm, W.

    2007-01-01

    Fusion is a very promising future energy option, which is characterized by almost unlimited fuel reserves, favourable safety features and environmental sustainability. The aim of the worldwide fusion research is a fusion power station which imitates the process taking place in the sun and thus gains energy from the fusion of light atomic nuclei. The experimental reactor ITER which will be built in Cadarache, France, marks a breakthrough in the worldwide fusion research: For the first time an energy multiplication factor of at least 10 will be achieved, the factor by which the fusion power exceeds the external plasma heating. Partners in this project are the European Union, Japan, the Russian Federation, USA, China, South Korea and India as well as Brazil as associated partner. The facility is supposed to demonstrate a long burning, reactor-typical plasma and to test techniques such as plasma heating, plasma confinement by superconducting magnets, fuel cycle as well as energy transition, tritium breeding and remote handling technologies. The next step beyond ITER will be the demonstration power station DEMO which requires further developments in order to create the basis for its design and construction. The roadmap to fusion energy is described. It consists of several elements which are needed to develop the knowledge required for a commercial fusion reactor. The DEMO time schedule depends on the efforts in terms of personnel and budget resources the society is willing to invest in fusion taking into account the long term energy supply and its environmental impact. (orig.)

  13. An FPGA computing demo core for space charge simulation

    Wu, Jinyuan; Huang, Yifei

    2009-01-01

    In accelerator physics, space charge simulation requires large amount of computing power. In a particle system, each calculation requires time/resource consuming operations such as multiplications, divisions, and square roots. Because of the flexibility of field programmable gate arrays (FPGAs), we implemented this task with efficient use of the available computing resources and completely eliminated non-calculating operations that are indispensable in regular micro-processors (e.g. instruction fetch, instruction decoding, etc.). We designed and tested a 16-bit demo core for computing Coulomb's force in an Altera Cyclone II FPGA device. To save resources, the inverse square-root cube operation in our design is computed using a memory look-up table addressed with nine to ten most significant non-zero bits. At 200 MHz internal clock, our demo core reaches a throughput of 200 M pairs/s/core, faster than a typical 2 GHz micro-processor by about a factor of 10. Temperature and power consumption of FPGAs were also lower than those of micro-processors. Fast and convenient, FPGAs can serve as alternatives to time-consuming micro-processors for space charge simulation.

  14. Enhancing the DEMO divertor target by interlayer engineering

    Barrett, T.R.; McIntosh, S.C.; Fursdon, M.; Hancock, D.; Timmis, W.; Coleman, M.; Rieth, M.; Reiser, J.

    2015-01-01

    Highlights: • The European ‘near-term’ DEMO forsees a water-cooled divertor. • Divertor targets typically use an interlayer between the armour and structure. • Engineering the properties of the interlayer can yield large gains in performance. • A response surface based design search and optimisation method is used. • A new design passes linear-elastic code rules up to applied heat flux of 18 MW/m"2. - Abstract: A robust water-cooled divertor target plate solution for DEMO has to date remained elusive. Common to all contemporary concepts is an interlayer at the boundary between the tungsten armour and the cooling structure. In this paper we show by design optimisation that an effectively designed interlayer can produce dramatic gains in power handling. By engineering the interlayer as part of the design study, it is found that divertor performance is enhanced by either a low conductivity ‘Thermal Break’ interlayer or an ‘Ultra-Compliant’ interlayer. For a 10 MW/m"2 surface heat flux we find that a thermal conductivity of 15 W/mK and elastic modulus of 1 GPa are effective. A design is proposed which passes linear-elastic code rules up to an applied heat flux of 18 MW/m"2.

  15. Enhancing the DEMO divertor target by interlayer engineering

    Barrett, T.R., E-mail: tom.barrett@ccfe.ac.uk [CCFE, Culham Science Centre, Oxfordshire OX14 3DB (United Kingdom); McIntosh, S.C.; Fursdon, M.; Hancock, D.; Timmis, W.; Coleman, M. [CCFE, Culham Science Centre, Oxfordshire OX14 3DB (United Kingdom); Rieth, M.; Reiser, J. [Karlsruhe Institute for Technology, IMF-I, D-7602 Karlsruhe (Germany)

    2015-10-15

    Highlights: • The European ‘near-term’ DEMO forsees a water-cooled divertor. • Divertor targets typically use an interlayer between the armour and structure. • Engineering the properties of the interlayer can yield large gains in performance. • A response surface based design search and optimisation method is used. • A new design passes linear-elastic code rules up to applied heat flux of 18 MW/m{sup 2}. - Abstract: A robust water-cooled divertor target plate solution for DEMO has to date remained elusive. Common to all contemporary concepts is an interlayer at the boundary between the tungsten armour and the cooling structure. In this paper we show by design optimisation that an effectively designed interlayer can produce dramatic gains in power handling. By engineering the interlayer as part of the design study, it is found that divertor performance is enhanced by either a low conductivity ‘Thermal Break’ interlayer or an ‘Ultra-Compliant’ interlayer. For a 10 MW/m{sup 2} surface heat flux we find that a thermal conductivity of 15 W/mK and elastic modulus of 1 GPa are effective. A design is proposed which passes linear-elastic code rules up to an applied heat flux of 18 MW/m{sup 2}.

  16. An assessment for the erosion rate of DEMO first wall

    Tokar, M. Z.

    2018-01-01

    In a fusion reactor a significant fraction of plasma particles lost from the confined volume will reach the vessel wall. The recombination of these charged species, electrons and ions of hydrogen isotopes, is a source of neutral molecules and atoms, recycling back into the plasma. Here they participate, in particular, in charge-exchange (c-x) collisions with the plasma ions and, as a result, atoms of high energies with chaotically oriented velocities are generated. A significant fraction of these hot neutrals will hit the wall, leading, as well as the outflowing fuel and impurity ions, to its erosion, limiting the reactor operation time. The rate of the wall erosion in DEMO is assessed by applying a one-dimensional model which takes into account the transport of charged and neutral species across the flux surfaces in the main part of the scrape-off layer, beyond the X-point vicinity and divertor, and by considering the shift of the centers of flux surfaces, their elongation and triangularity. Atoms generated by c-x of recycling neutrals are modeled kinetically to define firmly their energy spectrum, being of particular importance for the erosion assessment. It is demonstrated the erosion rate of the DEMO wall armor of tungsten will have a pronounced ballooning character with a significant maximum of 0.3 mm per full power year at the low field side, decreasing with an increase in the anomalous perpendicular transport in the ‘far’ SOL or the plasma density at the separatrix.

  17. An FPGA computing demo core for space charge simulation

    Wu, Jinyuan; Huang, Yifei; /Fermilab

    2009-01-01

    In accelerator physics, space charge simulation requires large amount of computing power. In a particle system, each calculation requires time/resource consuming operations such as multiplications, divisions, and square roots. Because of the flexibility of field programmable gate arrays (FPGAs), we implemented this task with efficient use of the available computing resources and completely eliminated non-calculating operations that are indispensable in regular micro-processors (e.g. instruction fetch, instruction decoding, etc.). We designed and tested a 16-bit demo core for computing Coulomb's force in an Altera Cyclone II FPGA device. To save resources, the inverse square-root cube operation in our design is computed using a memory look-up table addressed with nine to ten most significant non-zero bits. At 200 MHz internal clock, our demo core reaches a throughput of 200 M pairs/s/core, faster than a typical 2 GHz micro-processor by about a factor of 10. Temperature and power consumption of FPGAs were also lower than those of micro-processors. Fast and convenient, FPGAs can serve as alternatives to time-consuming micro-processors for space charge simulation.

  18. Neutronic analyses and tools development efforts in the European DEMO programme

    Fischer, U., E-mail: ulrich.fischer@kit.edu [Association KIT-Euratom, Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Bachmann, C. [European Fusion Development Agreement (EFDA), Garching (Germany); Bienkowska, B. [Association IPPLM-Euratom, IPPLM Warsaw/INP Krakow (Poland); Catalan, J.P. [Universidad Nacional de Educación a Distancia (UNED), Madrid (Spain); Drozdowicz, K.; Dworak, D. [Association IPPLM-Euratom, IPPLM Warsaw/INP Krakow (Poland); Leichtle, D. [Association KIT-Euratom, Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Fusion for Energy (F4E), Barcelona (Spain); Lengar, I. [MESCS-JSI, Ljubljana (Slovenia); Jaboulay, J.-C. [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Lu, L. [Association KIT-Euratom, Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Moro, F. [Associazione ENEA-Euratom, ENEA Fusion Division, Frascati (Italy); Mota, F. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Sanz, J. [Universidad Nacional de Educación a Distancia (UNED), Madrid (Spain); Szieberth, M. [Budapest University of Technology and Economics (BME), Budapest (Hungary); Palermo, I. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Pampin, R. [Fusion for Energy (F4E), Barcelona (Spain); Porton, M. [Euratom/CCFE Fusion Association, Culham Science Centre for Fusion Energy (CCFE), Culham (United Kingdom); Pereslavtsev, P. [Association KIT-Euratom, Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Ogando, F. [Universidad Nacional de Educación a Distancia (UNED), Madrid (Spain); Rovni, I. [Budapest University of Technology and Economics (BME), Budapest (Hungary); and others

    2014-10-15

    Highlights: •Evaluation of neutronic tools for application to DEMO nuclear analyses. •Generation of a DEMO model for nuclear analyses based on MC calculations. •Nuclear analyses of the DEMO reactor equipped with a HCLL-type blanket. -- Abstract: The European Fusion Development Agreement (EFDA) recently launched a programme on Power Plant Physics and Technology (PPPT) with the aim to develop a conceptual design of a fusion demonstration reactor (DEMO) addressing key technology and physics issues. A dedicated part of the PPPT programme is devoted to the neutronics which, among others, has to define and verify requirements and boundary conditions for the DEMO systems. The quality of the provided data depends on the capabilities and the reliability of the computational tools. Accordingly, the PPPT activities in the area of neutronics include both DEMO nuclear analyses and development efforts on neutronic tools including their verification and validation. This paper reports on first neutronics studies performed for DEMO, and on the evaluation and further development of neutronic tools.

  19. Impurity accumulation and performance of ITER and DEMO plasmas in the presence of transport barriers

    Chatthong, B; Promping, J; Onjun, T

    2017-01-01

    In this work, the impurity accumulations and their performance in the presence of both ITB and ETB in ITER and DEMO plasmas are investigated using a BALDUR integrated predictive modelling code. In these simulations, a combination of a neoclassical transport model NCLASS and an anomalous transport model Mixed Bohm/gyro-Bohm is used. The boundary condition is described at the top of the pedestal, which is calculated theoretically based on a combination of magnetic and flow shear stabilization pedestal width scaling and an infinite-n ballooning pressure gradient model. The toroidal flow is calculated based on the NTV (neoclassical toroidal viscosity) toroidal velocity model. The time evolution of plasma temperature and density profiles of ITER and DEMO (Korean K-DEMO and Japanese DEMO models A, B and C) plasmas are simulated in H -mode scenario with and without ITB formation. It is found that Japanese DEMO model C yields highest plasma temperature; while Korean DEMO yields the best plasma performance among those designs considered. Impurity accumulation is found to be highest in Japanese DEMO model B. (paper)

  20. Neutronic analyses and tools development efforts in the European DEMO programme

    Fischer, U.; Bachmann, C.; Bienkowska, B.; Catalan, J.P.; Drozdowicz, K.; Dworak, D.; Leichtle, D.; Lengar, I.; Jaboulay, J.-C.; Lu, L.; Moro, F.; Mota, F.; Sanz, J.; Szieberth, M.; Palermo, I.; Pampin, R.; Porton, M.; Pereslavtsev, P.; Ogando, F.; Rovni, I.

    2014-01-01

    Highlights: •Evaluation of neutronic tools for application to DEMO nuclear analyses. •Generation of a DEMO model for nuclear analyses based on MC calculations. •Nuclear analyses of the DEMO reactor equipped with a HCLL-type blanket. -- Abstract: The European Fusion Development Agreement (EFDA) recently launched a programme on Power Plant Physics and Technology (PPPT) with the aim to develop a conceptual design of a fusion demonstration reactor (DEMO) addressing key technology and physics issues. A dedicated part of the PPPT programme is devoted to the neutronics which, among others, has to define and verify requirements and boundary conditions for the DEMO systems. The quality of the provided data depends on the capabilities and the reliability of the computational tools. Accordingly, the PPPT activities in the area of neutronics include both DEMO nuclear analyses and development efforts on neutronic tools including their verification and validation. This paper reports on first neutronics studies performed for DEMO, and on the evaluation and further development of neutronic tools

  1. Thermal Hydraulic Analysis of K-DEMO Single Blanket Module for Preliminary Accident Analysis using MELCOR

    Moon, Sung Bo; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

    2016-05-15

    To develop the Korean fusion commercial reactor, preliminary design concept for K-DEMO (Korean fusion demonstration reactor) has been announced by NFRI (National Fusion Research Institute). This pre-conceptual study of K-DEMO has been introduced to identify technical details of a fusion power plant for the future commercialization of fusion reactor in Korea. Before this consideration, to build the K-DEMO, accident analysis is essential. Since the Fukushima accident, which is severe accident from unexpected disaster, safety analysis of nuclear power plant has become important. The safety analysis of both fission and fusion reactors is deemed crucial in demonstrating the low radiological effect of these reactors on the environment, during severe accidents. A risk analysis of K-DEMO should be performed, as a prerequisite for the construction of a fusion reactor. In this research, thermal-hydraulic analysis of single blanket module of K-DEMO is conducted for preliminary accident analysis for K-DEMO. Further study about effect of flow distributer is conducted. The normal K-DEMO operation condition is applied to the boundary condition and simulated to verify the material temperature limit using MELCOR. MELCOR is fully integrated, relatively fast-running code developed by Sandia National Laboratories. MELCOR had been used for Light Water Reactors and fusion reactor version of MELCOR was developed for ITER accident analysis. This study shows the result of thermal-hydraulic simulation of single blanket module with MELCOR which is severe accident code for nuclear fusion safety analysis. The difference of mass flow rate for each coolant channel with or without flow distributer is presented. With flow distributer, advantage of broadening temperature gradient in the K-DEMO blanket module and increase mass flow toward first wall is obtained. This can enhance the safety of K-DEMO blanket module. Most 13 .deg. C temperature difference in blanket module is obtained.

  2. Fuel cycle design for ITER and its extrapolation to DEMO

    Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Kyoto 611-0011 (Japan)], E-mail: s-konishi@iae.kyoto-u.ac.jp; Glugla, Manfred [Forschungszentrum Karlsruhe, P.O. Box 3640, D 76021 Karlsruhe (Germany); Hayashi, Takumi [Apan Atomic Energy AgencyTokai, Ibaraki 319-0015 Japan (Japan)

    2008-12-15

    ITER is the first fusion device that continuously processes DT plasma exhaust and supplies recycled fuel in a closed loop. All the tritium and deuterium in the exhaust are recovered, purified and returned to the tokamak with minimal delay, so that extended burn can be sustained with limited inventory. To maintain the safety of the entire facility, plant scale detritiation systems will also continuously run to remove tritium from the effluents at the maximum efficiency. In this entire tritium plant system, extremely high decontamination factor, that is the ratio of the tritium loss to the processing flow rate, is required for fuel economy and minimized tritium emissions, and the system design based on the state-of-the-art technology is expected to satisfy all the requirements without significant technical challenges. Considerable part of the fusion tritium system will be verified with ITER and its decades of operation experiences. Toward the DEMO plant that will actually generate energy and operate its closed fuel cycle, breeding blanket and power train that caries high temperature and pressure media from the fusion device to the generation system will be the major addition. For the tritium confinement, safety and environmental emission, particularly blanket, its coolant, and generation systems such as heat exchanger, steam generator and turbine will be the critical systems, because the tritium permeation from the breeder and handling large amount of high temperature, high pressure coolant will be further more difficult than that required for ITER. Detritiation of solid waste such as used blanket and divertor will be another issue for both tritium economy and safety. Unlike in the case of ITER that is regarded as experimental facility, DEMO will be expected to demonstrate the safety, reliability and social acceptance issue, even if economical feature is excluded. Fuel and environmental issue to be tested in the DEMO will determine the viability of the fusion as a

  3. Fuel cycle design for ITER and its extrapolation to DEMO

    Konishi, Satoshi; Glugla, Manfred; Hayashi, Takumi

    2008-01-01

    ITER is the first fusion device that continuously processes DT plasma exhaust and supplies recycled fuel in a closed loop. All the tritium and deuterium in the exhaust are recovered, purified and returned to the tokamak with minimal delay, so that extended burn can be sustained with limited inventory. To maintain the safety of the entire facility, plant scale detritiation systems will also continuously run to remove tritium from the effluents at the maximum efficiency. In this entire tritium plant system, extremely high decontamination factor, that is the ratio of the tritium loss to the processing flow rate, is required for fuel economy and minimized tritium emissions, and the system design based on the state-of-the-art technology is expected to satisfy all the requirements without significant technical challenges. Considerable part of the fusion tritium system will be verified with ITER and its decades of operation experiences. Toward the DEMO plant that will actually generate energy and operate its closed fuel cycle, breeding blanket and power train that caries high temperature and pressure media from the fusion device to the generation system will be the major addition. For the tritium confinement, safety and environmental emission, particularly blanket, its coolant, and generation systems such as heat exchanger, steam generator and turbine will be the critical systems, because the tritium permeation from the breeder and handling large amount of high temperature, high pressure coolant will be further more difficult than that required for ITER. Detritiation of solid waste such as used blanket and divertor will be another issue for both tritium economy and safety. Unlike in the case of ITER that is regarded as experimental facility, DEMO will be expected to demonstrate the safety, reliability and social acceptance issue, even if economical feature is excluded. Fuel and environmental issue to be tested in the DEMO will determine the viability of the fusion as a

  4. Automatic demand response referred to electricity spot price. Demo description

    Grande, Ove S.; Livik, Klaus; Hals, Arne

    2006-05-01

    This report presents background, technical solution and results from a test project (Demo I) developed in the DRR Norway) project. Software and technology from two different vendors, APAS and Powel ASA, are used to demonstrate a scheme for Automatic Demand Response (ADR) referred to spot price level and a system for documentation of demand response and cost savings. Periods with shortage of energy supply and hardly any investments in new production capacity have turned focus towards the need for increased price elasticity on the demand side in the Nordic power market. The new technology for Automatic Meter Reading (AMR) and Remote Load Control (RLC) provides an opportunity to improve the direct market participation from the demand side by introducing automatic schemes that reduce the need for customer attention to hourly market prices. The low prioritized appliances, and not the total load, are in this report defined as the Demand Response Objects, based on the assumption that there is a limit for what the customers are willing to pay for different uses of electricity. Only disconnection of residential water heaters is included in the demo, due to practical limitations. The test was performed for a group of single family houses over a period of 2 months. All the houses were equipped with a radio controlled 'Ebox' unit attached to the water heater socket. The settlement and invoicing were based on hourly metered values (kWh/h), which means that the customer benefit is equivalent to the accumulated changes in the electricity cost per hour. The actual load reduction is documented by comparison between the real meter values for the period and a reference curve. The curves show significant response to the activated control in the morning hours. In the afternoon it is more difficult to register the response, probably due to 'disturbing' activities like cooking etc. Demo I shows that load reduction referred to spot price level can be done in a smooth way. The experiences

  5. Design study of fusion Demo plant at JAERI

    Tobita, K.; Nishio, S.; Enoeda, M.

    2006-01-01

    Three options of fusion Demo plant are proposed characterized by functions of the center solenoid (Cs). The prime option uses a downsized CS, which does not provide sufficient V-s for plasma current ramp-up but supplies enough coil current for plasma shaping. This option produces a fusion output of 3 GW with a major radius of 5.5 m, aspect ratio of 2.6, normalized beta of 4.3 and maximum field of 16.4 T. The estimated reactor weight is lighter than that of other conventional tokamak reactors, suggesting an economic advantage. The plant uses rather conservative technologies such as Nb 3 Al superconductor, water-cooled solid breeder blanket, low activation ferritic steel as the structural material and tungsten monoblock divertor plate. The design philosophy and key issues related to the constituent technologies of the plant are described in the present paper

  6. Facebook pages as ’demo versions’ of issue publics

    Birkbak, Andreas

    ’political muscle’ through numbers. Second, these protests also focused on demonstrating harmful indirect consequences of a future payment ring by sharing news stories and other analyses that served to undermine the soundness of the payment ring. Third, these two kinds of demonstrations functioned as ’demoes...... of representative democracy are founded with a distinction between direct and indirect consequences of action (Dewey 1927), Facebook can be understood as an experimental issue public-generating device. In the payment ring controversy, several Facebook pages became spaces of ’demonstration’ in three senses...... is at stake in Facebook practices like these, then, it becomes useful to rethink publics as processes of on-going experimental inquiry into issues (Marres 2007)....

  7. Neutral beam deployment on DEMO and its influence on design

    Surrey, Elizabeth, E-mail: elizabeth.surrey@ccfe.ac.uk [EURATOM/CCFE, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); King, Damian; Lister, Jonathan; Porton, Michael; Timmis, William; Ward, David [EURATOM/CCFE, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom)

    2011-10-15

    The demands on the neutral beam heating and current drive system of a DEMO device exceed those of existing fusion experiments by several orders of magnitude. By predicting possible power waveforms it is possible to analyse the technological advances necessary to achieve a system relevant to deployment on a power plant. Achieving the necessary efficiency will require simultaneous improvements in beam current density, neutralization efficiency and beam transmission. Considering the deployment on the tokamak vessel shows no major disruption to the tritium breeder blanket and no requirement to reach a high packing density of injectors. The thermal management of components subjected to low heat flux for many hours is considered and it is shown that radiation cooling can be exploited to control the temperature of such items.

  8. Optimization and limitations of known DEMO divertor concepts

    Reiser, Jens; Rieth, Michael

    2012-01-01

    Highlights: ► Limitations of the materials. ► Improved H 2 O cooled divertor. ► Improved He cooled divertor. - Abstract: In this work we will introduce and discuss improvements for two types of DEMO divertors based on known designs: (i) gas cooled designs and (ii) liquid coolant concepts. In a first step, the advantages and disadvantages of gas cooling as well as the necessity of a jet impingement to increase the heat transfer coefficients will be discussed. Further discussion deals with the pros and cons of liquid coolant concepts, like for example, liquid metal or water cooling. Thereafter, we will present two rather contrary DEMO divertor concepts which are based on today's knowledge on refractory materials science, fabrication and joining technology. The first improved concept uses water flowing through steel pipes, typically made of Eurofer steel. It is well known that using Eurofer at low temperatures is critical due to its severe embrittlement under neutron irradiation. Here we make a proposal how it could be possible to use the Eurofer steel anyway: the solution could consist in a limited operation period followed by an annealing cycle at 550 °C for a few hours during any maintenance shut down phases. The second design is based on the known helium cooling concept using jet impingement. Drawbacks of the actual He-cooled divertor design are small scale parts as well as the necessary high helium inlet temperature of about 600–800 °C which leads to the question: How can we deal with such high helium temperatures? This paper shows a solution for large scale components as well as a new thermal management for the helium outlet gas that we call ‘cooling of the coolant’. Both concepts are discussed in terms of materials selection due to material limits and joining technology with a special focus on the material issue using already existing and available materials.

  9. European DEMO design strategy and consequences for materials

    Federici, G.; Biel, W.; Gilbert, M. R.; Kemp, R.; Taylor, N.; Wenninger, R.

    2017-09-01

    Demonstrating the production of net electricity and operating with a closed fuel-cycle remain unarguably the crucial steps towards the exploitation of fusion power. These are the aims of a demonstration fusion reactor (DEMO) proposed to be built after ITER. This paper briefly describes the DEMO design options that are being considered in Europe for the current conceptual design studies as part of the Roadmap to Fusion Electricity Horizon 2020. These are not intended to represent fixed and exclusive design choices but rather ‘proxies’ of possible plant design options to be used to identify generic design/material issues that need to be resolved in future fusion reactor systems. The materials nuclear design requirements and the effects of radiation damage are briefly analysed with emphasis on a pulsed ‘low extrapolation’ system, which is being used for the initial design integration studies, based as far as possible on mature technologies and reliable regimes of operation (to be extrapolated from the ITER experience), and on the use of materials suitable for the expected level of neutron fluence. The main technical issues arising from the plasma and nuclear loads and the effects of radiation damage particularly on the structural and heat sink materials of the vessel and in-vessel components are critically discussed. The need to establish realistic target performance and a development schedule for near-term electricity production tends to favour more conservative technology choices. The readiness of the technical (physics and technology) assumptions that are being made is expected to be an important factor for the selection of the technical features of the device.

  10. Neutronics experiments for DEMO blanket at JAERI/FNS

    Sato, Satoshi; Ochiai, K.; Hori, J.; Verzilov, Y.; Klix, A.; Wada, M.; Terada, Y.; Yamauchi, M.; Morimoto, Y.; Nishitani, T.

    2003-01-01

    In order to verify the accuracy of the tritium production rate (TPR), neutron irradiation experiments have been performed with a mockup relevant to the fusion DEMO blanket consisting of F82H blocks, Li 2 TiO 3 blocks with a 6 Li enrichment of 40 and 95%, and beryllium blocks. Sample pellets of Li 2 TiO 3 were irradiated and the TPR was measured by a liquid scintillation counter. The TPR was also calculated using the Monte Carlo code MCNP-4B with the nuclear data library JENDL-3.2 and ENDF-B/VI. The results agreed with experimental values within the statistical error (10%) of the experiment. Accordingly, it was clarified that the TPR could be evaluated within 10% uncertainty by the calculation code and the nuclear data. In order to estimate the induced activity caused by sequential reactions in cooling water pipes in the DEMO blanket, neutron irradiation experiments have been performed using test speciments simulating the pipes. Sample metals of Fe, W, Ti, Pb, Cu, V and reduced activation ferritic steels F82H were irradiated as typical fusion materials. The effective cross-sections for incident neutron flux to calculate the radioactive nuclei ( 56 Co, 184 Re, 48 V, 206 Bi, 65 Zn and 51 Cr) due to sequential reactions were measured. From the experimental results, it was found that the effective cross-sections remarkably increases with coming closer to polyethylene board that was a substitute of water. As a result of the present study, it has become clear that the sequential reaction rates are important factors to accurately evaluate the induced activity in fusion reactors design. (author)

  11. Neutronics experiments for DEMO blanket at JAERI/FNS

    Sato, S.; Ochiai, K.; Hori, J.; Verzilov, Y.; Klix, A.; Wada, M.; Terada, Y.; Yamauchi, M.; Morimoto, Y.; Nishitani, T.

    2003-01-01

    In order to verify the accuracy of the tritium production rate (TPR), neutron irradiation experiments have been performed with a mockup relevant to the fusion DEMO blanket consisting of F82H blocks, Li 2 TiO 3 blocks with a 6 Li enrichment of 40 and 95%, and beryllium blocks. Sample pellets of Li 2 TiO 3 were irradiated and the TPR was measured by a liquid scintillation counter. The TPR was also calculated using the Monte Carlo code MCNP-4B with the nuclear data library JENDL-3.2 and ENDF-B/VI. The results agreed with experimental values within the statistical error (10%) of the experiment. Accordingly, it was clarified that the TPR could be evaluated within 10% uncertainty by the calculation code and the nuclear data. In order to estimate the induced activity caused by sequential reactions in cooling water pipes in the DEMO blanket, neutron irradiation experiments have been performed using test specimens simulating the pipes. Sample metals of Fe, W, Ti, Pb, Cu, V and reduced activation ferritic steel F82H were irradiated as typical fusion materials. The effective cross- sections for incident neutron flux to calculate the radioactive nuclei ( 56 Co, 184 Re, 48 V, 206 Bi, 65 Zn and 51 Cr) due to sequential reactions were measured. From the experimental results, it was found that the effective cross-sections remarkably increases with coming closer to polyethylene board that was a substitute of water. As a result of the present study, it has become clear that the sequential reaction rates are important factors to accurately evaluate the induced activity in fusion reactors design. (author)

  12. Development of Tokamak Reactor System Code and Performance for Early Realization of DEMO

    Hong, B. G.; Lee, D. W.; Kim, Y.

    2006-01-01

    To develop the concepts of DEMO and identify the design parameters, dependence on performance objectives, design features and physical and technical constraints have to be considered. System analyses are necessary to find device variables which optimize figures of merit such as major radius, ignition margin, divertor heat load, neutron wall load, etc. Demonstration fusion power plant, DEMO is regarded as the last step before the development of a commercial fusion reactor in Korea National Basic Plan for the Development of Fusion Energy. The DEMO should demonstrate a net electric power generation, a tritium self sufficiency, and the safety aspect of a power plant. Performance of DEMO for early realization has been investigated with a limited extension from the plasma physics and technology in the 2nd phase of the ITER operation (EPP phase)

  13. Plasma regimes and research goals of JT-60SA towards ITER and DEMO

    Kamada, Y.; Ide, S.; Fujita, T.; Suzuki, T.; Matsunaga, G.; Yoshida, M.; Shinohara, K.; Urano, H.; Nakano, T.; Sakurai, S.; Kawashima, H.; Barabaschi, P.; Lackner, K.; Ishida, S.; Bolzonella, T.

    2011-01-01

    The JT-60SA device has been designed as a highly shaped large superconducting tokamak with a variety of plasma actuators (heating, current drive, momentum input, stability control coils, resonant magnetic perturbation coils, W-shaped divertor, fuelling, pumping, etc) in order to satisfy the central research needs for ITER and DEMO. In the ITER- and DEMO-relevant plasma parameter regimes and with DEMO-equivalent plasma shapes, JT-60SA quantifies the operation limits, plasma responses and operational margins in terms of MHD stability, plasma transport and confinement, high-energy particle behaviour, pedestal structures, scrape-off layer and divertor characteristics. By integrating advanced studies in these research fields, the project proceeds 'simultaneous and steady-state sustainment of the key performances required for DEMO' with integrated control scenario development applicable to the highly self-regulating burning high-β high bootstrap current fraction plasmas.

  14. A new meshless approach to map electromagnetic loads for FEM analysis on DEMO TF coil system

    Biancolini, Marco Evangelos; Brutti, Carlo; Giorgetti, Francesco; Muzzi, Luigi; Turtù, Simonetta; Anemona, Alessandro

    2015-01-01

    Graphical abstract: - Highlights: • Generation and mapping of magnetic load on DEMO using radial basis function. • Good agreement between RBF interpolation and EM TOSCA computations. • Resultant forces are stable with respect to the target mesh used. • Stress results are robust and accurate even if a coarse cloud is used for RBF interpolation. - Abstract: Demonstration fusion reactors (DEMO) are being envisaged to be able to produce commercial electrical power. The design of the DEMO magnets and of the constituting conductors is a crucial issue in the overall engineering design of such a large fusion machine. In the frame of the EU roadmap of the so-called fast track approach, mechanical studies of preliminary DEMO toroidal field (TF) coil system conceptual designs are being enforced. The magnetic field load acting on the DEMO TF coil conductor has to be evaluated as input in the FEM model mesh, in order to evaluate the stresses on the mechanical structure. To gain flexibility, a novel approach based on the meshless method of radial basis functions (RBF) has been implemented. The present paper describes this original and flexible approach for the generation and mapping of magnetic load on DEMO TF coil system.

  15. Recent technical progress on BA Program: DEMO activities and IFMIF/EVEDA

    Yamanishi, T.; Asakura, N.; Tobita, K.; Ohira, S. [Japan Atomic Energy Agency, Rokkasho, Aomori (Japan); Federici, G. [EFDA Close Support Unit, Garching (Germany); Heidinger, R. [Fusion for Energy, Garching (Germany); Knaster, J. [BA IFMIF/EVEDA Project Team, Rokkasho, Aomori (Japan); Clement, S. [Fusion for Energy, Barcelona (Spain); Nakajima, N. [BA IFERC Project Team, Rokkasho, Aomori (Japan)

    2016-11-01

    The Broader Approach (BA) activities consists of three major projects: the International Fusion Energy Research Center (IFERC) project, the International Fusion Materials Irradiation Facility/Engineering Validation and Engineering Design Activities (IFMIF/EVEDA) project, and the Satellite Tokamak Programme (STP, JT-60SA). These projects have been carried out to obtain basic data for the design of DEMO fusion reactor from 2007. For 8-year activities, the above projects could produce a set of fruitful results for the DEMO reactor. DEMO design activity has been conducted to build a set of DEMO design bases in accordance with a series of discussion between EU and JA. In the DEMO R&D activities, five basic R&D subjects for a DEMO blanket system have been selected, and been studies under close collaborations between EU and JA: structure materials (RAFM steels and SiC/SiC composites), functional materials (tritium breeders and neutron multipliers), and tritium technology. Some additional R&D subjects recommended by peer review comments have also been studied successfully in recent years. Regarding the IFMIF/EVEDA project, some main components of the accelerator facility been designed and tested. The validation test using EVEDA Lithium Test Loop (ELTL) was also completed successfully in October 2014.

  16. Pharmaceutical powder compaction technology

    Çelik, Metin

    2011-01-01

    ... through the compaction formulation process and application. Compaction of powder constituents both active ingredient and excipients is examined to ensure consistent and reproducible disintegration and dispersion profiles...

  17. Optimization and limitations of known DEMO divertor concepts

    Reiser, Jens, E-mail: Jens.Reiser@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, 76021 Karlsruhe (Germany); Rieth, Michael [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer Limitations of the materials. Black-Right-Pointing-Pointer Improved H{sub 2}O cooled divertor. Black-Right-Pointing-Pointer Improved He cooled divertor. - Abstract: In this work we will introduce and discuss improvements for two types of DEMO divertors based on known designs: (i) gas cooled designs and (ii) liquid coolant concepts. In a first step, the advantages and disadvantages of gas cooling as well as the necessity of a jet impingement to increase the heat transfer coefficients will be discussed. Further discussion deals with the pros and cons of liquid coolant concepts, like for example, liquid metal or water cooling. Thereafter, we will present two rather contrary DEMO divertor concepts which are based on today's knowledge on refractory materials science, fabrication and joining technology. The first improved concept uses water flowing through steel pipes, typically made of Eurofer steel. It is well known that using Eurofer at low temperatures is critical due to its severe embrittlement under neutron irradiation. Here we make a proposal how it could be possible to use the Eurofer steel anyway: the solution could consist in a limited operation period followed by an annealing cycle at 550 Degree-Sign C for a few hours during any maintenance shut down phases. The second design is based on the known helium cooling concept using jet impingement. Drawbacks of the actual He-cooled divertor design are small scale parts as well as the necessary high helium inlet temperature of about 600-800 Degree-Sign C which leads to the question: How can we deal with such high helium temperatures? This paper shows a solution for large scale components as well as a new thermal management for the helium outlet gas that we call 'cooling of the coolant'. Both concepts are discussed in terms of materials selection due to material limits and joining technology with a special focus on the material issue using already existing and

  18. Pre-conceptual design assessment of DEMO remote maintenance

    Loving, A., E-mail: antony.loving@ccfe.ac.uk [EURATOM/Culham Center Fusion Energy, Culham Science Centre, OX14 3DB, Abingdon (United Kingdom); Crofts, O.; Sykes, N.; Iglesias, D.; Coleman, M.; Thomas, J. [EURATOM/Culham Center Fusion Energy, Culham Science Centre, OX14 3DB, Abingdon (United Kingdom); Harman, J. [EFDA Close Support Unit Garching, Boltzmannstaße 2, D-85748 Garching Bei München (Germany); Fischer, U. [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Sanz, J. [Instituto de Fusión Nuclear/UPM, Madrid (Spain); Siuko, M. [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT (Finland); Mittwollen, M. [Karlsruhe Institute of Technology, Institut für Fördertechnik und Logistiksysteme, Gotthard-Franz-Straße 8, Geb.50.38, 76131 Karlsruhe (Germany)

    2014-10-15

    EDFA, as part of the Power Plant Physics and Technology programme, has been working on the pre-conceptual design of a Demonstration Power Plant (DEMO). As part of this programme, a review of the remote maintenance strategy considered maintenance solutions compatible with expected environmental conditions, whilst showing potential for meeting the plant availability targets. A key finding was that, for practical purposes, the expected radiation levels prohibit the use of complex remote handling operations to replace the first wall. In 2012/2013, these remote maintenance activities were further extended, providing an insight into the requirements, constraints and challenges. In particular, the assessment of blanket and divertor maintenance, in light of the expected radiation conditions and availability, has elaborated the need for a very different approach from that of ITER. This activity has produced some very informative virtual reality simulations of the blanket segments and pipe removal that are exceptionally valuable in communicating the complexity and scale of the required operations. Through these simulations, estimates of the maintenance task durations have been possible demonstrating that a full replacement of the blankets within 6 months could be achieved. The design of the first wall, including the need to use sacrificial limiters must still be investigated. In support of the maintenance operations, a first indication of the requirements of an Active Maintenance Facility (AMF) has been elaborated.

  19. Engineering options for the U.S. Fusion Demo

    Tillack, M.S.; El-Guebaly, L.; Wong, C.

    1995-01-01

    Through its successful operation, the US Fusion Demo must be sufficiently convincing that a utility or independent power producer will choose to purchase one as its next electric generating plant. A fusion power plant which is limited to the use of currently-proven technologies is unlikely to be sufficiently attractive to a utility unless fuel shortages and regulatory restrictions are far more crippling to competing energy sources than currently anticipated. In that case, the task of choosing an appropriate set of engineering technologies today involves trade-offs between attractiveness and technical risk. The design space for an attractive tokamak fusion power core is not unlimited; previous studies have shown that advanced low-activation ferritic steel, vanadium alloy, or SiC/SiC composites are the only candidates the authors have for the primary in-vessel structural material. An assessment of engineering design options has been performed using these three materials and the associated in-vessel component designs which are compatible with them

  20. Conceptual design of the beam source for the DEMO Neutral Beam Injectors

    Sonato, P.; Agostinetti, P.; Fantz, U.; Franke, T.; Furno, I.; Simonin, A.; Tran, M. Q.

    2016-12-01

    DEMO (DEMOnstration Fusion Power Plant) is a proposed nuclear fusion power plant that is intended to follow the ITER experimental reactor. The main goal of DEMO will be to demonstrate the possibility to produce electric energy from the fusion reaction. The injection of high energy neutral beams is one of the main tools to heat the plasma up to fusion conditions. A conceptual design of the Neutral Beam Injector (NBI) for the DEMO fusion reactor, is currently being developed by Consorzio RFX in collaboration with other European research institutes. High efficiency and low recirculating power, which are fundamental requirements for the success of DEMO, have been taken into special consideration for the DEMO NBI. Moreover, particular attention has been paid to the issues related to reliability, availability, maintainability and inspectability. A conceptual design of the beam source for the DEMO NBI is here presented featuring 20 sub-sources (two adjacent columns of 10 sub-sources each), following a modular design concept, with each sub-source featuring its radio frequency driver, capable of increasing the reliability and availability of the DEMO NBI. Copper grids with increasing size of the apertures have been adopted in the accelerator, with three main layouts of the apertures (circular apertures, slotted apertures and frame-like apertures for each sub-source). This design, permitting to significantly decrease the stripping losses in the accelerator without spoiling the beam optics, has been investigated with a self-consistent model able to study at the same time the magnetic field, the electrostatic field and the trajectory of the negative ions. Moreover, the status on the R&D carried out in Europe on the ion sources is presented.

  1. Safety studies on Korean fusion DEMO plant using integrated safety assessment methodology

    Oh, Kyemin; Kang, Myoung-suk; Heo, Gyunyoung; Kim, Hyoung-chan

    2014-01-01

    Highlights: •The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant. •The concepts of integrated safety assessment methodology (ISAM) that can be applied in addressing regulatory requirements and recognizing safety issues for K-DEMO were emphasized. •Phenomena identification and ranking table (PIRT) was proposed. It can recognize vulnerabilities of systems and identify the gaps in technical areas requiring additional researches. •This work is expected to contribute on the conceptual design of safety features for K-DEMO to design engineers and the guidance for regulatory requirements to licensers. -- Abstract: The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant (K-DEMO) as a part of R and D program through the National Fusion Research Institute of Korea. Even though nuclear regulation and licensing framework is well setup due to the operating and design experience of Pressurized Water Reactors (PWRs) since 1970s, the regulatory authority of South Korea has concerns on the challenge of facing new nuclear facilities including K-DEMO due to the differences in systems, materials, and inherent safety feature from conventional PWRs. Even though the follow-up of the ITER license process facilitates to deal with significant safety issues of fusion facilities, a licensee as well as a licenser should identify the gaps between ITER and DEMO in terms of safety issues. First we reviewed the methods of conducting safety analysis for unprecedented nuclear facilities such as Generation IV reactors, particularly very high temperature reactor (VHTR), which is called as integrated safety assessment methodology (ISAM). Second, the analysis for the conceptual design of K-DEMO on the basis of ISAM was conducted. The ISAM consists of five analytical tools to develop the safety requirements from licensee

  2. Safety studies on Korean fusion DEMO plant using integrated safety assessment methodology

    Oh, Kyemin; Kang, Myoung-suk [Kyung Hee University, Youngin-si, Gyeonggi-do 446-701 (Korea, Republic of); Heo, Gyunyoung, E-mail: gheo@khu.ac.kr [Kyung Hee University, Youngin-si, Gyeonggi-do 446-701 (Korea, Republic of); Kim, Hyoung-chan [National Fusion Research Institute, Daejeon-si 305-333 (Korea, Republic of)

    2014-10-15

    Highlights: •The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant. •The concepts of integrated safety assessment methodology (ISAM) that can be applied in addressing regulatory requirements and recognizing safety issues for K-DEMO were emphasized. •Phenomena identification and ranking table (PIRT) was proposed. It can recognize vulnerabilities of systems and identify the gaps in technical areas requiring additional researches. •This work is expected to contribute on the conceptual design of safety features for K-DEMO to design engineers and the guidance for regulatory requirements to licensers. -- Abstract: The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant (K-DEMO) as a part of R and D program through the National Fusion Research Institute of Korea. Even though nuclear regulation and licensing framework is well setup due to the operating and design experience of Pressurized Water Reactors (PWRs) since 1970s, the regulatory authority of South Korea has concerns on the challenge of facing new nuclear facilities including K-DEMO due to the differences in systems, materials, and inherent safety feature from conventional PWRs. Even though the follow-up of the ITER license process facilitates to deal with significant safety issues of fusion facilities, a licensee as well as a licenser should identify the gaps between ITER and DEMO in terms of safety issues. First we reviewed the methods of conducting safety analysis for unprecedented nuclear facilities such as Generation IV reactors, particularly very high temperature reactor (VHTR), which is called as integrated safety assessment methodology (ISAM). Second, the analysis for the conceptual design of K-DEMO on the basis of ISAM was conducted. The ISAM consists of five analytical tools to develop the safety requirements from licensee

  3. Design concept of K-DEMO for near-term implementation

    Kim, K.; Im, K.; Kim, H. C.; Oh, S.; Park, J. S.; Kwon, S.; Lee, Y. S.; Yeom, J. H.; Lee, C.; Lee, G.-S.; Neilson, G.; Kessel, C.; Brown, T.; Titus, P.; Mikkelsen, D.; Zhai, Y.

    2015-05-01

    A Korean fusion energy development promotion law (FEDPL) was enacted in 2007. As a following step, a conceptual design study for a steady-state Korean fusion demonstration reactor (K-DEMO) was initiated in 2012. After the thorough 0D system analysis, the parameters of the main machine characterized by the major and minor radii of 6.8 and 2.1 m, respectively, were chosen for further study. The analyses of heating and current drives were performed for the development of the plasma operation scenarios. Preliminary results on lower hybrid and neutral beam current drive are included herein. A high performance Nb3Sn-based superconducting conductor is adopted, providing a peak magnetic field approaching 16 T with the magnetic field at the plasma centre above 7 T. Pressurized water is the prominent choice for the main coolant of K-DEMO when the balance of plant development details is considered. The blanket system adopts a ceramic pebble type breeder. Considering plasma performance, a double-null divertor is the reference configuration choice of K-DEMO. For a high availability operation, K-DEMO incorporates a design with vertical maintenance. A design concept for K-DEMO is presented together with the preliminary design parameters.

  4. Activity inventories and decay heat calculations for a DEMO with HCPB and HCLL blanket modules

    Stankunas, Gediminas; Tidikas, Andrius; Pereslavstev, Pavel; Catalán, Juan; García, Raquel; Ogando, Francisco; Fischer, Ulrich

    2016-01-01

    Highlights: • The afterheat and activity inventories were calculated for Eurofer steel which is the reference structural material for DEMO. • The decay heat for the HCPB DEMO was found to be larger than for the HCLL both for short and longer cooling times. • The comparison calculations were performed for a single outboard blanket module of the HCLL DEMO assuming High-Temperature Ferritic–Martensitic (HT-FM) steel and SS-316 (LN) as structural material. - Abstract: Activation inventories, decay heat and radiation doses are important nuclear quantities which need to be assessed on a reliable basis for the safe operation of a fusion nuclear power reactor. The afterheat and activity inventories were shown to be dominated by the Eurofer steel which is the reference structural material for DEMO. The decay heat for the HCPB DEMO was found to be larger than for the HCLL both for short (a few days) and longer (more than a year) cooling times. As for the alternative steels, the induced radioactivity was turned out to be lowest for the SS-316 until about 200 years after shut-down. Afterwards, the activity level of SS-316 steel was found to be the highest. For these times, the activity of both Eurofer and the HT-FM steel is about one order of magnitude lower.

  5. Activity inventories and decay heat calculations for a DEMO with HCPB and HCLL blanket modules

    Stankunas, Gediminas, E-mail: gediminas.stankunas@lei.lt [Lithuanian Energy Institute, Laboratory of Nuclear Installation Safety, Breslaujos Str. 3, LT-44403 Kaunas (Lithuania); Tidikas, Andrius [Lithuanian Energy Institute, Laboratory of Nuclear Installation Safety, Breslaujos Str. 3, LT-44403 Kaunas (Lithuania); Pereslavstev, Pavel [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Catalán, Juan; García, Raquel; Ogando, Francisco [Departamento de Ingeniería Energética, UNED, 28040 Madrid (Spain); Fischer, Ulrich [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2016-11-01

    Highlights: • The afterheat and activity inventories were calculated for Eurofer steel which is the reference structural material for DEMO. • The decay heat for the HCPB DEMO was found to be larger than for the HCLL both for short and longer cooling times. • The comparison calculations were performed for a single outboard blanket module of the HCLL DEMO assuming High-Temperature Ferritic–Martensitic (HT-FM) steel and SS-316 (LN) as structural material. - Abstract: Activation inventories, decay heat and radiation doses are important nuclear quantities which need to be assessed on a reliable basis for the safe operation of a fusion nuclear power reactor. The afterheat and activity inventories were shown to be dominated by the Eurofer steel which is the reference structural material for DEMO. The decay heat for the HCPB DEMO was found to be larger than for the HCLL both for short (a few days) and longer (more than a year) cooling times. As for the alternative steels, the induced radioactivity was turned out to be lowest for the SS-316 until about 200 years after shut-down. Afterwards, the activity level of SS-316 steel was found to be the highest. For these times, the activity of both Eurofer and the HT-FM steel is about one order of magnitude lower.

  6. Neutronic analyses of design issues affecting the tritium breeding performance in different DEMO blanket concepts

    Pereslavtsev, Pavel, E-mail: pavel.pereslavtsev@kit.edu [Karlsruhe Institute for Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bachmann, Christian [EUROfusion – Programme Management Unit, Boltzmannstrasse 2, 85748 Garching (Germany); Fischer, Ulrich [Karlsruhe Institute for Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2016-11-01

    Highlights: • Realistic 3D MCNP model based on the CAD engineering model of DEMO. • Automated procedure for the generation and arrangement of the blanket modules for different DEMO concepts: HCPB, HCLL, WCLL, DCLL. • Several parameters affecting tritium breeding ratio (TBR) were investigated. • A set of practical guidelines was prepared for the designers developing the individual breeding blanket concepts. - Abstract: Neutronic analyses were performed to assess systematically the tritium breeding ratio (TBR) variations in the DEMO for the different blanket concepts HCPB, HCLL, WCLL and DCLL DEMOs due to modifications of the blanket configurations. A dedicated automated procedure was developed to fill the breeding modules in the common generic model in correspondence to the different concepts. The TBR calculations were carried out using the MCNP5 Monte Carlo code. The following parameters affecting the global TBR were investigated: TBR poloidal distribution, radial breeder zone depth, {sup 6}Li enrichment, steel content in the breeder modules, poloidal segmentation of the breeder blanket volume, size of gaps between blankets, thickness of the first wall and of the tungsten armour. Based on the results a set of practical guidelines was prepared for the designers developing the individual breeding blanket concepts with the goal to achieve the required tritium breeding performance in DEMO.

  7. Neutronic analyses of design issues affecting the tritium breeding performance in different DEMO blanket concepts

    Pereslavtsev, Pavel; Bachmann, Christian; Fischer, Ulrich

    2016-01-01

    Highlights: • Realistic 3D MCNP model based on the CAD engineering model of DEMO. • Automated procedure for the generation and arrangement of the blanket modules for different DEMO concepts: HCPB, HCLL, WCLL, DCLL. • Several parameters affecting tritium breeding ratio (TBR) were investigated. • A set of practical guidelines was prepared for the designers developing the individual breeding blanket concepts. - Abstract: Neutronic analyses were performed to assess systematically the tritium breeding ratio (TBR) variations in the DEMO for the different blanket concepts HCPB, HCLL, WCLL and DCLL DEMOs due to modifications of the blanket configurations. A dedicated automated procedure was developed to fill the breeding modules in the common generic model in correspondence to the different concepts. The TBR calculations were carried out using the MCNP5 Monte Carlo code. The following parameters affecting the global TBR were investigated: TBR poloidal distribution, radial breeder zone depth, "6Li enrichment, steel content in the breeder modules, poloidal segmentation of the breeder blanket volume, size of gaps between blankets, thickness of the first wall and of the tungsten armour. Based on the results a set of practical guidelines was prepared for the designers developing the individual breeding blanket concepts with the goal to achieve the required tritium breeding performance in DEMO.

  8. Concept for a vertical maintenance remote handling system for multi module blanket segments in DEMO

    Coleman, M.; Sykes, N.; Cooper, D.; Iglesias, D.; Bastow, R.; Loving, A.; Harman, J.

    2014-01-01

    Highlights: •A conceptual architectural model for a vertical maintenance DEMO is presented. •Novel concepts for a set of DEMO remote handling equipment are put forward. •Remote maintenance of a multi module segment blanket is found to be feasible. •The criticality of space in the vertical port is highlighted. -- Abstract: The anticipated high neutron flux, and the consequent damage to plasma-facing components in DEMO, results in the need to regularly replace the tritium breeding and radiation shielding blanket. The current European multi module segment (MMS) blanket concept favours a less invasive small port entry maintenance system over large sector transport concepts, because of the reduced impact on other tokamak systems – particularly the magnetic coils. This paper presents a novel conceptual remote maintenance strategy for a Vertical Maintenance Scheme DEMO, incorporating substantiated designs for an in-vessel mover, to detach and attach the blanket segments, and cask-housed vertical maintenance devices to open and close access ports, cut and join service connections, and extract blanket segments from the vessel. In addition, a conceptual architectural model for DEMO was generated to capture functional and spatial interfaces between the remote maintenance equipment and other systems. Areas of further study are identified in order to comprehensively establish the feasibility of the proposed maintenance system

  9. Design of a permeator against vacuum for tritium extraction from eutectic lithium-lead in a DCLL DEMO

    Garcinuño, Belit, E-mail: belit.garcinuno@ciemat.es [CIEMAT-LNF (Laboratorio Nacional de Fusión), Madrid (Spain); Rapisarda, David [CIEMAT-LNF (Laboratorio Nacional de Fusión), Madrid (Spain); Fernández, Iván [Fundación & Departamento de Ingeniería Energética, UNED, Madrid (Spain); CIEMAT-LNF (Laboratorio Nacional de Fusión), Madrid (Spain); Moreno, Carlos; Palermo, Iole; Ibarra, Ángel [CIEMAT-LNF (Laboratorio Nacional de Fusión), Madrid (Spain)

    2017-04-15

    Highlights: • A conceptual design of a Permeator Against Vacuum is presented. • The efficiency is dependent on geometry and Tritium transport. • The use of different membrane materials is discussed. • A squared PAV with alternated PbLi flowing and vacuum flat ducts is designed. • 80% efficiency of Tritium extraction is accomplished under DCLL-BB requirements. - Abstract: One of the most important issues in future fusion power plants is the extraction of tritium generated in the breeders in order to achieve self-sufficiency. When the breeder is a liquid metal one of the most promising techniques is the Permeation Against Vacuum, whose principle is based on tritium diffusion through a permeable membrane in contact with the liquid metal carrier and its further extraction by a vacuum pump. A conceptual design of permeator has been developed, taking into account the features of a DEMO reactor with a Dual Coolant Lithium Lead (DCLL) breeder blanket. The study is based on the analysis of different membranes and geometries aiming at the overall efficiency (extraction capability) of the device, as well as its compatibility with the breeder material. The permeator is based on a rectangular section multi-channel distribution where the liquid metal channels and vacuum channels are alternated in order to maximize the contact area and therefore to promote tritium transport from the bulk to the walls. The resulting permeator design has an excellent estimated extraction efficiency, of 80%, in a relatively compact device.

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

    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.

  11. Tritium Cycle Design for He-cooled Blankets for Demo

    Sedano, L. A.

    2007-01-01

    Final goal of COMPU task is to develop a reliable tritium Process Flow Diagram (PFD) modelling tool for DEMO tritium cycle. With this aim, the COMPU task is devoted to: (1) Review of existing available documentation related on configuration layouts, and systems and tritium control process key technologies. (2) To select those validated and considered relevant as basis for code development. (3) Implement results from (1), and (2) in the PFD TRICICLO. This fi rst deliverable focuses on item (1) and is conceived as a managerial tool to: (1) establish and discuss the correct inputs, (2) to identify existing lack of basic information and (3) to establish the general demands and characteristics for the development of an advanced PFD model. Thus, in order to discuss and determine the basic information required for future new developments of the task, this report presents a review of the documentation of: (1) The outline of total cycle and system configuration with the main tritium system design specifications. (2) The ultimate processing technologies with the associated design of their implementing units. (3) Key parameters needed to describe processes and modes of operation of the system units. (4) An overview of the existing models for cycle and units with a general analysis of their performances and limitations. Thus, this report is a direct review of the base information generated previously in the context of tasks of the EU FT Programmers (reported in EFDA Green Books) and available results in open fields literature provided by parallel Programmes abroad (JP, US, RF). (Author) 102 refs

  12. Tritium Cycle Design for He-cooled Blankets for Demo

    Sedano, L. A.

    2007-09-27

    Final goal of COMPU task is to develop a reliable tritium Process Flow Diagram (PFD) modelling tool for DEMO tritium cycle. With this aim, the COMPU task is devoted to: (1) Review of existing available documentation related on configuration layouts, and systems and tritium control process key technologies. (2) To select those validated and considered relevant as basis for code development. (3) Implement results from (1), and (2) in the PFD TRICICLO. This fi rst deliverable focuses on item (1) and is conceived as a managerial tool to: (1) establish and discuss the correct inputs, (2) to identify existing lack of basic information and (3) to establish the general demands and characteristics for the development of an advanced PFD model. Thus, in order to discuss and determine the basic information required for future new developments of the task, this report presents a review of the documentation of: (1) The outline of total cycle and system configuration with the main tritium system design specifications. (2) The ultimate processing technologies with the associated design of their implementing units. (3) Key parameters needed to describe processes and modes of operation of the system units. (4) An overview of the existing models for cycle and units with a general analysis of their performances and limitations. Thus, this report is a direct review of the base information generated previously in the context of tasks of the EU FT Programmers (reported in EFDA Green Books) and available results in open fields literature provided by parallel Programmes abroad (JP, US, RF). (Author) 102 refs.

  13. Modelling of DEMO core plasma consistent with SOL/divertor simulations for long-pulse scenarios with impurity seeding

    Pacher, G.W.; Pacher, H.D.; Janeschitz, G.; Kukushkin, A.S.; Kotov, V.; Reiter, D.

    2007-01-01

    The integrated core-pedestal-SOL model is applied to the simulation of a typical DEMO operation. Impurity seeding is used to reduce the power load on the divertor to acceptable levels. The influence on long-pulse operation of impurity seeding with various impurities is investigated. DEMO operation at acceptable peak power loads and long-pulse lengths is demonstrated

  14. Research at ITER towards DEMO: Specific reactor diagnostic studies to be carried out on ITER

    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.

  15. Research at ITER towards DEMO: Specific reactor diagnostic studies to be carried out on ITER

    Krasilnikov, A. V.; Kaschuck, Y. A.; Vershkov, V. A.; Petrov, A. A.; Petrov, V. G.; Tugarinov, S. N.

    2014-01-01

    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

  16. Research at ITER towards DEMO: Specific reactor diagnostic studies to be carried out on ITER

    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.

  17. Conceptual design of the DEMO neutral beam injectors: main developments and R&D achievements

    Sonato, P.; Agostinetti, P.; Bolzonella, T.; Cismondi, F.; Fantz, U.; Fassina, A.; Franke, T.; Furno, I.; Hopf, C.; Jenkins, I.; Sartori, E.; Tran, M. Q.; Varje, J.; Vincenzi, P.; Zanotto, L.

    2017-05-01

    The objectives of the nuclear fusion power plant DEMO, to be built after the ITER experimental reactor, are usually understood to lie somewhere between those of ITER and a ‘first of a kind’ commercial plant. Hence, in DEMO the issues related to efficiency and RAMI (reliability, availability, maintainability and inspectability) are among the most important drivers for the design, as the cost of the electricity produced by this power plant will strongly depend on these aspects. In the framework of the EUROfusion Work Package Heating and Current Drive within the Power Plant Physics and Development activities, a conceptual design of the neutral beam injector (NBI) for the DEMO fusion reactor has been developed by Consorzio RFX in collaboration with other European research institutes. In order to improve efficiency and RAMI aspects, several innovative solutions have been introduced in comparison to the ITER NBI, mainly regarding the beam source, neutralizer and vacuum pumping systems.

  18. Conceptual study of ECH/ECCD system for fusion DEMO plant

    Sakamoto, K.; Takahashi, K.; Kasugai, A.; Minami, R.; Kobayashi, N.; Nishio, S.; Sato, M.; Tobita, K.

