WorldWideScience

Sample records for ac transit demos

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

    Science.gov (United States)

    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.

  2. Alameda-Contra Costa Transit District (AC Transit) Fuel Cell Transit Buses: Preliminary Evaluation Results

    Energy Technology Data Exchange (ETDEWEB)

    Chandler, K.; Eudy, L.

    2007-03-01

    This report provides an evaluation of three prototype fuel cell-powered transit buses operating at AC Transit in Oakland, California, and six baseline diesel buses similar in design to the fuel cell buses.

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

    Science.gov (United States)

    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.

  4. National Fuel Cell Bus Program : Accelerated Testing Report, AC Transit

    Science.gov (United States)

    2009-01-01

    This is an evaluation of hydrogen fuel cell transit buses operating at AC Transit in revenue service since March 20, 2006 compared to similar diesel buses operating from the same depot. This evaluation report includes results from November 2007 throu...

  5. Fast electric dipole transitions in Ra-Ac nuclei

    International Nuclear Information System (INIS)

    Ahmad, I.

    1985-01-01

    Lifetime of levels in 225 Ra, 225 Ac, and 227 Ac have been measured by delayed coincidence techniques and these have been used to determine the E1 gamma-ray transition probabilities. The reduced E1 transition probabilities. The reduced E1 transition probabilities in 225 Ra and 225 Ac are about two orders of magnitude larger than the values in mid-actinide nuclei. On the other hand, the E1 rate in 227 Ac is similar to those measured in heavier actinides. Previous studies suggest the presence of octupole deformation in all the three nuclei. The present investigation indicates that fast E1 transitions occur for nuclei with octupole deformation. However, the studies also show that there is no one-to-one correspondence between E1 rate and octupole deformation. 13 refs., 4 figs

  6. DC and AC biasing of a transition edge sensor microcalorimeter

    International Nuclear Information System (INIS)

    Cunningham, M.F.; Ullom, J.N.; Miyazaki, T.; Drury, O.; Loshak, A.; Berg, M.L. van den; Labov, S.E.

    2002-01-01

    We are developing AC-biased transition edge sensor (TES) microcalorimeters for use in large arrays with frequency-domain multiplexing. Using DC bias, we have achieved a resolution of 17 eV FWHM at 2.6 keV with a decay time of 90 μs and an effective detector diameter of 300 μm. We have successfully measured thermal pulses with a TES microcalorimeter operated with an AC bias. We present here preliminary results from a single pixel detector operated under DC and AC bias conditions

  7. Transition towards DC micro grids: From an AC to a hybrid AC and DC energy infrastructure

    Directory of Open Access Journals (Sweden)

    Evi Ploumpidou

    2017-12-01

    Full Text Available Our electricity is predominantly powered by alternating current (AC, ever since the War of Currents ended in the favor of Nicola Tesla at the end of the 19th century. However, lots of the appliances we use, such as electronics and lights with light-emitting diode (LED technology, work internally on direct current (DC and it is projected that the number of these appliances will increase in the near future. Another contributor to the increase in DC consumption is the ongoing electrification of mobility (Electric Vehicles (EVs. At the same time, photovoltaics (PV generate DC voltages, while the most common storage technologies also use DC. In order to integrate all these appliances and technologies to the existing AC grid, there is a need for converters which introduce power losses. By distributing DC power to DC devices instead of converting it to AC first, it is possible to avoid substantial energy losses that occur every time electricity is converted. This situation initiated the concept for the implementation of the DC-Flexhouse project. A prototype DC installation will be developed and tested in one of the buildings of the developing living lab area called the District of Tomorrow (De Wijk van Morgen which is located in Heerlen, the Netherlands. A neighborhood cooperative (Vrieheide cooperatie is also part of the consortium in order to address the aspect of social acceptance. Although DC seems to be a promising solution for a more sustainable energy system, the business case is still debatable due to both technology- and market-related challenges. The current energy infrastructure is predominantly based on AC, manufacturers produce devices based on AC standards and people are using many AC products across a long life span. This Smart Energy Buildings & Cities (SEB&C PDEng project is a contribution to the DC-Flexhouse project. The aim is to analyze the challenges in the transition to DC micro grids, assess the market potential of DC

  8. National Fuel Cell Bus Program: Accelerated Testing Evaluation Report and Appendices, Alameda-Contra Costa Transit District (AC Transit)

    Energy Technology Data Exchange (ETDEWEB)

    Chandler, K.; Eudy, L.

    2009-01-01

    This is an evaluation of hydrogen fuel cell transit buses operating at AC Transit in revenue service since March 20, 2006 compared to similar diesel buses operating from the same depot. This evaluation report includes results from November 2007 through October 2008. Evaluation results include implementation experience, fueling station operation, fuel cell bus operations at Golden Gate Transit, and evaluation results at AC Transit (bus usage, availability, fuel economy, maintenance costs, and roadcalls).

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

    Czech Academy of Sciences Publication Activity Database

    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. Breeding blanket for Demo

    International Nuclear Information System (INIS)

    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

  11. Hidden transition in multiferroic and magnetodielectric CuCrO2 evidenced by ac-susceptibility

    Science.gov (United States)

    Shukla, Kaushak K.; Pal, Arkadeb; Singh, Abhishek; Singh, Rahul; Saha, J.; Sinha, A. K.; Ghosh, A. K.; Patnaik, S.; Awasthi, A. M.; Chatterjee, Sandip

    2017-04-01

    Ferroelectric polarization, magnetic-field dependence of the dielectric constant and ac and dc magnetizations of frustrated CuCrO2 have been measured. A new spin freezing transition below 32 K is observed which is thermally driven. The nature of the spin freezing is to be a single-ion process. Dilution by the replacements of Cr ions by magnetic Mn ions showed suppression of the spin freezing transition suggesting it to be fundamentally a single-ion freezing process. The observed freezing, which is seemingly associated to geometrical spin frustration, represents a novel form of magnetic glassy behavior.

  12. What must Demo do?

    International Nuclear Information System (INIS)

    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

  13. Effect of alkali content on AC conductivity of borate glasses containing two transition metals

    International Nuclear Information System (INIS)

    Kashif, I.; Rahman, Samy A.; Soliman, A.A.; Ibrahim, E.M.; Abdel-Khalek, E.K.; Mostafa, A.G.; Sanad, A.M.

    2009-01-01

    Sodium borate glasses containing iron and molybdenum ions with the total concentration of transition ions constant and gradual substitution of sodium oxide (network modifier) by borate oxide (network former) was prepared. Densities, molar volume, DC and AC conductivities are measured. The trends of these properties are attributed to changes in the glass network structure. Their DC and AC conductivity increased with increasing NaO concentration. The increase of AC conductivity of sodium borate glasses is attributed to the chemical composition and the hopping mechanism of conduction. Measurements of the dielectric constant (ε) and dielectric loss (tan δ) as a function of frequency (50 Hz-100 kHz) and temperature (RT-600 K) indicate that the increase in dielectric constant and loss (ε and tan δ) values with increasing sodium ion content could be attributed to the assumption that Fe and Mo ions tend to assume network-forming position in the glass compositions studied. The variation of the value of frequency exponent s for all glass samples as the function of temperature at a definite frequency indicates that the value of s decreases with increasing the temperature which agrees with the correlated barrier-hopping (CBH) model.

  14. Assessment of DEMO challenges in technology and physics

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. DEMO diagnostics and burn control

    NARCIS (Netherlands)

    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

  16. Introduction: Undoing the demos

    DEFF Research Database (Denmark)

    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....

  17. The DEMO Quasisymmetric Stellarator

    Directory of Open Access Journals (Sweden)

    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.

  18. DEMO diagnostics and burn control

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. Review of fusion DEMO reactor study

    International Nuclear Information System (INIS)

    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)

  20. Demo of Gaze Controlled Flying

    DEFF Research Database (Denmark)

    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....

  1. PSO-Ensemble Demo Application

    DEFF Research Database (Denmark)

    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...

  2. Steady State versus Pulsed Tokamak DEMO

    Energy Technology Data Exchange (ETDEWEB)

    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)

  3. The DEMO wall load challenge

    Czech Academy of Sciences Publication Activity Database

    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

  4. Experimental investigations on characteristics of boundary layer and control of transition on an airfoil by AC-DBD

    Science.gov (United States)

    Geng, Xi; Shi, Zhiwei; Cheng, Keming; Dong, Hao; Zhao, Qun; Chen, Sinuo

    2018-03-01

    Plasma-based flow control is one of the most promising techniques for aerodynamic problems, such as delaying the boundary layer transition. The boundary layer’s characteristics induced by AC-DBD plasma actuators and applied by the actuators to delay the boundary layer transition on airfoil at Ma = 0.3 were experimentally investigated. The PIV measurement was used to study the boundary layer’s characteristics induced by the plasma actuators. The measurement plane, which was parallel to the surface of the actuators and 1 mm above the surface, was involved in the test, including the perpendicular plane. The instantaneous results showed that the induced flow field consisted of many small size unsteady vortices which were eliminated by the time average. The subsequent oil-film interferometry skin friction measurement was conducted on a NASA SC(2)-0712 airfoil at Ma = 0.3. The coefficient of skin friction demonstrates that the plasma actuators successfully delay the boundary layer transition and the efficiency is better at higher driven voltage.

  5. High current superconductors for DEMO

    Energy Technology Data Exchange (ETDEWEB)

    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.

  6. Dual role of an ac driving force and the underlying two distinct order–disorder transitions in the vortex phase diagram of Ca3Ir4Sn13

    International Nuclear Information System (INIS)

    Kumar, Santosh; Singh, Ravi P.; Thamizhavel, A.; Tomy, C.V.; Grover, A.K.

    2014-01-01

    Highlights: • This work pertains to new findings related to a broad SMP anomaly. • Broad SMP prima facie encompasses two phase transformations in vortex matter. • We demarcated two phase boundaries pertaining to order–disorder transitions which have quasi first-order nature. - Abstract: We present distinct demarcation of the Bragg glass (BG) to multi-domain vortex glass (VG) transition line and the eventual amorphization of the VG phase in a weakly pinned single crystal of the superconducting compound Ca 3 Ir 4 Sn 13 on the basis of comprehension of the different yields about the second magnetization peak (SMP) anomaly in the dc magnetization and the corresponding anomalous feature in the ac susceptibility measurements. The shaking by a small ac magnetic field, inevitably present in the ac susceptibility measurements, is seen to result in contrasting responses in two different portions of the field-temperature (H, T) phase space of the multi-domain VG. In one of the portions, embracing the BG to VG transition across the onset of the SMP anomaly, the ac drive is surprisingly seen to assist the transformation of the well ordered BG phase to a lesser ordered VG phase. The BG phase exists as a superheated state over a small portion of the VG space and this attests to the first order nature of the BG to VG transition

  7. Parameters of DEMO DN and JET DN

    International Nuclear Information System (INIS)

    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.)

  8. Considerations on the DEMO pellet fuelling system

    Energy Technology Data Exchange (ETDEWEB)

    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.

  9. Considerations on the DEMO pellet fuelling system

    International Nuclear Information System (INIS)

    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.

  10. Initial DEMO tokamak design configuration studies

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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.)

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

    Science.gov (United States)

    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,…

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

    International Nuclear Information System (INIS)

    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.)

  14. Objective Provision Tree for K-DEMO

    International Nuclear Information System (INIS)

    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

  15. Constructing a Tibetan Demos in Exile

    DEFF Research Database (Denmark)

    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...

  16. Efficiency of water coolant for DEMO divertor

    International Nuclear Information System (INIS)

    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.

  17. Efficiency of water coolant for DEMO divertor

    Energy Technology Data Exchange (ETDEWEB)

    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.

  18. European blanket development for a demo reactor

    International Nuclear Information System (INIS)

    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

  19. Simulations with COREDIV Code of DEMO Discharges

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  1. Energy Storage System for a Pulsed DEMO

    International Nuclear Information System (INIS)

    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)

  2. Tritium extraction technologies and DEMO requirements

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. DEMO port plug design and integration studies

    Science.gov (United States)

    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

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

    NARCIS (Netherlands)

    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,

  5. Dual role of an ac driving force and the underlying two distinct order–disorder transitions in the vortex phase diagram of Ca{sub 3}Ir{sub 4}Sn{sub 13}

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Santosh, E-mail: santoshkumar@phy.iitb.ac.in [Department of Physics, Indian Institute of Technology Bombay, Mumbai 400076 (India); Singh, Ravi P.; Thamizhavel, A. [Department of Condensed Matter Physics and Materials Science, Tata Institute of Fundamental Research, Mumbai 400005 (India); Tomy, C.V., E-mail: tomy@phy.iitb.ac.in [Department of Physics, Indian Institute of Technology Bombay, Mumbai 400076 (India); Grover, A.K. [Department of Condensed Matter Physics and Materials Science, Tata Institute of Fundamental Research, Mumbai 400005 (India); Department of Physics, Panjab University, Chandigarh 160014 (India)

    2014-11-15

    Highlights: • This work pertains to new findings related to a broad SMP anomaly. • Broad SMP prima facie encompasses two phase transformations in vortex matter. • We demarcated two phase boundaries pertaining to order–disorder transitions which have quasi first-order nature. - Abstract: We present distinct demarcation of the Bragg glass (BG) to multi-domain vortex glass (VG) transition line and the eventual amorphization of the VG phase in a weakly pinned single crystal of the superconducting compound Ca{sub 3}Ir{sub 4}Sn{sub 13} on the basis of comprehension of the different yields about the second magnetization peak (SMP) anomaly in the dc magnetization and the corresponding anomalous feature in the ac susceptibility measurements. The shaking by a small ac magnetic field, inevitably present in the ac susceptibility measurements, is seen to result in contrasting responses in two different portions of the field-temperature (H, T) phase space of the multi-domain VG. In one of the portions, embracing the BG to VG transition across the onset of the SMP anomaly, the ac drive is surprisingly seen to assist the transformation of the well ordered BG phase to a lesser ordered VG phase. The BG phase exists as a superheated state over a small portion of the VG space and this attests to the first order nature of the BG to VG transition.

  6. Energy storage system for a pulsed DEMO

    International Nuclear Information System (INIS)

    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

  7. Neutronics requirements for a DEMO fusion power plant

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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.

  10. Phenomena Identification and Ranking Table for K-DEMO

    International Nuclear Information System (INIS)

    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

  11. Phenomena Identification and Ranking Table for K-DEMO

    Energy Technology Data Exchange (ETDEWEB)

    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.

  12. Japanese endeavors to establish technological bases for DEMO

    Energy Technology Data Exchange (ETDEWEB)

    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.

  13. He-cooled divertor development for DEMO

    International Nuclear Information System (INIS)

    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

  14. European DEMO BOT solid breeder blanket

    International Nuclear Information System (INIS)

    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

  15. DEMO development strategy based on China FPP program

    International Nuclear Information System (INIS)

    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.)

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

    Science.gov (United States)

    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.

  17. Technology and Plasma Physics Developments needed for DEMO

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

  20. Nuclear structure of 231Ac

    International Nuclear Information System (INIS)

    Boutami, R.; Borge, M.J.G.; Mach, H.; Kurcewicz, W.; Fraile, L.M.; Gulda, K.; Aas, A.J.; Garcia-Raffi, L.M.; Lovhoiden, G.; Martinez, T.; Rubio, B.; Tain, J.L.; Tengblad, O.

    2008-01-01

    The low-energy structure of 231 Ac has been investigated by means of γ ray spectroscopy following the β - decay of 231 Ra. Multipolarities of 28 transitions have been established by measuring conversion electrons with a MINI-ORANGE electron spectrometer. The decay scheme of 231 Ra → 231 Ac has been constructed for the first time. The Advanced Time Delayed βγγ(t) method has been used to measure the half-lives of five levels. The moderately fast B(E1) transition rates derived suggest that the octupole effects, albeit weak, are still present in this exotic nucleus

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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.

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

    International Nuclear Information System (INIS)

    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.

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. AC Initiation System.

    Science.gov (United States)

    An ac initiation system is described which uses three ac transmission signals interlocked for safety by frequency, phase, and power discrimination...The ac initiation system is pre-armed by the application of two ac signals have the proper phases, and activates a load when an ac power signal of the proper frequency and power level is applied. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  9. Relevance of NET first wall concept for DEMO DN

    International Nuclear Information System (INIS)

    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)

  10. A preliminary systems assessment of the Starlite Demo candidates

    International Nuclear Information System (INIS)

    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

  11. On tungsten technologies and qualification for DEMO

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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.

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

    International Nuclear Information System (INIS)

    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

  14. Configuration management of the EU DEMO conceptual design data

    International Nuclear Information System (INIS)

    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.

  15. Design and development of ceramic breeder demo blanket

    International Nuclear Information System (INIS)

    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)

  16. Configuration management of the EU DEMO conceptual design data

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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.)

  19. Massively Clustered CubeSats NCPS Demo Mission

    Science.gov (United States)

    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.

  20. Scoping studies for NBI launch geometries on DEMO

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

  2. The predicted 10.6 keV transition in Fr-221 from the alpha-decay of Ac-225 revealed

    Czech Academy of Sciences Publication Activity Database

    Yakushev, E. A.; Chumin, V. G.; Gorozhankin, VM.; Gromov, KY.; Kovalík, Alojz; Filosofov, D. V.; Norseev, YV.; Yurkova, LV.

    2002-01-01

    Roč. 28, č. 3 (2002), s. 463-467 ISSN 0954-3899 R&D Projects: GA ČR GA202/02/0157 Institutional research plan: CEZ:AV0Z1048901 Keywords : electromagnetic transition * electron conversion * alpha-decay Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 1.399, year: 2002

  3. Reel success creating demo reels and animation portfolios

    CERN Document Server

    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

  4. DEMO and fusion power plant conceptual studies in Europe

    International Nuclear Information System (INIS)

    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

  5. RAMI analysis for DEMO HCPB blanket concept cooling system

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  6. LTS and HTS high current conductor development for DEMO

    International Nuclear Information System (INIS)

    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.

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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.

  10. DEMO concepts and their roles within the fusion programme

    International Nuclear Information System (INIS)

    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.)

  11. Objectives and status of EUROfusion DEMO blanket studies

    Energy Technology Data Exchange (ETDEWEB)

    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.

  12. Model improvements for tritium transport in DEMO fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    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.

  13. Peltier ac calorimeter

    OpenAIRE

    Jung, D. H.; Moon, I. K.; Jeong, Y. H.

    2001-01-01

    A new ac calorimeter, utilizing the Peltier effect of a thermocouple junction as an ac power source, is described. This Peltier ac calorimeter allows to measure the absolute value of heat capacity of small solid samples with sub-milligrams of mass. The calorimeter can also be used as a dynamic one with a dynamic range of several decades at low frequencies.

  14. Low Offset AC Correlator.

    Science.gov (United States)

    This patent describes a low offset AC correlator avoids DC offset and low frequency noise by frequency operating the correlation signal so that low...noise, low level AC amplification can be substituted for DC amplification. Subsequently, the high level AC signal is demodulated to a DC level. (Author)

  15. ACAC Converters for UPS

    Directory of Open Access Journals (Sweden)

    Rusalin Lucian R. Păun

    2008-05-01

    Full Text Available This paper propose a new control technique forsingle – phase ACAC converters used for a on-line UPSwith a good dynamic response, a reduced-partscomponents, a good output characteristic, a good powerfactorcorrection(PFC. This converter no needs anisolation transformer. A power factor correction rectifierand an inverter with the proposed control scheme has beendesigned and simulated using Caspoc2007, validating theconcept.

  16. TRANSIT

    Indian Academy of Sciences (India)

    First page Back Continue Last page Overview Graphics. TRANSIT. SYSTEM: DETERMINE 2D-POSITION GLOBALLY BUT INTERMITTENT (POST-FACTO). IMPROVED ACCURACY. PRINCIPLE: POLAR SATELLITES WITH INNOVATIONS OF: GRAVITY-GRADIENT ATTITUDE CONTROL; DRAG COMPENSATION. WORKS ...

