WorldWideScience

Sample records for closed divertor configuration

  1. TCV divertor upgrade for alternative magnetic configurations

    Directory of Open Access Journals (Sweden)

    H. Reimerdes

    2017-08-01

    Full Text Available The Swiss Plasma Center (SPC is planning a divertor upgrade for the TCV tokamak. The upgrade aims at extending the research of conventional and alternative divertor configurations to operational scenarios and divertor regimes of greater relevance for a fusion reactor. The main elements of the upgrade are the installation of an in-vessel structure to form a divertor chamber of variable closure and enhanced diagnostic capabilities, an increase of the pumping capability of the divertor chamber and the addition of new divertor poloidal field coils. The project follows a staged approach and is carried out in parallel with an upgrade of the TCV heating system. First calculations using the EMC3-Eirene code indicate that realistic baffles together with the planned heating upgrade will allow for a significantly higher compression of neutral particles in the divertor, which is a prerequisite to test the power dissipation potential of various divertor configurations.

  2. Advanced divertor configurations with large flux expansion

    NARCIS (Netherlands)

    Soukhanovskii, V. A.; R.E. Bell,; Diallo, A.; S. Gerhardt,; S. Kaye,; E. Kolemen,; B.P. LeBlanc,; McLean, A.; Menard, J. E.; S.F. Paul,; Podesta, M.; Raman, R.; D.D. Ryutov,; F. Scotti,; Kaita, R.; Maingi, R.; D.M. Mueller,; Roquemore, A. L.; Reimerdes, H.; G.P. Canal,; Labit, B.; Vijvers, W.; Coda, S.; Duval, B. P.; Morgan, T.; Zielinski, J.; De Temmerman, G.; Tal, B.

    2013-01-01

    Experimental studies of the novel snowflake divertor concept (D. Ryutov, Phys. Plasmas 14 (2007) 064502) performed in the NSTX and TCV tokamaks are reviewed in this paper. The snowflake divertor enables power sharing between divertor strike points, as well as the divertor plasma-wetted area,

  3. The isotope effect on divertor conditions and neutral pumping in horizontal divertor configurations in JET-ILW Ohmic plasmas

    Directory of Open Access Journals (Sweden)

    J. Uljanovs

    2017-08-01

    Full Text Available Understanding the impact of isotope mass and divertor configuration on the divertor conditions and neutral pressures is critical for predicting the performance of the ITER divertor in DT operation. To address this need, ohmically heated hydrogen and deuterium plasma experiments were conducted in JET with the ITER-like wall in varying divertor configurations. In this study, these plasmas are simulated with EDGE2D-EIRENE outfitted with a sub-divertor model, to predict the neutral pressures in the plenum with similar fashion to the experiments. EDGE2D-EIRENE predictions show that the increased isotope mass results in up to a 25% increase in peak electron densities and 15% increase in peak ion saturation current at the outer target in deuterium when compared to hydrogen for all horizontal divertor configurations. Indicating that a change from hydrogen to deuterium as main fuel decreases the neutral mean free path, leading to higher neutral density in the divertor. Consequently, this mechanism also leads to higher neutral pressures in the sub-divertor. The experimental data provided by the hydrogen and deuterium ohmic discharges shows that closer proximity of the outer strike point to the pumping plenum results in a higher neutral pressure in the sub-divertor. The diaphragm capacitance gauge pressure measurements show that a two to three-fold increase in sub-divertor pressure was achieved in the corner and nearby horizontal configurations compared to the far-horizontal configurations, likely due to ballistic transport (with respect to the plasma facing components of the neutrals into the sub-divertor. The corner divertor configuration also indicates that a neutral expansion occurs during detachment, resulting in a sub-divertor neutral density plateau as a function of upstream density at the outer-mid plane.

  4. Divertor asymmetry and scrape-off layer flow in various divertor configurations in Experimental Advanced Superconducting Tokamak

    DEFF Research Database (Denmark)

    Liu, S. C.; Guo, H. Y.; Xu, Guandong

    2012-01-01

    plasmas exhibit the usual in-out asymmetry in particle and heat fluxes in LSN with the ion del B direction toward the lower X-point, favoring the outer divertor, especially at high density. The in-out asymmetry is reversed when changing the divertor configuration from LSN to USN, thus clearly...

  5. A review of radiative detachment studies in tokamak advanced magnetic divertor configurations

    Science.gov (United States)

    Soukhanovskii, V. A.

    2017-06-01

    The present vision for a plasma-material interface in the tokamak is an axisymmetric poloidal magnetic X-point divertor. Four tasks are accomplished by the standard poloidal X-point divertor: plasma power exhaust; particle control (D/T and He pumping); reduction of impurity production (source); and impurity screening by the divertor scrape-off layer. A low-temperature, low heat flux divertor operating regime called radiative detachment is viewed as the main option that addresses these tasks for present and future tokamaks. Advanced magnetic divertor configuration has the capability to modify divertor parallel and cross-field transport, radiative and dissipative losses, and detachment front stability. Advanced magnetic divertor configurations are divided into four categories based on their salient qualitative features: (1) multiple standard X-point divertors; (2) divertors with higher order nulls; (3) divertors with multiple X-points; and (4) long poloidal leg divertors (and also with multiple X-points). This paper reviews experiments and modeling in the area of radiative detachment in the advanced magnetic divertor configurations.

  6. Plasma transport in a simulated magnetic-divertor configuration

    Energy Technology Data Exchange (ETDEWEB)

    Strawitch, C. M.

    1981-03-01

    The transport properties of plasma on magnetic field lines that intersect a conducting plate are studied experimentally in the Wisconsin internal ring D.C. machine. The magnetic geometry is intended to simulate certain aspects of plasma phenomena that may take place in a tokamak divertor. It is found by a variety of measurements that the cross field transport is non-ambipolar; this may have important implications in heat loading considerations in tokamak divertors. The undesirable effects of nonambipolar flow make it preferable to be able to eliminate it. However, we find that though the non-ambipolarity may be reduced, it is difficult to eliminate entirely. The plasma flow velocity parallel to the magnetic field is found to be near the ion acoustic velocity in all cases. The experimental density and electron temperature profiles are compared to the solutions to a one dimensional transport model that is commonly used in divertor theory.

  7. 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.

  8. Results from recent detachment experiments in alternative divertor configurations on TCV

    Science.gov (United States)

    Theiler, C.; Lipschultz, B.; Harrison, J.; Labit, B.; Reimerdes, H.; Tsui, C.; Vijvers, W. A. J.; Boedo, J. A.; Duval, B. P.; Elmore, S.; Innocente, P.; Kruezi, U.; Lunt, T.; Maurizio, R.; Nespoli, F.; Sheikh, U.; Thornton, A. J.; van Limpt, S. H. M.; Verhaegh, K.; Vianello, N.; the TCV Team; the EUROfusion MST1 Team

    2017-07-01

    Divertor detachment is explored on the TCV tokamak in alternative magnetic geometries. Starting from typical TCV single-null shapes, the poloidal flux expansion at the outer strikepoint is varied by a factor of 10 to investigate the X-divertor characteristics, and the total flux expansion is varied by 70 % to study the properties of the super-X divertor. The effect of an additional X-point near the target is investigated in X-point target divertors. Detachment of the outer target is studied in these plasmas during Ohmic density ramps and with the ion \

  9. Results from recent detachment experiments in alternative divertor configurations on TCV

    NARCIS (Netherlands)

    Theiler, C.; Lipschultz, B.; Harrison, J.; Labit, B.; Reimerdes, H.; Tsui, C.; Vijvers, W. A. J.; Boedo, J. A.; Duval, B. P.; Elmore, S.; Innocente, P.; Kruezi, U.; Lunt, T.; Maurizio, R.; Nespoli, F.; Sheikh, U.; Thornton, A. J.; van Limpt, S. H. M.; Verhaegh, K.; Vianello, N.; TCV team,; EUROfusion MST1 Team,

    2017-01-01

    Divertor detachment is explored on the TCV tokamak in alternative magnetic geometries. Starting from typical TCV single-null shapes, the poloidal flux expansion at the outer strikepoint is varied by a factor of 10 to investigate the X-divertor characteristics, and the total flux expansion is varied

  10. Atomic and molecular processes in JT-60U divertor plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Takenaga, H.; Shimizu, K.; Itami, K. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

    1997-01-01

    Atomic and molecular data are indispensable for the understanding of the divertor characteristics, because behavior of particles in the divertor plasma is closely related to the atomic and molecular processes. In the divertor configuration, heat and particles escaping from the main plasma flow onto the divertor plate along the magnetic field lines. In the divertor region, helium ash must be effectively exhausted, and radiation must be enhanced for the reduction of the heat load onto the divertor plate. In order to exhaust helium ash effectively, the difference between behavior of neutral hydrogen (including deuterium and tritium) and helium in the divertor plasma should be understood. Radiation from the divertor plasma generally caused by the impurities which produced by the erosion of the divertor plate and/or injected by gas-puffing. Therefore, it is important to understand impurity behavior in the divertor plasma. The ions hitting the divertor plate recycle through the processes of neutralization, reflection, absorption and desorption at the divertor plates and molecular dissociation, charge-exchange reaction and ionization in the divertor plasma. Behavior of hydrogen, helium and impurities in the divertor plasmas can not be understood without the atomic and molecular data. In this report, recent results of the divertor study related to the atomic and molecular processes in JT-60U were summarized. Behavior of neural deuterium and helium was discussed in section 2. In section 3, the comparisons between the modelling of the carbon impurity transport and the measurements of C II and C IV were discussed. In section 4, characteristics of the radiative divertor using Ne puffing were reported. The new diagnostic method for the electron density and temperature in the divertor plasmas using the intensity ratios of He I lines was described in section 5. (author)

  11. Numerical study of the connection lengths for various magnetic configurations in Wendelstein 7-X to optimize the heat load on the divertor

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, Priyanjana; Hoelbe, Hauke; Sunn Pedersen, Thomas [Max Planck Institute of Plasma Physics, Greifswald (Germany)

    2016-07-01

    Fusion has the potential to play an important role as a future energy resource. It has the capacity to produce large-scale clean energy. The two main confinement concepts are the tokamak and the stellarator. The W7-X machine is based on stellarator principle and is using special form of coils to achieve steady-state plasma confinement. Divertors are used in tokamaks and stellarator to control the exhaust of waste gases and impurities from the machine. The divertor concept of W7-X is a so-called island divertor. The island chain isolates the confinement core from regions where the plasma-wall interaction takes place. The area of the divertor that receives the main part of the heat loads, the so-called wetted area, increases with the distance along the magnetic field from the outboard midplane to the divertor target. The connection length is relatively short in tokamaks with conventional divertors. In the stellarator island divertor, the connection length can be varied significantly, which should allow for optimization of the wetted area. We present here a numerical study of the achievable connection lengths in various W7-X configurations and discuss the possibilities for running dedicated experiments to understand the physics of what sets the wetted area.