    2006-01-01

    The conceptual study of the electron cyclotron heating and current drive (ECH/ECCD) system for a DEMO reactor was carried out. The ECH/ECCD system was considered on the basis of a design of the DEMO reactor by JAERI. The reactor is a low aspect ratio tokamak, and its size and magnetic field are similar to those of ITER. Therefore, many ECH/ECCD technologies developed at 170 GHz for ITER can be applied. Truly continuous operation is needed for DEMO, and the neutron fluence from the plasma is two orders of magnitude higher than that of ITER. An RF launcher that has reliability under the condition of high neutron fluence is, critically, important. For power deposition control in the plasma, a gyrotron frequency tuning system is considered as the primary candidate to realize a simple and robust launching system, but two RF beam steering systems are discussed as alternatives

  19. Report on the diagnostics for control of the fusion DEMO reactors

    2014-05-01

    The range of diagnostics that can be used in DEMO will be severely restricted compared to that used in the current experiments or to be used in ITER. Therefore, a study is planned on the technical feasibility of sensors and diagnostics on the basis of specific tokamak and helical DEMO designs, with the involvement of a wide range of specialists covering reactor design, diagnostics, neutronics, reactor structure, remote maintenance, plasma physics, plasma and machine control, and computer simulation. Topics included typical characteristic times of target plasma behavior, diagnostics tools with their resolution and lifetime, response time of actuators, and plasmas. Through these studies, possible candidates for DEMO diagnostics were identified. The outcome of two years of activities is summarized in this report with a recommendation to the government of Japan. (J.P.N.)

  20. He-cooled divertor for DEMO. Fabrication technology for tungsten cooling fingers

    Reiser, J.; Norajitra, P.; Widak, V.; Krauss, W. [Forschungszentrum Karlsruhe GmbH (Germany)

    2008-07-01

    A modular helium-cooled divertor design based on the multi-jet impingement concept (HEMJ) has been developed for the ''post-ITER'' demonstration reactor (DEMO) at the Forschungszentrum Karlsruhe [1, 2]. The main function of the divertor is to keep the plasma free from impurities by catching particles, such as fusion ash and eroded particles from the first wall. From the divertor surface, a maximum heat load of 10 MW/m{sup 2} at least has to be removed. The whole divertor is split up into a number of cassettes (48 according to the latest design studies [3]). Each cassette is cooled separately. The target plates are provided with several cooling fingers to keep the thermal stresses low. Each cooling finger consists of a tungsten tile which is brazed to a thimble-like cap made of a tungsten alloy W-1%La2O3 (WL10) underneath. The thimble has to be connected to the ODS EUROFER steel structure, which is accomplished by brazing again. The tungsten/tungsten brazing is exposed to 1200 C operation temperature while the tungsten/steel brazing joint must withstand 700 C operating temperature. Cooling of the finger is achieved by multi-jet impingement with helium. The inlet temperature of helium is 600 C and rises up to 700 C at the outlet. With this kind of cooling, a mean heat transfer coefficient of 35.000 W/(m{sup 2*}K) can be reached. This compact report will focus on the manufacturing of such a cooling finger unit at FZK. It will cover the machining of the tungsten tile as well as of the thimble and, the brazing of the parts. The major aim of this activity is, on the one hand, to obtain functioning mock-ups with high quality and high reliability, in particular in terms of minimising the surface roughness, cracks, and micro-cracks. On the other hand, effort should also be laid on realising the mass production from economic point of view. (orig.)

  1. Assessment of hypervapotron heat sink performance using CFD under DEMO relevant first wall conditions

    Domalapally, Phani, E-mail: p_kumar.domalapally@cvrez.cz

    2016-11-01

    Highlights: • Performance of Hypervapotron heat sink was tested for First wall limiter application. • Two different materials were tested Eurofer 97 and CuCrZr at PWR conditions. • Simulations were performed to see the effect of the different inlet conditions and materials on the maximum temperature. • It was found that CuCrZr heat sink performance is far better than Eurofer heat sink at the same operating conditions. - Abstract: Among the proposed First Wall (FW) cooling concepts for European Demonstration Fusion Power Plant (DEMO), water cooled FW is one of the options. The heat flux load distribution on the FW of the DEMO reactor is not yet precisely defined. But if the heat loads on the FW are extrapolated from ITER conditions, the numbers are quite high and have to be handled none the less. The design of the FW itself is challenging as the thermal conductivity ratio of heat sink materials in ITER (CuCrZr) and in DEMO (Eurofer 97) is ∼10–12 and the operating conditions are of Pressurized Water Reactor (PWR) in DEMO instead of 70 °C and 4 MPa as in ITER. This paper analyzes the performance of Hypervapotron (HV) heat sink for FW limiter application under DEMO conditions. Where different materials, temperatures, heat fluxes and velocities are considered to predict the performance of the HV, to establish its limits in handling the heat loads before reaching the upper limits from temperature point of view. In order to assess the performance, numerical simulations are performed using commercial CFD code, which was previously validated in predicting the thermal hydraulic performance of HV geometry. Based on the results the potential usage of HV heat sink for DEMO will be assessed.

  2. Progress in the RAMI analysis of a conceptual LHCD system for DEMO

    Mirizzi, F.

    2014-02-01

    Reliability, Availability, Maintainability and Inspectability (RAMI) concepts and techniques, that acquired great importance during the first manned space missions, have been progressively extended to industrial, scientific and consumer equipments to assure them satisfactory performances and lifetimes. In the design of experimental facilities, like tokamaks, mainly aimed at demonstrating validity and feasibility of scientific theories, RAMI analysis has been often left aside. DEMO, the future prototype fusion reactors, will be instead designed for steadily delivering electrical energy to commercial grids, so that the RAMI aspects will assume an absolute relevance since their initial design phases. A preliminary RAMI analysis of the LHCD system for the conceptual EU DEMO reactor is given in the paper.

  3. Progress in the RAMI analysis of a conceptual LHCD system for DEMO

    Mirizzi, F.

    2014-01-01

    Reliability, Availability, Maintainability and Inspectability (RAMI) concepts and techniques, that acquired great importance during the first manned space missions, have been progressively extended to industrial, scientific and consumer equipments to assure them satisfactory performances and lifetimes. In the design of experimental facilities, like tokamaks, mainly aimed at demonstrating validity and feasibility of scientific theories, RAMI analysis has been often left aside. DEMO, the future prototype fusion reactors, will be instead designed for steadily delivering electrical energy to commercial grids, so that the RAMI aspects will assume an absolute relevance since their initial design phases. A preliminary RAMI analysis of the LHCD system for the conceptual EU DEMO reactor is given in the paper

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

    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.)

  5. Overview of progress on the European DEMO remote maintenance strategy

    Crofts, Oliver, E-mail: oliver.crofts@ccfe.ac.uk [RACE/UKAEA, Culham Science Centre, Abingdon OX14 3DB, Oxon (United Kingdom); Loving, Antony; Iglesias, Daniel [RACE/UKAEA, Culham Science Centre, Abingdon OX14 3DB, Oxon (United Kingdom); Coleman, Matti [EUROfusion Consortium, PPP& T Department, Boltzmannstraße 2, 85748 Garching (Germany); Siuko, Mikko [VTT Technical Research Centre of Finland, PO Box 1000, FI-02044 VTT, Espoo (Finland); Mittwollen, Martin [KIT Institut für Fördertechnik und Logistiksysteme, Gotthard-Franz-Straße 8, Geb.50.38, 76131 Karlsruhe (Germany); Queral, Vicente [CIEMAT Laboratorio Nacional de Fusión, Edif. 66, Avenida Complutense 40, 28040 Madrid (Spain); Vale, Alberto [IST, Instituto Superior Técnico, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Villedieu, Eric [CEA, IRFM, 13108 St Paul lez Durance (France)

    2016-11-01

    Highlights: • The remote maintenance strategy is applicable to the range of tokamak and component options currently under consideration in Europe • The remote maintenance development work is concentrating on the application and limits of the immature technologies that pose the greatest risk to the feasibility of the maintenance strategy • Position control during the handling of the in-vessel components is one of the areas of high risk and a system is being developed and will be tested prior to concept design to demonstrate the feasibility and capability of a system capable of real time incorporation of changing kinematic data provided by a structural simulator running in parallel • In-vessel recovery and rescue and the pipe joining technology form two more of the high risk areas where developments are being concentrated - Abstract: The EU-DEMO remote maintenance strategy must be relevant for a range of in-vessel component design options. The remote maintenance project must provide an understanding of the limits of the strategy and technologies so as to inform the developing plant design of the maintenance constraints. A comprehensive set of maintenance requirements has been produced, in conjunction with the plant designers, against which design options can be assessed. The proposed maintenance solutions are based around a strategy that deploys casks above each of the vertical ports to exchange the blanket segments and at each of the divertor ports to exchange the divertor cassettes. The casks deploy remote handling equipment to open and close the vacuum vessel, remove and re-install pipework, and replace the in-vessel components. A technical design risk assessment has shown that the largest risks are common to all of the proposed solutions and that they are associated with two key issues, first; the ability to handle the large blanket and divertor components to the required positional accuracy with limited viewing and position feedback, and second; to

  6. Overview of progress on the European DEMO remote maintenance strategy

    Crofts, Oliver; Loving, Antony; Iglesias, Daniel; Coleman, Matti; Siuko, Mikko; Mittwollen, Martin; Queral, Vicente; Vale, Alberto; Villedieu, Eric

    2016-01-01

    Highlights: • The remote maintenance strategy is applicable to the range of tokamak and component options currently under consideration in Europe • The remote maintenance development work is concentrating on the application and limits of the immature technologies that pose the greatest risk to the feasibility of the maintenance strategy • Position control during the handling of the in-vessel components is one of the areas of high risk and a system is being developed and will be tested prior to concept design to demonstrate the feasibility and capability of a system capable of real time incorporation of changing kinematic data provided by a structural simulator running in parallel • In-vessel recovery and rescue and the pipe joining technology form two more of the high risk areas where developments are being concentrated - Abstract: The EU-DEMO remote maintenance strategy must be relevant for a range of in-vessel component design options. The remote maintenance project must provide an understanding of the limits of the strategy and technologies so as to inform the developing plant design of the maintenance constraints. A comprehensive set of maintenance requirements has been produced, in conjunction with the plant designers, against which design options can be assessed. The proposed maintenance solutions are based around a strategy that deploys casks above each of the vertical ports to exchange the blanket segments and at each of the divertor ports to exchange the divertor cassettes. The casks deploy remote handling equipment to open and close the vacuum vessel, remove and re-install pipework, and replace the in-vessel components. A technical design risk assessment has shown that the largest risks are common to all of the proposed solutions and that they are associated with two key issues, first; the ability to handle the large blanket and divertor components to the required positional accuracy with limited viewing and position feedback, and second; to

  7. Tritium transport in HCLL and WCLL DEMO blankets

    Candido, Luigi [DENERG, Politecnico di Torino, Corso Duca degli Abruzzi 24, 10129 Torino (Italy); Utili, Marco [ENEA UTIS- C.R. Brasimone, Bacino del Brasimone, Camugnano, BO (Italy); Nicolotti, Iuri [DENERG, Politecnico di Torino, Corso Duca degli Abruzzi 24, 10129 Torino (Italy); Zucchetti, Massimo, E-mail: massimo.zucchetti@polito.it [DENERG, Politecnico di Torino, Corso Duca degli Abruzzi 24, 10129 Torino (Italy)

    2016-11-01

    Highlights: • Tritium inventories and tritium losses are the main output of the presented model for HCLL and WCLL. • A parametric study has been performed, to show the behavior of the two systems when certain parameters are changed, in order to minimize inventories and/or losses. • An improved design is needed, in order to reduce the radiological hazard related to tritium activity. According to test number 7, HCLL-BB could be able to have a tritium inventory of 33.05 g and losses of 19.55 Ci/d. • WCLL-BB shows a very low radiological risk, much lower than that suggested (inventory: 17.48 g, losses: 3.2 Ci/d). An ptimization study has been performed aiming to minimize the water flow rate for an upgraded design. • Both for HCLL and WCLL, the most critical parameters able to produce relevant variations in inventories and losses are the helium/water fraction, the CPS/WDS and the permeation reduction factors. - Abstract: The Helium-Cooled Lithium Lead (HCLL) and Water-Cooled Lithium Lead (WCLL) Breeding Blankets are two of the four blanket designs proposed for DEMO reactor. The study of tritium transport inside the blankets is fundamental to assess their preliminary design and safety features. A mathematical model has been derived, in a new form making makes easier to determine the most critical components as far as tritium losses and tritium inventories are concerned, and to model the tritium performance of the whole system. Two cases have been studied, the former with tritium generation rate constant in time and the latter considering a typical pulsed operation for a time span of 100 h. Tritium inventories and tritium losses are the main output of the model. Tritium concentrations, inventories and losses are initially calculated and compared for the two blankets, in a reference case without permeation barriers or cold traps. A parametric study to show the behavior of the two systems when certain parameters are changed, in order to minimize inventories and

  8. Overview of Progress on the EU DEMO Reactor Magnet System Design

    Zani, L.; Bayer, C.M.; Biancolini, M.E.; Bonifetto, R.; Bruzzone, P.; Brutti, C.; Ciazynski, D.; Coleman, M.; Ďuran, Ivan; Eisterer, M.; Fietz, W.H.; Gade, P.V.; Gaio, E.; Giorgetti, F.; Goldacker, W.; Gömöry, F.; Granados, X.; Heller, R.; Hertout, P.; Hoa, C.; Kario, A.; Lacroix, B.; Lewandowska, M.; Maistrello, A.; Muzzi, L.; Nijhuis, A.; Nunio, F.; Panin, A.; Petrisor, T.; Poncet, J.-M.; Prokopec, R.; Sanmarti Cardona, M.; Savoldi, L.; Schlachter, S.I.; Sedlak, K.; Stepanov, B.; Tiseanu, I.; Torre, A.; Turtu, S.; Vallcorba, R.; Vojenciak, M.; Weiss, K.-P.; Wesche, R.; Yagotintsev, K.; Zanino, R.

    2016-01-01

    Roč. 26, č. 4 (2016), č. článku 4204505. ISSN 1051-8223 Institutional support: RVO:61389021 Keywords : DEMO * fusion * HTS * LTS * Nb3Sn * superconducting magnets Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.092, year: 2015

  9. Comparison over the nuclear analysis of the HCLL blanket for the European DEMO

    Jaboulay, Jean-Charles, E-mail: jean-charles.jaboulay@cea.fr [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Aiello, Giacomo; Aubert, Julien [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Villari, Rosaria [ENEA, UTFUS-TECN, Via E. Fermi 4, 00044 Frascati, Rome (Italy); Fischer, Ulrich [Karlsruhe Institute of Technology, 76344 Eggenstein-Leopoldshafen, Karlsruhe (Germany)

    2016-11-01

    Highlights: • A complete nuclear analysis of the DEMO HCLL has been carried out at CEA with the TRIPOLI-4{sup ®} Monte Carlo code. • The DEMO tokamak model was generated by the CAD import tool McCad. • The HCLL blankets were implemented using a previous MCNP model developed at ENEA. • A good agreement is observed between the results obtained at CEA with TRIPOLI-4 and JEFF-3.1.1 and whose obtained at ENEA with MCNP and FENDL-2.1. - Abstract: This paper presents the comparison over the nuclear analysis of the European DEMO with HCLL blanket carried out with the TRIPOLI-4{sup ®} Monte Carlo code and the JEFF-3.1.1 nuclear data library and with the MCNP5 Monte Carlo code and the FENDL-2.1 nuclear data library. The MCNP5 analysis was conducted firstly by ENEA with a detailed 3D model describing all the HCLL blanket internal structures. This MCNP5 model was converted into TRIPOLI-4{sup ®} representation for performing the nuclear analysis at CEA with the objective to demonstrate consistency between both analyses. A very good agreement was obtained for all of the relevant nuclear responses (neutron wall loading, tritium breeding ratio, nuclear heating, neutron flux distribution, etc.), validating CEA’s nuclear analysis approach, based on TRIPOLI-4{sup ®} Monte Carlo code and JEFF-3.1.1 nuclear data library, for the European DEMO.

  10. Comparison over the nuclear analysis of the HCLL blanket for the European DEMO

    Jaboulay, Jean-Charles; Aiello, Giacomo; Aubert, Julien; Villari, Rosaria; Fischer, Ulrich

    2016-01-01

    Highlights: • A complete nuclear analysis of the DEMO HCLL has been carried out at CEA with the TRIPOLI-4"® Monte Carlo code. • The DEMO tokamak model was generated by the CAD import tool McCad. • The HCLL blankets were implemented using a previous MCNP model developed at ENEA. • A good agreement is observed between the results obtained at CEA with TRIPOLI-4 and JEFF-3.1.1 and whose obtained at ENEA with MCNP and FENDL-2.1. - Abstract: This paper presents the comparison over the nuclear analysis of the European DEMO with HCLL blanket carried out with the TRIPOLI-4"® Monte Carlo code and the JEFF-3.1.1 nuclear data library and with the MCNP5 Monte Carlo code and the FENDL-2.1 nuclear data library. The MCNP5 analysis was conducted firstly by ENEA with a detailed 3D model describing all the HCLL blanket internal structures. This MCNP5 model was converted into TRIPOLI-4"® representation for performing the nuclear analysis at CEA with the objective to demonstrate consistency between both analyses. A very good agreement was obtained for all of the relevant nuclear responses (neutron wall loading, tritium breeding ratio, nuclear heating, neutron flux distribution, etc.), validating CEA’s nuclear analysis approach, based on TRIPOLI-4"® Monte Carlo code and JEFF-3.1.1 nuclear data library, for the European DEMO.

  11. Divertor remote handling for DEMO: Concept design and preliminary FMECA studies

    Carfora, D., E-mail: dario.carfora@gmail.com [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); ENEA/CREATE/Università degli studi Napoli Federico II, 80125 Napoli (Italy); Di Gironimo, G. [ENEA/CREATE/Università degli studi Napoli Federico II, 80125 Napoli (Italy); Järvenpää, J. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Huhtala, K. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Määttä, T.; Siuko, M. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland)

    2015-10-15

    Highlights: • Concept design of the RH system for the DEMO fusion power plant. • Divertor mover: hydraulic telescopic boom concept design. • An alternative solution to ITER rack and pinion divertor mover (CMM). • Divertor cassettes end effector studies. • FMECA studies started on the DEMO divertor mover. - Abstract: The paper describes a concept design of a remote handling (RH) system for replacing divertor cassettes and cooling pipes in future DEMO fusion power plant. In DEMO reactor design important considerations are the reactor availability and reliable maintenance operations. The proposed divertor mover is a hydraulic telescopic boom driven from the transportation cask through the maintenance tunnel of the reactor. The boom is divided in three sections and it is driving an end-effector in order to perform the scheduled operations of maintenance inside the vacuum vessel. Two alternative designs of the end effector to grip and manipulate the divertor cassette are presented in this work. Both concepts are hydraulically actuated, based on ITER previous studies. The divertor cassette end-effector consists of a lifting arm linked to the divertor mover, a tilting plate, a cantilever arm and a hook-plate. Taking advantage of the ITER RH background and experience, the proposed hydraulic RH system is compared with the rack and pinion system currently designed for ITER and is an object of simulations at Divertor Test Platform (DTP2) in VTT's Labs of Tampere, Finland. Pros and cons will be put in evidence.

  12. Dynamic modelling of balance of plant systems for a pulsed DEMO power plant

    Harrington, C., E-mail: Chris.Harrington@ccfe.ac.uk

    2015-10-15

    Highlights: • A fully dynamic model of the balance of plant systems for pulsed DEMO is presented. • An operating strategy for handling pulse/dwell transitions has been devised. • Operation of a water-cooled system without energy storage appears feasible. • Steam turbine cycling can be minimised if rotation speed is maintained. - Abstract: The current baseline concept for a European DEMO defines a pulsed reactor producing power for periods of 2–4 h at a time, interrupted by dwell periods of approximately half an hour, potentially leading to cyclic fatigue of the heat transfer system and power generation equipment. Thermal energy storage systems could mitigate pulsing issues; however, the requirements for such a system cannot be defined without first understanding the challenges for pulsed operation, while any system will simultaneously increase the cost and complexity of the balance of plant. This work therefore presents a dynamic model of the primary heat transfer system and associated steam plant for a water-cooled DEMO, without energy storage, capable of simulating pulsed plant operation. An operating regime is defined such that the primary coolant flows continuously throughout the dwell period while the secondary steam flow is reduced. Simulation results show minimised thermal and pressure transients in the primary circuit, and small thermally induced stresses on the steam turbine rotor. If the turbine can be kept spinning to also minimise mechanical cycling, pulsed operation of a water-cooled DEMO without thermal energy storage may be feasible.

  13. Non-destructive examination of the bonding interface in DEMO divertor fingers

    Richou, Marianne; Missirlian, Marc; Vignal, Nicolas; Cantone, Vincent; Hernandez, Caroline; Norajitra, Prachai; Spatafora, Luigi

    2013-01-01

    Highlights: • SATIR tests on DEMO divertor fingers (integrating or not He cooling system). • Millimeter size artificial defects were manufactured. • Detectability of millimeter size artificial defects was evaluated. • SATIR can detect defect in DEMO divertor fingers. • Simulations are well correlated to SATIR tests. -- Abstract: Plasma facing components (PFCs) with tungsten (W) armor materials for DEMO divertor require a high heat flux removal capability (at least 10 MW/m 2 in steady-state conditions). The reference divertor PFC concept is a finger with a tungsten tile as a protection and sacrificial layer brazed to a thimble made of tungsten alloy W – 1% La 2 O 3 (WL10). Defects may be located at the W thimble to W tile interface. As the number of fingers is considerable (>250,000), it is then a major issue to develop a reliable control procedure in order to control with a non-destructive examination the fabrication processes. The feasibility for detecting defect with infrared thermography SATIR test bed is presented. SATIR is based on the heat transient method and is used as an inspection tool in order to assess component heat transfer capability. SATIR tests were performed on fingers integrating or not the complex He cooling system (steel cartridge with jet holes). Millimeter size artificial defects were manufactured and their detectability was evaluated. Results of this study demonstrate that the SATIR method can be considered as a relevant non-destructive technique examination for the defect detection of DEMO divertor fingers

  14. RF DEMO ceramic helium cooled blanket, coolant and energy transformation systems

    Kovalenko, V.; Leshukov, A.; Poliksha, V.; Popov, A.; Strebkov, Yu.; Borisov, A.; Shatalov, G.; Demidov, V.; Kapyshev, V.

    2004-01-01

    RF DEMO-S reactor is a prototype of commercial fusion reactors for further generation. A blanket is the main element unit of the reactor design. The segment structure is the basis of the ceramic blanket. The segments mounting/dismounting operations are carried out through the vacuum vessel vertical port. The inboard/outboard blanket segment is the modules welded design, which are welded by back plate. The module contains the back plate, the first wall, lateral walls and breeding zone. The 9CrMoVNb steel is used as structural material. The module internal space formed by the first wall, lateral walls and back plate is used for breeding zone arrangement. The breeding zone design based upon the poloidal BIT (Breeder Inside Tube) concept. The beryllium is used as multiplier material and the lithium orthosilicate is used as breeder material. The helium at 0.1 MPa is used as purge gas. The cooling is provided by helium at 10 MPa. The coolant supply/return to the blanket modules are carrying out on the two independent circuits. The performed investigations of possible transformation schemes of DEMO-S blanket heat power into the electricity allowed to make a conclusion about the preferable using of traditional steam-turbine facility in the secondary circuit. (author)

  15. Overview of physics results from MAST towards ITER/DEMO and the MAST Upgrade

    Meyer, H.; Abel, I.G.; Akers, R.J.

    2013-01-01

    New diagnostic, modelling and plant capability on the Mega Ampère Spherical Tokamak (MAST) have delivered important results in key areas for ITER/DEMO and the upcoming MAST Upgrade, a step towards future ST devices on the path to fusion currently under procurement. Micro-stability analysis...

  16. The Monte Carlo approach to the economics of a DEMO-like power plant

    Bustreo, Chiara, E-mail: chiara.bustreo@igi.cnr.it; Bolzonella, Tommaso; Zollino, Giuseppe

    2015-10-15

    Highlights: • A steady state DEMO-like power plant is modelled with the FRESCO code. • The Monte Carlo method is used to assess the probability distribution of the COE. • Uncertainties on technical and economical aspects make the COE vary in a large range. • The COE can be nearly 2/3 to nearly 4 times the cost derived deterministically. - Abstract: An early assessment of the economics of a fusion power plant is a key step to ensure the technology viability in a future global energy system. The FRESCO code is here used to generate the technical, physical and economic model of a steady state DEMO-like power plant whose features are taken from the current European research activities on the DEMO design definition. The Monte Carlo method is used to perform stochastic analyses in order to assess the weight on the cost of electricity of uncertainties on technical and economical aspects. This study demonstrates that a stochastic approach offers a much better perspective over the spectrum of values that could be expected for the cost of electricity from fusion. Specifically, this analysis proves that the cost of electricity of the DEMO-like power plant studied could vary in quite large range, from nearly 2/3 to nearly 4 times the cost derived through a deterministic approach, by choosing reference values for all the stochastic parameters, taken from the literature.

  17. Overview of Progress on the EU DEMO Reactor Magnet System Design

    Zani, L.; Bayer, C.; biancolini, M.E.; Bonifetto, R.; Nijhuis, Arend; Yagotintsev, K.

    2016-01-01

    The DEMO reactor is expected to be the first application of fusion for electricity generation in the near future. To this aim, conceptual design activities are progressing in Europe (EU) under the lead of the EUROfusion Consortium in order to drive on the development of the major tokamak systems. In

  18. Preliminary concept design of the divertor remote handling system for DEMO power plant

    Carfora, D., E-mail: dario.carfora@gmail.com [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); ENEA/CREATE/University of Naples Federico II, 80125 Naples (Italy); Di Gironimo, G. [ENEA/CREATE/University of Naples Federico II, 80125 Naples (Italy); Järvenpää, J. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Huhtala, K. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Määttä, T.; Siuko, M. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland)

    2014-11-15

    Highlights: • Concept design of the RH system for the DEMO fusion power plant. • Divertor Mover: Hydraulic telescopic boom concept design. An alternative solution to ITER rack and pinion divertor mover (CMM). • Divertor cassettes end effector studies. • Transportation cask conceptual studies and logistic. - Abstract: This paper is based on the remote maintenance system project (WPRM) for the demonstration fusion power reactor (DEMO). Following ITER, DEMO aims to confirm the capability of generating several hundred of MW of net electricity by 2050. The main objective of these activities is to develop an efficient and reliable remote handling (RH) system for replacing the divertor cassettes. This paper presents the preliminary results of the concept design of the divertor RH system. The proposed divertor mover is a hydraulic telescopic boom driven from the transportation cask through the maintenance tunnel of the reactor. The boom is divided in three sections of 4 m each, and it is driving an end-effector in order to perform the scheduled operations of maintenance inside the vacuum vessel. Two alternative design of the end effector to grip and manipulate the divertor cassette are also presented in this work. Both the concepts are hydraulically actuated, basing on the ITER previous studies. The divertor cassette end-effector consists of a lifting arm linked to the divertor mover, a tilting plate, a cantilever arm and a hook-plate. The main objective of this paper is to illustrate the feasibility of DEMO divertor remote maintenance operations.

  19. DEMO relevance of the test blanket modules in ITER-Application to the European test blanket modules

    Magnani, E.; Gabriel, F.; Boccaccini, L.V.; Li-Puma, A.

    2010-01-01

    Test blanket module (TBM) testing programme in ITER as a support to DEMO design is a very important step on the road map to commercial fusion reactors although it is an ambitious task. Finding as much as possible DEMO relevant tests in view of the future DEMO blanket design is therefore a major goal since ITER and DEMO environment and loading conditions are different. To clarify and quantify the meaning of the DEMO relevance, criteria using a structural, functional and behavioural representation of the breeding blanket acting as a system are investigated. Then, a three-step strategy is proposed to carry out TBM DEMO relevant tests associated with a TBM design modification strategy. Key parameters should intensively be used as target for TBM characterization and numerical code validation. When assessing the relevance, on the other hand, not only the actual difference between DEMO and ITER values should be considered, but also whether the analyzed phenomena have a threshold and a range of applicability, as numerical simulations are usually permitted within these limits. The proposed methodology is at the end applied to the design of the HCLL TBM breeding unit configuration.

  20. Pre-conceptual studies and R and D for DEMO superconducting magnets

    Bruzzone, Pierluigi, E-mail: pierluigi.bruzzone@psi.ch

    2014-10-15

    Highlights: • Comparison of DEMO parameters vs. ITER for TF coils. • Hybridization of winding pack, Nb/Nb{sub 3}Sn, by graded layer winding. • Use of react and wind method opposite to wind and react with related advantages. • Feasibility, reliability and cost competitiveness for DEMO. - Abstract: The DEMO plant will demonstrate by mid century the feasibility of electric power generation by nuclear fusion. Since 2011, conceptual design studies are coordinated by the EFDA Power Plant Physics and Technology (PPPT) Division, with the aim of identifying requirements, propose design approaches and start R and D for the magnet system of DEMO. The input and generic boundary conditions are given by the system codes: the major radius of the tokamak is about 9 m. The proposed operating current at 13.6 T peak field is 82 kA, placing the DEMO TF conductor at substantially higher performance compared to ITER TF (68 kA/11.5 T). The innovative winding layout is a graded, layer wound with Nb{sub 3}Sn/NbTi hybridization, aiming at minimizing the size and the cost of the superconductor. Two options are considered for the Nb{sub 3}Sn conductor: one a “wind and react” cable-in-conduit (CICC) with reduced void fraction and rectangular shape. The other conductor is a “react and wind” flat cable with copper segregation and thick steel conduit assembled by longitudinal weld. The conductor designs were first drafted in 2012 and updated in 2013 based on a first round of assessments, which includes electromagnetic, thermal-hydraulic and mechanical analysis. The manufacture of full size prototype conductors is planned in 2014. The technical requirement of the DEMO superconducting magnets is highlighted in comparison to ITER and other fusion devices. The large size of the DEMO tokamak is the main challenge for the demonstration of the feasibility of power generation by fusion. Together with the technical issues, the cost of the superconducting magnets will be eventually the

  1. Preliminary analysis of K-DEMO thermal hydraulic system using MELCOR; Parametric study of hydrogen explosion

    Moon, Sung Bo; Lim, Soo Min; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

    2016-10-15

    K-DEMO (Korean fusion demonstration reactor) is future reactor for the commercializing the fusion power generation. The Design of K-DEMO is similar to that of ITER but the fusion energy generation is much bigger because ITER is experimental reactor. For this reason, K-DEMO uses more fusion reaction with bigger amount of tritium. Higher fusion power means more neutron generation that can irradiate the structure around fusion plasma. Fusion reactor can produce many kinds of radioactive material in the accident. Because of this hazard, preliminary safety analysis is mandatory before its construction. Concern for safety problem of accident of fusion/fission reactor has been growing after Fukushima accident which is severe accident from unexpected disaster. To model the primary heat transfer system, in this study, MARS-KS thermal hydraulic analysis is referred. Lee et al. and Kim et al. conducted thermal hydraulic analysis using MARS-KS and multiple module simulation to deal with the phenomena of first wall corrosion for each plasma pulse. This study shows the relationship between vacuum vessel rupture area and source term leakage after hydrogen explosion. For the conservative study, first wall heating is not terminated because the heating inside the vacuum vessel increase the pressure inside VV. Pressurizer, steam generator and turbine is not damaged. 6.69 kg of tritiated water (HTO) and 1 ton of dust is modeled which is ITER guideline. The entire system of K-DEMO is smaller than that of ITER. For this reason, lots of aerosol is release into environment although the safety system like DS is maintained. This result shows that the safety system of K-DEMO should use much more safety system.

  2. Updated conceptual design of helium cooling ceramic blanket for HCCB-DEMO

    Wang, Suhao [University of Science and Technology of China, Hefei, Anhui (China); Southwestern Institute of Physics, Chengdu, Sichuan (China); Cao, Qixiang; Wu, Xinghua; Wang, Xiaoyu; Zhang, Guoshu [Southwestern Institute of Physics, Chengdu, Sichuan (China); Feng, Kaiming, E-mail: fengkm@swip.ac.cn [Southwestern Institute of Physics, Chengdu, Sichuan (China)

    2016-11-15

    Highlights: • An updated design of Helium Cooled Ceramic breeder Blanket (HCCB) for HCCB-DEMO is proposed in this paper. • The Breeder Unit is transformed to TBM-like sub-modules, with double “banana” shape tritium breeder. Each sub-module is inserted in space formed by Stiffen Grids (SGs). • The performance analysis is performed based on the R&D development of material, fabrication technology and safety assessment in CN ITER TBM program. • Hot spots will be located at the FW bend side. - Abstract: The basic definition of the HCCB-DEMO plant and preliminary blanket designed by Southwestern Institution of Physics was proposed in 2009. The DEMO fusion power is 2550 MW and electric power is 800 MW. Based on development of R&D in breeding blanket, a conceptual design of helium cooled blanket with ceramic breeder in HCCB-DEMO was presented. The main design features of the HCCB-DEMO blanket were: (1) CLF-1 structure materials, Be multiplier and Li{sub 4}SiO{sub 4} breeder; (2) neutronic wall load is 2.3 MW/m{sup 2} and surface heat flux is 0.43 MW/m{sup 2} (2) TBR ≈ 1.15; (3) geometry of breeding units is ITER TBM-like segmentation; (4)Pressure of helium is 8 MPa and inlet/outlet temperature is 300/500 °C. On the basis of these design, some important analytical results are presented in aspects of (i) neutronic behavior of the blanket; (ii) design of 3D structure and thermal-hydraulic lay-out for breeding blanket module; (iii) structural-mechanical behavior of the blanket under pressurization. All of these assessments proved current stucture fulfill the design requirements.

  3. European DEMO divertor target: Operational requirements and material-design interface

    J.H. You

    2016-12-01

    Full Text Available Recently, an integrated program of conceptual design activities for the European DEMO reactor was launched in the framework of the EUROfusion Consortium, where reliable power handling capability was identified as one of the most critical scientific as well as technological challenges for a DEMO reactor. The divertor is the key in-vessel plasma-facing component being in charge of power exhaust and removal of impurity particles. The DEMO divertor target will have to withstand extreme thermal loads where the local peak heat flux is expected to reach up to 20 MW/m2 during slow transient events in DEMO. To assure sufficient heat removal capability of the divertor target against normal and transient operational scenarios under expected cumulative neutron dose of up to 13 dpa is one of the fundamental engineering challenges imposed on target design. To develop the design of the DEMO divertor and related technologies, an R&D work package ‘Divertor’ has been set up in this consortium. The subproject ‘Target Development’ is devoted to the development of the conceptual design and the core technologies of the plasma-facing target. Devising and implementing novel structural heat sink materials (e.g. W/Cu composites to advanced target design concepts is one of the major objectives of this subproject. In this paper, the underlying design requirements imposed by the envisaged power exhaust goal and the prominent material-design interface issues are discussed. In addition, the candidate design concepts being currently considered are presented together with the related material issues. Finally, the first results achieved so far are presented.

  4. Self-Compacting Concrete

    Okamura, Hajime; Ouchi, Masahiro

    2003-01-01

    Self-compacting concrete was first developed in 1988 to achieve durable concrete structures. Since then, various investigations have been carried out and this type of concrete has been used in practical structures in Japan, mainly by large construction companies. Investigations for establishing a rational mix-design method and self-compactability testing methods have been carried out from the viewpoint of making self-compacting concrete a standard concrete.

  5. Compact Polarimetry Potentials

    Truong-Loi, My-Linh; Dubois-Fernandez, Pascale; Pottier, Eric

    2011-01-01

    The goal of this study is to show the potential of a compact-pol SAR system for vegetation applications. Compact-pol concept has been suggested to minimize the system design while maximize the information and is declined as the ?/4, ?/2 and hybrid modes. In this paper, the applications such as biomass and vegetation height estimates are first presented, then, the equivalence between compact-pol data simulated from full-pol data and compact-pol data processed from raw data as such is shown. Finally, a calibration procedure using external targets is proposed.

  6. Pharmaceutical powder compaction technology

    Çelik, Metin

    2011-01-01

    "Revised to reflect modern pharmaceutical compacting techniques, this Second Edition guides pharmaceutical engineers, formulation scientists, and product development and quality assurance personnel...

  7. Compact Antenna Range

    Federal Laboratory Consortium — Facility consists of a folded compact antenna range including a computer controlled three axis position table, parabolic reflector and RF sources for the measurement...

  8. First disruption studies and simulations in view of the development of the DEMO Physics Basis

    Ramogida, G., E-mail: giuseppe.ramogida@enea.it [ENEA for EUROfusion, via E. Fermi 45, 00044 Frascati, Roma (Italy); Maddaluno, G. [ENEA for EUROfusion, via E. Fermi 45, 00044 Frascati, Roma (Italy); Villone, F. [University of Cassino Consorzio CREATE, Cassino (Italy); Albanese, R. [University Federico II Consorzio CREATE, Naples (Italy); Barbato, L. [University of Cassino Consorzio CREATE, Cassino (Italy); Crisanti, F. [ENEA for EUROfusion, via E. Fermi 45, 00044 Frascati, Roma (Italy); Mastrostefano, S. [University of Cassino Consorzio CREATE, Cassino (Italy); Mazzuca, R. [ENEA for EUROfusion, via E. Fermi 45, 00044 Frascati, Roma (Italy); Palmaccio, R. [University of Cassino Consorzio CREATE, Cassino (Italy); Rubinacci, G.; Ventre, S. [University Federico II Consorzio CREATE, Naples (Italy); Wenninger, R. [IPP, Garching (Germany); EFDA, Garching (Germany)

    2015-10-15

    Highlights: • The prediction of disruption features and loads is essential in the design of DEMO. • Different disruptions need to be simulated to evaluate the EM and thermal loads. • Extrapolation of the thermal quench duration to DEMO gives values from 0.8 to 1.1 ms. • Extrapolation of the current quench duration to DEMO gives values from 47 to 107 ms. • First CarMa0NL simulations points out the effect of large 3D conductive structures. - Abstract: In the development of the DEMO Physics Basis an important role is played by the prediction of the plasma disruption features and by the evaluation of the electro-magnetic (EM) and thermal loads associated with these events. Indeed, the kind and number of foreseen plasma disruptions drive the development of the DEMO operation scenarios and the design of vessel and in-vessel components. To characterize a plausible macroscopic plasma dynamics during these events, we will carry out an extrapolation from present-day machines of the main parameters characterizing the disruptions: thermal and current quench time, evolution of plasma current, β and l{sub i}, safety factor limits, halo current fraction and width, radiated heat fraction. In particular, we will focus on extrapolations for the thermal and current quench characteristic times, due to their importance for the subsequent simulations aimed at the evaluation of the EM and thermal loads. The different options for DEMO design will be taken into account and the possible range of variation of the parameters will be estimated. The 2D axysimmetric MAXFEA and the 3D CarMa0NL codes will be used to evaluate the effects of the induced currents and the EM loads during a disruptive event and to analyze the various design options obtained by the PROCESS code. The results of these simulations, modeled as worst expected events, will be used as input for the system level analysis and design of the vessel and relevant in-vessel components. First simulations with CarMa0NL code

  9. DEVELOPMENT AND ASSESSMENT OF A SCORE™ DEMO2.1 THERMO-ACOUSTIC ENGINE

    BAIMAN CHEN

    2013-04-01

    Full Text Available The early low-cost, wood burning Thermo-Acoustic Engine (TAE known as Demo2.0-build-1 was developed by SCORE™ at the UK Centre and was capable of achieving 22.7 Watts of electricity. This prototype was limited to an operating temperature of about 300oC and due to excessive leaks could not operate continuously above ambient pressure. To absorb a thermal heat input of 4.4 kW from the burning wood so as to fulfil the required acoustic power, the Hot Heat Exchanger (HHX requires heating to the highest possible temperature. Therefore, a corrugated stainless steel plate HHX design that maximises heating surface area was adopted to the current Demo2 TAE design. In addition, the system is often pressurised to achieve higher acoustic intensity. Rigorous sealing of the system at high temperature is also required. A Demo2.1 TAE design based on the Demo2 TAE design and its prototype which is developed recently by the SCORE™ Centre in Malaysia was successfully constructed and well integrated with the stove. During the early construction and assembly process, fabrication difficulties and serious leak problems around the HHX’s edges were found when the apparatus operated at high temperatures. This is because the uneven geometrical HHX (convolution profile makes it difficult and relatively costly to be sealed. The Demo2.1 TAE is focused on the sealing efficiency and effective manufacturing cost by meantime to allow further modification variation. The design was made to adopt the local manufacturing technologies and materials available or easy to access in Malaysia. It also aims to minimise the parasitic heat losses to lower the system onset temperature. By removing the Linear Alternator and Tuning Volume from the system, preliminary measurements shown that the apparatus was oscillating at the frequency of 70 Hz. A much lower onset temperature was observed at around 144oC for the new configuration when the apparatus was oscillating at approximately 200 Pa

  10. Status on DEMO Helium Cooled Lithium Lead breeding blanket thermo-mechanical analyses

    Aubert, J., E-mail: julien.aubert@cea.fr [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Aiello, G.; Jaboulay, J.-C. [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Kiss, B. [Institute of Nuclear Techniques, Budapest University of Technology and Economics, Budapest (Hungary); Morin, A. [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France)

    2016-11-01

    Highlights: • CEA with the support of Wigner-RCP and IPP-CR, is in charge of the design of the HCLL blanket for DEMO. The DEMO HCLL breeding blanket design capitalizes on the experience acquired on the HCLL Test Blanket Module designed for ITER. Design improvements are being implemented to adapt the design to DEMO specifications and performance objectives. • Thermal and mechanical analyses have been carried out in order to justify the design of the HCLL breeding blanket showing promising results for tie rods modules’ attachments system and relatively good behavior of the box in case of LOCA when comparing to RCC-MRx criteria. • CFD thermal analyses on generic breeding unit have enabled the consolidation of the results obtained with previous FEM design analyses. - Abstract: The EUROfusion Consortium develops a design of a fusion power demonstrator (DEMO) in the framework of the European “Horizon 2020” innovation and research program. One of the key components in the fusion reactor is the breeding blanket surrounding the plasma, ensuring tritium self-sufficiency, heat removal for conversion into electricity, and neutron shielding. The Helium Cooled Lithium Lead (HCLL) blanket is one of the concepts which is investigated for DEMO. It is made of a Eurofer structure and uses the eutectic liquid lithium–lead as tritium breeder and neutron multiplier, and helium gas as coolant. Within the EUROfusion organization, CEA with the support of Wigner-RCP and IPP-CR, is in charge of the design of the HCLL blanket for DEMO. This paper presents the status of the thermal and mechanical analyses carried out on the HCLL breeding blanket in order to justify the design. CFD thermal analyses on generic breeding unit including stiffening plates and cooling plates have been performed with ANSYS in order to consolidate results obtained with previous FEM design analyses. Moreover in order to expand the justification of the HCLL Breeding blanket design, the most loaded area of

  11. Uniaxial backfill block compaction

    Koskinen, V.

    2012-05-01

    The main parts of the project were: to make a literature survey of the previous uniaxial compaction experiments; do uniaxial compaction tests in laboratory scale; and do industrial scale production tests. Object of the project was to sort out the different factors affecting the quality assurance chain of the backfill block uniaxial production and solve a material sticking to mould problem which appeared during manufacturing the blocks of bentonite and cruched rock mixture. The effect of mineralogical and chemical composition on the long term functionality of the backfill was excluded from the project. However, the used smectite-rich clays have been tested for mineralogical consistency. These tests were done in B and Tech OY according their SOPs. The objective of the Laboratory scale tests was to find right material- and compaction parameters for the industrial scale tests. Direct comparison between the laboratory scale tests and industrial scale tests is not possible because the mould geometry and compaction speed has a big influence for the compaction process. For this reason the selected material parameters were also affected by the previous compaction experiments. The industrial scale tests were done in summer of 2010 in southern Sweden. Blocks were done with uniaxial compaction. A 40 tons of the mixture of bentonite and crushed rock blocks and almost 50 tons of Friedland-clay blocks were compacted. (orig.)

  12. Compaction properties of isomalt

    Bolhuis, Gerad K.; Engelhart, Jeffrey J. P.; Eissens, Anko C.

    Although other polyols have been described extensively as filler-binders in direct compaction of tablets, the polyol isomalt is rather unknown as pharmaceutical excipient, in spite of its description in all the main pharmacopoeias. In this paper the compaction properties of different types of

  13. Model Compaction Equation

    The currently proposed model compaction equation was derived from data sourced from the. Niger Delta and it relates porosity to depth for sandstones under hydrostatic pressure condition. The equation is useful in predicting porosity and compaction trend in hydrostatic sands of the. Niger Delta. GEOLOGICAL SETTING OF ...

  14. Stabilization of compactible waste

    Franz, E.M.; Heiser, J.H. III; Colombo, P.

    1990-09-01

    This report summarizes the results of series of experiments performed to determine the feasibility of stabilizing compacted or compactible waste with polymers. The need for this work arose from problems encountered at disposal sites attributed to the instability of this waste in disposal. These studies are part of an experimental program conducted at Brookhaven National Laboratory (BNL) investigating methods for the improved solidification/stabilization of DOE low-level wastes. The approach taken in this study was to perform a series of survey type experiments using various polymerization systems to find the most economical and practical method for further in-depth studies. Compactible dry bulk waste was stabilized with two different monomer systems: styrene-trimethylolpropane trimethacrylate (TMPTMA) and polyester-styrene, in laboratory-scale experiments. Stabilization was accomplished by wetting or soaking compactible waste (before or after compaction) with monomers, which were subsequently polymerized. Three stabilization methods are described. One involves the in-situ treatment of compacted waste with monomers in which a vacuum technique is used to introduce the binder into the waste. The second method involves the alternate placement and compaction of waste and binder into a disposal container. In the third method, the waste is treated before compaction by wetting the waste with the binder using a spraying technique. A series of samples stabilized at various binder-to-waste ratios were evaluated through water immersion and compression testing. Full-scale studies were conducted by stabilizing two 55-gallon drums of real compacted waste. The results of this preliminary study indicate that the integrity of compacted waste forms can be readily improved to ensure their long-term durability in disposal environments. 9 refs., 10 figs., 2 tabs

  15. ECN's torrefaction-based BO2-technology. From pilot to demo

    Kiel, J.H.A. [ECN Biomass, Coal and Environmental Research, Petten (Netherlands)

    2011-02-15

    The contents of this PowerPoint presentation are: Torrefaction design challenges; Initial small-scale R and D; ECN's torrefaction-based BO2-technology; Pilot-scale testing; and Demonstration and market introduction. The conclusions state that Torrefaction potentially allows cost-effective production of 2nd generation biomass pellets from a wide range of biomass/waste feedstock with a high energy efficiency (>90%); Torrefaction pellets show: High energy density, Water resistance, No/Limited biological degradation and heating, Excellent grindability, and Good combustion and gasification properties; Torrefaction is a separate thermal regime and requires dedicated reactor/process design; Torrefaction development is in pilot/demo-phase and shows strong market pull for torrefaction plants and torrefaction pellets; For ECN's BO2-technology a demo-plant is in preparation and industrial partnership for world-wide market introduction is nearly established.

  16. The German DEMO working group. Perspectives of a fusion power plant

    Hesch, Klaus

    2013-01-01

    Fusion development has many different challenges in the areas of plasma physics, fusion technologies, materials development and plasma wall interaction. For making fusion power a reality, a coherent approach is necessary, interlinking the different areas of work. To this end, the German fusion program started in 2010 the German DEMO Working Group, bringing together high-level experts from all the different fields, from the 3 German fusion centers Max-Planck-Institut fuer Plasmaphysik (IPP), Karlsruher Institut fuer Technologie (KIT) and Forschungszentrum Juelich (FZJ). An encompassing view of what will be needed with high priority, in plasma physics, in fusion technology and in the interrelation of the fields, to make fusion energy real, has been elaborated, and is presented here in a condensed way. On this basis, the 3 German fusion centers now are composing their work program, towards a fusion demonstration reactor DEMO. (orig.)

  17. A high frequency, high power CARM proposal for the DEMO ECRH system

    Mirizzi, Francesco; Spassovsky, Ivan; Ceccuzzi, Silvio; Dattoli, Giuseppe; Di Palma, Emanuele; Doria, Andrea; Gallerano, Gianpiero; Lampasi, Alessandro; Maffia, Giuseppe; Ravera, GianLuca; Sabia, Elio; Tuccillo, Angelo Antonio; Zito, Pietro

    2015-01-01

    Highlights: • ECRH system for DEMO. • Cyclotron Auto-Resonance Maser (CARM) devices. • Relativistic electron beams. • Bragg reflectors. • High voltage pulse modulators. - Abstract: ECRH&CD systems are extensively used on tokamak plasmas due to their capability of highly tailored power deposition, allowing very localised heating and non-inductive current drive, useful for MHD and profiles control. The high electron temperatures expected in DEMO will require ECRH systems with operating frequency in the 200–300 GHz range, equipped with a reasonable number of high power (P ≥ 1 MW) CW RF sources, for allowing central RF power deposition. In this frame the ENEA Fusion Department (Frascati) is coordinating a task force aimed at the study and realisation of a suitable high power, high frequency reliable source.

  18. A high frequency, high power CARM proposal for the DEMO ECRH system

    Mirizzi, Francesco, E-mail: francesco.mirizzi@enea.it [Consorzio CREATE, Via Claudio 21, I-80125 Napoli (Italy); Spassovsky, Ivan [Unità Tecnica Applicazioni delle Radiazioni – ENEA, C.R. Frascati, via E. Fermi 45, I-00044 Frascati (Italy); Ceccuzzi, Silvio [Unità Tecnica Fusione – ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati, Roma (Italy); Dattoli, Giuseppe; Di Palma, Emanuele; Doria, Andrea; Gallerano, Gianpiero [Unità Tecnica Applicazioni delle Radiazioni – ENEA, C.R. Frascati, via E. Fermi 45, I-00044 Frascati (Italy); Lampasi, Alessandro; Maffia, Giuseppe; Ravera, GianLuca [Unità Tecnica Fusione – ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati, Roma (Italy); Sabia, Elio [Unità Tecnica Applicazioni delle Radiazioni – ENEA, C.R. Frascati, via E. Fermi 45, I-00044 Frascati (Italy); Tuccillo, Angelo Antonio; Zito, Pietro [Unità Tecnica Fusione – ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati, Roma (Italy)

    2015-10-15

    Highlights: • ECRH system for DEMO. • Cyclotron Auto-Resonance Maser (CARM) devices. • Relativistic electron beams. • Bragg reflectors. • High voltage pulse modulators. - Abstract: ECRH&CD systems are extensively used on tokamak plasmas due to their capability of highly tailored power deposition, allowing very localised heating and non-inductive current drive, useful for MHD and profiles control. The high electron temperatures expected in DEMO will require ECRH systems with operating frequency in the 200–300 GHz range, equipped with a reasonable number of high power (P ≥ 1 MW) CW RF sources, for allowing central RF power deposition. In this frame the ENEA Fusion Department (Frascati) is coordinating a task force aimed at the study and realisation of a suitable high power, high frequency reliable source.

  19. Applicability of tungsten/EUROFER blanket module for the DEMO first wall

    Igitkhanov, Yu.; Bazylev, B.; Landman, I.; Boccaccini, L.

    2013-01-01

    In this paper we analyse a sandwich-type blanket configuration of W/EUROFER for DEMO first wall under steady-state normal operation and off-normal conditions, such as vertical displacements and runaway electrons. The heat deposition and consequent erosion of the tungsten armour is modelled under condition of helium cooling of the first wall blanket module and by taking into account the conversion of the magnetic energy stored in the runaway electron current into heat through the ohmic dissipation of the return current induced in the metallic armour structure. It is shown that under steady-state DEMO operation the first wall sandwich type module will tolerate heat loads up to ∼14 MW/m 2 . It will also sustain the off-normal events, apart from the hot vertical displacement events, which will melt the tungsten armour surface

  20. Thermal-hydraulic investigations on the CEA-ENEA DEMO relevant helium cooled poloidal blanket

    Dell'Orco, G.; Polazzi, G.; Vallette, F.; Proust, E.; Eid, M.

    1994-01-01

    The CEA-ENEA design of an Helium Cooled Solid Breeder Blanket (HCSBB) for the DEMO reactor, with a breeder in tube (BIT) poloidal arrangement, is based on the use of lithium ceramic pellets, the ENEA γ-LiAlO 2 or the CEA Li 2 ZrO 3 . Due to the geometry of the DEMO reactor plasma chamber, these breeder bundles are adapted to the Vacuum Vessel with a strong poloidal curvature. This curvature influences the thermal-hydraulic behaviour of the coolant flowing inside the bundle. The paper presents the CEA-ENEA first results of the experimental and theoretical programme, aiming at optimizing the breeder module thermal hydraulic design. (author) 6 refs.; 7 figs.; 1 tab

  1. Thermal-hydraulic analysis of LTS cables for the DEMO TF coil using simplified models

    Lewandowska Monika

    2017-03-01

    Full Text Available The conceptual design activities for the DEMOnstration reactor (DEMO – the prototype fusion power plant – are conducted in Europe by the EUROfusion Consortium. In 2015, three design concepts of the DEMO toroidal field (TF coil were proposed by Swiss Plasma Center (EPFL-SPC, PSI Villigen, Italian National Agency for New Technologies (ENEA Frascati, and Atomic Energy and Alternative Energies Commission (CEA Cadarache. The proposed conductor designs were subjected to complete mechanical, electromagnetic, and thermal-hydraulic analyses. The present study is focused on the thermal-hydraulic analysis of the candidate conductor designs using simplified models. It includes (a hydraulic analysis, (b heat removal analysis, and (c assessment of the maximum temperature and the maximum pressure in each conductor during quench. The performed analysis, aimed at verification whether the proposed design concepts fulfil the established acceptance criteria, provides the information for further improvements of the coil and conductors design.

  2. Initial results of NEXT-DEMO, a large-scale prototype of the NEXT-100 experiment

    Álvarez, V; Cárcel, S; Cervera, A; Díaz, J; Ferrario, P; Gil, A; Borges, F I G; Conde, C A N; Dias, T H V T; Fernandes, L M P; Freitas, E D C; Castel, J; Cebrián, S; Dafni, T; Egorov, M; Gehman, V M; Goldschmidt, A; Esteve, R; Evtoukhovitch, P; Ferreira, A L

    2013-01-01

    NEXT-DEMO is a large-scale prototype of the NEXT-100 detector, an electroluminescent time projection chamber that will search for the neutrinoless double beta decay of XE using 100–150 kg of enriched xenon gas. NEXT-DEMO was built to prove the expected performance of NEXT-100, namely, energy resolution better than 1% FWHM at 2.5 MeV and event topological reconstruction. In this paper we describe the prototype and its initial results. A resolution of 1.75% FWHM at 511 keV (which extrapolates to 0.8% FWHM at 2.5 MeV) was obtained at 10 bar pressure using a gamma-ray calibration source. Also, a basic study of the event topology along the longitudinal coordinate is presented, proving that it is possible to identify the distinct dE/dx of electron tracks in high-pressure xenon using an electroluminescence TPC.

  3. Applicability of tungsten/EUROFER blanket module for the DEMO first wall

    Igitkhanov, Yu.; Bazylev, B.; Landman, I.; Boccaccini, L.

    2013-07-01

    In this paper we analyse a sandwich-type blanket configuration of W/EUROFER for DEMO first wall under steady-state normal operation and off-normal conditions, such as vertical displacements and runaway electrons. The heat deposition and consequent erosion of the tungsten armour is modelled under condition of helium cooling of the first wall blanket module and by taking into account the conversion of the magnetic energy stored in the runaway electron current into heat through the ohmic dissipation of the return current induced in the metallic armour structure. It is shown that under steady-state DEMO operation the first wall sandwich type module will tolerate heat loads up to ˜14 MW/m2. It will also sustain the off-normal events, apart from the hot vertical displacement events, which will melt the tungsten armour surface.

  4. Applicability of tungsten/EUROFER blanket module for the DEMO first wall

    Igitkhanov, Yu., E-mail: juri.igitkhanov@lhm.fzk.de [Karlsruhe Institute of Technology, IHM, Karlsruhe (Germany); Bazylev, B.; Landman, I. [Karlsruhe Institute of Technology, IHM, Karlsruhe (Germany); Boccaccini, L. [Karlsruhe Institute of Technology, INR, Karlsruhe (Germany)

    2013-07-15

    In this paper we analyse a sandwich-type blanket configuration of W/EUROFER for DEMO first wall under steady-state normal operation and off-normal conditions, such as vertical displacements and runaway electrons. The heat deposition and consequent erosion of the tungsten armour is modelled under condition of helium cooling of the first wall blanket module and by taking into account the conversion of the magnetic energy stored in the runaway electron current into heat through the ohmic dissipation of the return current induced in the metallic armour structure. It is shown that under steady-state DEMO operation the first wall sandwich type module will tolerate heat loads up to ∼14 MW/m{sup 2}. It will also sustain the off-normal events, apart from the hot vertical displacement events, which will melt the tungsten armour surface.

  5. Diagnostics and required R and D for control of DEMO grade plasmas

    Park, Hyeon K., E-mail: hyeonpark@unist.ac.kr [Fusion Plasma Stability and Confinement Research Center, UNIST, 50 Unist-gil, Ulju-gun, Ulsan (Korea, Republic of)

    2014-08-21

    Even if the diagnostics of ITER performs as expected, installation and operation of the diagnostic systems in Demo device will be much harsher than those of the present ITER device. In order to operate the Demo grade plasmas, which may have a higher beta limit, safely with very limited number of simple diagnostic system, it requires a well defined predictable plasma modelling in conjunction with the reliable control system for burn control and potential harmful instabilities. Development of such modelling in ITER is too risky and the logical choice would be utilization of the present day steady state capable devices such as KSTAR and EAST. In order to fulfill this mission, sophisticated diagnostic systems such as 2D/3D imaging systems can validate the physics in the theoretical modeling and challenge the predictable capability.

  6. Diagnostics and control for the steady state and pulsed tokamak DEMO

    Orsitto, F.P.; Villari, R.; Moro, F.; Todd, T.N.; Lilley, S.; Jenkins, I.; Felton, R.; Biel, W.; Silva, A.; Scholz, M.; Rzadkiewicz, J.; Ďuran, Ivan; Tardocchi, M.; Gorini, G.; Morlock, C.; Federici, G.; Litnovsky, A.