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

    International Nuclear Information System (INIS)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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.

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

    International Nuclear Information System (INIS)

    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.

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

    Science.gov (United States)

    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…

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

    Science.gov (United States)

    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…

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

    International Nuclear Information System (INIS)

    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

  4. Constitutional Crowdsourcing to Reconcile Demos with Aristos and Nomos

    DEFF Research Database (Denmark)

    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...

  5. Nuclear structure of {sup 231}Ac

    Energy Technology Data Exchange (ETDEWEB)

    Boutami, R. [Instituto de Estructura de la Materia, CSIC, Serrano 113 bis, E-28006 Madrid (Spain); Borge, M.J.G. [Instituto de Estructura de la Materia, CSIC, Serrano 113 bis, E-28006 Madrid (Spain)], E-mail: borge@iem.cfmac.csic.es; Mach, H. [Department of Radiation Sciences, ISV, Uppsala University, SE-751 21 Uppsala (Sweden); Kurcewicz, W. [Department of Physics, University of Warsaw, Pl-00 681 Warsaw (Poland); Fraile, L.M. [Departamento Fisica Atomica, Molecular y Nuclear, Facultad CC. Fisicas, Universidad Complutense, E-28040 Madrid (Spain); ISOLDE, PH Department, CERN, CH-1211 Geneva 23 (Switzerland); Gulda, K. [Department of Physics, University of Warsaw, Pl-00 681 Warsaw (Poland); Aas, A.J. [Department of Chemistry, University of Oslo, PO Box 1033, Blindern, N-0315 Oslo (Norway); Garcia-Raffi, L.M. [Instituto de Fisica Corpuscular, CSIC - Universidad de Valencia, Apdo. 22805, E-46071 Valencia (Spain); Lovhoiden, G. [Department of Physics, University of Oslo, PO Box 1048, Blindern, N-0316 Oslo (Norway); Martinez, T.; Rubio, B.; Tain, J.L. [Instituto de Fisica Corpuscular, CSIC - Universidad de Valencia, Apdo. 22805, E-46071 Valencia (Spain); Tengblad, O. [Instituto de Estructura de la Materia, CSIC, Serrano 113 bis, E-28006 Madrid (Spain); ISOLDE, PH Department, CERN, CH-1211 Geneva 23 (Switzerland)

    2008-10-15

    The low-energy structure of {sup 231}Ac has been investigated by means of {gamma} ray spectroscopy following the {beta}{sup -} decay of {sup 231}Ra. Multipolarities of 28 transitions have been established by measuring conversion electrons with a MINI-ORANGE electron spectrometer. The decay scheme of {sup 231}Ra {yields}{sup 231}Ac has been constructed for the first time. The Advanced Time Delayed {beta}{gamma}{gamma}(t) method has been used to measure the half-lives of five levels. The moderately fast B(E1) transition rates derived suggest that the octupole effects, albeit weak, are still present in this exotic nucleus.

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

    International Nuclear Information System (INIS)

    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)

  7. An FPGA computing demo core for space charge simulation

    International Nuclear Information System (INIS)

    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.

  8. Enhancing the DEMO divertor target by interlayer engineering

    International Nuclear Information System (INIS)

    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.

  9. Enhancing the DEMO divertor target by interlayer engineering

    Energy Technology Data Exchange (ETDEWEB)

    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}.

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

    Science.gov (United States)

    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.

  11. An FPGA computing demo core for space charge simulation

    Energy Technology Data Exchange (ETDEWEB)

    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.

  12. FLUIDIC AC AMPLIFIERS.

    Science.gov (United States)

    Several fluidic tuned AC Amplifiers were designed and tested. Interstage tuning and feedback designs are considered. Good results were obtained...corresponding Q’s as high as 12. Element designs and test results of one, two, and three stage amplifiers are presented. AC Modulated Carrier Systems

  13. AC power supply systems

    International Nuclear Information System (INIS)

    Law, H.

    1987-01-01

    An ac power supply system includes a rectifier fed by a normal ac supply, and an inverter connected to the rectifier by a dc link, the inverter being effective to invert the dc output of the receiver at a required frequency to provide an ac output. A dc backup power supply of lower voltage than the normal dc output of the rectifier is connected across the dc link such that the ac output of the rectifier is derived from the backup supply if the voltage of the output of the inverter falls below that of the backup supply. The dc backup power may be derived from a backup ac supply. Use in pumping coolant in nuclear reactor is envisaged. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

  17. ACS Zero Point Verification

    Science.gov (United States)

    Dolphin, Andrew

    2005-07-01

    The uncertainties in the photometric zero points create a fundamental limit to the accuracy of photometry. The current state of the ACS calibration is surprisingly poor, with zero point uncertainties of 0.03 magnitudes. The reason for this is that the ACS calibrations are based primarily on semi-emprical synthetic zero points and observations of fields too crowded for accurate ground-based photometry. I propose to remedy this problem by obtaining ACS images of the omega Cen standard field with all nine broadband ACS/WFC filters. This will permit the direct determination of the ACS zero points by comparison with excellent ground-based photometry, and should reduce their uncertainties to less than 0.01 magnitudes. A second benefit is that it will facilitate the comparison of the WFPC2 and ACS photometric systems, which will be important as WFPC2 is phased out and ACS becomes HST's primary imager. Finally, three of the filters will be repeated from my Cycle 12 observations, allowing for a measurement of any change in sensitivity.

  18. First Spaceborne GNSS-Reflectometry Observations of Hurricanes From the UK TechDemoSat-1 Mission

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  3. Design study of fusion Demo plant at JAERI

    International Nuclear Information System (INIS)

    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

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

    DEFF Research Database (Denmark)

    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)....

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  6. AcMNPV

    African Journals Online (AJOL)

    USER

    2010-08-16

    Aug 16, 2010 ... biosynthesis pathway and plays an important role in insect growth and .... Construction and propagation of recombined AcMNPV. The recombined ... infected by virus increased with incubation time (Figure. 3). The growth of ...

  7. AC BREAKDOWN IN GASES

    Science.gov (United States)

    electron- emission (multipactor) region, and (3) the low-frequency region. The breakdown mechanism in each of these regions is explained. An extensive bibliography on AC breakdown in gases is included.

  8. THERMIONIC AC GENERATION

    Science.gov (United States)

    is shown that the maximum ac efficiency is equal to approximately 70% of the corresponding dc value. An illustrative example, including a proposed design for a rather unconventional transformer, is appended. (Author)

  9. Optimization and limitations of known DEMO divertor concepts

    International Nuclear Information System (INIS)

    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.

  10. European DEMO design strategy and consequences for materials

    Science.gov (United States)

    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.

  11. Neutronics experiments for DEMO blanket at JAERI/FNS

    International Nuclear Information System (INIS)

    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)

  12. Neutronics experiments for DEMO blanket at JAERI/FNS

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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.

  15. AC conductivity for a holographic Weyl semimetal

    Energy Technology Data Exchange (ETDEWEB)

    Grignani, Gianluca; Marini, Andrea; Peña-Benitez, Francisco; Speziali, Stefano [Dipartimento di Fisica e Geologia, Università di Perugia,I.N.F.N. Sezione di Perugia,Via Pascoli, I-06123 Perugia (Italy)

    2017-03-23

    We study the AC electrical conductivity at zero temperature in a holographic model for a Weyl semimetal. At small frequencies we observe a linear dependence in the frequency. The model shows a quantum phase transition between a topological semimetal (Weyl semimetal phase) with a non vanishing anomalous Hall conductivity and a trivial semimetal. The AC conductivity has an intermediate scaling due to the presence of a quantum critical region in the phase diagram of the system. The phase diagram is reconstructed using the scaling properties of the conductivity. We compare with the experimental data of https://www.doi.org/10.1103/PhysRevB.93.121110 obtaining qualitative agreement.

  16. Divertor Heat Flux Reduction by Resonant Magnetic Perturbations in the LHD-Type Helical DEMO Reactor

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    International Nuclear Information System (INIS)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. Optimization and limitations of known DEMO divertor concepts

    Energy Technology Data Exchange (ETDEWEB)

    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

  20. Critical Design Factors for Sector Transport Maintenance in DEMO

    Energy Technology Data Exchange (ETDEWEB)

    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

  1. Mass of AC Andromedae

    International Nuclear Information System (INIS)

    King, D.S.; Cox, A.N.; Hodson, S.W.

    1975-01-01

    Calculations indicate that AC Andromedae is population I rather than population II. A mass and radius for this star are calculated using a new set of opacities for the Kippenhahn Ia mixture. It is concluded that the mass is too high for an ordinary RR Lyrae star. (BJG)

  2. AC/RF Superconductivity

    Energy Technology Data Exchange (ETDEWEB)

    Ciovati, G [Jefferson Lab (United States)

    2014-07-01

    This contribution provides a brief introduction to AC/RF superconductivity, with an emphasis on application to accelerators. The topics covered include the surface impedance of normal conductors and superconductors, the residual resistance, the field dependence of the surface resistance, and the superheating field.

  3. AC/RF Superconductivity

    Energy Technology Data Exchange (ETDEWEB)

    Ciovati, Gianluigi [JLAB

    2015-02-01

    This contribution provides a brief introduction to AC/RF superconductivity, with an emphasis on application to accelerators. The topics covered include the surface impedance of normal conductors and superconductors, the residual resistance, the field dependence of the surface resistance, and the superheating field.

  4. Conceptual design of the SlimCS fusion DEMO reactor

    International Nuclear Information System (INIS)

    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)

  5. Pre-conceptual design assessment of DEMO remote maintenance

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

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

    Science.gov (United States)

    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.

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    Science.gov (United States)

    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.

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

    International Nuclear Information System (INIS)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

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

    Science.gov (United States)

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

    2014-08-01

    In ITER diagnostics will operate in the very hard radiation environment of fusion reactor. Extensive technology studies are carried out during development of the ITER diagnostics and procedures of their calibration and remote handling. Results of these studies and practical application of the developed diagnostics on ITER will provide the direct input to DEMO diagnostic development. The list of DEMO measurement requirements and diagnostics will be determined during ITER experiments on the bases of ITER plasma physics results and success of particular diagnostic application in reactor-like ITER plasma. Majority of ITER diagnostic already passed the conceptual design phase and represent the state of the art in fusion plasma diagnostic development. The number of related to DEMO results of ITER diagnostic studies such as design and prototype manufacture of: neutron and γ-ray diagnostics, neutral particle analyzers, optical spectroscopy including first mirror protection and cleaning technics, reflectometry, refractometry, tritium retention measurements etc. are discussed.

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Krasilnikov, A. V.; Kaschuck, Y. A.; Vershkov, V. A.; Petrov, A. A.; Petrov, V. G.; Tugarinov, S. N. [Institution Project center ITER, Moscow (Russian Federation)

    2014-08-21

    In ITER diagnostics will operate in the very hard radiation environment of fusion reactor. Extensive technology studies are carried out during development of the ITER diagnostics and procedures of their calibration and remote handling. Results of these studies and practical application of the developed diagnostics on ITER will provide the direct input to DEMO diagnostic development. The list of DEMO measurement requirements and diagnostics will be determined during ITER experiments on the bases of ITER plasma physics results and success of particular diagnostic application in reactor-like ITER plasma. Majority of ITER diagnostic already passed the conceptual design phase and represent the state of the art in fusion plasma diagnostic development. The number of related to DEMO results of ITER diagnostic studies such as design and prototype manufacture of: neutron and γ–ray diagnostics, neutral particle analyzers, optical spectroscopy including first mirror protection and cleaning technics, reflectometry, refractometry, tritium retention measurements etc. are discussed.

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

    Science.gov (United States)

    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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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.)

  7. Increased Ac excision (iae): Arabidopsis thaliana mutations affecting Ac transposition

    International Nuclear Information System (INIS)

    Jarvis, P.; Belzile, F.; Page, T.; Dean, C.

    1997-01-01

    The maize transposable element Ac is highly active in the heterologous hosts tobacco and tomato, but shows very much reduced levels of activity in Arabidopsis. A mutagenesis experiment was undertaken with the aim of identifying Arabidopsis host factors responsible for the observed low levels of Ac activity. Seed from a line carrying a single copy of the Ac element inserted into the streptomycin phosphotransferase (SPT) reporter fusion, and which displayed typically low levels of Ac activity, were mutagenized using gamma rays. Nineteen mutants displaying high levels of somatic Ac activity, as judged by their highly variegated phenotypes, were isolated after screening the M2 generation on streptomycin-containing medium. The mutations fall into two complementation groups, iae1 and iae2, are unlinked to the SPT::Ac locus and segregate in a Mendelian fashion. The iae1 mutation is recessive and the iae2 mutation is semi-dominant. The iae1 and iae2 mutants show 550- and 70-fold increases, respectively, in the average number of Ac excision sectors per cotyledon. The IAE1 locus maps to chromosome 2, whereas the SPT::Ac reporter maps to chromosome 3. A molecular study of Ac activity in the iae1 mutant confirmed the very high levels of Ac excision predicted using the phenotypic assay, but revealed only low levels of Ac re-insertion. Analyses of germinal transposition in the iae1 mutant demonstrated an average germinal excision frequency of 3% and a frequency of independent Ac re-insertions following germinal excision of 22%. The iae mutants represents a possible means of improving the efficiency of Ac/Ds transposon tagging systems in Arabidopsis, and will enable the dissection of host involvement in Ac transposition and the mechanisms employed for controlling transposable element activity

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Science.gov (United States)

    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.

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  12. Superconducting ac cable

    Science.gov (United States)

    Schmidt, F.

    1980-11-01

    The components of a superconducting 110 kV ac cable for power ratings or = 2000 MVA were developed. The cable design is of the semiflexible type, with a rigid cryogenic envelope containing a flexible hollow coaxial cable core. The cable core consists of spirally wound Nb-A1 composite wires electrically insulated by high pressure polyethylene tape wrappings. A 35 m long single phase test cable with full load terminals rated at 110 kV and 10 kA was constructed and successfully tested. The results obtained prove the technical feasibility and capability of this cable design.

  13. ac superconducting articles

    International Nuclear Information System (INIS)

    Meyerhoff, R.W.

    1977-01-01

    A noval ac superconducting cable is described. It consists of a composite structure having a superconducting surface along with a high thermally conductive material wherein the superconducting surface has the desired physical properties, geometrical shape and surface finish produced by the steps of depositing a superconducting layer upon a substrate having a predetermined surface finish and shape which conforms to that of the desired superconducting article, depositing a supporting layer of material on the superconducting layer and removing the substrate, the surface of the superconductor being a replica of the substrate surface

  14. Superconducting ac cable

    International Nuclear Information System (INIS)

    Schmidt, F.

    1980-01-01

    The components of a superconducting 110 kV ac cable for power ratings >= 2000 MVA have been developed. The cable design especially considered was of the semiflexible type, with a rigid cryogenic envelope and flexible hollow coaxial cable cores pulled into the former. The cable core consists of spirally wound Nb-Al composite wires and a HDPE-tape wrapped electrical insulation. A 35 m long single phase test cable with full load terminations for 110 kV and 10 kA was constructed and successfully tested. The results obtained prove the technical feasibility and capability of our cable design. (orig.) [de

  15. AC Calorimetric Design for Dynamic of Biological Materials

    OpenAIRE

    Shigeo Imaizumi

    2006-01-01

    We developed a new AC calorimeter for the measurement of dynamic specific heat capacity in liquids, including aqueous suspensions of biological materials. This method has several advantages. The first is that a high-resolution measurement of heat capacity, inmillidegrees, can be performed as a function of temperature, even with a very small sample. Therefore, AC calorimeter is a powerful tool to study critical behavior a tphase transition in biological materials. The second advantage is that ...

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    International Nuclear Information System (INIS)

    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

  18. Tritium transport in HCLL and WCLL DEMO blankets

    Energy Technology Data Exchange (ETDEWEB)

    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

  19. ACS Postflash Characterization

    Science.gov (United States)

    Smith, Linda

    2011-10-01

    This program will evaluate the in-flight performance of the ACS/WFC post-flash lamp. A series of observations of Omega Cen will be taken using short and long exposures. The short exposures will be post-flashed using pre-determined exposure times to produce backgrounds from 0 to 125 e-. The data will be used to {1} make an empirical study of the effectiveness in preserving counts for faint stars on various post-flash backgrounds; {2} validate that our current mechanisms for formula-based and pixel-based corrections provide good fixes for whatever CTE remains; and {3} probe a fine enough range of backgrounds that users will be able to pick the level that optimizes their science, which will be a straightforward compromise between the noise added and the signal preserved.

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

    Czech Academy of Sciences Publication Activity Database

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

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

    DEFF Research Database (Denmark)

    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...

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    NARCIS (Netherlands)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  12. Alpha decay 225 Ac → 221Fr

    International Nuclear Information System (INIS)

    Gromov, K. Ya.; Gorozhankin, V.M.; Malov, L.A.; Fominykh, V.I.; Tsupko-Sitnikov, V.V.; Chumin, V.G.; Jakushev, E.A.; Kudrya, S.A.; Sergienko, V.A.; Malikov, Sh.R.

    2004-01-01

    Full text: Considerable attention has been given to nuclei with A = 220 - 230 recently. In this region there occurs transition from the spherical to the deformed nuclear shape, which gives rise to some specific features in the nuclear structure. In particular, negative parity levels with low excitation energies have been found in even-even nuclei from this region [1, 2]. One of the nuclei allowing experimental investigation of the above properties is 221 Fr. The nuclide 221 Fr is from the region of isotopes which does not include stable nuclei and thus it cannot be studied in several-nucleon transfer reactions. In addition, the neutron excess in this nucleus makes it impossible to study the nucleus in reactions with heavy ions. Experimental information on the 221 Fr level structure can only be gained from investigation of the 225 Ac (T 1/2 = 10 days) alpha decay or the 221 Rn (T 1/2 = 25 min) beta decay. In the latter case the possibilities of the investigation are restricted by difficulties in making of 221 Rn sources. Therefore, most information on the structure and properties of 221 Fr is derived from investigation of the 225 Ac α -decay [3]. In-depth investigation of ( α - γ )- coincidences at the 225 Ac decay is carried out. Twenty-one new weak γ - rays are found; 18 γ-rays earlier ascribed to the 225 Ac decay are not confirmed. The quantitative analysis of the ( α - γ )- coincidences makes it possible to find the intensity of 221 Fr levels by the decay and multipolarities of five weak γ -transitions. The conversion electron spectrum is investigated in the range of 5 † 24 keV with a high (some 20 eV) energy resolution. A new M1 type 10.6-keV γ-transition is found. The proposed 225 Ac decay scheme includes 31 excited 221 Fr states. Parities are established for 16 of them. Possible spin values are proposed for 221 Fr levels. Properties of excited 221 Fr states are satisfactorily described by the quasiparticle-phonon nuclear model without the

  13. ITC18: 18th international Toki conference. Development of physics and technology of stellarators/heliotrons 'en route to DEMO'. Proceedings

    International Nuclear Information System (INIS)

    2009-02-01

    18th International Toki Conference (ITC18) was held in Toki (Japan) December 9-12 2008 organized by the National Institute for Fusion Science (NIFS). More than 150 experts in fusion research, especially in stellarator/heliotron research from Australia, Belgium, China, France, Germany, Hungary, India, Iran, Italy, Japan, Korea, Serbia, Spain, Sweden, Switzerland, and the United States of America gathered at the conference. The International Organizing Committee (IOC) chaired by O. Motojima, the International Program Committee (IPC) chaired by Y. Ogawa and the Local Organizing Committee (LOC) chaired by T. Mutoh have played the leading role in the elaboration of the scientific program of the conference. NIFS has organized the ITC as an annual meeting for fusion related sciences since its establishment in 1989. The IPC arranged 2 plenary talks, 1 review talk, 34 invited talks in addition to 109 contributed presentations including 6 oral talks. Recent developments in the experimental, theoretical and technical research show the clear route to the realization of a stellarator/heliotron type demo fusion reactor. ITC18 was devoted to review the recent developments and to discuss the next steps forward to the demo reactor realization of stellarator/heliotron type. In the conference, recent experimental results from both tokamak and stellarator/heliotron devices are reviewed and the experimental and theoretical physics of plasma confinement in toroidal devices are also discussed and confirmed that the physical base of the fusion reactor is well developed. The development of steady state operation, heating, fueling, divertors, plasma wall interaction and wall materials, advanced diagnostics for reactor relevant plasma, blanket materials as well as super conducting magnets are discussed as inevitable key physics and technologies for the DEMO reactor. Slides of all oral presentations as well as the proceedings are available at http://itc.nifs.ac.jp/. Extended papers of major

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Directory of Open Access Journals (Sweden)

    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.