  12. Preliminary comparison of the conventional and quasi-snowflake divertor configurations with the 2D code EDGE2D/EIRENE in the FAST tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Viola, B.; Maddaluno, G.; Pericoli Ridolfini, V. [EURATOM-ENEA Association, C.R. Frascati, Via E. Fermi 45, 00044 Frascati (Rome) (Italy); Corrigan, G.; Harting, D. [Culham Centre of Fusion Energy, EURATOM-Association, Abingdon (United Kingdom); Mattia, M. [Dipartimento di Informatica, Sistemi e Produzione, Universita di Roma, Tor Vergata, Via del Politecnico, 00133 Roma (Italy); Zagorski, R. [Institute of Plasma Physics and Laser Microfusion-EURATOM Association, 01-497 Warsaw (Poland)

    2014-06-15

    The new magnetic configurations for tokamak divertors, snowflake and super-X, proposed to mitigate the problem of the power exhaust in reactors have clearly evidenced the need for an accurate and reliable modeling of the physics governing the interaction with the plates. The initial effort undertaken jointly by ENEA and IPPLM has been focused to exploit a simple and versatile modeling tool, namely the 2D TECXY code, to obtain preliminary comparison between the conventional and snowflake configurations for the proposed new device FAST that should realize an edge plasma with properties quite close to those of a reactor. The very interesting features found for the snowflake, namely a power load mitigation much larger than expected directly from the change of the magnetic topology, has further pushed us to check these results with the more sophisticated computational tool EDGE2D coupled with the neutral code module EIRENE. After a preparatory work that has been carried out in order to adapt this code combination to deal with non-conventional, single null equilibria and in particular with second order nulls in the poloidal field generated in the snowflake configuration, in this paper we describe the first activity to compare these codes and discuss the first results obtained for FAST. The outcome of these EDGE2D runs is in qualitative agreement with those of TECXY, confirming the potential benefit obtainable from a snowflake configuration. (copyright 2014 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  13. Impurity radiation modulations in an ergodic divertor

    Energy Technology Data Exchange (ETDEWEB)

    Laugier, F. E-mail: laugier@pegase.cad.cea.fr; Becoulet, M.; De Michelis, C.; Ghendrih, Ph.; Gunn, J.P.; Monier-Garbet, P.; Reichle, R.; Vallet, J.C

    2001-03-01

    The 3-D geometry of radiation losses is investigated in the Tore Supra ergodic divertor. Measurements from passive bolometers located on the divertor coils show evidence of toroidal and poloidal radiation modulations. They were interpreted using a 3-D code solving heat transport equation that gives the whole geometry of plasma radiation in a divertor configuration close to Tore Supra. The results of the code are in qualitative agreement with the measurements and they show that the total radiated power is underestimated when inferred from standard bolometers located between divertor modules. Maximum of radiation in front of the modules is explained by the multiplication of radiative zones at this place due to the intersection of field lines with the vessel wall. This effect leads to non-monotonic temperature profiles along field lines in the boundary plasma.

  14. Plasma{endash}neutral interaction in tokamak divertor for {open_quote}{open_quote}gas box{close_quote}{close_quote} neutral model

    Energy Technology Data Exchange (ETDEWEB)

    Krasheninnikov, S.I. [Massachusetts Institute of Technology, Plasma Fusion Center, Cambridge, Massachusetts 02139 (United States); Soboleva, T.K. [Instituto de Ciencias Nucleares, UNAM, Mexico D.F. (Mexico)

    1996-06-01

    Plasma flow through the gas cloud in a tokamak divertor for {open_quote}{open_quote}gas box{close_quote}{close_quote} divertor geometry and Knudsen regime of neutral transport is investigated. It is shown that similar to the neutral models that have considered previously, (i) plasma parameters near the target is sensitive to the energy flux into the hydrogen recycling region and can change rapidly, resulting in bifurcation-like behavior, which might be interpreted as a transition to detached regime, (ii) plasma flux onto the target starts to decrease at a very low plasma temperature near the target, while a strong pressure drop already occurs. At low plasma temperature near the target the recombination processes can significantly alter the plasma flux onto the target. {copyright} {ital 1996 American Institute of Physics.}

  15. Monte-Carlo fluid approaches to detached plasmas in non-axisymmetric divertor configurations

    Science.gov (United States)

    Feng, Y.; Frerichs, H.; Kobayashi, M.; Reiter, D.

    2017-03-01

    Fluid transport modeling in three-dimensional boundaries of toroidal confinement devices is reviewed with the emphasis on a Monte-Carlo approach to simulate detached plasmas. The loss of axisymmetry in such configurations presents a major challenge for numerical implementation of the standard fluid model widely applied to fusion experimental devices. A large-scale effort has been made to address this problem under complementary aspects including different magnetic topologies and numerical techniques. In this paper, we give a brief review of the different strategies pioneered and the challenges involved. A more detailed description is provided for the Monte-Carlo code—EMC3-Eirene, where the physics model and the basic idea behind the applied Monte-Carlo method are presented. The focus is put on its applications to detachment studies for stellarators and tokamaks. Here, major achievements and difficulties encountered are described. Model limitations and further development plans are discussed.

  16. Establishment of a low recycling state with full density control by active pumping of the closed helical divertor at LHD

    Science.gov (United States)

    Motojima, G.; Masuzaki, S.; Tanaka, H.; Morisaki, T.; Sakamoto, R.; Murase, T.; Tsuchibushi, Y.; Kobayashi, M.; Schmitz, O.; Shoji, M.; Tokitani, M.; Yamada, H.; Takeiri, Y.; The LHD Experiment Group

    2018-01-01

    Superior control of particle recycling and hence full governance of plasma density has been established in the Large Helical Device (LHD) using largely enhanced active pumping of the closed helical divertor (CHD). In-vessel cryo-sorption pumping systems inside the CHD in five out of ten inner toroidal divertor sections have been developed and installed step by step in the LHD. The total effective pumping speed obtained was 67  ±  5 m3 s‑1 in hydrogen, which is approximately seven times larger than previously obtained. As a result, a low recycling state was observed with CHD pumping for the first time in LHD featuring excellent density control even under intense pellet fueling conditions. A global particle confinement time (τ p* ) is used for comparison of operation with and without the CHD pumping. The τ p* was evaluated from the density decay after the fueling of hydrogen pellet injection or gas puffing in NBI plasmas. A reliably low base density before the fueling and short τ p* after the fueling were obtained during the CHD pumping, demonstrating for the first time full control of the particle balance with active pumping in the CHD.

  17. Analysis of particle transport in a gas target divertor

    Energy Technology Data Exchange (ETDEWEB)

    Ohtsu, Shigeki; Tanaka, Satoru [Tokyo Univ. (Japan). Faculty of Engineering

    1996-10-01

    2-dimensional modelling of divertor plasma was performed with three types of the divertor geometry configuration. Pumping is effective to reduce neutral recycling to core region in the configuration without baffle. In baffle configuration, a good shielding of neutrals in the divertor region can be achieved. The dome configuration reduces plasma density near the null region and flow shear near the separatrix. (author)

  18. Divertor detachment

    Science.gov (United States)

    Krasheninnikov, Sergei

    2015-11-01

    The heat exhaust is one of the main conceptual issues of magnetic fusion reactor. In a standard operational regime the large heat flux onto divertor target reaches unacceptable level in any foreseeable reactor design. However, about two decades ago so-called ``detached divertor'' regimes were found. They are characterized by reduced power and plasma flux on divertor targets and look as a promising solution for heat exhaust in future reactors. In particular, it is envisioned that ITER will operate in a partly detached divertor regime. However, even though divertor detachment was studied extensively for two decades, still there are some issues requiring a new look. Among them is the compatibility of detached divertor regime with a good core confinement. For example, ELMy H-mode exhibits a very good core confinement, but large ELMs can ``burn through'' detached divertor and release large amounts of energy on the targets. In addition, detached divertor regimes can be subject to thermal instabilities resulting in the MARFE formation, which, potentially, can cause disruption of the discharge. Finally, often inner and outer divertors detach at different plasma conditions, which can lead to core confinement degradation. Here we discuss basic physics of divertor detachment including different mechanisms of power and momentum loss (ionization, impurity and hydrogen radiation loss, ion-neutral collisions, recombination, and their synergistic effects) and evaluate the roles of different plasma processes in the reduction of the plasma flux; detachment stability; and an impact of ELMs on detachment. We also evaluate an impact of different magnetic and divertor geometries on detachment onset, stability, in- out- asymmetry, and tolerance to the ELMs. Supported by the U.S. Department of Energy Office of Science, Office of Fusion Energy Sciences under Award Number DE-DE-FG02-04ER54739 at UCSD.

  19. Simulation of DIII-D experiments on detachment and divertor closure

    Science.gov (United States)

    Meier, Eric; Moser, Auna; Leonard, Tony

    2017-10-01

    While divertor detachment is necessary to control the heat flux to divertor targets in ITER and future tokamak fusion devices, detachment is often associated with high pedestal density, which can be problematic for core plasma performance. Divertor closure experiments on DIII-D have shown that the pedestal electron density at detachment is reduced by 35% for a configuration with a high degree of outer divertor closure, compared to an open outer divertor configuration. In this work, SOLPS-ITER modeling, with full drift physics engaged, is used to evaluate the experimental open and closed configurations. Realistic power and particle fluxes are assumed at the core boundary. Predicted 2D ionization profiles will be presented, and sensitivity of detachment behavior to particle and thermal diffusivities, cryopump efficiency, and wall pumping assumptions will be addressed. Initial simulations show a 20% decrease in pedestal density at detachment for the closed configuration, and a similar reduction in the pedestal ionization source. Work supported by US DOE under DE-SC0007880, DE-SC0010434, and DE-FC02-04ER54698.

  20. Fourier-spectral element approximation of the ion–electron Braginskii system with application to tokamak edge plasma in divertor configuration

    Energy Technology Data Exchange (ETDEWEB)

    Minjeaud, Sebastian [Lab. J. A. Dieudonné, UMR CNRS 7351, Université de Nice-Sophia Antipolis, F-06108 Nice (France); INRIA project CASTOR (France); Pasquetti, Richard, E-mail: richard.pasquetti@unice.fr [Lab. J. A. Dieudonné, UMR CNRS 7351, Université de Nice-Sophia Antipolis, F-06108 Nice (France); INRIA project CASTOR (France)

    2016-09-15

    Due to the extreme conditions required to produce energy by nuclear fusion in tokamaks, simulating the plasma behavior is an important but challenging task. We focus on the edge part of the plasma, where fluid approaches are probably the best suited, and our approach relies on the Braginskii ion–electron model. Assuming that the electric field is electrostatic, this yields a set of 10 strongly coupled and non-linear conservation equations that exhibit multiscale and anisotropy features. The computational domain is a torus of complex geometrical section, that corresponds to the divertor configuration, i.e. with an “X-point” in the magnetic surfaces. To capture the complex physics that is involved, high order methods are used: The time-discretization is based on a Strang splitting, that combines implicit and explicit high order Runge–Kutta schemes, and the space discretization makes use of the spectral element method in the poloidal plane together with Fourier expansions in the toroidal direction. The paper thoroughly describes the algorithms that have been developed, provides some numerical validations of the key algorithms and exhibits the results of preliminary numerical experiments. In particular, we point out that the highest frequency of the system is intermediate between the ion and electron cyclotron frequencies.