    2016-01-01

    Roč. 56, č. 2 (2016), č. článku 026009. ISSN 0029-5515 Institutional support: RVO:61389021 Keywords : measurement systems, fusion reactor, fusion plasma diagnostics * fusion reactor * fusion plasma diagnostics * DEMO * Hall sensors * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 3.307, year: 2016 http://iopscience.iop.org/article/10.1088/0029-5515/56/2/026009

  7. Conceptual design studies for the European DEMO divertor: Rationale and first results

    You, J.H.; Mazzone, F.; Visca, E.; Bachmann, Ch.; Autissier, E.; Barrett, T.; Cocilovo, V.; Crescenzi, F.; Domalapally, P.K.; Dongiovanni, D.; Entler, Slavomír; Federici, G.; Frosi, P.; Fursdon, M.; Greuner, H.; Hancock, D.; Marzullo, D.; McIntosh, S.; Müller, A.V.; Porfiri, M.T.; Ramogida, G.; Reiser, J.; Richou, M.; Rieth, M.; Rydzy, A.; Villari, R.; Widak, V.

    109-111, November (2016), s. 1598-1603 ISSN 0920-3796. [International Symposium on Fusion Nuclear Technology (ISFNT-12)/12./. Jeju, 14.09.2015-18.09.2015] EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : DEMO * Tokamak * Divertor * Plasma-facing component * Conceptual design * Eurofusiona Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S0920379615303331

  8. Study of the cooling systems with S-CO2 for the DEMO fusion power reactor.

    Veselý, L.; Dostál, V.; Entler, Slavomír

    2017-01-01

    Roč. 124, November (2017), s. 244-247 ISSN 0920-3796. [SOFT 2016: Symposium on Fusion Technology /29./. Prague, 05.09.2016-09.09.2016] Institutional support: RVO:61389021 Keywords : DEMO * Cooling * Energy conversion * Thermal cycle * Carbon dioxide * SCO2a Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S0920379617305719

  9. Overview of physics results from MAST towards ITER/DEMO and the MAST Upgrade

    Meyer, H.; Abel, I.G.; Akers, R.J.; Allan, A.; Allan, S.Y.; Appel, L.C.; Asunta, O.; Barnes, M.; Barratt, N.C.; Ben Ayed, N.; Bradley, J.W.; Canik, J.; Cahyna, Pavel; Cecconelo, M.; Challis, C.D.; Chapman, I.T.; Ciric, D.; Colyer, G.; Conway, N.J.; Cox, M.; Crowley, B.J.; Cowley, S.C.; Cunningham, G.; Danilov, A.; Darke, A.; De Bock, M.F.M.; De Temmerman, G.; Dendy, R.O.; Denner, P.; Dickinson, D.; Dnestrovsky, A.Y.; Dnestrovsky, Y.; Driscoll, M.D.; Dudson, B.; Dunai, D.; Dunstan, M.; Dura, P.; Elmore, S.; Field, A.R.; Fishpool, G.; Freethy, S.; Fundameski, W.; Garzotti, L.; Ghim, Y.C.; Gibson, K.J.; Gryaznevich, M.P.; Harrison, J.; Havlíčková, E.; Hawkes, N.C.; Heidbrink, W.W.; Hender, T.C.; Highcock, E.; Higgins, D.; Hill, P.; Hnat, B.; Hole, M.J.; Horáček, Jan; Howell, D.F.; Imada, K.; Jones, O.; Kaveeva, E.; Keeling, D.; Kirk, A.; Kočan, M.; Lake, R.J.; Lehnen, M.; Leggate, H.J.; Liang, Y.; Lilley, M.K.; Lisgo, S.W.; Liu, Y.Q.; Lloyd, B.; Maddison, G.P.; Mailloux, J.; Martin, R.; McArdle, G.J.; McClements, K.G.; McMillan, B.; Michael, C.; Militello, F.; Molchanov, P.; Mordijck, S.; Morgan, T.; Morris, A.W.; Muir, D.G.; Nardon, E.; Naulin, V.; Naylor, G.; Nielsen, A.H.; O’Brien, M.R.; O’Gorman, T.; Pamela, S.; Parra, F.I.; Patel, A.; Pinches, S.D.; Price, M.N.; Roach, C.M.; Robinson, J.R.; Romanelli, M.; Rozhansky, V.; Saarelma, S.; Sangaroon, S.; Saveliev, A.; Scannell, R.; Seidl, J.; Sharapov, S.E.; Schekochihin, A.A.; Shevchenko, V.; Shibaev, S.; Stork, D.; Storrs, J.; Sykes, A.; Tallents, G. J.; Tamain, P.; Taylor, D.; Temple, D.; Thomas-Davies, N.; Thornton, A.; Turnyanskiy, M.R.; Valovič, M.; Vann, R.G.L.; Verwichte, E.; Voskoboynikov, P.; Voss, G.; Warder, S.E.V.; Wilson, H. R.; Wodniak, I.; Zoletnik, S.; Zagórski, R.

    2013-01-01

    Roč. 53, č. 10 (2013), s. 104008-104008 ISSN 0029-5515. [IAEA Fusion Energy Conference/24./. San Diego, 08.10.2012-13.10.2012] Institutional support: RVO:61389021 Keywords : ITER * DEMO * MAST * spherical tokamak * JET Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.243, year: 2013 http://iopscience.iop.org/0029-5515/53/10/104008/pdf/0029-5515_53_10_104008.pdf

  10. Non-linear failure analysis of HCPB blanket for DEMO taking into account high dose irradiation

    Aktaa, J., E-mail: jarir.aktaa@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Kecskés, S.; Pereslavtsev, P.; Fischer, U.; Boccaccini, L.V. [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: • First non-linear structural analysis for the European Helium Cooled Pebble Bed Blanket Module taking into account high dose irradiation. • Most critical areas were identified and analyzed with regard to the effect of irradiation on predicted damage at these areas. • Despite the extensive computing time 100 cycles were simulated by using the sub-modelling technique investigating damage at most critical area. • The results show a positive effect of irradiation on calculated damage which is mainly attributed to the irradiation induced hardening. - Abstract: For the European helium cooled pebble bed (HCPB) blanket of DEMO the reduced activation ferritic martensitic steel EUROFER has been selected as structural material. During operation the HCPB blanket will be subjected to complex thermo-mechanical loadings and high irradiation doses. Taking into account the material and structural behaviour under these conditions is a precondition for a reliable blanket design. For considering high dose irradiation in structural analysis of the DEMO blanket, the coupled deformation damage model, extended recently taking into account the influence of high dose irradiation on the material behaviour of EUROFER and implemented in the finite element code ABAQUS, has been used. Non-linear finite element (FE) simulations of the DEMO HCPB blanket have been performed considering the design of the HCPB Test Blanket Module (TBM) as reference and the thermal and mechanical boundary conditions of previous analyses. The irradiation dose rate required at each position in the structure as an additional loading parameter is estimated by extrapolating the results available for the TBM in ITER scaling the value calculated in neutronics and activation analysis for ITER boundary conditions to the DEMO boundary conditions. The results of the FE simulations are evaluated considering damage at most critical highly loaded areas of the structure and discussed with regard to the impact of

  11. Non-linear failure analysis of HCPB blanket for DEMO taking into account high dose irradiation

    Aktaa, J.; Kecskés, S.; Pereslavtsev, P.; Fischer, U.; Boccaccini, L.V.

    2014-01-01

    Highlights: • First non-linear structural analysis for the European Helium Cooled Pebble Bed Blanket Module taking into account high dose irradiation. • Most critical areas were identified and analyzed with regard to the effect of irradiation on predicted damage at these areas. • Despite the extensive computing time 100 cycles were simulated by using the sub-modelling technique investigating damage at most critical area. • The results show a positive effect of irradiation on calculated damage which is mainly attributed to the irradiation induced hardening. - Abstract: For the European helium cooled pebble bed (HCPB) blanket of DEMO the reduced activation ferritic martensitic steel EUROFER has been selected as structural material. During operation the HCPB blanket will be subjected to complex thermo-mechanical loadings and high irradiation doses. Taking into account the material and structural behaviour under these conditions is a precondition for a reliable blanket design. For considering high dose irradiation in structural analysis of the DEMO blanket, the coupled deformation damage model, extended recently taking into account the influence of high dose irradiation on the material behaviour of EUROFER and implemented in the finite element code ABAQUS, has been used. Non-linear finite element (FE) simulations of the DEMO HCPB blanket have been performed considering the design of the HCPB Test Blanket Module (TBM) as reference and the thermal and mechanical boundary conditions of previous analyses. The irradiation dose rate required at each position in the structure as an additional loading parameter is estimated by extrapolating the results available for the TBM in ITER scaling the value calculated in neutronics and activation analysis for ITER boundary conditions to the DEMO boundary conditions. The results of the FE simulations are evaluated considering damage at most critical highly loaded areas of the structure and discussed with regard to the impact of

  12. Preliminary analysis of the efficiency of non-standard divertor configurations in DEMO

    F. Subba

    2017-08-01

    Full Text Available The standard Single Null (SN divertor is currently expected to be installed in DEMO. However, a number of alternative configurations are being evaluated in parallel as backup solutions, in case the standard divertor does not extrapolate successfully from ITER to a fusion power plant. We used the SOLPS code to produce a preliminary analysis of two such configurations, the X-Divertor (XD and the Super X-Divertor (SX, and compare them to the SN solution. Considering the nominal power flowing into the SOL (PSOL = 150 MW, we estimated the amplitude of the acceptable DEMO operational space. The acceptability criterion was chosen as plasma temperature at the target lower than 5eV, providing low sputtering and at least partial detachment, while the operational space was defined in terms of the electron density at the outboard mid-plane separatrix and of the seeded impurity (Ar only in the present study concentration. It was found that both the XD and the SXD extend the DEMO operational space, although the advantages detected so far are not dramatic. The most promising configuration seems to be the XD, which can produce acceptable target temperatures at moderate outboard mid-plane electron density (nomp=4.5×1019 m−3 and Zeff= 1.3.

  13. Overview of the design approach and prioritization of R&D activities towards an EU DEMO

    Federici, G., E-mail: gianfranco.federici@euro-fusion.org [EUROfusion Consortium, Boltzmannstr. 2, Garching 85748 (Germany); Bachmann, C. [EUROfusion Consortium, Boltzmannstr. 2, Garching 85748 (Germany); Biel, W. [Institute of Energy and Climate Research, Forschungszentrum Jülich GmbH, Jülich (Germany); Department of Applied Physics, Ghent University, Ghent (Belgium); Boccaccini, L. [Karlsruhe Institute of Technology (KIT), Campus Nord Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Cismondi, F.; Ciattaglia, S.; Coleman, M. [EUROfusion Consortium, Boltzmannstr. 2, Garching 85748 (Germany); Day, C. [Karlsruhe Institute of Technology (KIT), Campus Nord Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Diegele, E.; Franke, T. [EUROfusion Consortium, Boltzmannstr. 2, Garching 85748 (Germany); Grattarola, M. [Ansaldo Nucleare, Corso Perrone 25, 16152 Genova (Italy); Hurzlmeier, H. [EUROfusion Consortium, Boltzmannstr. 2, Garching 85748 (Germany); Ibarra, A. [CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain); Loving, A. [CCFE Culham Science Centre, Abingdon OX14-3DB, Oxon (United Kingdom); Maviglia, F.; Meszaros, B.; Morlock, C. [EUROfusion Consortium, Boltzmannstr. 2, Garching 85748 (Germany); Rieth, M. [Karlsruhe Institute of Technology (KIT), Campus Nord Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Shannon, M. [EUROfusion Consortium, Boltzmannstr. 2, Garching 85748 (Germany); Taylor, N. [CCFE Culham Science Centre, Abingdon OX14-3DB, Oxon (United Kingdom); and others

    2016-11-01

    Highlights: • An important objective of the EU fusion roadmap Horizon 2020 is to lay the foundation of a DEMO Fusion Power Reactor to follow ITER. • This paper describes the progress of the DEMO design and R&D activities in Europe in the EUROfusion Consortium. • Focus is on a systems engineering/design integration approach to identify technology & physics R&D requirements and address design challenges. • Preliminary design choices/sensitivity studies to explore the design space and identify/select attractive design points are described. • Initial results of work conducted by distributed project teams involving EU labs, universities, and industries in Europe are presented. - Abstract: This paper describes the progress of the DEMO design and R&D activities in Europe. The focus is on a systems engineering and design integration approach, which is recognized to be essential from an early stage to identify and address the engineering and operational challenges, and the requirements for technology and physics R&D. We present some of the preliminary design choices/sensitivity studies to explore and narrow down the design space and identify/select attractive design points. We also discuss some of the initial results of work being executed in the EUROfusion Consortium by a geographically distributed project team involving many EU laboratories, universities, and industries in Europe.

  14. On the EU approach for DEMO architecture exploration and dealing with uncertainties

    Coleman, M., E-mail: matti.coleman@euro-fusion.org [EUROfusion Consortium, Boltzmannstraße 2, 85748 Garching (Germany); CCFE Fusion Association, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Maviglia, F.; Bachmann, C. [EUROfusion Consortium, Boltzmannstraße 2, 85748 Garching (Germany); Anthony, J. [CCFE Fusion Association, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Federici, G. [EUROfusion Consortium, Boltzmannstraße 2, 85748 Garching (Germany); Shannon, M. [EUROfusion Consortium, Boltzmannstraße 2, 85748 Garching (Germany); CCFE Fusion Association, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Wenninger, R. [EUROfusion Consortium, Boltzmannstraße 2, 85748 Garching (Germany); Max-Planck-Institut für Plasmaphysik, 85748 Garching (Germany)

    2016-11-01

    Highlights: • The issue of epistemic uncertainties in the DEMO design basis is described. • An approach to tackle uncertainty by investigating plant architectures is proposed. • The first wall heat load uncertainty is addressed following the proposed approach. - Abstract: One of the difficulties inherent in designing a future fusion reactor is dealing with uncertainty. As the major step between ITER and the commercial exploitation of nuclear fusion energy, DEMO will have to address many challenges – the natures of which are still not fully known. Unlike fission reactors, fusion reactors suffer from the intrinsic complexity of the tokamak (numerous interdependent system parameters) and from the dependence of plasma physics on scale – prohibiting design exploration founded on incremental progression and small-scale experimentation. For DEMO, this means that significant technical uncertainties will exist for some time to come, and a systems engineering design exploration approach must be developed to explore the reactor architecture when faced with these uncertainties. Important uncertainties in the context of fusion reactor design are discussed and a strategy for dealing with these is presented, treating the uncertainty in the first wall loads as an example.

  15. Transient analyses on the cooling channels of the DEMO HCPB blanket concept under accidental conditions

    Chen, Yuming; Ghidersa, Bradut-Eugen; Jin, Xue Zhou

    2016-01-01

    Highlights: • This paper presents transient CFD analyses on the cooling channels of the DEMO HCPB FW for accidental scenarios LOCA and LOFA. • In both LOCA & LOFA, the wall temperature increases quickly to an unacceptable level within seconds. • If the coolant flow rate is maintained at a half of nominal value in case of LOFA (partial LOFA), the wall temperature rises much slower, but will still leads to a damage of structure within minutes. • The simulated heat transfer coefficients were compared with empirical correlations. - Abstract: Helium Cooled Pebble Bed (HCPB) blanket concept is one of the DEMO (Demonstration Power Plant) blanket concepts running for the final DEMO design selection. In this paper, transient analyses on the cooling channels of the FW are carried out by means of CFD simulations for the selected accidental scenarios loss-of-coolant-accident (LOCA) and loss-of-flow-accident (LOFA). ANSYS-CFX is used for the simulations. The simulation results help to understand how fast the temperature of the FW can increase and what is the time window that is available until the temperature of the structural material reaches the design limit in order to be able to define a suitable protection strategy for the system. In view of later developments of the models, the heat transfer coefficients calculated with CFD are compared with the values predicted by two widely used correlations for turbulent pipe flows.

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

    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.

  17. Use cases and DEMO: aligning functional features of ICT-infrastructure to business processes.

    Maij, E; Toussaint, P J; Kalshoven, M; Poerschke, M; Zwetsloot-Schonk, J H M

    2002-11-12

    The proper alignment of functional features of the ICT-infrastructure to business processes is a major challenge in health care organisations. This alignment takes into account that the organisational structure not only shapes the ICT-infrastructure, but that the inverse also holds. To solve the alignment problem, relevant features of the ICT-infrastructure should be derived from the organisational structure and the influence of this envisaged ICT to the work practices should be pointed out. The objective of our study was to develop a method to solve this alignment problem. In a previous study we demonstrated the appropriateness of the business process modelling methodology Dynamic Essential Modelling of Organizations (DEMO). A proven and widely used modelling language for expressing functional features is Unified Modelling Language (UML). In the context of a specific case study at the University Medical Centre Utrecht in the Netherlands we investigated if the combined use of DEMO and UML could solve the alignment problem. The study demonstrated that the DEMO models were suited as a starting point in deriving system functionality by using the use case concept of UML. Further, the case study demonstrated that in using this approach for the alignment problem, insight is gained into the mutual influence of ICT-infrastructure and organisation structure: (a) specification of independent, re-usable components-as a set of related functionalities-is realised, and (b) a helpful representation of the current and future work practice is provided for in relation to the envisaged ICT support.

  18. The Studsvik power transient programs Demo-Ramp II and Trans-Ramp I

    Bergenlid, U.; Lysell, G.; Mogard, H.; Roennberg, G.

    1984-01-01

    The Studsvik Demo-Ramp II och Trans-Ramp I are internationally sponsored research programs. The main objectives are similar in both programs: to study the effects on the PCI/SCC failure process of short time power transients, above the failure threshold where cladding failure (FP leakage) is expected to occur after a sufficient hold time. Demo-Ramp II is completed, whereas, at present, Trans-Ramp I is in progress. Test fuel rods of standard BWR design are used. The fuel rods have been base-irradiated in a power reactor (burn-up in the range 18 to 29 MWd/kg U) and subsequently ramp tested in the R2 reactor. Extensive examinations of the rods have been performed. In the Demo-Ramp II program a large number of incipient cladding cracks were observed to be formed more rapidly than expected, based on previous knowledge. It was possible to operate one rod for a very short time above the failure threshold without SCC crack formation. One objective of the Trans-Ramp I program is to define more closely the power-time region above the failure threshold where the rods remain intact after power transients. (author)

  19. Transient analyses on the cooling channels of the DEMO HCPB blanket concept under accidental conditions

    Chen, Yuming, E-mail: Yuming.chen@kit.edu; Ghidersa, Bradut-Eugen; Jin, Xue Zhou

    2016-11-01

    Highlights: • This paper presents transient CFD analyses on the cooling channels of the DEMO HCPB FW for accidental scenarios LOCA and LOFA. • In both LOCA & LOFA, the wall temperature increases quickly to an unacceptable level within seconds. • If the coolant flow rate is maintained at a half of nominal value in case of LOFA (partial LOFA), the wall temperature rises much slower, but will still leads to a damage of structure within minutes. • The simulated heat transfer coefficients were compared with empirical correlations. - Abstract: Helium Cooled Pebble Bed (HCPB) blanket concept is one of the DEMO (Demonstration Power Plant) blanket concepts running for the final DEMO design selection. In this paper, transient analyses on the cooling channels of the FW are carried out by means of CFD simulations for the selected accidental scenarios loss-of-coolant-accident (LOCA) and loss-of-flow-accident (LOFA). ANSYS-CFX is used for the simulations. The simulation results help to understand how fast the temperature of the FW can increase and what is the time window that is available until the temperature of the structural material reaches the design limit in order to be able to define a suitable protection strategy for the system. In view of later developments of the models, the heat transfer coefficients calculated with CFD are compared with the values predicted by two widely used correlations for turbulent pipe flows.

  20. Mouse Embryo Compaction.

    White, M D; Bissiere, S; Alvarez, Y D; Plachta, N

    2016-01-01

    Compaction is a critical first morphological event in the preimplantation development of the mammalian embryo. Characterized by the transformation of the embryo from a loose cluster of spherical cells into a tightly packed mass, compaction is a key step in the establishment of the first tissue-like structures of the embryo. Although early investigation of the mechanisms driving compaction implicated changes in cell-cell adhesion, recent work has identified essential roles for cortical tension and a compaction-specific class of filopodia. During the transition from 8 to 16 cells, as the embryo is compacting, it must also make fundamental decisions regarding cell position, polarity, and fate. Understanding how these and other processes are integrated with compaction requires further investigation. Emerging imaging-based techniques that enable quantitative analysis from the level of cell-cell interactions down to the level of individual regulatory molecules will provide a greater understanding of how compaction shapes the early mammalian embryo. © 2016 Elsevier Inc. All rights reserved.

  1. Small Valdivia compact spaces

    Kubi's, W; Kubi\\'s, Wieslaw; Michalewski, Henryk

    2005-01-01

    We prove a preservation theorem for the class of Valdivia compact spaces, which involves inverse sequences of ``simple'' retractions. Consequently, a compact space of weight $\\loe\\aleph_1$ is Valdivia compact iff it is the limit of an inverse sequence of metric compacta whose bonding maps are retractions. As a corollary, we show that the class of Valdivia compacta of weight at most $\\aleph_1$ is preserved both under retractions and under open 0-dimensional images. Finally, we characterize the class of all Valdivia compacta in the language of category theory, which implies that this class is preserved under all continuous weight preserving functors.

  2. Sensisivity and Uncertainty analysis for the Tritium Breeding Ratio of a DEMO Fusion reactor with a Helium cooled pebble bed blanket

    Nunnenmann, Elena; Fischer, Ulrich; Stieglitz, Robert

    2016-01-01

    An uncertainty analysis was performed for the tritium breeding ratio (TBR) of a fusion power plant of the European DEMO type using the MCSEN patch to the MCNP Monte Carlo code. The breeding blanket was of the type Helium Cooled Pebble Bed (HCPB), currently under development in the European Power Plant Physics and Technology (PPPT) programme for a fusion power demonstration reactor (DEMO). A suitable 3D model of the DEMO reactor with HCPB blanket modules, as routinely used for blanket design c...

  3. A Fusion Nuclear Science Facility for a fast-track path to DEMO

    Garofalo, A.M., E-mail: garofalo@fusion.gat.com [General Atomics, San Diego, CA (United States); Abdou, M.A. [University of California, Los Angeles, Los Angeles, CA (United States); Canik, J.M. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Chan, V.S.; Hyatt, A.W. [General Atomics, San Diego, CA (United States); Hill, D.N. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Morley, N.B. [University of California, Los Angeles, Los Angeles, CA (United States); Navratil, G.A. [Columbia University, New York, NY (United States); Sawan, M.E. [University of Wisconsin Madison, Madison, WI (United States); Taylor, T.S.; Wong, C.P.C.; Wu, W. [General Atomics, San Diego, CA (United States); Ying, A. [University of California, Los Angeles, Los Angeles, CA (United States)

    2014-10-15

    Highlights: • A FNSF is needed to reduce the knowledge gaps to a fusion DEMO and accelerate progress toward fusion energy. • FNSF will test and qualify first-wall/blanket components and materials in a DEMO-relevant fusion environment. • The Advanced Tokamak approach enables reduced size and risks, and is on a direct path to an attractive target power plant. • Near term research focus on specific tasks can enable starting FNSF construction within the next ten years. - Abstract: An accelerated fusion energy development program, a “fast-track” approach, requires proceeding with a nuclear and materials testing program in parallel with research on burning plasmas, ITER. A Fusion Nuclear Science Facility (FNSF) would address many of the key issues that need to be addressed prior to DEMO, including breeding tritium and completing the fuel cycle, qualifying nuclear materials for high fluence, developing suitable materials for the plasma-boundary interface, and demonstrating power extraction. The Advanced Tokamak (AT) is a strong candidate for an FNSF as a consequence of its mature physics base, capability to address the key issues, and the direct relevance to an attractive target power plant. The standard aspect ratio provides space for a solenoid, assuring robust plasma current initiation, and for an inboard blanket, assuring robust tritium breeding ratio (TBR) >1 for FNSF tritium self-sufficiency and building of inventory needed to start up DEMO. An example design point gives a moderate sized Cu-coil device with R/a = 2.7 m/0.77 m, κ = 2.3, B{sub T} = 5.4 T, I{sub P} = 6.6 MA, β{sub N} = 2.75, P{sub fus} = 127 MW. The modest bootstrap fraction of ƒ{sub BS} = 0.55 provides an opportunity to develop steady state with sufficient current drive for adequate control. Proceeding with a FNSF in parallel with ITER provides a strong basis to begin construction of DEMO upon the achievement of Q ∼ 10 in ITER.

  4. Neutronics studies for the design of the European DEMO vacuum vessel

    Flammini, Davide, E-mail: davide.flammini@enea.it [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Villari, Rosaria; Moro, Fabio; Pizzuto, Aldo [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Bachmann, Christian [EUROfusion Consortium, Boltzmannstr. 2, 85748 Garching (Germany)

    2016-11-01

    Highlights: • MCNP calculation of nuclear heating, damage, helium production and neutron flux in DEMO HCLL and HCPB vacuum vessel at the inboard equatorial plane. • Study of impact of the poloidal gap between blanket modules, for several gap width, on vacuum vessel nuclear quantities. • Effect of the gap on nuclear heating result to be moderate, however high values of nuclear heating are found, even far from the gap with HCLL blanket. • Radiation damage limit of 2.75 DPA is met with a 1 cm wide gap. Helium production results very sensitive to the gap width. • Comparison between HCLL and HCPB blankets is shown for nuclear heating and neutron flux in the vacuum vessel. - Abstract: The DEMO vacuum vessel, a massive water cooled double-walled steel vessel, is located behind breeding blankets and manifolds and it will be subjected to an intense neutron and photon irradiation. Therefore, a proper evaluation of the vessel nuclear heat loads is required to assure adequate cooling and, given the significant lifetime neutron fluence of DEMO, the radiation damage limit of the vessel needs to be carefully controlled. In the present work nuclear heating, radiation damage (DPA), helium production, neutron and photon fluxes have been calculated on the vacuum vessel at the inboard by means of MCNP5 using a 3D Helium Cooled Lithium Lead (HCLL) DEMO model with 1572 MW of fusion power. In particular, the effect of the poloidal gap between the breeding-blanket segments on vacuum vessel nuclear loads has been estimated varying the gap width from 0 to 5 cm. High values of the nuclear heating (≈1 W/cm{sup 3}), which might cause intense thermal stresses, were obtained in inboard equatorial zone. The effect of the poloidal gap on the nuclear heating resulted to be moderate (within 30%). The radiation damage limit of 2.75 DPA on the vessel is almost met with 1 cm of poloidal gap over DEMO lifetime. A comparison with Helium Cooled Pebble Bed blanket is also provided.

  5. Compact turbidity meter

    Hirschberg, J. G.

    1979-01-01

    Proposed monitor that detects back-reflected infrared radiation makes in situ turbidity measurements of lakes, streams, and other bodies of water. Monitor is compact, works well in daylight as at night, and is easily operated in rough seas.

  6. Compaction of FGD-gypsum

    Stoop, B.T.J.; Larbi, J.A.; Heijnen, W.M.M.

    1996-01-01

    It is shown that it is possible to produce compacted gypsum with a low porosity and a high strength on a laboratory scale by uniaxial compaction of flue gas desulphurization (FGD-) gypsum powder. Compacted FGD-gypsum cylinders were produced at a compaction pres-sure between 50 and 500 MPa yielding

  7. Physically detached 'compact groups'

    Hernquist, Lars; Katz, Neal; Weinberg, David H.

    1995-01-01

    A small fraction of galaxies appear to reside in dense compact groups, whose inferred crossing times are much shorter than a Hubble time. These short crossing times have led to considerable disagreement among researchers attempting to deduce the dynamical state of these systems. In this paper, we suggest that many of the observed groups are not physically bound but are chance projections of galaxies well separated along the line of sight. Unlike earlier similar proposals, ours does not require that the galaxies in the compact group be members of a more diffuse, but physically bound entity. The probability of physically separated galaxies projecting into an apparent compact group is nonnegligible if most galaxies are distributed in thin filaments. We illustrate this general point with a specific example: a simulation of a cold dark matter universe, in which hydrodynamic effects are included to identify galaxies. The simulated galaxy distribution is filamentary and end-on views of these filaments produce apparent galaxy associations that have sizes and velocity dispersions similar to those of observed compact groups. The frequency of such projections is sufficient, in principle, to explain the observed space density of groups in the Hickson catalog. We discuss the implications of our proposal for the formation and evolution of groups and elliptical galaxies. The proposal can be tested by using redshift-independent distance estimators to measure the line-of-sight spatial extent of nearby compact groups.

  8. Optimization of the first wall for the DEMO water cooled lithium lead blanket

    Aubert, Julien, E-mail: julien.aubert@cea.fr [CEA Saclay, F-91191 Gif-Sur-Yvette (France); Aiello, Giacomo [CEA Saclay, F-91191 Gif-Sur-Yvette (France); Bachmann, Christian [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Di Maio, Pietro Alessandro [Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Giammusso, Rosario [ENEA C.R. Brasimone, 40032 Camugnano, Bologna (Italy); Li Puma, Antonella; Morin, Alexandre [CEA Saclay, F-91191 Gif-Sur-Yvette (France); Tincani, Amelia [ENEA C.R. Brasimone, 40032 Camugnano, Bologna (Italy)

    2015-10-15

    Highlights: • This paper presents the optimization of the first wall of the water cooled lithium lead DEMO blanket with pressurized water reactor condition and circular channels in order to find the best geometry that can allow the maximum heat flux considering design criteria since an estimate of the engineering limit of the first wall heat load capacity is an essential input for the decision to implement limiters in DEMO. • An optimization study was carried out for the flat first wall design of the DEMO Water-Cooled Lithium Lead considering thermal and mechanical constraint functions, assuming T{sub inlet}/T{sub outlet} equal to 285 °C/325 °C, based on geometric design parameters. • It became clear that through the optimization the advantages of a waved First Wall are diminished. • The analysis shows that the maximum heat load could achieve 2.53 MW m{sup −2}, but considering assumptions such as a coolant velocity ≤8 m/s, pipe diameter ≥5 mm and a total first wall thickness ≤22 mm, heat flux is limited to 1.57 MW m{sup −2}. - Abstract: The maximum heat load capacity of a DEMO First Wall (FW) of reasonable cost may impact the decision of the implementation of limiters in DEMO. An estimate of the engineering limit of the FW heat load capacity is an essential input for this decision. This paper describes the work performed to optimize the FW of the Water Cooled Lithium-Lead (WCLL) blanket concept for DEMO fusion reactor in order to increase its maximum heat load capacity. The optimization is based on the use of water at typical Pressurised Water Reactors conditions as coolant. The present WCLL FW with a waved plasma-faced surface and with circular channels was studied and the heat load limit has been predicted with FEM analysis equal to 1.0 MW m{sup −2} with respect to the Eurofer temperature limit. An optimization study was then carried out for a flat FW design considering thermal and mechanical constraints assuming inlet and outlet

  9. Failure rate modeling using fault tree analysis and Bayesian network: DEMO pulsed operation turbine study case

    Dongiovanni, Danilo Nicola, E-mail: danilo.dongiovanni@enea.it [ENEA, Nuclear Fusion and Safety Technologies Department, via Enrico Fermi 45, Frascati 00040 (Italy); Iesmantas, Tomas [LEI, Breslaujos str. 3 Kaunas (Lithuania)

    2016-11-01

    Highlights: • RAMI (Reliability, Availability, Maintainability and Inspectability) assessment of secondary heat transfer loop for a DEMO nuclear fusion plant. • Definition of a fault tree for a nuclear steam turbine operated in pulsed mode. • Turbine failure rate models update by mean of a Bayesian network reflecting the fault tree analysis in the considered scenario. • Sensitivity analysis on system availability performance. - Abstract: Availability will play an important role in the Demonstration Power Plant (DEMO) success from an economic and safety perspective. Availability performance is commonly assessed by Reliability Availability Maintainability Inspectability (RAMI) analysis, strongly relying on the accurate definition of system components failure modes (FM) and failure rates (FR). Little component experience is available in fusion application, therefore requiring the adaptation of literature FR to fusion plant operating conditions, which may differ in several aspects. As a possible solution to this problem, a new methodology to extrapolate/estimate components failure rate under different operating conditions is presented. The DEMO Balance of Plant nuclear steam turbine component operated in pulse mode is considered as study case. The methodology moves from the definition of a fault tree taking into account failure modes possibly enhanced by pulsed operation. The fault tree is then translated into a Bayesian network. A statistical model for the turbine system failure rate in terms of subcomponents’ FR is hence obtained, allowing for sensitivity analyses on the structured mixture of literature and unknown FR data for which plausible value intervals are investigated to assess their impact on the whole turbine system FR. Finally, the impact of resulting turbine system FR on plant availability is assessed exploiting a Reliability Block Diagram (RBD) model for a typical secondary cooling system implementing a Rankine cycle. Mean inherent availability

  10. Failure rate modeling using fault tree analysis and Bayesian network: DEMO pulsed operation turbine study case

    Dongiovanni, Danilo Nicola; Iesmantas, Tomas

    2016-01-01

    Highlights: • RAMI (Reliability, Availability, Maintainability and Inspectability) assessment of secondary heat transfer loop for a DEMO nuclear fusion plant. • Definition of a fault tree for a nuclear steam turbine operated in pulsed mode. • Turbine failure rate models update by mean of a Bayesian network reflecting the fault tree analysis in the considered scenario. • Sensitivity analysis on system availability performance. - Abstract: Availability will play an important role in the Demonstration Power Plant (DEMO) success from an economic and safety perspective. Availability performance is commonly assessed by Reliability Availability Maintainability Inspectability (RAMI) analysis, strongly relying on the accurate definition of system components failure modes (FM) and failure rates (FR). Little component experience is available in fusion application, therefore requiring the adaptation of literature FR to fusion plant operating conditions, which may differ in several aspects. As a possible solution to this problem, a new methodology to extrapolate/estimate components failure rate under different operating conditions is presented. The DEMO Balance of Plant nuclear steam turbine component operated in pulse mode is considered as study case. The methodology moves from the definition of a fault tree taking into account failure modes possibly enhanced by pulsed operation. The fault tree is then translated into a Bayesian network. A statistical model for the turbine system failure rate in terms of subcomponents’ FR is hence obtained, allowing for sensitivity analyses on the structured mixture of literature and unknown FR data for which plausible value intervals are investigated to assess their impact on the whole turbine system FR. Finally, the impact of resulting turbine system FR on plant availability is assessed exploiting a Reliability Block Diagram (RBD) model for a typical secondary cooling system implementing a Rankine cycle. Mean inherent availability

  11. Status of advanced tritium breeder development for DEMO in the broader approach activities in Japan

    Hoshino, Tsuyoshi; Oikawa, Fumiaki; Nishitani, Takeo

    2010-01-01

    DEMO reactors require ' 6 Li-enriched ceramic tritium breeders' which have high tritium breeding ratios (TBRs) in the blanket designs of both EU and JA. Both parties have been promoting the development of fabrication technologies of Li 2 TiO 3 pebbles and of Li 4 SiO 4 pebbles including the reprocessing. However, the fabrication techniques of tritium breeders pebbles have not been established for large quantities. Therefore, these parties launch a collaborative project on scaleable and reliable production routes of advanced tritium breeders. In addition, this project aims to develop fabrication techniques allowing effective reprocessing of 6 Li. The development of the production and 6 Li reprocessing techniques includes preliminary fabrication tests of breeder pebbles, reprocessing of lithium, and suitable out-of-pile characterizations. The R and D on the fabrication technologies of the advanced tritium breeders and the characterization of developed materials has been started between the EU and Japan in the DEMO R and D of the International Fusion Energy Research Centre (IFERC) project as a part of the Broader Approach activities from 2007 to 2016. The equipment for production of advanced breeder pebbles is planned will be installed in the DEMO R and D building at Rokkasho, Japan. The design work in this facility was carried out. The specifications of the pebble production apparatuses and related equipment in this facility were fixed, and the basic data of these apparatuses was obtained. In this design work, the preliminary investigations of the dissolution and purification process of tritium breeders were carried out. From the results of the preliminary investigations, lithium resources of 90% above were recovered by the aqueous dissolving methods using HNO 3 and H 2 O 2 . The removal efficiency of 60 Co by the addition in the dissolved solutions of lithium ceramics were 97-99.9% above using activated carbon impregnated with 8-hydroxyquinolinol. In this report

  12. Conceptual design studies for the European DEMO divertor: Rationale and first results

    You, J.H.; Mazzone, G.; Visca, E.; Bachmann, Ch.; Autissier, E.; Barrett, T.; Cocilovo, V.; Crescenzi, F.; Domalapally, P.K.; Dongiovanni, D.; Entler, S.; Federici, G.; Frosi, P.; Fursdon, M.; Greuner, H.; Hancock, D.; Marzullo, D.; McIntosh, S.; Müller, A.V.; Porfiri, M.T.

    2016-01-01

    Highlights: • A brief overview is given on the overall R&D activities of the work package Divertor which is a project of the EUROfusion Consortium. • The rationale of the hydraulic, thermal and structural design scheme is described. • The first results obtained for the preliminary DEMO divertor cassette model are presented. - Abstract: In the European fusion roadmap, reliable power handling has been defined as one of the most critical challenges for realizing a commercially viable fusion power. In this context, the divertor is the key in-vessel component, as it is responsible for power exhaust and impurity removal for which divertor target is subjected to very high heat flux loads. To this end, an integrated R&D project was launched in the EUROfusion Consortium in order to deliver a holistic conceptual design solution together with the core technologies for the entire divertor system of a DEMO reactor. The work package ‘Divertor’ consists of two project areas: ‘Cassette design and integration’ and ‘Target development’. The essential mission of the project is to develop and verify advanced design concepts and the required technologies for a divertor system being capable of meeting the physical and system requirements defined for the next-generation European DEMO reactor. In this contribution, a brief overview is presented of the works from the first project year (2014). Focus is put on the loads specification, design boundary conditions, materials requirements, design approaches, and R&D strategy. Initial ideas and first estimates are presented.

  13. Conceptual design studies for the European DEMO divertor: Rationale and first results

    You, J.H., E-mail: you@ipp.mpg.de [Max Planck Institute for Plasma Physics, Boltzmann Str. 2, 85748 Garching (Germany); Mazzone, G.; Visca, E. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); Bachmann, Ch. [EUROfusion PMU, c/o IPP, Boltzmann Str. 2, 85748 Garching (Germany); Autissier, E. [CEA, IRFM, F-13108 Saint Paul Lez Durance (France); Barrett, T. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Cocilovo, V.; Crescenzi, F. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); Domalapally, P.K. [Research Cnter Rez, Hlavní 130, 250 68 Husinec–Řež (Czech Republic); Dongiovanni, D. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); Entler, S. [Institute of Plasma Physics CAS, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Federici, G. [EUROfusion PMU, c/o IPP, Boltzmann Str. 2, 85748 Garching (Germany); Frosi, P. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); Fursdon, M. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Greuner, H. [Max Planck Institute for Plasma Physics, Boltzmann Str. 2, 85748 Garching (Germany); Hancock, D. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Marzullo, D. [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy); McIntosh, S. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Müller, A.V. [Max Planck Institute for Plasma Physics, Boltzmann Str. 2, 85748 Garching (Germany); Porfiri, M.T. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); and others

    2016-11-01

    Highlights: • A brief overview is given on the overall R&D activities of the work package Divertor which is a project of the EUROfusion Consortium. • The rationale of the hydraulic, thermal and structural design scheme is described. • The first results obtained for the preliminary DEMO divertor cassette model are presented. - Abstract: In the European fusion roadmap, reliable power handling has been defined as one of the most critical challenges for realizing a commercially viable fusion power. In this context, the divertor is the key in-vessel component, as it is responsible for power exhaust and impurity removal for which divertor target is subjected to very high heat flux loads. To this end, an integrated R&D project was launched in the EUROfusion Consortium in order to deliver a holistic conceptual design solution together with the core technologies for the entire divertor system of a DEMO reactor. The work package ‘Divertor’ consists of two project areas: ‘Cassette design and integration’ and ‘Target development’. The essential mission of the project is to develop and verify advanced design concepts and the required technologies for a divertor system being capable of meeting the physical and system requirements defined for the next-generation European DEMO reactor. In this contribution, a brief overview is presented of the works from the first project year (2014). Focus is put on the loads specification, design boundary conditions, materials requirements, design approaches, and R&D strategy. Initial ideas and first estimates are presented.

  14. Effect on the Tritium Breeding Ratio due to a distributed ICRF antenna in a DEMO reactor

    Garcia, A.; Noterdaeme, J.-M.; Fischer, U.; Dies, J.

    2016-01-01

    This thesis reports results of MCNP-5 calculations, with the nuclear data library FENDL-2.1, to assess the effect on the Tritium Breeding Ratio (TBR) due to a distributed Ion Cyclotron Range of Frequencies (ICRF) antenna integrated in the blanket of a DEMO fusion power reactor. A preliminary design of the antenna with a reference configuration of the DEMO reactor was used together with a parametric analysis for different parameters that strongly affect the TBR. These are the type of breeding blanket (Helium Cooled Pebble Bed, Helium Cooled Lithium Lead and Water Cooled Lithium Lead), the covering ratio of the straps of the antenna (the ratio between the surface of all the straps and the projected surface of the antenna slot: 0.49, 0.72 and 0.94), the antenna radial thickness (20 cm and 40 cm), the thickness of the straps (2 cm, 4 cm and a double layer of 0.2 cm plus 2.5 cm with the composition of the First Wall), and finally the poloidal position of the antenna (0°, which is the equatorial port, 40° and 90°, which is the upper port). For an antenna with a full toroidal circumference of 360°, located poloidaly at 40° with a poloidal extension of 1 m and a total First Wall surface of 67 m"2, the reduction of the TBR is −0.35% for a HCPB blanket concept, −0.53% for a HCLL blanket concept and −0.51% for a WCLL blanket concept. In all cases covered by the parametric analysis, the loss of TBR remains below 0.61%. Such a distributed ICRF antenna has thus only a marginal effect on the TBR for a DEMO reactor.

  15. Modelling of mitigation of the power divertor loading for the EU DEMO through Ar injection

    Subba, Fabio; Aho-Mantila, Leena; Coster, David; Maddaluno, Giorgio; Nallo, Giuseppe F.; Sieglin, Bernard; Wenninger, Ronald; Zanino, Roberto

    2018-03-01

    In this paper we present a computational study on the divertor heat load mitigation through impurity injection for the EU DEMO. The study is performed by means of the SOLPS5.1 code. The power crossing the separatrix is considered fixed and corresponding to H-mode operation, whereas the machine operating condition is defined by the outboard mid-plane upstream electron density and the impurity level. The selected impurity for this study is Ar, based on its high radiation efficiency at SOL characteristic temperatures. We consider a conventional vertical target geometry for the EU DEMO and monitor target conditions for different operational points, considering as acceptability criteria the target electron temperature (≤5 eV to provide sufficiently low W sputtering rate) and the peak heat flux (below 5-10 MW m-2 to guarantee safe steady-state cooling conditions). Our simulations suggest that, neglecting the radiated power deposition on the plate, it is possible to satisfy the desired constraints. However, this requires an upstream density of the order of at least 50% of the Greenwald limit and a sufficiently high argon fraction. Furthermore, if the radiated power deposition is taken into account, the peak heat flux on the outer plate could not be reduced below 15 MW m-2 in these simulations. As these simulations do not take into account neutron loading, they strongly indicate that the vertical target divertor solution with a radiative front distributed along the divertor leg has a very marginal operational space in an EU DEMO sized reactor.

  16. Effect on the Tritium Breeding Ratio due to a distributed ICRF antenna in a DEMO reactor

    Garcia, A., E-mail: albert.garcia.hp@gmail.com [Max-Planck-Institut für Plasmaphysik (IPP), Garching (Germany); Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Polytechnic University of Catalonia (UPC), Barcelona (Spain); Department of Applied Physics, Ghent University, Ghent (Belgium); Noterdaeme, J.-M. [Max-Planck-Institut für Plasmaphysik (IPP), Garching (Germany); Department of Applied Physics, Ghent University, Ghent (Belgium); Fischer, U. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Dies, J. [Polytechnic University of Catalonia (UPC), Barcelona (Spain)

    2016-11-15

    This thesis reports results of MCNP-5 calculations, with the nuclear data library FENDL-2.1, to assess the effect on the Tritium Breeding Ratio (TBR) due to a distributed Ion Cyclotron Range of Frequencies (ICRF) antenna integrated in the blanket of a DEMO fusion power reactor. A preliminary design of the antenna with a reference configuration of the DEMO reactor was used together with a parametric analysis for different parameters that strongly affect the TBR. These are the type of breeding blanket (Helium Cooled Pebble Bed, Helium Cooled Lithium Lead and Water Cooled Lithium Lead), the covering ratio of the straps of the antenna (the ratio between the surface of all the straps and the projected surface of the antenna slot: 0.49, 0.72 and 0.94), the antenna radial thickness (20 cm and 40 cm), the thickness of the straps (2 cm, 4 cm and a double layer of 0.2 cm plus 2.5 cm with the composition of the First Wall), and finally the poloidal position of the antenna (0°, which is the equatorial port, 40° and 90°, which is the upper port). For an antenna with a full toroidal circumference of 360°, located poloidaly at 40° with a poloidal extension of 1 m and a total First Wall surface of 67 m{sup 2}, the reduction of the TBR is −0.35% for a HCPB blanket concept, −0.53% for a HCLL blanket concept and −0.51% for a WCLL blanket concept. In all cases covered by the parametric analysis, the loss of TBR remains below 0.61%. Such a distributed ICRF antenna has thus only a marginal effect on the TBR for a DEMO reactor.

  17. BA DEMO R and D, activities on advanced tritium breeders in EU

    Knitter, Regina; Kolb, Matthias H.H.; Leys, Oliver H.J.B. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany). Inst. for Applied Materials (IAM-WPT)

    2013-07-01

    Within the Broader Approach (BA) activities on DEMO R and D, EU and Japan have launched a collaborative project on scalable and reliable production routes for advanced tritium breeders. Besides the development of the fabrication process, the reprocessing as well as the long-term stability of advanced breeder is to be investigated. In the EU, a modified melt-based process for the fabrication of lithium orthosilicate pebbles have been developed. Besides the optimization of process parameters, the chemical composition of the pebbles was altered by additions of titania in order to increase the mechanical properties by the formation of lithium metatitanate as a secondary, strengthening phase. (orig.)

  18. Demos Center, Militsiia mezhdu Rossiei i Chechnei. Veterany konflikta v rossiiskom obshchestve

    Elisabeth Sieca-Kozlowski

    2009-03-01

    Full Text Available The Demos study on policemen who are veterans of the Chechen war is the Centre’s second in-depth study. The first dealt with the phenomenon of arbitrariness (“proizvol” in the police force. This new study focuses on one of the contributing factors of this arbitrariness, the fact of having gone through the Chechen war. Although the two Chechen conflicts (1994-1996 and 1999 to the present involved the dispatch of tens of thousands of military and members of “power” ministries to the combat zo...

  19. Assembling a game development scene? Uncovering Finland’s largest demo party

    Heikki Tyni

    2014-03-01

    Full Text Available The study takes look at Assembly, a large-scale LAN and demo party founded in 1992 and organized annually in Helsinki, Finland. Assembly is used as a case study to explore the relationship between computer hobbyism – including gaming, demoscene and other related activities – and professional game development. Drawing from expert interviews, a visitor query and news coverage we ask what kind of functions Assembly has played for the scene in general, and on the formation and fostering of the Finnish game industry in particular. The conceptual contribution of the paper is constructed around the interrelated concepts of scene, technicity and gaming capital.

  20. Evaluation of European blanket concepts for DEMO from availability and reliability point of view

    Nardi, C.

    1995-12-01

    This technical report is concerned with the ENEA activities relating to reliability and availability for the selection among two of the four European blanket concepts for the DEMO reactor. The activities on the BIT concept, the one proposed by ENEA, are emphasized. In spite of the lack of data relating to the behaviour of structures in an environment similar to that of a fusion reactor, it is evidenced that the available data are relevant to the BIT concept geometry. Moreover, it is evidenced that the qualitative reliability evaluations, compared to the quantitative ones, can lead to a better understanding of the typical problems of a structure to be used in a fusion reactor

  1. Inhomogeneous compact extra dimensions

    Bronnikov, K.A. [Center of Gravity and Fundamental Metrology, VNIIMS, 46 Ozyornaya st., Moscow 119361 (Russian Federation); Budaev, R.I.; Grobov, A.V.; Dmitriev, A.E.; Rubin, Sergey G., E-mail: kb20@yandex.ru, E-mail: buday48@mail.ru, E-mail: alexey.grobov@gmail.com, E-mail: alexdintras@mail.ru, E-mail: sergeirubin@list.ru [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), 115409 Moscow (Russian Federation)

    2017-10-01

    We show that an inhomogeneous compact extra space possesses two necessary features— their existence does not contradict the observable value of the cosmological constant Λ{sub 4} in pure f ( R ) theory, and the extra dimensions are stable relative to the 'radion mode' of perturbations, the only mode considered. For a two-dimensional extra space, both analytical and numerical solutions for the metric are found, able to provide a zero or arbitrarily small Λ{sub 4}. A no-go theorem has also been proved, that maximally symmetric compact extra spaces are inconsistent with 4D Minkowski space in the framework of pure f ( R ) gravity.

  2. Interview of Tanya Lokshina, President of the Demos center, conducted by Olga Filippova, Moscow, 11 May 2007

    2007-12-01

    Full Text Available Demos Veterans ProjectPIPSS.ORG – Could you please retrace for us the history of the research program untitled “Veterans of Chechnya” and, inside it, the sub-program “Drawing Public Attention to the Chechen Conflict through the Prism of Issues Associated with Social Adaptation and Professional Activities of Veterans”? Tanya Lokshina: Demos provides informative and expert-analytical work on current issues in Russia. On the basis of our informative work, we carry out in-depth research into the ...

  3. Characterization of ceramic powder compacts

    Yanai, K.; Ishimoto, S.; Kubo, T.; Ito, K.; Ishikawa, T.; Hayashi, H.

    1995-01-01

    UO 2 and Al 2 O 3 powder packing structures in cylindrical powder compacts are observed by scanning electron microscopy using polished cross sections of compacts fixed by low viscosity epoxy resin. Hard aggregates which are not destroyed during powder compaction are observed in some of the UO 2 powder compacts. A technique to measure local density in powder compacts is developed based on counting characteristic X-ray intensity by energy dispersive X-ray analysis (EDX). The local density of the corner portion of the powder compact fabricated by double-acting dry press is higher than that of the inner portion. ((orig.))

  4. Concept of DT fuel cycle for a fusion neutron source DEMO-FNS

    Ananyev, Sergey S., E-mail: Ananyev_SS@nrcki.ru; Spitsyn, Alexander V.; Kuteev, Boris V.

    2016-11-01

    Highlights: • We presented the concept of a deuterium-tritium fuel cycle of stationary thermonuclear reactor. • Data of fuel cycles for nuclear facility (DEMO-FNS) with 2 variants of the fuel mixture for NBI system are presented. • The amount of tritium which is required for operation of DEMO-FNS is estimated. - Abstract: The paper describes the concept of a deuterium-tritium fuel cycle of a steady-state thermonuclear reactor with a fusion power over 10 MW. Parameters of fuel cycle for nuclear facility (JET scale) with different types of fuel mixtures for neutral beam injection system are presented. Optimization of fuel cycle characteristics was aimed at reducing flows and inventory of hydrogen isotopes and tritium in fuel cycle subsystems. The calculations were carried out using computer code TC-FNS to estimate tritium distribution in fusion reactor systems and components of “tritium plant”. The code enables calculations of tritium flows and inventory in the tokamak systems. Calculations of tritium flows and accumulation have been carried out for two different cases of the fuel mixture for neutral beam injection (NBI) system. The amounts of tritium which is required for operation of all fuel cycle systems in two different cases of the fuel mixture for NBI are 0.45 “” kg (D:T = 1:0) and 0.9 kg (D:T = 1:1) respectively.

  5. Numerical analysis of tungsten erosion and deposition processes under a DEMO divertor plasma

    Yuki Homma

    2017-08-01

    Full Text Available Erosion reduction of tungsten (W divertor target is one of the most important research subjects for the DEMO fusion reactor design, because the divertor target has to sustain large fluence of incident particles, composed mainly of fuel ions and seeded impurities, during year-long operation period. Rate of net erosion and deposition on outer divertor target has been studied by using the integrated SOL/divertor plasma code SONIC and the kinetic full-orbit impurity transport code IMPGYRO. Two background plasmas have been used: one is lower density ni and higher temperature case and the other is higher ni and lower temperature case. Net erosion has been seen in the lower ni case. But in the higher ni case, the net erosion has been almost suppressed due to increased return rate and reduced self-sputtering yield. Following two factors are important to understand the net erosion formation: (i ratio of the 1st ionization length of sputtered W atom to the Larmor gyro radius of W+ ion, (ii balance between the friction force and the thermal force exerted on W ions. DEMO divertor design should take into account these factors to prevent target erosion.

  6. Parametric analyses of DEMO Divertor using two dimensional transient thermal hydraulic modelling

    Domalapally, Phani; Di Caro, Marco

    2018-05-01

    Among the options considered for cooling of the Plasma facing components of the DEMO reactor, water cooling is a conservative option because of its high heat removal capability. In this work a two-dimensional transient thermal hydraulic code is developed to support the design of the divertor for the projected DEMO reactor with water as a coolant. The mathematical model accounts for transient 2D heat conduction in the divertor section. Temperature-dependent properties are used for more accurate analysis. Correlations for single phase flow forced convection, partially developed subcooled nucleate boiling, fully developed subcooled nucleate boiling and film boiling are used to calculate the heat transfer coefficients on the channel side considering the swirl flow, wherein different correlations found in the literature are compared against each other. Correlation for the Critical Heat Flux is used to estimate its limit for a given flow conditions. This paper then investigates the results of the parametric analysis performed, whereby flow velocity, diameter of the coolant channel, thickness of the coolant pipe, thickness of the armor material, inlet temperature and operating pressure affect the behavior of the divertor under steady or transient heat fluxes. This code will help in understanding the basic parameterś effect on the behavior of the divertor, to achieve a better design from a thermal hydraulic point of view.

  7. Neutronic design analyses for a dual-coolant blanket concept: Optimization for a fusion reactor DEMO

    Palermo, I.; Gómez-Ros, J.M.; Veredas, G.; Sanz, J.; Sedano, L.

    2012-01-01

    Highlights: ► Dual-Coolant He/Pb15.7Li breeding blanket for a DEMO fusion reactor is studied. ► An iterative process optimizes neutronic responses minimizing reactor dimension. ► A 3D toroidally symmetric geometry has been generated from the CAD model. ► Overall TBR values support the feasibility of the conceptual model considered. ► Power density in TF coils is below load limit for quenching. - Abstract: The generation of design specifications for a DEMO reactor, including breeding blanket (BB), vacuum vessel (VV) and magnetic field coils (MFC), requires a consistent neutronic optimization of structures between plasma and MFC. This work targets iteratively to generate these neutronic specifications for a Dual-Coolant He/Pb15.7Li breeding blanket design. The iteration process focuses on the optimization of allowable space between plasma scrapped-off-layer and VV in order to generate a MFC/VV/BB/plasma sustainable configuration with minimum global system volumes. Two VV designs have been considered: (1) a double-walled option with light-weight stiffeners and (2) a thick massive one. The optimization process also involves VV materials, looking to warrant radiation impact operational limits on the MFC. The resulting nuclear responses: peak nuclear heating in toroidal field (TF) coil, tritium breeding ratio (TBR), power amplification factor and helium production in the structural material are provided.

  8. Analysis of displacement damage in materials in nuclear fusion facilities (DEMO, IFMIF and TechnoFusion)

    Mota, F.; Vila, R.; Ortiz, C.; Garcia, A.; Casal, N.; Ibarra, A.; Rapisarda, D.; Queral, V.

    2011-01-01

    Present pathway to fusion reactors includes a rigorous material testing program. To reach this objective, irradiation facilities must produce the displacement damage per atom (dpa), primary knock-on atom (PKA) spectrum and gaseous elements by transmutation reactions (He, H) as closely as possible to the ones expected in the future fusion reactors (as DEMO).The irradiation parameters (PKA spectra and damage function) of some candidate materials for fusion reactors (Al 2 O 3 , SiC and Fe) have been studied and then, the suitability of some proposed experimental facilities, such as IFMIF and TechnoFusion, to perform relevant tests with these materials has been assessed.The following method has been applied: neutron fluxes present in different irradiation modules of IFMIF have been calculated by the neutron transport McDeLicious code. In parallel, the energy differential cross sections of PKA have been calculated by using the NJOY code. After that, the damage generated by the PKA spectra was analyzed using the MARLOWE code (binary collision approximation) and custom analysis codes. Finally, to analyze the ions effects in different irradiation conditions in the TechnoFusion irradiation area, the SRIM and Marlowe codes have been used. The results have been compared with the expected ones for a DEMO HCLL reactor.

  9. Evaluation of EM loads distribution on DEMO blanket segments and their effect on mechanical integrity

    Maione, Ivan Alessio; Zeile, Christian; Boccaccini, Lorenzo V.; Vaccaro, Alessandro

    2016-01-01

    Highlights: • Two DEMO 2015 ANSYS FEM models (for EM and structural analysis) have been implemented based on the EU-HCPB concept. • Lorentz’s forces have been calculated and their impact on the segment structure has been evaluated. • EM loads show a predominant total radial moment due to the high toroidal magnetic field (in comparison with the poloidal one). • A preliminary assessment of the primary stresses according the RCC-MRx code indicates the ability of the segments to resist the EM forces. - Abstract: This work is aimed to analyze the EM internal forces distribution on the blanket system (blankets modules and segment back supporting structure) of the EU PPPT DEMO 2015 reactor configuration. In order to validate their impact on the segment structure, an EM analysis is conducted using a simplified plasma central disruption. The calculated Lorentz’s forces distributions are then used as input for structural analyses focusing on the mechanical integrity of the segment back supporting structure. In particular, the electrical and structural assumptions used in this work are based on the HCPB blanket design developed at the Karlsruhe Institute of Technology. A preliminary assessment of the primary stresses according the design code RCC-MRx indicates the ability of the segments to resist the EM forces, where the lowest margin is given by the immediate plastic instability criterion on the inboard segment with 14%.