  17. Superconductive AC current limiter

    International Nuclear Information System (INIS)

    Bekhaled, M.

    1987-01-01

    This patent describes an AC current limiter for a power transport line including a power supply circuit and feeding a load circuit via an overload circuit-breaker member. The limiter comprises a transformer having a primary winding connected in series between the power supply circuit and the load circuit and at least one secondary winding of superconductor material contained in a cryogenic enclosure and short-circuited on itself. The leakage reactance of the transformer as seen from the primary winding is low, and the resistance of the at least one secondary winding when in the non-superconducting state and as seen from the primary is much greater than the nominal impedance of the transformer. The improvement whereby the at least one secondary winding of the transformer comprises an active winding in association with a set of auxiliary windings. The set of auxiliary windings is constituted by an even number of series-connected auxiliary windings wound in opposite directions, with the total number of turns in one direction being equal to the total number of turns in the opposite direction, and with the thermal capacity of the secondary winding as a whole being sufficiently high to limit the expansion thereof to a value which remains small during the time it takes the circuit-breaking member to operate

  18. ACS Photometric Zero Point Verification

    Science.gov (United States)

    Dolphin, Andrew

    2003-07-01

    The uncertainties in the photometric zero points create a fundamental limit to the accuracy of photometry. The current state of the ACS calibration is surprisingly poor, with zero point uncertainties of 0.03 magnitudes in the Johnson filters. The reason for this is that ACS observations of excellent ground-based standard fields, such as the omega Cen field used for WFPC2 calibrations, have not been obtained. Instead, the ACS photometric calibrations are based primarily on semi-emprical synthetic zero points and observations of fields too crowded for accurate ground-based photometry. I propose to remedy this problem by obtaining ACS broadband images of the omega Cen standard field with both the WFC and HRC. This will permit the direct determination of the ACS transformations, and is expected to double the accuracy to which the ACS zero points are known. A second benefit is that it will facilitate the comparison of the WFPC2 and ACS photometric systems, which will be important as WFPC2 is phased out and ACS becomes HST's primary imager.

  19. Introduction to AC machine design

    CERN Document Server

    Lipo, Thomas A

    2018-01-01

    AC electrical machine design is a key skill set for developing competitive electric motors and generators for applications in industry, aerospace, and defense. This book presents a thorough treatment of AC machine design, starting from basic electromagnetic principles and continuing through the various design aspects of an induction machine. Introduction to AC Machine Design includes one chapter each on the design of permanent magnet machines, synchronous machines, and thermal design. It also offers a basic treatment of the use of finite elements to compute the magnetic field within a machine without interfering with the initial comprehension of the core subject matter. Based on the author's notes, as well as after years of classroom instruction, Introduction to AC Machine Design: * Brings to light more advanced principles of machine design--not just the basic principles of AC and DC machine behavior * Introduces electrical machine design to neophytes while also being a resource for experienced designers * ...

  20. Sea Ice Detection Based on Differential Delay-Doppler Maps from UK TechDemoSat-1

    Directory of Open Access Journals (Sweden)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    Directory of Open Access Journals (Sweden)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  4. Aislamiento acústico

    Directory of Open Access Journals (Sweden)

    Tobío, J. M.

    1970-07-01

    Full Text Available This is a very specific subject in the field of architectural acoustics, namely, insulation'. Emphasis is placed on the theoretical foundations of this phenomenon, and the most simple formula are developed to calculate easily the transmission losses of a material or the constructional insulating arrangements. The practical aspect of insulation can be considered by means of several graphs and charts, without the use of mathematics, and utilising common materials, that will not substantially increase the cost of the project. Finally this papers offers a critical discussion of building codes, and their reference to the acoustical insulation of dwellings, and data is included on the new regulations of the Madrid Municipality.Se trata un tema muy concreto de la Acústica Arquitectónica, el aislamiento, haciendo hincapié en los fundamentos teóricos del fenómeno y estableciendo las fórmulas más sencillas que permiten calcular fácilmente las pérdidas de transmisión de un material o disposición constructiva aislante. Varias gráficas y abacos permiten abordar, sin ningún tratamiento matemático, el problema práctico del aislamiento, aprovechando los materiales comunes y sin ocasionar gastos que graven sustancialmente el importe del proyecto. Por último, se hace un estudio crítico de las normas y su incidencia en los problemas del aislamiento de viviendas, incluyendo datos referentes a la nueva Ordenanza del Ayuntamiento de Madrid.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    Directory of Open Access Journals (Sweden)

    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.

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

    International Nuclear Information System (INIS)

    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.

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Czech Academy of Sciences Publication Activity Database

    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

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

    Czech Academy of Sciences Publication Activity Database

    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

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

    Czech Academy of Sciences Publication Activity Database

    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

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

    Czech Academy of Sciences Publication Activity Database

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    International Nuclear Information System (INIS)

    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

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

    Directory of Open Access Journals (Sweden)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-15

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

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

    Science.gov (United States)

    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.

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    OpenAIRE

    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...

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  14. Hopping models and ac universality

    DEFF Research Database (Denmark)

    Dyre, Jeppe; Schrøder, Thomas

    2002-01-01

    Some general relations for hopping models are established. We proceed to discuss the universality of the ac conductivity which arises in the extreme disorder limit of the random barrier model. It is shown that the relevant dimension entering into the diffusion cluster approximation (DCA) is the h......Some general relations for hopping models are established. We proceed to discuss the universality of the ac conductivity which arises in the extreme disorder limit of the random barrier model. It is shown that the relevant dimension entering into the diffusion cluster approximation (DCA......) is the harmonic (fracton) dimension of the diffusion cluster. The temperature scaling of the dimensionless frequency entering into the DCA is discussed. Finally, some open problems regarding ac universality are listed....

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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.

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  4. AC ignition of HID lamps

    NARCIS (Netherlands)

    Sobota, A.; Kanters, J.H.M.; Manders, F.; Veldhuizen, van E.M.; Haverlag, M.

    2010-01-01

    Our aim was to examine the starting behaviour of mid-pressure argon discharges in pin-pin (point-to-point) geometry, typically used in HID lamps. We focused our work on AC ignition of 300 and 700 mbar Ar discharges in Philips 70W standard burners. Frequency was varied between 200 kHz and 1 MHz. In

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

    Energy Technology Data Exchange (ETDEWEB)

    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.)

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

    Directory of Open Access Journals (Sweden)

    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...

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

    Directory of Open Access Journals (Sweden)

    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.

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

    International Nuclear Information System (INIS)

    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

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

    Directory of Open Access Journals (Sweden)

    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 ...

  10. AcEST: DK954361 [AcEST

    Lifescience Database Archive (English)

    Full Text Available in 5-4 OS=Homo sap... 33 1.1 sp|Q9DBY1|SYVN1_MOUSE E3 ubiquitin-protein ligase synoviolin OS=... 33 1.4 sp|Q...86TM6|SYVN1_HUMAN E3 ubiquitin-protein ligase synoviolin OS=... 33 1.4 sp|O55188|DMP1_MOUSE Dentin matrix ac

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Directory of Open Access Journals (Sweden)

    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.

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

    Science.gov (United States)

    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.

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

    International Nuclear Information System (INIS)

    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.

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

    International Nuclear Information System (INIS)

    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.

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

    International Nuclear Information System (INIS)

    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%.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Science.gov (United States)

    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.

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

    Directory of Open Access Journals (Sweden)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  8. Aperture measurements with AC dipole

    CERN Document Server

    Fuster Martinez, Nuria; Dilly, Joschua Werner; Nevay, Laurence James; Bruce, Roderik; Tomas Garcia, Rogelio; Redaelli, Stefano; Persson, Tobias Hakan Bjorn; CERN. Geneva. ATS Department

    2018-01-01

    During the MDs performed on the 15th of September and 29th of November 2017, we measured the LHC global aperture at injection with a new AC dipole method as well as using the Transverse Damper (ADT) blow-up method used during the 2017 LHC commissioning for benchmarking. In this note, the MD procedure is presented as well as the analysis of the comparison between the two methods. The possible benefits of the new method are discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  12. Simultaneous distribution of AC and DC power

    Science.gov (United States)

    Polese, Luigi Gentile

    2015-09-15

    A system and method for the transport and distribution of both AC (alternating current) power and DC (direct current) power over wiring infrastructure normally used for distributing AC power only, for example, residential and/or commercial buildings' electrical wires is disclosed and taught. The system and method permits the combining of AC and DC power sources and the simultaneous distribution of the resulting power over the same wiring. At the utilization site a complementary device permits the separation of the DC power from the AC power and their reconstruction, for use in conventional AC-only and DC-only devices.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    Science.gov (United States)

    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

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

    International Nuclear Information System (INIS)

    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.

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

    CERN Document Server

    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.

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

    DEFF Research Database (Denmark)

    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....

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

    International Nuclear Information System (INIS)

    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 )

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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.)

  2. Performance of AC/graphite capacitors at high weight ratios of AC/graphite

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Hongyu [IM and T Ltd., Advanced Research Center, Saga University, 1341 Yoga-machi, Saga 840-0047 (Japan); Yoshio, Masaki [Advanced Research Center, Department of Applied Chemistry, Saga University, 1341 Yoga-machi, Saga 840-0047 (Japan)

    2008-03-01

    The effect of negative to positive electrode materials' weight ratio on the electrochemical performance of both activated carbon (AC)/AC and AC/graphite capacitors has been investigated, especially in the terms of capacity and cycle-ability. The limited capacity charge mode has been proposed to improve the cycle performance of AC/graphite capacitors at high weight ratios of AC/graphite. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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).

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    Directory of Open Access Journals (Sweden)

    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.

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  15. Nuclear structure effects in the exotic decay of $^{225}$Ac via $^{14}$C emission

    CERN Document Server

    Bonetti, R; Guglielmetti, A; Matheoud, R; Migliorino, C; Pasinetti, A L; Ravn, H L

    1993-01-01

    By using a $^{225}$Ac source produced at the electromagnetic separator Isolde we collected on our track-recording glass detectors 305 $^{14}$C events from the radioactive decays of $^{225}$Ac and its daughter $^{221}$Fr and obtained, for $^{225}$Ac, a branching ratio B($^{14}$C/$\\alpha$)=(6.0 $\\pm$ 1.3) x 10$^{-12}$. Our result suggests that such a decay from an odd proton nucleus is dominated by transition to the ground or to the first excited state of daughter nucleus.

  16. SNS AC Power Distribution and Reliability of AC Power Supply

    CERN Document Server

    Holik, Paul S

    2005-01-01

    The SNS Project has 45MW of installed power. A design description under the Construction Design and Maintenance (CDM) with regard to regulations (OSHA, NFPA, NEC), reliability issues and maintenance of the AC power distribution system are herewith presented. The SNS Project has 45MW of installed power. The Accelerator Systems are Front End (FE)and LINAC KLYSTRON Building (LK), Central Helium Liquefier (CHL), High Energy Beam Transport (HEBT), Accumulator Ring and Ring to Target Beam Transport (RTBT) Support Buildings have 30MW installed power. FELK has 16MW installed, majority of which is klystron and magnet power supply system. CHL, supporting the super conducting portion of the accelerator has 7MW installed power and the RING Systems (HEBT, RING and RTBT) have also 7MW installed power.*

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

    International Nuclear Information System (INIS)

    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)

  18. Universality of ac conduction in disordered solids

    DEFF Research Database (Denmark)

    Dyre, Jeppe; Schrøder, Thomas

    2000-01-01

    The striking similarity of ac conduction in quite different disordered solids is discussed in terms of experimental results, modeling, and computer simulations. After giving an overview of experiment, a macroscopic and a microscopic model are reviewed. For both models the normalized ac conductivity...... as a function of a suitably scaled frequency becomes independent of details of the disorder in the extreme disorder limit, i.e., when the local randomly varying mobilities cover many orders of magnitude. The two universal ac conductivities are similar, but not identical; both are examples of unusual non......-power-law universalities. It is argued that ac universality reflects an underlying percolation determining dc as well as ac conductivity in the extreme disorder limit. Three analytical approximations to the universal ac conductivities are presented and compared to computer simulations. Finally, model predictions...

  19. The AC photovoltaic module is here!

    Science.gov (United States)

    Strong, Steven J.; Wohlgemuth, John H.; Wills, Robert H.

    1997-02-01

    This paper describes the design, development, and performance results of a large-area photovoltaic module whose electrical output is ac power suitable for direct connection to the utility grid. The large-area ac PV module features a dedicated, integrally mounted, high-efficiency dc-to-ac power inverter with a nominal output of 250 watts (STC) at 120 Vac, 60 H, that is fully compatible with utility power. The module's output is connected directly to the building's conventional ac distribution system without need for any dc wiring, string combiners, dc ground-fault protection or additional power-conditioning equipment. With its advantages, the ac photovoltaic module promises to become a universal building block for use in all utility-interactive PV systems. This paper discusses AC Module design aspects and utility interface issues (including islanding).

  20. Study on ac losses of HTS coil carrying ac transport current

    International Nuclear Information System (INIS)

    Dai Taozhen; Tang Yuejin; Li Jingdong; Zhou Yusheng; Cheng Shijie; Pan Yuan

    2005-01-01

    Ac loss has an important influence on the thermal performances of HTS coil. It is necessary to quantify ac loss to ascertain its impact on coil stability and for sizing the coil refrigeration system. In this paper, we analyzed in detail the ac loss components, hysteresis loss, eddy loss and flux flow loss in the pancake HTS coil carrying ac transport current by finite element method. We also investigated the distribution of the ac losses in the coil to study the effects of magnetic field distribution on ac losses

  1. RHIC spin flipper AC dipole controller

    Energy Technology Data Exchange (ETDEWEB)

    Oddo, P.; Bai, M.; Dawson, C.; Gassner, D.; Harvey, M.; Hayes, T.; Mernick, K.; Minty, M.; Roser, T.; Severino, F.; Smith, K.

    2011-03-28

    The RHIC Spin Flipper's five high-Q AC dipoles which are driven by a swept frequency waveform require precise control of phase and amplitude during the sweep. This control is achieved using FPGA based feedback controllers. Multiple feedback loops are used to and dynamically tune the magnets. The current implementation and results will be presented. Work on a new spin flipper for RHIC (Relativistic Heavy Ion Collider) incorporating multiple dynamically tuned high-Q AC-dipoles has been developed for RHIC spin-physics experiments. A spin flipper is needed to cancel systematic errors by reversing the spin direction of the two colliding beams multiple times during a store. The spin flipper system consists of four DC-dipole magnets (spin rotators) and five AC-dipole magnets. Multiple AC-dipoles are needed to localize the driven coherent betatron oscillation inside the spin flipper. Operationally the AC-dipoles form two swept frequency bumps that minimize the effect of the AC-dipole dipoles outside of the spin flipper. Both AC bumps operate at the same frequency, but are phase shifted from each other. The AC-dipoles therefore require precise control over amplitude and phase making the implementation of the AC-dipole controller the central challenge.

  2. Lamin A/C might be involved in the EMT signalling pathway.

    Science.gov (United States)

    Zuo, Lingkun; Zhao, Huanying; Yang, Ronghui; Wang, Liyong; Ma, Hui; Xu, Xiaoxue; Zhou, Ping; Kong, Lu

    2018-07-15

    We have previously reported a heterogeneous expression pattern of the nuclear membrane protein lamin A/C in low- and high-Gleason score (GS) prostate cancer (PC) tissues, and we have now found that this change is not associated with LMNA mutations. This expression pattern appears to be similar to the process of epithelial to mesenchymal transition (EMT) or to that of mesenchymal to epithelial transition (MET). The role of lamin A/C in EMT or MET in PC remains unclear. Therefore, we first investigated the expression levels of and the associations between lamin A/C and several common EMT markers, such as E-cadherin, N-cadherin, β-catenin, snail, slug and vimentin in PC tissues with different GS values and in different cell lines with varying invasion abilities. Our results suggest that lamin A/C might constitute a type of epithelial marker that better signifies EMT and MET in PC tissue, since a decrease in lamin A/C expression in GS 4 + 5 cases is likely associated with the EMT process, while the re-expression of lamin A/C in GS 5 + 4 cases is likely linked with MET. The detailed GS better exhibited the changes in lamin A/C and the EMT markers examined. Lamin A/C overexpression or knockdown had an impact on EMT biomarkers in a cell model by direct regulation of β-catenin. Hence, we suggest that lamin A/C might serve as a reliable epithelial biomarker for the distinction of PC cell differentiation and might also be a fundamental factor in the occurrence of EMT or MET in PC. Copyright © 2018. Published by Elsevier B.V.

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

    International Nuclear Information System (INIS)

    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

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

    Directory of Open Access Journals (Sweden)

    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.

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

    CERN Document Server

    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.

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

    OpenAIRE

    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...

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-02-15

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

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  1. Bioinformatics and Astrophysics Cluster (BinAc)

    Science.gov (United States)

    Krüger, Jens; Lutz, Volker; Bartusch, Felix; Dilling, Werner; Gorska, Anna; Schäfer, Christoph; Walter, Thomas

    2017-09-01

    BinAC provides central high performance computing capacities for bioinformaticians and astrophysicists from the state of Baden-Württemberg. The bwForCluster BinAC is part of the implementation concept for scientific computing for the universities in Baden-Württemberg. Community specific support is offered through the bwHPC-C5 project.

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

    International Nuclear Information System (INIS)

    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)

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

    Directory of Open Access Journals (Sweden)

    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.

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

    International Nuclear Information System (INIS)

    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.

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

    International Nuclear Information System (INIS)

    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.)

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Directory of Open Access Journals (Sweden)

    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.

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

    International Nuclear Information System (INIS)

    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.)

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

    Science.gov (United States)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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.

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

    Science.gov (United States)

    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.

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

    Science.gov (United States)

    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.