  1. Effect of 3D magnetic perturbations on divertor conditions and detachment in tokamak and stellarator

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, J-W; Briesemester, A. R.; Kobayashi, M.; Lore, J. D.; Schmitz, O.; Diallo, A.; Gray, T. K.; Lasnier, C. J.; LeBlanc, B. P.; Maingi, R.; McLean, A. G.; Sabbagh, S. A.; Soukhanovskii, V. A.

    2017-06-22

    Enhanced perpendicular heat and momentum transport induces parallel pressure loss leading to divertor detachment, which can be produced by the increase of density in 2D tokamaks. However, in the 3D configurations such as tokamaks with 3D fields and stellarators, the fraction of perpendicular transport can be higher even in a lower density regime, which could lead to the early transition to detachment without passing through the high-recycling regime. 3D fields applied to the limiter tokamak plasmas produce edge stochastic layers close to the last closed flux surface (LCFS), which can allow for enhanced perpendicular transport and indeed the absence of high recycling regime and early detachment have been observed in TEXTOR and Tore Supra. However, in the X-point divertor tokamaks with the applied 3D fields, the parallel transport is still dominant and the detachment facilitation has not been observed yet. Rather, 3D fields affected detachment adversely under certain conditions, either by preventing detachment onset as seen in DIII-D or by re-attaching the existing detached plasma as shown in NSTX. The possible way for strong 3D effects to induce access to the early detachment in divertor tokamaks appears to be via significant perpendicular loss of parallel momentum by frictional force for the counter-streaming flows between neighboring flow channels in the divertor. In principle, the adjacent lobes in the 3D divertor tokamak may generate the counter-streaming flow channels. However, an EMC3-EIRENE simulation for ITER H-mode plasmas demonstrated that screened RMP leads to significantly reduced counter-flows near the divertor target, therefore the momentum loss effect leading to detachment facilitation is expected to be small. This is consistent with the observation in LHD, which showed screening (amplification) of RMP fields in the attachment (stable detachment) case. Work for optimal parameter window for best divertor operation scenario is needed particularly for

  2. Effect of 3D magnetic perturbations on divertor conditions and detachment in tokamak and stellarator

    Science.gov (United States)

    Ahn, J.-W.; Briesemester, A. R.; Kobayashi, M.; Lore, J. D.; Schmitz, O.; Diallo, A.; Gray, T. K.; Lasnier, C. J.; LeBlanc, B. P.; Maingi, R.; McLean, A. G.; Sabbagh, S. A.; Soukhanovskii, V. A.

    2017-08-01

    Enhanced perpendicular heat and momentum transport induces parallel pressure loss leading to divertor detachment, which can be produced by the increase of density in 2D tokamaks. However, in the 3D configurations such as tokamaks with 3D fields and stellarators, the fraction of perpendicular transport can be higher even in a lower density regime, which could lead to the early transition to detachment without passing through the high-recycling regime. 3D fields applied to the limiter tokamak plasmas produce edge stochastic layers close to the last closed flux surface (LCFS), which can allow for enhanced perpendicular transport and indeed the absence of high recycling regime and early detachment have been observed in TEXTOR and Tore Supra. However, in the X-point divertor tokamaks with the applied 3D fields, the parallel transport is still dominant and the detachment facilitation has not been observed yet. Rather, 3D fields affected detachment adversely under certain conditions, either by preventing detachment onset as seen in DIII-D or by re-attaching the existing detached plasma as shown in NSTX. The possible way for strong 3D effects to induce access to the early detachment in divertor tokamaks appears to be via significant perpendicular loss of parallel momentum by frictional force for the counter-streaming flows between neighboring flow channels in the divertor. In principle, the adjacent lobes in the 3D divertor tokamak may generate the counter-streaming flow channels. However, an EMC3-EIRENE simulation for ITER H-mode plasmas demonstrated that screened RMP leads to significantly reduced counter-flows near the divertor target, therefore the momentum loss effect leading to detachment facilitation is expected to be small. This is consistent with the observation in LHD, which showed screening (amplification) of RMP fields in the attachment (stable detachment) case. Work for optimal parameter window for best divertor operation scenario is needed particularly for

  3. Experimental studies of the snowflake divertor in TCV

    Directory of Open Access Journals (Sweden)

    B. Labit

    2017-08-01

    Full Text Available To address the risk that, in a fusion reactor, the conventional single-null divertor (SND configuration may not be able to handle the power exhaust, alternative divertor configurations, such as the Snowflake divertor (SFD, are investigated in TCV. The expected benefits of the SFD-minus in terms of power load and peak heat flux are discussed and compared to experimental measurements. In addition, key results obtained during the last years are summarized.

  4. Hydrogen recycling and transport in the helical divertor of TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Clever, Meike

    2010-07-01

    The aim of this thesis was to investigate the hydrogen recycling at the target plates of the helical divertor in TEXTOR and by this the capability of this divertor configuration to access such favourable operational regimes. In order to study the different divertor density regimes in TEXTOR, discharges were performed in which the total plasma density was increased continuously up to the density limit. The recycling was investigated in a fixed helical divertor structure where four helical strike points with a poloidal width of about 8-10 cm are created at the divertor target plates. The experimental investigation of the hydrogen recycling was carried out using mainly spectroscopic methods supplemented by Langmuir probe, interferometric and atomic beam measurements. In the framework of this thesis a spectroscopic multi camera system has been built that facilitates the simultaneous observation of four different spectral lines, recording images of the divertor target plates and the plasma volume close to the target. The system facilitates the simultaneous measurement of the poloidal and toroidal pattern of the recycling flux at the divertor target without the need for sweeping the plasma structure. The simultaneous observation of different spectral lines reduces the uncertainty in the analysis based on several lines, as the contribution from uncertainties in the reproducibility of plasma parameters in different discharges are eliminated and only the uncertainty of the measurement method limits the accuracy. The spatial resolution of the system in poloidal and toroidal direction (0.8 mm{+-}0.01 mm) is small compared to the separation of the helical strike points, the capability of the measurement method to resolve these structures is therefore limited by the line-of-sight integration and the penetration depth of the light emitting species. The measurements showed that the recycling flux increases linearly with increasing plasma density, a high recycling regime is not

  5. Application of modern optical fiber technology to the study of plasmas of closed divertors and pump limiters in reactor-relevant conditions

    Science.gov (United States)

    Klepper, C. C.; Simpkins, J. E.; Hills, D. L.; Mioduszewski, P. K.; Moyer, R. A.; Gray, D.; Dippel, K. H.; Pospieszczyk, A.

    1990-10-01

    Modern optical fibers, through control of the purity of the materials and the tolerances of the core and clad diameters, provide very good light transmission in the visible and near-ultraviolet regions of the spectrum. This makes it possible to use them in place of traditional optical systems without large losses in light intensity at the detectors. In addition, the same control of the quality of the fiber materials, coupled with novel jacket materials, makes it possible to use the fibers inside vacuum chambers and at elevated temperatures. A fiber-optic bundle recently installed in the TEXTOR tokamak is an example of the use of modern fiber technology. The bundle was made of 80 100-μm fibers held together with a polyimide organic material that has good outgassing specifications up to 400 °C. This fiber bundle has been used for recent measurements of the recycling in the throat region of one of the blades of the Advanced Limiter Test-II (ALT-II) belt pump limiter. Another system presently under design and testing employs individual fibers that are gold plated. These fibers are fed through holes in a vacuum blank flange and silver soldered to the flange. This system is designed to transmit the light from the strike point inside the closed divertor of the DIII-D tokamak out to a spectrometer. There, the spectral profile of the Hα line is analyzed to determine the energy distribution of the recycling particles.

  6. DTT: a divertor tokamak test facility for the study of the power exhaust issues in view of DEMO

    Science.gov (United States)

    Albanese, R.; WPDTT2 Team; DTT Project Proposal Contributors, the

    2017-01-01

    In parallel with the programme to optimize the operation with a conventional divertor based on detached conditions to be tested on the ITER device, a project has been launched to investigate alternative power exhaust solutions for DEMO, aimed at the definition and the design of a divertor tokamak test facility (DTT). The DTT project proposal refers to a set of parameters selected so as to have edge conditions as close as possible to DEMO, while remaining compatible with DEMO bulk plasma performance in terms of dimensionless parameters and given constraints. The paper illustrates the DTT project proposal, referring to a 6 MA plasma with a major radius of 2.15 m, an aspect ratio of about 3, an elongation of 1.6-1.8, and a toroidal field of 6 T. This selection will guarantee sufficient flexibility to test a wide set of divertor concepts and techniques to cope with large heat loads, including conventional tungsten divertors; liquid metal divertors; both conventional and advanced magnetic configurations (including single null, snow flake, quasi snow flake, X divertor, double null); internal coils for strike point sweeping and control of the width of the scrape-off layer in the divertor region; and radiation control. The Poloidal Field system is planned to provide a total flux swing of more than 35 Vs, compatible with a pulse length of more than 100 s. This is compatible with the mission of studying the power exhaust problem and is obtained using superconducting coils. Particular attention is dedicated to diagnostics and control issues, especially those relevant for plasma control in the divertor region, designed to be as compatible as possible with a DEMO-like environment. The construction is expected to last about seven years, and the selection of an Italian site would be compatible with a budget of 500 M€.

  7. Modeling detachment physics in the NSTX snowflake divertor

    Energy Technology Data Exchange (ETDEWEB)

    Meier, E.T., E-mail: emeier@wm.edu [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Soukhanovskii, V.A. [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Bell, R.E.; Diallo, A.; Kaita, R.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, Princeton, NJ 08540 (United States); McLean, A.G. [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Podestà, M. [Princeton Plasma Physics Laboratory, Princeton, NJ 08540 (United States); Rognlien, T.D.; Scotti, F. [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States)

    2015-08-15

    The snowflake divertor is a proposed technique for coping with the tokamak power exhaust problem in next-step experiments and eventually reactors, where extreme power fluxes to material surfaces represent a leading technological and physics challenge. In lithium-conditioned National Spherical Torus Experiment (NSTX) discharges, application of the snowflake divertor typically induced partial outer divertor detachment and severalfold heat flux reduction. UEDGE is used to analyze and compare conventional and snowflake divertor configurations in NSTX. Matching experimental upstream profiles and divertor measurements in the snowflake requires target recycling of 0.97 vs. 0.91 in the conventional case, implying partial saturation of the lithium-based pumping mechanism. Density scans are performed to analyze the mechanisms that facilitate detachment in the snowflake, revealing that increased divertor volume provides most of the parallel heat flux reduction. Also, neutral gas power loss is magnified by the increased wetted area in the snowflake, and plays a key role in generating volumetric recombination.