  10. Analysis of displacement damage in materials in nuclear fusion facilities (DEMO, IFMIF and TechnoFusion)

    Mota, F., E-mail: fernando.mota@ciemat.es [Laboratorio Nacional de Fusion por Confinamiento Magnetico-CIEMAT, 28040 Madrid (Spain); Vila, R.; Ortiz, C.; Garcia, A.; Casal, N.; Ibarra, A.; Rapisarda, D.; Queral, V. [Laboratorio Nacional de Fusion por Confinamiento Magnetico-CIEMAT, 28040 Madrid (Spain)

    2011-10-15

    Present pathway to fusion reactors includes a rigorous material testing program. To reach this objective, irradiation facilities must produce the displacement damage per atom (dpa), primary knock-on atom (PKA) spectrum and gaseous elements by transmutation reactions (He, H) as closely as possible to the ones expected in the future fusion reactors (as DEMO).The irradiation parameters (PKA spectra and damage function) of some candidate materials for fusion reactors (Al{sub 2}O{sub 3}, SiC and Fe) have been studied and then, the suitability of some proposed experimental facilities, such as IFMIF and TechnoFusion, to perform relevant tests with these materials has been assessed.The following method has been applied: neutron fluxes present in different irradiation modules of IFMIF have been calculated by the neutron transport McDeLicious code. In parallel, the energy differential cross sections of PKA have been calculated by using the NJOY code. After that, the damage generated by the PKA spectra was analyzed using the MARLOWE code (binary collision approximation) and custom analysis codes. Finally, to analyze the ions effects in different irradiation conditions in the TechnoFusion irradiation area, the SRIM and Marlowe codes have been used. The results have been compared with the expected ones for a DEMO HCLL reactor.

  11. Research and development plan of fusion technologies in JAERI toward DEMO reactors

    Nishitani, Takeo; Hayashi, Takumi; Abe, Tetsuya; Akiba, Masato; Isono, Takaaki; Inoue, Takashi; Enoeda, Mikio; Okuno, Kiyoshi; Koizumi, Norikiyo; Sakamoto, Keishi; Sato, Satoshi; Jitsukawa, Shiro; Sugimoto, Masayoshi; Suzuki, Satoshi; Seki, Shogo; Takatsu, Hideyuki; Tanzawa, Sadamitsu; Tsuchiya, Kunihiko; Nishi, Masataka; Hayashi, Kimio; Matsui, Hideki; Yamanishi, Toshihiko; Watanabe, Kazuhiro

    2005-03-01

    In accordance with the 'Third Phase Basic Program on Fusion Research and Development' established by the Fusion Council of the Japan Atomic Energy Commission, research and development (R and D) of fusion technologies aim at realization of two elements: development of ITER key components and their improvement for higher performances; and construction of sound technical basis of fusion nuclear technologies essential for fusion energy utilization. JAERI has been assigned in the Third Phase Basic Program as a responsible institute for developing the above two elements, and accordingly has been implementing technology R and Ds categorized in the following three areas: R and D for ITER construction and operation; R and D for ITER utilization (blanket testing in ITER) and toward DEMO; and R and D on basic fusion technologies. The present report reviews the status and the plan of fusion technology R and Ds in the latter two areas, and presents the technical objectives, technical issues, status of R and D and near-term R and D plans for: breeding blankets; structural materials; the IFMIF program; improvements of the key ITER components for higher performances toward DEMO; and basic fusion technologies. (author)

  12. Post-examination of helium-cooled tungsten components exposed to DEMO specific cyclic thermal loads

    Ritz, G.; Hirai, T.; Linke, J.; Norajitra, P.; Giniyatulin, R.; Singheiser, L.

    2009-01-01

    A concept of helium-cooled tungsten finger module was developed for the European DEMO divertor. The concept was realized and tested under DEMO specific cyclic thermal loads up to 10 MW/m 2 . The modules were examined carefully before and after loading by metallography and microstructural analyses. While before loading mainly discrete and shallow cracks were found on the tungsten surface due to the manufacturing process, dense crack networks were observed at the loaded surfaces due to the thermal stress. In addition, cracks occurred in the structural, heat sink part and propagated along the grains orientation of the deformed tungsten material. Facilitated by cracking, the molten brazing metal between the tungsten plasma facing material and the W-La 2 O 3 heat sink, that could not withstand the operational temperatures, infiltrated the tungsten components and, due to capillary forces, even reached the plasma facing surface through the cracks. The formed cavity in the brazed layer reduced the heat conduction and the modules were further damaged due to overheating during the applied heat loads. Based on this detailed characterization and possible improvements of the design and of the manufacturing routes are discussed.

  13. Development status of the integrated tokamak simulator for K-DEMO

    Kang, J. S.; Wang, J.; Hwang, Y. S. [Seoul National University, Seoul (Korea, Republic of); Jung, L. [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Korean fusion demonstration reactor (K-DEMO) study has been conducted to investigate the feasibility of an electricity generation, self-sustained tritium cycle, and component test facility. To estimate its capability, the integrated fusion operation simulator called INFRA has been developed by organizing relevant computational codes with standard data models and framework. The different modules of the integrated simulator are chosen among well-validated codes. Standard data models are directly linked with KSTAR experimental data so that the integrated simulator can be used for interpretative simulations but also for predictive simulations. In this study, the current status of code development and some examples of KSTAR interpretative simulations are reported. ITER integrated modelling and analysis suite is imported to K-DEMO data model to take over ITER experience and to accelerate collaboration with international IMAS community. Standardized rules and guideline have been developed by ITER team for many years. Based on strict policy, this data model has been established and updated. This data model is used for experimental and simulation results. The INFRA system has been utilized to be an alpha version of a KDEMO simulator. Database, framework, and module integration are conducted. A test equilibrium run for KSTAR is done by filling the database with experiment results. More modules will be incorporated in a near future. Validation with KSTAR data and benchmarking previous modelling activity is also planned in order to confirm the feasibility of this system.

  14. Preliminary structural assessment of DEMO vacuum vessel against a vertical displacement event

    Mozzillo, Rocco; Tarallo, Andrea; Marzullo, Domenico; Bachmann, Christian; Di Gironimo, Giuseppe; Mazzone, Giuseppe

    2016-01-01

    Highlights: • The paper focuses on a preliminary structural analysis of the current concept design of DEMO vacuum vessel. • The Vacuum Vessel was checked against the VDE in combinations with the weight force of all components that the vessel shall bear. • Different configurations for the vacuum vessel supports are considered, showing that the best solution is VV supported at the lower port. • The analyses evaluated the “P damage” according to RCC-MRx code. - Abstract: This paper focuses on a preliminary structural analysis of the current concept design of DEMO vacuum vessel (VV). The VV structure is checked against a vertical load due to a Vertical Displacement Event in combination with the weight force of all components that the main vessel shall bear. Different configurations for the supports are considered. Results show that the greatest safety margins are reached when the tokamak is supported through the lower ports rather than the equatorial ports, though all analyzed configurations are compliant with RCC-MRx design rules.

  15. Preliminary structural assessment of DEMO vacuum vessel against a vertical displacement event

    Mozzillo, Rocco, E-mail: rocco.mozzillo@unina.it [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy); Tarallo, Andrea; Marzullo, Domenico [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy); Bachmann, Christian [EUROfusion PMU, Boltzmannstraße 2, 85748 Garching (Germany); Di Gironimo, Giuseppe [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy); Mazzone, Giuseppe [Unità Tecnica Fusione - ENEA C.R. Frascati, Via E. Fermi 45, 00044 Frascati (Italy)

    2016-11-15

    Highlights: • The paper focuses on a preliminary structural analysis of the current concept design of DEMO vacuum vessel. • The Vacuum Vessel was checked against the VDE in combinations with the weight force of all components that the vessel shall bear. • Different configurations for the vacuum vessel supports are considered, showing that the best solution is VV supported at the lower port. • The analyses evaluated the “P damage” according to RCC-MRx code. - Abstract: This paper focuses on a preliminary structural analysis of the current concept design of DEMO vacuum vessel (VV). The VV structure is checked against a vertical load due to a Vertical Displacement Event in combination with the weight force of all components that the main vessel shall bear. Different configurations for the supports are considered. Results show that the greatest safety margins are reached when the tokamak is supported through the lower ports rather than the equatorial ports, though all analyzed configurations are compliant with RCC-MRx design rules.

  16. Neutronic analyses of the preliminary design of a DCLL blanket for the EUROfusion DEMO power plant

    Palermo, Iole, E-mail: iole.palermo@ciemat.es; Fernández, Iván; Rapisarda, David; Ibarra, Angel

    2016-11-01

    Highlights: • We perform neutronic calculations for the preliminary DCLL Blanket design. • We study the tritium breeding capability of the reactor. • We determine the nuclear heating in the main components. • We verify if the shielding of the TF coil is maintained. - Abstract: In the frame of the newly established EUROfusion WPBB Project for the period 2014–2018, four breeding blanket options are being investigated to be used in the fusion power demonstration plant DEMO. CIEMAT is leading the development of the conceptual design of the Dual Coolant Lithium Lead, DCLL, breeding blanket. The primary role of the blanket is of energy extraction, tritium production, and radiation shielding. With this aim the DCLL uses LiPb as primary coolant, tritium breeder and neutron multiplier and Eurofer as structural material. Focusing on the achievement of the fundamental neutronic responses a preliminary blanket model has been designed. Thus detailed 3D neutronic models of the whole blanket modules have been generated, arranged in a specific DCLL segmentation and integrated in the generic DEMO model. The initial design has been studied to demonstrate its viability. Thus, the neutronic behaviour of the blanket and of the shield systems in terms of tritium breeding capabilities, power generation and shielding efficiency has been assessed in this paper. The results demonstrate that the primary nuclear performances are already satisfactory at this preliminary stage of the design, having obtained the tritium self-sufficiency and an adequate shielding.

  17. Conceptual design studies of the Electron Cyclotron launcher for DEMO reactor

    Moro, Alessandro; Bruschi, Alex; Franke, Thomas; Garavaglia, Saul; Granucci, Gustavo; Grossetti, Giovanni; Hizanidis, Kyriakos; Tigelis, Ioannis; Tran, Minh-Quang; Tsironis, Christos

    2017-10-01

    A demonstration fusion power plant (DEMO) producing electricity for the grid at the level of a few hundred megawatts is included in the European Roadmap [1]. The engineering design and R&D for the electron cyclotron (EC), ion cyclotron and neutral beam systems for the DEMO reactor is being performed by Work Package Heating and Current Drive (WPHCD) in the framework of EUROfusion Consortium activities. The EC target power to the plasma is about 50 MW, in which the required power for NTM control and burn control is included. EC launcher conceptual design studies are here presented, showing how the main design drivers of the system have been taken into account (physics requirements, reactor relevant operations, issues related to its integration as in-vessel components). Different options for the antenna are studied in a parameters space including a selection of frequencies, injection angles and launch points to get the best performances for the antenna configuration, using beam tracing calculations to evaluate plasma accessibility and deposited power. This conceptual design studies comes up with the identification of possible limits, constraints and critical issues, essential in the selection process of launcher setup solution.

  18. Conceptual design studies of the Electron Cyclotron launcher for DEMO reactor

    Moro Alessandro

    2017-01-01

    Full Text Available A demonstration fusion power plant (DEMO producing electricity for the grid at the level of a few hundred megawatts is included in the European Roadmap [1]. The engineering design and R&D for the electron cyclotron (EC, ion cyclotron and neutral beam systems for the DEMO reactor is being performed by Work Package Heating and Current Drive (WPHCD in the framework of EUROfusion Consortium activities. The EC target power to the plasma is about 50 MW, in which the required power for NTM control and burn control is included. EC launcher conceptual design studies are here presented, showing how the main design drivers of the system have been taken into account (physics requirements, reactor relevant operations, issues related to its integration as in-vessel components. Different options for the antenna are studied in a parameters space including a selection of frequencies, injection angles and launch points to get the best performances for the antenna configuration, using beam tracing calculations to evaluate plasma accessibility and deposited power. This conceptual design studies comes up with the identification of possible limits, constraints and critical issues, essential in the selection process of launcher setup solution.

  19. Weakly compact operators and interpolation

    Maligranda, Lech

    1992-01-01

    The class of weakly compact operators is, as well as the class of compact operators, a fundamental operator ideal. They were investigated strongly in the last twenty years. In this survey, we have collected and ordered some of this (partly very new) knowledge. We have also included some comments, remarks and examples. The class of weakly compact operators is, as well as the class of compact operators, a fundamental operator ideal. They were investigated strongly in the last twenty years. I...

  20. Compact stellarators as reactors

    Lyon, J.F.; Valanju, P.; Zarnstorff, M.C.; Hirshman, S.; Spong, D.A.; Strickler, D.; Williamson, D.E.; Ware, A.

    2001-01-01

    Two types of compact stellarators are examined as reactors: two- and three-field-period (M=2 and 3) quasi-axisymmetric devices with volume-average =4-5% and M=2 and 3 quasi-poloidal devices with =10-15%. These low-aspect-ratio stellarator-tokamak hybrids differ from conventional stellarators in their use of the plasma-generated bootstrap current to supplement the poloidal field from external coils. Using the ARIES-AT model with B max =12T on the coils gives Compact Stellarator reactors with R=7.3-8.2m, a factor of 2-3 smaller R than other stellarator reactors for the same assumptions, and neutron wall loadings up to 3.7MWm -2 . (author)

  1. Compact torsatron reactors

    Lyon, J.F.; Carreras, B.A.; Lynch, V.E.; Tolliver, J.S.; Sviatoslavsky, I.N.

    1988-05-01

    Low-aspect-ratio torsatron configurations could lead to compact stellarator reactors with R 0 = 8--11m, roughly one-half to one-third the size of more conventional stellarator reactor designs. Minimum-size torsatron reactors are found using various assumptions. Their size is relatively insensitive to the choice of the conductor parameters and depends mostly on geometrical constraints. The smallest size is obtained by eliminating the tritium breeding blanket under the helical winding on the inboard side and by reducing the radial depth of the superconducting coil. Engineering design issues and reactor performance are examined for three examples to illustrate the feasibility of this approach for compact reactors and for a medium-size (R 0 ≅ 4 m,/bar a/ /approx lt/ 1 m) copper-coil ignition experiment. 26 refs., 11 figs., 7 tabs

  2. Compact Spreader Schemes

    Placidi, M.; Jung, J. -Y.; Ratti, A.; Sun, C.

    2014-07-25

    This paper describes beam distribution schemes adopting a novel implementation based on low amplitude vertical deflections combined with horizontal ones generated by Lambertson-type septum magnets. This scheme offers substantial compactness in the longitudinal layouts of the beam lines and increased flexibility for beam delivery of multiple beam lines on a shot-to-shot basis. Fast kickers (FK) or transverse electric field RF Deflectors (RFD) provide the low amplitude deflections. Initially proposed at the Stanford Linear Accelerator Center (SLAC) as tools for beam diagnostics and more recently adopted for multiline beam pattern schemes, RFDs offer repetition capabilities and a likely better amplitude reproducibility when compared to FKs, which, in turn, offer more modest financial involvements both in construction and operation. Both solutions represent an ideal approach for the design of compact beam distribution systems resulting in space and cost savings while preserving flexibility and beam quality.

  3. Compact fusion reactors

    CERN. Geneva

    2015-01-01

    Fusion research is currently to a large extent focused on tokamak (ITER) and inertial confinement (NIF) research. In addition to these large international or national efforts there are private companies performing fusion research using much smaller devices than ITER or NIF. The attempt to achieve fusion energy production through relatively small and compact devices compared to tokamaks decreases the costs and building time of the reactors and this has allowed some private companies to enter the field, like EMC2, General Fusion, Helion Energy, Lawrenceville Plasma Physics and Lockheed Martin. Some of these companies are trying to demonstrate net energy production within the next few years. If they are successful their next step is to attempt to commercialize their technology. In this presentation an overview of compact fusion reactor concepts is given.

  4. Compact nuclear fuel storage

    Kiselev, V.V.; Churakov, Yu.A.; Danchenko, Yu.V.; Bylkin, B.K.; Tsvetkov, S.V.

    1983-01-01

    Different constructions of racks for compact storage of spent fuel assemblies (FA) in ''coolin''g pools (CP) of NPPs with the BWR and PWR type reactors are described. Problems concerning nuclear and radiation safety and provision of necessary thermal conditions arising in such rack design are discussed. It is concluded that the problem of prolonged fuel storage at NPPs became Very actual for many countries because of retapdation of the rates of fuel reprocessing centers building. Application of compact storage racks is a promising solution of the problem of intermediate FA storage at NPPs. Such racks of stainless boron steel and with neutron absorbers in the from of boron carbide panels enable to increase the capacity of the present CP 2-2.6 times, and the period of FA storage in them up to 5-10 years

  5. Analysis of laboratory compaction methods of roller compacted concrete

    Trtík, Tomáš; Chylík, Roman; Bílý, Petr; Fládr, Josef

    2017-09-01

    Roller-Compacted Concrete (RCC) is an ordinary concrete poured and compacted with machines typically used for laying of asphalt road layers. One of the problems connected with this technology is preparation of representative samples in the laboratory. The aim of this work was to analyse two methods of preparation of RCC laboratory samples with bulk density as the comparative parameter. The first method used dynamic compaction by pneumatic hammer. The second method of compaction had a static character. The specimens were loaded by precisely defined force in laboratory loading machine to create the same conditions as during static rolling (in the Czech Republic, only static rolling is commonly used). Bulk densities obtained by the two compaction methods were compared with core drills extracted from real RCC structure. The results have shown that the samples produced by pneumatic hammer tend to overestimate the bulk density of the material. For both compaction methods, immediate bearing index test was performed to verify the quality of compaction. A fundamental difference between static and dynamic compaction was identified. In static compaction, initial resistance to penetration of the mandrel was higher, after exceeding certain limit the resistance was constant. This means that the samples were well compacted just on the surface. Specimens made by pneumatic hammer actively resisted throughout the test, the whole volume was uniformly compacted.

  6. Compaction of cereal grain

    Wychowaniec, J.; Griffiths, I.; Gay, A.; Mughal, A.

    2013-01-01

    We report on simple shaking experiments to measure the compaction of a column of Firth oat grain. Such grains are elongated anisotropic particles with a bimodal polydispersity. In these experiments, the particle configurations start from an initially disordered, low-packing-fraction state and under vertical shaking evolve to a dense state with evidence of nematic-like structure at the surface of the confining tube. This is accompanied by an increase in the packing fraction of the grain.

  7. Compact nuclear reactor

    Juric, S.I.

    1975-01-01

    A compact nuclear reactor of the pressurized-water variety is described which has two separate parts separably engageable for ease of inspection, maintenance and repair. One of the parts is a pressure vessel having an active core and the other of the parts is a closure adapted on its lower surface with an integral steam generator. An integral pump, external pressurizer and control rods are provided which communicate with the active core when engaged to form a total unit. (U.S.)

  8. Compact power reactor

    Wetch, J.R.; Dieckamp, H.M.; Wilson, L.A.

    1978-01-01

    There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector

  9. CMS (Compact Muon Solenoid)

    Anon.

    1995-01-01

    The milestone workshops on LHC experiments in Aachen in 1990 and at Evian in 1992 provided the first sketches of how LHC detectors might look. The concept of a compact general-purpose LHC experiment based on a solenoid to provide the magnetic field was first discussed at Aachen, and the formal Expression of Interest was aired at Evian. It was here that the Compact Muon Solenoid (CMS) name first became public. Optimizing first the muon detection system is a natural starting point for a high luminosity (interaction rate) proton-proton collider experiment. The compact CMS design called for a strong magnetic field, of some 4 Tesla, using a superconducting solenoid, originally about 14 metres long and 6 metres bore. (By LHC standards, this warrants the adjective 'compact'.) The main design goals of CMS are: 1 - a very good muon system providing many possibilities for momentum measurement (physicists call this a 'highly redundant' system); 2 - the best possible electromagnetic calorimeter consistent with the above; 3 - high quality central tracking to achieve both the above; and 4 - an affordable detector. Overall, CMS aims to detect cleanly the diverse signatures of new physics by identifying and precisely measuring muons, electrons and photons over a large energy range at very high collision rates, while also exploiting the lower luminosity initial running. As well as proton-proton collisions, CMS will also be able to look at the muons emerging from LHC heavy ion beam collisions. The Evian CMS conceptual design foresaw the full calorimetry inside the solenoid, with emphasis on precision electromagnetic calorimetry for picking up photons. (A light Higgs particle will probably be seen via its decay into photon pairs.) The muon system now foresaw four stations. Inner tracking would use silicon microstrips and microstrip gas chambers, with over 10 7 channels offering high track finding efficiency. In the central CMS barrel, the tracking elements are

  10. Compact Information Representations

    2016-08-02

    Department of Defense, Executive Services, Directorate (0704-0188).   Respondents should be aware that notwithstanding any other provision of law, no person...which lies in the mission of AFOSR. 15.  SUBJECT TERMS sparse sampling , principal components analysis 16.  SECURITY CLASSIFICATION OF: 17...approved for public release Contents 1 Training for Ph.D. Students and Postdoc Researchers 2 2 Papers 2 3 Summary of Proposed Research: Compact

  11. Blanket/first wall challenges and required R&D on the pathway to DEMO

    Abdou, Mohamed, E-mail: abdou@fusion.ucla.edu; Morley, Neil B.; Smolentsev, Sergey; Ying, Alice; Malang, Siegfried; Rowcliffe, Arthur; Ulrickson, Mike

    2015-11-15

    The breeding blanket with integrated first wall (FW) is the key nuclear component for power extraction, tritium fuel sustainability, and radiation shielding in fusion reactors. The ITER device will address plasma burn physics and plasma support technology, but it does not have a breeding blanket. Current activities to develop “roadmaps” for realizing fusion power recognize the blanket/FW as one of the principal remaining challenges. Therefore, a central element of the current planning activities is focused on the question: what are the research and major facilities required to develop the blanket/FW to a level which enables the design, construction and successful operation of a fusion DEMO? The principal challenges in the development of the blanket/FW are: (1) the Fusion Nuclear Environment – a multiple-field environment (neutrons, heat/particle fluxes, magnetic field, etc.) with high magnitudes and steep gradients and transients; (2) Nuclear Heating in a large volume with sharp gradients – the nuclear heating drives most blanket phenomena, but accurate simulation of this nuclear heating can be done only in a DT-plasma based facility; and (3) Complex Configuration with blanket/first wall/divertor inside the vacuum vessel – the consequence is low fault tolerance and long repair/replacement time. These blanket/FW development challenges result in critical consequences: (a) non-fusion facilities (laboratory experiments) need to be substantial to simulate multiple fields/multiple effects and must be accompanied by extensive modeling; (b) results from non-fusion facilities will be limited and will not fully resolve key technical issues. A DT-plasma based fusion nuclear science facility (FNSF) is required to perform “multiple effects” and “integrated” experiments in the fusion nuclear environment; and (c) the Reliability/Availability/Maintainability/Inspectability (RAMI) of fusion nuclear components is a major challenge and is one of the primary reasons

  12. Blanket/first wall challenges and required R&D on the pathway to DEMO

    Abdou, Mohamed; Morley, Neil B.; Smolentsev, Sergey; Ying, Alice; Malang, Siegfried; Rowcliffe, Arthur; Ulrickson, Mike

    2015-01-01

    The breeding blanket with integrated first wall (FW) is the key nuclear component for power extraction, tritium fuel sustainability, and radiation shielding in fusion reactors. The ITER device will address plasma burn physics and plasma support technology, but it does not have a breeding blanket. Current activities to develop “roadmaps” for realizing fusion power recognize the blanket/FW as one of the principal remaining challenges. Therefore, a central element of the current planning activities is focused on the question: what are the research and major facilities required to develop the blanket/FW to a level which enables the design, construction and successful operation of a fusion DEMO? The principal challenges in the development of the blanket/FW are: (1) the Fusion Nuclear Environment – a multiple-field environment (neutrons, heat/particle fluxes, magnetic field, etc.) with high magnitudes and steep gradients and transients; (2) Nuclear Heating in a large volume with sharp gradients – the nuclear heating drives most blanket phenomena, but accurate simulation of this nuclear heating can be done only in a DT-plasma based facility; and (3) Complex Configuration with blanket/first wall/divertor inside the vacuum vessel – the consequence is low fault tolerance and long repair/replacement time. These blanket/FW development challenges result in critical consequences: (a) non-fusion facilities (laboratory experiments) need to be substantial to simulate multiple fields/multiple effects and must be accompanied by extensive modeling; (b) results from non-fusion facilities will be limited and will not fully resolve key technical issues. A DT-plasma based fusion nuclear science facility (FNSF) is required to perform “multiple effects” and “integrated” experiments in the fusion nuclear environment; and (c) the Reliability/Availability/Maintainability/Inspectability (RAMI) of fusion nuclear components is a major challenge and is one of the primary reasons

  13. Prospective performances in JT-60SA towards the ITER and DEMO relevant plasmas

    Tamai, H.; Fujita, T.; Kikuchi, M.

    2006-01-01

    JT-60SA, the former JT-60SC and NCT, a superconducting tokamak positioned as the satellite machine of ITER, collaborating with Japan and EU fusion community, aims at contribution to ITER and DEMO through the demonstration of advanced plasma operation scenario and the plasma applicability test with advanced materials. After the discussions in JA-EU Satellite Tokamak Working Group in 2005, the increased heating power, higher heat removal capacity for the plasma facing components, improvement of the radiation shielding, the remote handling system for the maintenance of in-vessel components, and related equipment are planed to be additionally installed. With such full equipment towards the increased heating power of 41 MW (34 MW-NBI and 7 MW-ECH) with 100 s, the prospective plasma performances, analysed by the equilibrium and transport analysis codes, are rather improved in the view point of the contribution to ITER and DEMO relevant research. Accessibility for higher heating power in a higher density region enables the lower normalized Larmor radius and normalized collision frequency close to the reactor relevant plasma with the ITER-similar configuration of single null divertor plasma with the aspect ratio of A = 3.1, elongation of k95 = 1.7, triangularity of d95 (q95) in the plasma current of I p = 3.5 MA, toroidal magnetic field of B T = 2.59 T and the major radius of Rp=3.16 m. The increases in the electron temperature, beam driven and bootstrap current fraction by the increase of the power of Negative ion based NBI (10 MW) reduce the loop voltage and contribute to increase the maximum plasma current of ITER similar shape. Full non-inductive current drive controllability is also extended into the high density and high plasma current operation towards high beta plasma. Flexibility in aspect ratio and shape parameter is kept the same as NCT, i.e. a double null divertor plasma with A = 2.6, k95 = 1.83, d95 = 0.57, I p = 5.5 MA, B T = 2.72 T, and R p = 3.01 m which

  14. Diffusion through statically compacted clay

    Ho, C.L.; Shebl, M.A.A.

    1994-01-01

    This paper presents experimental work on the effect of compaction on contaminant flow through clay liners. The experimental program included evaluation of soil properties, compaction, permeability and solute diffusion. A permeameter was built of non reactive materials to test samples compacted at different water contents and compactive efforts. The flow of a permeating solute, LiCl, was monitored. Effluent samples were collected for solute concentration measurements. The concentrations were measured by performing atomic adsorption tests. The analyzed results showed different diffusion characteristics when compaction conditions changed. At each compactive effort, permeability decreased as molding water content increased. Consequently, transit time (measured at relative concentration 50%) increased and diffusivity decreased. As compactive effort increased for soils compacted dry of optimum, permeability and diffusion decreased. On the other hand, as compactive effort increased for soils compacted wet of optimum, permeability and diffusivity increased. Tortuosity factor was indirectly measured from the diffusion and retardation rate. Tortuosity factor also decreased as placement water content was increased from dry of optimum to wet of optimum. Then decreases were more pronounced for low compactive effort tests. 27 refs., 7 figs., 5 tabs

  15. MECHANICS OF DYNAMIC POWDER COMPACTION PROCESS

    Nurettin YAVUZ

    1996-01-01

    In recent years, interest in dynamic compaction methods of metal powders has increased due to the need to improve compaction properties and to increase production rates of compacts. In this paper, review of dynamic and explosive compaction of metal powders are given. An attempt is made to get a better understanding of the compaction process with the mechanicis of powder compaction.

  16. Japanese contributions to ITER testing program of solid breeder blankets for DEMO

    Kuroda, Toshimasa; Yoshida, Hiroshi; Takatsu, Hideyuki; Maki, Koichi; Mori, Seiji; Kobayashi, Takeshi; Suzuki, Tatsushi; Hirata, Shingo; Miura, Hidenori.

    1991-04-01

    ITER Conceptual Design Activity (CDA), which has been conducted by four parties (Japan, EC, USA and USSR) since May 1988, has been finished on December 1990 with a great achievement of international design work of the integrated fusion experimental reactor. Numerous issues of physics and technology have been clarified for providing a framework of the next phase of ITER (Engineering Design Activity; EDA). Establishment of an ITER testing program, which includes technical test issues of neutronics, solid breeder blankets, liquid breeder blankets, plasma facing components, and materials, has been one of the goals of the CDA. This report describes Japanese proposal for the testing program of DEMO/power reactor blanket development. For two concepts of solid breeder blanket (helium-cooled and water-cooled), identification of technical issues, scheduling of test program, and conceptual design of test modules including required test facility such as cooling and tritium recovery systems have been carried out as the Japanese contribution to the CDA. (author)

  17. Behaviors of impurity in ITER and DEMOs using BALDUR integrated predictive modeling code

    Onjun, Thawatchai; Buangam, Wannapa; Wisitsorasak, Apiwat

    2015-01-01

    The behaviors of impurity are investigated using self-consistent modeling of 1.5D BALDUR integrated predictive modeling code, in which theory-based models are used for both core and edge region. In these simulations, a combination of NCLASS neoclassical transport and Multi-mode (MMM95) anomalous transport model is used to compute a core transport. The boundary is taken to be at the top of the pedestal, where the pedestal values are described using a theory-based pedestal model. This pedestal temperature model is based on a combination of magnetic and flow shear stabilization pedestal width scaling and an infinite-n ballooning pressure gradient model. The time evolution of plasma current, temperature and density profiles is carried out for ITER and DEMOs plasmas. As a result, the impurity behaviors such as impurity accumulation and impurity transport can be investigated. (author)

  18. DeMO: An Ontology for Discrete-event Modeling and Simulation

    Silver, Gregory A; Miller, John A; Hybinette, Maria; Baramidze, Gregory; York, William S

    2011-01-01

    Several fields have created ontologies for their subdomains. For example, the biological sciences have developed extensive ontologies such as the Gene Ontology, which is considered a great success. Ontologies could provide similar advantages to the Modeling and Simulation community. They provide a way to establish common vocabularies and capture knowledge about a particular domain with community-wide agreement. Ontologies can support significantly improved (semantic) search and browsing, integration of heterogeneous information sources, and improved knowledge discovery capabilities. This paper discusses the design and development of an ontology for Modeling and Simulation called the Discrete-event Modeling Ontology (DeMO), and it presents prototype applications that demonstrate various uses and benefits that such an ontology may provide to the Modeling and Simulation community. PMID:22919114

  19. Qualification of MHD effects in dual-coolant DEMO blanket and approaches to their modelling

    Mas de les Valls, E.; Batet, L.; Medina, V. de; Fradera, J.; Sedano, L.A.

    2011-01-01

    Design refinements of vertical insulated banana-shaped liquid metal channels are being considered as a progress of conceptual design of dual-coolant liquid metal blankets (DEMO specifications). Among them: (a) optimised channel geometry and (b) improvements on flow channel inserts. Progress of channel conceptual design is conducted in parallel with underlying physics of MHD models in diverse aspects: (1) MHD models, (2) MHD turbulence, (3) LM buoyancy effects, (4) three-dimensional flows, and (5) LM/FCI/wall electrical and thermal coupling; in order to progress on common liquid metal flow characterisation, pressure drop and three-dimensional flows. The analyses are assumed as extension of those previous carried out for the DCLL blankets for new design refinements. At the present stage of the conceptual design progress, a preliminary thermofluid MHD study is of crucial interest for further design improvements and future detailed modelling. The paper overviews the ongoing modelling studies, making model refinements explicit, and anticipates some modelling results.

  20. LBNO-DEMO (WA105): a large demonstrator of the Liquid Argon double phase TPC

    Trzaska, Wladyslaw Henryk

    2015-01-01

    LBNO-DEMO (WA105) is a large demonstrator of the double phase liquid argon TPC intended to develop and test the main elements of the GLACIER-based design for the purpose of scaling it up to the 10–50 kton size needed for Long Baseline Neutrino Oscillation studies. The crucial components of the design are: ultra-high argon purity in non-evacuable tank, long drifts, very high drift voltages, large area Micro Pattern Gas Detectors, and cold preamplifiers. The active volume of the demonstrator is 666 m3 (approximately 300t). WA105 is under construction at CERN and will be exposed to charged particle beams (0.5-20 GeV/c) in the North Area in 2018. The data will provide the necessary calibration of the detector performance and benchmark reconstruction algorithms. This project is a crucial milestone for the long baseline neutrino program, including projects like LBNO and DUNE.

  1. Demo - Talk2Me: A Framework for Device–to–Device Augmented Reality Social Network

    Shu, Jiayu; Kosta, Sokol; Zheng, Rui

    2018-01-01

    –to–Device fashion. When a user looks at nearby persons through her camera–enabled wearable devices (e.g., Google Glass), the framework automatically extracts the face–signature of the person of interest, compares it with the previously captured signatures, and presents the information shared by this person......In this demo, we present Talk2Me, an augmented reality social network framework that enables users to disseminate information in a distributed way and view others’ information instantly. Talk2Me advertises users’ messages, together with their face–signatures, to every nearby device in a Device...... to the user. We design a lightweight and yet accurate face recognition algorithm, together with an efficient distributed dissemination protocol. We integrate their implementations in an Android prototype....

  2. Operation and first results of the NEXT-DEMO prototype using a silicon photomultiplier tracking array

    Álvarez, V; Cárcel, S; Cervera, A; Díaz, J; Ferrario, P; Gil, A; Borges, F I G; Conde, C A N; Dias, T H V T; Fernandes, L M P; Freitas, E D C; Castel, J; Cebrián, S; Dafni, T; Egorov, M; Gehman, V M; Goldschmidt, A; Esteve, R; Evtoukhovitch, P; Ferreira, A L

    2013-01-01

    NEXT-DEMO is a high-pressure xenon gas TPC which acts as a technological test-bed and demonstrator for the NEXT-100 neutrinoless double beta decay experiment. In its current configuration the apparatus fully implements the NEXT-100 design concept. This is an asymmetric TPC, with an energy plane made of photomultipliers and a tracking plane made of silicon photomultipliers (SiPM) coated with TPB. The detector in this new configuration has been used to reconstruct the characteristic signature of electrons in dense gas, demonstrating the ability to identify the MIP and ''blob'' regions. Moreover, the SiPM tracking plane allows for the definition of a large fiducial region in which an excellent energy resolution of 1.82% FWHM at 511 keV has been measured (a value which extrapolates to 0.83% at the xenon Q ββ )

  3. Design study of superconducting coils for the fusion DEMO plant at JAERI

    Isono, T.; Koizumi, N.; Okuno, K.; Kurihara, R.; Nishio, S.; Tobita, K.

    2006-01-01

    A design study of the TF coil for the fusion DEMO plant at JAERI is in progress. A major issue is to estimate the maximum fields generated by the TF coils for three tokamak options and two conductor options. Three tokamak options are proposed varying the aspect ratio and the role of the CS coil. Two kinds of conductors using advanced superconducting materials are candidates for the TF coils: Nb 3 Al and high temperature superconductor (HTS). In order to evaluate achievable magnetic fields, a simple method was adopted to calculate mechanical properties. The estimated maximum fields are 17-20 T by the HTS conductor and 16-17 T by the Nb 3 Al conductor. There is a possibility of a 0.7 T enhancement using grading of Nb 3 Al winding

  4. Diffusion bonding of reduced activation ferritic steel F82H for demo blanket application

    Kurasawa, T.; Tamura, M.

    1996-01-01

    A reduced activation ferritic steel, a grade F82H developed by JAERI, is a promising candidate structural material for the blanket and the first wall of DEMO reactors. In the present study, diffusion bonding of F82H has been investigated to develop the fabrication procedures of the blanket box and the first wall panel with cooling channels embedded by F82H. The parameters examined are the bonding temperature (810-1050 C), bonding pressure (2-10 MPa) and roughness of the bonding surface (0.5-12.8 μR max ), and metallurgical examination and mechanical tests of the diffusion bonded joints have been conducted. From the tests, sufficient bonding was obtained under the temperatures of 840-1 050 C (compressive stress of 3-12 MPa), and it was found that heat treatment following diffusion bonding is essential to obtain the mechanical properties similar to that of the base metal. (orig.)

  5. Development of a zonal applicability tool for remote handling equipment in DEMO

    Madzharov, Vladimir, E-mail: vladimir.madzharov@kit.edu [Karlsruhe Institute of Technology, Institute for Material Handling and Logistics, Karlsruhe (Germany); Mittwollen, Martin [Karlsruhe Institute of Technology, Institute for Material Handling and Logistics, Karlsruhe (Germany); Leichtle, Dieter [Fusion for Energy F4E, Barcelona (Spain); Hermon, Gary [Culham Center for Fusion Energy, Culham Science Centre, OX14 3DB Abingdon (United Kingdom)

    2015-10-15

    Highlights: • Radiation-hardness assessment of remote handling (RH) components used in DEMO. • A radiation assessment tool for supporting remote handling engineers. • Connecting data from the radiation field analysis to the radiation hardness data. • Output is the expected lifetime of the selected RH component used for maintenance. - Abstract: A radiation-induced damage caused by the ionizing radiation can induce a malfunctioning of the remote handling equipment (RHE) used during maintenance in fusion power plants, other nuclear power stations and high-energy accelerators facilities like e.g. IFMIF. Therefore to achieve a sufficient length of operational time inside future fusion power plants, a suitable radiation tolerant RHE for maintenance operations in radiation environments is inevitably required. To assess the influence of the radiation on remote handling equipment (RHE), an investigation about radiation hardness assessment of typically used RHE components, has been performed. Additionally, information about the absorbed total dose that every component can withstand before failure was collected. Furthermore, the development of a zonal applicability tool for supporting RHE designers has been started using Excel VBA. The tool connects the data from the radiation field analysis (3-D radiation map) to the radiation hardness data of the planned RHE for DEMO remote maintenance. The intelligent combination of the available information for the radiation behaviour and radiation level at certain time and certain location may help with the taking of decisions about the application of RHE in radiation environment. The user inputs the following parameters: the specific device used in the RHE, the planned location and the maintenance period. The output is the expected lifetime of the selected RHE component at the given location and maintenance period. Planned action times have to be also considered. After having all the parameters it can be decided, if specific RHE

  6. Study of dynamic amplification factor of DEMO blanket caused by a gap at the supporting key

    Frosi, Paolo; Mazzone, Giuseppe

    2015-01-01

    Highlights: • With the preliminary hypothesis established, the dynamic displacements are not so high and the state of stress (not reported) does not exhibit large region with plastic strain. • The dynamic displacements show a certain dependency from the mesh adopted, and the geometry chosen. • The energy (kinetic or strain) of the whole structure gives useful information about the key behavior during impact. • In order to better understand the overall phenomenon other details (non-linear material, better evaluation of damping, other disruption rise-times and so on. - Abstract: Among the design activities of the in vessel components for DEMO promoted by European Fusion Development Agreement (EFDA) organization, this work deals with the gap required at the supporting keys of the blanket. Due to its higher operating temperatures compared to the vacuum vessel (VV) ones, this gap will increase during operation. The electro magnetic (EM) loads due to fast disruptions occur on a short time and might accelerate the blanket significantly before it touches the supporting keys, causing an impact of the blanket itself onto the keys. Depending on their stiffness, the EM loads with their short time scale could excite the structure's natural frequencies, causing dynamic amplification. Both phenomena (impact and dynamic amplification) can cause stresses in the structure significantly higher than the static ones. This work develops a finite element model of DEMO blanket to study its non-linear transient dynamic behavior under impact loadings. A VV sector, the ribs between the inner and outer VV, the backward manifolds and the supporting keys of the blanket have been modeled. The analyses have been performed with Abaqus [1] and Ansys [2] FEM codes focused on the displacements of the keys in their housing on the blanket. The dynamic amplification factor has been evaluated as the ratio of dynamic to static displacements in meaningful points of the structure for a growing gap

  7. Status of EC solid breeder blanket designs and R and D for demo fusion reactors

    Proust, E.; Anzidei, L.; Moons, F.

    1994-01-01

    Within the European Community Fusion Technology Program two solid breeder blankets for a DEMO reactor are being developed. The two blankets have various features in common: helium as coolant and as tritium purge gas, the martensitic steel MANET as structural material and beryllium as neutron multiplier. The configurations of the two blankets are however different: in the B.I.T. (Breeder Inside Tube) concept the breeder materials are LiAlO 2 or Li 2 ZrO 3 in the form of annular pellets contained in tubes surrounded by beryllium blocks, the coolant helium being outside the tubes, whereas in the B.O.T. (Breeder out of Tube) the breeder and multiplier material are Li 4 SiO 4 and beryllium pebbles forming a mixed bed placed outside the tubes containing the coolant helium. The main critical issues for both blankets are the behavior of the breeder ceramics and of beryllium under irradiation and the tritium control. Other issues are the low temperature irradiation induced embrittlement of MANET, the mechanical effects caused by major plasma disruptions, and safety and reliability. The R and D work concentrate on these issues. The development of martensitic steels including MANET is part of a separate program. Breeder ceramics and beryllium irradiations have been so far performed for conditions which do not cover the peak values injected in the DEMO blankets. Further irradiations in thermal reactors and in fast reactors, especially for beryllium, are required. An effective tritium control requires the development of permeation barriers and/or of methods of oxidation of the tritium in the main helium cooling systems. First promising results have been obtained also in field of mechanical effects from plasma disruptions and safety and reliability, however further work is required in the reliability field and to validate the codes for the calculations of the plasma disruption effects. (authors). 8 figs., 2 tabs., 53 refs

  8. [The theory of the demographic transition as a reference for demo-economic models].

    Genne, M

    1981-01-01

    The aim of the theory of demographic transition (TTD) is to better understand the behavior and interrelationship of economic and demographic variables. There are 2 types of demo-economic models: 1) the malthusian models, which consider demographic variables as pure exogenous variables, and 2) the neoclassical models, which consider demographic variables as strictly endogenous. If TTD can explore the behavior of exogenous and endogenous demographic variables, it cannot demonstrate neither the relation nor the order of causality among the various demographic and economic variables, but it is simply the theoretical framework of a complex social and economic phenomenon which started in Europe in the 19th Century, and which today can be extended to developing countries. There are 4 stages in the TTD; the 1st stage is characterized by high levels of fecundity and mortality; the 2nd stage is characterized by high fecundity levels and declining mortality levels; the 3rd stage is characterized by declining fecundity levels and low mortality levels; the 4th stage is characterized by low fertility and mortality levels. The impact of economic variables over mortality and birth rates is evident for mortality rates, which decline earlier and at a greater speed than birth rates. According to reliable mathematical predictions, around the year 1987 mortality rates in developing countries will have reached the low level of European countries, and growth rate will be only 1.5%. If the validity of demo-economic models has not yet been established, TTD has clearly shown that social and economic development is the factor which influences demographic expansion.

  9. Tokamak DEMO-FNS: Concept of magnet system and vacuum chamber

    Azizov, E. A., E-mail: Azizov-EA@nrcki.ru; Ananyev, S. S. [National Research Center Kurchatov Institute (Russian Federation); Belyakov, V. A.; Bondarchuk, E. N.; Voronova, A. A. [D.V. Efremov Scientific Research Institute of Electrophysical Apparatus (Russian Federation); Golikov, A. A. [National Research Center Kurchatov Institute (Russian Federation); Goncharov, P. R. [Peter the Great St. Petersburg Polytechnic University (Russian Federation); Dnestrovskij, A. Yu. [National Research Center Kurchatov Institute (Russian Federation); Zapretilina, E. R. [D.V. Efremov Scientific Research Institute of Electrophysical Apparatus (Russian Federation); Ivanov, D. P. [National Research Center Kurchatov Institute (Russian Federation); Kavin, A. A.; Kedrov, I. V. [D.V. Efremov Scientific Research Institute of Electrophysical Apparatus (Russian Federation); Klischenko, A. V.; Kolbasov, B. N. [National Research Center Kurchatov Institute (Russian Federation); Krasnov, S. V. [D.V. Efremov Scientific Research Institute of Electrophysical Apparatus (Russian Federation); Krylov, A. I. [National Research Center Kurchatov Institute (Russian Federation); Krylov, V. A.; Kuzmin, E. G. [D.V. Efremov Scientific Research Institute of Electrophysical Apparatus (Russian Federation); Kuteev, B. V. [National Research Center Kurchatov Institute (Russian Federation); Labusov, A. N. [D.V. Efremov Scientific Research Institute of Electrophysical Apparatus (Russian Federation); and others

    2016-12-15

    The level of knowledge accumulated to date in the physics and technologies of controlled thermonuclear fusion (CTF) makes it possible to begin designing fusion—fission hybrid systems that would involve a fusion neutron source (FNS) and which would admit employment for the production of fissile materials and for the transmutation of spent nuclear fuel. Modern Russian strategies for CTF development plan the construction to 2023 of tokamak-based demonstration hybrid FNS for implementing steady-state plasma burning, testing hybrid blankets, and evolving nuclear technologies. Work on designing the DEMO-FNS facility is still in its infancy. The Efremov Institute began designing its magnet system and vacuum chamber, while the Kurchatov Institute developed plasma-physics design aspects and determined basic parameters of the facility. The major radius of the plasma in the DEMO-FNS facility is R = 2.75 m, while its minor radius is a = 1 m; the plasma elongation is k{sub 95} = 2. The fusion power is P{sub FUS} = 40 MW. The toroidal magnetic field on the plasma-filament axis is B{sub t0} = 5 T. The plasma current is I{sub p} = 5 MA. The application of superconductors in the magnet system permits drastically reducing the power consumed by its magnets but requires arranging a thick radiation shield between the plasma and magnet system. The central solenoid, toroidal-field coils, and poloidal-field coils are manufactured from, respectively, Nb{sub 3}Sn, NbTi and Nb{sub 3}Sn, and NbTi. The vacuum chamber is a double-wall vessel. The space between the walls manufactured from 316L austenitic steel is filled with an iron—water radiation shield (70% of stainless steel and 30% of water).

  10. Limitations of transient power loads on DEMO and analysis of mitigation techniques

    Maviglia, F., E-mail: francesco.maviglia@euro-fusion.org [EUROfusion Consortium, PPPT Department, Boltzmannstr. 2, Garching (Germany); Consorzio CREATE, University Napoli Federico II – DIETI, 80125 Napoli (Italy); Federici, G. [EUROfusion Consortium, PPPT Department, Boltzmannstr. 2, Garching (Germany); Strohmayer, G. [Max-Planck-Institut fur Plasmaphysik, Boltzmannstr. 2, Garching (Germany); Wenninger, R. [EUROfusion Consortium, PPPT Department, Boltzmannstr. 2, Garching (Germany); Max-Planck-Institut fur Plasmaphysik, Boltzmannstr. 2, Garching (Germany); Bachmann, C. [EUROfusion Consortium, PPPT Department, Boltzmannstr. 2, Garching (Germany); Albanese, R. [Consorzio CREATE, University Napoli Federico II – DIETI, 80125 Napoli (Italy); Ambrosino, R. [Consorzio CREATE University Napoli Parthenope, Naples (Italy); Li, M. [Max-Planck-Institut fur Plasmaphysik, Boltzmannstr. 2, Garching (Germany); Loschiavo, V.P. [Consorzio CREATE, University Napoli Federico II – DIETI, 80125 Napoli (Italy); You, J.H. [Max-Planck-Institut fur Plasmaphysik, Boltzmannstr. 2, Garching (Germany); Zani, L. [CEA, IRFM, F-13108 St Paul-Lez-Durance (France)

    2016-11-01

    Highlights: • A parametric thermo-hydraulic analysis of the candidate DEMO divertor is presented. • The operational space assessment is presented under static and transient heat loads. • Strike points sweeping is analyzed as a divertor power exhaust mitigation technique. • Results are presented on sweeping installed power required, AC losses and thermal fatigue. - Abstract: The present European standard DEMO divertor target technology is based on a water-cooled tungsten mono-block with a copper alloy heat sink. This paper presents the assessment of the operational space of this technology under static and transient heat loads. A transient thermo-hydraulic analysis was performed using the code RACLETTE, which allowed a broad parametric scan of the target geometry and coolant conditions. The limiting factors considered were the coolant critical heat flux (CHF), and the temperature limits of the materials. The second part of the work is devoted to the study of the plasma strike point sweeping as a mitigation technique for the divertor power exhaust. The RACLETTE code was used to evaluate the impact of a large range of sweeping frequencies and amplitudes. A reduced subset of cases, which complied with the constraints, was benchmarked with a 3D FEM model. A reduction of the heat flux to the coolant, up to a factor ∼4, and lower material temperatures were found for an incident heat flux in the range (15–30) MW/m{sup 2}. Finally, preliminary assessments were performed on the installed power required for the sweeping, the AC losses in the superconductors and thermal fatigue analysis. No evident show stoppers were found.

  11. Gyrotron development at KIT: FULGOR test facility and gyrotron concepts for DEMO

    Schmid, M., E-mail: martin.schmid@kit.edu [Institute for Pulsed Power and Microwave Technology (IHM), Karlsruhe Institute of Technology, Association EURATOM-KIT, Karlsruhe (Germany); Franck, J.; Kalaria, P.; Avramidis, K.A.; Gantenbein, G.; Illy, S. [Institute for Pulsed Power and Microwave Technology (IHM), Karlsruhe Institute of Technology, Association EURATOM-KIT, Karlsruhe (Germany); Jelonnek, J. [Institute for Pulsed Power and Microwave Technology (IHM), Karlsruhe Institute of Technology, Association EURATOM-KIT, Karlsruhe (Germany); Institute of High Frequency Techniques and Electronics (IHE), Karlsruhe Institute of Technology, Association EURATOM-KIT, Karlsruhe (Germany); Pagonakis, I. Gr.; Rzesnicki, T. [Institute for Pulsed Power and Microwave Technology (IHM), Karlsruhe Institute of Technology, Association EURATOM-KIT, Karlsruhe (Germany); Thumm, M. [Institute for Pulsed Power and Microwave Technology (IHM), Karlsruhe Institute of Technology, Association EURATOM-KIT, Karlsruhe (Germany); Institute of High Frequency Techniques and Electronics (IHE), Karlsruhe Institute of Technology, Association EURATOM-KIT, Karlsruhe (Germany)

    2015-10-15

    Highlights: • Substantial extension of the KIT gyrotron test facility FULGOR has started. • FULGOR will be able to test gyrotrons with continuous RF output power up to 4 MW. • Design of 240 GHz gyrotrons for efficient electron cyclotron current drive is progressing. • Output power of 240 GHz gyrotrons with conventional cavity up to 830 kW, with coaxial cavity up to 2 MW is feasible. • Multi-frequency operation with gyrotrons is also possible (170–267 GHz). - Abstract: At the Karlsruhe Institute of Technology (KIT), theoretical and experimental foundations for the development of future gyrotrons for fusion applications are being laid down. This includes the construction of the new Fusion Long Pulse Gyrotron Laboratory (FULGOR) test facility as well as physical design studies towards DEMO-compatible gyrotrons. Initially FULGOR will comprise of a 10 MW CW power supply, a 5 MW water cooling system (upgradeable to 10 MW), a superconducting 10 T magnet, one or two 2 MW ECRH test loads and a new control and data acquisition system for all these elements. The test facility will then be equipped to test the conventional 1 MW or coaxial 2 MW gyrotrons for DEMO, currently under design, as well as possible upgraded gyrotrons for W7-X and ITER. The design of the new high voltage DC power supply (HVDCPS) is flexible enough to handle gyrotrons with 4 MW CW output power (conceivably up to 170 GHz), but also test gyrotrons with higher frequencies (>250 GHz) which, due to physical limitations in the gyrotron design, will require less power but have more stringent demands on voltage stability.

  12. Effect of heat loads on the plasma facing components of demo

    Igitkhanov, Yu., E-mail: juri.igitkhanov@partner.kit.edu [ITEP, Karlsruhe Institute of Technology (Germany); Fetzer, R. [IHM, Karlsruhe Institute of Technology (Germany); Bazylev, B. [INR, Karlsruhe Institute of Technology (Germany)

    2016-11-01

    Highlights: • Under the DEMO1 stationary operation the nominal power fluxes along the magnetic field at the FW blanket modules is expected about 50 MW/m{sup 2}. • In the current design and averaged incident angle about 3–4.5° (similar to ITER) the engineering power load to the FW is expected within 2.5÷3.9 MW/m{sup 2}. • In the case of the unmitigated Type I ELMs unavoidable in the higher confinement H-mode of operation energy load per ELM is about 20 MJ/m{sup 2} along the field line, arriving at a frequency of 0.8 Hz with deposition time of 0.6 ms per each ELM. - Abstract: In this paper we analyse a thermo-hydraulic performance of the first wall blanket module during the stationary DEMO operation with the edge localized mode (ELM). Heat loads are estimated based on scaling arguments and predictions from the peeling-ballooning ELM model. Effect of parallel heat fluxes intersecting with the first wall panels and avoidance of overheating by inclination of the panels are considered. The material temperatures of the W/EUROFER sandwich type module with water cooling stainless steel tube and Cu alloy compliance embedded into EUROFER is calculated by using the MEMOS code. The calculations were carried out indicating the required geometric parameters as well as the cooling conditions which allow keeping materials temperatures within allowable engineering limits. Effect of inclination of the first wall plates to avoid the misalignment problems is considered.

  13. Study of dynamic amplification factor of DEMO blanket caused by a gap at the supporting key

    Frosi, Paolo, E-mail: paolo.frosi@enea.it; Mazzone, Giuseppe

    2015-10-15

    Highlights: • With the preliminary hypothesis established, the dynamic displacements are not so high and the state of stress (not reported) does not exhibit large region with plastic strain. • The dynamic displacements show a certain dependency from the mesh adopted, and the geometry chosen. • The energy (kinetic or strain) of the whole structure gives useful information about the key behavior during impact. • In order to better understand the overall phenomenon other details (non-linear material, better evaluation of damping, other disruption rise-times and so on. - Abstract: Among the design activities of the in vessel components for DEMO promoted by European Fusion Development Agreement (EFDA) organization, this work deals with the gap required at the supporting keys of the blanket. Due to its higher operating temperatures compared to the vacuum vessel (VV) ones, this gap will increase during operation. The electro magnetic (EM) loads due to fast disruptions occur on a short time and might accelerate the blanket significantly before it touches the supporting keys, causing an impact of the blanket itself onto the keys. Depending on their stiffness, the EM loads with their short time scale could excite the structure's natural frequencies, causing dynamic amplification. Both phenomena (impact and dynamic amplification) can cause stresses in the structure significantly higher than the static ones. This work develops a finite element model of DEMO blanket to study its non-linear transient dynamic behavior under impact loadings. A VV sector, the ribs between the inner and outer VV, the backward manifolds and the supporting keys of the blanket have been modeled. The analyses have been performed with Abaqus [1] and Ansys [2] FEM codes focused on the displacements of the keys in their housing on the blanket. The dynamic amplification factor has been evaluated as the ratio of dynamic to static displacements in meaningful points of the structure for a growing

  14. Find the weakest link. A comparison between demographic, genetic and demo-genetic metapopulation extinction times

    Robert Alexandre

    2011-09-01

    Full Text Available Abstract Background While the ultimate causes of most species extinctions are environmental, environmental constraints have various secondary consequences on evolutionary and ecological processes. The roles of demographic, genetic mechanisms and their interactions in limiting the viabilities of species or populations have stirred much debate and remain difficult to evaluate in the absence of demography-genetics conceptual and technical framework. Here, I computed projected times to metapopulation extinction using (1 a model focusing on the effects of species properties, habitat quality, quantity and temporal variability on the time to demographic extinction; (2 a genetic model focusing on the dynamics of the drift and inbreeding loads under the same species and habitat constraints; (3 a demo-genetic model accounting for demographic-genetic processes and feedbacks. Results Results indicate that a given population may have a high demographic, but low genetic viability or vice versa; and whether genetic or demographic aspects will be the most limiting to overall viability depends on the constraints faced by the species (e.g., reduction of habitat quantity or quality. As a consequence, depending on metapopulation or species characteristics, incorporating genetic considerations to demographically-based viability assessments may either moderately or severely reduce the persistence time. On the other hand, purely genetically-based estimates of species viability may either underestimate (by neglecting demo-genetic interactions or overestimate (by neglecting the demographic resilience true viability. Conclusion Unbiased assessments of the viabilities of species may only be obtained by identifying and considering the most limiting processes (i.e., demography or genetics, or, preferentially, by integrating them.

  15. Activation, decay heat, and waste classification studies of the European DEMO concept

    Gilbert, M. R.; Eade, T.; Bachmann, C.; Fischer, U.; Taylor, N. P.

    2017-04-01

    Inventory calculations have a key role to play in designing future fusion power plants because, for a given irradiation field and material, they can predict the time evolution in chemical composition, activation, decay heat, gamma-dose, gas production, and even damage (dpa) dose. For conceptual designs of the European DEMO fusion reactor such calculations provide information about the neutron shielding requirements, maintenance schedules, and waste disposal prospects; thereby guiding future development. Extensive neutron-transport and inventory calculations have been performed for a reference DEMO reactor model with four different tritium-breeding blanket concepts. The results have been used to chart the post-operation variation in activity and decay heat from different vessel components, demonstrating that the shielding performance of the different blanket concepts—for a given blanket thickness—varies significantly. Detailed analyses of the simulated nuclide inventories for the vacuum vessel (VV) and divertor highlight the most dominant radionuclides, potentially suggesting how changes in material composition could help to reduce activity. Minor impurities in the raw composition of W used in divertor tiles, for example, are shown to produce undesirable long-lived radionuclides. Finally, waste classifications, based on UK regulations, and a recycling potential limit, have been applied to estimate the time-evolution in waste masses for both the entire vessel (including blanket modules, VV, divertor, and some ex-vessel components) and individual components, and also to suggest when a particular component might be suitable for recycling. The results indicate that the large mass of the VV will not be classifiable as low level waste on the 100 year timescale, but the majority of the divertor will be, and that both components will be potentially recyclable within that time.