  16. Assessing the allelotypic effect of two aminocyclopropane carboxylic acid synthase-encoding genes MdACS1 and MdACS3a on fruit ethylene production and softening in Malus

    Science.gov (United States)

    Dougherty, Laura; Zhu, Yuandi; Xu, Kenong

    2016-01-01

    Phytohormone ethylene largely determines apple fruit shelf life and storability. Previous studies demonstrated that MdACS1 and MdACS3a, which encode 1-aminocyclopropane-1-carboxylic acid synthases (ACS), are crucial in apple fruit ethylene production. MdACS1 is well-known to be intimately involved in the climacteric ethylene burst in fruit ripening, while MdACS3a has been regarded a main regulator for ethylene production transition from system 1 (during fruit development) to system 2 (during fruit ripening). However, MdACS3a was also shown to have limited roles in initiating the ripening process lately. To better assess their roles, fruit ethylene production and softening were evaluated at five time points during a 20-day post-harvest period in 97 Malus accessions and in 34 progeny from 2 controlled crosses. Allelotyping was accomplished using an existing marker (ACS1) for MdACS1 and two markers (CAPS866 and CAPS870) developed here to specifically detect the two null alleles (ACS3a-G289V and Mdacs3a) of MdACS3a. In total, 952 Malus accessions were allelotyped with the three markers. The major findings included: The effect of MdACS1 was significant on fruit ethylene production and softening while that of MdACS3a was less detectable; allele MdACS1–2 was significantly associated with low ethylene and slow softening; under the same background of the MdACS1 allelotypes, null allele Mdacs3a (not ACS3a-G289V) could confer a significant delay of ethylene peak; alleles MdACS1–2 and Mdacs3a (excluding ACS3a-G289V) were highly enriched in M. domestica and M. hybrid when compared with those in M. sieversii. These findings are of practical implications in developing apples of low and delayed ethylene profiles by utilizing the beneficial alleles MdACS1-2 and Mdacs3a. PMID:27231553

  17. DIAGNOSTIC FEATURES RESEARCH OF AC ELECTRIC POINT MOTORS

    Directory of Open Access Journals (Sweden)

    S. YU. Buryak

    2014-05-01

    Full Text Available Purpose.Considerable responsibility for safety of operation rests on signal telephone and telegraph department of railway. One of the most attackable nodes (both automation systems, and railway in whole is track switches. The aim of this investigation is developing such system for monitoring and diagnostics of track switches, which would fully meet the requirements of modern conditions of high-speed motion and heavy trains and producing diagnostics, collection and systematization of data in an automated way. Methodology. In order to achieve the desired objectives research of a structure and the operating principle description of the switch electric drive, sequence of triggering its main units were carried out. The operating characteristics and settings, operating conditions, the causes of failures in the work, andrequirements for electric drives technology and their service were considered and analyzed. Basic analysis principles of dependence of nature of the changes the current waveform, which flows in the working circuit of AC electric point motor were determined. Technical implementation of the monitoring and diagnosing system the state of AC electric point motors was carried out. Findings. Signals taken from serviceable and defective electric turnouts were researched. Originality. Identified a strong interconnectionbetween the technical condition of the track switchand curve shape that describes the current in the circuit of AC electric point motor during operation which is based on the research processes that have influence on it during operation. Practical value. Shown the principles of the technical approach to the transition from scheduled preventive maintenance to maintenance of real condition for a more objective assessment and thus more rapid response to emerging or failures when they occur gradually, damages and any other shortcomings in the work track switch AC drives.

  18. 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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  19. 78 FR 49318 - Availability of Draft Advisory Circular (AC) 90-106A and AC 20-167A

    Science.gov (United States)

    2013-08-13

    ...] Availability of Draft Advisory Circular (AC) 90-106A and AC 20- 167A AGENCY: Federal Aviation Administration... of draft Advisory Circular (AC) 90-106A, Enhanced Flight Vision Systems and draft AC 20- 167A... Federal holidays. FOR FURTHER INFORMATION CONTACT: For technical questions concerning draft AC 90-106A...

  20. AC distribution system for TFTR pulsed loads

    International Nuclear Information System (INIS)

    Carroll, R.F.; Ramakrishnan, S.; Lemmon, G.N.; Moo, W.I.

    1977-01-01

    This paper outlines the AC distribution system associated with the Tokamak Fusion Test Reactor and discusses the significant areas related to design, protection, and equipment selection, particularly where there is a departure from normal utility and industrial applications

  1. Nonlinear AC susceptibility, surface and bulk shielding

    Science.gov (United States)

    van der Beek, C. J.; Indenbom, M. V.; D'Anna, G.; Benoit, W.

    1996-02-01

    We calculate the nonlinear AC response of a thin superconducting strip in perpendicular field, shielded by an edge current due to the geometrical barrier. A comparison with the results for infinite samples in parallel field, screened by a surface barrier, and with those for screening by a bulk current in the critical state, shows that the AC response due to a barrier has general features that are independent of geometry, and that are significantly different from those for screening by a bulk current in the critical state. By consequence, the nonlinear (global) AC susceptibility can be used to determine the origin of magnetic irreversibility. A comparison with experiments on a Bi 2Sr 2CaCu 2O 8+δ crystal shows that in this material, the low-frequency AC screening at high temperature is mainly due to the screening by an edge current, and that this is the unique source of the nonlinear magnetic response at temperatures above 40 K.

  2. Logistics Reduction: Advanced Clothing System (ACS)

    Data.gov (United States)

    National Aeronautics and Space Administration — The goal of the Advanced Exploration System (AES) Logistics Reduction (LR) project's Advanced Clothing System (ACS) is to use advanced commercial off-the-shelf...

  3. Marketingová komunikace AC Sparta Praha

    OpenAIRE

    Fanta, Jan

    2016-01-01

    Title: Marketing communications of AC Sparta Praha Objectives: The main objective of this thesis is to analyze contemporary state of marketing communications with the audience of AC Sparta Praha, identify deficiencies and develop a proposal to improve the marketing communications with fans of this club. Methods: In this thesis have been used methods of case study, analysis of available documents and texts, structured interview with director od marketing, and director of communications and pub...

  4. Cooperative Frequency Control for Autonomous AC Microgrids

    DEFF Research Database (Denmark)

    Shafiee, Qobad; Quintero, Juan Carlos Vasquez; Guerrero, Josep M.

    2015-01-01

    Distributed secondary control strategies have been recently studied for frequency regulation in droop-based AC Microgrids. Unlike centralized secondary control, the distributed one might fail to provide frequency synchronization and proportional active power sharing simultaneously, due to having...... not require measuring the system frequency as compared to the other presented methods. An ac Microgrid with four sources is used to verify the performance of the proposed control methodology....

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  8. Safety issues related to the intermediate heat storage for the EU DEMO

    Energy Technology Data Exchange (ETDEWEB)

    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.

  9. Transport AC losses in YBCO coated conductors

    Energy Technology Data Exchange (ETDEWEB)

    Majoros, M [Ohio State University, Columbus, OH 43210 (United States); Ye, L [IRC in Superconductivity, University of Cambridge, Madingley Road, Cambridge CB3 0HE (United Kingdom); Velichko, A V [IRC in Superconductivity, University of Cambridge, Madingley Road, Cambridge CB3 0HE (United Kingdom); Coombs, T A [IRC in Superconductivity, University of Cambridge, Madingley Road, Cambridge CB3 0HE (United Kingdom); Sumption, M D [Ohio State University, Columbus, OH 43210 (United States); Collings, E W [Ohio State University, Columbus, OH 43210 (United States)

    2007-09-15

    Transport AC loss measurements have been made on YBCO-coated conductors prepared on two different substrate templates-RABiTS (rolling-assisted biaxially textured substrate) and IBAD (ion-beam-assisted deposition). RABiTS samples show higher losses compared with the theoretical values obtained from the critical state model, with constant critical current density, at currents lower than the critical current. An origin of this extra AC loss was demonstrated experimentally by comparison of the AC loss of two samples with different I-V curves. Despite a difference in I-V curves and in the critical currents, their measured losses, as well as the normalized losses, were practically the same. However, the functional dependence of the losses was affected by the ferromagnetic substrate. An influence of the presence of a ferromagnetic substrate on transport AC losses in YBCO film was calculated numerically by the finite element method. The presence of a ferromagnetic substrate increases transport AC losses in YBCO films depending on its relative magnetic permeability. The two loss contributions-transport AC loss in YBCO films and ferromagnetic loss in the substrate-cannot be considered as mutually independent.

  10. Proportional-Integral-Resonant AC Current Controller

    Directory of Open Access Journals (Sweden)

    STOJIC, D.

    2017-02-01

    Full Text Available In this paper an improved stationary-frame AC current controller based on the proportional-integral-resonant control action (PIR is proposed. Namely, the novel two-parameter PIR controller is applied in the stationary-frame AC current control, accompanied by the corresponding parameter-tuning procedure. In this way, the proportional-resonant (PR controller, common in the stationary-frame AC current control, is extended by the integral (I action in order to enable the AC current DC component tracking, and, also, to enable the DC disturbance compensation, caused by the voltage source inverter (VSI nonidealities and by nonlinear loads. The proposed controller parameter-tuning procedure is based on the three-phase back-EMF-type load, which corresponds to a wide range of AC power converter applications, such as AC motor drives, uninterruptible power supplies, and active filters. While the PIR controllers commonly have three parameters, the novel controller has two. Also, the provided parameter-tuning procedure needs only one parameter to be tuned in relation to the load and power converter model parameters, since the second controller parameter is directly derived from the required controller bandwidth value. The dynamic performance of the proposed controller is verified by means of simulation and experimental runs.

  11. Optimization of the breeder zone cooling tubes of the DEMO Water-Cooled Lithium Lead breeding blanket

    Energy Technology Data Exchange (ETDEWEB)

    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

  12. The Effect of the Feedback Controller on Superconducting Tokamak AC Losses + AC-CRPP user manual

    International Nuclear Information System (INIS)

    Schaerz, B.; Bruzzone, P.; Favez, J.Y.; Lister, J.B.; Zapretilina, E.

    2001-11-01

    Superconducting coils in a Tokamak are subject to AC losses when the field transverse to the coil current varies. A simple model to evaluate the AC losses has been derived and benchmarked against a complete model used in the ITER design procedure. The influence of the feedback control strategy on the AC losses is examined using this model. An improved controller is proposed, based on this study. (author)

  13. Comparison study on neutronic analysis of the K-DEMO water cooled ceramic breeder blanket using MCNP and ATTILA

    Energy Technology Data Exchange (ETDEWEB)

    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

  14. Consequences of the technology survey and gap analysis on the EU DEMO R&D programme in tritium, matter injection and vacuum

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. The Ethylene Biosynthesis Gene CitACS4 Regulates Monoecy/Andromonoecy in Watermelon (Citrullus lanatus).

    Science.gov (United States)

    Manzano, Susana; Aguado, Encarnación; Martínez, Cecilia; Megías, Zoraida; García, Alicia; Jamilena, Manuel

    2016-01-01

    Monoecious and andromonoecious cultivars of watermelon are characterised by the production of male and female flower or male and hermaphrodite flowers, respectively. The segregation analysis in the offspring of crosses between monoecious and andromonoecious lines has demonstrated that this trait is controlled by a single gene pair, being the monoecious allele M semi-dominant to the andromonoecious allele A. The two studied F1 hybrids (MA) had a predominantly monoecious phenotype since both produced not only female flowers, but also bisexual flowers with incomplete stamens, and hermaphrodite flowers with pollen. Given that in other cucurbit species andromonoecy is conferred by mutations in the ethylene biosynthesis genes CmACS7, CsACS2 and CpACS27A we have cloned and characterised CitACS4, the watermelon gene showing the highest similarity with the formers. CitACS4 encoded for a type ACS type III enzyme that is predominantly expressed in pistillate flowers of watermelon. In the andromonoecious line we have detected a missense mutation in a very conserved residue of CitACS4 (C364W) that cosegregates with the andromonoecious phenotype in two independent F2 populations, concomitantly with a reduction in ethylene production in the floral buds that will develop as hermaphrodite flowers. The gene does not however co-segregates with other sex expression traits regulated by ethylene in this species, including pistillate flowering transition and the number of pistillate flowers per plant. These data indicate that CitAC4 is likely to be involved in the biosynthesis of the ethylene required for stamen arrest during the development of female flowers. The C364W mutation would reduce the production of ethylene in pistillate floral buds, promoting the conversion of female into hermaphrodite flowers, and therefore of monoecy into andromonoecy.

  16. Design and synthesis of 225Ac radioimmunopharmaceuticals

    International Nuclear Information System (INIS)

    McDevitt, Michael R.; Ma, Dangshe; Simon, Jim; Frank, R. Keith; Scheinberg, David A.

    2002-01-01

    The alpha-particle-emitting radionuclides 213 Bi, 211 At, 224 Ra are under investigation for the treatment of leukemias, gliomas, and ankylosing spondylitis, respectively. 213 Bi and 211 At were attached to monoclonal antibodies and used as targeted immunotherapeutic agents while unconjugated 224 Ra chloride selectively seeks bone. 225 Ac possesses favorable physical properties for radioimmunotherapy (10 d half-life and 4 net alpha particles), but has a history of unfavorable radiolabeling chemistry and poor metal-chelate stability. We selected functionalized derivatives of DOTA as the most promising to pursue from out of a group of potential 225 Ac chelate compounds. A two-step synthetic process employing either MeO-DOTA-NCS or 2B-DOTA-NCS as the chelating moiety was developed to attach 225 Ac to monoclonal antibodies. This method was tested using several different IgG systems. The chelation reaction yield in the first step was 93±8% radiochemically pure (n=26). The second step yielded 225 Ac-DOTA-IgG constructs that were 95±5% radiochemically pure (n=27) and the mean percent immunoreactivity ranged from 25% to 81%, depending on the antibody used. This process has yielded several potential novel targeted 225 Ac-labeled immunotherapeutic agents that may now be evaluated in appropriate model systems and ultimately in humans

  17. Approaches to building single-stage AC/AC conversion switch-mode audio power amplifiers

    DEFF Research Database (Denmark)

    Ljusev, Petar; Andersen, Michael Andreas E.

    2004-01-01

    This paper discusses the possible topologies and promising approaches towards direct single-phase AC-AC conversion of the mains voltage for audio applications. When compared to standard Class-D switching audio power amplifiers with a separate power supply, it is expected that direct conversion...

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. Progress in the engineering design and assessment of the European DEMO first wall and divertor plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    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.

  20. Thermal-hydraulic analysis of water cooled breeding blanket of K-DEMO using MARS-KS code

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.)

  2. Neutronics Design of Helical Type DEMO Reactor FFHR-d1

    Energy Technology Data Exchange (ETDEWEB)

    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

  3. Current design of the European TBM systems and implications on DEMO breeding blanket

    Energy Technology Data Exchange (ETDEWEB)

    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.

  4. Structural integrity for DEMO: An opportunity to close the gap from materials science to engineering needs

    Energy Technology Data Exchange (ETDEWEB)

    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

  5. Pressurizing Behavior on Ingress of Coolant into Pebble Bed of Blanket of Fusion DEMO Reactor

    International Nuclear Information System (INIS)

    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

  6. Adaptation of the HCPB DEMO TBM as breeding blanket for ITER : Neutronic and thermal analyses

    International Nuclear Information System (INIS)

    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

  7. ac propulsion system for an electric vehicle

    Science.gov (United States)

    Geppert, S.

    1980-01-01

    It is pointed out that dc drives will be the logical choice for current production electric vehicles (EV). However, by the mid-80's, there is a good chance that the price and reliability of suitable high-power semiconductors will allow for a competitive ac system. The driving force behind the ac approach is the induction motor, which has specific advantages relative to a dc shunt or series traction motor. These advantages would be an important factor in the case of a vehicle for which low maintenance characteristics are of primary importance. A description of an EV ac propulsion system is provided, taking into account the logic controller, the inverter, the motor, and a two-speed transmission-differential-axle assembly. The main barrier to the employment of the considered propulsion system in EV is not any technical problem, but inverter transistor cost.

  8. Superconducting three element synchronous ac machine

    International Nuclear Information System (INIS)

    Boyer, L.; Chabrerie, J.P.; Mailfert, A.; Renard, M.

    1975-01-01

    There is a growing interest in ac superconducting machines. Of several new concepts proposed for these machines in the last years one of the most promising seems to be the ''three elements'' concept which allows the cancellation of the torque acting on the superconducting field winding, thus overcoming some of the major contraints. This concept leads to a device of induction-type generator. A synchronous, three element superconducting ac machine is described, in which a room temperature, dc fed rotating winding is inserted between the superconducting field winding and the ac armature. The steady-state machine theory is developed, the flux linkages are established, and the torque expressions are derived. The condition for zero torque on the field winding, as well as the resulting electrical equations of the machine, are given. The theoretical behavior of the machine is studied, using phasor diagrams and assuming for the superconducting field winding either a constant current or a constant flux condition

  9. 21 CFR 886.4440 - AC-powered magnet.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false AC-powered magnet. 886.4440 Section 886.4440 Food... DEVICES OPHTHALMIC DEVICES Surgical Devices § 886.4440 AC-powered magnet. (a) Identification. An AC-powered magnet is an AC-powered device that generates a magnetic field intended to find and remove...

  10. Mapa acústico parcial de Benetusser

    OpenAIRE

    MORILLA CASTELLANOS, EMILIO

    2012-01-01

    Se establece el mapa de ruido del municipio de Benetússer para evaluar y conocer su exposición al ruido ambiental y así poder dar cumplimiento a la Directiva Europea sobre Gestión y Evaluación de Ruido Ambiental (2002/49/CE) y a la Ley nacional 37/2003 del Ruido. Los mapas estratégicos de ruido nos aportan la información fundamental para diagnosticar la situación acústica y para la gestión del ruido ambiental. Morilla Castellanos, E. (2012). Mapa acústico parcial de Benetusser. http://h...

  11. Preliminary study on AC superconducting machines

    International Nuclear Information System (INIS)

    Yamamoto, M.; Ishigohka, T.; Shimohka, T.; Mizukami, N.; Yamaguchi, M.

    1988-01-01

    This paper describes the issues involved in developing AC superconducting machines. In the first phase, as a preliminary experiment, a 4kVa AC superconducting coil which employs 100A class 50/60Hz superconductors is made and tested. And, in the second phase, as an extension of the 4kVa coil, a model superconducting transformer is made and examined. The transformer has a novel quench protection system with an auxiliary coil only in the low voltage side. The behavior of the overcurrent protection system is confirmed

  12. Control of Power Converters in AC Microgrids

    DEFF Research Database (Denmark)

    Rocabert, Joan; Luna, Alvaro; Blaabjerg, Frede

    2012-01-01

    The enabling of ac microgrids in distribution networks allows delivering distributed power and providing grid support services during regular operation of the grid, as well as powering isolated islands in case of faults and contingencies, thus increasing the performance and reliability of the ele......The enabling of ac microgrids in distribution networks allows delivering distributed power and providing grid support services during regular operation of the grid, as well as powering isolated islands in case of faults and contingencies, thus increasing the performance and reliability...

  13. Statistical time lags in ac discharges

    International Nuclear Information System (INIS)

    Sobota, A; Kanters, J H M; Van Veldhuizen, E M; Haverlag, M; Manders, F

    2011-01-01

    The paper presents statistical time lags measured for breakdown events in near-atmospheric pressure argon and xenon. Ac voltage at 100, 400 and 800 kHz was used to drive the breakdown processes, and the voltage amplitude slope was varied between 10 and 1280 V ms -1 . The values obtained for the statistical time lags are roughly between 1 and 150 ms. It is shown that the statistical time lags in ac-driven discharges follow the same general trends as the discharges driven by voltage of monotonic slope. In addition, the validity of the Cobine-Easton expression is tested at an alternating voltage form.

  14. Statistical time lags in ac discharges

    Energy Technology Data Exchange (ETDEWEB)

    Sobota, A; Kanters, J H M; Van Veldhuizen, E M; Haverlag, M [Eindhoven University of Technology, Department of Applied Physics, Postbus 513, 5600MB Eindhoven (Netherlands); Manders, F, E-mail: a.sobota@tue.nl [Philips Lighting, LightLabs, Mathildelaan 1, 5600JM Eindhoven (Netherlands)

    2011-04-06

    The paper presents statistical time lags measured for breakdown events in near-atmospheric pressure argon and xenon. Ac voltage at 100, 400 and 800 kHz was used to drive the breakdown processes, and the voltage amplitude slope was varied between 10 and 1280 V ms{sup -1}. The values obtained for the statistical time lags are roughly between 1 and 150 ms. It is shown that the statistical time lags in ac-driven discharges follow the same general trends as the discharges driven by voltage of monotonic slope. In addition, the validity of the Cobine-Easton expression is tested at an alternating voltage form.