  8. Control of divertor geometry and performance of the ergodic divertor of Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Ghendrih, Ph. E-mail: ghendrih@drfc.cad.cea.fr; Becoulet, M.; Costanzo, L.; Corre, Y.; Grisolia, C.; Grosman, A.; Guirlet, R.; Gunn, J.; Loarer, T.; Monier-Garbet, P.; Mank, G.; Reichle, R.; Vallet, J.-C.; Zabiego, M.; Azeroual, A.; Bucalossi, J.; Devynck, P.; De Michelis, C; Finken, K.H.; Hogan, J.; Laugier, F.; Nguyen, F.; Pegourie, B.; Saint-Laurent, F.; Schunke, B

    2001-03-01

    Experimental evidence of the location of the ergodic divertor separatrix is shown to agree with the predicted value given by codes. Variation of this position modifies the divertor tightness, defined as the ratio of the divertor to core density. This effect is governed by laminar transport, i.e., transport proportional to the magnitude of the perturbation. Operation with feedback control of the divertor temperature allows one to optimise the choice of injected impurity species. At 10 eV divertor temperature, nitrogen is shown to lead to the largest decrease in energy flux to the divertor at lowest contribution to Z{sub eff}. Parallel energy fluxes as low as 2 MW m{sup -2} are thus achieved on the target plates. For this impurity, radiation is localised in the divertor volume thus leading to radiation compression close to 10. The ergodic divertor appears as a powerful tool to control plasma-wall interaction with no loss of core confinement or plasma current.

  9. X-point target divertor concept and the Alcator DX high power divertor test facility

    Science.gov (United States)

    Labombard, B.; Marmar, E.; Irby, J.; Vieria, R.; Wolfe, S.; Bonoli, P.; Fiore, C.; Granetz, R.; Greenwald, M.; Hutchinson, I.; Hubbard, A.; Hughes, J.; Lin, Y.; Lipschultz, B.; Parker, R.; Porkolab, M.; Reinke, M.; Rice, J.; Shiraiwa, S.; Terry, J.; Theiler, C.; Wallace, G.; White, A.; Whyte, D.; Wukitch, S.

    2013-10-01

    Three critical challenges must be met before a steady-state, power-producing fusion reactor can be realized: (1) safely handle extreme plasma exhaust power, (2) completely suppress material erosion at divertor targets and (3) do this while maintaining a burning plasma core. Advanced divertors such as `Super X' and `X-point target' have the potential to solve all three challenges by producing a stable, fully detached, low temperature plasma in the divertor while maintaining a hot boundary layer around a clean plasma core. The X-point target divertor may be particularly effective. It places a second X-point in the pathway of the peak parallel heat flux with the intention of forming an X-point MARFE in the divertor volume, well away from the primary X-point that defines the last closed flux surface and at larger major radius, providing detachment front stability. Divertor heat dissipation is via volumetric processes (radiation, ion-neutral collisions), virtually eliminating erosion by ion bombardment and reducing peak heat flux and neutron fluence on remote divertor target components. Alcator DX is conceived as a national facility to test these ideas. It employs the high magnetic field technology of Alcator combined with high-power ICRH to investigate advanced divertors at reactor-level parallel heat flux densities.

  10. High flux expansion divertor studies in NSTX

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V A; Maingi, R; Bell, R E; Gates, D A; Kaita, R; Kugel, H W; LeBlanc, B P; Maqueda, R; Menard, J E; Mueller, D; Paul, S F; Raman, R; Roquemore, A L

    2009-06-29

    Projections for high-performance H-mode scenarios in spherical torus (ST)-based devices assume low electron collisionality for increased efficiency of the neutral beam current drive. At lower collisionality (lower density), the mitigation techniques based on induced divertor volumetric power and momentum losses may not be capable of reducing heat and material erosion to acceptable levels in a compact ST divertor. Divertor geometry can also be used to reduce high peak heat and particle fluxes by flaring a scrape-off layer (SOL) flux tube at the divertor plate, and by optimizing the angle at which the flux tube intersects the divertor plate, or reduce heat flow to the divertor by increasing the length of the flux tube. The recently proposed advanced divertor concepts [1, 2] take advantage of these geometry effects. In a high triangularity ST plasma configuration, the magnetic flux expansion at the divertor strike point (SP) is inherently high, leading to a reduction of heat and particle fluxes and a facilitated access to the outer SP detachment, as has been demonstrated recently in NSTX [3]. The natural synergy of the highly-shaped high-performance ST plasmas with beneficial divertor properties motivated a further systematic study of the high flux expansion divertor. The National Spherical Torus Experiment (NSTX) is a mid-sized device with the aspect ratio A = 1.3-1.5 [4]. In NSTX, the graphite tile divertor has an open horizontal plate geometry. The divertor magnetic configuration geometry was systematically changed in an experiment by either (1) changing the distance between the lower divertor X-point and the divertor plate (X-point height h{sub X}), or by (2) keeping the X-point height constant and increasing the outer SP radius. An initial analysis of the former experiment is presented below. Since in the divertor the poloidal field B{sub {theta}} strength is proportional to h{sub X}, the X-point height variation changed the divertor plasma wetted area due to

  11. Divertor Heat Flux Mitigation in the National Spherical Torus Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V A; Maingi, R; Gates, D A; Menard, J E; Paul, S F; Raman, R; Roquemore, A L; Bell, M G; Bell, R E; Boedo, J A; Bush, C E; Kaita, R; Kugel, H W; LeBlanc, B P; Mueller, D

    2008-08-04

    Steady-state handling of divertor heat flux is a critical issue for both ITER and spherical torus-based devices with compact high power density divertors. Significant reduction of heat flux to the divertor plate has been achieved simultaneously with favorable core and pedestal confinement and stability properties in a highly-shaped lower single null configuration in the National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40, 557 2000] using high magnetic flux expansion at the divertor strike point and the radiative divertor technique. A partial detachment of the outer strike point was achieved with divertor deuterium injection leading to peak flux reduction from 4-6 MW m{sup -2} to 0.5-2 MW m{sup -2} in small-ELM 0.8-1.0 MA, 4-6 MW neutral beam injection-heated H-mode discharges. A self-consistent picture of outer strike point partial detachment was evident from divertor heat flux profiles and recombination, particle flux and neutral pressure measurements. Analytic scrape-off layer parallel transport models were used for interpretation of NSTX detachment experiments. The modeling showed that the observed peak heat flux reduction and detachment are possible with high radiated power and momentum loss fractions, achievable with divertor gas injection, and nearly impossible to achieve with main electron density, divertor neutral density or recombination increases alone.

  12. Controlling marginally detached divertor plasmas

    Science.gov (United States)

    Eldon, D.; Kolemen, E.; Barton, J. L.; Briesemeister, A. R.; Humphreys, D. A.; Leonard, A. W.; Maingi, R.; Makowski, M. A.; McLean, A. G.; Moser, A. L.; Stangeby, P. C.

    2017-06-01

    A new control system at DIII-D has stabilized the inter-ELM detached divertor plasma state for H-mode in close proximity to the threshold for reattachment, thus demonstrating the ability to maintain detachment with minimal gas puffing. When the same control system was instead ordered to hold the plasma at the threshold (here defined as T e  =  5 eV near the divertor target plate), the resulting T e profiles separated into two groups with one group consistent with marginal detachment, and the other with marginal attachment. The plasma dithers between the attached and detached states when the control system attempts to hold at the threshold. The control system is upgraded from the one described in Kolemen et al (2015 J. Nucl. Mater. 463 1186) and it handles ELMing plasmas by using real time D α measurements to remove during-ELM slices from real time T e measurements derived from divertor Thomson scattering. The difference between measured and requested inter-ELM T e is passed to a PID (proportional-integral-derivative) controller to determine gas puff commands. While some degree of detachment is essential for the health of ITER’s divertor, more deeply detached plasmas have greater radiative losses and, at the extreme, confinement degradation, making it desirable to limit detachment to the minimum level needed to protect the target plate (Kolemen et al 2015 J. Nucl. Mater. 463 1186). However, the observed bifurcation in plasma conditions at the outer strike point with the ion B   ×  \

  13. Radiative divertor optimization for NSTX Upgrade based on open geometry standard divertor experiments in NSTX

    Science.gov (United States)

    Soukhanovskii, V. A.; Izacard, O.; Scotti, F.; Maingi, R.; Bell, R. E.; Kaita, R.; Kaye, S. M.; Leblanc, B. P.; Menard, J. E.; Mueller, D.

    2017-10-01

    Recent analyses of NSTX divertor experiments suggest a way to optimize the standard open geometry divertor configuration for partial detachment with deuterium puffing and intrinsic carbon radiation. Results from the NSTX experiments and the divertor transport and radiation model obtained with the multi-fluid code UEDGE are used to show that detachment onset and properties are sensitive to 1) placing the neutral gas source in the vicinity of the strike point, 2) directing the recycling neutrals toward the separatrix by adjusting the poloidal separatrix angle, and 3) entrapping neutrals by plasma plugging via the high poloidal magnetic flux expansion configuration. These findings will be tested in NSTX Upgrade, where H-mode scenarios with 2 MA, 1 T, 10 MW NBI-heated discharges and 5 s flattop are predicted to produce unmitigated peak divertor heat fluxes above 10 MW/m2, necessitating the scrape-off layer power sharing between upper and lower divertors and inducing dissipative losses. Supported by the US DOE under Contracts DE-AC52-07NA27344 and DE-AC02-09CH11466.

  14. Progress in snowflake divertor research in DIII-D, NSTX and NSTX-U

    Science.gov (United States)

    Soukhanovskii, V. A.; Allen, S.; Fenstermacher, M.; Izacard, O.; Lasnier, C.; Makowski, M.; McLean, A.; Myer, W.; Ryutov, D.; Scotti, F.; Eldon, D.; Kolemen, E.; Vail, P.; Canal, G.; Groebner, R.; Hyatt, A.; Leonard, A.; Osborne, T.; Bell, R.; Diallo, A.; Gerhardt, S.; Kaye, S.; Leblanc, B.; Menard, J.; Podesta, M.

    2016-10-01

    Recent snowflake (SF) divertor DIII-D experiments focused on divertor heat transport under attached and radiative divertor conditions, incl 1-understanding of increased scrape-off layer width in SF-plus configuration at lower densities; 2-particle, heat and radiation distribution in the SF divertor with CD4 seeding. NSTX data was analyzed to understand the link between SF divertor and ELM (de)stabilization with and without CD4 seeding and lithium conditioning. Prep for SF divertor experiments in NSTX-U include 1-equilibria modeling with ISOLVER code using various sets of divertor coils and L- and H-mode plasma scenarios; 2-transport and impurity radiation modeling with UEDGE code; 3-new diagnostics (ie-a 100-200 kHz camera for null-region mode observations). Supported by DOE under DE-AC52-07NA27344, DE-AC02-09CH11466, DE-FC02-04ER54698.