  16. Concept design of DEMO divertor cassette remote handling: Simply supported beam approach

    Mozzillo, Rocco [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy); Di Gironimo, Giuseppei, E-mail: peppe.digironimo@gmail.com [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy); Mäkinen, Harri [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Miccichè, Gioacchino [ENEA – CR Brasimone, I-40032 Camugnano, BO (Italy); Määttä, Timo [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland)

    2017-03-15

    Highlights: • The present work focused on a new approach to the design of DEMO Divertor Cassette Remote Handling Equipment. • The work provides an alternative approach to the design based on the concept of a simply supported beam. • The approach proposed focuses a Divertor Cassette mover that performs the maintenance of the three cassettes at each port. • First rough dimensioning of the main components has been provided and demonstrating the feasibility of the design solutions. • The main idea of the work consisted on a design capable to use knowledge already adopted in industrial contexts. - Abstract: The present work focused on the development of a new approach to the concept design of DEMO Divertor Cassette (DC) Remote Handling Equipment (RHE). The approach is based on three main assumptions: the DC remote handling activities and the equipment shall be simplified as much as possible; technologies well known and consolidated in the industrial context can be adopted also in the nuclear fusion field; the design of the RHE should be based on a simply supported beam approach instead of cantilever approach. In detail, during the maintenance activities the barycentre of the DC is centred with respect to DC supports. This solution could simplify the design of RHE with a consequent reduction of the design and development costs. Moreover also the DC remote handling tasks shall be simplified in order to better manage the DC maintenance processes. For this reason the DC assembly and disassembly process has been simplified dividing the main sequences in basic movements. For each movement a dedicated tool has been conceived.

  17. Evaluation of remote maintenance schemes by plasma equilibrium analysis in Tokamak DEMO reactor

    Utoh, Hiroyasu; Tobita, Kenji; Asakura, Nobuyuki; Sakamoto, Yoshiteru

    2014-01-01

    Highlights: • The remote maintenance schemes in DEMO reactor were evaluated by the plasma equilibrium analysis. • Horizontal sector transport maintenance scheme requires the largest total PF coil current. • The difference of total PF coil current for MHD equilibrium in between the large segmented divertor maintenance and the segmentalized divertor maintenance was about 10%. - Abstract: The remote maintenance schemes in a DEMO reactor are categorized by insertion direction, blanket segmentation, and divertor maintenance scheme, and are quantitatively evaluated by analysing the plasma equilibrium. The positions of the poloidal field (PF) coil are limited by the size of the toroidal field (TF) coil and the maintenance port layout of each remote maintenance scheme. Because the PF coils are located near the larger TF coil and far from the plasma surface, the horizontal sector transport maintenance scheme requires the largest part of total PF coil current, 25% larger than that required for separated sector transport using vertical maintenance ports with segmented divertor maintenance (SDM). In the unsegmented divertor maintenance (UDM) scheme, the total magnetic stored energy in the PF coils at plasma equilibrium is about 30% larger than that stored in the SDM scheme, but the time required for removal and installation of all the divertor cassettes in the UDM scheme is roughly a third of that required in the SDM scheme because the number of divertor cassettes in the UDM scheme is a third of that in the SDM scheme. From the viewpoint of simple maintenance operations, the merit of the UDM scheme has more merit than the SDM scheme

  18. Radioactive waste produced by DEMO and commerical fusion reactors extrapolated from ITER and advanced data bases

    Stacey, W.M.; Hertel, N.E.; Hoffman, E.A.

    1994-01-01

    The potential for providing energy with minimal environmental impact is a powerful motivation for the development of fusion and is the long-term objective of most fusion programs. However, the societal acceptability of magnetic fusion may well be decided in the near-term when decisions are taken on the construction of DEMO to follow ITER (if not when the construction decision is taken on ITER). Component wastes were calculated for DEMOs based on each data base by first calculating reactor sizes needed to satisfy the physics, stress and radiation attenuation requirements, and then calculating component replacement rates based on radiation damage and erosion limits. Then, radioactive inventories were calculated and compared to a number of international criteria for open-quote near-surface close-quote burial. None of the components in either type of design would meet the Japanese LLW criterion ( 3 ) within 10 years of shutdown, although the advanced (V/Li) blanket would do so soon afterwards. The vanadium first wall, divertor and blanket would satisfy the IAEA LLW criterion (<2 mSv/h contact dose) within about 10 years after shutdown, but none of the stainless steel or copper components would. All the components in the advanced data base designs except the stainless steel vacuum vessel and shield readily satisfy the US extended 10CFR61 intruder dose criterion, but none of the components in the open-quotes ITER data baseclose quotes designs do so. It seems unlikely that a stainless steel first wall or a copper divertor plate could satisfy the US (class C) criterion for near surface burial, much less the more stringent international, criteria. On the other hand, the first wall, divertor and blanket of the V/Li system would still satisfy the intruder dose concentration limits even if the dose criterion was reduced by two orders of magnitude

  19. Compact neutron flux monitor

    Madhavi, V.; Phatak, P.R.; Bahadur, C.; Bayala, A.K.; Jakati, R.K.; Sathian, V.

    2003-01-01

    Full text: A compact size neutron flux monitor has been developed incorporating standard boards developed for smart radiation monitors. The sensitivity of the monitors is 0.4cps/nV. It has been tested up to 2075 nV flux with standard neutron sources. It shows convincing results even in high flux areas like 6m away from the accelerator in RMC (Parel) for 106/107 nV. These monitors have a focal and remote display, alarm function with potential free contacts for centralized control and additional provision of connectivity via RS485/Ethernet. This paper describes the construction, working and results of the above flux monitor

  20. Compact ignition experiments

    Angelini, A.; Coppi, B.; Nassi, M.

    1992-01-01

    This paper reports on high magnetic field experiments which can be designed to investigate D-T ignition conditions based on present-day experimental results and theoretical understanding of plasma phenomena. The key machine elements are: large plasma currents, compact dimensions, tight aspect ratios, moderate elongations and significant triangularities of the plasma column. High plasma densities, strong ohmic heating, the needed degree of energy confinement, good plasma purity and robust stability against ideal and resistive instabilities can be achieved simultaneously. The Ignitor design incorporates all these characteristics and involves magnet technology developments, started with the Alcator experiment, that use cryogenically cooled normal conductors

  1. Compact LINAC for deuterons

    Kurennoy, S.S.; O'Hara, J.F.; Rybarcyk, L.J.

    2008-01-01

    We are developing a compact deuteron-beam accelerator up to the deuteron energy of a few MeV based on room-temperature inter-digital H-mode (IH) accelerating structures with the transverse beam focusing using permanent-magnet quadrupoles (PMQ). Combining electromagnetic 3-D modeling with beam dynamics simulations and thermal-stress analysis, we show that IHPMQ structures provide very efficient and practical accelerators for light-ion beams of considerable currents at the beam velocities around a few percent of the speed of light. IH-structures with PMQ focusing following a short RFQ can also be beneficial in the front end of ion linacs.

  2. Compact electron storage rings

    Williams, G.P.

    1987-01-01

    There have been many recent developments in the area of compact storage rings. Such rings would have critical wavelengths of typically 10 A, achieved with beam energies of several hundreds of MeV and superconducting dipole fields of around 5 Tesla. Although the primary motivation for progress in this area is that of commercial x-ray lithography, such sources might be an attractive source for college campuses to operate. They would be useful for many programs in materials science, solid state, x-ray microscopy and other biological areas. We discuss the properties of such sources and review developments around the world, primarily in the USA, japan and W. Germany

  3. Compact synchrotron radiation source

    Liu, N.; Wang, T.; Tian, J.; Lin, Y.; Chen, S.; He, W.; Hu, Y.; Li, Q.

    1985-01-01

    A compact 800 MeV synchrotron radiation source is discussed. The storage ring has a circumference of 30.3 m, two 90 degree and four 45 degree bending magnet sections, two long straight sections and four short straight sections. The radius of the bending magnet is 2.224m. The critical wave length is 24A. The injector is a 15 Mev Microtron Electrons are accelerated from 15 Mev to 800 Mev by ramping the field of the ring. The expected stored current will be around 100 ma

  4. LASL Compact Torus Program

    Linford, R.K.; Armstrong, W.T.; Bartsch, R.R.

    1981-01-01

    The Compact Torus (CT) concept includes any axisymmetric toroidal plasma configuration, which does not require the linking of any material through the hole in the torus. Thus, the magnet coils, vacuum vessel, etc., have a simple cylindrical or spherical geometry instead of the toroidal geometry required for Tokamaks and RFP's. This simplified geometry results in substantial engineering advantages in CT reactor embodiments while retaining the good confinement properties afforded by an axisymmetric toroidal plasma-field geometry. CT's can be classified into three major types by using the ion gyro radius rho/sub i/ and the magnitude of the maximum toroidal field B/sub tm/

  5. Compact Q-balls

    Bazeia, D., E-mail: bazeia@fisica.ufpb.br [Departamento de Física, Universidade Federal da Paraíba, 58051-970 João Pessoa, PB (Brazil); Losano, L.; Marques, M.A. [Departamento de Física, Universidade Federal da Paraíba, 58051-970 João Pessoa, PB (Brazil); Menezes, R. [Departamento de Ciências Exatas, Universidade Federal da Paraíba, 58297-000 Rio Tinto, PB (Brazil); Departamento de Física, Universidade Federal de Campina Grande, 58109-970 Campina Grande, PB (Brazil); Rocha, R. da [Centro de Matemática, Computação e Cognição, Universidade Federal do ABC, 09210-580 Santo André (Brazil)

    2016-07-10

    In this work we deal with non-topological solutions of the Q-ball type in two space–time dimensions, in models described by a single complex scalar field that engenders global symmetry. The main novelty is the presence of stable Q-balls solutions that live in a compact interval of the real line and appear from a family of models controlled by two distinct parameters. We find analytical solutions and study their charge and energy, and show how to control the parameters to make the Q-balls classically and quantum mechanically stable.

  6. Scalable Nonlinear Compact Schemes

    Ghosh, Debojyoti [Argonne National Lab. (ANL), Argonne, IL (United States); Constantinescu, Emil M. [Univ. of Chicago, IL (United States); Brown, Jed [Univ. of Colorado, Boulder, CO (United States)

    2014-04-01

    In this work, we focus on compact schemes resulting in tridiagonal systems of equations, specifically the fifth-order CRWENO scheme. We propose a scalable implementation of the nonlinear compact schemes by implementing a parallel tridiagonal solver based on the partitioning/substructuring approach. We use an iterative solver for the reduced system of equations; however, we solve this system to machine zero accuracy to ensure that no parallelization errors are introduced. It is possible to achieve machine-zero convergence with few iterations because of the diagonal dominance of the system. The number of iterations is specified a priori instead of a norm-based exit criterion, and collective communications are avoided. The overall algorithm thus involves only point-to-point communication between neighboring processors. Our implementation of the tridiagonal solver differs from and avoids the drawbacks of past efforts in the following ways: it introduces no parallelization-related approximations (multiprocessor solutions are exactly identical to uniprocessor ones), it involves minimal communication, the mathematical complexity is similar to that of the Thomas algorithm on a single processor, and it does not require any communication and computation scheduling.

  7. Compact magnetic fusion systems

    Linford, R.K.

    1983-12-01

    If the core (first wall, blanket, shield, and magnet coils) of fusion reactor systems could be made smaller in mass and volume for a given net electric power output than is usually predicted for the mainline tokamak/sup 1/ and mirror concepts, the cost of the technological development of the core and the construction of power plants might be significantly reduced. Although progress in plasma physics and engineering approaches should continue to yield improvements in reactor designs, certain physics features of the mainline concepts may prevent major reductions in the size of the core without straining the limits of technology. However, more than a factor of ten reduction in volume and mass of the core, at constant output power, may be possible for a class of toroidal confinement concepts in which the confining magnetic fields are supported more by currents flowing in the plasma than those in the external coils. In spite of this dramatic increase in power density (ratio of total thermal output power to the volume of the core), the design of compact systems need not rely on any materials requirements that are qualitatively more difficult than those proposed for the lower-power-density mainline fusion concepts. In some respects compact systems require less of an extension of existing technology, e.g. magnetics.

  8. Compact magnetic fusion systems

    Linford, R.K.

    1983-01-01

    If the core (first wall, blanket, shield, and magnet coils) of fusion reactor systems could be made smaller in mass and volume for a given net electric power output than is usually predicted for the mainline tokamak 1 and mirror concepts, the cost of the technological development of the core and the construction of power plants might be significantly reduced. Although progress in plasma physics and engineering approaches should continue to yield improvements in reactor designs, certain physics features of the mainline concepts may prevent major reductions in the size of the core without straining the limits of technology. However, more than a factor of ten reduction in volume and mass of the core, at constant output power, may be possible for a class of toroidal confinement concepts in which the confining magnetic fields are supported more by currents flowing in the plasma than those in the external coils. In spite of this dramatic increase in power density (ratio of total thermal output power to the volume of the core), the design of compact systems need not rely on any materials requirements that are qualitatively more difficult than those proposed for the lower-power-density mainline fusion concepts. In some respects compact systems require less of an extension of existing technology, e.g. magnetics

  9. Diffusion in compacted betonite

    Muurinen, A.; Rantanen, J.

    1985-01-01

    The objective of this report is to collect the literature bearing on the diffusion in compacted betonite, which has been suggested as possible buffer material for the disposal of spent fuel. Diffusion in a porous, water-saturated material is usually described as diffusion in the pore-water where sorption on the solid matter can delay the migration in the instationary state. There are also models which take into consideration that the sorbed molecules can also move while being sorbed. Diffusion experiments in compacted bentonite have been reported by many authors. Gases, anions, cations and actinides have been used as diffusing molecules. The report collects the results and the information on the measurement methods. On the basis of the results can be concluded that different particles possibly follow different diffusion mechanisms. The parameters which affect the diffusion seem to be for example the size, the electric charge and the sorption properties of the diffusing molecule. The report also suggest the parameters to be used in the diffusion calculation of the safety analyses of spent fuel disposal. (author)

  10. Compact Infrasonic Windscreen

    Zuckerwar, Allan J.; Shams, Qamar A.; Sealey, Bradley S.; Comeaux, Toby

    2005-01-01

    A compact windscreen has been conceived for a microphone of a type used outdoors to detect atmospheric infrasound from a variety of natural and manmade sources. Wind at the microphone site contaminates received infrasonic signals (defined here as sounds having frequencies <20 Hz), because a microphone cannot distinguish between infrasonic pressures (which propagate at the speed of sound) and convective pressure fluctuations generated by wind turbulence. Hence, success in measurement of outdoor infrasound depends on effective screening of the microphone from the wind. The present compact windscreen is based on a principle: that infrasound at sufficiently large wavelength can penetrate any barrier of practical thickness. Thus, a windscreen having solid, non-porous walls can block convected pressure fluctuations from the wind while transmitting infrasonic acoustic waves. The transmission coefficient depends strongly upon the ratio between the acoustic impedance of the windscreen and that of air. Several materials have been found to have impedance ratios that render them suitable for use in constructing walls that have practical thicknesses and are capable of high transmission of infrasound. These materials (with their impedance ratios in parentheses) are polyurethane foam (222), space shuttle tile material (332), balsa (323), cedar (3,151), and pine (4,713).

  11. Compact electrostatic comb actuator

    Rodgers, M. Steven; Burg, Michael S.; Jensen, Brian D.; Miller, Samuel L.; Barnes, Stephen M.

    2000-01-01

    A compact electrostatic comb actuator is disclosed for microelectromechanical (MEM) applications. The actuator is based upon a plurality of meshed electrostatic combs, some of which are stationary and others of which are moveable. One or more restoring springs are fabricated within an outline of the electrostatic combs (i.e. superposed with the moveable electrostatic combs) to considerably reduce the space required for the actuator. Additionally, a truss structure is provided to support the moveable electrostatic combs and prevent bending or distortion of these combs due to unbalanced electrostatic forces or external loading. The truss structure formed about the moveable electrostatic combs allows the spacing between the interdigitated fingers of the combs to be reduced to about one micron or less, thereby substantially increasing the number of active fingers which can be provided in a given area. Finally, electrostatic shields can be used in the actuator to substantially reduce unwanted electrostatic fields to further improve performance of the device. As a result, the compact electrostatic comb actuator of the present invention occupies only a fraction of the space required for conventional electrostatic comb actuators, while providing a substantial increase in the available drive force (up to one-hundred times).

  12. Development task of compact reactor

    Kurushima, Morihiro

    1982-01-01

    In the Ministry of International Trade and Industry, studies proceed on the usage of compact medium and small LWRs. As such, the reactors from 100 to 200 MW may meet varieties of demands in scale and kind in view of the saving of petroleum and the economy of nuclear power. In this case, the technology of light water reactors with already established safety will be suitable for the development of compact reactors. The concept of ''nuclear power community'' using the compact reactors in local society and industrial zones was investigated. The following matters are described: need for the introduction of compact reactors, the survey on the compact reactor systems, and the present status and future problems for compact reactor usage. (J.P.N.)

  13. The United Nations Global Compact

    Rasche, Andreas; Waddock, Sandra; McIntosh, Malcolm

    2013-01-01

    This article reviews the interdisciplinary literature on the UN Global Compact. The review identifies three research perspectives, which scholars have used to study the UN Global Compact so far: a historical perspective discussing the Global Compact in the context of UN-business relations...... key empirical as well as conceptual scholarly contributions. The remainder of this article contains focused summaries of the articles selected for this Special Issue. All articles are introduced and evaluated against the background of the three research perspectives....

  14. DEMOS PLUS. Robot for decontaminating soils and cavity walls of the reactor and fuel pools NPP primarily during periods of recharging fuel

    Lacalle Bayo, J.; Vaquer Perez, J. I.; Rosello Garcia, J. I.

    2014-01-01

    In this work the robot Plus Demos, equipment that has been developed by GD Energy Services from the redesign and development of robot Demos show, which took place on last year. This evolution has given the team greater capabilities, highlighting the decontamination of vertical surfaces. The main objective pursued is to minimize operational doses to workers operating in cavity as well as the risk of surface contamination during them. (Author)

  15. Compact particle accelerator

    Elizondo-Decanini, Juan M.

    2017-08-29

    A compact particle accelerator having an input portion configured to receive power to produce particles for acceleration, where the input portion includes a switch, is provided. In a general embodiment, a vacuum tube receives particles produced from the input portion at a first end, and a plurality of wafer stacks are positioned serially along the vacuum tube. Each of the plurality of wafer stacks include a dielectric and metal-oxide pair, wherein each of the plurality of wafer stacks further accelerate the particles in the vacuum tube. A beam shaper coupled to a second end of the vacuum tube shapes the particles accelerated by the plurality of wafer stacks into a beam and an output portion outputs the beam.

  16. Compact vacuum insulation embodiments

    Benson, D.K.; Potter, T.F.

    1992-04-28

    An ultra-thin compact vacuum insulation panel is comprised of two hard, but bendable metal wall sheets closely spaced apart from each other and welded around the edges to enclose a vacuum chamber. Glass or ceramic spacers hold the wall sheets apart. The spacers can be discrete spherical beads or monolithic sheets of glass or ceramic webs with nodules protruding therefrom to form essentially point' or line' contacts with the metal wall sheets. In the case of monolithic spacers that form line' contacts, two such spacers with the line contacts running perpendicular to each other form effectively point' contacts at the intersections. Corrugations accommodate bending and expansion, tubular insulated pipes and conduits, and preferred applications are also included. 26 figs.

  17. Compact vacuum insulation

    Benson, D.K.; Potter, T.F.

    1993-01-05

    An ultra-thin compact vacuum insulation panel is comprised of two hard, but bendable metal wall sheets closely spaced apart from each other and welded around the edges to enclose a vacuum chamber. Glass or ceramic spacers hold the wall sheets apart. The spacers can be discrete spherical beads or monolithic sheets of glass or ceramic webs with nodules protruding therefrom to form essentially point'' or line'' contacts with the metal wall sheets. In the case of monolithic spacers that form line'' contacts, two such spacers with the line contacts running perpendicular to each other form effectively point'' contacts at the intersections. Corrugations accommodate bending and expansion, tubular insulated pipes and conduits, and preferred applications are also included.

  18. The Compact Ignition Tokamak

    Schmidt, J.

    1987-01-01

    The author discusses his lab's plan for completing the Compact Ignition Tokamak (CIT) conceptual design during calendar year 1987. Around July 1 they froze the subsystem envelopes on the device to continue with the conceptual design. They did this by formalizing a general requirements document. They have been developing the management plan and submitted a version to the DOE July 10. He describes a group of management activities. They released the vacuum vessel Request For Proposals (RFP) on August 5. An RFP to do a major part of the system engineering on the device is being developed. They intend to assemble the device outside of the test cell, then move it into the the test cell, install it there, and bring to the test cell many of the auxiliary facilities from TFTR, for example, power supplies

  19. Compact cryocooler heat exchangers

    Luna, J.; Frederking, T.H.K.

    1991-01-01

    Compact heat exchangers are subject to different constraints as a room temperature gas is cooled down by a cold stream returning from a JT valve (or a similar cryoprocess component). In particular, the optimization of exchangers for liquid helium systems has to cover a wide range in temperature and density of the fluid. In the present work we address the following thermodynamic questions: 1. The optimization of intermediate temperatures which optimize stage operation (a stage is assumed to have a constant cross section); 2. The optimum temperature difference available for best overall economic performance values. The results are viewed in the context of porous media concepts applied to rather low speeds of fluid flow in narrow passages. In this paper examples of fluid/solid constraints imposed in this non-classical low temperature area are presented

  20. Compact semiconductor lasers

    Yu, Siyuan; Lourtioz, Jean-Michel

    2014-01-01

    This book brings together in a single volume a unique contribution by the top experts around the world in the field of compact semiconductor lasers to provide a comprehensive description and analysis of the current status as well as future directions in the field of micro- and nano-scale semiconductor lasers. It is organized according to the various forms of micro- or nano-laser cavity configurations with each chapter discussing key technical issues, including semiconductor carrier recombination processes and optical gain dynamics, photonic confinement behavior and output coupling mechanisms, carrier transport considerations relevant to the injection process, and emission mode control. Required reading for those working in and researching the area of semiconductors lasers and micro-electronics.

  1. Effect of activation cross-section uncertainties on the radiological assessment of the MFE/DEMO first wall

    Cabellos, O.; Reyes, S.; Sanz, J.; Rodriguez, A.; Youssef, M.; Sawan, M.

    2006-01-01

    A Monte Carlo procedure has been applied in this work in order to address the impact of activation cross-sections (XS) uncertainties on contact dose rate and decay heat calculations for the outboard first wall (FW) of a magnetic fusion energy (MFE) demonstration (DEMO) reactor. The XSs inducing the major uncertainty in the prediction of activation related quantities have been identified. Results have shown that for times corresponding to maintenance activities the uncertainties effect is insignificant since the dominant XSs involved in these calculations are based on accurate experimental data evaluations. However, for times corresponding to waste management/recycling activities, the errors induced by the XSs uncertainties, which in this case are evaluated using systematic models, must be considered. It has been found that two particular isotopes, 6 Co and 94 Nb, are key contributors to the global DEMO FW activation uncertainty results. In these cases, the benefit from further improvements in the accuracy of the critical reaction XSs is discussed

  2. The Logic-Based Supervisor Control for Sun-Tracking System of 1 MW HCPV Demo Plant: Study Case

    Hong-Yih Yeh

    2012-02-01

    Full Text Available This paper presents a logic-based supervisor controller designed for trackers for a 1MW HCPV demo plant in Taiwan. A sun position sensor on the tracker is used to detect the sun position, as the sensor is sensitive to the intensity of sun light. The signal output of the sensor is partially affected by the cloud, which has a hard control position with the traditional PID control. Therefore we have used logic-based supervisor (LBS control which permits switching the PID control to sun trajectory under sunny or cloudy conditions. To verify the stability of the proposed control, an experiment was performed and the results show that the proposed control can efficiently achieve stabilization of the trackers of the 1MW HCPV demo plant.

  3. Characterization of high temperature superconductor cables for magnet toroidal field coils of the DEMO fusion power plant

    Bayer, Christoph M

    2017-01-01

    Nuclear fusion is a key technology to satisfy the basic demand for electric energy sustainably. The official EUROfusion schedule foresees a first industrial DEMOnstration Fusion Power Plant for 2050. In this work several high temperature superconductor sub-size cables are investigated for their applicability in large scale DEMO toroidal field coils. Main focus lies on the electromechanical stability under the influence of high Lorentz forces at peak magnetic fields of up to 12 T.

  4. Analysis of the WCLL European demo blanket concept in terms of activation and decay heat after exposure to neutron irradiation

    Stankunas Gediminas; Tidikas Andrius

    2017-01-01

    This comparative paper describes the activation and decay heat calculations for water-cooled lithium-lead performed part of the EURO fusion WPSAE programme and specifications in comparison to other European DEMO blanket concepts on the basis of using a three-dimensional neutronics calculation model. Results are provided for a range of decay times of interest for maintenance activities, safety and waste management assessments. The study revealed that water-c...

  5. Progress of R&D on water cooled ceramic breeder for ITER test blanket system and DEMO

    Kawamura, Yoshinori, E-mail: kawamura.yoshinori@jaea.go.jp [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Tanigawa, Hisashi; Hirose, Takanori; Enoeda, Mikio [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Sato, Satoshi [Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai, Ibaraki 319-1195 (Japan); Ochiai, Kentaro [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan); Konno, Chikara; Edao, Yuki; Hayashi, Takumi [Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai, Ibaraki 319-1195 (Japan); Hoshino, Tsuyoshi; Nakamichi, Masaru; Tanigawa, Hiroyasu [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan); Nishi, Hiroshi; Suzuki, Satoshi; Ezato, Koichiro; Seki, Yohji [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Yamanishi, Toshihiko [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan)

    2016-11-01

    Highlights: • Thermo-hydraulic calculation in the TBM at the water ingress event has been done. • Shielding calculations for the ITER equatorial port #18 were conducted by using C-lite model. • Prototypic pebbles of Be{sub 17}Ti{sub 2} and Be{sub 12}V had a good oxidation property similar to Be{sub 12}Ti pebble. • Li rich Li{sub 2}TiO{sub 3} pebbles were successfully fabricated using the emulsion method by controlling sintering atmosphere. • New tritium production/recovery experiments at FNS have been started by using ionization chamber as on-line gas monitor. - Abstract: 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 development of DEMO blanket, R&D has been performed on the module fabrication technology, breeder and multiplier pebble fabrication technology, tritium production rate evaluation, as well as structural and safety design activities. The fabrication of full-scale first wall, side walls, breeder pebble bed box and back wall was completed, and assembly of TBM with box structure was successfully achieved. Development of advanced breeder and multiplier pebbles for higher chemical stability was continued for future DEMO blanket application. From the view point of TBM test result evaluation and DEMO blanket performance design, the development of the blanket tritium transport simulation technology, investigation of the TBM neutron measurement technology and the evaluation of the tritium production and recovery test using D-T neutron in the fusion neutron source (FNS) facility has been performed. This paper provides an overview of the recent achievements of the development of the WCCB Blanket in Japan.

  6. Characterization of high temperature superconductor cables for magnet toroidal field coils of the DEMO fusion power plant

    Bayer, Christoph M.

    2017-05-01

    Nuclear fusion is a key technology to satisfy the basic demand for electric energy sustainably. The official EUROfusion schedule foresees a first industrial DEMOnstration Fusion Power Plant for 2050. In this work several high temperature superconductor sub-size cables are investigated for their applicability in large scale DEMO toroidal field coils. Main focus lies on the electromechanical stability under the influence of high Lorentz forces at peak magnetic fields of up to 12 T.

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

    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.

  8. Compact magnetic fusin reactor concepts

    Chung, K.M.

    1984-01-01

    Compact, high-power-density approaches to fusion power represent alternatives to main-line fusion concepts, Tokamaks and mirrors. If technological issues are resolved, theses approaches would yield small, low-cost fusion power plants. This survey reviews the principal physics and technology employed by leading compact magnetic fusion plants. (Author)

  9. Solid targetry for compact cyclotrons

    Comor, J.

    2004-01-01

    In this presentation authors present experimental results of solid targetry for compact cyclotrons. It is concluded: Solid targetry is not restricted to large accelerator centers anymore; Small and medium scale radioisotope production is feasible with compact cyclotrons; The availability of versatile solid target systems is expected to boost the radiochemistry of 'exotic' positron emitters

  10. Roller-compacted concrete pavements.

    2010-09-01

    Roller-compacted concrete (RCC) gets its name from the heavy vibratory steel drum and rubber-tired rollers used to help compact it into its final form. RCC has similar strength properties and consists of the same basic ingredients as conventional con...

  11. Machine for compacting solid residues

    Herzog, J.

    1981-11-01

    Machine for compacting solid residues, particularly bulky radioactive residues, constituted of a horizontally actuated punch and a fixed compression anvil, in which the residues are first compacted horizontally and then vertically. Its salient characteristic is that the punch and the compression anvil have embossments on the compression side and interpenetrating plates in the compression position [fr

  12. Concept design of the DEMO divertor cassette-to-vacuum vessel locking system adopting a systems engineering approach

    Di Gironimo, G.; Carfora, D.; Esposito, G.; Lanzotti, A.; Marzullo, D.; Siuko, M.

    2015-01-01

    Highlights: • An iterative and incremental design process for cassette-to-VV locking system of DEMO divertor is presented. • Three different concepts have been developed with a systematic design approach. • The final concept has been selected with Fuzzy-Analytic Hierarchy Process in virtual reality. - Abstract: This paper deals with pre-concept studies of DEMO divertor cassette-to-vacuum vessel locking system under the work program WP13-DAS-07-T06: Divertor Remote Maintenance System pre-concept study. An iterative design process, consistent with Systems Engineering guidelines and named Iterative and Participative Axiomatic Design Process (IPADeP), is used in this paper to propose new innovative solutions for divertor locking system, which can overcome the difficulties in applying the ITER principles to DEMO. The solutions conceived have been analysed from the structural point of view using the software Ansys and, eventually, evaluated using the methodology known as Fuzzy-Analytic Hierarchy Process. Due to the lack and the uncertainty of the requirements in this early conceptual design stage, the aim is to cover a first iteration of an iterative and incremental process to propose an innovative design concept to be developed in more details as the information will be completed

  13. Benchmarking Reactor Systems Studies by Comparison of EU and Japanese System Code Results for Different DEMO Concepts

    Kemp, R.; Ward, D.J., E-mail: richard.kemp@ccfe.ac.uk [EURATOM/CCFE Association, Culham Centre for Fusion Energy, Abingdon (United Kingdom); Nakamura, M.; Tobita, K. [Japan Atomic Energy Agency, Rokkasho (Japan); Federici, G. [EFDA Garching, Max Plank Institut fur Plasmaphysik, Garching (Germany)

    2012-09-15

    Full text: Recent systems studies work within the Broader Approach framework has focussed on benchmarking the EU systems code PROCESS against the Japanese code TPC for conceptual DEMO designs. This paper describes benchmarking work for a conservative, pulsed DEMO and an advanced, steady-state, high-bootstrap fraction DEMO. The resulting former machine is an R{sub 0} = 10 m, a = 2.5 m, {beta}{sub N} < 2.0 device with no enhancement in energy confinement over IPB98. The latter machine is smaller (R{sub 0} = 8 m, a = 2.7 m), with {beta}{sub N} = 3.0, enhanced confinement, and high bootstrap fraction f{sub BS} = 0.8. These options were chosen to test the codes across a wide range of parameter space. While generally in good agreement, some of the code outputs differ. In particular, differences have been identified in the impurity radiation models and flux swing calculations. The global effects of these differences are described and approaches to identifying the best models, including future experiments, are discussed. Results of varying some of the assumptions underlying the modelling are also presented, demonstrating the sensitivity of the solutions to technological limitations and providing guidance for where further research could be focussed. (author)

  14. Overview of the European Union fusion nuclear technologies development and essential elements on the way to DEMO

    Andreani, R.; Diegele, E.; Gulden, W.; Laesser, R.; Maisonnier, D.; Murdoch, D.; Pick, M.; Poitevin, Y.

    2006-01-01

    EU is strongly preparing ITER construction involving the system of EU Associations, universities and industry. The European programme has been steered to be in line with the present conception of a future power reactor. Thirty percent of the fusion research budget has been spent on long-term related activities managed by EFDA. These include Power Plant Conceptual Studies (PPCS), the recently undertaken DEMO Conceptual Studies, design and R and D for breeder blankets, low activation materials and IFMIF. Developments on fuel cycle, neutronics, safety and socio-economics complement those specifically performed for ITER. Two EU helium-cooled DEMO blankets will be tested in ITER, using liquid lithium-lead and solid ceramics as breeders. The blanket structures will use EUROFER. Irradiations to 70-80 dpa will qualify EUROFER for DEMO. Advanced materials, in particular SiC f /SiC, under development, could provide more thermodynamically efficient blankets. Even with a fully successful ITER, a number of issues will remain open in technology. The application of high temperature superconductors, essential progress in materials, blanket design and remote handling, are required to produce environmentally safe and economically competitive fusion. A fully integrated world wide international programme is the best way to efficiently progress in these fields

  15. Concept design of the DEMO divertor cassette-to-vacuum vessel locking system adopting a systems engineering approach

    Di Gironimo, G., E-mail: giuseppe.digironimo@unina.it [Università degli Studi di Napoli “Federico II”, Dipartimento di Ingegneria Industriale, Piazzale Tecchio 80, 80135 Napoli (Italy); Carfora, D. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); VTT Technical Research Centre of Finland, Tekniikankatu 1, PO Box 1300, FI-33101 Tampere (Finland); Università degli Studi di Napoli “Federico II”, Dipartimento di Ingegneria Industriale, Piazzale Tecchio 80, 80135 Napoli (Italy); Esposito, G.; Lanzotti, A.; Marzullo, D. [Università degli Studi di Napoli “Federico II”, Dipartimento di Ingegneria Industriale, Piazzale Tecchio 80, 80135 Napoli (Italy); Siuko, M. [VTT Technical Research Centre of Finland, Tekniikankatu 1, PO Box 1300, FI-33101 Tampere (Finland)

    2015-05-15

    Highlights: • An iterative and incremental design process for cassette-to-VV locking system of DEMO divertor is presented. • Three different concepts have been developed with a systematic design approach. • The final concept has been selected with Fuzzy-Analytic Hierarchy Process in virtual reality. - Abstract: This paper deals with pre-concept studies of DEMO divertor cassette-to-vacuum vessel locking system under the work program WP13-DAS-07-T06: Divertor Remote Maintenance System pre-concept study. An iterative design process, consistent with Systems Engineering guidelines and named Iterative and Participative Axiomatic Design Process (IPADeP), is used in this paper to propose new innovative solutions for divertor locking system, which can overcome the difficulties in applying the ITER principles to DEMO. The solutions conceived have been analysed from the structural point of view using the software Ansys and, eventually, evaluated using the methodology known as Fuzzy-Analytic Hierarchy Process. Due to the lack and the uncertainty of the requirements in this early conceptual design stage, the aim is to cover a first iteration of an iterative and incremental process to propose an innovative design concept to be developed in more details as the information will be completed.

  16. AC loss, interstrand resistance and mechanical properties of prototype EU DEMO TF conductors up to 30 000 load cycles

    Yagotintsev, K.; Nijhuis, A.

    2018-07-01

    Two prototype Nb3Sn cable-in-conduit conductors conductors were designed and manufactured for the toroidal field (TF) magnet system of the envisaged European DEMO fusion reactor. The AC loss, contact resistance and mechanical properties of two sample conductors were tested in the Twente Cryogenic Cable Press under cyclic load up to 30 000 cycles. Though both conductors were designed to operate at 82 kA in a background magnetic field of 13.6 T, they reflect different approaches with respect to the magnet winding pack assembly. The first approach is based on react and wind technology while the second is the more common wind and react technology. Each conductor was tested first for AC loss in virgin condition without handling. The impact of Lorentz load during magnet operation was simulated using the cable press. In the press each conductor specimen was subjected to transverse cyclic load up to 30 000 cycles in liquid helium bath at 4.2 K. Here a summary of results for AC loss, contact resistance, conductor deformation, mechanical heat production and conductor stiffness evolution during cycling of the load is presented. Both conductors showed similar mechanical behaviour but quite different AC loss. In comparison with previously tested ITER TF conductors, both DEMO TF conductors possess very low contact resistance resulting in high coupling loss. At the same time, load cycling has limited impact on properties of DEMO TF conductors in comparison with ITER TF conductors.

  17. Exploration of one-dimensional plasma current density profile for K-DEMO steady-state operation

    Kang, J.S. [Seoul National University, Seoul 151-742 (Korea, Republic of); Jung, L. [National Fusion Research Institute, Daejeon (Korea, Republic of); Byun, C.-S.; Na, D.H.; Na, Y.-S. [Seoul National University, Seoul 151-742 (Korea, Republic of); Hwang, Y.S., E-mail: yhwang@snu.ac.kr [Seoul National University, Seoul 151-742 (Korea, Republic of)

    2016-11-01

    Highlights: • One-dimensional current density and its optimization for the K-DEMO are explored. • Plasma current density profile is calculated with an integrated simulation code. • The impact of self and external heating profiles is considered self-consistently. • Current density is identified as a reference profile by minimizing heating power. - Abstract: Concept study for Korean demonstration fusion reactor (K-DEMO) is in progress, and basic design parameters are proposed by targeting high magnetic field operation with ITER-sized machine. High magnetic field operation is a favorable approach to enlarge relative plasma performance without increasing normalized beta or plasma current. Exploration of one-dimensional current density profile and its optimization process for the K-DEMO steady-state operation are reported in this paper. Numerical analysis is conducted with an integrated plasma simulation code package incorporating a transport code with equilibrium and current drive modules. Operation regimes are addressed with zero-dimensional system analysis. One-dimensional plasma current density profile is calculated based on equilibrium, bootstrap current analysis, and thermal transport analysis. The impact of self and external heating profiles on those parameters is considered self-consistently, where thermal power balance and 100% non-inductive current drive are the main constraints during the whole exploration procedure. Current and pressure profiles are identified as a reference steady-state profile by minimizing the external heating power with desired fusion power.

  18. Options for a high heat flux enabled helium cooled first wall for DEMO

    Arbeiter, Frederik, E-mail: f.arbe@kit.edu; Chen, Yuming; Ghidersa, Bradut-Eugen; Klein, Christine; Neuberger, Heiko; Ruck, Sebastian; Schlindwein, Georg; Schwab, Florian; Weth, Axel von der

    2017-06-15

    Highlights: • Design challenges for helium cooled first wall reviewed and otimization approaches explored. • Application of enhanced heat transfer surfaces to the First Wall cooling channels. • Demonstrated a design point for 1 MW/m{sup 2} with temperatures <550 °C and acceptable stresses. • Feasibility of several manufacturing processes for ribbed surfaces is shown. - Abstract: Helium is considered as coolant in the plasma facing first wall of several blanket concepts for DEMO fusion reactors, due to the favorable properties of flexible temperature range, chemical inertness, no activation, comparatively low effort to remove tritium from the gas and no chemical corrosion. Existing blanket designs have shown the ability to use helium cooled first walls with heat flux densities of 0.5 MW/m{sup 2}. Average steady state heat loads coming from the plasma for current EU DEMO concepts are expected in the range of 0.3 MW/m{sup 2}. The definition of peak values is still ongoing and depends on the chosen first wall shape, magnetic configuration and assumptions on the fraction of radiated power and power fall off lengths in the scrape off layer of the plasma. Peak steady state values could reach and excess 1 MW/m{sup 2}. Higher short-term transient loads are expected. Design optimization approaches including heat transfer enhancement, local heat transfer tuning and shape optimization of the channel cross section are discussed. Design points to enable a helium cooled first wall capable to sustain heat flux densities of 1 MW/m{sup 2} at an average shell temperature lower than 500 °C are developed based on experimentally validated heat transfer coefficients of structured channel surfaces. The required pumping power is in the range of 3–5% of the collected thermal power. The FEM stress analyses show code-acceptable stress intensities. Several manufacturing methods enabling the application of the suggested heat transfer enhanced first wall channels are explored. An

  19. Safety studies of plasma-wall events with AINA code for Japanese DEMO

    Rivas, J.C., E-mail: jose.carlos.rivas@upc.edu [International Fusion Energy Research Centre (IFERC) (Japan); Fusion Energy Engineering Laboratory (FEEL), Technical University of Catalonia-BarcelonaTech (Spain); Nakamura, M.; Someya, Y.; Hoshino, K.; Asakura, N. [Japan Atomic Energy Agency (JAEA) (Japan); Takase, H. [International Fusion Energy Research Centre (IFERC) (Japan); Miyoshi, Y.; Utoh, H.; Tobita, K. [Japan Atomic Energy Agency (JAEA) (Japan); Dies, J.; Blas, A. de; Riego, A.; Fabbri, M. [Fusion Energy Engineering Laboratory (FEEL), Technical University of Catalonia-BarcelonaTech (Spain)

    2016-11-01

    Highlights: • Work done in AINA code during 2014 and 2015 at IFERC to develop a version for safety studies of a Japanese DEMO design. • A thermal model for a WCPB breeding blanket has been developed based in parametric input data from neutronics calculations. • A breakthrough for the safety studies of plasma-divertor transients: An integrated SOL-pedestal model + using melting time as objective variable + using optimization algorithm. • The results for the case of divertor show that both loss of plasma control (LOPC) transients and ex-vessel LOCA transient can induce severe melting. The difference is that while in the first case melting happens at PFC surface, in the second case it happens at copper heat sink. • Conclusions suggest that, because the minimum melting times are same order of magnitude than the energy confinement time, recovery time for plasma control system should be lower order. - Abstract: In this contribution, the work done in AINA code during 2014 and 2015 at IFERC is presented. The main motivation of this work was to adapt the code and to perform safety studies for a Japanese DEMO design. Related to AINA code, the work has supposed major changes in plasma models. Significant is the addition of an integrated SOL-pedestal model that allows the estimation of heat loads at divertor. Also, a thermal model for a WCPB (water cooled pebble bed) breeding blanket has been developed based in parametric input data from neutronics calculations. Related to safety studies, a major breakthrough in the study of LOPC (loss of plasma control) transients has been the use of an optimization method to determine the most severe transients in terms of the shortest melting times. The results of the safety study show that LOPC transients are not likely to be severe for breeding blanket, but for the case of divertor can induce severe melting. For ex-vessel LOCA (loss of coolant accident) analysis, it is severe for both blanket and divertor, but in the first case

  20. Safety studies of plasma-wall events with AINA code for Japanese DEMO

    Rivas, J.C.; Nakamura, M.; Someya, Y.; Hoshino, K.; Asakura, N.; Takase, H.; Miyoshi, Y.; Utoh, H.; Tobita, K.; Dies, J.; Blas, A. de; Riego, A.; Fabbri, M.

    2016-01-01

    Highlights: • Work done in AINA code during 2014 and 2015 at IFERC to develop a version for safety studies of a Japanese DEMO design. • A thermal model for a WCPB breeding blanket has been developed based in parametric input data from neutronics calculations. • A breakthrough for the safety studies of plasma-divertor transients: An integrated SOL-pedestal model + using melting time as objective variable + using optimization algorithm. • The results for the case of divertor show that both loss of plasma control (LOPC) transients and ex-vessel LOCA transient can induce severe melting. The difference is that while in the first case melting happens at PFC surface, in the second case it happens at copper heat sink. • Conclusions suggest that, because the minimum melting times are same order of magnitude than the energy confinement time, recovery time for plasma control system should be lower order. - Abstract: In this contribution, the work done in AINA code during 2014 and 2015 at IFERC is presented. The main motivation of this work was to adapt the code and to perform safety studies for a Japanese DEMO design. Related to AINA code, the work has supposed major changes in plasma models. Significant is the addition of an integrated SOL-pedestal model that allows the estimation of heat loads at divertor. Also, a thermal model for a WCPB (water cooled pebble bed) breeding blanket has been developed based in parametric input data from neutronics calculations. Related to safety studies, a major breakthrough in the study of LOPC (loss of plasma control) transients has been the use of an optimization method to determine the most severe transients in terms of the shortest melting times. The results of the safety study show that LOPC transients are not likely to be severe for breeding blanket, but for the case of divertor can induce severe melting. For ex-vessel LOCA (loss of coolant accident) analysis, it is severe for both blanket and divertor, but in the first case

  1. A new fully automatic PIM tool to replicate two component tungsten DEMO divertor parts

    Antusch, Steffen; Commin, Lorelei; Heneka, Jochen; Piotter, Volker; Plewa, Klaus; Walter, Heinz

    2013-01-01

    Highlights: • Development of a fully automatic 2C-PIM tool. • Replicate fusion relevant components in one step without additional brazing. • No cracks or gaps in the seam of the joining zone visible. • For both material combinations a solid bond of the material interface was achieved. • PIM is a powerful process for mass production as well as for joining even complex shaped parts. -- Abstract: At Karlsruhe Institute of Technology (KIT), divertor design concepts for future nuclear fusion power plants beyond ITER are intensively investigated. One promising KIT divertor design concept for the future DEMO power reactor is based on modular He-cooled finger units. The manufacturing of such parts by mechanical machining such as milling and turning, however, is extremely cost and time intensive because tungsten is very hard and brittle. Powder Injection Molding (PIM) has been adapted to tungsten processing at KIT since a couple of years. This production method is deemed promising in view of large-scale production of tungsten parts with high near-net-shape precision, hence, offering an advantage of cost-saving process compared to conventional machining. The properties of the effectively and successfully manufactured divertor part tile consisting only of pure tungsten are a microstructure without cracks and a high density (>98% T.D.). Based on the achieved results a new fully automatic multicomponent PIM tool was developed and allows the replication and joining without brazing of fusion relevant components of different materials in one step and the creation of composite materials. This contribution describes the process route to design and engineer a new fully automatic 2C-PIM tool, including the filling simulation and the implementing of the tool. The complete technological fabrication process of tungsten 2C-PIM, including material and feedstock (powder and binder) development, injection molding, and heat-treatment of real DEMO divertor parts is outlined

  2. Evaluation of energy and particle impact on the plasma facing components in DEMO

    Igitkhanov, Yuri; Bazylev, Boris

    2012-01-01

    Highlights: ► We analyze the first wall blanket W/EUROFER configuration for DEMO under steady-state normal operation and off-normal conditions, such as vertical displacements events (VDE) and runaway electrons (RE). The main issue is to find the optimal thickness of the W armor which will prevent tungsten surface from evaporation and melting and, on the other hand, will keep EUROFER below the critical thermal stresses. ► The minimum thickness of the tungsten amour about 3 mm for W/EUROFER sandwich structure will keep the maximum EUROFER temperature below the critical limit for EUROFER steel under steady-state operation and ITER like cooling conditions. ► The W surface temperature and the max. EUROFER temperature are increasing with incoming heat flux. For reference conditions (Dw ∼3 mm, DEUROFER ∼4 mm) the maximum tolerable heat flux which does not causes in thermal stresses in structural material is about ∼13.5 MW/m 2 . ► The RE deposit their energy deeper into W armour and for energies ≥50 MJ/m 2 and deposition times ≤0.1 s, the minimum armor thickness required to prevent EUROFER from thermal distraction is ≥1.4 cm. ► However, at this thickness the W surface melts. For higher RE energy deposition rates (≥100 MJ/m 2 in 0.1 s), the required armor thickness to prevent thermal destruction is even larger so that the bulk of the armor layer will melt and evaporate. - Abstract: We analyze the first wall blanket W/EUROFER configuration for DEMO under steady-state normal operation and off-normal conditions, such as vertical displacement events (VDE) and runaway electrons (RE). The main issue is to find the optimal thickness of the W armor which will prevent tungsten surface from evaporation and melting and, on the other hand, will keep EUROFER below the critical thermal stresses. Under steady-state operation heat transfer into the coolant must remain below the critical heat flux (CHF) to avoid the possible severe degradation of the coolant heat

  3. Evaluation of energy and particle impact on the plasma facing components in DEMO

    Igitkhanov, Yuri, E-mail: juri.gitkhanov@ihm.fzk.de [Karlsruhe Institute of Technology, IHM, Karlsruhe (Germany); Bazylev, Boris [Karlsruhe Institute of Technology, IHM, Karlsruhe (Germany)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer We analyze the first wall blanket W/EUROFER configuration for DEMO under steady-state normal operation and off-normal conditions, such as vertical displacements events (VDE) and runaway electrons (RE). The main issue is to find the optimal thickness of the W armor which will prevent tungsten surface from evaporation and melting and, on the other hand, will keep EUROFER below the critical thermal stresses. Black-Right-Pointing-Pointer The minimum thickness of the tungsten amour about 3 mm for W/EUROFER sandwich structure will keep the maximum EUROFER temperature below the critical limit for EUROFER steel under steady-state operation and ITER like cooling conditions. Black-Right-Pointing-Pointer The W surface temperature and the max. EUROFER temperature are increasing with incoming heat flux. For reference conditions (Dw {approx}3 mm, DEUROFER {approx}4 mm) the maximum tolerable heat flux which does not causes in thermal stresses in structural material is about {approx}13.5 MW/m{sup 2}. Black-Right-Pointing-Pointer The RE deposit their energy deeper into W armour and for energies {>=}50 MJ/m{sup 2} and deposition times {<=}0.1 s, the minimum armor thickness required to prevent EUROFER from thermal distraction is {>=}1.4 cm. Black-Right-Pointing-Pointer However, at this thickness the W surface melts. For higher RE energy deposition rates ({>=}100 MJ/m{sup 2} in 0.1 s), the required armor thickness to prevent thermal destruction is even larger so that the bulk of the armor layer will melt and evaporate. - Abstract: We analyze the first wall blanket W/EUROFER configuration for DEMO under steady-state normal operation and off-normal conditions, such as vertical displacement events (VDE) and runaway electrons (RE). The main issue is to find the optimal thickness of the W armor which will prevent tungsten surface from evaporation and melting and, on the other hand, will keep EUROFER below the critical thermal stresses. Under steady

  4. Compact radio sources

    Altschuler, D.R.

    1975-01-01

    Eighty-seven compact radio sources were monitored between 1971 and 1974 with the National Radio Astronomy Observatory interferometer. Both flux density and polarization were measured at intervals of about one month at wavelengths of 3.7 and 11.1 cms. Forty-four sources showed definite variability in their total and/or polarized flux density. The variations in polarization were of a shorter time scale than the corresponding flux density variations. Some of the qualitative features of an expanding source model were observed. The data suggest that some form of injection of relativistic electrons is taking place. The absence of significant depolarization in the variable sources indicates that only a small fraction of the mass of the radio outburst is in the form of non-relativistic plasma. Some of the objects observed belong to the BL-Lacertal class. It is shown that this class is very inhomogeneous in its radio properties. For the violently variable BL-Lacertal type objects the spectrum, flux variations and polarization data strongly suggest that these are very young objects

  5. Compact Dexterous Robotic Hand

    Lovchik, Christopher Scott (Inventor); Diftler, Myron A. (Inventor)

    2001-01-01

    A compact robotic hand includes a palm housing, a wrist section, and a forearm section. The palm housing supports a plurality of fingers and one or more movable palm members that cooperate with the fingers to grasp and/or release an object. Each flexible finger comprises a plurality of hingedly connected segments, including a proximal segment pivotally connected to the palm housing. The proximal finger segment includes at least one groove defining first and second cam surfaces for engagement with a cable. A plurality of lead screw assemblies each carried by the palm housing are supplied with power from a flexible shaft rotated by an actuator and output linear motion to a cable move a finger. The cable is secured within a respective groove and enables each finger to move between an opened and closed position. A decoupling assembly pivotally connected to a proximal finger segment enables a cable connected thereto to control movement of an intermediate and distal finger segment independent of movement of the proximal finger segment. The dexterous robotic hand closely resembles the function of a human hand yet is light weight and capable of grasping both heavy and light objects with a high degree of precision.

  6. Compact stellarator coils

    Pomphrey, N.; Berry, L.A.; Boozer, A.H.

    2001-01-01

    Experimental devices to study the physics of high-beta (β>∼4%), low aspect ratio (A<∼4.5) stellarator plasmas require coils that will produce plasmas satisfying a set of physics goals, provide experimental flexibility, and be practical to construct. In the course of designing a flexible coil set for the National Compact Stellarator Experiment, we have made several innovations that may be useful in future stellarator design efforts. These include: the use of Singular Value Decomposition methods for obtaining families of smooth current potentials on distant coil winding surfaces from which low current density solutions may be identified; the use of a Control Matrix Method for identifying which few of the many detailed elements of the stellarator boundary must be targeted if a coil set is to provide fields to control the essential physics of the plasma; the use of Genetic Algorithms for choosing an optimal set of discrete coils from a continuum of potential contours; the evaluation of alternate coil topologies for balancing the tradeoff between physics objective and engineering constraints; the development of a new coil optimization code for designing modular coils, and the identification of a 'natural' basis for describing current sheet distributions. (author)

  7. Compact neutron generator

    Leung, Ka-Ngo; Lou, Tak Pui

    2005-03-22

    A compact neutron generator has at its outer circumference a toroidal shaped plasma chamber in which a tritium (or other) plasma is generated. A RF antenna is wrapped around the plasma chamber. A plurality of tritium ion beamlets are extracted through spaced extraction apertures of a plasma electrode on the inner surface of the toroidal plasma chamber and directed inwardly toward the center of neutron generator. The beamlets pass through spaced acceleration and focusing electrodes to a neutron generating target at the center of neutron generator. The target is typically made of titanium tubing. Water is flowed through the tubing for cooling. The beam can be pulsed rapidly to achieve ultrashort neutron bursts. The target may be moved rapidly up and down so that the average power deposited on the surface of the target may be kept at a reasonable level. The neutron generator can produce fast neutrons from a T-T reaction which can be used for luggage and cargo interrogation applications. A luggage or cargo inspection system has a pulsed T-T neutron generator or source at the center, surrounded by associated gamma detectors and other components for identifying explosives or other contraband.

  8. Compact tokamak reactors

    Wootton, A.J.; Wiley, J.C.; Edmonds, P.H.; Ross, D.W.

    1997-01-01

    The possible use of tokamaks for thermonuclear power plants is discussed, in particular tokamaks with low aspect ratio and copper toroidal field coils. Three approaches are presented. First, the existing literature is reviewed and summarized. Second, using simple analytic estimates, the size of the smallest tokamak to produce an ignited plasma is derived. This steady state energy balance analysis is then extended to determine the smallest tokamaks power plant, by including the power required to drive the toroidal field and by considering two extremes of plasma current drive efficiency. Third, the analytic results are augmented by a numerical calculation that permits arbitrary plasma current drive efficiency and different confinement scaling relationships. Throughout, the importance of various restrictions is emphasized, in particular plasma current drive efficiency, plasma confinement, plasma safety factor, plasma elongation, plasma beta, neutron wall loading, blanket availability and recirculation of electric power. The latest published reactor studies show little advantage in using low aspect ratios to obtain a more compact device (and a low cost of electricity) unless either remarkably high efficiency plasma current drive and low safety factor are combined, or unless confinement (the H factor), the permissible elongation and the permissible neutron wall loading increase as the aspect ratio is reduced. These results are reproduced with the analytic model. (author). 22 refs, 3 figs

  9. Towards a reduced activation structural materials database for fusion DEMO reactors

    Moeslang, A.; Diegele, E.; Laesser, R.; Klimiankou, M.; Lindau, R.; Materna-Morris, E.; Rieth, M.; Lucon, E.; Petersen, C.; Schneider, H.-C.; Pippan, R.; Rensman, J.W.; Schaaf, B. van der; Tavassoli, F.

    2005-01-01

    The development of First Wall, Blanket and Divertor materials which are capable of withstanding many years the high neutron and heat fluxes, is a critical path to fusion power. Therefore, the timely availability of a sound materials database has become an indispensable element in international fusion road maps. In order to provide materials design data for short term needs of ITER Test Blanket Modules and for a DEMOnstration fusion reactor, a wealth of R and D results on the European reduced activation ferritic-martensitic steel EUROFER, and on oxide dispersion strengthened variants are being characterized, mainly in the temperature window 250-650 deg. C. The characterisation includes irradiations up to 15 dpa in the mixed spectrum reactor HFR and up to 75 dpa in the fast breeder reactor BOR60. Industrial EUROFER-batches of 3.5 and 7.5 tons have been produced with a variety of semi-finished, quality-assured product forms. To increase thermal efficiency of blankets, high temperature resistant SiC f /SiC channel inserts for liquid metal coolant tubes are also developed. Regarding radiation damage resistance, a broad based reactor irradiation programs counts several steps from ≤5dpa (ITER TBMs) up to 75 dpa (DEMO). For the European divertor designers, a materials data base is presently being set up for pure W and W alloys, and related reactor irradiations are foreseen with temperatures from 650-1000 deg. C. (author)

  10. Linguistic Justice for which Demos? The Democratic Legitimacy of Language Regime Choices

    Garcia Núria

    2016-10-01

    Full Text Available In the European Union language regime debate, theorists of multiculturalism and cosmopolitanism have framed their arguments in reference to different theories of justice and democracy. Philippe Van Parijs advocates the diffusion of a lingua franca, namely English, as means of changing the scale of the justificatory community to the European level and allowing the creation of a transnational demos. Paradoxically, one key dimension of democracy has hardly been addressed in this discussion: the question of the democratic legitimacy of language regime choices and citizens’ preferences on the different language regime scenarios. Addressing the question of the congruence of language policy choices operated by national and European elites and ordinary citizens’ preferences, this paper argues first that the dimension of democratic legitimacy is crucial and needs to be taken into account in discussions around linguistic justice. Criticizing the assumption of a direct correspondence between individuals’ language learning choices and citizens’ language regime preferences made by different authors, the analysis shows the ambivalence of citizens’ preferences measured by survey data. The article secondly raises the question of the boundaries of the political community at which the expression of citizens’ preferences should be measured and demonstrates that the outcome and the fairness of territorial linguistic regimes may vary significantly according to the level at which this democratic legitimacy is taken into account.

  11. Neutronic analysis of a dual He/LiPb coolant breeding blanket for DEMO

    Catalan, J.P.; Ogando, F.; Sanz, J.; Palermo, I.; Veredas, G.; Gomez-Ros, J.M.; Sedano, L.