  15. Magnetic images of the disintegration process of tablets in the human stomach by ac biosusceptometry

    International Nuclear Information System (INIS)

    Cora, L A; Andreis, U; Romeiro, F G; Americo, M F; Oliveira, R B; Baffa, O; Miranda, J R A

    2005-01-01

    Oral administration of solid dosage forms is usually preferred in drug therapy. Conventional imaging methods are essential tools to investigate the in vivo performance of these formulations. The non-invasive technique of ac biosusceptometry has been introduced as an alternative in studies focusing on gastrointestinal motility and, more recently, to evaluate the behaviour of magnetic tablets in vivo. The aim of this work was to employ a multisensor ac biosusceptometer system to obtain magnetic images of disintegration of tablets in vitro and in the human stomach. The results showed that the transition between the magnetic marker and the magnetic tracer characterized the onset of disintegration (t 50 ) and occurred in a short time interval (1.1 ± 0.4 min). The multisensor ac biosusceptometer was reliable to monitor and analyse the in vivo performance of magnetic tablets showing accuracy to quantify disintegration through the magnetic images and to characterize the profile of this process

  16. Magnetic images of the disintegration process of tablets in the human stomach by ac biosusceptometry

    Energy Technology Data Exchange (ETDEWEB)

    Cora, L A [Departamento de Fisica e BioFisica, IBB, UNESP, Botucatu, SP (Brazil); Andreis, U [Departamento de Fisica e BioFisica, IBB, UNESP, Botucatu, SP (Brazil); Romeiro, F G [Departamento de ClInica Medica, FMB, UNESP, Botucatu, SP (Brazil); Americo, M F [Departamento de ClInica Medica, FMRP, USP, Ribeirao Preto, SP (Brazil); Oliveira, R B [Departamento de ClInica Medica, FMRP, USP, Ribeirao Preto, SP (Brazil); Baffa, O [Departamento de Fisica e Matematica, FFCLRP, USP, Ribeirao Preto, SP (Brazil); Miranda, J R A [Departamento de Fisica e BioFisica, IBB, UNESP, Botucatu, SP (Brazil)

    2005-12-07

    Oral administration of solid dosage forms is usually preferred in drug therapy. Conventional imaging methods are essential tools to investigate the in vivo performance of these formulations. The non-invasive technique of ac biosusceptometry has been introduced as an alternative in studies focusing on gastrointestinal motility and, more recently, to evaluate the behaviour of magnetic tablets in vivo. The aim of this work was to employ a multisensor ac biosusceptometer system to obtain magnetic images of disintegration of tablets in vitro and in the human stomach. The results showed that the transition between the magnetic marker and the magnetic tracer characterized the onset of disintegration (t{sub 50}) and occurred in a short time interval (1.1 {+-} 0.4 min). The multisensor ac biosusceptometer was reliable to monitor and analyse the in vivo performance of magnetic tablets showing accuracy to quantify disintegration through the magnetic images and to characterize the profile of this process.

  17. Magnetic images of the disintegration process of tablets in the human stomach by ac biosusceptometry

    Science.gov (United States)

    Corá, L. A.; Andreis, U.; Romeiro, F. G.; Américo, M. F.; Oliveira, R. B.; Baffa, O.; Miranda, J. R. A.

    2005-12-01

    Oral administration of solid dosage forms is usually preferred in drug therapy. Conventional imaging methods are essential tools to investigate the in vivo performance of these formulations. The non-invasive technique of ac biosusceptometry has been introduced as an alternative in studies focusing on gastrointestinal motility and, more recently, to evaluate the behaviour of magnetic tablets in vivo. The aim of this work was to employ a multisensor ac biosusceptometer system to obtain magnetic images of disintegration of tablets in vitro and in the human stomach. The results showed that the transition between the magnetic marker and the magnetic tracer characterized the onset of disintegration (t50) and occurred in a short time interval (1.1 ± 0.4 min). The multisensor ac biosusceptometer was reliable to monitor and analyse the in vivo performance of magnetic tablets showing accuracy to quantify disintegration through the magnetic images and to characterize the profile of this process.

  18. AC measurements on uranium doped high temperature superconductors

    International Nuclear Information System (INIS)

    Eisterer, M.

    1999-11-01

    The subject of this thesis is the influence of fission tracks on the superconducting properties of melt textured Y-123. The critical current densities, the irreversibility lines and the transition temperature were determined by means of ac measurements. The corresponding ac techniques are explored in detail. Deviations of the ac signal from the expectations according to the Bean model were explained by the dependence of the shielding currents on the electric field. This explanation is supported by the influence of the ac amplitude and frequency on the critical current density but also by a comparison of the obtained data with other experimental techniques. Y-123 has to be doped with uranium in order to induce fission tracks. Uranium forms normal conducting clusters, which are nearly spherical, with a diameter of about 300 nm. Fission of uranium-235 by thermal neutrons creates two high energy ions with a total energy of about 160 MeV. Each of these fission products induces a linear defect with a diameter of about 10 nm. The length of one fission track is 2-4 μm. At 77 K the critical current density is enhanced by the pinning action of the uranium clusters, compared to undoped samples. With decreasing temperature this influence becomes negligible. The critical current densities are strongly enhanced due to the irradiation. At low magnetic fields we find extremely high values for melt textured materials, e.g. 2.5x10 9 Am -2 at 77 K and 0.25 T or 6x10 10 Am -2 at 5 K. Since the critical current was found to be inverse proportional to the square root of the applied magnetic field it decreases rapidly as the field increases. This behavior is predicted by simple theoretical considerations, but is only valid at low temperatures as well as in low magnetic fields at high temperatures. At high fields the critical current drops more rapidly. The irreversibility lines are only slightly changed by this irradiation technique. Only a small shift to higher fields and temperatures

  19. Power Electronic Transformer based Three-Phase PWM AC Drives

    Science.gov (United States)

    Basu, Kaushik

    A Transformer is used to provide galvanic isolation and to connect systems at different voltage levels. It is one of the largest and most expensive component in most of the high voltage and high power systems. Its size is inversely proportional to the operating frequency. The central idea behind a power electronic transformer (PET) also known as solid state transformer is to reduce the size of the transformer by increasing the frequency. Power electronic converters are used to change the frequency of operation. Steady reduction in the cost of the semiconductor switches and the advent of advanced magnetic materials with very low loss density and high saturation flux density implies economic viability and feasibility of a design with high power density. Application of PET is in generation of power from renewable energy sources, especially wind and solar. Other important application include grid tied inverters, UPS e.t.c. In this thesis non-resonant, single stage, bi-directional PET is considered. The main objective of this converter is to generate adjustable speed and magnitude pulse width modulated (PWM) ac waveforms from an ac or dc grid with a high frequency ac link. The windings of a high frequency transformer contains leakage inductance. Any switching transition of the power electronic converter connecting the inductive load and the transformer requires commutation of leakage energy. Commutation by passive means results in power loss, decrease in the frequency of operation, distortion in the output voltage waveform, reduction in reliability and power density. In this work a source based partially loss-less commutation of leakage energy has been proposed. This technique also results in partial soft-switching. A series of converters with novel PWM strategies have been proposed to minimize the frequency of leakage inductance commutation. These PETs achieve most of the important features of modern PWM ac drives including 1) Input power factor correction, 2) Common

  20. Multi-phase AC/AC step-down converter for distribution systems

    Science.gov (United States)

    Aeloiza, Eddy C.; Burgos, Rolando P.

    2017-10-25

    A step-down AC/AC converter for use in an electric distribution system includes at least one chopper circuit for each one of a plurality of phases of the AC power, each chopper circuit including a four-quadrant switch coupled in series between primary and secondary sides of the chopper circuit and a current-bidirectional two-quadrant switch coupled between the secondary side of the chopper circuit and a common node. Each current-bidirectional two-quadrant switch is oriented in the same direction, with respect to the secondary side of the corresponding chopper circuit and the common node. The converter further includes a control circuit configured to pulse-width-modulate control inputs of the switches, to convert a first multiphase AC voltage at the primary sides of the chopper circuits to a second multiphase AC voltage at the secondary sides of the chopper circuits, the second multiphase AC voltage being lower in voltage than the first multiphase AC voltage.

  1. AC loss in superconducting tapes and cables

    NARCIS (Netherlands)

    Oomen, M.P.

    2000-01-01

    The present study discusses the AC loss in high-temperature superconductors. Superconducting materials with a relatively high critical temperature were discovered in 1986. They are presently developed for use in large-scale power-engineering devices such as power-transmission cables, transformers

  2. Composite Based EHV AC Overhead Transmission Lines

    DEFF Research Database (Denmark)

    Sørensen, Thomas Kjærsgaard

    and analysed with regard to the possibilities, limitations and risks widespread application of composite materials on EHV AC overhead transmission lines may present. To form the basis for evaluation of the useability of composite materials, dierent overhead line projects aimed at reducing the environmental...

  3. Ac-dc converter firing error detection

    International Nuclear Information System (INIS)

    Gould, O.L.

    1996-01-01

    Each of the twelve Booster Main Magnet Power Supply modules consist of two three-phase, full-wave rectifier bridges in series to provide a 560 VDC maximum output. The harmonic contents of the twelve-pulse ac-dc converter output are multiples of the 60 Hz ac power input, with a predominant 720 Hz signal greater than 14 dB in magnitude above the closest harmonic components at maximum output. The 720 Hz harmonic is typically greater than 20 dB below the 500 VDC output signal under normal operation. Extracting specific harmonics from the rectifier output signal of a 6, 12, or 24 pulse ac-dc converter allows the detection of SCR firing angle errors or complete misfires. A bandpass filter provides the input signal to a frequency-to-voltage converter. Comparing the output of the frequency-to-voltage converter to a reference voltage level provides an indication of the magnitude of the harmonics in the ac-dc converter output signal

  4. THE ACS NEARBY GALAXY SURVEY TREASURY

    International Nuclear Information System (INIS)

    Dalcanton, Julianne J.; Williams, Benjamin F.; Rosema, Keith; Gogarten, Stephanie M.; Christensen, Charlotte; Gilbert, Karoline; Hodge, Paul; Seth, Anil C.; Dolphin, Andrew; Holtzman, Jon; Skillman, Evan D.; Weisz, Daniel; Cole, Andrew; Girardi, Leo; Karachentsev, Igor D.; Olsen, Knut; Freeman, Ken; Gallart, Carme; Harris, Jason; De Jong, Roelof S.

    2009-01-01

    The ACS Nearby Galaxy Survey Treasury (ANGST) is a systematic survey to establish a legacy of uniform multi-color photometry of resolved stars for a volume-limited sample of nearby galaxies (D 4 in luminosity and star formation rate. The survey data consist of images taken with the Advanced Camera for Surveys (ACS) on the Hubble Space Telescope (HST), supplemented with archival data and new Wide Field Planetary Camera 2 (WFPC2) imaging taken after the failure of ACS. Survey images include wide field tilings covering the full radial extent of each galaxy, and single deep pointings in uncrowded regions of the most massive galaxies in the volume. The new wide field imaging in ANGST reaches median 50% completenesses of m F475W = 28.0 mag, m F606W = 27.3 mag, and m F814W = 27.3 mag, several magnitudes below the tip of the red giant branch (TRGB). The deep fields reach magnitudes sufficient to fully resolve the structure in the red clump. The resulting photometric catalogs are publicly accessible and contain over 34 million photometric measurements of >14 million stars. In this paper we present the details of the sample selection, imaging, data reduction, and the resulting photometric catalogs, along with an analysis of the photometric uncertainties (systematic and random), for both ACS and WFPC2 imaging. We also present uniformly derived relative distances measured from the apparent magnitude of the TRGB.

  5. Predicting AC loss in practical superconductors

    International Nuclear Information System (INIS)

    Goemoery, F; Souc, J; Vojenciak, M; Seiler, E; Klincok, B; Ceballos, J M; Pardo, E; Sanchez, A; Navau, C; Farinon, S; Fabbricatore, P

    2006-01-01

    Recent progress in the development of methods used to predict AC loss in superconducting conductors is summarized. It is underlined that the loss is just one of the electromagnetic characteristics controlled by the time evolution of magnetic field and current distribution inside the conductor. Powerful methods for the simulation of magnetic flux penetration, like Brandt's method and the method of minimal magnetic energy variation, allow us to model the interaction of the conductor with an external magnetic field or a transport current, or with both of them. The case of a coincident action of AC field and AC transport current is of prime importance for practical applications. Numerical simulation methods allow us to expand the prediction range from simplified shapes like a (infinitely high) slab or (infinitely thin) strip to more realistic forms like strips with finite rectangular or elliptic cross-section. Another substantial feature of these methods is that the real composite structure containing an array of superconducting filaments can be taken into account. Also, the case of a ferromagnetic matrix can be considered, with the simulations showing a dramatic impact on the local field. In all these circumstances, it is possible to indicate how the AC loss can be reduced by a proper architecture of the composite. On the other hand, the multifilamentary arrangement brings about a presence of coupling currents and coupling loss. Simulation of this phenomenon requires 3D formulation with corresponding growth of the problem complexity and computation time

  6. Meso Mechanical Analysis of AC Mixture Response

    NARCIS (Netherlands)

    Woldekidan, M.F.; Huurman, M.; Vaccari, E.; Poot, M.

    2012-01-01

    Ongoing research into performance modeling of Asphalt Concrete (AC) mixtures using meso mechanics approaches is being undertaken at Delft University of Technology (TUD). The approach has already been successfully employed for evaluating the long term performance of porous asphalt concrete. The work

  7. Analysis of AC Low-Voltage Energy Harvesting

    Science.gov (United States)

    2014-09-01

    We have seen such efforts in car manufacturing, such as the Prius, that returns energy to the battery through the use of its regenerative brake ... system . Power electronics is the critical technology that makes harvesting this unused energy possible. Piezoelectricity is a material property that...Demo Circuit 1459b quick start guide [user’s guide]. Milpitas, CA: Linear Technology, April 2010. [13] Piezo Systems , “Piezoelectric Energy

  8. Investigate Transient Behaviours and Select Appropriate Fault Protection Solutions of Uni-grounded AC Microgrids

    Directory of Open Access Journals (Sweden)

    Duong Minh Bui

    2016-03-01

    Full Text Available Transient situations of a uni-grounded low-voltage AC microgrid are simulated in this paper, which include different fault tests and operation transition test between the grid-connected and islanded modes of the uni-grounded microgrid. Based on transient simulation results, available fault protection methods are proposed for the main and back-up protection of a uni-grounded AC microgrid. Main contributions of this paper are (i analysing transient responses of a typically uni-grounded lowvoltage AC microgrid from line-to-line, single line-to-ground, three-phase faults and a microgrid operation transition test; and (ii proposing available fault protection methods for uni-grounded AC microgrids, such as non-directional/directional overcurrent protection solutions, under/over voltage protection solutions, differential protection, voltage-restrained overcurrent protection, and other protection principles not based on high fault currents (e.g. total harmonic distortion detection of phase currents and voltages, or protection methods using symmetrical sequence components of current and voltage.

  9. A Generalised Fault Protection Structure Proposed for Uni-grounded Low-Voltage AC Microgrids

    Science.gov (United States)

    Bui, Duong Minh; Chen, Shi-Lin; Lien, Keng-Yu; Jiang, Jheng-Lun

    2016-04-01

    This paper presents three main configurations of uni-grounded low-voltage AC microgrids. Transient situations of a uni-grounded low-voltage (LV) AC microgrid (MG) are simulated through various fault tests and operation transition tests between grid-connected and islanded modes. Based on transient simulation results, available fault protection methods are proposed for main and back-up protection of a uni-grounded AC microgrid. In addition, concept of a generalised fault protection structure of uni-grounded LVAC MGs is mentioned in the paper. As a result, main contributions of the paper are: (i) definition of different uni-grounded LVAC MG configurations; (ii) analysing transient responses of a uni-grounded LVAC microgrid through line-to-line faults, line-to-ground faults, three-phase faults and a microgrid operation transition test, (iii) proposing available fault protection methods for uni-grounded microgrids, such as: non-directional or directional overcurrent protection, under/over voltage protection, differential current protection, voltage-restrained overcurrent protection, and other fault protection principles not based on phase currents and voltages (e.g. total harmonic distortion detection of currents and voltages, using sequence components of current and voltage, 3I0 or 3V0 components), and (iv) developing a generalised fault protection structure with six individual protection zones to be suitable for different uni-grounded AC MG configurations.

  10. AC susceptibility of thin Pb films in intermediate and mixed state

    Energy Technology Data Exchange (ETDEWEB)

    Janu, Zdenek, E-mail: janu@fzu.cz [Institute of Physics of the AS CR, v.v.i., Na Slovance 2, CZ-182 21 Prague 8 (Czech Republic); Svindrych, Zdenek [Institute of Physics of the AS CR, v.v.i., Na Slovance 2, CZ-182 21 Prague 8 (Czech Republic); Trunecek, Otakar [Charles University in Prague, Faculty of Mathematics and Physics, Ke Karlovu 3, CZ-121 16 Prague 2 (Czech Republic); Kus, Peter; Plecenik, Andrej [Komenius University in Bratislava, Faculty of Mathematics, Physics, and Informatics, Mlynska dolina, 842 48 Bratislava 4 (Slovakia)

    2011-12-15

    Thickness dependent transition in AC susceptibility between intermediate and mixed state in type-I superconducting films. The temperature induced crossover between reversible and irreversible behavior was observed in the thicker film. The temperature dependence of the AC susceptibility in mixed state follows prediction of model based on Bean critical state. The temperature dependence of the harmonics of the complex AC susceptibility in the intermediate state is explained. Thin films of type I superconductors of a thickness comparable or less than a flux penetration length behave like type II superconductors in a mixed state. With decreasing film thickness normal domains carrying a magnetic flux get smaller with smaller number of flux quanta per domain and finally transform into single quantum flux lines, i.e. quantum vortices similar to those found in type II superconductors. We give an evidence of this behavior from the measurements of the nonlinear response of a total magnetic moment to an applied AC magnetic field, directly from the temperature dependence of an AC susceptibility.

  11. A CFD analysis of flow blockage phenomena in ALFRED LFR demo fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Di Piazza, Ivan, E-mail: ivan.dipiazza@enea.it [Italian National Agency for New Technologies, Energy and Sustainable Economic Development, C.R. ENEA Brasimone (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

  12. Assay Methods for ACS Activity and ACS Phosphorylation by MAP Kinases In Vitro and In Vivo.

    Science.gov (United States)

    Han, Xiaomin; Li, Guojing; Zhang, Shuqun

    2017-01-01

    Ethylene, a gaseous phytohormone, has profound effects on plant growth, development, and adaptation to the environment. Ethylene-regulated processes begin with the induction of ethylene biosynthesis. There are two key steps in ethylene biosynthesis. The first is the biosynthesis of 1-aminocyclopropane-1-carboxylic acid (ACC) from S-Adenosyl-Methionine (SAM), a common precursor in many metabolic pathways, which is catalyzed by ACC synthase (ACS). The second is the oxidative cleavage of ACC to form ethylene under the action of ACC oxidase (ACO). ACC biosynthesis is the committing and generally the rate-limiting step in ethylene biosynthesis. As a result, characterizing the cellular ACS activity and understanding its regulation are important. In this chapter, we detail the methods used to measure, (1) the enzymatic activity of both recombinant and native ACS proteins, and (2) the phosphorylation of ACS protein by mitogen-activated protein kinases (MAPKs) in vivo and in vitro.

  13. High voltage AC/AC electrochemical capacitor operating at low temperature in salt aqueous electrolyte

    Science.gov (United States)

    Abbas, Qamar; Béguin, François

    2016-06-01

    We demonstrate that an activated carbon (AC)-based electrochemical capacitor implementing aqueous lithium sulfate electrolyte in 7:3 vol:vol water/methanol mixture can operate down to -40 °C with good electrochemical performance. Three-electrode cell investigations show that the faradaic contributions related with hydrogen chemisorption in the negative AC electrode are thermodynamically unfavored at -40 °C, enabling the system to work as a typical electrical double-layer (EDL) capacitor. After prolonged floating of the AC/AC capacitor at 1.6 V and -40°C, the capacitance, equivalent series resistance and efficiency remain constant, demonstrating the absence of ageing related with side redox reactions at this temperature. Interestingly, when temperature is increased back to 24 °C, the redox behavior due to hydrogen storage reappears and the system behaves as a freshly prepared one.