  15. A study on the fusion reactor - Numerical analyses of MHD equilibrium and= edge plasma transport in tokamak fusion reactor with divertor configurations

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sang Hee; Kang, Kyung Doo; Ryu, Ji Myung; Kim, Deok Kyu; Chung, TaeKyun; Chung, Mo Se [Seoul National University, Seoul (Korea, Republic of); Cho, Su Won [Kyungki University, Suwon (Korea, Republic of)

    1995-08-01

    In the present project for developing the numerical codes of 2-D MHD equilibrium, edge plasma transport and neutral particle transport for the tokamak plasmas, we computed the MHD equilibria of single and double null configurations and determined the external coil currents and the plasma parameters used for operation and control data. Also we numerically acquired the distributions of edge plasma parameters in poloidal and radial directions= and the design-related values according to the various operating conditions using the developed plasma transport code. Furthermore, a neutral particle transport code for the edge region is developed and them used for the analysis of the neutral particle behavior yielding the source terms in the fluid transport equations, and expected to supply the input parameters for the edge plasma transport code. 53 refs., 12 tabs., 44 figs. (author)

  16. Investigation of parameter space for fully detached long-legged divertor operation

    Science.gov (United States)

    Umansky, M. V.; Labombard, B.; Rensink, M. E.; Rognlien, T. D.

    2017-10-01

    Recently it was found in numerical modeling that passively-stable fully detached divertor regimes exist in a broad range of input power from the core, for divertor configurations with radially or vertically extended, tightly baffled, outer divertor legs, with or without a secondary X-point in the leg volume. This report presents a comparative computational study of detached divertor operation carried out for a variety of divertor configurations, expanding on the initial work reported in Ref.. The parameters are based on those of the ADX tokamak design, and the simulations are carried out with the tokamak edge transport code UEDGE. The simulations show that long-legged divertors have a large increase of the peak power handling capability, by up to an order of magnitude, compared to conventional divertors. For the detached divertor regime in these simulations, important physics combines interplay of strong convective plasma transport to the outer wall, confinement of neutral gas in the divertor volume, geometric effects including secondary X-point, and atomic radiation. As the power from the core is varied, the detachment front merely shifts up or down in the leg but remains stable. The present work addresses sensitivity of the detached divertor regime to various parameters used in the model, including the anomalous plasma transport, neutral transport, impurity radiation, and geometry of plasma-facing material surfaces. Work performed for U.S. DOE by LLNL under contract DE-AC52-07NA27344.

  17. Estimation of peak heat flux onto the targets for CFETR with extended divertor leg

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Chuanjia; Chen, Bin [School of Nuclear Science and Technology, University of Science and Technology of China, 96 Jinzhai Road, Hefei, Anhui 230026 (China); Xing, Zhe [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Wu, Haosheng [School of Nuclear Science and Technology, University of Science and Technology of China, 96 Jinzhai Road, Hefei, Anhui 230026 (China); Mao, Shifeng, E-mail: sfmao@ustc.edu.cn [School of Nuclear Science and Technology, University of Science and Technology of China, 96 Jinzhai Road, Hefei, Anhui 230026 (China); Luo, Zhengping; Peng, Xuebing [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Ye, Minyou [School of Nuclear Science and Technology, University of Science and Technology of China, 96 Jinzhai Road, Hefei, Anhui 230026 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)

    2016-11-01

    Highlights: • A hypothetical geometry is assumed to extend the outer divertor leg in CFETR. • Density scan SOLPS simulation is done to study the peak heat flux onto target. • Attached–detached regime transition in out divertor occurs at lower puffing rate. • Unexpected delay of attached–detached regime transition occurs in inner divertor. - Abstract: China Fusion Engineering Test Reactor (CFETR) is now in conceptual design phase. CFETR is proposed as a good complement to ITER for demonstrating of fusion energy. Divertor is a crucial component which faces the plasmas and handles huge heat power for CFETR and future fusion reactor. To explore an effective way for heat exhaust, various methods to reduce the heat flux to divertor target should be considered for CFETR. In this work, the effect of extended out divertor leg on the peak heat flux is studied. The magnetic configuration of the long leg divertor is obtained by EFIT and Tokamak Simulation Code (TSC), while a hypothetical geometry is assumed to extend the out divertor leg as long as possible inside vacuum vessel. A SOLPS simulation is performed to study peak heat flux of the long leg divertor for CFETR. D{sub 2} gas puffing is used and increasing of the puffing rate means increase of plasma density. Both peak heat flux onto inner and outer targets are below 10 MW/m{sup 2} is achieved. A comparison between the peak heat flux between long leg and conventional divertor shows that an attached–detached regime transition of out divertor occurs at lower gas puffing gas puffing rate for long leg divertor. While for the inner divertor, even the configuration is almost the same, the situation is opposite.

  18. Numerical analysis of particle recycling in the TEXTOR helical divertor

    Science.gov (United States)

    Frerichs, H.; Clever, M.; Feng, Y.; Lehnen, M.; Reiter, D.; Schmitz, O.

    2012-02-01

    The TEXTOR helical divertor is a magnetic configuration created by the application of external resonant magnetic perturbations with the intention to control plasma edge transport and the resulting particle and heat fluxes to the divertor target. It is confirmed by 3D computer simulations that no high-recycling-like regime is established under TEXTOR relevant conditions, despite the fact that a transition to detachment (i.e. a saturation or even a roll-over of the recycling flux) is observed at high densities. The driving mechanisms are, distinct from apparently similar observations in poloidal divertors and stellarator divertors, a combination of volumetric power losses and enhanced upstream-to-downstream heat transport, but with no significant role of the momentum balance.

  19. Influence of helium puff on divertor asymmetry in experimental advanced superconducting tokamak

    DEFF Research Database (Denmark)

    Liu, S. C.; Guo, H. Y.; Xu, G. S.

    2014-01-01

    Divertor asymmetries with helium puffing are investigated in various divertor configurations on Experimental Advanced Superconducting Tokamak (EAST). The outer divertor electron temperature decreases significantly during the gas injection at the outer midplane. As soon as the gas is injected...... parameters are measured by reciprocating probes at the outer midplane, showing that the electron temperature and density increase but the parallel Mach number decreases significantly due to the gas injection. Effects of poloidal E × B drifts and parallel SOL flows on the divertor asymmetry observed in EAST...

  20. L-H power threshold studies with tungsten/carbon divertor on the EAST tokamak

    DEFF Research Database (Denmark)

    Chen, L.; Xu, G. S.; Gao, W.

    2016-01-01

    The power threshold for low (L) to high (H) confinement mode transition achieved by radio-frequency heating and molybdenum first wall with lithium coating has been experimentally investigated on the EAST tokamak for two sets of divertor geometries and materials: tungsten/carbon divertor and full...... carbon divertor. For both sets of divertors, the power threshold was found to decrease with gradual accumulation of the lithium wall coating, suggesting the important role played by the low Z impurities and/or the edge neutral density on the L-H power threshold. When operating in the upper single null...... configuration, with the ion grad-B drift direction away from the primary X-point, a lower normalized power threshold is observed in EAST with the tungsten/carbon divertor, compared to the carbon divertor after intensive lithium wall coating. A newly installed cryopump increasing the pumping efficiency also...

  1. Comparative studies of inner and outer divertor discharges and a fueling study in QUEST

    Energy Technology Data Exchange (ETDEWEB)

    Mitarai, O., E-mail: omitarai@ktmail.tokai-u.jp [Kumamoto Liberal Arts Education Center, Tokai University, 9-1-1 Toroku, Higashi-ku, Kumamoto 862-8652 (Japan); Nakamura, K.; Hasegawa, M.; Onchi, T.; Idei, H.; Fujisawa, A.; Hanada, K.; Zushi, H.; Higashijima, A.; Nakashima, H.; Kawasaki, S. [Research Institute for Applied Mechanics, Kyushu University, 6-1 Kasugakoen, Kasuga 816-8580 Japan (Japan); Matsuoka, K. [National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292 (Japan); Koike, S.; Takahashi, T. [Division of Electronics and Informatics, Faculty of Science and Technology, Gunma University, 1-5-1 Tenjin-cho, Kiryu, Gunma 376-8515 (Japan); Tsutsui, H. [Research Laboratory for Nuclear Reactors, Tokyo Inst. Tech, 2-12-1 Ookayama, Tokyo 152-8550 (Japan)

    2016-11-01

    Highlights: • Central solenoid has a small flux in QUEST. • Large plasma current is obtained when the position is shifted to the inboard side. • Two types of divertor operation are compared. • Novel merging fueling methods are proposed. • Coaxial helicity injection (CHI) fueling was examined in QUEST divertor configuration. - Abstract: As QUEST has a small central solenoid (CS), a larger Ohmic discharge current has been obtained when the plasma shifts to the inboard side. This tendency restricts a divertor operation to the smaller plasma current regime. As the inner divertor coil has a smaller mutual inductance, it would be expected that its utilization seems to be better for easier plasma current ramp-up for a divertor operation. In this work, we made comparative studies on the plasma current ramp-up for two divertor coils. It is found that while the inner divertor coil with smaller mutual inductance needs a larger coil current, the outer divertor coil with larger mutual inductance needs a smaller coil current for divertor operation. Thus we have found that the plasma current ramp-up characteristics are almost similar for both configurations. We also propose a new fueling method for spherical tokamak (ST) using the coaxial helicity injection (CHI). The main plasma current would be generated at first, and then the CHI plasma current is created between bottom two electrode plates and merged into the main plasma current for fueling.

  2. Comparison of 2D simulations of detached divertor plasmas with divertor Thomson measurements in the DIII-D tokamak

    Directory of Open Access Journals (Sweden)

    T.D. Rognlien

    2017-08-01

    Full Text Available A modeling study is reported using new 2D data from DIII-D tokamak divertor plasmas and improved 2D transport model that includes large cross-field drifts for the numerically difficult low anomalous transport regime associated with the H-mode. The data set, which spans a range of plasma densities for both forward and reverse toroidal magnetic field (Bt, is provided by divertor Thomson scattering (DTS. Measurements utilizing X-point sweeping give corresponding 2D profiles of electron temperature (Te and density (ne across both divertor legs for individual discharges. The simulations focus on the open magnetic field-line regions, though they also include a small region of closed field lines. The calculations show the same features of in/out divertor plasma asymmetries as measured in the experiment, with the normal Bt direction (ion ∇B drift toward the X-point having higher ne and lower Te in the inner divertor leg than outer. Corresponding emission data for total radiated power shows a strong inner-divertor/outer-divertor asymmetry that is reproduced by the simulations. These 2D UEDGE transport simulations are enabled for steep-gradient H-mode conditions by newly implemented algorithms to control isolated grid-scale irregularities.