    2011-01-01

    A conceptual design of a DEMO fusion reactor is being developed under the Spanish Breeding Blanket Technology Programme: TECNO F US based on a He/LiPb dual coolant blanket as reference design option. The following issues have been analyzed to address the demonstration of the neutronic reliability of this conceptual blanket design: power amplification capacity of the blanket, tritium breeding capability for fuel self-sufficiency, power deposition due to nuclear heating in superconducting coils and material damage (dpa, gas production) to estimate the operational life of the steel-made structural components in the blanket and vacuum vessel (VV). In order to optimize the shielding of the coils different combinations of water and steel have been considered for the gap of the VV. The used neutron source is based on an axi-symmetric 2D fusion reaction profile for the given plasma equilibrium configuration. MCNPX has been used for transport calculations and ACAB has been used to handle gas production and damage energy cross sections.

  12. Comprehensive structural analysis of the HCPB demo blanket under thermal, mechanical, electromagnetic and radiation induced loads

    Boccaccini, L.V.; Norajitra, P.; Ruatto, P.; Scaffidi-Argentina, F.

    1998-01-01

    For the helium-cooled pebble bed (HCPB) blanket, which is one of the two reference concepts studied within the European Demo Development Program, a comprehensive finite element (FEM) structural analysis has been performed. The analysis refers to the steady-state operating conditions of an outboard blanket segment. On the basis of a three-dimensional model of radial-toroidal sections of the segment box, thermal stresses caused by the temperature gradients in the blanket structure have been calculated. Furthermore, the mechanical loads due to coolant pressure in normal operating conditions as well as an accidental over-pressurization of the blanket box have been accounted for. The stresses caused by a central plasma major disruption from an initial current of 20 MA to zero in 20 ms have been also taken into account. Radiation-induced dimensional changes of breeder and multiplier material caused by both helium production and neutron damage, have also been evaluated and discussed. All the above loads have been combined as input for a FEM stress analysis and the resulting stress distribution has been evaluated according to the American Society of Mechanical Engineers (ASME) norms. (orig.)

  13. Development of DEMO-FNS tokamak for fusion and hybrid technologies

    Kuteev, B. V.; Azizov, E. A.; Alexeev, P. N.; Ignatiev, V. V.; Subbotin, S. A.; Tsibulskiy, V. F.

    2015-07-01

    The history of fusion-fission hybrid systems based on a tokamak device as an extremely efficient DT-fusion neutron source has passed through several periods of ample research activity in the world since the very beginning of fusion research in the 1950s. Recently, a new roadmap of the hybrid program has been proposed with the goal to build a pilot hybrid plant (PHP) in Russia by 2030. Development of the DEMO-FNS tokamak for fusion and hybrid technologies, which is planned to be built by 2023, is the key milestone on the path to the PHP. This facility is in the phase of conceptual design aimed at providing feasibility studies for a full set of steady state tokamak technologies at a fusion energy gain factor Q ˜ 1, fusion power of ˜40 MW and opportunities for testing a wide range of hybrid technologies with the emphasis on continuous nuclide processing in molten salts. This paper describes the project motivations, its current status and the key issues of the design.

  14. Results of availability imposed configuration details developed for K-DEMO

    Brown, Tom, E-mail: tbrown@pppl.gov [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Titus, Peter; Brooks, Art; Zhang, Han; Neilson, Hutch [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Im, Kihak; Kim, Keeman [National Fusion Research Institute, Daejeon 305-806 (Korea, Republic of)

    2016-11-01

    The Korean fusion demonstration reactor (K-DEMO) has completed a two year study looking at key Tokamak components and configuration options in preparation of a conceptual design phase. A key part of a device configuration centers on defining an arrangement that enhances the ability to reach high availability values by defining design solutions that foster simplified maintenance operations. To maximize the size and minimize the number of in-vessel components enlarged TF coils were defined that incorporate a pair of windings within each coil to mitigate pressure drop issues and to reduce the cost of the coils. A semi-permanent shield structure was defined to develop labyrinth interfaces between double-null plasma contoured shield modules, provide an entity to align blanket components and provide support against disruption loads—with a load path that equilibrates blanket, TF and PF loads through a base structure. Blanket piping services and auxiliary systems that interface with in-vessel components have played a major role in defining the overall device arrangement—concept details will be presented along with general arrangement features and preliminary results obtained from disruption analysis.

  15. Effect of design geometry of the demo first wall on the plasma heat load

    Yu. Igitkhanov

    2016-12-01

    Full Text Available In this work we analyse the effect of W armour surface shaping on the heat load on the W/EUROFER DEMO sandwich type first wall blanket module with the water coolant. The armour wetted area is varied by changing the inclination and height of the «roof» type armor surface. The deleterious effect of leading edge at the tiles corner caused by misalignment is replaced in current design by rounded corners. Analysis has been carried out by means of the MEMOS code to assess the influence of the thickness of the layers and effect of the magnetic field inclination. Calculations show the evolution of the maximum temperatures in the tungsten, EUROFER, Cu allow and the stainless-steel water tube for different level of surface inclination (chamfering and in the case of rounded corners used in the current design. It is shown that the blanket module materials remain within a proper temperature range only at shallow incident angle if the width of EUROFER is reduced at list twice compare with the reference case.

  16. Tritium transport in the water cooled Pb-17Li blanket concept of DEMO

    Reiter, F.; Tominetti, S.; Perujo, A.

    1992-01-01

    The code TIRP has been used to calculate the time dependence of tritium inventory and tritium permeation into the coolant and into the first wall boxes in the water cooled Pb-17Li blanket concept of DEMO. The calculations have been performed for the martensitic steel MANET and the austenitic steel AISI 316L as blanket structure materials, for water or helium cooling and for convective or no motion of the liquid breeder in the blanket. Tritium inventories are rather low in blankets with MANET structure and higher in those with AISI 316L structure. Tritium permeation rates are too high in both blankets. Further calculations on tritium inventory and permeation are therefore presented for blankets with TiC permeation barriers of 1 μm thickness on various surfaces of the blanket structure and for blankets with any permeation barriers in function of their thickness, tritium diffusivities, tritium surface recombination rates and atomic densities. These last calculations have been performed for a blanket with coatings on the outer surfaces of the blanket and with a tritium residence time of 10 4 s and for a blanket with coatings on both sides of the cooling tubes and stagnant Pb-17Li in the blanket. The second case for a blanket with MANET structure presents a very interesting solution for tritium recovery by permeation into and pumping from the first wall boxes. (orig.)

  17. Participation as Post-Fordist Politics: Demos, New Labour, and Science Policy

    2010-01-01

    In recent years, British science policy has seen a significant shift ‘from deficit to dialogue’ in conceptualizing the relationship between science and the public. Academics in the interdisciplinary field of Science and Technology Studies (STS) have been influential as advocates of the new public engagement agenda. However, this participatory agenda has deeper roots in the political ideology of the Third Way. A framing of participation as a politics suited to post-Fordist conditions was put forward in the magazine Marxism Today in the late 1980s, developed in the Demos thinktank in the 1990s, and influenced policy of the New Labour government. The encouragement of public participation and deliberation in relation to science and technology has been part of a broader implementation of participatory mechanisms under New Labour. This participatory program has been explicitly oriented toward producing forms of social consciousness and activity seen as essential to a viable knowledge economy and consumer society. STS arguments for public engagement in science have gained influence insofar as they have intersected with the Third Way politics of post-Fordism. PMID:21258426

  18. Nuclear and thermal analyses of supercritical-water-cooled solid breeder blanket for fusion DEMO reactor

    Yanagi, Yoshihiko; Sato, Satoshi; Enoeda, Mikio; Hatano, Toshihisa; Kikuchi, Shigeto; Kuroda, Toshimasa; Kosaku, Yasuo; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    2001-11-01

    Within a design study of a fusion DEMO reactor aiming at demonstrating technologies of fusion power plant, supercritical water is applied as a coolant of solid breeder blanket to attain high thermal efficiency. The blanket has multi-layer composed of solid breeder pebbles (Li{sub 2}O) and neutron multiplier pebbles (Be) which are radially separated by cooling panels. The first wall and the breeding region are cooled by supercritical water below and above the pseudo-critical temperature, respectively. Temperature distribution and tritium breeding ratio (TBR) have been estimated by one-dimensional nuclear and thermal calculations. The local TBR as high as 1.47 has been obtained after optimization of temperature distribution in the breeder region under the following conditions: neutron wall loading of 5 MW/m{sup 2}, {sup 6}Li enrichment of 30% and coolant temperature at inlet of breeder region of 380degC. In the case of the higher coolant temperature 430degC of the breeder region the local TBR was reduced to be 1.40. This means that the net TBR higher than 1.0 could be expected with the supercritical-water-cooled blanket, whose temperature distribution in the breeder region would be optimized by following the coolant temperature, and where a coverage of the breeder region is assumed to be 70%. (author)

  19. Integrated core-SOL simulations of L-mode plasma in ITER and Indian demo

    Wisitsorasak, Apiwat; Onjun, Thawatchai; Kanjanaput, Wittawat

    2015-01-01

    Core-SOL simulations are carried out using 1.5D BALDUR integrated predictive modeling code to investigate tokamak plasma in ITER and Indian DEMO reactors operating in low confinement mode (L-Mode). In each simulation, the plasma current, temperature, and density profiles in both core and SOL region are evolved self-consistency. The SOL is simulated by integrating the fluid equations, including sources, along the field lines. The solutions in SOL subsequently provide as the boundary conditions of core plasma region on low-confinement mode. The core plasma transport model is described using a combination of anomalous transport by Multi-Mode-Model version 2001 (MMM2001) and neoclassical transport calculated by NCLASS module together with the toroidal velocity based on the torque due to Neoclassical Toroidal Viscosity (NTV). In addition, a sensitivity analysis is explored by varying plasma parameters, such as plasma density and auxiliary heating power. Furthermore, the ignition tests are conducted to observed plasma response in each design after shutting down an auxiliary heating. (author)

  20. HCPB TBM thermo mechanical design: Assessment with respect codes and standards and DEMO relevancy

    Cismondi, F.; Kecskes, S.; Aiello, G.

    2011-01-01

    In the frame of the activities of the European TBM Consortium of Associates the Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) is developed in Karlsruhe Institute of Technology (KIT). After performing detailed thermal and fluid dynamic analyses of the preliminary HCPB TBM design, the thermo mechanical behaviour of the TBM under typical ITER loads has to be assessed. A synthesis of the different design options proposed has been realized building two different assemblies of the HCPB-TBM: these two assemblies and the analyses performed on them are presented in this paper. Finite Element thermo-mechanical analyses of two detailed 1/4 scaled models of the HCPB-TBM assemblies proposed have been performed, with the aim of verifying the accordance of the mechanical behaviour with the criteria of the design codes and standards. The structural design limits specified in the codes and standard are discussed in relation with the EUROFER available data and possible damage modes. Solutions to improve the weak structural points of the present design are identified and the DEMO relevancy of the present thermal and structural design parameters is discussed.

  1. Role of the lower hybrid spectrum in the current drive modeling for DEMO scenarios

    Cardinali, A.; Castaldo, C.; Cesario, R.; Santini, F.; Amicucci, L.; Ceccuzzi, S.; Galli, A.; Mirizzi, F.; Napoli, F.; Panaccione, L.; Schettini, G.; Tuccillo, A. A.

    2017-07-01

    The active control of the radial current density profile is one of the major issues of thermonuclear fusion energy research based on magnetic confinement. The lower hybrid current drive could in principle be used as an efficient tool. However, previous understanding considered the electron temperature envisaged in a reactor at the plasma periphery too large to allow penetration of the coupled radio frequency (RF) power due to strong Landau damping. In this work, we present new numerical results based on quasilinear theory, showing that the injection of power spectra with different {n}// widths of the main lobe produce an RF-driven current density profile spanning most of the outer radial half of the plasma ({n}// is the refractive index in a parallel direction to the confinement magnetic field). Plasma kinetic profiles envisaged for the DEMO reactor are used as references. We demonstrate the robustness of the modeling results concerning the key role of the spectral width in determining the lower hybrid-driven current density profile. Scans of plasma parameters are extensively carried out with the aim of excluding the possibility that any artefact of the utilised numerical modeling would produce any novelty. We neglect here the parasitic effect of spectral broadening produced by linear scattering due to plasma density fluctuations, which mainly occurs for low magnetic field devices. This effect will be analyzed in other work that completes the report on the present breakthrough.

  2. Metal Hall sensors for the new generation fusion reactors of DEMO scale

    Bolshakova, I.; Bulavin, M.; Kargin, N.; Kost, Ya.; Kuech, T.; Kulikov, S.; Radishevskiy, M.; Shurygin, F.; Strikhanov, M.; Vasil'evskii, I.; Vasyliev, A.

    2017-11-01

    For the first time, the results of on-line testing of metal Hall sensors based on nano-thickness (50-70) nm gold films, which was conducted under irradiation by high-energy neutrons up to the high fluences of 1 · 1024 n · m-2, are presented. The testing has been carried out in the IBR-2 fast pulsed reactor in the neutron flux with the intensity of 1.5 · 1017 n · m-2 · s-1 at the Joint Institute for Nuclear Research. The energy spectrum of neutron flux was very close to that expected for the ex-vessel sensors locations in the ITER experimental reactor. The magnetic field sensitivity of the gold sensors was stable within the whole fluence range under research. Also, sensitivity values at the start and at the end of irradiation session were equal within the measurement error (<1%). The results obtained make it possible to recommend gold sensors for magnetic diagnostics in the new generation fusion reactors of DEMO scale.

  3. Compact Holographic Data Storage

    Chao, T. H.; Reyes, G. F.; Zhou, H.

    2001-01-01

    NASA's future missions would require massive high-speed onboard data storage capability to Space Science missions. For Space Science, such as the Europa Lander mission, the onboard data storage requirements would be focused on maximizing the spacecraft's ability to survive fault conditions (i.e., no loss in stored science data when spacecraft enters the 'safe mode') and autonomously recover from them during NASA's long-life and deep space missions. This would require the development of non-volatile memory. In order to survive in the stringent environment during space exploration missions, onboard memory requirements would also include: (1) survive a high radiation environment (1 Mrad), (2) operate effectively and efficiently for a very long time (10 years), and (3) sustain at least a billion write cycles. Therefore, memory technologies requirements of NASA's Earth Science and Space Science missions are large capacity, non-volatility, high-transfer rate, high radiation resistance, high storage density, and high power efficiency. JPL, under current sponsorship from NASA Space Science and Earth Science Programs, is developing a high-density, nonvolatile and rad-hard Compact Holographic Data Storage (CHDS) system to enable large-capacity, high-speed, low power consumption, and read/write of data in a space environment. The entire read/write operation will be controlled with electrooptic mechanism without any moving parts. This CHDS will consist of laser diodes, photorefractive crystal, spatial light modulator, photodetector array, and I/O electronic interface. In operation, pages of information would be recorded and retrieved with random access and high-speed. The nonvolatile, rad-hard characteristics of the holographic memory will provide a revolutionary memory technology meeting the high radiation challenge facing the Europa Lander mission. Additional information is contained in the original extended abstract.

  4. Compact instantaneous water heater

    Azevedo, Jorge G.W.; Machado, Antonio R.; Ferraz, Andre D.; Rocha, Ivan C.C. da; Konishi, Ricardo [Companhia de Gas de Santa Catarina (SCGAS), Florianopolis, SC (Brazil); Lehmkuhl, Willian A.; Francisco Jr, Roberto W.; Hatanaka, Ricardo L.; Pereira, Fernando M.; Oliveira, Amir A.M. [Universidade Federal de Santa Catarina (UFSC), Florianopolis, SC (Brazil)

    2012-07-01

    This paper presents an experimental study of combustion in an inert porous medium in a liquid heating device application. This project aims to increase efficiency in the application of natural gas in residential and commercial sectors with the use of advanced combustion and heat transfer. The goal is to facilitate the development of a high performance compact water heater allowing hot water supply for up to two simultaneous showers. The experiment consists in a cylindrical porous burner with an integrated annular water heat exchanger. The reactants were injected radially into the burner and the flame stabilizes within the porous matrix. The water circulates in a coiled pipe positioned at the center of the burner. This configuration allows for heat transfer by conduction and radiation from the solid matrix to the heat exchanger. This article presented preliminary experimental results of a new water heater based on an annular porous burner. The range of equivalence ratios tested varied from 0.65 to 0.8. The power range was varied from 3 to 5 kW. Increasing the equivalence ratio or decreasing the total power input of the burner resulted in increased thermal efficiencies of the water heater. Thermal efficiencies varying from 60 to 92% were obtained. The condition for the goal of a comfortable bath was 20 deg C for 8-12 L/min. This preliminary prototype has achieved water temperature of 11deg C for 5 L/min. Further optimizations will be necessary in order to achieve intense heating with high thermal efficiency. (author)

  5. What Is Business's Social Compact?

    Avishai, Bernard

    1994-01-01

    Under the "new" social compact, businesses must focus on continuous learning and thus have both an obligation to support teaching and an opportunity to profit from it. Learning organizations must also be teaching organizations. (SK)

  6. Collapse settlement in compacted soils

    Booth, AR

    1977-01-01

    Full Text Available Research into collapse settlement in compacted soils is described, with special reference to recent cases in Southern Africa where collapse settlement occurred in road embankments following wetting of the soil. The laboratory work described...

  7. DEMOS PLUS. Robot for decontaminating soils and cavity walls of the reactor and fuel pools NPP primarily during periods of recharging fuel; DEMOS PLUS. Robot para la descontaminacion de suelos y paredes de la cavidad de reactor y piscinas de combustible de CC.NN. principalmente durante los periodos de recarga de combustible

    Lacalle Bayo, J.; Vaquer Perez, J. I.; Rosello Garcia, J. I.

    2014-07-01

    In this work the robot Plus Demos, equipment that has been developed by GD Energy Services from the redesign and development of robot Demos show, which took place on last year. This evolution has given the team greater capabilities, highlighting the decontamination of vertical surfaces. The main objective pursued is to minimize operational doses to workers operating in cavity as well as the risk of surface contamination during them. (Author)

  8. Lecture-Room Interference Demo Using a Glass Plate and a Laser Beam Focused on It

    Ageev, Leonid A.; Yegorenkov, Vladimir D.

    2010-01-01

    We describe a simple case of non-localized interference produced with a glass plate and a laser beam focused on it. The proposed setup for observing interference is compact when semiconductor lasers are employed, and it is well suited for demonstration and comparison of interference in reflected and transmitted light in a large lecture-room. This…

  9. Maximal design basis accident of fusion neutron source DEMO-TIN

    Kolbasov, B. N., E-mail: Kolbasov-BN@nrcki.ru [National Research Center Kurchatov Institute (Russian Federation)

    2015-12-15

    When analyzing the safety of nuclear (including fusion) facilities, the maximal design basis accident at which the largest release of activity is expected must certainly be considered. Such an accident is usually the failure of cooling systems of the most thermally stressed components of a reactor (for a fusion facility, it is the divertor or the first wall). The analysis of safety of the ITER reactor and fusion power facilities (including hybrid fission–fusion facilities) shows that the initial event of such a design basis accident is a large-scale break of a pipe in the cooling system of divertor or the first wall outside the vacuum vessel of the facility. The greatest concern is caused by the possibility of hydrogen formation and the inrush of air into the vacuum chamber (VC) with the formation of a detonating mixture and a subsequent detonation explosion. To prevent such an explosion, the emergency forced termination of the fusion reaction, the mounting of shutoff valves in the cooling systems of the divertor and the first wall or blanket for reducing to a minimum the amount of water and air rushing into the VC, the injection of nitrogen or inert gas into the VC for decreasing the hydrogen and oxygen concentration, and other measures are recommended. Owing to a continuous feed-out of the molten-salt fuel mixture from the DEMO-TIN blanket with the removal period of 10 days, the radioactivity release at the accident will mainly be determined by tritium (up to 360 PBq). The activity of fission products in the facility will be up to 50 PBq.

  10. Low cycle fatigue behavior of ITER-like divertor target under DEMO-relevant operation conditions

    Li, Muyuan; Werner, Ewald [Lehrstuhl für Werkstoffkunde und Werkstoffmechanik, Technische Universität München, Boltzmannstr. 15, 85748 Garching (Germany); You, Jeong-Ha, E-mail: you@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany)

    2015-01-15

    Highlights: • LCF behavior of the cooling tube and the interlayer of an ITER-like divertor target is studied. • For the cooling tube, LCF failure will not be an issue under an HHF load of up to 18 MW/m{sup 2}. • Plastic strain in the interlayer is concentrated at the free surface edge of the bond interface. • The predicted LCF lifetime of the interlayer may not meet the design requirement. - Abstract: In this work the low cycle fatigue (LCF) behavior of the copper alloy cooling tube and the copper interlayer of an ITER-like divertor target is reported for nine different combinations of loading and cooling conditions relevant to DEMO divertor operation. The LCF lifetime is presented as a function of loading and cooling conditions considered here by means of cyclic plasticity simulation and using LCF data of materials relevant for ITER. The numerical predictions indicate, that fatigue failure will not be an issue for the copper alloy tube under a high heat flux (HHF) load of up to 18 MW/m{sup 2} as long as it preserves its initial strength. In contrast, the copper interlayer exhibits significant plastic dissipation at the free surface edge of the bond interface adjacent to the cooling tube, where the LCF lifetime is predicted to be below 3000 load cycles for HHF loads higher than 15 MW/m{sup 2}. Most of the bulk region of the copper interlayer away from the free surface edge does not experience severe plastic fatigue and hence does not pose any critical concern as the LCF lifetime is predicted to be at least 7000 load cycles. LCF lifetime decreases as HHF load is increased or coolant temperature is decreased.

  11. Safety issues related to the intermediate heat storage for the EU DEMO

    Carpignano, Andrea [NEMO group, Dipartimento Energia, Politecnico di Torino, C.so Duca degli Abruzzi 24, 10129 Torino (Italy); Pinna, Tonio [ENEA, 00044 Frascati (Italy); Savoldi, Laura; Sobrero, Giulia; Uggenti, Anna Chiara [NEMO group, Dipartimento Energia, Politecnico di Torino, C.so Duca degli Abruzzi 24, 10129 Torino (Italy); Zanino, Roberto, E-mail: roberto.zanino@polito.it [NEMO group, Dipartimento Energia, Politecnico di Torino, C.so Duca degli Abruzzi 24, 10129 Torino (Italy)

    2016-11-01

    Highlights: • IHS affects only the PHTS and the BoP (Balance of Plant). • PIEs list does not change but IHS influences PIEs evolution. • Additional issues to be addressed in PIEs study due to the implementation of HIS. • No safety/operational major obstacles were found for IHS concept. - Abstract: The functional deviations able to compromise system safety in the EU DEMO Primary Heat Transfer System (PHTS) with intermediate heat storage (IHS) based on molten salts are identified and compared to the deviations identified with PHTS without IHS. The resulting safety issues for the Balance of Plant (BoP) have been taken into account. Functional Failure Mode and Effects Analysis (FFMEA) is used to highlight the Postulated Initiating Events (PIE) of incident/accident sequences and to provide some safety insights during the preliminary design. The architecture of the system with IHS does not introduce new PIE with respect to the case without IHS, but it modifies some of them. In particular the two Postulated Initiating Events that are affected by the presence of IHS are the LOCA in the tubes of the HX between primary and intermediate circuit and the loss of heat sink for the first wall or the breeding zone. In fact the IHS introduces some advantages concerning the stability of the secondary circuit, but some weaknesses are associated to the physical-chemical nature of molten salts, especially oxidizing power, corrosive nature and risk of solidification. These issues can be managed in the design by the introduction of new safety functions.

  12. Overview of physics results from MAST towards ITER/DEMO and the MAST Upgrade

    Meyer, H.; Akers, R.J.; Allan, S.Y.; Appel, L.C.; Ben Ayed, N.; Challis, C.D.; Chapman, I.T.; Ciric, D.; Colyer, G.; Conway, N.J.; Cox, M.; Abel, I.G.; Barnes, M.; Allan, A.; Barratt, N.C.; Asunta, O.; Bradley, J.W.; Canik, J.; Cahyna, P.; Cecconello, M.

    2013-01-01

    New diagnostic, modelling and plant capability on the Mega Ampère Spherical Tokamak (MAST) have delivered important results in key areas for ITER/DEMO and the upcoming MAST Upgrade, a step towards future ST devices on the path to fusion currently under procurement. Micro-stability analysis of the pedestal highlights the potential roles of micro-tearing modes and kinetic ballooning modes for the pedestal formation. Mitigation of edge localized modes (ELM) using resonant magnetic perturbation has been demonstrated for toroidal mode numbers n = 3, 4, 6 with an ELM frequency increase by up to a factor of 9, compatible with pellet fuelling. The peak heat flux of mitigated and natural ELMs follows the same linear trend with ELM energy loss and the first ELM-resolved T i measurements in the divertor region are shown. Measurements of flow shear and turbulence dynamics during L–H transitions show filaments erupting from the plasma edge whilst the full flow shear is still present. Off-axis neutral beam injection helps to strongly reduce the redistribution of fast-ions due to fishbone modes when compared to on-axis injection. Low-k ion-scale turbulence has been measured in L-mode and compared to global gyro-kinetic simulations. A statistical analysis of principal turbulence time scales shows them to be of comparable magnitude and reasonably correlated with turbulence decorrelation time. T e inside the island of a neoclassical tearing mode allow the analysis of the island evolution without assuming specific models for the heat flux. Other results include the discrepancy of the current profile evolution during the current ramp-up with solutions of the poloidal field diffusion equation, studies of the anomalous Doppler resonance compressional Alfvén eigenmodes, disruption mitigation studies and modelling of the new divertor design for MAST Upgrade. The novel 3D electron Bernstein synthetic imaging shows promising first data sensitive to the edge current profile and flows

  13. Optimization of the breeder zone cooling tubes of the DEMO Water-Cooled Lithium Lead breeding blanket

    Di Maio, P.A.; Arena, P.; Bongiovì, G. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, 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, Palermo (Italy); Del Nevo, A. [ENEA Brasimone, Camugnano, BO (Italy); Forte, R. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, Palermo (Italy)

    2016-11-01

    Highlights: • Determination of an optimal configuration for the breeder zone cooling tubes. • Attention has been focused on the toroidal–radial breeder zone cooling tubes lay out. • A theoretical-computational approach based on the Finite Element Method (FEM) has been followed, adopting a qualified commercial FEM code. • Five different configurations have been investigated to optimize the breeder zone cooling tubes arrangement fulfilling all the rules prescribed by safety codes. - Abstract: The determination of an optimal configuration for the breeder zone (BZ) cooling tubes is one of the most important issues in the DEMO Water-Cooled Lithium Lead (WCLL) breeding blanket R&D activities, since BZ cooling tubes spatial distribution should ensure an efficient heat power removal from the breeder, avoiding hotspots occurrence in the thermal field. Within the framework of R&D activities supported by the HORIZON 2020 EUROfusion Consortium action on the DEMO WCLL breeding blanket design, a campaign of parametric analyses has been launched at the Department of Energy, Information Engineering and Mathematical Models of the University of Palermo (DEIM), in close cooperation with ENEA-Brasimone, in order to assess the potential influence of BZ cooling tubes number on the thermal performances of the DEMO WCLL outboard breeding blanket equatorial module under the nominal steady state operative conditions envisaged for it, optimizing their geometric configuration and taking also into account that a large number of cooling pipes can deteriorate the tritium breeding performances of the module. In particular, attention has been focused on the toroidal-radial option for the BZ tube bundles lay-out and a parametric study has been carried out taking into account different tube bundles arrangement within the module. The study has been carried out following a numerical approach, based on the finite element method (FEM), and adopting a qualified commercial FEM code. Results

  14. Compact Intracloud Discharges

    Smith, David A. [Univ. of Colorado, Boulder, CO (United States)

    1998-11-01

    In November of 1993, mysterious signals recorded by a satellite-borne broadband VHF radio science experiment called Blackboard led to a completely unexpected discovery. Prior to launch of the ALEXIS satellite, it was thought that its secondary payload, Blackboard, would most often detect the radio emissions from lightning when its receiver was not overwhelmed by noise from narrowband communication carriers. Instead, the vast majority of events that triggered the instrument were isolated pairs of pulses that were one hundred times more energetic than normal thunderstorm electrical emissions. The events, which came to be known as TIPPs (for transionospheric pulse pairs), presented a true mystery to the geophysics community. At the time, it was not even known whether the events had natural or anthropogenic origins. After two and one half years of research into the unique signals, two ground-based receiver arrays in New Mexico first began to detect and record thunderstorm radio emissions that were consistent with the Blackboard observations. On two occasions, the ground-based systems and Blackboard even recorded emissions that were produced by the same exact events. From the ground based observations, it has been determined that TIPP events areproduced by brief, singular, isolated, intracloud electrical discharges that occur in intense regions of thunderstorms. These discharges have been dubbed CIDS, an acronym for compact intracloud discharges. During the summer of 1996, ground-based receiver arrays were used to record the electric field change signals and broadband HF emissions from hundreds of CIDS. Event timing that was accurate to within a few microseconds made possible the determination of source locations using methods of differential time of arrival. Ionospheric reflections of signals were recorded in addition to groundwave/line-of-sight signals and were used to determine accurate altitudes for the discharges. Twenty-four CIDS were recorded from three

  15. Comparison study on neutronic analysis of the K-DEMO water cooled ceramic breeder blanket using MCNP and ATTILA

    Park, Jong Sung, E-mail: jspark@nfri.re.kr; Kwon, Sungjin; Im, Kihak

    2016-11-01

    Highlights: • A comparison study of main parameter calculations: neutron wall loading (NWL), tritium breeding ratio (TBR), and nuclear heating, on a Korean fusion demonstration reactor (K-DEMO) neutronic analysis model using MCNP and ATTILA was performed to investigate the feasibility of using ATTILA. • The calculation results of this study indicates that ATTILA showed close agreement with MCNP within ranges (3.3–28%). • Partly high discrepancy (17–28%) results between two codes existed to the nuclear heating calculation in high attenuating materials and radially thick structure regions. • The rest of the results showed small differences of NWL calculation (3.3%) and TBR distribution (3.9%). • ATTILA could be acceptable for K-DEMO neutronic analysis considering discrepancy (3.3–28%). - Abstract: A comparison study of main parameter calculations: neutron wall loading (NWL), tritium breeding ratio (TBR), and nuclear heating, on a Korean fusion demonstration reactor (K-DEMO) neutronic analysis model using MCNP and ATTILA was performed to investigate the feasibility of using ATTILA for the main parameter calculations. The model was created by commercial CAD program (Pro-Engineer™) as a 22.5° sector of tokamak consisting of major components such as blankets, shields, divertors, vacuum vessels (VV), toroidal field (TF) coils, and others, which was directly imported into ATTILA by Parasolid file. The discretizing in space, angle, and energy variables were refined for application of the K-DEMO neutronic analysis model through an iterative process since these variables greatly impact on accuracy, solution times, and memory consumptions in ATTILA. The main parameter calculations using ATTILA and the result of comparison studies indicate that the NWL distributions by two codes were almost agreed within discrepancy of 3.3%; the TBR distribution using ATTILA was slightly bigger than MCNP with a difference 3.9%; the nuclear heating values on TF coils and VV

  16. Consequences of the technology survey and gap analysis on the EU DEMO R&D programme in tritium, matter injection and vacuum

    Day, Chr., E-mail: Christian.Day@kit.edu [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Butler, B. [Culham Science Centre (CCFE), Abingdon (United Kingdom); Giegerich, T. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Lang, P.T. [Max-Planck-Institute of Plasma Physics (IPP), Garching (Germany); Lawless, R. [Culham Science Centre (CCFE), Abingdon (United Kingdom); Meszaros, B. [EUROfusion Consortium, Programme Management Unit, Garching (Germany)

    2016-11-01

    Highlights: • The inner fuel cycle architecture of DEMO is developed in a systems engineering approach as a functional break-down diagram, driven by the need for inventory minimisation. • Technologies to fulfil the required functions are discussed and ranked. • Prime technologies are identified and an associated R&D programme is developed. • The core challenges of a DEMO fuel cycle beyond those already addressed in ITER are discussed. - Abstract: In the framework of the EUROfusion Programme, EU is preparing the conceptual design of the inner fuel cycle of a pulsed tokamak DEMO. This paper illustrates a quantified process to shape a R&D programme that exploits as much as possible previous R&D. In an initial step, the high-level requirements are collected and a novel DEMO inner fuel cycle architecture with its three sub-systems vacuum pumping, matter injection (fuelling and injection of plasma enhancement gases) and tritium systems (tritium plant and breeder coolant purification) is delineated, driven by the DEMO key challenge to reduce tritium inventory. Then, a technology survey is carried out to review potential existing solutions for the required process functions and to assess their maturity and risks. Finally, a decision-making scheme is applied to select the most promising candidates. ITER technology is exploited where possible. As a primary result, a fuel cycle architecture is suggested with an advanced tritium plant that avoids full isotope separation in the main loop and with a Direct Internal Recycling path in the vacuum systems to shorten cycle times. For core fuelling, classical inboard pellet injection technology is selected, in principle similar to that proposed for ITER but aiming for higher launch speeds to achieve deep fuelling of the DEMO plasma. Based on these findings, a tailored R&D programme is shaped that tackles the key questions until 2020.

  17. Summary of Self-compacting Concrete Workability

    GUO Gui-xiang; Duan Hong-jun

    2015-01-01

    On the basis of a large number of domestic and foreign literature, situation and development of self-compacting concrete is introduced. Summary of the compacting theory of self-compacting concrete. And some of the factors affecting the workability of self-compacting concrete were discussed and summarized to a certain extent. Aims to further promote the application and research of self-compacting concrete

  18. Clustering of near clusters versus cluster compactness

    Yu Gao; Yipeng Jing

    1989-01-01

    The clustering properties of near Zwicky clusters are studied by using the two-point angular correlation function. The angular correlation functions for compact and medium compact clusters, for open clusters, and for all near Zwicky clusters are estimated. The results show much stronger clustering for compact and medium compact clusters than for open clusters, and that open clusters have nearly the same clustering strength as galaxies. A detailed study of the compactness-dependence of correlation function strength is worth investigating. (author)

  19. Compact magnetic confinement fusion: Spherical torus and compact torus

    Zhe Gao

    2016-05-01

    Full Text Available The spherical torus (ST and compact torus (CT are two kinds of alternative magnetic confinement fusion concepts with compact geometry. The ST is actually a sub-category of tokamak with a low aspect ratio; while the CT is a toroidal magnetic configuration with a simply-connected geometry including spheromak and field reversed pinch. The ST and CT have potential advantages for ultimate fusion reactor; while at present they can also provide unique fusion science and technology contributions for mainstream fusion research. However, some critical scientific and technology issues should be extensively investigated.

  20. Professional Windows Embedded Compact 7

    Phung, Samuel; Joubert, Thierry; Hall, Mike

    2011-01-01

    Learn to program an array of customized devices and solutions As a compact, highly efficient, scalable operating system, Windows Embedded Compact 7 (WEC7) is one of the best options for developing a new generation of network-enabled, media-rich, and service-oriented devices. This in-depth resource takes you through the benefits and capabilities of WEC7 so that you can start using this performance development platform today. Divided into several major sections, the book begins with an introduction and then moves on to coverage of OS design, application development, advanced application developm

  1. Modeling of compact loop antennas

    Baity, F.W.

    1987-01-01

    A general compact loop antenna model which treats all elements of the antenna as lossy transmission lines has been developed. In addition to capacitively-tuned resonant double loop (RDL) antennas the model treats stub-tuned resonant double loop antennas. Calculations using the model have been compared with measurements on full-scale mockups of resonant double loop antennas for ATF and TFTR in order to refine the transmission line parameters. Results from the model are presented for RDL antenna designs for ATF, TFTR, Tore Supra, and for the Compact Ignition Tokamak

  2. Compact accelerator for medical therapy

    Caporaso, George J.; Chen, Yu-Jiuan; Hawkins, Steven A.; Sampayan, Stephen E.; Paul, Arthur C.

    2010-05-04

    A compact accelerator system having an integrated particle generator-linear accelerator with a compact, small-scale construction capable of producing an energetic (.about.70-250 MeV) proton beam or other nuclei and transporting the beam direction to a medical therapy patient without the need for bending magnets or other hardware often required for remote beam transport. The integrated particle generator-accelerator is actuable as a unitary body on a support structure to enable scanning of a particle beam by direction actuation of the particle generator-accelerator.

  3. Compact toroid refueling of reactors

    Gouge, M.J.; Hogan, J.T.; Milora, S.L.; Thomas, C.E.

    1988-04-01

    The feasibility of refueling fusion reactors and devices such as the International Thermonuclear Engineering Reactor (ITER) with high-velocity compact toroids is investigated. For reactors with reasonable limits on recirculating power, it is concluded that the concept is not economically feasible. For typical ITER designs, the compact toroid fueling requires about 15 MW of electrical power, with about 5 MW of thermal power deposited in the plasma. At these power levels, ideal ignition (Q = ∞) is not possible, even for short-pulse burns. The pulsed power requirements for this technology are substantial. 6 ref., 1 figs

  4. Co-compact Gabor Systems on Locally Compact Abelian Groups

    Jakobsen, Mads Sielemann; Lemvig, Jakob

    2016-01-01

    In this work we extend classical structure and duality results in Gabor analysis on the euclidean space to the setting of second countable locally compact abelian (LCA) groups. We formulate the concept of rationally oversampling of Gabor systems in an LCA group and prove corresponding characteriz...

  5. Isometric coactions of compact quantum groups on compact ...

    a compact quantum metric space in the framework of Rieffel, where the ... This problem can be formulated and studied in various settings. ... The spaces we are interested in this paper are metric spaces, both classical and quantum. ... He has given a definition for a quantum symmetry of a classical ...... by the construction of I.

  6. Effect of activation cross-section uncertainties on the radiological assessment of the MFE/DEMO first wall

    Cabellos, O. [Instituto de Fusion Nuclear, Universidad Politecnica de Madrid, Madrid (Spain)]. E-mail: cabellos@din.upm.es; Reyes, S. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Sanz, J. [Instituto de Fusion Nuclear, Universidad Politecnica de Madrid, Madrid (Spain); University Nacional Educacion a Distancia, Dep. Ingenieria Energetica, Juan del Rosal 12, 28040 Madrid (Spain); Rodriguez, A. [University Nacional Educacion a Distancia, Dep. Ingenieria Energetica, Juan del Rosal 12, 28040 Madrid (Spain); Youssef, M. [University of California, Los Angeles, CA (United States); Sawan, M. [University of Wisconsin, Madison, WI (United States)

    2006-02-15

    A Monte Carlo procedure has been applied in this work in order to address the impact of activation cross-sections (XS) uncertainties on contact dose rate and decay heat calculations for the outboard first wall (FW) of a magnetic fusion energy (MFE) demonstration (DEMO) reactor. The XSs inducing the major uncertainty in the prediction of activation related quantities have been identified. Results have shown that for times corresponding to maintenance activities the uncertainties effect is insignificant since the dominant XSs involved in these calculations are based on accurate experimental data evaluations. However, for times corresponding to waste management/recycling activities, the errors induced by the XSs uncertainties, which in this case are evaluated using systematic models, must be considered. It has been found that two particular isotopes, {sup 6}Co and {sup 94}Nb, are key contributors to the global DEMO FW activation uncertainty results. In these cases, the benefit from further improvements in the accuracy of the critical reaction XSs is discussed.

  7. Progress in the engineering design and assessment of the European DEMO first wall and divertor plasma facing components

    Barrett, Thomas R., E-mail: tom.barrett@ukaea.uk [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Ellwood, G.; Pérez, G.; Kovari, M.; Fursdon, M.; Domptail, F.; Kirk, S.; McIntosh, S.C.; Roberts, S.; Zheng, S. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Boccaccini, L.V. [KIT, INR, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); You, J.-H. [Max Planck Institute for Plasma Physics, Boltzmannstr. 2, 85748 Garching (Germany); Bachmann, C. [EUROfusion, PPPT, Boltzmann Str. 2, 85748 Garching (Germany); Reiser, J.; Rieth, M. [KIT, IAM, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Visca, E.; Mazzone, G. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); Arbeiter, F. [KIT, INR, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Domalapally, P.K. [Research Center Rez, Hlavní 130, 250 68 Husinec – Řež (Czech Republic)

    2016-11-01

    Highlights: • The engineering of the plasma facing components for DEMO is an extreme challenge. • PFC overall requirements, methods for assessment and designs status are described. • Viable divertor concepts for 10 MW/m{sup 2} surface heat flux appear to be within reach. • The first wall PFC concept will need to vary poloidally around the wall. • First wall coolant, structural material and PFC topology are open design choices. - Abstract: The European DEMO power reactor is currently under conceptual design within the EUROfusion Consortium. One of the most critical activities is the engineering of the plasma-facing components (PFCs) covering the plasma chamber wall, which must operate reliably in an extreme environment of neutron irradiation and surface heat and particle flux, while also allowing sufficient neutron transmission to the tritium breeding blankets. A systems approach using advanced numerical analysis is vital to realising viable solutions for these first wall and divertor PFCs. Here, we present the system requirements and describe bespoke thermo-mechanical and thermo-hydraulic assessment procedures which have been used as tools for design. The current first wall and divertor designs are overviewed along with supporting analyses. The PFC solutions employed will necessarily vary around the wall, depending on local conditions, and must be designed in an integrated manner by analysis and physical testing.

  8. Thermal-hydraulic analysis of water cooled breeding blanket of K-DEMO using MARS-KS code

    Lee, Jeong-Hun; Park, Il Woong; Kim, Geon-Woo; Park, Goon-Cherl [Seoul National University, Seoul (Korea, Republic of); Cho, Hyoung-Kyu, E-mail: chohk@snu.ac.kr [Seoul National University, Seoul (Korea, Republic of); Im, Kihak [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Highlights: • The thermal design of breeding blanket for the K-DEMO is evaluated using MARS-KS. • To confirm the prediction capability of MARS, the results were compared with the CFD. • The results of MARS-KS calculation and CFD prediction are in good agreement. • A transient simulation was carried out so as to show the applicability of MARS-KS. • A methodology to simulate the entire blanket system is proposed. - Abstract: The thermal design of a breeding blanket for the Korean Fusion DEMOnstration reactor (K-DEMO) is evaluated using the Multidimensional Analysis of Reactor Safety (MARS-KS) code in this study. The MARS-KS code has advantages in simulating transient two-phase flow over computational fluid dynamics (CFD) codes. In order to confirm the prediction capability of the code for the present blanket system, the calculation results were compared with the CFD prediction. The results of MARS-KS calculation and CFD prediction are in good agreement. Afterwards, a transient simulation for a conceptual problem was carried out so as to show the applicability of MARS-KS for a transient or accident condition. Finally, a methodology to simulate the multiple blanket modules is proposed.

  9. Neutronics Design of Helical Type DEMO Reactor FFHR-d1

    Tanaka, T.; Sagara, A.; Goto, T.; Yanagi, N.; Masuzaki, S.; Tamura, H.; Miyazawa, J.; Muroga, T., E-mail: teru@nifs.ac.jp [National Institute for Fusion Science, Toki (Japan)

    2012-09-15

    Full text: Neutronics design study has been performed in a newly started conceptual design activity for a helical type DEMO reactor FFHR-d1. Features of the FFHR-d1 design are enlargement of the basic configurations of reactor components and extrapolation of plasma parameters from those of the helical type plasma experimental machine Large Helical Device (LHD) to achieve the highest feasibility. From the neutronics point of view, a blanket space of FFHR-d1 is severely limited at the inboard of the torus. This is due to the core plasma position shifting to the inboard side under the confinement condition extrapolated from LHD. The first step of the neutronics investigation using the MCNP code has been performed with a simple torus model simulating thin inboard blanket space. A Flibe+Be/Ferritic steel breeding blanket showed preferable performances for both tritium breeding and shielding, and has been adapted as a reference blanket system for FFHR-d1. The investigations indicate that a combination of a 15 cm thick breeding blanket, 55 cm thick WC+B4C shield, i.e., the blanket space of 70 cm, could suppress the fast neutron flux and nuclear heating in the helical coils to the design targets for the neutron wall loading of 1.5 MW/m{sup 2}. Since the outboard side can provide a large space for a 60 cm thick breeding blanket, a fully-covered tritium breeding ratio (TBR) of 1.31 has been obtained in the simple torus model. The neutronics design study has proceeded to the second step using a 3-D helical reactor model. The most important issue in the 3-D neutronics design is a compatibility with the helical divertor design. To achieve a higher TBR and shielding performance, the core plasma has to be covered by the breeding blanket layers as possible. However, the dimensions of the blanket layers are limited by magnetic field lines connecting an edge of the core plasma and divertor pumping ports. After repeating modification of the blanket configuration, the global TBR of 1

  10. Divertor Heat Flux Reduction by Resonant Magnetic Perturbations in the LHD-Type Helical DEMO Reactor

    Yanagi, N.; Sagara, A.; Goto, T.; Masuzaki, S.; Miyazawa, J., E-mail: yanagi@lhd.nifs.ac.jp [National Institute for Fusion Science, Toki (Japan)

    2012-09-15

    Full text: The conceptual design studies of the LHD-type helical fusion DEMO reactor, FFHR-d1, are progressing steadfastly. The LHD-type heliotron magnetic configuration equipped with the built- in helical divertors has a potential to realize low divertor heat flux in spatial average. However, the toroidal asymmetry may give more than a couple of times higher peak heat flux at some locations, as has been experimentally observed in LHD and confirmed by magnetic field-line tracing. By providing radiation dispersion accompanied with a plasma detachment, the heat flux may decrease significantly though the compatibility with a good core plasma confinement is an important issue to be explored. Whereas the engineering difficulties for developing materials to be used under the neutron environment require even further decrease of the heat flux (even though the heliotron is a unique configuration that divertor plates be largely shielded from the direct irradiation of neutrons by breeder blankets). In this respect, we proposed, in the last IAEA FEC, a new strike point sweeping scheme using a set of auxiliary helical coils, termed helical divertor (HD) coils. The HD coils carrying a few percent of the current amplitude of the main helical coils sweep the divertor strike points without altering the core plasma. Though this scheme is effective in dispersing the heat flux in the poloidal direction, the toroidal asymmetry still remains. The AC operation may also give unforeseen engineering difficulties. We here propose that the peak heat flux be mitigated using RMP fields in steady-state. The magnetic field-lines are numerically traced in the vacuum configuration and their footprints coming to the divertor regions are counted. Their fraction plotted as a function of the toroidal angle indicates that the peak heat flux be mitigated to {approx} 20 MW per square meters at 3 GW fusion power generation without having radiation dispersion when an RMP field is applied. We note that the

  11. Current design of the European TBM systems and implications on DEMO breeding blanket

    Ricapito; Calderoni, P. [Fusion for Energy, 08019 Barcelona (Spain); Aiello, A. [ENEA, Bacino del Brasimone, I-40032 Camugnano, Bo (Italy); Ghidersa, B. [Karlsruher Institut für Technologie, D-76021 Karlsruhe (Germany); Poitevin, Y.; Pacheco, J. [Fusion for Energy, 08019 Barcelona (Spain)

    2016-11-01

    -going R&D activities carried out in Europe are presented and discussed. In the last part, different considerations are proposed about the impact of the design and operation of the main HCLL and HCPB-TBM ancillary systems technologies on the design of a DEMO BB.

  12. Structural integrity for DEMO: An opportunity to close the gap from materials science to engineering needs

    Porton, M., E-mail: michael.porton@ccfe.ac.uk [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Wynne, B.P. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); University of Sheffield, Sheffield, South Yorkshire S10 2TN (United Kingdom); Bamber, R.; Hardie, C.D.; Kalsey, M. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)

    2016-11-01

    Highlights: • Key shortfalls in the current approaches to verification of structural integrity are outlined. • Case studies for high integrity applications in other demanding environments are examined. • Relevant lessons are drawn from fission and space for the design stage and through service life. • Future efforts are suggested to align materials and engineering for DEMO structural integrity. - Abstract: It is clear that fusion demonstration devices offer unique challenges due to the myriad, interacting material degradation effects and the numerous, conflicting requirements that must be addressed in order for in-vessel components to deliver satisfactory performance over the required lifetime. The link between mechanical engineering and materials science is pivotal to assure the timely realisation and exploitation of successful fusion power. A key aspect of this link is the verification of structural integrity, achieved at the design stage via structural design criteria against which designs are judged to be sufficiently resilient (or not) to failure, for a given set of loading conditions and desired lifetime. As various demonstration power plant designs progress through their current conceptual design phases, this paper seeks to highlight key shortfalls in this vital link between engineering needs and materials science, offering a perspective on where future attention can be prioritised to maximise impact. Firstly, issues in applying existing structural design criteria to demonstration power plant designs are identified. Whilst fusion offers particular challenges, there are significant insights to be gained from attempts to address such issues for high performance, high integrity applications in other demanding environments. Therefore case studies from beyond fusion are discussed. These offer examples where similar shortfalls have been successfully addressed, via approaches at the design stage and through service lifetime in order to deliver significant

  13. Pressurizing Behavior on Ingress of Coolant into Pebble Bed of Blanket of Fusion DEMO Reactor

    Daigo Tsuru; Mikio Enoeda; Masato Akiba

    2006-01-01

    Solid breeder blankets are being developed as candidate blankets for the Fusion DEMO reactor in Japan. JAEA is performing the development of the water cooled and helium cooled solid breeder blankets. The blanket utilizes ceramic breeder pebbles and multiplier pebbles beds cooled by high pressure water or high pressure helium in the cooling tubes placed in the blanket box structure. In the development of the blanket, it is very important to incorporate the safety technology as well as the performance improvement on tritium production and energy conversion. In the safety design and technology, coolant ingress in the blanket box structure is one of the most important events as the initiators. Especially the thermal hydraulics in the pebble bed in the case of the high pressure coolant ingress is very important to evaluate the pressure propagation and coolant flow behavior. This paper presents the preliminary results of the pressure loss characteristics by the coolant ingress in the pebble bed. Experiments have been performed by using alumina pebble bed (4 litter maximum volume of the pebble bed) and nitrogen gas to simulate the helium coolant ingress into breeder and multiplier pebble beds. Reservoir tank of 10 liter is filled with 1.0 MPa nitrogen. The nitrogen gas is released at the bottom part of the alumina pebble bed whose upper part is open to the atmosphere. The pressure change in the pebble bed is measured to identify the pressure loss. The measured values are compared with the predicted values by Ergun's equation, which is the correlation equation on pressure loss of the flow through porous medium. By the results of the experiments with no constraint on the alumina pebble bed, it was clarified that the measured value agreed in the lower flow rate. However, in the higher flow rate where the pressure loss is high, the measured value is about half of the predicted value. The differences between the measured values and the predicted values will be discussed from

  14. Adaptation of the HCPB DEMO TBM as breeding blanket for ITER : Neutronic and thermal analyses

    Aquaro, D.; Morellini, D.; Cerullo, N.

    2006-01-01

    Two breeding blanket are presently developed in Europe for the DEMO reactor: the first one, the Helium Cooled Lithium Lead (HCLL) uses a liquid breeder while the other , the Helium Cooled Pebble Bed (HCPB), uses a solid breeder in form of pebble bed. The modules of these blankets, called Test Blanket Modules (TBM) will be located in correspondence of the equatorial ports of ITER in order to be tested. ITER FEAT was designed with shielding blankets, therefore in the final stage of the experiment, in the foreseen tritium -deuterium operation phase, the tritium will be supplied to the reactor and not produced inside it. Since the production of tritium is of main importance for the feasibility of a nuclear fusion reactor, perhaps in the ITER final stage, the shielding blanket could be substituted by means of a breeding blanket. The geometry and composition of this breeding blanket would be, of course, similar to that of TBM which demonstrated to have the best performances. This paper illustrates a neutronic and thermal analysis of an hypothetical triziogen blanket for ITER FEAT made similar to a HCPB test module. The main aims of the performed analyses are to determine the Tritium Breeding Ratio (TBR) considering different solid breeders (Li 4 SiO 4 and Li 2 TiO 3 ) with different enrichment in 6 Li and different structural materials (a 9%CRWVTa reduced activation ferritic martensitic steel (EUROFER) or ceramic matrix composites like SiCf/SiC). The breeding blanket design is compared considering the highest value of TBR and the verification of the temperature constraints ( 550 o C for the steel, 950 o C for the breeder and 650 o C for the Beryllium). The neutronic analyses have been performed by means of MCNP-4C code and the thermal analyses using the MSC-MARC code. A TBR about equal 1 was obtained with a SiCf/SiC structural material and a Li 4 SiO 4 breeder. The performed analyses have to be considered preliminary and an academic exercise, nevertheless they could give

  15. Compaction dynamics of crunchy granular material

    Guillard François

    2017-01-01

    Full Text Available Compaction of brittle porous material leads to a wide variety of densification patterns. Static compaction bands occurs naturally in rocks or bones, and have important consequences in industry for the manufacturing of powder tablets or metallic foams for example. Recently, oscillatory compaction bands have been observed in brittle porous media like snow or cereals. We will discuss the great variety of densification patterns arising during the compaction of puffed rice, including erratic compaction at low velocity, one or several travelling compaction bands at medium velocity and homogeneous compaction at larger velocity. The conditions of existence of each pattern are studied thanks to a numerical spring lattice model undergoing breakage and is mapped to the phase diagram of the patterns based on dimensionless characteristic quantities. This also allows to rationalise the evolution of the compaction behaviour during a single test. Finally, the localisation of compaction bands is linked to the strain rate sensitivity of the material.

  16. Compaction dynamics of crunchy granular material

    Guillard, François; Golshan, Pouya; Shen, Luming; Valdès, Julio R.; Einav, Itai

    2017-06-01

    Compaction of brittle porous material leads to a wide variety of densification patterns. Static compaction bands occurs naturally in rocks or bones, and have important consequences in industry for the manufacturing of powder tablets or metallic foams for example. Recently, oscillatory compaction bands have been observed in brittle porous media like snow or cereals. We will discuss the great variety of densification patterns arising during the compaction of puffed rice, including erratic compaction at low velocity, one or several travelling compaction bands at medium velocity and homogeneous compaction at larger velocity. The conditions of existence of each pattern are studied thanks to a numerical spring lattice model undergoing breakage and is mapped to the phase diagram of the patterns based on dimensionless characteristic quantities. This also allows to rationalise the evolution of the compaction behaviour during a single test. Finally, the localisation of compaction bands is linked to the strain rate sensitivity of the material.

  17. Compact objects and accretion disks

    Blandford, Roger; Agol, Eric; Broderick, Avery; Heyl, Jeremy; Koopmans, Leon; Lee, Hee-Won

    2002-01-01

    Recent developments in the spectropolarimetric study of compact objects, specifically black holes (stellar and massive) and neutron stars are reviewed. The lectures are organized around five topics: disks, jets, outflows, neutron stars and black holes. They emphasize physical mechanisms and are

  18. Engineering aspects of compact stellarators

    Nelson, B.E.; Benson, R.D.; Brooks, A.

    2003-01-01

    Compact stellarators could combine the good confinement and high beta of a tokamak with the inherently steady state, disruption-free characteristics of a stellarator. Two U.S. compact stellarator facilities are now in the conceptual design phase: the National Compact Stellarator Experiment (NCSX) and the Quasi- Poloidal Stellarator (QPS). NCSX has a major radius of 1.4 m and a toroidal field up to 2 T. The primary feature of both NCSX and QPS is the set of modular coils that provide the basic magnetic configuration. These coils represent a major engineering challenge due to the complex shape, precise geometric accuracy, and high current density of the windings. The winding geometry is too complex for conventional hollow copper conductor construction. Instead, the modular coils will be wound with flexible, multi strand cable conductor that has been compacted to a 75% copper packing fraction. Inside the NCSX coil set and surrounding the plasma is a highly contoured vacuum vessel. The vessel consists of three identical, 120 deg. segments that are bolted together at double sealed joints. The QPS device has a major radius of 0.9 m, a toroidal field of 1 T, and an aspect ratio of only 2.7. Instead of an internal vacuum vessel, the QPS modular coils will operate in an external vacuum tank. (author)

  19. Compact Circuit Preprocesses Accelerometer Output

    Bozeman, Richard J., Jr.

    1993-01-01

    Compact electronic circuit transfers dc power to, and preprocesses ac output of, accelerometer and associated preamplifier. Incorporated into accelerometer case during initial fabrication or retrofit onto commercial accelerometer. Made of commercial integrated circuits and other conventional components; made smaller by use of micrologic and surface-mount technology.

  20. Compaction and relaxation of biofilms

    Valladares Linares, R.

    2015-06-18

    Operation of membrane systems for water treatment can be seriously hampered by biofouling. A better characterization of biofilms in membrane systems and their impact on membrane performance may help to develop effective biofouling control strategies. The objective of this study was to determine the occurrence, extent and timescale of biofilm compaction and relaxation (decompaction), caused by permeate flux variations. The impact of permeate flux changes on biofilm thickness, structure and stiffness was investigated in situ and non-destructively with optical coherence tomography using membrane fouling monitors operated at a constant crossflow velocity of 0.1 m s−1 with permeate production. The permeate flux was varied sequentially from 20 to 60 and back to 20 L m−2 h−1. The study showed that the average biofilm thickness on the membrane decreased after elevating the permeate flux from 20 to 60 L m−2 h−1 while the biofilm thickness increased again after restoring the original flux of 20 L m−2 h−1, indicating the occurrence of biofilm compaction and relaxation. Within a few seconds after the flux change, the biofilm thickness was changed and stabilized, biofilm compaction occurred faster than the relaxation after restoring the original permeate flux. The initial biofilm parameters were not fully reinstated: the biofilm thickness was reduced by 21%, biofilm stiffness had increased and the hydraulic biofilm resistance was elevated by 16%. Biofilm thickness was related to the hydraulic biofilm resistance. Membrane performance losses are related to the biofilm thickness, density and morphology, which are influenced by (variations in) hydraulic conditions. A (temporarily) permeate flux increase caused biofilm compaction, together with membrane performance losses. The impact of biofilms on membrane performance can be influenced (increased and reduced) by operational parameters. The article shows that a (temporary) pressure increase leads to more

  1. Rate type isotach compaction of consolidated sandstone

    Waal, J.A. de; Thienen-Visser, K. van; Pruiksma, J.P.

    2015-01-01

    Laboratory experiments on samples from a consolidated sandstone reservoir are presented that demonstrate rate type compaction behaviour similar to that observed on unconsolidated sands and soils. Such rate type behaviour can have large consequences for reservoir compaction, surface subsidence and

  2. Siting actions in compacts and nonmember states

    Tullis, J.