  14. Digital model for harmonic interactions in AC/DC/AC systems

    Energy Technology Data Exchange (ETDEWEB)

    Guarini, A P; Rangel, R D; Pilotto, L A.S.; Pinto, R J; Passos, Junior, R [Centro de Pesquisas de Energia Eletrica (CEPEL), Rio de Janeiro, RJ (Brazil)

    1994-12-31

    The main purpose of this paper is to present a model for calculation of HVdc converter harmonics taking into account the influence of the harmonic interactions between the ac systems in dc link transmissions. The ideas and methodologies used in the model development take into account the dc current ripple and ac voltage distortion in the ac systems. The theory of switching functions is applied to contemplate for the frequency conversions between the ac and dc sides, in an iterative process. It is possible then to obtain, even in balanced situations, non-characteristic harmonics that are produced by frequencies originated in the other terminal, which can be significant in a strongly coupled system, such as back-to-back configuration. (author) 9 refs., 3 figs.

  15. Approaches to building single-stage AC/AC conversion switch-mode audio power amplifiers

    Energy Technology Data Exchange (ETDEWEB)

    Ljusev, P.; Andersen, Michael A.E.

    2005-07-01

    This paper discusses the possible topologies and promising approaches towards direct single-phase AC-AC conversion of the mains voltage for audio applications. When compared to standard Class-D switching audio power amplifiers with a separate power supply, it is expected that direct conversion will provide better efficiency and higher level of integration, leading to lower component count, volume and cost, but at the expense of a minor performance deterioration. (au)

  16. Low-temperature heat capacity of small Nb3Sn polycrystals by ac calorimetry

    International Nuclear Information System (INIS)

    Viswanathan, R.; Johnston, D.C.

    1976-01-01

    It is shown by an ac calorimetry technique that the multiple heat capacity anomalies which occur below the superconducting transition temperature for small polycrystalline Nb 3 Sn samples are intrinsic to these samples. The recent suggestions that shear stresses can account for these results are analyzed for their validity. The dependence of the occurrence of these multiple anomalies upon the thermal history of the samples was investigated

  17. Modelling of fission gas release in rods from the International DEMO-RAMP-II Project at Studsvik

    International Nuclear Information System (INIS)

    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)

  18. The enhanced pellet centrifuge launcher at ASDEX Upgrade: Advanced operation and application as technology test facility for ITER and DEMO

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. Creep behavior of 8Cr2WVTa martensitic steel designed for fusion DEMO reactor. An assessment on helium embrittlement resistance

    International Nuclear Information System (INIS)

    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)

  20. Structural design of DEMO Divertor Cassette Body: provisional FEM analysis and introductive application of RCC-MRx design rules

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  2. Development of Tokamak reactor system code and conceptual studies of DEMO with He Cooled Molten Li blanket

    International Nuclear Information System (INIS)

    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.)

  3. Faradaic AC Electrokinetic Flow and Particle Traps

    Science.gov (United States)

    Ben, Yuxing; Chang, Hsueh-Chia

    2004-11-01

    Faradaic reaction at higher voltages can produce co-ion polarization at AC electrodes instead of counter-ion polarization due to capacitive charging from the bulk. The Faradaic co-ion polarization also does not screen the external field and hence can produce large net electro-kinetic flows at frequencies lower than the inverse RC time of the double layer. Due to the opposite polarization of capacitve and Faradaic charging, we can reverse the direction of AC flows on electrodes by changing the voltage and frequency. Particles and bacteria are trapped and then dispersed at stagnation lines, at locations predicted by our theory, by using these two flows sequentially. This technique offers a good way to concentrate and detect bacteria.

  4. AC application of second generation HTS wire

    Science.gov (United States)

    Thieme, C. L. H.; Gagnon, K.; Voccio, J.; Aized, D.; Claassen, J.

    2008-02-01

    For the production of Second Generation (2G) YBCO High Temperature Superconductor wire American Superconductor uses a wide-strip MOD-YBCO/RABiTSTM process, a low-cost approach for commercial manufacturing. It can be engineered with a high degree of flexibility to manufacture practical 2G conductors with architectures and properties tailored for specific applications and operating conditions. For ac applications conductor and coil design can be geared towards low hysteretic losses. For applications which experience high frequency ac fields, the stabilizer needs to be adjusted for low eddy current losses. For these applications a stainless-steel laminate is used. An example is a Low Pass Filter Inductor which was developed and built in this work.

  5. Aging, Counterfeiting Configuration Control (AC3)

    Science.gov (United States)

    2010-01-31

    Systems Intergrated Into AC3 CABS - Common As-Built System PRISM - Process Re-inventing Integration Systems for Manufacturing PDM - Product Data...looks forward to deploying the completed tool at Raytheon in a true production environment, for as much as we like the challenge associated with...performance of DoD systems. DoD systems are particularly susceptible to intrusion of counterfeit parts, especially during surge and extended production

  6. The LHC AC Dipole system: an introduction

    CERN Document Server

    Serrano, J; CERN. Geneva. BE Department

    2010-01-01

    The LHC AC Dipole is an instrument to study properties of the LHC lattice by inducing large transverse displacements in the beam. These displacements are generated by exciting the beam with an oscillating magnetic field at a frequency close to the tune. This paper presents the system requirements and the technical solution chosen to meet them, based of high-power audio amplifiers and a resonant parallel RLC circuit.

  7. Modeling photovoltaic systems for AC appliances

    Directory of Open Access Journals (Sweden)

    Andreea Maria Neaca

    2009-10-01

    Full Text Available In this paper is described the development of a model which can simulate the performance of a photovoltaic (PV system under specific meteorological conditions and transforming the DC current into AC current. In this model, the accent stands on the design of a series charge regulator. It is treated also the benefit of creating a circuit, with different methods, that can test the maximum power point trackers (MPPT for different photovoltaic applications.

  8. Control of grid interactive AC microgrids

    DEFF Research Database (Denmark)

    Wang, Xiongfei; Guerrero, Josep M.; Chen, Zhe

    2010-01-01

    Over the last decade, distributed energy resources (DER) technology has undergone a fast development. Increased penetration of DER units and wide spread use of renewable energy sources challenge the entire architecture of traditional power system. Microgrid, characterizing higher flexibility......, microgrid controls and power management strategies are presented. Future trends of microgrid are discussed pointing out how this concept can be a key to achieve a more intelligent and flexible AC grid....

  9. CTE Corrections for WFPC2 and ACS

    Science.gov (United States)

    Dolphin, Andrew

    2003-07-01

    The error budget for optical broadband photometry is dominated by three factors: CTE corrections, long-short anomaly corrections, and photometric zero points. Questions about the dependencies of the CTE have largely been resolved, and my CTE corrections have been included in the WFPC2 handbook and tutorial. What remains to be done is the determination of the "final" CTE correction at the end of the WFPC2 mission, which will increase the accuracy of photometry obtained in the final few cycles. The long-short anomaly is still the subject of much debate, as it remains unclear whethere or not this effect is real and, if so, what its size and nature is. Photometric zero points have likewise varied by over 0.05 magnitudes in the literature, and will likely remain unresolved until the long-short anomaly is addressed {given that most calibration exposures are short while most science exposures are long}. It is also becoming apparent that similar issues will affect the accuracy of ACS photometry, and consequently that an ACS CTE study analogous to my WFPC2 work would significantly improve the calibration of ACS. I therefore propose to use archival WFPC2 images of omega Cen and ACS images of 47 Tuc to continue my HST calibration work. I also propose to begin work on "next-generation" CTE corrections, in which corrections are applied to the images based on accurate charge-trapping models rather than to the reduced photometry. This technique will allow for more accurate CTE corrections in certain cases {such as a star above a bright star or on a variable background}, improved PSF-fitting photometry of faint stars, and image restoration for accurate analysis of extended objects.

  10. New three-phase ac-ac converter incorporating three-phase boost integrated ZVT bridge and single-phase HF link

    International Nuclear Information System (INIS)

    Abdelhamid, Tamer H.; Sabzali, Ahmad J.

    2008-01-01

    This paper presents a new zero voltage transition (ZVT), power factor corrected three phase ac-ac converter with single phase high frequency (HF) link. It is a two stage converter; the first stage is a boost integrated bridge converter (combination of a 3 ph boost converter and a bridge converter) operated at fixed frequency and that operates in two modes at ZVT for all switches and establishes a 1 ph square wave HF link. The second stage is a bi-directional pulse width modulation (PWM) 3 ph bridge that converts the 1 ph HF link to a 3 ph voltage using a novel switching strategy. The converter modes of operation and key equations are outlined. Simulation of the overall system is conducted using Simulink. The switching strategy and its corresponding control circuit are clearly described. Experimental verification of the simulation is conducted for a prototype of 100 V, 500 W at 10 kHz link frequency

  11. Direct amplitude detuning measurement with ac dipole

    Directory of Open Access Journals (Sweden)

    S. White

    2013-07-01

    Full Text Available In circular machines, nonlinear dynamics can impact parameters such as beam lifetime and could result in limitations on the performance reach of the accelerator. Assessing and understanding these effects in experiments is essential to confirm the accuracy of the magnetic model and improve the machine performance. A direct measurement of the machine nonlinearities can be obtained by characterizing the dependency of the tune as a function of the amplitude of oscillations (usually defined as amplitude detuning. The conventional technique is to excite the beam to large amplitudes with a single kick and derive the tune from turn-by-turn data acquired with beam position monitors. Although this provides a very precise tune measurement it has the significant disadvantage of being destructive. An alternative, nondestructive way of exciting large amplitude oscillations is to use an ac dipole. The perturbation Hamiltonian in the presence of an ac dipole excitation shows a distinct behavior compared to the free oscillations which should be correctly taken into account in the interpretation of experimental data. The use of an ac dipole for direct amplitude detuning measurement requires careful data processing allowing one to observe the natural tune of the machine; the feasibility of such a measurement is demonstrated using experimental data from the Large Hadron Collider. An experimental proof of the theoretical derivations based on measurements performed at injection energy is provided as well as an application of this technique at top energy using a large number of excitations on the same beam.

  12. Flame spread over inclined electrical wires with AC electric fields

    KAUST Repository

    Lim, Seung J.; Park, Sun H.; Park, Jeong; Fujita, Osamu; Keel, Sang I.; Chung, Suk-Ho

    2017-01-01

    Flame spread over polyethylene-insulated electrical wires was studied experimentally with applied alternating current (AC) by varying the inclination angle (θ), applied voltage (VAC), and frequency (fAC). For the baseline case with no electric field

  13. The Hubble Legacy Archive ACS grism data

    Science.gov (United States)

    Kümmel, M.; Rosati, P.; Fosbury, R.; Haase, J.; Hook, R. N.; Kuntschner, H.; Lombardi, M.; Micol, A.; Nilsson, K. K.; Stoehr, F.; Walsh, J. R.

    2011-06-01

    A public release of slitless spectra, obtained with ACS/WFC and the G800L grism, is presented. Spectra were automatically extracted in a uniform way from 153 archival fields (or "associations") distributed across the two Galactic caps, covering all observations to 2008. The ACS G800L grism provides a wavelength range of 0.55-1.00 μm, with a dispersion of 40 Å/pixel and a resolution of ~80 Å for point-like sources. The ACS G800L images and matched direct images were reduced with an automatic pipeline that handles all steps from archive retrieval, alignment and astrometric calibration, direct image combination, catalogue generation, spectral extraction and collection of metadata. The large number of extracted spectra (73,581) demanded automatic methods for quality control and an automated classification algorithm was trained on the visual inspection of several thousand spectra. The final sample of quality controlled spectra includes 47 919 datasets (65% of the total number of extracted spectra) for 32 149 unique objects, with a median iAB-band magnitude of 23.7, reaching 26.5 AB for the faintest objects. Each released dataset contains science-ready 1D and 2D spectra, as well as multi-band image cutouts of corresponding sources and a useful preview page summarising the direct and slitless data, astrometric and photometric parameters. This release is part of the continuing effort to enhance the content of the Hubble Legacy Archive (HLA) with highly processed data products which significantly facilitate the scientific exploitation of the Hubble data. In order to characterize the slitless spectra, emission-line flux and equivalent width sensitivity of the ACS data were compared with public ground-based spectra in the GOODS-South field. An example list of emission line galaxies with two or more identified lines is also included, covering the redshift range 0.2 - 4.6. Almost all redshift determinations outside of the GOODS fields are new. The scope of science projects

  14. Importance of Attenuation Correction (AC) for Small Animal PET Imaging

    DEFF Research Database (Denmark)

    El Ali, Henrik H.; Bodholdt, Rasmus Poul; Jørgensen, Jesper Tranekjær

    2012-01-01

    was performed. Methods: Ten NMRI nude mice with subcutaneous implantation of human breast cancer cells (MCF-7) were scanned consecutively in small animal PET and CT scanners (MicroPETTM Focus 120 and ImTek’s MicroCATTM II). CT-based AC, PET-based AC and uniform AC methods were compared. Results: The activity...

  15. Development of a hardware-based AC microgrid for AC stability assessment

    Science.gov (United States)

    Swanson, Robert R.

    As more power electronic-based devices enable the development of high-bandwidth AC microgrids, the topic of microgrid power distribution stability has become of increased interest. Recently, researchers have proposed a relatively straightforward method to assess the stability of AC systems based upon the time-constants of sources, the net bus capacitance, and the rate limits of sources. In this research, a focus has been to develop a hardware test system to evaluate AC system stability. As a first step, a time domain model of a two converter microgrid was established in which a three phase inverter acts as a power source and an active rectifier serves as an adjustable constant power AC load. The constant power load can be utilized to create rapid power flow transients to the generating system. As a second step, the inverter and active rectifier were designed using a Smart Power Module IGBT for switching and an embedded microcontroller as a processor for algorithm implementation. The inverter and active rectifier were designed to operate simultaneously using a synchronization signal to ensure each respective local controller operates in a common reference frame. Finally, the physical system was created and initial testing performed to validate the hardware functionality as a variable amplitude and variable frequency AC system.

  16. Pixel-based CTE Correction of ACS/WFC: Modifications To The ACS Calibration Pipeline (CALACS)

    Science.gov (United States)

    Smith, Linda J.; Anderson, J.; Armstrong, A.; Avila, R.; Bedin, L.; Chiaberge, M.; Davis, M.; Ferguson, B.; Fruchter, A.; Golimowski, D.; Grogin, N.; Hack, W.; Lim, P. L.; Lucas, R.; Maybhate, A.; McMaster, M.; Ogaz, S.; Suchkov, A.; Ubeda, L.

    2012-01-01

    The Advanced Camera for Surveys (ACS) was installed on the Hubble Space Telescope (HST) nearly ten years ago. Over the last decade, continuous exposure to the harsh radiation environment has degraded the charge transfer efficiency (CTE) of the CCDs. The worsening CTE impacts the science that can be obtained by altering the photometric, astrometric and morphological characteristics of sources, particularly those farthest from the readout amplifiers. To ameliorate these effects, Anderson & Bedin (2010, PASP, 122, 1035) developed a pixel-based empirical approach to correcting ACS data by characterizing the CTE profiles of trails behind warm pixels in dark exposures. The success of this technique means that it is now possible to correct full-frame ACS/WFC images for CTE degradation in the standard data calibration and reduction pipeline CALACS. Over the past year, the ACS team at STScI has developed, refined and tested the new software. The details of this work are described in separate posters. The new code is more effective at low flux levels (repair ACS electronics) and pixel-based CTE correction. In addition to the standard cosmic ray corrected, flat-fielded and drizzled data products (crj, flt and drz files) there are three new equivalent files (crc, flc and drc) which contain the CTE-corrected data products. The user community will be able to choose whether to use the standard or CTE-corrected products.

  17. Examining the Link Between Public Transit Use and Active Commuting

    Directory of Open Access Journals (Sweden)

    Melissa Bopp

    2015-04-01

    Full Text Available Background: An established relationship exists between public transportation (PT use and physical activity. However, there is limited literature that examines the link between PT use and active commuting (AC behavior. This study examines this link to determine if PT users commute more by active modes. Methods: A volunteer, convenience sample of adults (n = 748 completed an online survey about AC/PT patterns, demographic, psychosocial, community and environmental factors. t-test compared differences between PT riders and non-PT riders. Binary logistic regression analyses examined the effect of multiple factors on AC and a full logistic regression model was conducted to examine AC. Results: Non-PT riders (n = 596 reported less AC than PT riders. There were several significant relationships with AC for demographic, interpersonal, worksite, community and environmental factors when considering PT use. The logistic multivariate analysis for included age, number of children and perceived distance to work as negative predictors and PT use, feelings of bad weather and lack of on-street bike lanes as a barrier to AC, perceived behavioral control and spouse AC were positive predictors. Conclusions: This study revealed the complex relationship between AC and PT use. Further research should investigate how AC and public transit use are related.

  18. Transitional Justice

    DEFF Research Database (Denmark)

    Gissel, Line Engbo

    This presentation builds on an earlier published article, 'Contemporary Transitional Justice: Normalising a Politics of Exception'. It argues that the field of transitional justice has undergone a shift in conceptualisation and hence practice. Transitional justice is presently understood to be th...... to be the provision of ordinary criminal justice in contexts of exceptional political transition.......This presentation builds on an earlier published article, 'Contemporary Transitional Justice: Normalising a Politics of Exception'. It argues that the field of transitional justice has undergone a shift in conceptualisation and hence practice. Transitional justice is presently understood...

  19. Aplicación móvil para la visualización y ejecución de demos en IPOL

    OpenAIRE

    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...

  20. Combination of helical ferritic-steel inserts and flux-tube-expansion divertor for the heat control in tokamak DEMO reactor

    International Nuclear Information System (INIS)

    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

  1. Design concept of conducting shell and in-vessel components suitable for plasma vertical stability and remote maintenance scheme in DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    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.

  2. MATHEMATICAL MODELING OF AC ELECTRIC POINT MOTOR

    Directory of Open Access Journals (Sweden)

    S. YU. Buryak

    2014-03-01

    Full Text Available Purpose. In order to ensure reliability, security, and the most important the continuity of the transportation process, it is necessary to develop, implement, and then improve the automated methods of diagnostic mechanisms, devices and rail transport systems. Only systems that operate in real time mode and transmit data on the instantaneous state of the control objects can timely detect any faults and thus provide additional time for their correction by railway employees. Turnouts are one of the most important and responsible components, and therefore require the development and implementation of such diagnostics system.Methodology. Achieving the goal of monitoring and control of railway automation objects in real time is possible only with the use of an automated process of the objects state diagnosing. For this we need to know the diagnostic features of a control object, which determine its state at any given time. The most rational way of remote diagnostics is the shape and current spectrum analysis that flows in the power circuits of railway automatics. Turnouts include electric motors, which are powered by electric circuits, and the shape of the current curve depends on both the condition of the electric motor, and the conditions of the turnout maintenance. Findings. For the research and analysis of AC electric point motor it was developed its mathematical model. The calculation of parameters and interdependencies between the main factors affecting the operation of the asynchronous machine was conducted. The results of the model operation in the form of time dependences of the waveform curves of current on the load on engine shaft were obtained. Originality. During simulation the model of AC electric point motor, which satisfies the conditions of adequacy was built. Practical value. On the basis of the constructed model we can study the AC motor in various mode of operation, record and analyze current curve, as a response to various changes

  3. AC susceptibility enhancement studies in magnetic systems

    International Nuclear Information System (INIS)

    Mukherjee, S.; Ranganathan, R.; Chakravarti, A.; Sil, S.