  3. Divertor IR thermography on Alcator C-Moda)

    Science.gov (United States)

    Terry, J. L.; LaBombard, B.; Brunner, D.; Payne, J.; Wurden, G. A.

    2010-10-01

    Alcator C-Mod is a particularly challenging environment for thermography. It presents issues that will similarly face ITER, including low-emissivity metal targets, low-Z surface films, and closed divertor geometry. In order to make measurements of the incident divertor heat flux using IR thermography, the C-Mod divertor has been modified and instrumented. A 6° toroidal sector has been given a 2° toroidal ramp in order to eliminate magnetic field-line shadowing by imperfectly aligned divertor tiles. This sector is viewed from above by a toroidally displaced IR camera and is instrumented with thermocouples and calorimeters. The camera provides time histories of surface temperatures that are used to compute incident heat-flux profiles. The camera sensitivity is calibrated in situ using the embedded thermocouples, thus correcting for changes and nonuniformities in surface emissivity due to surface coatings.

  4. Taming the plasma-material interface with the snowflake divertor.

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V A

    2015-04-24

    Experiments in several tokamaks have provided increasing support for the snowflake configuration as a viable tokamak heat exhaust concept. This white paper summarizes the snowflake properties predicted theoretically and studied experimentally, and identifies outstanding issues to be resolved in existing and future facilities before the snowflake divertor can qualify for the reactor interface.

  5. Proposal of an alternative upper divertor in ASDEX Upgrade supported by EMC3-EIRENE simulations

    Directory of Open Access Journals (Sweden)

    T. Lunt

    2017-08-01

    Full Text Available We discuss the benefits of installing a pair of in-vessel coils with currents |Ifx| ≲ 50 kAt in the upper divertor of ASDEX Upgrade (AUG to study a series of ‘alternative’ divertor configurations, like the Snowflake (SF and the X-divertor (XD, that are currently considered as alternative solutions for the power exhaust problem. The possibility of operating the standard lower single-null (SN and double-null (DN would be preserved. Potential effects to reduce the peak parallel- and/or perpendicular heat flux are predicted from a simple geometrical-diffusive model as well as by numerical EMC3-EIRENE simulations for pure deuterium attached conditions with spatially constant diffusion coefficients. Beyond that a series of other potential transport- and radiation related heat flux mitigation effects are identified and could be studied experimentally with the modified upper divertor in the high-power divertor Tokamak AUG.

  6. Divertor and midplane materials evaluation system in DIII-D

    Science.gov (United States)

    Wong, C. P. C.; Rudakov, D. L.; Allain, J. P.; Bastasz, R. J.; Brooks, N. H.; Brooks, J. N.; Doerner, R. P.; Evans, T. E.; Hassanein, A.; Jacob, W.; Krieger, K.; Litnovsky, A.; McLean, A. G.; Philipps, V.; Pigarov, A. Yu.; Wampler, W. R.; Watkins, J. G.; West, W. P.; Whaley, J.; Wienhold, P.

    2007-06-01

    The Divertor Materials Evaluation System (DiMES) at General Atomics has successfully advanced the understanding of plasma surface interaction phenomena involving ITER-relevant materials and has been utilized for advanced diagnostic designs in the lower divertor of DIII-D. This paper describes a series of recent successful experiments. These include the study of carbon deposition in gaps and metallic mirrors as a function of temperature, study of dust migration from the divertor, study of methane injection in order to benchmark chemical sputtering diagnostics, and the measurement of charge exchange neutrals with a hydrogen sensor. In concert with the modification of the lower divertor of DIII-D, the DiMES sample vertical location was modified to match the raised divertor floor. The new Mid-plane Material Exposure Sample (MiMES) design will also be presented. MiMES will allow the study and measurement of erosion and redeposition of material at the outboard mid-plane of DIII-D, including effects from convective transport. We will continue to expose relevant materials and advanced diagnostics to different plasma configurations under various operational regimes, including material erosion and redeposition experiments, and gaps and mirror exposures at elevated temperature.

  7. The Effect of Magnetic Balance and Particle Drifts on Radiating Divertor Behavior in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Petrie, T; Porter, G; Brooks, N; Fenstermacher, M; Ferron, J; Groth, M; Hyatt, A; La Haye, R; Lasnier, C; Leonard, A; Politzer, P; Rensink, M; Schaffer, M; Wade, M; Watkins, J; West, W

    2008-10-14

    Success of the puff-and-pump radiating divertor approach depends sensitively on both the divertor magnetic geometry and the ion B x {del}B drift direction. In the puff-and-pump scenario used in this study, argon impurities were injected into the private flux region, while plasma flows into both the inner and outer divertors were enhanced by a combination of particle pumping near both divertor targets and deuterium gas puffing upstream of the divertor targets. For single-null (SN) configurations, argon accumulation was 2-3 times lower in the main plasma when the ion B x {del}B drift was directed away from the divertor. The puff-and-pump approach was much less effective in screening argon from the main plasma of double-null (DN) discharges than of SN discharges, such that argon impurities accumulated in the main plasma of DNs at a rate {approx}2-3 times higher than in corresponding SNs. Regardless of which divertor in DN had argon injection, argon accumulated in the divertor that was opposite the B x {del}B drift direction. The argon density in the main plasma during puff-and-pump operation fell by a factor of three for dRsep {ge} +0.4 cm when the ion B x {del}B drift was directed away from the dominant divertor, and this represents the transition from DN to SN behavior during puff-and-pump application. Comparison of identically-prepared SN H-mode plasmas showed that core density control of deuterium and the argon was far more sensitive to the ion B x {del}B drift direction than to divertor closure in DIII-D.

  8. An innovative small angle slot divertor concept for long pulse advanced tokamaks

    Science.gov (United States)

    Guo, Houyang

    2017-10-01

    A new Small Angle Slot (SAS) divertor is being developed in DIII-D to address the challenge of efficient divertor heat dispersal at the relatively low plasma density required for non-inductive current drive in future advanced tokamaks. SAS features a small incident angle near the plasma strike point on the divertor target plate with a progressively opening slot. SOLPS (B2-Eirene) edge code analysis finds that SAS can achieve strong plasma cooling when the strike point is placed near the small angle target plate in the slot, leading to low electron temperature Te across the entire divertor target. This is enabled by strong coupling between a gas tight slot and directed neutral recycling by the small angle target to enhance neutral buildup near the target. SOLPS analysis reveals a strong correlation between Te and D2 density at the target for various divertor configurations including the flat target, slanted target, and lower single null divertor. The strong correlation suggests that achievement of low Te may reduce essentially to identifying the divertor baffle geometry that achieves the highest target gas density at a given upstream condition. The SAS divertor concept has recently been tested in DIII-D for a range of plasma configurations and conditions with precise control of slot strike point location. In confirmation of SOLPS predictions, a sharp transition is observed when the strike point is moved to the critical outer corner of SAS. A set of Langmuir probes imbedded in SAS show that the Te radial profile, which is peaked at the strike point when it is located away from the SAS corner, becomes low across the target when the strike point is located near the corner. With further increase in density, deep-slot detachment occurs with Te 1 eV, measured by the unique DIII-D divertor Thomson Scattering diagnostic. Work supported by US DOE under DE-FC02-04ER54698.

  9. Closed form analytical inverse solutions for Risley-prism-based beam steering systems in different configurations

    Science.gov (United States)

    Li, Yajun

    2011-08-01

    Nonparaxial ray tracing through Risley prisms of four different configurations is performed to give the exact solution of the inverse problem arisen from applications of Risley prisms to free space communications. Predictions of the exact solution and the third-order theory [Appl. Opt.50, 679 (2011)APOPAI0003-693510.1364/AO.50.000679] are compared and results are shown by curves for systems using prisms of different materials. The exact solution for the problem of precision pointing is generalized to investigate the synthesis of the scan pattern, i.e., to create a desirable scan pattern on some plane perpendicular to the optical axis of the system by controlling the circular motion of the two prisms.

  10. A New Scaling for Divertor Detachment

    Science.gov (United States)

    Goldston, Robert

    2017-10-01

    The ITER design and future fusion power plant designs depend on divertor detachment, whether partial, pronounced or complete, both to limit heat flux to plasma-facing components and to limit surface erosion due to sputtering. Generally the parallel heat flux, estimated as proportional to Psep / R or Psep B / R , is used as a proxy for the difficulty of achieving detachment. Here we argue that the impurity cooling required for detachment is strongly dependent on the upstream separatrix density, which is limited by Greenwald scaling. Taking this into account self-consistently, along with the Heuristic Drift (HD) model for the SOL width, and using a Lengyel radiation model that includes non-coronal effects, we find that the relative impurity concentration, cz ≡nz /ne , required for detachment scales dominantly as cz Psep /Bp(nsep /nGW) 2 . The absence of any explicit favorable size scaling is concerning, as Psep must increase by an order of magnitude from present experiments to an economic fusion power system, while increases in the poloidal magnetic field strength are limited by magnet technology and MHD stability. This result should not be surprising, as it follows from the simplest scaling, Psep czne2VSOL , taking into account the Greenwald density limit and the HD SOL volume scaling. Reinke has combined a similar approach with the requirement to maintain H-mode, which sets a lower limit on Psep, and also arrives at an incentive for high field and disincentive for large size. These results should be challenged by comparison with 2D divertor codes and with measurements on existing experiments. In particular measurements are required for extrinsic divertor impurity concentration over a range of power and density conditions far from the regime where detachment can be achieved with deuterium puffing and intrinsic impurities alone. Nonetheless, these results suggest that higher magnetic field, stronger shaping, double-null operation, `advanced' divertor magnetic and

  11. Detached divertor plasmas in JET

    Energy Technology Data Exchange (ETDEWEB)

    Horton, L.D.; Borrass, K.; Corrigan, G.; Gottardi, N.; Lingertat, J.; Loarte, A.; Simonini, R.; Stamp, M.F.; Taroni, A. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking; Stangeby, P.C. [Toronto Univ., ON (Canada). Inst. for Aerospace Studies

    1994-07-01

    In simulations with high radiated power fractions, it is possible to produce the drop in ion current to the divertor targets typical of detached plasmas. Despite the fact that these experiments are performed on beryllium target tiles, radiation from deuterium and beryllium cannot account for the measured power losses. The neutral deuterium levels in the SOL in these plasmas are higher than the model predicts. This may be due to leakage from the divertor or to additional wall sources related to the non-steady nature of these plasmas. In contrast, a surprisingly high level of carbon is present in these discharges; higher even than would be predicted are the divertor target tiles pure carbon. This level may well be large enough to produce the measured radiation. (authors). 6 refs., 2 figs., 1 tab.