    1986-05-01

    This paper examines the status of siting actions in those compacts and states currently progressing with siting studies. The efforts of the Central Compact Commission, Texas, California, Colorado and Illinois are highlighted to illustrate progress, methodology, and problems encountered

  3. Powder compaction in systems of bimodal distribution

    Chattopadhyay, A. K.; Whittemore, O. J., Jr.

    1973-01-01

    The compaction of mixtures involving different particle sizes is discussed. The various stages of the compaction process include the rearrangement of particles, the filling of the interstices of the large particles by the smaller ones, and the change in particle size and shape upon further densification through the application of pressure. Experimental approaches and equipment used for compacting material are discussed together with the theoretical relations of the compacting process.

  4. UV written compact broadband optical couplers

    Olivero, Massimo; Svalgaard, Mikael

    2005-01-01

    In this paper the first demonstration of compact asymmetric directional couplers made by UV writing is presented. The combined performance in terms bandwidth, loss and compactness exceeds that reported using other, more elaborate fabrication techniques.......In this paper the first demonstration of compact asymmetric directional couplers made by UV writing is presented. The combined performance in terms bandwidth, loss and compactness exceeds that reported using other, more elaborate fabrication techniques....

  5. A CFD analysis of flow blockage phenomena in ALFRED LFR demo fuel assembly

    Di Piazza, Ivan, E-mail: ivan.dipiazza@enea.it [Italian National Agency for New Technologies, Energy and Sustainable Economic Development, C.R. ENEA Brasimone (Italy); Magugliani, Fabrizio [Ansaldo Nucleare, ANN, Corso Perrone n.25, Genova (Italy); Tarantino, Mariano [Italian National Agency for New Technologies, Energy and Sustainable Economic Development, C.R. ENEA Brasimone (Italy); Alemberti, Alessandro [Ansaldo Nucleare, ANN, Corso Perrone n.25, Genova (Italy)

    2014-09-15

    Highlights: • URANS simulations were performed on internal flow blockage in HLM fuel assemblies. • Comparison with RELAP results for foot blockage shows a very good agreement. • The temperature peak behind the blockage is dominant for large blockages. • A blockage of ∼15% leads to a maximum clad temperature around 800 °C in 3–4 s. • Local clad temperatures around 1000 °C are reached for blockages of 30% or more. - Abstract: A CFD study was carried out on fluid flow and heat transfer in the Lead-cooled Fuel Pin Bundle of the ALFRED LFR DEMO. In the context of GEN-IV heavy liquid metal-cooled reactors safety studies, the flow blockage in a fuel sub-assembly is considered one of the main issues to be addressed and the most important and realistic accident for LFR fuel assembly. The present paper is a first step toward a detailed analysis of such phenomena, and a CFD model and approach are presented to have a detailed thermo-fluid dynamic picture in the case of blockage. In particular the closed hexagonal, grid-spaced fuel assembly of the LFR ALFRED was modeled and computed. At this stage, the details of the spacer grids were not included, but a conservative analysis has been carried out based on the current main geometrical and physical features. Reactivity feedback, as well as axial power profile, were not included in this analysis. Results indicate that critical conditions, with clad temperatures around ∼900 °C, are reached with blockage larger than 30% in terms of area fraction. Two main effects can be distinguished: a local effect in the wake/recirculation region downstream the blockage and a global effect due to the lower mass flow rate in the blocked subchannels; the former effect gives rise to a temperature peak behind the blockage and it is dominant for large blockages (>20%), while the latter effect determines a temperature peak at the end of the active region and it is dominant for small blockages (<10%). The blockage area was placed at

  6. Invariant subsets under compact quantum group actions

    Huang, Huichi

    2012-01-01

    We investigate compact quantum group actions on unital $C^*$-algebras by analyzing invariant subsets and invariant states. In particular, we come up with the concept of compact quantum group orbits and use it to show that countable compact metrizable spaces with infinitely many points are not quantum homogeneous spaces.

  7. Equationally Compact Acts : Coproducts / Peeter Normak

    Normak, Peeter

    1998-01-01

    In this article equational compactness of acts and its generalizations are discussed. As equational compactness does not carry over to coproducts a slight generalization of c-equational campactness is introduced. It is proved that a coproduct of acts is c-equationally compact if and only if all components are c-equationally campact

  8. Formation and evolution of compact binaries

    Sluijs, Marcel Vincent van der

    2006-01-01

    In this thesis we investigate the formation and evolution of compact binaries. Chapters 2 through 4 deal with the formation of luminous, ultra-compact X-ray binaries in globular clusters. We show that the proposed scenario of magnetic capture produces too few ultra-compact X-ray binaries to explain

  9. Compact sources for eyesafe illumination

    Baranova, Nadia; Pu, Rui; Stebbins, Kenneth; Bystryak, Ilya; Rayno, Michael; Ezzo, Kevin; DePriest, Christopher

    2018-02-01

    Q-peak has demonstrated a compact, pulsed eyesafe laser architecture operating with >10 mJ pulse energies at repetition rates as high as 160 Hz. The design leverages an end-pumped solid-state laser geometry to produce adequate eyesafe beam quality (M2˜4), while also providing a path toward higher-density laser architectures for pulsed eyesafe applications. The baseline discussed in this paper has shown a unique capability for high-pulse repetition rates in a compact package, and offers additional potential for power scaling based on birefringence compensation. The laser consists of an actively Q-switched oscillator cavity producing pulse widths designed to fit within a volume of 3760 cm3. We will discuss details of the optical system design, modeled thermal effects and stress-induced birefringence, as well as experimental advantages of the end-pumped laser geometry, along with proposed paths to higher eyesafe pulse energies.

  10. Magnetohydrodynamical processes near compact objects

    Bisnovatyi Kogan, G.S.

    1979-01-01

    Magnetohydrodynamical processes near compact objects are reviewed in this paper. First the accretion of the magnetized matter into a single black hole and spectra of radiation are considered. Then the magnetic-field phenomena in the disk accretion, when the black hole is in a pair are discussed. Furthermore, the magnetohydrodynamics phenomena during supernova explosion are considered. Finally the magnetohydrodynamics in the accretion of a neutron star is considered in connection With x-ray sources

  11. Compact toroids with Alfvenic flows

    Wang Zhehui; Tang, X.Z.

    2004-01-01

    The Chandrasekhar equilibria form a class of stationary ideal magnetohydrodynamics equilibria stabilized by magnetic-field-aligned Alfvenic flows. Analytic solutions of the Chandrasekhar equilibria are explicitly constructed for both field-reversed configurations and spheromaks. Favorable confinement property of nested closed flux surfaces and the ideal magnetohydrodynamic stability of the compact toroids are of interest for both magnetic trapping of high energy electrons in astrophysics and confinement of high temperature plasmas in laboratory

  12. Durability of Self Compacting Concrete

    Benmarce, A.; Boudjehem, H.; Bendjhaiche, R.

    2011-01-01

    Self compacting concrete (SCC) seem to be a very promising materials for construction thanks to their properties in a fresh state. Studying of the influence of the parameters of specific designed mixes to their mechanical, physical and chemical characteristics in a state hardened is an important stage so that it can be useful for new-to-the-field researchers and designers (worldwide) beginning studies and work involving self compacting concrete. The objective of this research is to study the durability of self compacting concrete. The durability of concrete depends very much on the porosity; the latter determines the intensity of interactions with aggressive agents. The pores inside of concrete facilitate the process of damage, which began generally on the surface. We are interested to measure the porosity of concrete on five SCC with different compositions (w/c, additives) and vibrated concrete to highlight the influence of the latter on the porosity, thereafter on the compressive strength and the transfer properties (oxygen permeability, chloride ion diffusion, capillary absorption). (author)

  13. Comminution circuits for compact itabirites

    Pedro Ferreira Pinto

    Full Text Available Abstract In the beneficiation of compact Itabirites, crushing and grinding account for major operational and capital costs. As such, the study and development of comminution circuits have a fundamental importance for feasibility and optimization of compact Itabirite beneficiation. This work makes a comparison between comminution circuits for compact Itabirites from the Iron Quadrangle. The circuits developed are: a crushing and ball mill circuit (CB, a SAG mill and ball mill circuit (SAB and a single stage SAG mill circuit (SSSAG. For the SAB circuit, the use of pebble crushing is analyzed (SABC. An industrial circuit for 25 million tons of run of mine was developed for each route from tests on a pilot scale (grinding and industrial scale. The energy consumption obtained for grinding in the pilot tests was compared with that reported by Donda and Bond. The SSSAG route had the lowest energy consumption, 11.8kWh/t and the SAB route had the highest energy consumption, 15.8kWh/t. The CB and SABC routes had a similar energy consumption of 14.4 kWh/t and 14.5 kWh/t respectively.

  14. Strange matter in compact stars

    Klähn, Thomas; Blaschke, David B.

    2018-02-01

    We discuss possible scenarios for the existence of strange matter in compact stars. The appearance of hyperons leads to a hyperon puzzle in ab-initio approaches based on effective baryon-baryon potentials but is not a severe problem in relativistic mean field models. In general, the puzzle can be resolved in a natural way if hadronic matter gets stiffened at supersaturation densities, an effect based on the quark Pauli quenching between hadrons. We explain the conflict between the necessity to implement dynamical chiral symmetry breaking into a model description and the conditions for the appearance of absolutely stable strange quark matter that require both, approximately masslessness of quarks and a mechanism of confinement. The role of strangeness in compact stars (hadronic or quark matter realizations) remains unsettled. It is not excluded that strangeness plays no role in compact stars at all. To answer the question whether the case of absolutely stable strange quark matter can be excluded on theoretical grounds requires an understanding of dense matter that we have not yet reached.

  15. Strange matter in compact stars

    Klähn Thomas

    2018-01-01

    Full Text Available We discuss possible scenarios for the existence of strange matter in compact stars. The appearance of hyperons leads to a hyperon puzzle in ab-initio approaches based on effective baryon-baryon potentials but is not a severe problem in relativistic mean field models. In general, the puzzle can be resolved in a natural way if hadronic matter gets stiffened at supersaturation densities, an effect based on the quark Pauli quenching between hadrons. We explain the conflict between the necessity to implement dynamical chiral symmetry breaking into a model description and the conditions for the appearance of absolutely stable strange quark matter that require both, approximately masslessness of quarks and a mechanism of confinement. The role of strangeness in compact stars (hadronic or quark matter realizations remains unsettled. It is not excluded that strangeness plays no role in compact stars at all. To answer the question whether the case of absolutely stable strange quark matter can be excluded on theoretical grounds requires an understanding of dense matter that we have not yet reached.

  16. Modelling of fission gas release in rods from the International DEMO-RAMP-II Project at Studsvik

    Malen, K.

    1983-01-01

    The DEMO-RAMP-II rods had a burn-up of 25-30 MWd/kg U. They were ramped to powers in the range 40-50 kW/m with hold times between 10 s and 4.5 minutes. In spite of the short hold times the fission gas release at the higher powers was more than 1%. With these short hold times it is natural to assume that mixing of released gas with plenum gas is limited. Modelling has been performed using GAPCONSV (a modified GAPCON-THERMAL-2) both with and without mixing of released gas with plenum gas. In particular for the high power-short duration ramps only the ''no mixing'' modelling yields release fractions comparable to the experimental values. (author)

  17. The enhanced pellet centrifuge launcher at ASDEX Upgrade: Advanced operation and application as technology test facility for ITER and DEMO

    Ploeckl, B., E-mail: bernhard.ploeckl@ipp.mpg.de [Max Planck Institute for Plasma Physics, EURATOM Association, Boltzmannstr. 2, 85748 Garching (Germany); Day, Chr. [Karlsruhe Institute of Technology (KIT), 76021 Karlsruhe (Germany); Lamalle, Ph. [ITER Organization, Route de Vinon sur Verdon, CS 90046, 13067 Saint-Paul-lez-Durance (France); Lang, P.T.; Rohde, V.; Viezzer, E. [Max Planck Institute for Plasma Physics, EURATOM Association, Boltzmannstr. 2, 85748 Garching (Germany)

    2015-10-15

    The pellet centrifuge at ASDEX Upgrade has served for more than 20 years as a powerful tool for plasma control. Its recently enhanced control system provides more thorough control over parameters and a detailed view of all measured values. A study has recently been initiated on the conceptual design of an optimized DEMO core particle fuelling system. For this approach, first technical tests aimed on an optimized pellet transfer with respect to the preparation of the solid fuel and the transfer systems have been performed. An investigation of the temperature dependence of transfer efficiency (mass loss due to erosion and broken pellets) has revealed a weak dependence. For ITER, in which it is intended to operate a heating scheme with ICRF minority heating of He-3, test injections are performed using D{sub 2}-pellets as carriers for He-4. Admixing of N{sub 2} was investigated as well.

  18. Creep behavior of 8Cr2WVTa martensitic steel designed for fusion DEMO reactor. An assessment on helium embrittlement resistance

    Yamamoto, Norikazu; Murase, Yoshiharu; Nagakawa, Johsei; Shiba, Kiyoyuki

    2001-01-01

    Mechanical response against transmutational helium production, alternatively susceptibility to helium embrittlement, in a nuclear fusion reactor was examined on 8Cr2WVTa martensitic steel, a prominent structural candidate for advanced fusion systems. In order to simulate DEMO (demonstrative) reactor environments, helium was implanted into the material at 823 K with concentrations up to 1000 appmHe utilizing an α-beam from a cyclotron. Creep rupture properties were subsequently determined at the same temperature and were compared with those of the material without helium. It has been proved that helium caused no meaningful deterioration in terms of both the creep lifetime and rupture elongation. Furthermore, failure occurred completely in a transgranular and ductile manner even after high concentration helium introduction and there was no symptom of grain boundary decohesion which very often arises in helium bearing materials. These facts would mirror preferable resistance of this steel toward helium embrittlement. (author)

  19. Structural design of DEMO Divertor Cassette Body: provisional FEM analysis and introductive application of RCC-MRx design rules

    Frosi, Paolo, E-mail: paolo.frosi@enea.it [Unità Tecnica Fusione-ENEA C.R. Frascati, Via E. Fermi 45, 00044 Frascati (Italy); Mazzone, Giuseppe [Unità Tecnica Fusione-ENEA C.R. Frascati, Via E. Fermi 45, 00044 Frascati (Italy); You, Jeong-Ha [Max Planck Institute of Plasma Physics, Boltzmann Str. 2, 85748 Garching (Germany)

    2016-11-01

    This paper deals with the early steps in developing a structural fem model of DEMO Divertor. The study is focused on the thermal and structural analysis of the Cassette Body: a new geometry has been developed for this component: it is foreseen that the plasma facing component (PFC) will be directly placed on the cassette but for the Dome no choice has been adopted yet. For now the model contains only a suitable schematization of the Cassette Body and its objective is to analyze the effect produced by the main loads (electromagnetic loads, coolant pressure, thermal neutron and convective loads) on itself: an available estimate of loads is that one derived from ITER: for a proper translation some assumptions have been made and they are described in the paper. Now it is not a primary purpose to obtain some definitive statements about stresses, displacements, temperatures and so on; the authors want to construct a set of FEM models that will help all the decisions of DEMO Divertor design in its future development. This set is conceived as a tool that shall be improved to account for all the main enhancements that will be found in geometry, in material properties data and in load evaluations. Moreover, the main design variables (loads, material properties, some geometric items, mesh element size) are defined as parameters. This work considers also an introductive approach for future structural verification of the Divertor Cassette Body: so a concern of the Design and Construction Rules for Mechanical Components of Nuclear Installation (RCC-MRx) has been implemented. The FEM code used is Ansys rel. 15.

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

    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.

  1. Development of Tokamak reactor system code and conceptual studies of DEMO with He Cooled Molten Li blanket

    Hong, B.G.; Lee, Dong Won; Kim, Yong Hi

    2007-01-01

    To develop the concepts of fusion power plants and identify the design parameters, we have been developing the tokamak reactor system code. The system code can take into account a wide range of plasma physics and technology effects simultaneously and it can be used to find design parameters which optimize the given figure of merits. The outcome of the system studies using the system code is to identify which areas of plasma physics and technologies and to what extent should be developed for realization of a given fusion power plant concepts. As an application of the tokamak reactor system code, we investigate the performance of DEMO for early realization with a limited extension from the plasma physics and technology used in the design of the ITER. Main requirements for DEMO are selected as: 1) to demonstrate tritium self-sufficiency, 2) to generate net electricity, and 3) for steady-state operation. The size of plasma is assumed to be same as that of ITER and the plasma parameters which characterize the performance, i.e. normalized β value, β N , confinement improvement factor for the H-mode, H and the ratio of the Greenwald density limit n/n G are assumed to be improved beyond those of ITER: β N >2.0, H>1.0 and n/n G >1.0. Tritium self-sufficiency is provided by the He Cooled Molten Lithium (HCML) blanket with the total thickness of 2.5 m including the shield. With n/n G >1.2, net electric power bigger than 500 MW is possible with β N >4.0 andH>1.2. To access operation space for higher electric power, main restrictions are given by the divertor heat load and the steady-state operation requirements. Developments in both plasma physics and technology are required to handle high heat load and to increase the current drive efficiency. (orig.)

  2. Response Of Lowland Rice To Soil Compaction

    Idawati; Haryanto

    2000-01-01

    Soil compaction, as a new tillage practice for paddy soil, is to substitute pudding in order to reduce land preparation cost. To study response of lowland rice to soil compaction, a pot experiment has been conducted which took place in the greenhouse of P3TIR-BATAN. Soil for experiment was taken from pusakanegara. Two factors (degree of soil compaction and rice variety) were combined. Degree of compaction was split into 3 levels (DI = normal; D215% more compact than normal; 30 % more compact than normal), and rice variety into 2 levels (IR64 and Atomita IV). KH 2 32 PO 4 solution was injected into the soil surrounding rice clump to test the root activity at blooming stage of rice plant. Data resulted from this experiment is presented together with additional data from some other experiments of fertilization in the research s erie to study soil compaction. Some information's from experiment results are as following. Both rice varieties tested gave the same response to soil compaction. Root activity, according to data of 32 P absorbed by plant, was not harmed by soil compaction at the degree tested in the experiment. This prediction is supported by the growth by rice observed at generative growth stage, in pot experiment as well as in field experiment, which showed that soil compaction tested did not decrease rice yield but in opposite in tended to increase the yield. In practising soil compaction in land preparation, fertilizers should be applied by deep placement to have higher increasing is rice yield

  3. Prediction of reservoir compaction and surface subsidence

    De Waal, J.A.; Smits, R.M.M.

    1988-06-01

    A new loading-rate-dependent compaction model for unconsolidated clastic reservoirs is presented that considerably improves the accuracy of predicting reservoir rock compaction and surface subsidence resulting from pressure depletion in oil and gas fields. The model has been developed on the basis of extensive laboratory studies and can be derived from a theory relating compaction to time-dependent intergranular friction. The procedure for calculating reservoir compaction from laboratory measurements with the new model is outlined. Both field and laboratory compaction behaviors appear to be described by one single normalized, nonlinear compaction curve. With the new model, the large discrepancies usually observed between predictions based on linear compaction models and actual (nonlinear) field behavior can be explained.

  4. Diverse Formation Mechanisms for Compact Galaxies

    Kim, Jin-Ah; Paudel, Sanjaya; Yoon, Suk-Jin

    2018-01-01

    Compact, quenched galaxies such as M32 are unusual ones located off the mass - size scaling relation defined by normal galaxies. Still, their formation mechanisms remain unsolved. Here we investigate the evolution of ~100 compact, quenched galaxies at z = 0 identified in the Illustris cosmological simulation. We identify three ways for a galaxy to become a compact one and, often, multiple mechanisms operate in a combined manner. First, stripping is responsible for making about a third of compact galaxies. Stripping removes stars from galaxies, usually while keeping their sizes intact. About one third are galaxies that cease their growth early on after entering into more massive, gigantic halos. Finally, about half of compact galaxies, ~ 35 % of which turn out to undergo stripping, experience the compaction due to the highly centrally concentrated star formation. We discuss the evolutionary path of compact galaxies on the mass – size plane for each mechanism in a broader context of dwarf galaxy formation and evolution.

  5. Implementation of KoHLT-EB DAQ System using compact RIO with EPICS

    Chang, Dae-Sik; Kim, Suk-Kwon; Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    EPICS (Experimental Physics and Industrial Control System) is a collection of software tools collaboratively developed which can be integrated to provide a comprehensive and scalable control system. Currently there is an increase in use of such systems in large Physics experiments like KSTAR, ITER and DAIC (Daejeon Accelerator Ion Complex). The Korean heat load test facility (KoHLT-EB) was installed at KAERI. This facility is utilized for a qualification test of the plasma facing component (PFC) for the ITER first wall and DEMO divertor, and the thermo-hydraulic experiments. The existing data acquisition device was Agilent 34980A multifunction switch and measurement unit and controlled by Agilent VEE. In the present paper, we report the EPICS based newly upgraded KoHLT-EB DAQ system which is the advanced data acquisition system using FPGA-based reconfigurable DAQ devices like compact RIO. The operator interface of KoHLT-EB DAQ system is composed of Control-System Studio (CSS) and another server is able to archive the related data using the standalone archive tool and the archiveviewer can retrieve that data at any time in the infra-network.

  6. Self-compacting concrete (SCC)

    Geiker, Mette Rica

    2008-01-01

    In many aspects Self-Compacting Concrete (SCC, “Self-Consolidating Concrete” in North America) can be considered the concrete of the future. SCC is a family of tailored concretes with special engineered properties in the fresh state. SCC flows into the formwork and around even complicated...... reinforcement arrangements under its own weight. Thus, SCC is not vibrated like conventional concrete. This drastically improves the working environment during construction, the productivity, and potentially improves the homogeneity and quality of the concrete. In addition SCC provides larger architectural...

  7. Portable compact multifunction IR calibrator

    Wyatt, C.L.; Jacobsen, L.; Steed, A.

    1988-01-01

    A compact portable multifunction calibrator designed for future sensor systems is described which enables a linearity calibration for all detectors simultaneously using a near small-area source, a high-resolution mapping of the focal plane with 10 microrad setability and with a blur of less than 100 microrad, system spectral response calibration (radiometer) using a Michelson interferometer source, relative spectral response (spectrometer) using high-temperature external commercial blackbody simulators, and an absolute calibration using an internal low-temperature extended-area source. 5 references

  8. Thermal evolution of compact stars

    Schaab, C.; Glendenning, N.K.

    1996-01-01

    A collection of modern, field-theoretical equations of state is applied to the investigation of cooling properties of compact stars. These comprise neutron stars as well as hypothetical strange-matter stars, made up of absolutely stable 3-flavor strange-quark matter. Various uncertainties in the behavior of matter at supernuclear densities, e.g., hyperonic degrees of freedom, behavior of coupling strengths in matter, pion and meson condensation, superfluidity, transition to quark matter, absolute stability of strange-quark matter, and last but not least the many-body technique itself are tested against the body of observed cooling data. (orig.)

  9. Shock compaction of molybdenum powder

    Ahrens, T. J.; Kostka, D.; Vreeland, T., Jr.; Schwarz, R. B.; Kasiraj, P.

    1983-01-01

    Shock recovery experiments which were carried out in the 9 to 12 GPa range on 1.4 distension Mo and appear adequate to compact to full density ( 45 (SIGMA)m) powders were examined. The stress levels, however, are below those calculated to be from 100 to approx. 22 GPa which a frictional heating model predicts are required to consolidate approx. 10 to 50 (SIGMA)m particles. The model predicts that powders that have a distension of m=1.6 shock pressures of 14 to 72 GPa are required to consolidate Mo powders in the 50 to 10 (SIGMA)m range.

  10. Simplified compact containment BWR plant

    Heki, H.; Nakamaru, M.; Tsutagawa, M.; Hiraiwa, K.; Arai, K.; Hida, T.

    2004-01-01

    The reactor concept considered in this paper has a small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. The Compact Containment Boiling Water Reactor (CCR), which is being developed with matured BWR technologies together with innovative systems/components, is expected to prove attractive in the world energy markets due to its flexibility in regard to both energy demands and site conditions, its high potential for reducing investment risk and its safety features facilitating public acceptance. The flexibility is achieved by CCR's small power output of 300 MWe class and capability of long operating cycle (refueling intervals). CCR is expected to be attractive from view point of investment due to its simplification/innovation in design such as natural circulation core cooling with the bottom located short core, internal upper entry control rod drives (CRDs) with ring-type dryers and simplified ECCS system with high pressure containment concept. The natural circulation core eliminates recirculation pumps and the maintenance of such pumps. The internal upper entry CRDs reduce the height of the reactor vessel (RPV) and consequently reduce the height of the primary containment vessel (PCV). The safety features mainly consist of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), passive auto catalytic recombiner and in-vessel retention (IVR) capability. The large inventory increases the system response time in the case of design-base accidents, including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. The recombiner decreases hydrogen concentration in the PCV in the case of a severe accident. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. The feasibility of CCR safety system has been confirmed by LOCA

  11. Porewater chemistry in compacted bentonite

    Muurinen, A.; Lehikoinen, J. [VTT Chemical Technology, Espoo (Finland)

    1999-03-01

    In this study, the porewater chemistry in compacted bentonite, considered as an engineered barrier in the repository of spent fuel, has been studied in interaction experiments. Many parameters, like the composition and density of bentonite, composition of the solution, bentonite-to-water ratio (B/W), surrounding conditions and experimental time have been varied in the experiments. At the end of the interaction the equilibrating solution, the porewaters squeezed out of the bentonite samples, and bentonites themselves were analyzed to give information for the interpretation and modelling of the interaction. Equilibrium modelling was performed with the HYDRAQL/CE computer code 33 refs.

  12. Compact inertial confinement multireactor concepts

    Pendergrass, J.H.

    1985-01-01

    Inertial confinement fusion (ICF) commercial-applications plant-optimum driver pulse repetition rates may exceed reactor pulse-repetition-rate capabilities. Thus, more than one reactor may be required for low-cost production of electric power, process heat, fissionable fuels, etc., in ICF plants. Substantial savings in expensive reactor containment cells and blankets can be realized by placing more than one reactor in a cell and by surrounding more than one reactor cavity with a single blanket system. There are also some potential disadvantages associated with close coupling in compact multicavity blankets and multireactor cells. Tradeoffs associated with several scenarios have been studied

  13. Main activities in Kazakhstan aimed to substantiate ITER and demo reactors safety

    Shestakov, V.; Chikhray, Y.; Tazhibayeva, I.; Kenzhin, Ye.; Dzhakishev, M.; Goryaev, G.; Gagarin, A.; Shakhvorostov, Yr.; Savchuk, V.

    2004-01-01

    The first stage of such activity is examinations of physicochemical properties of compact beryllium. This work is carrying out Ulba Plant - worl known beryllium producer. Quality control of compact beryllium products includes step-by-step operational control and attestation control of final products for compliancy with customer's requirements. Step-by-step control is carried out along the whole production process and includes the control of the following: temperature, pressure, duration of the process and other process parameter, listed in the in-plant documentation; quality of intermediate semi products (chemical, physical and mechanical properties, defects, appearance, dimension, etc). The process control is carried out by personnel and by an independent inspection service. The attestation control of final products is carried out for compliancy of products with requirement of consumers and includes the following: chemical analysis, mechanical testing, radiographic testing, ultrasonic testing, appearance inspection, dimension inspection, density testing, and metallographic inspection. The attestation control is carried out by a special service independent of technologists. This is the service that makes a final report on the compliancy of the product with requirements of customers and gives permission for shipping the products. The process and attestation control is carried out with the use of equipment, apparatuses and devices, which are checked regularly by special instrumentation service. If they do not meet the requirements in precision, reliability and stability they are removed from service and not approved for measurements. Methods of control of specific values and characteristics, the apparatuses used and allowed classes of accuracy are specified in state standards, tensile specifications of products and in-plant standards or in agreements between a producer and a customer. The next stage will be manufacturing of mock-ups of reactor's first wall elements

  14. Spectrometers for compact neutron sources

    Voigt, J.; Böhm, S.; Dabruck, J. P.; Rücker, U.; Gutberlet, T.; Brückel, T.

    2018-03-01

    We discuss the potential for neutron spectrometers at novel accelerator driven compact neutron sources. Such a High Brilliance Source (HBS) relies on low energy nuclear reactions, which enable cryogenic moderators in very close proximity to the target and neutron optics at comparably short distances from the moderator compared to existing sources. While the first effect aims at increasing the phase space density of a moderator, the second allows the extraction of a large phase space volume, which is typically requested for spectrometer applications. We find that competitive spectrometers can be realized if (a) the neutron production rate can be synchronized with the experiment repetition rate and (b) the emission characteristics of the moderator can be matched to the phase space requirements of the experiment. MCNP simulations for protons or deuterons on a Beryllium target with a suitable target/moderator design yield a source brightness, from which we calculate the sample fluxes by phase space considerations for different types of spectrometers. These match closely the figures of todays spectrometers at medium flux sources. Hence we conclude that compact neutron sources might be a viable option for next generation neutron sources.

  15. Manufacturability of compact synchrotron mirrors

    Douglas, Gary M.

    1997-11-01

    While many of the government funded research communities over the years have put their faith and money into increasingly larger synchrotrons, such as Spring8 in Japan, and the APS in the United States, a viable market appears to exist for smaller scale, research and commercial grade, compact synchrotrons. These smaller, and less expensive machines, provide the research and industrial communities with synchrotron radiation beamline access at a portion of the cost of their larger and more powerful counterparts. A compact synchrotron, such as the Aurora-2D, designed and built by Sumitomo Heavy Industries, Ltd. of japan (SHI), is a small footprint synchrotron capable of sustaining 20 beamlines. Coupled with a Microtron injector, with 150 MeV of injection energy, an entire facility fits within a 27 meter [88.5 ft] square floorplan. The system, controlled by 2 personal computers, is capable of producing 700 MeV electron energy and 300 mA stored current. Recently, an Aurora-2D synchrotron was purchased from SHI by the University of Hiroshima. The Rocketdyne Albuquerque Operations Beamline Optics Group was approached by SHI with a request to supply a group of 16 beamline mirrors for this machine. These mirrors were sufficient to supply 3 beamlines for the Hiroshima machine. This paper will address engineering issues which arose during the design and manufacturing of these mirrors.

  16. Compact Visualisation of Video Summaries

    Janko Ćalić

    2007-01-01

    Full Text Available This paper presents a system for compact and intuitive video summarisation aimed at both high-end professional production environments and small-screen portable devices. To represent large amounts of information in the form of a video key-frame summary, this paper studies the narrative grammar of comics, and using its universal and intuitive rules, lays out visual summaries in an efficient and user-centered way. In addition, the system exploits visual attention modelling and rapid serial visual presentation to generate highly compact summaries on mobile devices. A robust real-time algorithm for key-frame extraction is presented. The system ranks importance of key-frame sizes in the final layout by balancing the dominant visual representability and discovery of unanticipated content utilising a specific cost function and an unsupervised robust spectral clustering technique. A final layout is created using an optimisation algorithm based on dynamic programming. Algorithm efficiency and robustness are demonstrated by comparing the results with a manually labelled ground truth and with optimal panelling solutions.

  17. Prediction for swelling characteristics of compacted bentonite

    Komine, H.; Ogata, N.

    1996-01-01

    Compacted bentonites are attracting greater attention as back-filling (buffer) materials for high-level nuclear waste repositories. For this purpose, it is very important to quantitatively evaluate the swelling characteristics of compacted bentonite. New equations for evaluating the relationship between the swelling deformation of compacted bentonite and the distance between two montmorillonite layers are derived. New equations for evaluating the ion concentration of pore water and the specific surface of bentonite, which significantly influence the swelling characteristics of compacted bentonite, are proposed. Furthermore, a prediction method for the swelling characteristics of compacted bentonite is presented by combining the new equations with the well-known theoretical equations of repulsive and attractive forces between two montmorillonite layers. The applicability of this method was investigated by comparing the predicted results with laboratory test results on the swelling deformation and swelling pressure of compacted bentonites. (author) 31 refs., 8 tabs., 12 figs

  18. Aplicación móvil para la visualización y ejecución de demos en IPOL

    Ramírez Ravelo, Miguel Isaías

    2014-01-01

    [ES] IPOL es una revista científica de procesamiento digital de imágenes y diversos métodos de análisis de imágenes. En cada publicación se incorpora una demo donde cualquier persona puede probar, vía web, el funcionamiento del método descrito en dicha publicación. De esta forma, se puede usar el método sin tener conocimiento de programación ni tener que instalarlo en su ordenador. En este proyecto fin de carrera se quiere desarrollar una aplicación que permita la ejecución de las demos desde...

  19. Combination of helical ferritic-steel inserts and flux-tube-expansion divertor for the heat control in tokamak DEMO reactor

    Takizuka, T.; Tokunaga, S.; Hoshino, K.; Shimizu, K.; Asakura, N.

    2015-01-01

    Edge localized modes (ELMs) in the H-mode operation of tokamak reactors may be suppressed/mitigated by the resonant magnetic perturbation (RMP), but RMP coils are considered incompatible with DEMO reactors under the strong neutron flux. We propose an innovative concept of the RMP without installing coils but inserting ferritic steels of the helical configuration. Helically perturbed field is naturally formed in the axisymmetric toroidal field through the helical ferritic steel inserts (FSIs). When ELMs are avoided, large stationary heat load on divertor plates can be reduced by adopting a flux-tube-expansion (FTE) divertor like an X divertor. Separatrix shape and divertor-plate inclination are similar to those of a simple long-leg divertor configuration. Combination of the helical FSIs and the FTE divertor is a suitable method for the heat control to avoid transient ELM heat pulse and to reduce stationary divertor heat load in a tokamak DEMO reactor

  20. Design concept of conducting shell and in-vessel components suitable for plasma vertical stability and remote maintenance scheme in DEMO reactor

    Utoh, Hiroyasu, E-mail: uto.hiroyasu@jaea.go.jp [Japan Atomic Energy Agency, Obuchi, Rokkasho-mura, Aomori-ken 039-3212 (Japan); International Fusion Energy Research Centre, 2-166, Obuchi, Rokkasho, Aomori 039-3212 (Japan); Takase, Haruhiko [Japan Atomic Energy Agency, Obuchi, Rokkasho-mura, Aomori-ken 039-3212 (Japan); International Fusion Energy Research Centre, 2-166, Obuchi, Rokkasho, Aomori 039-3212 (Japan); Sakamoto, Yoshiteru; Tobita, Kenji [Japan Atomic Energy Agency, Obuchi, Rokkasho-mura, Aomori-ken 039-3212 (Japan); Mori, Kazuo; Kudo, Tatsuya [Japan Atomic Energy Agency, Obuchi, Rokkasho-mura, Aomori-ken 039-3212 (Japan); International Fusion Energy Research Centre, 2-166, Obuchi, Rokkasho, Aomori 039-3212 (Japan); Someya, Youji; Asakura, Nobuyuki; Hoshino, Kazuo; Nakamura, Makoto; Tokunaga, Shinsuke [Japan Atomic Energy Agency, Obuchi, Rokkasho-mura, Aomori-ken 039-3212 (Japan)

    2016-02-15

    Highlights: • Conceptual design of in-vessel component including conducting shell has been investigated. • The conducting shell design for plasma vertical stability was clarified from the plasma vertical stability analysis. • The calculation results showed that the double-loop shell has the most effect on plasma vertical stability. - Abstract: In order to realize a feasible DEMO, we designed an in-vessel component including the conducting shell. The project is affiliated with the broader approach DEMO design activities and is conceptualized from a plasma vertical stability and engineering viewpoint. The dependence of the plasma vertical stability on the conducing shell parameters and the electromagnetic force at plasma disruption were investigated in numerical simulations (programmed in the 3D eddy current analysis code and a plasma position control code). The simulations assumed the actual shape and position of the vacuum vessel and in-vessel components. The plasma vertical stability was most effectively maintained by the double-loop shell.

  1. Comparative study of the tungsten irradiation conditions in IFMIF and DEMO

    Simakov, S.P.; Pereslavtsev, P.; Fischer, U. [Karlsruhe Institute of Technology, Eggenstein-Leopoldshafen (Germany). Inst. for Neutron Physics and Reactor Technology; Moeslang, A. [Karlsruhe Institute of Technology, Eggenstein-Leopoldshafen (Germany). Inst. for Material Research I

    2010-05-15

    The International Fusion Material Irradiation Facility (IFMIF) [1] will provide an accelerator based intense neutron source with a white spectrum extending up to 55 MeV for high fluence irradiations of fusion power reactor (FPR) candidate materials. Material samples located in test modules will be subjected to a radiation load anticipated for a fusion power reactor. The highest neutron flux is expected in the High Flux Test Module, which is considered in the IFMIF design to host around 1000 compactly packed stainless steel samples - the main structure materials of power fusion reactors. Another material subjected to the highest loads in a FPR is a tungsten. It is planned to be used as armour tiles for the divertor or the first wall. It turned out that no specific effort has been undertaken so far to search for a suitable irradiation location in the IFMIF Test Cell which provides a reasonable representation of the irradiation conditions in the divertor of a fusion power reactors. (orig.)

  2. The Future Concrete: Self-Compacting Concrete

    Liana Iureş

    2010-01-01

    Full Text Available The paper presents the characteristics of the self-compacting concretes, their advantages and disadvantages when they are used in buildings. Due to its properties and composition, the self-compacting concrete is described here as being one of the future friendly enviromental material for buildings. Tests concerning to obtaining a self-compacting concrete, together with the specific fresh concrete properties tests, are described.

  3. Starbursts in Blue compact dwarf galaxies

    Thuan, T.X.

    1987-01-01

    We summarize all the arguments for a bursting mode of star formation in blue compact dwarf galaxies. We show in particular how spectral synthesis of far ultraviolet spectra of Blue compact dwarf galaxy constitutes a powerful way for studying the star formation history in these galaxies. Blue compact dwarf galaxy luminosity functions show jumps and discontinuities. These jumps act like fossil records of the star-forming bursts, helping us to count and date the bursts

  4. Effect of slash on forwarder soil compaction

    Timothy P. McDonald; Fernando Seixas

    1997-01-01

    A study of the effect of slash on forwarder soil compaction was carried out. The level of soil compaction at two soil moisture contents, three slash densities (0, 10, and 20 kg/m2), and two levels of traffic (one and five passes) were measured. Results indicated that, on dry, loamy sand soils, the presence of slash did not decrease soil compaction after one forwarder...

  5. The Future Concrete: Self-Compacting Concrete

    Iureş, Liana; Bob, Corneliu

    2010-01-01

    The paper presents the characteristics of the self-compacting concretes, their advantages and disadvantages when they are used in buildings. Due to its properties and composition, the self-compacting concrete is described here as being one of the future friendly enviromental material for buildings. Tests concerning to obtaining a self-compacting concrete, together with the specific fresh concrete properties tests, are described.

  6. Compact approach to fusion power reactors

    Hagenson, R.L.; Krakowski, R.A.; Bathke, C.G.; Miller, R.L.

    1984-01-01

    The potential of the Reversed-Field Pinch (RFP) for development into an efficient, compact, copper-coil fusion reactor has been quantified by comprehensive parametric tradeoff studies. These compact systems promise to be competitive in size, power density, and cost to alternative energy sources. Conceptual engineering designs that largely substantiate these promising results have since been completed. This 1000-MWe(net) design is described along with a detailed rationale and physics/technology assessment for the compact approach to fusion

  7. Multicriteria selection in concept design of a divertor remote maintenance port in the EU DEMO reactor using an AHP participative approach

    Carfora, D. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Gironimo, G. Di, E-mail: giuseppe.digironimo@unina.it [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Esposito, G. [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Huhtala, K. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Määttä, T.; Mäkinen, H. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Miccichè, G. [ENEA Brasimone, I:40032 Camugnano (Italy); Mozzillo, R. [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy)

    2016-11-15

    Highlights: • Concept Studies in Divertor Remote Handling. • Prioritization of concept alternatives. • Comparison and evaluation of product alternatives using AHP. - Abstract: The work behind this paper took place in the Eurofusion remote maintenance system project (WPRM) for the EU Demonstration Fusion Power Reactor (DEMO). Following ITER, the aim of DEMO is to demonstrate the capability of generating several hundreds of MW of net electricity by 2050. The main objective of this paper was the study of the most efficient design of the maintenance port for replacing the divertor cassettes in a Remote Handling (RH) point of view. In DEMO overall design, one important consideration is the availability and short down time operations. The inclination of the divertor port has a very important impact on all the RH tasks such as the design of the divertor mover, the divertor locking systems and the end effectors. The current reference scenario of the EU DEMO foresees a 45° inclined port for the remote maintenance (RM) of the divertor in the lower part of the reactor. Nevertheless, in the optic of the systems engineering (SE) approach, in early concept design phase, all possible configurations shall be taken into account. Even the solutions which seem not feasible at all need to be investigated, because they could lead to new and innovative engineering proposals. The different solutions were compared using an approach based on the Analytic Hierarchy Process (AHP). The technique is a multi-criteria decision making approach in which the factors that are important in making a decision are arranged in a hierarchic structure. The results of these studies show how the application of the AHP improved and focused the selection on the concept which is closer to the requirements arose from technical meetings with the experts of the RH field.

  8. Multicriteria selection in concept design of a divertor remote maintenance port in the EU DEMO reactor using an AHP participative approach

    Carfora, D.; Gironimo, G. Di; Esposito, G.; Huhtala, K.; Määttä, T.; Mäkinen, H.; Miccichè, G.; Mozzillo, R.

    2016-01-01

    Highlights: • Concept Studies in Divertor Remote Handling. • Prioritization of concept alternatives. • Comparison and evaluation of product alternatives using AHP. - Abstract: The work behind this paper took place in the Eurofusion remote maintenance system project (WPRM) for the EU Demonstration Fusion Power Reactor (DEMO). Following ITER, the aim of DEMO is to demonstrate the capability of generating several hundreds of MW of net electricity by 2050. The main objective of this paper was the study of the most efficient design of the maintenance port for replacing the divertor cassettes in a Remote Handling (RH) point of view. In DEMO overall design, one important consideration is the availability and short down time operations. The inclination of the divertor port has a very important impact on all the RH tasks such as the design of the divertor mover, the divertor locking systems and the end effectors. The current reference scenario of the EU DEMO foresees a 45° inclined port for the remote maintenance (RM) of the divertor in the lower part of the reactor. Nevertheless, in the optic of the systems engineering (SE) approach, in early concept design phase, all possible configurations shall be taken into account. Even the solutions which seem not feasible at all need to be investigated, because they could lead to new and innovative engineering proposals. The different solutions were compared using an approach based on the Analytic Hierarchy Process (AHP). The technique is a multi-criteria decision making approach in which the factors that are important in making a decision are arranged in a hierarchic structure. The results of these studies show how the application of the AHP improved and focused the selection on the concept which is closer to the requirements arose from technical meetings with the experts of the RH field.

  9. European DEMO BOT Solid Breeder Blanket: the concept based on the use of cooling plates and beds of beryllium and Li4SiO4 pebbles

    Dalle Donne, M.; Fischer, U.; Norajitra, P.; Reimann, G.; Reiser, H.

    1995-01-01

    The paper presents an important modification of the European DEMO BOT Solid Breeder Blanket. The new design uses cooling plates rather than tubes. This allows a considerable simplification of the blanket and the separation of the beryllium from the Li 4 SiO 4 pebbles. The neutronic, thermohydraulic and tritium performance of the new design is quite good and equivalent to that of the previous one. (orig.)

  10. The DEMO trial: a randomized, parallel-group, observer-blinded clinical trial of strength versus aerobic versus relaxation training for patients with mild to moderate depression

    Krogh, Jesper; Saltin, Bengt; Gluud, Christian

    2009-01-01

    OBJECTIVE: To assess the benefit and harm of exercise training in adults with clinical depression. METHOD: The DEMO trial is a randomized pragmatic trial for patients with unipolar depression conducted from January 2005 through July 2007. Patients were referred from general practitioners or psych......: Our findings do not support a biologically mediated effect of exercise on symptom severity in depressed patients, but they do support a beneficial effect of strength training on work capacity. TRIAL REGISTRATION: (ClinicalTrials.gov) Identifier: NCT00103415....

  11. Performance of tungsten-based materials and components under ITER and DEMO relevant steady-state thermal loads

    Ritz, Guillaume Henri

    2011-07-01

    In nuclear fusion devices the surfaces directly facing the plasma are irradiated with high energy fluxes. The most intense loads are deposited on the divertor located at the bottom of the plasma chamber, which has to withstand continuous heat loads with a power density of several MW . m{sup -2} as well as transient events. These are much shorter (in the millisecond and sub-millisecond regime) but deposit a higher power densities of a few GW . m{sup -2}. The search for materials that can survive to those severe loading conditions led to the choice of tungsten which possesses advantageous attributes such as a high melting point, high thermal conductivity, low thermal expansion and an acceptable activation rate. These properties made it an attractive and promising candidate as armor material for divertors of future fusion devices such as ITER and DEMO. For the DEMO divertor, conceptual studies on helium-cooled tungsten plasma-facing components were performed. The concept was realized and tested under DEMO specific cyclic thermal loads. The examination of the plasma-facing components by microstructural analyses before and after thermal loading enabled to determine the mechanisms for components failure. Among others, it clearly showed the impact of the tungsten grade and the thermal stress induced crack formation on the performance of the armor material and in general of the plasma-facing component under high heat loads. A tungsten qualification program was launched to study the behaviour of various tungsten grades, in particular the crack formation, under fusion relevant steady-state thermal loads. In total, seven commercially available materials from two industrial suppliers were investigated. As the material's thermal response is strongly related to its microstructure, this program comprised different material geometries and manufacturing technologies. It also included the utilization of an actively cooled specimen holder which has been designed to perform

  12. Performance of tungsten-based materials and components under ITER and DEMO relevant steady-state thermal loads

    Ritz, Guillaume Henri

    2011-01-01

    In nuclear fusion devices the surfaces directly facing the plasma are irradiated with high energy fluxes. The most intense loads are deposited on the divertor located at the bottom of the plasma chamber, which has to withstand continuous heat loads with a power density of several MW . m -2 as well as transient events. These are much shorter (in the millisecond and sub-millisecond regime) but deposit a higher power densities of a few GW . m -2 . The search for materials that can survive to those severe loading conditions led to the choice of tungsten which possesses advantageous attributes such as a high melting point, high thermal conductivity, low thermal expansion and an acceptable activation rate. These properties made it an attractive and promising candidate as armor material for divertors of future fusion devices such as ITER and DEMO. For the DEMO divertor, conceptual studies on helium-cooled tungsten plasma-facing components were performed. The concept was realized and tested under DEMO specific cyclic thermal loads. The examination of the plasma-facing components by microstructural analyses before and after thermal loading enabled to determine the mechanisms for components failure. Among others, it clearly showed the impact of the tungsten grade and the thermal stress induced crack formation on the performance of the armor material and in general of the plasma-facing component under high heat loads. A tungsten qualification program was launched to study the behaviour of various tungsten grades, in particular the crack formation, under fusion relevant steady-state thermal loads. In total, seven commercially available materials from two industrial suppliers were investigated. As the material's thermal response is strongly related to its microstructure, this program comprised different material geometries and manufacturing technologies. It also included the utilization of an actively cooled specimen holder which has been designed to perform sophisticated

  13. Soil compaction and growth of woody plants

    Kozlowski, T.T. [Univ. of California, Berkeley (United States). Dept. of Environmental Science, Policy and Management

    1999-07-01

    Although soil compaction in the field may benefit or inhibit the growth of plants, the harmful effects are much more common. This paper emphasizes the deleterious effects of predominantly high levels of soil compaction on plant growth and yield. High levels of soil compaction are common in heavily used recreation areas, construction sites, urban areas, timber harvesting sites, fruit orchards, agroforestry systems and tree nurseries. Compaction can occur naturally by settling or slumping of soil or may be induced by tillage tools, heavy machinery, pedestrian traffic, trampling by animals and fire. Compaction typically alters soil structure and hydrology by increasing soil bulk density; breaking down soil aggregates; decreasing soil porosity, aeration and infiltration capacity; and by increasing soil strength, water runoff and soil erosion. Appreciable compaction of soil leads to physiological dysfunctions in plants. Often, but not always, reduced water absorption and leaf water deficits develop. Soil compaction also induces changes in the amounts and balances of growth hormones in plants, especially increases in abscisic acid and ethylene. Absorption of the major mineral nutrients is reduced by compaction of both surface soils and subsoils. The rate of photosynthesis of plants growing in very compacted soil is decreased by both stomatal and non-stomatal inhibition. Total photosynthesis is reduced as a result of smaller leaf areas. As soils become increasingly compacted respiration of roots shifts toward an anaerobic state. Severe soil compaction adversely influences regeneration of forest stands by inhibiting seed germination and growth of seedlings, and by inducing seedling mortality. Growth of woody plants beyond the seedling stage and yields of harvestable plant products also are greatly decreased by soil compaction because of the combined effects of high soil strength, decreased infiltration of water and poor soil aeration, all of which lead to a decreased

  14. Peculiarities of powder brittle media compaction

    Perel'nam, V.E.; Aristarkhov, A.I.

    1981-01-01

    The paper is concerned with theoretical and practical aspects of the compaction process for powders of almost unstrained materials. Consideration from the standpoint of compressible body strain mechanics shows that such porous media may have a certain ''threshold'' density. Ductile characteristics of the porous material compacted up to this extent are identical with properties of compacrat bodies, i.e. there is a theoretically substantiated ban on a possibility of their further compaction without changing the state of the powder particle material. Theoretical conclusions are confirmed by results of experimental studies in compaction of titanium- containing ceramics [ru

  15. Soil compaction and growth of woody plants

    Kozlowski, T.T.

    1999-01-01

    Although soil compaction in the field may benefit or inhibit the growth of plants, the harmful effects are much more common. This paper emphasizes the deleterious effects of predominantly high levels of soil compaction on plant growth and yield. High levels of soil compaction are common in heavily used recreation areas, construction sites, urban areas, timber harvesting sites, fruit orchards, agroforestry systems and tree nurseries. Compaction can occur naturally by settling or slumping of soil or may be induced by tillage tools, heavy machinery, pedestrian traffic, trampling by animals and fire. Compaction typically alters soil structure and hydrology by increasing soil bulk density; breaking down soil aggregates; decreasing soil porosity, aeration and infiltration capacity; and by increasing soil strength, water runoff and soil erosion. Appreciable compaction of soil leads to physiological dysfunctions in plants. Often, but not always, reduced water absorption and leaf water deficits develop. Soil compaction also induces changes in the amounts and balances of growth hormones in plants, especially increases in abscisic acid and ethylene. Absorption of the major mineral nutrients is reduced by compaction of both surface soils and subsoils. The rate of photosynthesis of plants growing in very compacted soil is decreased by both stomatal and non-stomatal inhibition. Total photosynthesis is reduced as a result of smaller leaf areas. As soils become increasingly compacted respiration of roots shifts toward an anaerobic state. Severe soil compaction adversely influences regeneration of forest stands by inhibiting seed germination and growth of seedlings, and by inducing seedling mortality. Growth of woody plants beyond the seedling stage and yields of harvestable plant products also are greatly decreased by soil compaction because of the combined effects of high soil strength, decreased infiltration of water and poor soil aeration, all of which lead to a decreased

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

    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.

  17. On the optimization of the first wall of the DEMO water-cooled lithium lead outboard breeding blanket equatorial module

    Di Maio, P.A., E-mail: pietroalessandro.dimaio@unipa.it; Arena, P.; Bongiovì, G.; Chiovaro, P.; Forte, R.; Garitta, S.

    2016-11-01

    Highlights: • The geometric optimization of the DEMO WCLL blanket module first wall has been performed, maximizing the heat flux it may safely undergo. • Attention has been focused on the FW flat concept endowed with square cooling channels. • A theoretical-computational approach based on the finite element method (FEM) has been followed, adopting a qualified commercial FEM code. • Four optimized FW configurations have been found to safely withstand a heat flux up to 2 MW/m{sup 2} fulfilling all the rules prescribed by safety codes. - Abstract: Within the framework of EUROfusion R&D activities a research campaign has been carried out at the University of Palermo in order to investigate the thermo-mechanical performances of the DEMO water-cooled lithium lead (WCLL) breeding blanket first wall (FW). The research campaign has been mainly focused on the optimization of the FW geometric configuration in order to maximize the heat flux it may safely withstand fulfilling all the thermal, hydraulic and mechanical requirements foreseen by safety codes. Attention has been focused on the FW flat concept endowed with square cooling channels and the potential influence of its four main geometrical parameters on its thermo-mechanical performances has been assessed performing a parametric analysis by means of a qualified commercial finite element method code. A set of 5929 different FW geometric configurations has been considered and the thermal performances of each one of them have been numerically assessed in case it undergoes 26 different values of heat flux on its plasma-facing surface. The resulting 154154 thermal analyses have allowed to select those cases fulfilling the adopted thermal-hydraulic requirements, whose thermo-mechanical performances have been numerically assessed under both normal operation and over-pressurization steady state loading scenarios to check whether they met the mechanical requirements prescribed by the pertaining SDC-IC safety rules. Four

  18. Multifarious Physics Analyses of the Core Plasma Properties in a Helical DEMO Reactor FFHR-d1

    Miyazawa, J.; Satake, S.; Goto, T.; Seki, R.; Nunami, M.; Funaba, H.; Yamada, I.; Suzuki, C.; Sakamoto, R.; Motojima, G.; Yamada, H.; Sagara, A., E-mail: miyazawa@lhd.nifs.ac.jp [National Institute for Fusion Science, Toki (Japan); Yokoyama, M.; Suzuki, Y.; Masaoka, Y.; Murakami, S. [Departement Nuclear Engineering, Kyoto University, Kyoto (Japan)

    2012-09-15

    Full text: Theoretical analyses on the MHD equilibrium, the neoclassical transport, and the alpha particle transport, etc., are being carried out for a helical fusion DEMO reactor named FFHR- d1, using radial profiles extrapolated from LHD. FFHR-d1 is a heliotron type DEMO reactor of which the conceptual design activity has been launched since 2010. It is possible to sustain the burning plasma without auxiliary heating (i.e., self-ignition) in FFHR-d1, since there is no need of plasma current drive in heliotron plasmas. The device size is 4 times enlarged from LHD, i.e., the major radius of helical coil center is 15.6 m, the magnetic field strength at the helical coil center is 4.7 T, and the fusion output is {approx} 3 GW. One of the distinguished subjects in FFHR-d1 compared with the former FFHR design series is the robust similarity with LHD. The arrangement of superconducting magnet coils in FFHR-d1 is similar to that of LHD, except a pair of planar poloidal coils omitted to maximize the maintenance ports. This makes reasonable to assume a similar MHD equilibrium as observed in LHD for FFHR-d1, as long as the beta profiles in these two are similar. In FFHR-d1, radial profiles of density and temperature are determined by multiplying proper enhancement factors on those obtained in LHD, according to the DPE (Direct Profile Extrapolation) method. The enhancement factors are calculated consistently with the gyro-Bohm model. Therefore, the global confinement properties as expressed in ISS95 or ISS04 are kept in FFHR-d1. A large Shafranov shift is foreseen in FFHR-d1 due to its high-beta property. This leads to deterioration in the neoclassical transport and alpha particle confinement. Effectiveness of plasma position control and/or magnetic configuration optimization has been examined to solve this problem and to check the validity of extrapolated profiles. According to these analyses, it is concluded that the self-ignition condition can be achieved in FFHR-d1 by

  19. DemoMinga: O compromisso social da extensão na área da saúde

    César Augusto Radice Oviedo

    2016-10-01

    Full Text Available A FACISA-UNE, no quadro da sua política de "contribuir para a melhoria da qualidade de vida da comunidade circundante", implementa o Projeto DemoMinga: Primeira área de demonstração nacional no Paraguai, no município de Minga Guaçu, Alto Paraná. No DemoMinga é executado o "Programa Integrado de Intervenção de base Comunitária para o Desenvolvimento Integral". Objetivo: Melhorar o estado de saúde e qualidade de vida dos moradores do bairro Norma Luisa do município de Minga Guaçu, para reduzir custos e aumentar benefícios em saúde pública. Metodologia: Constituída por duas fases: a de investigação, para a construção de uma linha de base, e b planejamento de estratégias de intervenção em diferentes eixos temáticos. Resultados: são realizados cursos de curta duração de cozinha saudável e atividade física, consultas médicas e odontológicas, atividades de promoção e prevenção, em local próprio. No âmbito da promoção, foi assinado um convênio com a empresa "Chiperia Leticia S.A." para produzir e comercializar a "chipa saudável", sem gorduras trans. Participação: nos cursos participam, em média, 25 pessoas. Desde setembro de 2013 foram realizados 24 cursos; em 2014, 96 e, de fevereiro a junho de 2015, foram realizados 48. Conclusões: O programa constitui uma área de desenvolvimento multi e interdisciplinar, interinstitucional que permite a articulação da docência, investigação e extensão universitária, no quadro da sua responsabilidade social.

  20. On the optimization of the first wall of the DEMO water-cooled lithium lead outboard breeding blanket equatorial module

    Di Maio, P.A.; Arena, P.; Bongiovì, G.; Chiovaro, P.; Forte, R.; Garitta, S.

    2016-01-01

    Highlights: • The geometric optimization of the DEMO WCLL blanket module first wall has been performed, maximizing the heat flux it may safely undergo. • Attention has been focused on the FW flat concept endowed with square cooling channels. • A theoretical-computational approach based on the finite element method (FEM) has been followed, adopting a qualified commercial FEM code. • Four optimized FW configurations have been found to safely withstand a heat flux up to 2 MW/m"2 fulfilling all the rules prescribed by safety codes. - Abstract: Within the framework of EUROfusion R&D activities a research campaign has been carried out at the University of Palermo in order to investigate the thermo-mechanical performances of the DEMO water-cooled lithium lead (WCLL) breeding blanket first wall (FW). The research campaign has been mainly focused on the optimization of the FW geometric configuration in order to maximize the heat flux it may safely withstand fulfilling all the thermal, hydraulic and mechanical requirements foreseen by safety codes. Attention has been focused on the FW flat concept endowed with square cooling channels and the potential influence of its four main geometrical parameters on its thermo-mechanical performances has been assessed performing a parametric analysis by means of a qualified commercial finite element method code. A set of 5929 different FW geometric configurations has been considered and the thermal performances of each one of them have been numerically assessed in case it undergoes 26 different values of heat flux on its plasma-facing surface. The resulting 154154 thermal analyses have allowed to select those cases fulfilling the adopted thermal-hydraulic requirements, whose thermo-mechanical performances have been numerically assessed under both normal operation and over-pressurization steady state loading scenarios to check whether they met the mechanical requirements prescribed by the pertaining SDC-IC safety rules. Four

  1. (U) Influence of Compaction Model Form on Planar and Cylindrical Compaction Geometries

    Fredenburg, David A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Carney, Theodore Clayton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fichtl, Christopher Allen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Ramsey, Scott D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2018-01-05

    The dynamic compaction response of CeO2 is examined within the frameworks of the Ramp and P-a compaction models. Hydrocode calculations simulating the dynamic response of CeO2 at several distinct pressures within the compaction region are investigated in both planar and cylindrically convergent geometries. Findings suggest additional validation of the compaction models is warranted under complex loading configurations.