    2001-01-01

    Enhancement of AC susceptibility has been observed for typical ferromagnets (Gd), reentrant spin glasses like (Fe 1.5 Mn 1.5 Si) and canted spin systems (Ce(Fe 0.96 Al 0.04 ) 2 ). The data have been interpreted with the help of a simulation model based on dry friction-like pinning of domain walls for systems having ferromagnetic domain structures. A strong pinning mechanism appears in the reentrant spin glass like and canted spin systems at low temperatures in addition to the intrinsic one in the ferromagnetic phase. The temperature variation of the pinning potential has been given qualitatively for the reentrant spin glass like systems

  4. Protection of AC and DC Microgrids

    DEFF Research Database (Denmark)

    Beheshtaein, Siavash; Savaghebi, Mehdi; Quintero, Juan Carlos Vasquez

    2015-01-01

    and DC microgrids, and then investigates the existing and promising solutions for the corresponding challenges. To the authors’ knowledge, three parts of smart grids are required to be developed to facilitate implementation of protection scheme in microgrids. The main requirements and open issues......In future, distributed energy resources (RESs) will be utilized at consumption points. As a consequence, power flow and fault current would be bidirectional and topologydependent; and hence the conventional protection strategies would be inefficient. This paper categorizes the main challenges in AC...

  5. Flexible AC transmission systems modelling and control

    CERN Document Server

    Zhang, Xiao-Ping; Pal, Bikash

    2012-01-01

    The extended and revised second edition of this successful monograph presents advanced modeling, analysis and control techniques of Flexible AC Transmission Systems (FACTS). The book covers comprehensively a range of power-system control problems: from steady-state voltage and power flow control, to voltage and reactive power control, to voltage stability control, to small signal stability control using FACTS controllers. In the six years since the first edition of the book has been published research on the FACTS has continued to flourish while renewable energy has developed into a mature and

  6. DC injection into low voltage AC networks

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    This report summarises the results of a study investigating the impact of levels of injected DC current injections on a low voltage AC distribution network systems in order to recommend acceptable limits of DC from microgeneration. Relevant literature is reviewed, and the impact of DC levels in distribution transformers, transformer modelling, and instrumental transformers are discussed. The impact of DC in residual current devices (RCD) and in domestic electricity watt hour meters is examined along with DC enhanced corrosion, corrosion failure, and the measurement of DC current injection. Sources of DC injection outlined include DC from computer power supplies, network faults, geomagnetic phenomena, lighting circuits/dimmers, and embedded generators.

  7. Three-Level AC-DC-AC Z-Source Converter Using Reduced Passive Component Count

    DEFF Research Database (Denmark)

    Loh, Poh Chiang; Gao, Feng; Tan, Pee-Chin

    2009-01-01

    This paper presents a three-level ac-dc-ac Z-source converter with output voltage buck-boost capability. The converter is implemented by connecting a low-cost front-end diode rectifier to a neutral-point-clamped inverter through a single X-shaped LC impedance network. The inverter is controlled...... to switch with a three-level output voltage, where the middle neutral potential is uniquely tapped from the star-point of a wye-connected capacitive filter placed before the front-end diode rectifier for input current filtering. Through careful control, the resulting converter can produce the correct volt...

  8. Multicriteria selection in concept design of a divertor remote maintenance port in the EU DEMO reactor using an AHP participative approach

    Energy Technology Data Exchange (ETDEWEB)

    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.

  9. Multicriteria selection in concept design of a divertor remote maintenance port in the EU DEMO reactor using an AHP participative approach

    International Nuclear Information System (INIS)

    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.

  10. European DEMO BOT Solid Breeder Blanket: the concept based on the use of cooling plates and beds of beryllium and Li4SiO4 pebbles

    International Nuclear Information System (INIS)

    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.)

  11. 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

    DEFF Research Database (Denmark)

    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....

  12. A novel wireless power and data transmission AC to DC converter for an implantable device.

    Science.gov (United States)

    Liu, Jhao-Yan; Tang, Kea-Tiong

    2013-01-01

    This article presents a novel AC to DC converter implemented by standard CMOS technology, applied for wireless power transmission. This circuit combines the functions of the rectifier and DC to DC converter, rather than using the rectifier to convert AC to DC and then supplying the required voltage with regulator as in the transitional method. This modification can reduce the power consumption and the area of the circuit. This circuit also transfers the loading condition back to the external circuit by the load shift keying(LSK), determining if the input power is not enough or excessive, which increases the efficiency of the total system. The AC to DC converter is fabricated with the TSMC 90nm CMOS process. The circuit area is 0.071mm(2). The circuit can produce a 1V DC voltage with maximum output current of 10mA from an AC input ranging from 1.5V to 2V, at 1MHz to 10MHz.

  13. Electrodeformation of multi-bilayer spherical concentric membranes by AC electric fields

    Science.gov (United States)

    Lira-Escobedo, J.; Arauz-Lara, J.; Aranda-Espinoza, H.; Adlerz, K.; Viveros-Mendez, P. X.; Aranda-Espinoza, S.

    2017-09-01

    It is now well established that external stresses alter the behaviour of cells, where such alterations can be as profound as changes in gene expression. A type of stresses of particular interest are those due to alternating-current (AC) electric fields. The effect of AC fields on cells is still not well understood, in particular it is not clear how these fields affect the cell nucleus and other organelles. Here, we propose that one possible mechanism is through the deformation of the membranes. In order to investigate the effect of AC fields on the morphological changes of the cell organelles, we modelled the cell as two concentric bilayer membranes. This model allows us to obtain the deformations induced by the AC field by balancing the elastic energy and the work done by the Maxwell stresses. Morphological phase diagrams are obtained as a function of the frequency and the electrical properties of the media and membranes. We demonstrate that the organelle shapes can be changed without modifying the shape of the external cell membrane and that the organelle deformation transitions can be used to measure, for example, the conductivity of the nucleus.

  14. Performance of tungsten-based materials and components under ITER and DEMO relevant steady-state thermal loads

    Energy Technology Data Exchange (ETDEWEB)

    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

  15. Performance of tungsten-based materials and components under ITER and DEMO relevant steady-state thermal loads

    International Nuclear Information System (INIS)

    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

  16. Transcranial Alternating Current Stimulation (tACS Mechanisms and Protocols

    Directory of Open Access Journals (Sweden)

    Amir V. Tavakoli

    2017-09-01

    Full Text Available Perception, cognition and consciousness can be modulated as a function of oscillating neural activity, while ongoing neuronal dynamics are influenced by synaptic activity and membrane potential. Consequently, transcranial alternating current stimulation (tACS may be used for neurological intervention. The advantageous features of tACS include the biphasic and sinusoidal tACS currents, the ability to entrain large neuronal populations, and subtle control over somatic effects. Through neuromodulation of phasic, neural activity, tACS is a powerful tool to investigate the neural correlates of cognition. The rapid development in this area requires clarity about best practices. Here we briefly introduce tACS and review the most compelling findings in the literature to provide a starting point for using tACS. We suggest that tACS protocols be based on functional brain mechanisms and appropriate control experiments, including active sham and condition blinding.

  17. Practical Observations of the Transit of Venus

    Indian Academy of Sciences (India)

    Home; Journals; Resonance – Journal of Science Education; Volume 9; Issue 5. Practical Observations of the Transit of Venus. B S Shyalaja. Classroom Volume 9 Issue 5 May 2004 pp 79-83. Fulltext. Click here to view fulltext PDF. Permanent link: https://www.ias.ac.in/article/fulltext/reso/009/05/0079-0083 ...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  19. On the optimization of the first wall of the DEMO water-cooled lithium lead outboard breeding blanket equatorial module

    Energy Technology Data Exchange (ETDEWEB)

    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

  20. Multifarious Physics Analyses of the Core Plasma Properties in a Helical DEMO Reactor FFHR-d1

    Energy Technology Data Exchange (ETDEWEB)

    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

  1. DemoMinga: O compromisso social da extensão na área da saúde

    Directory of Open Access Journals (Sweden)

    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.

  2. On the optimization of the first wall of the DEMO water-cooled lithium lead outboard breeding blanket equatorial module

    International Nuclear Information System (INIS)

    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

  3. Bifurcation theory of ac electric arcing

    International Nuclear Information System (INIS)

    Christen, Thomas; Peinke, Emanuel

    2012-01-01

    The performance of alternating current (ac) electric arcing devices is related to arc extinction or its re-ignition at zero crossings of the current (so-called ‘current zero’, CZ). Theoretical investigations thus usually focus on the transient behaviour of arcs near CZ, e.g. by solving the modelling differential equations in the vicinity of CZ. This paper proposes as an alternative approach to investigate global mathematical properties of the underlying periodically driven dynamic system describing the electric circuit containing the arcing device. For instance, the uniqueness of the trivial solution associated with the insulating state indicates the extinction of any arc. The existence of non-trivial attractors (typically a time-periodic state) points to a re-ignition of certain arcs. The performance regions of arcing devices, such as circuit breakers and arc torches, can thus be identified with the regions of absence and existence, respectively, of non-trivial attractors. Most important for applications, the boundary of a performance region in the model parameter space is then associated with the bifurcation of the non-trivial attractors. The concept is illustrated for simple black-box arc models, such as the Mayr and the Cassie model, by calculating for various cases the performance boundaries associated with the bifurcation of ac arcs. (paper)

  4. A nonlinear model for AC induced corrosion

    Directory of Open Access Journals (Sweden)

    N. Ida

    2012-09-01

    Full Text Available The modeling of corrosion poses particular difficulties. The understanding of corrosion as an electrochemical process has led to simple capacitive-resistive models that take into account the resistance of the electrolytic cell and the capacitive effect of the surface potential at the interface between conductors and the electrolyte. In some models nonlinear conduction effects have been added to account for more complex observed behavior. While these models are sufficient to describe the behavior in systems with cathodic protection, the behavior in the presence of induced AC currents from power lines and from RF sources cannot be accounted for and are insufficient to describe the effects observed in the field. Field observations have shown that a rectifying effect exists that affects the cathodic protection potential and this effect is responsible for corrosion in the presence of AC currents. The rectifying effects of the metal-corrosion interface are totally missing from current models. This work proposes a nonlinear model based on finite element analysis that takes into account the nonlinear behavior of the metal-oxide interface and promises to improve modeling by including the rectification effects at the interface.

  5. Measuring Gravitational Flexion in ACS Clusters

    Science.gov (United States)

    Goldberg, David

    2005-07-01

    We propose measurement of the gravitational "Flexion" signal in ACS cluster images. The flexion, or "arciness" of a lensed background galaxy arises from variations in the lensing field. As a result, it is extremely sensitive to small scale perturbations in the field, and thus, to substructure in clusters. Moreover, because flexion represents gravitationally induced asymmetries in the lensed image, it is completely separable from traditional measurements of shear, which focus on the induced ellipticity of the image, and thus, the two signals may be extracted simultaneously. Since typical galaxies are roughly symmetric upon 180 degree rotation, even a small induced flexion can potentially produce a noticeable effect {Goldberg & Bacon, 2005}. We propose the measurement of substructure within approximately 4 clusters with high-quality ACS data, and will further apply a test of a new tomographic technique whereby comparisons of lensed arcs at different redshifts may be used to estimate the background cosmology, and thus place constraints on the equation of state of dark energy.

  6. Deletion of the AcMNPV core gene ac109 results in budded virions that are non-infectious

    International Nuclear Information System (INIS)

    Fang Minggang; Nie, Yingchao; Theilmann, David A.

    2009-01-01

    Autographa californica multiple nucleopolyhedrovirus (AcMNPV) ac109 is a core gene and its function in the virus life cycle is unknown. To determine its role in the baculovirus life cycle, we used the AcMNPV bacmid system to generate an ac109 deletion virus (vAc 109KO ). Fluorescence and light microscopy showed that transfection of vAc 109KO results in a single-cell infection phenotype. Viral DNA replication is unaffected and the development of occlusion bodies in vAc 109KO -transfected cells evidenced progression to the very late phases of viral infection. Western blot and confocal immunofluorescence analysis showed that AC109 is expressed in the cytoplasm and nucleus throughout infection. In addition, AC109 is a structural protein as it was detected in both budded virus (BV) and occlusion derived virus in both the envelope and nucleocapsid fractions. Titration assays by qPCR and TCID 50 showed that vAc 109KO produced BV but the virions are non-infectious. The vAc 109KO BV were indistinguishable from the BV of repaired and wild type control viruses as determined by negative staining and electron microscopy.

  7. Automatic Control Systems (ACS for Generation and Sale of Electric Power Under Conditions of Industry-Sector Liberalization

    Directory of Open Access Journals (Sweden)

    Yu. S. Petrusha

    2013-01-01

    Full Text Available Possible risks pertaining to transition of electric-power industry to market relations have been considered in the paper. The paper presents an integrated ACS for generation and sale of electric power as an improvement of methodology for organizational and technical management. The given system is based on integration of operating Automatic Dispatch Control System (ADCS and developing Automatic Electricity Meter Reading System (AEMRS. The paper proposes to form an inter-branch sector of ACS PLC (Automatic Control System for Prolongation of Life Cycle users which is oriented on provision of development strategy.

  8. Modeling and reliability analysis of three phase z-source AC-AC converter

    Directory of Open Access Journals (Sweden)

    Prasad Hanuman

    2017-12-01

    Full Text Available This paper presents the small signal modeling using the state space averaging technique and reliability analysis of a three-phase z-source ac-ac converter. By controlling the shoot-through duty ratio, it can operate in buck-boost mode and maintain desired output voltage during voltage sag and surge condition. It has faster dynamic response and higher efficiency as compared to the traditional voltage regulator. Small signal analysis derives different control transfer functions and this leads to design a suitable controller for a closed loop system during supply voltage variation. The closed loop system of the converter with a PID controller eliminates the transients in output voltage and provides steady state regulated output. The proposed model designed in the RT-LAB and executed in a field programming gate array (FPGA-based real-time digital simulator at a fixedtime step of 10 μs and a constant switching frequency of 10 kHz. The simulator was developed using very high speed integrated circuit hardware description language (VHDL, making it versatile and moveable. Hardware-in-the-loop (HIL simulation results are presented to justify the MATLAB simulation results during supply voltage variation of the three phase z-source ac-ac converter. The reliability analysis has been applied to the converter to find out the failure rate of its different components.

  9. AC Electric Field Activated Shape Memory Polymer Composite

    Science.gov (United States)

    Kang, Jin Ho; Siochi, Emilie J.; Penner, Ronald K.; Turner, Travis L.

    2011-01-01

    Shape memory materials have drawn interest for applications like intelligent medical devices, deployable space structures and morphing structures. Compared to other shape memory materials like shape memory alloys (SMAs) or shape memory ceramics (SMCs), shape memory polymers (SMPs) have high elastic deformation that is amenable to tailored of mechanical properties, have lower density, and are easily processed. However, SMPs have low recovery stress and long response times. A new shape memory thermosetting polymer nanocomposite (LaRC-SMPC) was synthesized with conductive fillers to enhance its thermo-mechanical characteristics. A new composition of shape memory thermosetting polymer nanocomposite (LaRC-SMPC) was synthesized with conductive functionalized graphene sheets (FGS) to enhance its thermo-mechanical characteristics. The elastic modulus of LaRC-SMPC is approximately 2.7 GPa at room temperature and 4.3 MPa above its glass transition temperature. Conductive FGSs-doped LaRC-SMPC exhibited higher conductivity compared to pristine LaRC SMP. Applying an electric field at between 0.1 Hz and 1 kHz induced faster heating to activate the LaRC-SMPC s shape memory effect relative to applying DC electric field or AC electric field at frequencies exceeding1 kHz.

  10. AC and DC electrical properties of graphene nanoplatelets reinforced epoxy syntactic foam

    Science.gov (United States)

    Zegeye, Ephraim; Wicker, Scott; Woldesenbet, Eyassu

    2018-04-01

    Benefits of employing graphene nanopletlates (GNPLs) in composite structures include mechanical as well as multifunctional properties. Understanding the impedance behavior of GNPLs reinforced syntactic foams may open new applications for syntactic foam composites. In this work, GNPLs reinforced syntactic foams were fabricated and tested for DC and AC electrical properties. Four sets of syntactic foam samples containing 0, 0.1, 0.3, and 0.5 vol% of GNPLs were fabricated and tested. Significant increase in conductivity of syntactic foams due to the addition of GNPLs was noted. AC impedance measurements indicated that the GNPLs syntactic foams become frequency dependent as the volume fraction of GNPLs increases. With addition of GNPLs, the characteristic of the syntactic foams are also observed to transition from dominant capacitive to dominant resistive behavior. This work was carried out at Southern University, Mechanical Engineering Department, Baton Rouge, LA 70802, United States of America.

  11. Computer aided design of operational units for tritium recovery from Li17Pb83 blanket of a DEMO fusion reactor

    International Nuclear Information System (INIS)

    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

  12. The concept of system for chips production need to work demo CHP plant in company 'AGROSAVA' from Šimanovci

    Directory of Open Access Journals (Sweden)

    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.

  13. Phase change of First Wall in Water-Cooled Breeding Blankets of K-DEMO for Vertical

    Energy Technology Data Exchange (ETDEWEB)

    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.

  14. First spaceborne phase altimetry over sea ice using TechDemoSat-1 GNSS-R signals

    Science.gov (United States)

    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

    International Nuclear Information System (INIS)

    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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  17. Neutronic design studies of a conceptual DCLL fusion reactor for a DEMO and a commercial power plant

    Science.gov (United States)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Directory of Open Access Journals (Sweden)

    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.

  20. Design of a Multi-Tube Pd-Membrane Module for Tritium Recovery from He in DEMO

    Directory of Open Access Journals (Sweden)

    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.

  1. Supporting Transition

    Science.gov (United States)

    Qureshi, Asima; Petrucco, James

    2018-01-01

    Meadowbrook Primary School has explored the use of The Teacher Assessment in Primary Science (TAPS) to support transition, initially for transfer to secondary school and now for transition from Early Years Foundation Stage (EYFS) into Key Stage 1 (ages 5-7). This article will consider an example of a secondary transition project and discuss the…

  2. AC losses in high Tc superconductors

    International Nuclear Information System (INIS)

    Campbell, A.M.

    1998-01-01

    Full text: Although in principle the AC losses in high Tc superconductors can be calculated from the critical current density, a number of complications make this difficult. The Jc is very field dependent, there are intergranular and intragranular critical currents, the material is anisotropic and there is usually a large demagnetising factor. Care must be taken in interpreting electrical measurements since the voltage depends on the position of the contacts. In spite of these complications the simple theory of Norris has proved surprisingly successful and arguments will be presented as to why this is the case. Results on a range of tapes will be compared with theory and numerical methods for predicting losses discussed. Finally a theory for coupling losses will be given for a composite conductor with high resistance barriers round the filaments

  3. Transcranial alternating current stimulation (tACS

    Directory of Open Access Journals (Sweden)

    Andrea eAntal

    2013-06-01

    Full Text Available Transcranial alternating current stimulation (tACS seems likely to open a new era of the field of noninvasive electrical stimulation of the human brain by directly interfering with cortical rhythms. It is expected to synchronize (by one single resonance frequency or desynchronize (e.g. by the application of several frequencies cortical oscillations. If applied long enough it may cause neuroplastic effects. In the theta range it may improve cognition when applied in phase. Alpha rhythms could improve motor performance, whereas beta intrusion may deteriorate them. TACS with both alpha and beta frequencies has a high likelihood to induce retinal phosphenes. Gamma intrusion can possibly interfere with attention. Stimulation in the ripple range induces intensity dependent inhibition or excitation in the motor cortex most likely by entrainment of neuronal networks, whereas stimulation in the low kHz range induces excitation by neuronal membrane interference. TACS in the 200 kHz range may have a potential in oncology.

  4. Ac loss measurement of SSC dipole magnets

    International Nuclear Information System (INIS)

    Delchamps, S.; Hanft, R.; Jaffery, T.; Kinney, W.; Koska, W.; Lamm, M.J.; Mazur, P.O.; Orris, D.; Ozelis, J.P.; Strait, J.; Wake, M.