  12. Reconstruction of equilibrium magnetic configurations in the Globus-M spherical tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Sakharov, N. V., E-mail: nikolay.sakharov@mail.ioffe.ru; Voronin, A. V.; Gusev, V. K. [Russian Academy of Sciences, Ioffe Physical Technical Institute (Russian Federation); Kavin, A. A.; Kamenshchikov, S. N.; Lobanov, K. M. [Efremov Research Institute of Electrophysical Apparatus (Russian Federation); Minaev, V. B.; Novokhatsky, A. N.; Patrov, M. I., E-mail: michael.patrov@mail.ioffe.ru; Petrov, Yu. V.; Shchegolev, P. B. [Russian Academy of Sciences, Ioffe Physical Technical Institute (Russian Federation)

    2015-12-15

    The results of reconstruction of equilibrium magnetic configurations in the Globus-M spherical tokamak by means of the EFIT code and by the method of movable filaments with the use of the data from magnetic measurements are compared. The EFIT code allows one to completely reconstruct the magnetic configuration by solving the Grad−Shafranov equation. In the method of movable filaments, the distribution of the toroidal current flowing through the plasma is described by a set of infinitely thin current-carrying rings. In this method, the last closed magnetic surface (LCMS) and the open surfaces lying beyond the LCMS are calculated. Using both methods, the coordinates of the regions where the separatrix strikes the divertor plates were determined. The results obtained agree well with the distributions of the temperature over the tungsten divertor tiles measured using an IR camera.

  13. Intensity, frequency and spatial configuration of winter temperature inversions in the closed La Brevine valley, Switzerland

    Science.gov (United States)

    Vitasse, Yann; Klein, Geoffrey; Kirchner, James W.; Rebetez, Martine

    2017-11-01

    Some of the world's valleys are famous for having particularly cold microclimates. The La Brevine valley, in the Swiss Jura Mountains, holds the record for the lowest temperature ever measured in an inhabited location in Switzerland. We studied cold air pools (CAPs) in this valley during the winter of 2014-2015 using 44 temperature data loggers distributed between 1033 and 1293 m asl. Our goals were to (i) describe the climatic conditions under which CAPs form in the valley, (ii) examine the spatial configuration and the temperature structure of the CAPs and (iii) quantify how often temperature inversions occur in winter using long-term series of temperature from the valley floor. Our results show that CAPs occurred every second night, on average, during the winter of 2014-2015 and were typically formed under cloudless, windless and high-pressure conditions. Strong temperature inversions up to 28 °C were detected between the valley floor and the surrounding hills. The spatial temperature structure of the CAPs varies among the different inversion days, with the upper boundary of the cold pool generally situated at about 1150 m asl. Although mean temperatures have increased in this area over the period 1960-2015 in connection with climate change, the occurrences of extreme cold temperatures did not decrease in winter and are highly correlated with the North Atlantic Oscillation and the East Atlantic indices. This suggests that CAPs in sheltered valleys are largely decoupled from the free atmosphere temperature and will likely continue to occur in the next decades under warmer conditions.

  14. Development of database for the divertor recycling in JT-60U and its analysis

    Energy Technology Data Exchange (ETDEWEB)

    Takizuka, Tomonori; Shimizu, Katsuhiro; Hayashi, Nobuhiko; Asakura, Nobuyuki [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Arakawa, Kazuya [Komatsu, Ltd., Tokyo (Japan)

    2003-05-01

    We have developed a database for the divertor recycling in JT-60U plasmas. This database makes it possible to investigate behaviors of the neutral-particle flux in plasmas and the ion flux to divertor plates under a condition for core-plasma parameters, such as electron density and heating power. The correlation between the electron density and the heating power is not strong in this database, and parameter scans for the density and the power in wide ranges are realized. On the basis of this database, we have analyzed the ion flux to divertor plates. The divertor-plate ion flux amplified by the recycling grows nonlinearly with the increase of the electron density n{sub e}. Its averaged dependence is a linear growth ({approx}n{sub e}{sup 1.0}) at the low density, and becomes a nonlinear growth ({approx}n{sub e}{sup 1.5}) at the high density. The spread of dependence from the averaged one is very large. This spread is caused mainly by complex physical characteristics of divertor plasmas, though it is little dependent on the heating power. The behavior of ion flux depends strongly on divertor configurations and divertor-plate/first-wall conditions. It is confirmed that the bifurcated transition takes place from the low-recycling divertor plasma at the low density to the high-recycling divertor plasma at the high density. The density at the transition is nearly proportional to the 1/4 power of the heating power. (author)

  15. Photon trapping effects in DEMO divertor plasma

    Energy Technology Data Exchange (ETDEWEB)

    Hoshino, K.; Tokunaga, S.; Asakura, N. [Japan Atomic Energy Agency, Rokkasho, Aomori (Japan); Sawada, K.; Idei, R. [Faculty of Engineering, Shinshu Univ., Nagano (Japan); Shimizu, K. [Japan Atomic Energy Agency, Naka, Ibaraki (Japan); Ohno, N. [Graduate School of Engineering, Nagoya Univ, Aichi (Japan)

    2016-08-15

    In the DEMO divertor, the neutral density becomes high to produce the full detachment and therefore the photon trapping can become important. In this paper, effects of the photon trapping on the DEMO divertor plasma has been studied. The pre-evaluation of the photon trapping effects on the fixed background plasma profile was carried out by using an iterative self-consistent collisional radiative model. The recombining plasma near the inner target and the private region changed to the ionizing plasma by the photon-excitation. Based on the preevaluation result, the database of the effective ionization rate coefficient including the photon transport inside a 2 mm sphere. Advantage of the 2 mm sphere approximation is that the extra calculation cost is not necessary. Iterative calculation of the SONIC including the photon trapping effects was carried out. While the electron density increased and the neutral density decreased in the wide region, the electron density decreases close to the inner strike point. This may be due to decrease in the ionization rate by decrease in the neutral density. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  16. Progress in ergodic divertor operation on Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Ghendrih, Ph.; Becoulet, M.; Colas, L.; Grosman, A.; Guirlet, R.; Gunn, J.; Loarer, T.; Azeroual, A.; Basiuk, V.; Beaumont, B.; Becoulet, A.; Bremond, S.; Bucalossi, J.; Capes, H.; Corre, Y.; Costanzo, L.; Michelis, C. de; Devynck, P.; Feron, S.; Friant, C.; Garbet, X.; Giannella, R.; Grisolia, C.; Hess, W.; Hogan, J.; Ladurelle, L.; Laugier, F.; Martin, G.; Mattioli, M.; Meslin, B.; Monier-Garbet, P.; Moulin, D.; Nguyen, F.; Pascal, J.Y.; Pecquet, A.L.; Pegourie, B.; Reichle, R.; Saint-Laurent, F.; Vallet, J.C.; Zabiego, M

    1999-09-01

    Upgrade of the Tore ergodic divertor has led to significant progress in ergodic divertor physics. The disruptive limit governed by the stochastization of the outer magnetic surfaces is found to occur for a value of the Chirikov parameter reaching 2 on the magnetic surface q = 2 + 3 / 12. This experimentally observed robustness allows one to operate at very low safety factor on the separatrix (q {approx} 2). Numerical analysis of ballooning turbulence in a stochastic layer indicates that the decay of the density fluctuations is in associated with an increase of the fluctuating electric drift velocity. The bottom line is then an enhanced cross-field transport in the vicinity of the target plates. This lowering of confinement appears to be compensated by an intrinsic transport barrier on the electron temperature. The 3-D response of the temperature field is computed with a fluid code. The intrinsic transport barrier at the separatrix, reported experimentally, can be recovered together with small amplitude temperature modulations in the divertor volume. Experimental evidence of the 3 density regimes (linear, high recycling and detachment) is reported. The low critical density values for these transitions indicate that similar parallel physics govern the axisymmetric and ergodic divertor, despite the open configuration of the latter. Measurement and understanding of these density regimes provide a means for feedback control of plasma density and an improvement in ICRH coupling scenarios. Experimental data also indicated that particle control with the vented target plates is effective. Increase of impurity control and radiation efficiency are recalled. Global power balance has been analysed. These results confirm the enhanced radiation capacity of the ergodic divertor. (author)

  17. The edge plasma and divertor in TIBER

    Energy Technology Data Exchange (ETDEWEB)

    Barr, W.L.

    1987-10-16

    An open divertor configuration has been adopted for TIBER. Most recent designs, including DIII-D, NET and CIT use open configurations and rely on a dense edge plasma to shield the plasma from the gas produced at the neutralizer plate. Experiments on ASDEX, PDX, D-III, and recently on DIII-D have shown that a dense edge plasma can be produced by re-ionizing most of the gas produced at the plate. This high recycling mode allows a large flux of particles to carry the heat to the plate, so that the mean energy per particle can be low. Erosion of the plate can be greatly reduced if the average impact energy of the ions at the plate can be reduced to near or below the threshold for sputtering of the plate material. The present configuration allows part of the flux of edge plasma ions to be neutralized at the entrance to the pumping duct so that helium is pumped as well as hydrogen. 7 refs., 3 figs.

  18. Thermal-hydraulic analysis of the HL-2M divertor using an homogeneous equilibrium model

    Science.gov (United States)

    Lu, Yong; Cai, Lijun; Liu, Yuxiang; Liu, Jian; Yuan, Yinglong; Zheng, Guoyao; Liu, Dequan

    2017-09-01

    The heat flux of the HL-2M divertor would reach 10 MW m-2 or more at the local area when the device operates at high parameters. Subcooled boiling could occur at high thermal load, which would be simulated based on the homogeneous equilibrium model. The results show that the current design of the HL-2M divertor could withstand the local heat flux 10 MW m-2 at a plasma pulse duration of 5 s, inlet coolant pressure of 1.5 MPa and flow velocity of 4 m s-1. The pulse duration that the HL-2M divertor could withstand is closely related to the coolant velocity. In addition, at the time of 2 min after plasma discharge, the flow velocity decreased from 4 m s-1 to 1 m s-1, and the divertor could also be cooled to the initial temperature before the next plasma discharge commences.