  2. Acceleration of a compact torus

    Hartmann, C.W.; Eddleman, J.L.; Hammer, J.H.; Kusse, B.

    1987-01-01

    The authors report the first results of a study of acceleration of spheromak-type compact toruses in the RACE experiment (plasma Ring ACceleration Experiment). The RACE apparatus consists of (1) a magnetized, coaxial plasma gun 50 cm long, 35 cm OD, 20 cm ID, (2) 600 cm long coaxial acceleration electrodes 50 cm OD, 20 cm ID, (3) a 250 kJ electrolytic capacitor bank to drive the gun solenoid for initial magnetization, (4) a 200 kJ gun bank, (5) a 260 kJ accelerator bank, and (6) magnetic probes and other diagnostics, and vacuum apparatus. To outer acceleration electrode is an extension, at larger OD, of the gun outer electrode, and the inner acceleration electrode is supported and fed by a coaxial insert in the gun center electrode as shown

  3. Experimental studies of compact toroids

    1991-01-01

    The Berkeley Compact Toroid Experiment (BCTX) device is a plasma device with a Marshall-gun generated, low aspect ratio toroidal plasma. The device is capable of producing spheromak-type discharges and may, with some modification, produce low-aspect ratio tokamak configurations. A unique aspect of this experimenal devie is its large lower hybrid (LH) heating system, which consists of two 450MHz klystron tubes generating 20 megawatts each into a brambilla-type launching structure. Successful operation with one klystron at virtually full power (18 MW) has been accomplished with 110 μs pulse length. A second klystron is currently installed in its socket and magnet but has not been added to the RF drive system. This report describes current activities and accomplishments and describes the anticipated results of next year's activity

  4. Quasistatic evolution of compact toroids

    Sgro, A.G.; Spencer, R.L.; Lilliequist, C.

    1981-01-01

    Some results are presented of simulations of the post formation evolution of compact toroids. The simulations were performed with a 1-1/2 D transport code. Such a code makes explicit use of the fact that the shapes of the flux surfaces in the plasma change much more slowly than do the profiles of the physical variables across the flux surfaces. Consequently, assuming that the thermodynamic variables are always equilibrated on a flux surface, one may calculate the time evolution of these profiles as a function of a single variable that labels the flux surfaces. Occasionally, during the calculation these profiles are used to invert the equilibrium equation to update the shapes of the flux surfaces. In turn, these shapes imply certain geometric cofficients, such as A = 2 >, which contain the geometric information required by the 1-D equations

  5. Compact RFID Enabled Moisture Sensor

    U. H. Khan

    2016-09-01

    Full Text Available This research proposes a novel, low-cost RFID tag sensor antenna implemented using commercially available Kodak photo-paper. The aim of this paper is to investigate the possibility of stable, RFID centric communication under varying moisture levels. Variation in the frequency response of the RFID tag in presence of moisture is used to detect different moisture levels. Combination of unique jaw shaped contours and T-matching network is used for impedance matching which results in compact size and minimal ink consumption. Proposed tag is 1.4 × 9.4 cm2 in size and shows optimum results for various moisture levels upto 45% in FCC band with a bore sight read range of 12.1 m.

  6. COMPACTION STUDIES OF TORREFIED WILLOW

    Michał Rejdak

    2017-01-01

    Full Text Available The article presents the results of studies of torrefied willow (Salix viminalis L. compaction. Densification tests were performed using a hydraulic press with a maximum pressure of 216 MPa. The effect of basic parameters of the briquetting process (pressure and temperature on mechanical parameters of manufactured briquettes were determined. On the basis of the research, it was found that the increase in pressure and temperature of the densification process increases the density and strength of pressed briquettes. The positive effect of temperature is particularly noticeable at lower pressing pressures (36 MPa – 72 MPa. In the case of a temperature of 300 °C, the increase in a pressure from 144 MPa to 216 MPa resulted in the decrease in the density and strength of the briquette. It was also found that the briquettes manufactured at this temperature are characterized by lower density and strength than the briquettes obtained at a temperature of 200 °C.

  7. A compact mobile neutron generator

    Zhou Changgeng; Li Yan; Hu Yonghong; Lou Benchao; Wu Chunlei

    2007-06-01

    Through fitting the high voltage terminal from introducing overseas and pulse system et al. from oneself developing together, a compact mobile neutron generator is established. The length and weight of this neutron generator are 2 500 mm and less than 1 t, respectively. It can be expediently moved to the location which is required by experimental people. It is consisted of RF ion source, acceleration tube, high voltage generator, focus device, microsecond pulse system, gas leak system, control system, vacuum system and experimental target. It can produce 150 μA continuous deuterium ion beam current, also can produce the pulse deuterium ion beam current. The pulse widths are 10-100 μs and frequencies 10 Hz, 1 000 Hz, 10 000 Hz. The D-T neutron yields of the neutron generator may arrive 1.5 x 10 10 s -1 . The working principle and the structure of the main parts of this neutron generator are described. (authors)

  8. Compact torus compression of microwaves

    Hewett, D.W.; Langdon, A.B.

    1985-01-01

    The possibility that a compact torus (CT) might be accelerated to large velocities has been suggested by Hartman and Hammer. If this is feasible one application of these moving CTs might be to compress microwaves. The proposed mechanism is that a coaxial vacuum region in front of a CT is prefilled with a number of normal electromagnetic modes on which the CT impinges. A crucial assumption of this proposal is that the CT excludes the microwaves and therefore compresses them. Should the microwaves penetrate the CT, compression efficiency is diminished and significant CT heating results. MFE applications in the same parameters regime have found electromagnetic radiation capable of penetrating, heating, and driving currents. We report here a cursory investigation of rf penetration using a 1-D version of a direct implicit PIC code

  9. Anisotropic models for compact stars

    Maurya, S.K.; Dayanandan, Baiju [University of Nizwa, Department of Mathematical and Physical Sciences, College of Arts and Science, Nizwa (Oman); Gupta, Y.K. [Jaypee Institute of Information Technology University, Department of Mathematics, Noida, Uttar Pradesh (India); Ray, Saibal [Government College of Engineering and Ceramic Technology, Department of Physics, Kolkata, West Bengal (India)

    2015-05-15

    In the present paper we obtain an anisotropic analog of the Durgapal and Fuloria (Gen Relativ Gravit 17:671, 1985) perfect fluid solution. The methodology consists of contraction of the anisotropic factor Δ with the help of both metric potentials e{sup ν} and e{sup λ}. Here we consider e{sup λ} the same as Durgapal and Fuloria (Gen Relativ Gravit 17:671, 1985) did, whereas e{sup ν} is as given by Lake (Phys Rev D 67:104015, 2003). The field equations are solved by the change of dependent variable method. The solutions set mathematically thus obtained are compared with the physical properties of some of the compact stars, strange star as well as white dwarf. It is observed that all the expected physical features are available related to the stellar fluid distribution, which clearly indicates the validity of the model. (orig.)

  10. Compact oleic acid in HAMLET.

    Fast, Jonas; Mossberg, Ann-Kristin; Nilsson, Hanna; Svanborg, Catharina; Akke, Mikael; Linse, Sara

    2005-11-07

    HAMLET (human alpha-lactalbumin made lethal to tumor cells) is a complex between alpha-lactalbumin and oleic acid that induces apoptosis in tumor cells, but not in healthy cells. Heteronuclear nuclear magnetic resonance (NMR) spectroscopy was used to determine the structure of 13C-oleic acid in HAMLET, and to study the 15N-labeled protein. Nuclear Overhauser enhancement spectroscopy shows that the two ends of the fatty acid are in close proximity and close to the double bond, indicating that the oleic acid is bound to HAMLET in a compact conformation. The data further show that HAMLET is a partly unfolded/molten globule-like complex under physiological conditions.

  11. Studies of accelerated compact toruses

    Hartman, C.W.; Eddleman, J.; Hammer, J.H.

    1983-01-01

    In an earlier publication we considered acceleration of plasma rings (Compact Torus). Several possible accelerator configurations were suggested and the possibility of focusing the accelerated rings was discussed. In this paper we consider one scheme, acceleration of a ring between coaxial electrodes by a B/sub theta/ field as in a coaxial rail-gun. If the electrodes are conical, a ring accelerated towards the apex of the cone undergoes self-similar compression (focusing) during acceleration. Because the allowable acceleration force, F/sub a/ = kappaU/sub m//R where (kappa - 2 , the accelerating distance for conical electrodes is considerably shortened over that required for coaxial electrodes. In either case, however, since the accelerating flux can expand as the ring moves, most of the accelerating field energy can be converted into kinetic energy of the ring leading to high efficiency

  12. Magnetohydodynamics stability of compact stellarators

    Fu, G.Y.; Ku, L.P.; Cooper, W.A.; Hirshman, S.H.

    2000-01-01

    Recent stability results of external kink modes and vertical modes in compact stellarators are presented. The vertical mode is found to be stabilized by externally generated poloidal flux. A simple stability criterion is derived in the limit of large aspect ratio and constant current density. For a wall at infinite distance from the plasma, the amount of external flux needed for stabilization is given by Fi = (k2 minus k)=(k2 + 1), where k is the axisymmetric elongation and Fi is the fraction of the external rotational transform. A systematic parameter study shows that the external kink mode in QAS can be stabilized at high beta (approximately 5%) without a conducting wall by magnetic shear via 3D shaping. It is found that external kinks are driven by both parallel current and pressure gradient. The pressure contributes significantly to the overall drive through the curvature term and the Pfirsch-Schluter current

  13. A Compact UWB Diversity Antenna

    Hui Zhao

    2014-01-01

    Full Text Available A compact printed ultrawideband (UWB diversity antenna with a size of 30 mm × 36 mm operating at a frequency range of 3.1–10.6 GHz is proposed. The antenna is composed of two semielliptical monopoles fed by two microstrip lines. Two semicircular slots, two rectangular slots, and one stub are introduced in the ground plane to adjust the impedance bandwidth of the antenna and improve the isolation between two feeding ports. The simulated and measured results show that impedance bandwidth of the proposed antenna can cover the whole UWB band with a good isolation of < −15 dB. The radiation patterns, peak antenna gain, and envelope correlation coefficient are also measured and discussed. The measured results show that the proposed antenna can be a good candidate for some portable MIMO/diversity UWB applications.

  14. Compact Digital High Voltage Charger

    Li, Ge

    2005-01-01

    The operation of classical resonant circuit developed for the pulse energizing is investigated. The HV pulse or generator is very compact by a soft switching circuit made up of IGBT working at over 30 kHZ. The frequencies of macro pulses andμpulses can be arbitrarily tuned below resonant frequency to digitalize the HV pulse power. Theμpulses can also be connected by filter circuit to get the HVDC power. The circuit topology is given and its novel control logic is analyzed by flowchart. The circuit is part of a system consisting of a AC or DC LV power supply, a pulse transformer, the pulse generator implemented by LV capacitor and leakage inductance of the transformer, a HV DC or pulse power supply and the charged HV capacitor of the modulators.

  15. Compact Microwave Fourier Spectrum Analyzer

    Savchenkov, Anatoliy; Matsko, Andrey; Strekalov, Dmitry

    2009-01-01

    A compact photonic microwave Fourier spectrum analyzer [a Fourier-transform microwave spectrometer, (FTMWS)] with no moving parts has been proposed for use in remote sensing of weak, natural microwave emissions from the surfaces and atmospheres of planets to enable remote analysis and determination of chemical composition and abundances of critical molecular constituents in space. The instrument is based on a Bessel beam (light modes with non-zero angular momenta) fiber-optic elements. It features low power consumption, low mass, and high resolution, without a need for any cryogenics, beyond what is achievable by the current state-of-the-art in space instruments. The instrument can also be used in a wide-band scatterometer mode in active radar systems.

  16. Optimal shapes of compact strings

    Maritan, A.; Micheletti, C.; Trovato, A.; Banavar, J.R.

    2000-07-01

    Optimal geometrical arrangements, such as the stacking of atoms, are of relevance in diverse disciplines. A classic problem is the determination of the optimal arrangement of spheres in three dimensions in order to achieve the highest packing fraction; only recently has it been proved that the answer for infinite systems is a face-centred-cubic lattice. This simply stated problem has had a profound impact in many areas, ranging from the crystallization and melting of atomic systems, to optimal packing of objects and subdivision of space. Here we study an analogous problem-that of determining the optimal shapes of closely packed compact strings. This problem is a mathematical idealization of situations commonly encountered in biology, chemistry and physics, involving the optimal structure of folded polymeric chains. We find that, in cases where boundary effects are not dominant, helices with a particular pitch-radius ratio are selected. Interestingly, the same geometry is observed in helices in naturally-occurring proteins. (author)

  17. Physics of Compact Advanced Stellarators

    Zarnstorff, M.C.; Berry, L.A.; Brooks, A.; Fredrickson, E.; Fu, G.-Y.; Hirshman, S.; Hudson, S.; Ku, L.-P.; Lazarus, E.; Mikkelsen, D.; Monticello, D.; Neilson, G.H.; Pomphrey, N.; Reiman, A.; Spong, D.; Strickler, D.; Boozer, A.; Cooper, W.A.; Goldston, R.; Hatcher, R.; Isaev, M.; Kessel, C.; Lewandowski, J.; Lyon, J.; Merkel, P.; Mynick, H.; Nelson, B.E.; Nuehrenberg, C.; Redi, M.; Reiersen, W.; Rutherford, P.; Sanchez, R.; Schmidt, J.; White, R.B.

    2001-01-01

    Compact optimized stellarators offer novel solutions for confining high-beta plasmas and developing magnetic confinement fusion. The 3-D plasma shape can be designed to enhance the MHD stability without feedback or nearby conducting structures and provide drift-orbit confinement similar to tokamaks. These configurations offer the possibility of combining the steady-state low-recirculating power, external control, and disruption resilience of previous stellarators with the low-aspect ratio, high beta-limit, and good confinement of advanced tokamaks. Quasi-axisymmetric equilibria have been developed for the proposed National Compact Stellarator Experiment (NCSX) with average aspect ratio 4-4.4 and average elongation of approximately 1.8. Even with bootstrap-current consistent profiles, they are passively stable to the ballooning, kink, vertical, Mercier, and neoclassical-tearing modes for beta > 4%, without the need for external feedback or conducting walls. The bootstrap current generates only 1/4 of the magnetic rotational transform at beta = 4% (the rest is from the coils), thus the equilibrium is much less nonlinear and is more controllable than similar advanced tokamaks. The enhanced stability is a result of ''reversed'' global shear, the spatial distribution of local shear, and the large fraction of externally generated transform. Transport simulations show adequate fast-ion confinement and thermal neoclassical transport similar to equivalent tokamaks. Modular coils have been designed which reproduce the physics properties, provide good flux surfaces, and allow flexible variation of the plasma shape to control the predicted MHD stability and transport properties

  18. Compact autonomous navigation system (CANS)

    Hao, Y. C.; Ying, L.; Xiong, K.; Cheng, H. Y.; Qiao, G. D.

    2017-11-01

    Autonomous navigation of Satellite and constellation has series of benefits, such as to reduce operation cost and ground station workload, to avoid the event of crises of war and natural disaster, to increase spacecraft autonomy, and so on. Autonomous navigation satellite is independent of ground station support. Many systems are developed for autonomous navigation of satellite in the past 20 years. Along them American MANS (Microcosm Autonomous Navigation System) [1] of Microcosm Inc. and ERADS [2] [3] (Earth Reference Attitude Determination System) of Honeywell Inc. are well known. The systems anticipate a series of good features of autonomous navigation and aim low cost, integrated structure, low power consumption and compact layout. The ERADS is an integrated small 3-axis attitude sensor system with low cost and small volume. It has the Earth center measurement accuracy higher than the common IR sensor because the detected ultraviolet radiation zone of the atmosphere has a brightness gradient larger than that of the IR zone. But the ERADS is still a complex system because it has to eliminate many problems such as making of the sapphire sphere lens, birefringence effect of sapphire, high precision image transfer optical fiber flattener, ultraviolet intensifier noise, and so on. The marginal sphere FOV of the sphere lens of the ERADS is used to star imaging that may be bring some disadvantages., i.e. , the image energy and attitude measurements accuracy may be reduced due to the tilt image acceptance end of the fiber flattener in the FOV. Besides Japan, Germany and Russia developed visible earth sensor for GEO [4] [5]. Do we have a way to develop a cheaper/easier and more accurate autonomous navigation system that can be used to all LEO spacecraft, especially, to LEO small and micro satellites? To return this problem we provide a new type of the system—CANS (Compact Autonomous Navigation System) [6].

  19. General Relativity and Compact Stars

    Glendenning, Norman K.

    2005-01-01

    Compact stars--broadly grouped as neutron stars and white dwarfs--are the ashes of luminous stars. One or the other is the fate that awaits the cores of most stars after a lifetime of tens to thousands of millions of years. Whichever of these objects is formed at the end of the life of a particular luminous star, the compact object will live in many respects unchanged from the state in which it was formed. Neutron stars themselves can take several forms--hyperon, hybrid, or strange quark star. Likewise white dwarfs take different forms though only in the dominant nuclear species. A black hole is probably the fate of the most massive stars, an inaccessible region of spacetime into which the entire star, ashes and all, falls at the end of the luminous phase. Neutron stars are the smallest, densest stars known. Like all stars, neutron stars rotate--some as many as a few hundred times a second. A star rotating at such a rate will experience an enormous centrifugal force that must be balanced by gravity or else it will be ripped apart. The balance of the two forces informs us of the lower limit on the stellar density. Neutron stars are 10 14 times denser than Earth. Some neutron stars are in binary orbit with a companion. Application of orbital mechanics allows an assessment of masses in some cases. The mass of a neutron star is typically 1.5 solar masses. They can therefore infer their radii: about ten kilometers. Into such a small object, the entire mass of our sun and more, is compressed

  20. On compact multipliers of topological algebras

    Mohammad, N.

    1994-08-01

    It is shown that if the maximal ideal space Δ(A) of a semisimple commutative complete metrizable locally convex algebra contains no isolated points, then every compact multiplier is trivial. Particularly, compact multipliers on semisimple commutative Frechet algebras whose maximal ideal space has no isolated points are identically zero. (author). 5 refs

  1. Compaction Characteristics of Igumale Shale | Iorliam | Global ...

    This paper reports the outcome of an investigation into the effect of different compactive energies on the compaction characteristics of Igumale shale, to ascertain its suitability as fill material in highway ... The study showed that Igumale shale is not suitable for use as base, subbase and filling materials in road construction.

  2. Compact Process Development at Babcock & Wilcox

    Eric Shaber; Jeffrey Phillips

    2012-03-01

    Multiple process approaches have been used historically to manufacture cylindrical nuclear fuel compacts. Scale-up of fuel compacting was required for the Next Generation Nuclear Plant (NGNP) project to achieve an economically viable automated production process capable of providing a minimum of 10 compacts/minute with high production yields. In addition, the scale-up effort was required to achieve matrix density equivalent to baseline historical production processes, and allow compacting at fuel packing fractions up to 46% by volume. The scale-up approach of jet milling, fluid-bed overcoating, and hot-press compacting adopted in the U.S. Advanced Gas Reactor (AGR) Fuel Development Program involves significant paradigm shifts to capitalize on distinct advantages in simplicity, yield, and elimination of mixed waste. A series of compaction trials have been completed to optimize compaction conditions of time, temperature, and forming pressure using natural uranium oxycarbide (NUCO) fuel at packing fractions exceeding 46% by volume. Results from these trials are included. The scale-up effort is nearing completion with the process installed and operable using nuclear fuel materials. Final process testing is in progress to certify the process for manufacture of qualification test fuel compacts in 2012.

  3. Investigation of pressing of molybdenum powder compacts

    Mymrin, S.A.; Kuznetsov, V.Eh.; Yampol'skij, M.L.; Leonov, S.A.; Mikhridinov, R.M.; Korzukhin, V.A.

    1990-01-01

    Results of an experimental investigation into pressing of compacts of MCh type molybdenum powders using the industrial equipment are presented. To measure the density of powder molybdenum billets a radioisotopic density meter with cesium-137 is used as radioactive gamma radiation source. The dependence of the produced billet density on the specific compacting pressure at different values of the powder bulk density is ascertained

  4. Compact fuel storage rack for fuel pools

    Parras, F.; Louvat, J.P.

    1986-01-01

    ETS LEMER and FRAMATOME propose a new compact storage rack. This rack permits a considerable increase of the storage capacity of cooling pools. A short description of the structure and the components is presented, to propose racks that are: . Inalterable, . Compact, . Insensitive to earthquakes. Installation in pools already in operation is simplified by their light structure and the bearing device [fr

  5. Quantification of the compactibility of pharmaceutical powders

    Sonnergaard, Jørn

    2006-01-01

    The purpose of this study is to investigate and to quantify the compactibility of pharmaceutical powders by a simple linear relationship between the diametral compressive strength of tablets and the applied compaction pressure. The mechanical strength of the tablets is characterized as the crushing...

  6. Feature Based Control of Compact Disc Players

    Odgaard, Peter Fogh

    Two servo control loops are used to keep the Optical Pick-up Unit focused and radially on the information track of the Compact Disc. These control servos have problems handling surface faults on the Compact Disc. In this Ph.D thesis a method is proposed to improve the handling of these surface...

  7. Computing Decoupled Residuals for Compact Disc Players

    Odgaard, Peter Fogh; Stoustrup, Jakob; Andersen, Palle

    2006-01-01

    a pair of residuals generated by Compact Disc Player. However, these residuals depend on the performance of position servos in the Compact Disc Player. In other publications of the same authors a pair of decoupled residuals is derived. However, the computation of these alternative residuals has been...

  8. The classification of 2-compact groups

    K. S. Andersen, Kasper; Grodal, Jesper

    2009-01-01

    with Moeller and Viruel for p odd, this establishes the full classification of p-compact groups, stating that, up to isomorphism, there is a one-to-one correspondence between connected p-compact groups and root data over the p-adic integers. As a consequence we prove the maximal torus conjecture, giving a one...

  9. Computer aided design of operational units for tritium recovery from Li17Pb83 blanket of a DEMO fusion reactor

    Malara, C.; Viola, A.

    1995-01-01

    The problem of tritium recovery from Li 17 Pb 83 blanket of a DEMO fusion reactor is analyzed with the objective of limiting tritium permeation into the cooling water to acceptable levels. To this aim, a mathematical model describing the tritium behavior in blanket/recovery unit circuit has been formulated. By solving the model equations, tritium permeation rate into the cooling water and tritium inventory in the blanket are evaluated as a function of dimensionless parameters describing the combined effects of overall resistance for tritium transfer from Li 17 Pb 83 alloy to cooling water, circulating rate of the molten alloy in blanket/recovery unit circuit and extraction efficiency of tritium recovery unit. The extraction efficiency is, in turn, evaluated as a function of the operating conditions of recovery unit. The design of tritium recovery unit is then optimized on the basis of the above parametric analysis and the results are herein reported and discussed for a tritium permeation limit of 10 g/day into the cooling water. 14 refs., 9 figs., 2 tabs

  10. Sea Ice Detection Based on Differential Delay-Doppler Maps from UK TechDemoSat-1

    Yongchao Zhu

    2017-07-01

    Full Text Available Global Navigation Satellite System (GNSS signals can be exploited to remotely sense atmosphere and land and ocean surface to retrieve a range of geophysical parameters. This paper proposes two new methods, termed as power-summation of differential Delay-Doppler Maps (PS-D and pixel-number of differential Delay-Doppler Maps (PN-D, to distinguish between sea ice and sea water using differential Delay-Doppler Maps (dDDMs. PS-D and PN-D make use of power-summation and pixel-number of dDDMs, respectively, to measure the degree of difference between two DDMs so as to determine the transition state (water-water, water-ice, ice-ice and ice-water and hence ice and water are detected. Moreover, an adaptive incoherent averaging of DDMs is employed to improve the computational efficiency. A large number of DDMs recorded by UK TechDemoSat-1 (TDS-1 over the Arctic region are used to test the proposed sea ice detection methods. Through evaluating against ground-truth measurements from the Ocean Sea Ice SAF, the proposed PS-D and PN-D methods achieve a probability of detection of 99.72% and 99.69% respectively, while the probability of false detection is 0.28% and 0.31% respectively.

  11. The concept of system for chips production need to work demo CHP plant in company 'AGROSAVA' from Šimanovci

    Dedić Aleksandar Đ.

    2014-01-01

    Full Text Available In this paper according to the calculation of chips productivity needs for gasification in the demo CHP plant for co-generation: electricity and heat, chippers were analyzed due to: the type of mobility, running for chipping and the method of delivering chips to temporary yard. The plant was planned to generate electricity power up to 200kWelec. First, in consideration were taken the chippers with medium capacity, which mainly served for chipping brushwood and leaves that remain after harvest plantations on mostly flat terrain and parks. Later, the comparative characteristics of the world's three largest manufacturers of machinery for the production of wood chips significantly larger amounts (up to 30m3/h were given. These chippers were particularly suitable for the higher density of crops and stationed yard, in which brushwood would be brought and chip. At the end, the types of convective dryers were analyzed that could be successfully used for drying wood chips (drum and pneumatic dryer and based on the calculation proposed the types of dryers that were available in the local market.

  12. Phase change of First Wall in Water-Cooled Breeding Blankets of K-DEMO for Vertical

    Kim, Geon Woo; Lee, Jeong Hun; Cho, Hyoung Kyu; Park, Goon Cherl [Seoul National University, Seoul (Korea, Republic of); Im, Ki Hak [NFRI, Daejeon (Korea, Republic of)

    2016-05-15

    The purpose of this study is to simulate thermal-hydraulic behavior of a single blanket module when plasma disruption occurs. Plasma disruptions, such as vertical displacement events (VDE), with high heat flux can cause melting and vaporization of plasma facing materials and also burnout of coolant channels. The thermal design, evaluation and validation have been performed in order to establish the conceptual design guidelines of the water-cooled breeding blanket for the K-DEMO reactor. As a part of the NFRI research, Seoul National University (SNU) is conducting transient thermal-hydraulic analysis to confirm the integrity of blanket system for plasma disruption events. Vertical displacement events (VDE) with high heat flux can cause melting and vaporization of plasma facing materials (PFCs) and also burnout of coolant channels. In order to simulate melting of first wall in blanket module when VDE occurs, one-dimensional heat conduction equations were solved numerically with modification of the specific heat of the first wall materials using effective heat capacity method. Temperature profiles in first wall for VDE are shown in fig 7 - 9. At first, temperature of tungsten rapidly raised and even exceeded its melting temperature. When VDE just ended at 0.1 second, 0.83 mm thick of tungsten melted. But the other materials including vanadium and RAFM didn't exceed their melting temperatures after 500 seconds.

  13. First Spaceborne GNSS-Reflectometry Observations of Hurricanes From the UK TechDemoSat-1 Mission

    Foti, Giuseppe; Gommenginger, Christine; Srokosz, Meric

    2017-12-01

    We present the first examples of Global Navigation Satellite Systems-Reflectometry (GNSS-R) observations of hurricanes using spaceborne data from the UK TechDemoSat-1 (TDS-1) mission. We confirm that GNSS-R signals can detect ocean condition changes in very high near-surface ocean wind associated with hurricanes. TDS-1 GNSS-R reflections were collocated with International Best Track Archive for Climate Stewardship (IBTrACS) hurricane data, MetOp ASCAT A/B scatterometer winds, and two reanalysis products. Clear variations of GNSS-R reflected power (σ0) are observed as reflections travel through hurricanes, in some cases up to and through the eye wall. The GNSS-R reflected power is tentatively inverted to estimate wind speed using the TDS-1 baseline wind retrieval algorithm developed for low to moderate winds. Despite this, TDS-1 GNSS-R winds through the hurricanes show closer agreement with IBTrACS estimates than winds provided by scatterometers and reanalyses. GNSS-R wind profiles show realistic spatial patterns and sharp gradients that are consistent with expected structures around the eye of tropical cyclones.

  14. First spaceborne phase altimetry over sea ice using TechDemoSat-1 GNSS-R signals

    Li, Weiqiang; Cardellach, Estel; Fabra, Fran; Rius, Antonio; Ribó, Serni; Martín-Neira, Manuel

    2017-08-01

    A track of sea ice reflected Global Navigation Satellite System (GNSS) signal collected by the TechDemoSat-1 mission is processed to perform phase altimetry over sea ice. High-precision carrier phase measurements are extracted from coherent GNSS reflections at a high angle of elevation (>57°). The altimetric results show good consistency with a mean sea surface (MSS) model, and the root-mean-square difference is 4.7 cm with an along-track sampling distance of ˜140 m and a spatial resolution of ˜400 m. The difference observed between the altimetric results and the MSS shows good correlation with the colocated sea ice thickness data from Soil Moisture and Ocean Salinity. This is consistent with the reflecting surface aligned with the bottom of the ice-water interface, due to the penetration of the GNSS signal into the sea ice. Therefore, these high-precision altimetric results have potential to be used for determination of sea ice thickness.

  15. Phase change of First Wall in Water-Cooled Breeding Blankets of K-DEMO for Vertical

    Kim, Geon Woo; Lee, Jeong Hun; Cho, Hyoung Kyu; Park, Goon Cherl; Im, Ki Hak

    2016-01-01

    The purpose of this study is to simulate thermal-hydraulic behavior of a single blanket module when plasma disruption occurs. Plasma disruptions, such as vertical displacement events (VDE), with high heat flux can cause melting and vaporization of plasma facing materials and also burnout of coolant channels. The thermal design, evaluation and validation have been performed in order to establish the conceptual design guidelines of the water-cooled breeding blanket for the K-DEMO reactor. As a part of the NFRI research, Seoul National University (SNU) is conducting transient thermal-hydraulic analysis to confirm the integrity of blanket system for plasma disruption events. Vertical displacement events (VDE) with high heat flux can cause melting and vaporization of plasma facing materials (PFCs) and also burnout of coolant channels. In order to simulate melting of first wall in blanket module when VDE occurs, one-dimensional heat conduction equations were solved numerically with modification of the specific heat of the first wall materials using effective heat capacity method. Temperature profiles in first wall for VDE are shown in fig 7 - 9. At first, temperature of tungsten rapidly raised and even exceeded its melting temperature. When VDE just ended at 0.1 second, 0.83 mm thick of tungsten melted. But the other materials including vanadium and RAFM didn't exceed their melting temperatures after 500 seconds

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

    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.

  17. Neutronic design studies of a conceptual DCLL fusion reactor for a DEMO and a commercial power plant

    Palermo, I.; Veredas, G.; Gómez-Ros, J. M.; Sanz, J.; Ibarra, A.

    2016-01-01

    Neutronic analyses or, more widely, nuclear analyses have been performed for the development of a dual-coolant He/LiPb (DCLL) conceptual design reactor. A detailed three-dimensional (3D) model has been examined and optimized. The design is based on the plasma parameters and functional materials of the power plant conceptual studies (PPCS) model C. The initial radial-build for the detailed model has been determined according to the dimensions established in a previous work on an equivalent simplified homogenized reactor model. For optimization purposes, the initial specifications established over the simplified model have been refined on the detailed 3D design, modifying material and dimension of breeding blanket, shield and vacuum vessel in order to fulfil the priority requirements of a fusion reactor in terms of the fundamental neutronic responses. Tritium breeding ratio, energy multiplication factor, radiation limits in the TF coils, helium production and displacements per atom (dpa) have been calculated in order to demonstrate the functionality and viability of the reactor design in guaranteeing tritium self-sufficiency, power efficiency, plasma confinement, and re-weldability and structural integrity of the components. The paper describes the neutronic design improvements of the DCLL reactor, obtaining results for both DEMO and power plant operational scenarios.

  18. Analysis of the WCLL European demo blanket concept in terms of activation and decay heat after exposure to neutron irradiation

    Stankunas Gediminas

    2017-01-01

    Full Text Available This comparative paper describes the activation and decay heat calculations for water-cooled lithium-lead performed part of the EURO fusion WPSAE programme and specifications in comparison to other European DEMO blanket concepts on the basis of using a three-dimensional neutronics calculation model. Results are provided for a range of decay times of interest for maintenance activities, safety and waste management assessments. The study revealed that water-cooled lithium-lead has the highest total decay heat at longer decay times in comparison to the helium-cooled design which has the lowest total decay heat. In addition, major nuclides were identified for water-cooled lithium-lead in W armour, Eurofer, and LiPb. In addition, great attention has been dedicated to the analysis of the decay heat and activity both from the different water-cooled lithium-lead blanket modules for the entire reactor and from each water-cooled lithium-lead blanket module separately. The neutron induced activation and decay heat at shutdown were calculated by the FISPACT code, using the neutron flux densities and spectra that were provided by the preceding MCNP neutron transport calculations.

  19. Design of a Multi-Tube Pd-Membrane Module for Tritium Recovery from He in DEMO

    Marco Incelli

    2016-10-01

    Full Text Available Dense self-supported Pd-alloy membranes are used to selectively separate hydrogen and hydrogen isotopes. In particular, deuterium (D and tritium (T are currently identified as the main elements for the sustainability of the nuclear fusion reaction aimed at carbon free power generation. In the fusion nuclear reactors, a breeding blanket produces the tritium that is extracted and purified before being sent to the plasma chamber in order to sustain the fusion reaction. In this work, the application of Pd-alloy membranes has been tested for recovering tritium from a solid breeding blanket through a helium purge stream. Several simulations have been performed in order to optimize the design of a Pd-Ag multi-tube module in terms of geometry, operating parameters, and membrane module configuration (series vs. parallel. The results demonstrate that a pre-concentration stage before the Pd-membrane unit is mandatory because of the very low tritium concentration in the He which leaves the breeding blanket of the fusion reactor. The most suitable operating conditions could be reached by: (i increasing the hydrogen partial pressure in the lumen side and (ii decreasing the shell pressure. The preliminary design of a membrane unit has been carried out for the case of the DEMO fusion reactor: the optimized membrane module consists of an array of 182 Pd-Ag tubes of 500 mm length, 10 mm diameter, and 0.100 mm wall thickness (total active area of 2.85 m2.

  20. Correlating particle hardness with powder compaction performance.

    Cao, Xiaoping; Morganti, Mikayla; Hancock, Bruno C; Masterson, Victoria M

    2010-10-01

    Assessing particle mechanical properties of pharmaceutical materials quickly and with little material can be very important to early stages of pharmaceutical research. In this study, a wide range of pharmaceutical materials were studied using atomic force microscopy (AFM) nanoindentation. A significant amount of particle hardness and elastic modulus data were provided. Moreover, powder compact mechanical properties of these materials were investigated in order to build correlation between the particle hardness and powder compaction performance. It was found that the materials with very low or high particle hardness most likely exhibit poor compaction performance while the materials with medium particle hardness usually have good compaction behavior. Additionally, the results from this study enriched Hiestand's special case concept on particle hardness and powder compaction performance. This study suggests that the use of AFM nanoindentation can help to screen mechanical properties of pharmaceutical materials at early development stages of pharmaceutical research.

  1. On compact galaxies in the UGC catalogue

    Kogoshvili, N.G.

    1980-01-01

    A problem of separation of compact galaxies in the UGC Catalogue is considered. Value of surface brightness equal to or less than 21sup(m) was used as compactness criterion from a square second of arc. 96 galaxies, which are brighter than 14sup(m)5 satisfy this criterion. Among compact galaxies discovered in the UGC Catalogue 7% are the Zwicky galaxies, 15% belong to the Markarian galaxies and 27% of galaxies are part of a galaxy list with high surface brightness. Considerable divergence in estimates of total share of compact galaxies in the B.A. Worontsov-Veljaminov Morphological Catalogue of Galaxies (MCG) and the UGC Catalogue is noted. This divergence results from systematical underestimation of visible sizes of compact galaxies in the MCG Catalogue as compared with the UGC Catalogue [ru

  2. Cooling of hypernuclear compact stars

    Raduta, Adriana R.; Sedrakian, Armen; Weber, Fridolin

    2018-04-01

    We study the thermal evolution of hypernuclear compact stars constructed from covariant density functional theory of hypernuclear matter and parametrizations which produce sequences of stars containing two-solar-mass objects. For the input in the simulations, we solve the Bardeen-Cooper-Schrieffer gap equations in the hyperonic sector and obtain the gaps in the spectra of Λ, Ξ0, and Ξ- hyperons. For the models with masses M/M⊙ ≥ 1.5 the neutrino cooling is dominated by hyperonic direct Urca processes in general. In the low-mass stars the (Λp) plus leptons channel is the dominant direct Urca process, whereas for more massive stars the purely hyperonic channels (Σ-Λ) and (Ξ-Λ) are dominant. Hyperonic pairing strongly suppresses the processes on Ξ-s and to a lesser degree on Λs. We find that intermediate-mass 1.5 ≤ M/M⊙ ≤ 1.8 models have surface temperatures which lie within the range inferred from thermally emitting neutron stars, if the hyperonic pairing is taken into account. Most massive models with M/M⊙ ≃ 2 may cool very fast via the direct Urca process through the (Λp) channel because they develop inner cores where the S-wave pairing of Λs and proton is absent.

  3. Spherically symmetric charged compact stars

    Maurya, S.K. [University of Nizwa, Department of Mathematical and Physical Sciences, College of Arts and Science, Nizwa (Oman); Gupta, Y.K. [Jaypee Institute of Information Technology University, Department of Mathematics, Noida, Uttar Pradesh (India); Ray, Saibal [Government College of Engineering and Ceramic Technology, Department of Physics, Kolkata, West Bengal (India); Chowdhury, Sourav Roy [Seth Anandaram Jaipuria College, Department of Physics, Kolkata, West Bengal (India)

    2015-08-15

    In this article we consider the static spherically symmetric metric of embedding class 1. When solving the Einstein-Maxwell field equations we take into account the presence of ordinary baryonic matter together with the electric charge. Specific new charged stellar models are obtained where the solutions are entirely dependent on the electromagnetic field, such that the physical parameters, like density, pressure etc. do vanish for the vanishing charge. We systematically analyze altogether the three sets of Solutions I, II, and III of the stellar models for a suitable functional relation of ν(r). However, it is observed that only the Solution I provides a physically valid and well-behaved situation, whereas the Solutions II and III are not well behaved and hence not included in the study. Thereafter it is exclusively shown that the Solution I can pass through several standard physical tests performed by us. To validate the solution set presented here a comparison has also been made with that of the compact stars, like RX J 1856 - 37, Her X - 1, PSR 1937+21, PSRJ 1614-2230, and PSRJ 0348+0432, and we have shown the feasibility of the models. (orig.)

  4. Compact 3D quantum memory

    Xie, Edwar; Deppe, Frank; Renger, Michael; Repp, Daniel; Eder, Peter; Fischer, Michael; Goetz, Jan; Pogorzalek, Stefan; Fedorov, Kirill G.; Marx, Achim; Gross, Rudolf

    2018-05-01

    Superconducting 3D microwave cavities offer state-of-the-art coherence times and a well-controlled environment for superconducting qubits. In order to realize at the same time fast readout and long-lived quantum information storage, one can couple the qubit to both a low-quality readout and a high-quality storage cavity. However, such systems are bulky compared to their less coherent 2D counterparts. A more compact and scalable approach is achieved by making use of the multimode structure of a 3D cavity. In our work, we investigate such a device where a transmon qubit is capacitively coupled to two modes of a single 3D cavity. External coupling is engineered so that the memory mode has an about 100 times larger quality factor than the readout mode. Using an all-microwave second-order protocol, we realize a lifetime enhancement of the stored state over the qubit lifetime by a factor of 6 with a fidelity of approximately 80% determined via quantum process tomography. We also find that this enhancement is not limited by fundamental constraints.

  5. Does soil compaction increase floods? A review

    Alaoui, Abdallah; Rogger, Magdalena; Peth, Stephan; Blöschl, Günter

    2018-02-01

    Europe has experienced a series of major floods in the past years which suggests that flood magnitudes may have increased. Land degradation due to soil compaction from crop farming or grazing intensification is one of the potential drivers of this increase. A literature review suggests that most of the experimental evidence was generated at plot and hillslope scales. At larger scales, most studies are based on models. There are three ways in which soil compaction affects floods at the catchment scale: (i) through an increase in the area affected by soil compaction; (ii) by exacerbating the effects of changes in rainfall, especially for highly degraded soils; and (iii) when soil compaction coincides with soils characterized by a fine texture and a low infiltration capacity. We suggest that future research should focus on better synthesising past research on soil compaction and runoff, tailored field experiments to obtain a mechanistic understanding of the coupled mechanical and hydraulic processes, new mapping methods of soil compaction that combine mechanical and remote sensing approaches, and an effort to bridge all disciplines relevant to soil compaction effects on floods.

  6. Evaluation of automatic vacuum- assisted compaction solutions

    M. Brzeziński

    2011-01-01

    Full Text Available Currently on the mould-making machines market the companies like: DiSA, KUENKEL WAGNER, HAFLINGER, HEINRICH WAGNER SINTO, HUNTER, SAVELLI AND TECHNICAL play significant role. These companies are the manufacturers of various solutions in machines and instalations applied in foundry engineering. Automatic foundry machines for compaction of green sand have the major role in mechanisation and automation processes of making the mould. The concept of operation of automatic machines is based on the static and dynamic methods of compacting the green sand. The method which gains the importance is the compacting method by using the energy of the air pressure. It's the initial stage or the supporting process of compacting the green sand. However in the automatic mould making machines using this method it's essential to use the additional compaction of the mass in order to receive the final parameters of the form. In the constructional solutions of the machines there is the additional division which concerns the method of putting the sand into the mould box. This division distinquishes the transport of the sand with simultaneous compaction or the putting of the sand without the pre-compaction. As the solutions of the major manufacturers are often the subject for application in various foundries, the authors of the paper would like/have the confidence to present their own evaluation process confirmed by their own researches and independent analysis of the producers' solutions.

  7. Soil compaction and fertilization in soybean productivity

    Beutler Amauri Nelson

    2004-01-01

    Full Text Available Soil compaction and fertilization affect soybean development. This study evaluated the effects of soil compaction and fertilization on soybean (Glycine max cv. Embrapa 48 productivity in a Typic Haplustox under field conditions in Jaboticabal, SP, Brazil. A completely randomized design with a 5 x 2 factorial layout (compaction vs. fertilization, with four replications in each treatment, was employed. Each experimental unit (replicate consisted of a 3.6 m² useful area. After the soil was prepared by cultivation, an 11 Mg tractor passed over it a variable number of times to create five levels of compaction. Treatments were: T0= no compaction, T1= one tractor pass, T2= two, T4= four, and T6= six passes, and no fertilizer and fertilizer to give soybean yields of 2.5 to 2.9 Mg ha-1. Soil was sampled at depths of 0.02-0.05, 0.07-0.10, and 0.15-0.18 m to determine macro and microporosity, penetration resistance (PR, and bulk density (Db. After 120 days growing under these conditions, the plants were analyzed in terms of development (plant height, number of pods, shoot dry matter per plant and weight of 100 seeds and seed productivity per hectare. Soil compaction decreased soybean development and productivity, but this effect was decreased by soil fertilization, showing that such fertilization increased soybean tolerance to soil compaction.

  8. Topological entropy of continuous actions of compactly generated groups

    Schneider, Friedrich Martin

    2015-01-01

    We introduce a notion of topological entropy for continuous actions of compactly generated topological groups on compact Hausdorff spaces. It is shown that any continuous action of a compactly generated topological group on a compact Hausdorff space with vanishing topological entropy is amenable. Given an arbitrary compactly generated locally compact Hausdorff topological group $G$, we consider the canonical action of $G$ on the closed unit ball of $L^{1}(G)' \\cong L^{\\infty}(G)$ endowed with...

  9. Strategy Guideline. Compact Air Distribution Systems

    Burdick, Arlan [IBACOS, Inc., Pittsburgh, PA (United States)

    2013-06-01

    This guideline discusses the benefits and challenges of using a compact air distribution system to handle the reduced loads and reduced air volume needed to condition the space within an energy efficient home. The decision criteria for a compact air distribution system must be determined early in the whole-house design process, considering both supply and return air design. However, careful installation of a compact air distribution system can result in lower material costs from smaller equipment, shorter duct runs, and fewer outlets; increased installation efficiencies, including ease of fitting the system into conditioned space; lower loads on a better balanced HVAC system, and overall improved energy efficiency of the home.

  10. 'Crescent'-shaped tokamak for compact ignition

    Yamazaki, K.; Reiersen, W.T.

    1985-12-01

    A compact high-beta tokamak configuration with ''crescent''-shaped (or ''boomerang''-shaped) cross-section is proposed as a next-generation ignition machine. This configuration with a small indentation but a large triangularity is more compact than the normal dee-shaped design because of its high-beta characteristics in the first-second transition regime of stability. This may also be a more reliable next-generation compact device than the bean-shaped design with large indentation and small triangularity, because this design dose not rely on the second stability and is easily extendable from the present dee-shaped design. (author)

  11. 'Crescent'-shaped tokamak for compact ignition

    Yamazaki, K.; Reiersen, W.T.

    1986-01-01

    A compact high-beta tokamak configuration with ''crescent''-shaped (or ''boomerang''-shaped) cross section is proposed as a next-generation ignition machine. This configuration with a small indentation but a large triangularity is more compact than the normal dee-shaped design because of its high-beta characteristics in the first-second transition regime of stability. This may also be a more reliable next-generation compact device than the bean-shaped design with large indentation and small triangularity, because this design does not rely on the second stability and is easily extendable from the present dee-shaped design. (author)

  12. Generalised model for anisotropic compact stars

    Maurya, S.K. [University of Nizwa, Department of Mathematical and Physical Sciences College of Arts and Science, Nizwa (Oman); Gupta, Y.K. [Raj Kumar Goel Institute of Technology, Department of Mathematics, Ghaziabad, Uttar Pradesh (India); Ray, Saibal [Government College of Engineering and Ceramic Technology, Department of Physics, Kolkata, West Bengal (India); Deb, Debabrata [Indian Institute of Engineering Science and Technology, Shibpur, Department of Physics, Howrah, West Bengal (India)

    2016-12-15

    In the present investigation an exact generalised model for anisotropic compact stars of embedding class 1 is sought with a general relativistic background. The generic solutions are verified by exploring different physical aspects, viz. energy conditions, mass-radius relation, stability of the models, in connection to their validity. It is observed that the model presented here for compact stars is compatible with all these physical tests and thus physically acceptable as far as the compact star candidates RXJ 1856-37, SAX J 1808.4-3658 (SS1) and SAX J 1808.4-3658 (SS2) are concerned. (orig.)

  13. Ion diffusion in compacted bentonite

    Lehikoinen, J. [VTT Chemical Technology, Espoo (Finland)

    1999-03-01

    In the study, a two-dimensional molecular-level diffusion model, based on a modified form of the Gouy-Chapman (GC) theory of the electrical double layers, for hydrated ionic species in compacted bentonite was developed. The modifications to the GC theory, which forms the very kernel of the diffusion model, stem from various non-conventional features: ionic hydration, dielectric saturation, finite ion-sizes and specific adsorption. The principal objectives of the study were met. With the aid of the consistent diffusion model, it is a relatively simple matter to explain the experimentally observed macroscopic exclusion for anions as well as the postulated, but greatly controversial, surface diffusion for cations. From purely theoretical grounds, it was possible to show that the apparent diffusivities of cations, anions and neutral molecules (i) do not exhibit order-or-magnitude differences, and (ii) are practically independent of the solution ionic strength used and, consequently, of the distribution coefficient, K{sub d}, unless they experience specific binding onto the substrate surface. It was also of interest to investigate the equilibrium anionic concentration distribution in the pore geometry of the GMM model as a function of the solution ionic strength, and to briefly speculate its consequences to diffusion. An explicit account of the filter-plate effect was taken by developing a computerised macroscopic diffusion model, which is based upon the very robust and efficient Laplace Transform Finite-Difference technique. Finally, the inherent limitations as well as the potential fields of applications of the models were addressed. (orig.) 45 refs.

  14. Development of compact nuclear simulator

    Ham, Chang Shik; Kwon, Kee Choon; Lyu, Sung Phil; Kim, Jung Taek; Jung, Chul Hwan; Lee, Dong Young; Hwang, In Koo; Kim, Young Gil; Kim, Jung Soo; Park, Won Man

    1988-12-01

    Compact nuclear simulator is designed to carry out the various operational modes as real nuclear power plant, start-up, preoperational test, preheating, hot start-up, cold shutdown, power control and the operational conditions in steady and accident states. It can be used for the fundamental training of the operators, maintenance personnel, inspectors of regulatory body, system or component designers, NSSS designers, safety analysis by transient analysis and for the making questions for an operator qualifying examination and the training of research fellows in the Nuclear Training Center of KAERI. Everyone knows that the TMI accident resulted from the defect of the man-machine interface of main control room and of the quality of the operators. No proper action on the malfunction of small part in a system can cause severe accident like TMI-2 accident, so it is very important urgent to upgrade the operators' capability and to train operators for the understanding of dynamic transient phenomena in plant system. So it is necessary to develop CNS which is very efficient to train operators, operation and maintenance supervisors, maintenance personnel and inspectors of regulation committee to understand the dynamic transient phenomena. This report is the final report of KAERI-CNS project which was designed and manufactured in '85.7-'88.12. This CNS was designed and fabricated in conjunction with STUDSVIK, Sweden and installed at KAERI-NTC, and entitled KAERI-CNS. KAERI and STUDSVIK have developed math. modeling software. Many parts of CNS hardware were supplied by local firms.The followings are major parts of this project performed in '85.7-'88.12. 1.Contract with STUDSVIK for joint design and manufacturing CNS 2.Selection of malfunctions and design and manufacture of console panel 3.Manufacture of interface card and graphic display system 4.Software module development 5.S/W and H/W integration 6.Factory acceptance test and Site acceptance test 7.Running test. (Author)

  15. Ion diffusion in compacted bentonite

    Lehikoinen, J.

    1999-03-01

    In the study, a two-dimensional molecular-level diffusion model, based on a modified form of the Gouy-Chapman (GC) theory of the electrical double layers, for hydrated ionic species in compacted bentonite was developed. The modifications to the GC theory, which forms the very kernel of the diffusion model, stem from various non-conventional features: ionic hydration, dielectric saturation, finite ion-sizes and specific adsorption. The principal objectives of the study were met. With the aid of the consistent diffusion model, it is a relatively simple matter to explain the experimentally observed macroscopic exclusion for anions as well as the postulated, but greatly controversial, surface diffusion for cations. From purely theoretical grounds, it was possible to show that the apparent diffusivities of cations, anions and neutral molecules (i) do not exhibit order-or-magnitude differences, and (ii) are practically independent of the solution ionic strength used and, consequently, of the distribution coefficient, K d , unless they experience specific binding onto the substrate surface. It was also of interest to investigate the equilibrium anionic concentration distribution in the pore geometry of the GMM model as a function of the solution ionic strength, and to briefly speculate its consequences to diffusion. An explicit account of the filter-plate effect was taken by developing a computerised macroscopic diffusion model, which is based upon the very robust and efficient Laplace Transform Finite-Difference technique. Finally, the inherent limitations as well as the potential fields of applications of the models were addressed. (orig.)

  16. Compact Ceramic Microchannel Heat Exchangers

    Lewinsohn, Charles [Ceramatec, Inc., Salt Lake City, UT (United States)

    2016-10-31

    The objective of the proposed work was to demonstrate the feasibility of a step change in power plant efficiency at a commercially viable cost, by obtaining performance data for prototype, compact, ceramic microchannel heat exchangers. By performing the tasks described in the initial proposal, all of the milestones were met. The work performed will advance the technology from Technology Readiness Level 3 (TRL 3) to Technology Readiness Level 4 (TRL 4) and validate the potential of using these heat exchangers for enabling high efficiency solid oxide fuel cell (SOFC) or high-temperature turbine-based power plants. The attached report will describe how this objective was met. In collaboration with The Colorado School of Mines (CSM), specifications were developed for a high temperature heat exchanger for three commercial microturbines. Microturbines were selected because they are a more mature commercial technology than SOFC, they are a low-volume and high-value target for market entry of high-temperature heat exchangers, and they are essentially scaled-down versions of turbines used in utility-scale power plants. Using these specifications, microchannel dimensions were selected to meet the performance requirements. Ceramic plates were fabricated with microchannels of these dimensions. The plates were tested at room temperature and elevated temperature. Plates were joined together to make modular, heat exchanger stacks that were tested at a variety of temperatures and flow rates. Although gas flow rates equivalent to those in microturbines could not be achieved in the laboratory environment, the results showed expected efficiencies, robust operation under significant temperature gradients at high temperature, and the ability to cycle the stacks. Details of the methods and results are presented in this final report.

  17. Investigation of wetting property between liquid lead lithium alloy and several structural materials for Chinese DEMO reactor

    Lu, Wei; Wang, Weihua; Jiang, Haiyan; Zuo, Guizhong; Pan, Baoguo; Xu, Wei; Chu, Delin; Hu, Jiansheng; Qi, Junli

    2017-10-01

    The dual-cooled lead lithium (PbLi) blanket is considered as one of the main options for the Chinese demonstration reactor (DEMO). Liquid PbLi alloy is used as the breeder material and coolant. Reduced activation ferritic/martensitic (RAFM) steel, stainless steel and the silicon carbide ceramic matrix composite (SiCf) are selected as the substrate materials for different use. To investigate the wetting property and inter-facial interactions of PbLi/RAFM steel, PbLi/SS316L, PbLi/SiC and PbLi/SiCf couples, in this paper, the special vacuum experimental device is built, and the 'dispensed droplet' modification for the classic sessile droplet technique is made. Contact angles are measured between the liquid PbLi and the various candidate materials at blanket working temperature from 260 to 480 °C. X-ray photoelectron spectroscopy (XPS) is used to characterize the surface components of PbLi droplets and substrate materials, in order to study the element trans-port and corrosion mechanism. Results show that SiC composite (SiCf) and SiC ceramic show poor wetting properties with the liquid PbLi alloy. Surface roughness and testing temperature only provide tiny improvements on the wetting property below 480 °C. RAFM steel performs better wetting properties and corrosion residence when contacted with molten PbLi, while SS316L shows low corrosion residence above 420 °C for the decomposition of protective surface film mainly consisted of chromic sesquioxide. The results could provide meaningful compatibility database of liquid PbLi alloy and valuable reference in engineering design of candidate structural and functional materials for future fusion blanket.

  18. Influence of thermal performance on design parameters of a He/LiPb dual coolant DEMO concept blanket design

    Mas de les Valls, E., E-mail: elisabet.masdelesvalls@gits.ws [Technical University of Catalonia (UPC), Jordi Girona 1-3, 08034 Barcelona (Spain); Technology for Fusion (T4F) Research Group, GREENER, Department of Heat Engines, Barcelona (Spain); Batet, L. [Technical University of Catalonia (UPC), Jordi Girona 1-3, 08034 Barcelona (Spain); Technology for Fusion (T4F) Research Group, GREENER, Department of Physics and Nuclear Engineering, Barcelona (Spain); Medina, V. de [Technical University of Catalonia (UPC), Jordi Girona 1-3, 08034 Barcelona (Spain); Sediment Transport Research Group, Department of Engineering Hydraulic, Marine and Environmental Engineering, Barcelona (Spain); Fradera, J. [Technical University of Catalonia (UPC), Jordi Girona 1-3, 08034 Barcelona (Spain); Technology for Fusion (T4F) Research Group, GREENER, Department of Physics and Nuclear Engineering, Barcelona (Spain); Sanmarti, M. [bFUS-IREC, Jardins de les Dones de Negre 1, 08930 Sant Adria del Besos (Spain); Sedano, L.A. [EURATOM-CIEMAT Association, 28040 Madrid (Spain)

    2012-08-15

    Spanish Breeding Blanket Technology Programme TECNO{sub F}US is exploring the technological capabilities of a Dual-Coolant He/Pb15.7Li breeding blanket for DEMO and studying new breeding blanket design specifications. The progress of the channel conceptual design is being conducted in parallel with the extension of MHD computational capabilities of CFD tools and the underlying physics of MHD models. A qualification of MHD effects under present blanket design specifications and some approaches to their modelling were proposed by the authors in . The analysis was accomplished with the 2D transient algorithm from Sommeria and Moreau and implemented in the OpenFOAM CFD toolbox . The thermal coupling was implemented by means of the Boussinesq hypothesis. Previous analyses showed the need of improvement of FCI thickness and thermal properties in order to obtain a desirable liquid metal temperature gain of 300 Degree-Sign C. In the present study, an assessment through sensitivity and parametric analyses of the required FCI thickness is performed. Numerical simulations have been carried out considering a Robin-type thermal boundary condition which assumes 1D steady state thermal balance across the solid FCI and Eurofer layers. Such boundary condition has been validated with a fluid-solid coupled domain analysis. Results for the studied flow conditions and channel dimensions show that, in order to obtain a liquid metal temperature gain of about 300 Degree-Sign C, the required FCI material should have a very small effective heat transfer coefficient ((k/{delta}) {<=} 1 W/m{sup 2}K) and fluid velocities should be about 0.2 m/s or less. Moreover, special attention has to be placed on the temperature difference across the FCI layer. However, for a maximised liquid metal thermal gain, higher velocities would be preferable, what would also imply a reduced temperature difference across the FCI layer.

  19. Diagnostics for the National Compact Stellarator Experiment

    Stratton, B.C.; Johnson, D.; Feder, R.; Fredrickson, E.; Neilson, H.; Takahashi, H.; Zarnstorf, M.; Cole, M.; Goranson, P.; Lazarus, E.; Nelson, B.

    2003-01-01

    The status of planning of the National Compact Stellarator Experiment (NCSX) diagnostics is presented, with the emphasis on resolution of diagnostics access issues and on diagnostics required for the early phases of operation

  20. Observing Compact Stars with AstroSat

    Dipankar Bhattacharya

    2017-09-12

    Sep 12, 2017 ... based observatory for compact star research. An account is given of ... unprecedented capability to study such rapid variability simultaneously at all ..... Physical Research Laboratory, University of Leicester and the Canadian ...