    1992-09-01

    AC losses in full length and 1.5 m model SSC collider dipoles were successfully measured by the direct observation of energy flow into and out of magnets during a ramp cycle. The measurement was performed by using two double-integrating type digital volt meters (DVM's) for current and voltage measurement. Measurements were performed for six is m long ASST magnets and five 1.5 m long model magnets, inducting one 40 mm diameter magnet. There were large variations in the eddy current losses. Since these magnets use conductors with slight deviations in their internal structures and processing of the copper surface depending on the manufacturer, it is likely that there are differences in the contact resistance between strands. Correlation between the ramp rate dependence of the,quench current and the eddy current loss was evident

  5. Ac irreversibility line of bismuth-based high temperature superconductors

    Energy Technology Data Exchange (ETDEWEB)

    Mehdaoui, A. [Laboratoire de Physique et de Spectroscopie Electronique, URA 1435 Faculte des Sciences, Universite de Haute Alsace 4, rue des Freres Lumiere, 68093 Mulhouse Cedex (France); Beille, J. [Laboratoire Louis Neel, CNRS, BP 166, 38042 Grenoble Cedex 9 (France); Berling, D.; Loegel, B. [Laboratoire de Physique et de Spectroscopie Electronique, URA 1435 Faculte des Sciences, Universite de Haute Alsace 4, rue des Freres Lumiere, 68093 Mulhouse Cedex (France); Noudem, J.G.; Tournier, R. [EPM-MATFORMAG, Laboratoire dElaboration par Procede Magnetique, CNRS, BP 166, 38042 Grenoble Cedex 9 (France)

    1997-09-01

    We discuss the magnetic properties of lead doped Bi-2223 bulk samples obtained through combined magnetic melt texturing and hot pressing (MMTHP). The ac complex susceptibility measurements are achieved over a broad ac field range (1 Oe{lt}h{sub ac}{lt}100 Oe) and show highly anisotropic properties. The intergranular coupling is improved in the direction perpendicular to the applied stress and magnetic field direction, and an intragranular loss peak is observed for the first time. A comparison is made with other bismuth-based compounds and it is shown that the MMTHP process shifts the ac irreversibility line (ac IL) toward higher fields. It is also shown that all the ac IL{close_quote}s for quasi 2D bismuth-based compounds show a nearly quadratic temperature dependence and deviate therefore strongly from the linear behavior observed in quasi 3D compounds and expected from a critical state model.{copyright} {ital 1997 Materials Research Society.}

  6. Ac irreversibility line of bismuth-based high temperature superconductors

    International Nuclear Information System (INIS)

    Mehdaoui, A.; Beille, J.; Berling, D.; Loegel, B.; Noudem, J.G.; Tournier, R.

    1997-01-01

    We discuss the magnetic properties of lead doped Bi-2223 bulk samples obtained through combined magnetic melt texturing and hot pressing (MMTHP). The ac complex susceptibility measurements are achieved over a broad ac field range (1 Oe ac <100 Oe) and show highly anisotropic properties. The intergranular coupling is improved in the direction perpendicular to the applied stress and magnetic field direction, and an intragranular loss peak is observed for the first time. A comparison is made with other bismuth-based compounds and it is shown that the MMTHP process shifts the ac irreversibility line (ac IL) toward higher fields. It is also shown that all the ac IL close-quote s for quasi 2D bismuth-based compounds show a nearly quadratic temperature dependence and deviate therefore strongly from the linear behavior observed in quasi 3D compounds and expected from a critical state model.copyright 1997 Materials Research Society

  7. AC conductivity and dielectric behavior of bulk Furfurylidenemalononitrile

    Science.gov (United States)

    El-Nahass, M. M.; Ali, H. A. M.

    2012-06-01

    AC conductivity and dielectric behavior for bulk Furfurylidenemalononitrile have been studied over a temperature range (293-333 K) and frequency range (50-5×106 Hz). The frequency dependence of ac conductivity, σac, has been investigated by the universal power law, σac(ω)=Aωs. The variation of the frequency exponent (s) with temperature was analyzed in terms of different conduction mechanisms, and it was found that the correlated barrier hopping (CBH) model is the predominant conduction mechanism. The temperature dependence of σac(ω) showed a linear increase with the increase in temperature at different frequencies. The ac activation energy was determined at different frequencies. Dielectric data were analyzed using complex permittivity and complex electric modulus for bulk Furfurylidenemalononitrile at various temperatures.

  8. Magnetic irreversibility in granular superconductors: ac susceptibility study

    International Nuclear Information System (INIS)

    Perez, F.; Obradors, X.; Fontcuberta, J.; Vallet, M.; Gonzalez-Calbet, J.

    1991-01-01

    Ac susceptibility measurements of a ceramic weak-coupled superconductor in very low ac fields (2mG, 111Hz) are reported. We present evidence for the observation of the magnetic irreversibility following a ZFC-FC thermal cycling by means of ac susceptibilty measurements. It is shown that this technique also reflect local magnetic field effects in granular superconductors, as previously suggested in microwave surface resistance and I-V characteristics. (orig.)

  9. AC Own Motion Percentage of Randomly Sampled Cases

    Data.gov (United States)

    Social Security Administration — Longitudinal report detailing the numbers and percentages of Appeals Council (AC) own motion review actions taken on un-appealed favorable hearing level decisions...

  10. AC electric motors control advanced design techniques and applications

    CERN Document Server

    Giri, Fouad

    2013-01-01

    The complexity of AC motor control lies in the multivariable and nonlinear nature of AC machine dynamics. Recent advancements in control theory now make it possible to deal with long-standing problems in AC motors control. This text expertly draws on these developments to apply a wide range of model-based control designmethods to a variety of AC motors. Contributions from over thirty top researchers explain how modern control design methods can be used to achieve tight speed regulation, optimal energetic efficiency, and operation reliability and safety, by considering online state var

  11. Improved Design Methods for Robust Single- and Three-Phase ac-dc-ac Power Converters

    DEFF Research Database (Denmark)

    Qin, Zian

    . The approaches for improving their performance, in terms of the voltage stress, efficiency, power density, cost, loss distribution, and temperature, will be studied. The structure of the thesis is as follows, Chapter 1 presents the introduction and motivation of the whole project as well as the background...... becomes a emerging challenge. Accordingly, installation of sustainable power generators like wind turbines and solar panels has experienced a large increase during the last decades. Meanwhile, power electronics converters, as interfaces in electrical system, are delivering approximately 80 % electricity...... back-to-back, and meanwhile improve the harmonics, control flexibility, and thermal distribution between the switches. Afterwards, active power decoupling methods for single-phase inverters or rectifiers that are similar to the single-phase ac-dc-ac converter, are studied in Chapter 4...

  12. Wind-powered asynchronous AC/DC/AC converter system. [for electric power supply regulation

    Science.gov (United States)

    Reitan, D. K.

    1973-01-01

    Two asynchronous ac/dc/ac systems are modelled that utilize wind power to drive a variable or constant hertz alternator. The first system employs a high power 60-hertz inverter tie to the large backup supply of the power company to either supplement them from wind energy, storage, or from a combination of both at a preset desired current; rectifier and inverter are identical and operate in either mode depending on the silicon control rectifier firing angle. The second system employs the same rectification but from a 60-hertz alternator arrangement; it provides mainly dc output, some sinusoidal 60-hertz from the wind bus and some high harmonic content 60-hertz from an 800-watt inverter.

  13. Transition radiation and transition scattering

    International Nuclear Information System (INIS)

    Ginzburg, V.L.

    1982-01-01

    Transition radiation is a process of a rather general character. It occurs when some source, which does not have a proper frequency (for example, a charge) moves at a constant velocity in an inhomogeneous and (or) nonstationary medium or near such a medium. The simplest type of transition radiation takes place when a charge crosses a boundary between two media (the role of one of the media may be played by vacuum). In the case of periodic variation of the medium, transition radiation possesses some specific features (resonance transition radiation or transition scattering). Transition scattering occurs, in particular, when a permittivity wave falls onto an nonmoving (fixed) charge. Transition scattering is closely connected with transition bremsstrahlung radiation. All these transition processes are essential for plasma physics. Transition radiation and transition scattering have analogues outside the framework of electrodynamics (like in the case of Vavilov-Cherenkov radiation). In the present report the corresponding range of phenomena is elucidated, as far as possible, in a generally physical aspect. (Auth.)

  14. The AC Stark Effect, Time-Dependent Born-Oppenheimer Approximation, and Franck-Condon Factors

    CERN Document Server

    Hagedorn, G A; Jilcott, S W

    2005-01-01

    We study the quantum mechanics of a simple molecular system that is subject to a laser pulse. We model the laser pulse by a classical oscillatory electric field, and we employ the Born--Oppenheimer approximation for the molecule. We compute transition amplitudes to leading order in the laser strength. These amplitudes contain Franck--Condon factors that we compute explicitly to leading order in the Born--Oppenheimer parameter. We also correct an erroneous calculation in the mathematical literature on the AC Stark effect for molecular systems.

  15. Coordinated control of three-phase AC and DC type EV–ESSs for efficient hybrid microgrid operations

    International Nuclear Information System (INIS)

    Rahman, Md Shamiur; Hossain, M.J.; Lu, Junwei

    2016-01-01

    Highlights: • A coordinated control is proposed for three-phase AC and DC type electric vehicles. • A four-quadrant interlinking converter is designed for hybrid microgrid operations. • Concurrent real irradiation data and commercial load profile are used for testing. • Unbalanced scenario due to single-phase electric vehicle charging is considered. • Improved AC and DC bus voltages and frequency regulations are achieved. - Abstract: This paper presents a three-layered coordinated control to incorporate three-phase (3P) alternating current (AC) and direct current (DC) type electric vehicle energy storage systems (EV–ESSs) for improved hybrid AC/DC microgrid operations. The first layer of the algorithm ensures DC subgrid management by regulating the DC bus voltage and DC side power management. The second and third layer manages AC subgrid by regulating the AC bus voltage and the frequency by managing reactive and active power respectively. The multi-layered coordination is embedded into the microgrid central controller (MGCC) which controls the interlinking controller in between AC and DC microgrid and the interfacing controllers of the participating electric vehicles (EVs) and distributed generation (DG) units. The whole system is designed in MATLAB/SIMULINK® environment resembling the under construction microgrid at Griffith University, Australia. Extensive case studies are performed using real life irradiation data and commercial loads of the campus buildings. Impacts of homogeneous and heterogeneous single-phase EV charging are investigated to observe both balanced and unbalanced scenarios. Synchronization during the transition from the islanded to grid-tied mode is tested considering a contingency situation. From the comparative simulation results it is evident that the proposed controller exhibits effective, reliable and robust performance for all the cases.

  16. Investigation of wetting property between liquid lead lithium alloy and several structural materials for Chinese DEMO reactor

    Science.gov (United States)

    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.

  17. Influence of thermal performance on design parameters of a He/LiPb dual coolant DEMO concept blanket design

    Energy Technology Data Exchange (ETDEWEB)

    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.

  18. Ac and dc motor flooding times

    International Nuclear Information System (INIS)

    Crowley, D.A.; Hinton, J.H.

    1988-01-01

    Reactor safety studies, such as the emergency cooling system (ECS) limits analyses and the probabilistic risk assessment, require that the flood-out times be calculated for the ac and dc motors at the -40 foot level. New calculations are needed because dams of an improved design have been installed between the pump room and motor room, and because updated leak rate calculations have shown that the maximum possible leak rate is larger than that which had been previously calculated. The methodology for calculating the motor flood-out times has also been improved. A computer program has been written to calculate flood-out times for various leak rates and sump pump operabilities. For ECS limits analyses, the worst case dc motor flood-out times are 161 and 297 seconds in LKC and P-areas, respectively. These times are for a 135,468 gpm leak that first flows to the motor room and all of the sump pumps are off

  19. Improving Power Quality in AC Supply Grids

    Directory of Open Access Journals (Sweden)

    Piotr Fabijański

    2015-12-01

    Full Text Available This paper describes a digital and actual model of the UPQC (Unified Power Quality Conditioner integrated system for power quality improvement. The UPQC’s design and its connection to an AC supply grid, 1-phase and 3-phase alike, provide effective compensation of unwanted interferences in the waveforms of load supply voltages and non-linear load currents. This article presents an overview of topologies and control strategies. The study of the UPQC confirmed its positive impact on the power quality. The electricity parameters were significantly improved. Total harmonic distortion in supply voltage THDu decreased six-fold to 1.89%, and total harmonic distortion in load current THDi decreased more than ten-fold to 2.38% for a non-linear load (uncontrolled bridge rectifier with load L. Additionally, symmetrisation of supply voltages and reactive power compensation Q of linear load was obtained. The UPQC integrated system for power quality improvement can be used wherever high-quality and PN-EN 50160 standard – compliant electricity is required.

  20. Neurinoma central do nervo acústico

    Directory of Open Access Journals (Sweden)

    Paulo Pinto Pupo

    1950-03-01

    Full Text Available O autor apresenta o caso de uma paciente com 45 anos, com hipertensão arterial, queixando-se de tonturas e surdez progressiva à esquerda que, ao exame neurológico, apresentava síndrome protuberancial, com hemi-anestesia táctil e dolorosa à direita respeitando a face, hemiparesia direita, ataxia de tipo sensitivo nos membros da direita, paralisia facial de tipo periférico, hipoacusia, paresia de motor ocular externo à esquerda, síndrome vertiginosa e nistagmo horizontal ao olhar para a direita. À necrópsia foi encontrado um tumor na hemicalota protuberancial esquerda e foco malácico adjacente, secundário a distúrbio circulatório. O tumor, intimamente dependente das raízes intraprotuberanciais do nervo acústico, se apresentava com as características histológicas dos neurinomas. Além dessas particularidades, a lesão do feixe central da calota e conseqüente degeneração "hipertrófica" da oliva bulbar constituem outro aspecto de grande interêsse dêste caso.

  1. Cosmic Shear With ACS Pure Parallels

    Science.gov (United States)

    Rhodes, Jason

    2002-07-01

    Small distortions in the shapes of background galaxies by foreground mass provide a powerful method of directly measuring the amount and distribution of dark matter. Several groups have recently detected this weak lensing by large-scale structure, also called cosmic shear. The high resolution and sensitivity of HST/ACS provide a unique opportunity to measure cosmic shear accurately on small scales. Using 260 parallel orbits in Sloan textiti {F775W} we will measure for the first time: beginlistosetlength sep0cm setlengthemsep0cm setlengthopsep0cm em the cosmic shear variance on scales Omega_m^0.5, with signal-to-noise {s/n} 20, and the mass density Omega_m with s/n=4. They will be done at small angular scales where non-linear effects dominate the power spectrum, providing a test of the gravitational instability paradigm for structure formation. Measurements on these scales are not possible from the ground, because of the systematic effects induced by PSF smearing from seeing. Having many independent lines of sight reduces the uncertainty due to cosmic variance, making parallel observations ideal.

  2. Introduction of hvdc transmission into a predominantly ac network

    Energy Technology Data Exchange (ETDEWEB)

    Casson, W; Last, F H; Huddart, K W

    1966-02-01

    Methods for reinforcing the supply network, including systems employing dc links, without introducing a new primary network are briefly described. The arrangement for dc links is outlined and the application to an existing ac system is considered. The economics of ac and dc for reinforcement schemes are briefly mentioned.

  3. Low ac loss geometries in YBCO coated conductors

    International Nuclear Information System (INIS)

    Duckworth, R.C.; List, F.A.; Paranthaman, M.P.; Rupich, M.W.; Zhang, W.; Xie, Y.Y.; Selvamanickam, V.

    2007-01-01

    Reduction of ac losses in applied ac fields can be accomplished through either the creation of filaments and bridging in YBCO coated conductors or by an assembly of narrow width YBCO tapes. The ac losses for each of these geometries were measured at 77 K in perpendicular ac fields up to 100 mT. Despite physical isolation of the filaments, coupling losses were still present in the samples when compared to the expected hysteretic loss. In addition to filamentary conductors the assembly of stacked YBCO conductor provides an alternative method of ac loss reduction. When compared to a 4-mm wide YBCO coated conductor with a critical current of 60 A, the ac loss in a stack of 2-mm wide YBCO coated conductors with a similar total critical current was reduced. While the reduction in ac loss in a 2-mm wide stack coincided with the reduction in the engineering current density of the conductor, further reduction of ac loss was obtained through the splicing of the 2-mm wide tapes with low resistance solders

  4. Low ac loss geometries in YBCO coated conductors

    Energy Technology Data Exchange (ETDEWEB)

    Duckworth, R.C. [Oak Ridge National Laboratory, One Bethel Valley Road, P.O. Box 2008, MS-6305, Oak Ridge, TN 37831-6305 (United States)], E-mail: duckworthrc@ornl.gov; List, F.A.; Paranthaman, M.P. [Oak Ridge National Laboratory, One Bethel Valley Road, P.O. Box 2008, MS-6305, Oak Ridge, TN 37831-6305 (United States); Rupich, M.W.; Zhang, W. [American Superconductor, Two Technology Drive, Westborough, MA 01581 (United States); Xie, Y.Y.; Selvamanickam, V. [SuperPower, 450 Duane Ave, Schenectady, NY 12304 (United States)

    2007-10-01

    Reduction of ac losses in applied ac fields can be accomplished through either the creation of filaments and bridging in YBCO coated conductors or by an assembly of narrow width YBCO tapes. The ac losses for each of these geometries were measured at 77 K in perpendicular ac fields up to 100 mT. Despite physical isolation of the filaments, coupling losses were still present in the samples when compared to the expected hysteretic loss. In addition to filamentary conductors the assembly of stacked YBCO conductor provides an alternative method of ac loss reduction. When compared to a 4-mm wide YBCO coated conductor with a critical current of 60 A, the ac loss in a stack of 2-mm wide YBCO coated conductors with a similar total critical current was reduced. While the reduction in ac loss in a 2-mm wide stack coincided with the reduction in the engineering current density of the conductor, further reduction of ac loss was obtained through the splicing of the 2-mm wide tapes with low resistance solders.

  5. Flexible AC transmission systems: the state of the art

    Energy Technology Data Exchange (ETDEWEB)

    Edris, Abdel-Aty [Electric Power Research Inst., Palo Alto, CA (United States). Electric Systems Division

    1994-12-31

    Flexible AC transmission systems (FACTS) is a concept promoting the use of power electronic controllers to enhance the controllability and usable capacity of AC transmission. This paper presents the state of the art of FACTS and the status of the current projects for the application of the FACTS controllers in transmission systems. (author) 8 refs., 8 figs.

  6. Ammonia treated Mo/AC catalysts for CO hydrogenation with ...

    Indian Academy of Sciences (India)

    SHARIF F ZAMAN

    the influence of acid treated AC as a support with K-Ni-. Mo active ... K-Ni-Mo/AC catalyst was more selective to oxygenates. (>40% ... mineral impurities (K, Si, Sn and Fe) <1%. ...... edge technical support with thanks Science and Technology.

  7. ELECTRONIC SYSTEM FOR EXPERIMENTATION IN AC ELECTROGRAVIMETRY I: TECHNIQUE FUNDAMENTALS

    Directory of Open Access Journals (Sweden)

    Róbinson Torres

    Full Text Available Basic fundamentals of AC electrogravimetry are introduced. Their main requirements and characteristics are detailed to establish the design of an electronic system that allows the appropriate extraction of data needed to determine the electrogravimetric transfer function (EGTF and electrochemical impedance (EI, in an experimental set-up for the AC electrogravimetry technique.

  8. Operation of AC Adapters Visualized Using Light-Emitting Diodes

    Science.gov (United States)

    Regester, Jeffrey

    2016-01-01

    A bridge rectifier is a diamond-shaped configuration of diodes that serves to convert alternating current(AC) into direct current (DC). In our world of AC outlets and DC electronics, they are ubiquitous. Of course, most bridge rectifiers are built with regular diodes, not the light-emitting variety, because LEDs have a number of disadvantages. For…