  19. Magnetic field models and their application in optimal magnetic divertor design

    Energy Technology Data Exchange (ETDEWEB)

    Blommaert, M.; Reiter, D. [Institute of Energy and Climate Research (IEK-4), FZ Juelich GmbH, Juelich (Germany); Baelmans, M. [KU Leuven, Department of Mechanical Engineering, Leuven (Belgium); Heumann, H. [TEAM CASTOR, INRIA Sophia Antipolis (France); Marandet, Y.; Bufferand, H. [Aix-Marseille Universite, CNRS, PIIM, Marseille (France); Gauger, N.R. [TU Kaiserslautern, Chair for Scientific Computing, Kaiserslautern (Germany)

    2016-08-15

    In recent automated design studies, optimal design methods were introduced to successfully reduce the often excessive heat loads that threaten the divertor target surface. To this end, divertor coils were controlled to improve the magnetic configuration. The divertor performance was then evaluated using a plasma edge transport code and a ''vacuum approach'' for magnetic field perturbations. Recent integration of a free boundary equilibrium (FBE) solver allows to assess the validity of the vacuum approach. It is found that the absence of plasma response currents significantly limits the accuracy of the vacuum approach. Therefore, the optimal magnetic divertor design procedure is extended to incorporate full FBE solutions. The novel procedure is applied to obtain first results for the new WEST (Tungsten Environment in Steady-state Tokamak) divertor currently under construction in the Tore Supra tokamak at CEA (Commissariat a l'Energie Atomique, France). The sensitivities and the related divertor optimization paths are strongly affected by the extension of the magnetic model. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  20. Plasma facing components integration studies for the WEST divertor

    Energy Technology Data Exchange (ETDEWEB)

    Ferlay, Fabien, E-mail: fabien.ferlay@cea.fr; Missirlian, Marc; Guilhem, Dominique; Firdaouss, Mehdi; Richou, Marianne; Doceul, Louis; Faisse, Frédéric; Languille, Pascal; Larroque, Sébastien; Martinez, André; Proust, Maxime; Louison, Céphise; Jeanne, Florian; Saille, Alain; Samaille, Frank; Verger, Jean-Marc; Bucalossi, Jérôme

    2015-10-15

    Highlights: • The divertor PFU integration has been studied regarding existing environment. • Magnetic, electric, thermal, hydraulic, mechanical loads and assembly are considered. - Abstract: In the context of the Tokamak Tore-Supra evolution, the CEA aims at transforming it into a test bench for ITER actively cooled tungsten (ACW) plasma facing components (PFC). This project named WEST (Tungsten Environment in Steady state Tokamak) is especially focused on the divertor target. The modification of the machine, by adding two axisymmetric divertors will make feasible an H-mode with an X-point close to the lower divertor. This environment will allow exposing the divertor ACW components up to 20 MW/m{sup 2} heat flux during long pulse. These specifications are well suited to test the ITER-like ACW target elements, respecting the ITER design. One challenge in such machine evolution is to integrate components in an existing vacuum vessel in order to obtain the best achievable performance. This paper deals with the design integration of ITER ACW target elements into the WEST environment considering magnetic, electric, thermal and mechanical loads. The feasibility of installation and maintenance has to be strongly considered as these PFC could be replaced several times. The ports size allows entering a 30° sector of pre-installed tungsten targets which will be plugged as quickly and easily as possible. The main feature of steady state operation is the active cooling, which leads to have many embedded cooling channels and bulky pipes on the PFC module including many connections and sealings between vacuum and water channels. The 30° sector design is now finalized regarding the ITER ACW elements specifications. No major modifications are expected.

  1. A Lithium Vapor Box Divertor Similarity Experiment

    Science.gov (United States)

    Cohen, Robert A.; Emdee, Eric D.; Goldston, Robert J.; Jaworski, Michael A.; Schwartz, Jacob A.

    2017-10-01

    A lithium vapor box divertor offers an alternate means of managing the extreme power density of divertor plasmas by leveraging gaseous lithium to volumetrically extract power. The vapor box divertor is a baffled slot with liquid lithium coated walls held at temperatures which increase toward the divertor floor. The resulting vapor pressure differential drives gaseous lithium from hotter chambers into cooler ones, where the lithium condenses and returns. A similarity experiment was devised to investigate the advantages offered by a vapor box divertor design. We discuss the design, construction, and early findings of the vapor box divertor experiment including vapor can construction, power transfer calculations, joint integrity tests, and thermocouple data logging. Heat redistribution of an incident plasma-based heat flux from a typical linear plasma device is also presented. This work supported by DOE Contract No. DE-AC02-09CH11466 and The Princeton Environmental Institute.

  2. Experimental study of the topological aspect of the ergodic divertor in Tore-supra tokamak; Etude experimentale des aspects topologiques du divertor ergodique de Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Costanzo, L

    2001-10-01

    The control of power deposition onto plasma facing components in tokamaks is a determining factor for future thermonuclear fusion reactors. Plasma surface interaction can be performed using limiters or divertors. The ergodic divertor installed on Tore Supra is an atypical example of a magnetic divertor. It consists in applying a magnetic perturbation which establishes a particular topology of the plasma in contact with the wall (edge plasma). We carried out dedicated experiments in order to study parallel heat flux which strike the divertor neutralizers. This quantitative and qualitative analysis of heat flux as a function of experimental conditions allows to determine the profiles of power deposition along the neutralizers. The influence of plasma electron density, additional heating, impurities and injected gas was established. An experimental study of the sheath heat transmission factor {gamma} was carried out by correlating measurements made with Langmuir probes and infrared imaging. This study gave rise to a major conclusion: for ohmic discharges with deuterium injection and most of the time with helium, it was experimentally confirmed that {gamma}=7 in agreement with classical sheath theory. However, an increase of this factor with additional power has been shown. Detached plasma, which is an attractive regime in order to reduce the power deposition, requires an optimized control. A new measurement of the detachment onset has been developed. It is based on the variation of heat flux onto the plates derived from infrared measurements. A detachment cartography with the determination of a new 2D 'IR' Degree of Detachment was carried out allowing to locate the zone where the detachment starts. We can apply this concept both to other tokamaks such as JET and ITER. A comparison between the axisymmetric divertor and the ergodic divertor is also presented concerning the power deposition in the two configurations. Low heat flux with the ergodic divertor is a

  3. Effect of the magnetic topology of a tokamak divertor on the power exhaust properties

    Science.gov (United States)

    Pericoli Ridolfini, V.; Ambrosino, R.; Calabrò, G.; Crisanti, F.; Lombroni, R.; Mastrostefano, S.; Rubino, G.; Zagórski, R.

    2017-08-01

    The peculiarities of various advanced divertor magnetic configurations that could be adopted for a tokamak reactor are investigated with the 2D edge code TECXY applied to the different divertor options of the projected tokamak DTT (Divertor Test Tokamak). The analysis highlights very interesting features for those configurations that realize a wide region with significantly depressed poloidal field in between the main X point and the target. Here, the energy cross-field diffusion can become so fast to extend up to ≈10 times the width of the power flow channel, in terms of the poloidal flux coordinates. This can spread the power over a long length and then drop the peak heat load below the technologically safe value, even with no help from impurities. Furthermore, the strongly enlarged effective divertor volume can favour the dissipative processes and lead to plasma detachment from the associated target. The driving mechanism appears to rest on the strongly increased connection lengths. This reduces the parallel thermal gradient and then slows down the power streaming, hence forcing the flow channel to widen in order to convey the same amount of power. However, the other target can be significantly penalized by an unbalance in the power sharing between the two divertor plates. Similarly, modifying the topology of this region also could overcome this problem.

  4. Particle transport in the vicinity of divertor separatrix

    Science.gov (United States)

    Nishimura, Y.; Lyu, J. C.

    2017-10-01

    Guiding center orbit following code in a tokamak edge geometry is developed which connects straight field line coordinate system (away from the separatrix) and Cartesian coordinate system (in the vicinity of the separatrix) smoothly in the equation of motion. In the presence of magnetic stochasticity charged particles in the closed magnetic field line region can be transported to the open field line region and then hit the divertor plates within several toroidal transits. Our preliminary studies suggest finite heat load both on the inner and outer divertor plates. Energy spectrum of particles reaching the plates (which differs from that of the bulk plasma) as function of imposed magnetic stochasticity, is analyzed. This work is supported by Taiwan MOST 104-2112-M-006-019.

  5. Moving Divertor Plates in a Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    S.J. Zweben, H. Zhang

    2009-02-12

    Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions.

  6. Rapidly Moving Divertor Plates In A Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    S. Zweben

    2011-05-16

    It may be possible to replace conventional actively cooled tokamak divertor plates with a set of rapidly moving, passively cooled divertor plates on rails. These plates would absorb the plasma heat flux with their thermal inertia for ~10-30 sec, and would then be removed from the vessel for processing. When outside the tokamak, these plates could be cooled, cleaned, recoated, inspected, and then returned to the vessel in an automated loop. This scheme could provide nearoptimal divertor surfaces at all times, and avoid the need to stop machine operation for repair of damaged or eroded plates. We describe various possible divertor plate designs and access geometries, and discuss an initial design for a movable and removable divertor module for NSTX-U.

  7. Vlasov Fokker Planck Study of Electron Dynamics in the Scrape Off Layer with Expander Divertor

    Science.gov (United States)

    Gupta, S.; Yushmanov, P.; Tae Team, The; Barnes, D. C.

    2017-10-01

    Control of electron heat losses in the open field region surrounding a Field Reversed Configuration (FRC) is important for sustaining higher temperatures in the FRC core, for favorable beam energy deposition, and for reducing loads on divertor plates. At TAE, a magnetic expander will be used to attain these objectives in the new C-2W machine and to comprehensively study expander divertor physics. The electron dynamics and electrostatic potential formation in the expanding magnetic field is analyzed using a 3-D (2 velocity and 1 spatial) Vlasov Fokker Planck code (Ksol). Numerical results showing the effect of collisionality, current, Zeff, incoming distribution etc., on the formation of electrostatic potentials will be presented.

  8. Toroidally symmetric/asymmetric effect on the divertor flux due to neon/nitrogen seeding in LHD

    Directory of Open Access Journals (Sweden)

    H. Tanaka

    2017-08-01

    Full Text Available Toroidal distributions of divertor particle flux during neon (Ne and nitrogen (N2 seeded discharges were investigated in the Large Helical Device (LHD. By using 14 toroidally distributed divertor probe arrays, which were positioned at radially inner side where the divertor flux concentrates in the inward-shifted magnetic axis configuration, it is found that Ne puffing leads to toroidally quasi-uniform reduction of divertor particle fluxes; whereas toroidally localized reductions were observed with N2 puffing. The toroidally asymmetric reduction pattern with N2 puffing is strongly related to the magnetic field structure around the N2 puffing port. Assuming that nitrogen particles do not recycle, EMC3-EIRENE simulation shows similar reduction pattern with the experiment around the N2 puffing port.

  9. Flute instability in the tandem mirror with the divertor/dipole regions

    Energy Technology Data Exchange (ETDEWEB)

    Katanuma, I.; Masaki, S.; Sato, S.; Sekiya, K.; Ichimura, M.; Imai, T. [Plasma Research Center, University of Tsukuba, Tsukuba, Ibaraki 305-8577 (Japan)

    2011-11-15

    The numerical simulation is performed in GAMMA10 A-divertor magnetic configuration, which is a candidate of remodeled device of the GAMMA10 tandem mirror [M. Inutake et al., Phys. Rev. Lett. 55, 939 (1985)]. Both divertor and dipole regions are included in the numerical calculation, which is a new point. The electron short circuit effect along x-point, therefore, is not assumed so that it is not used the boundary condition of the electrostatic perturbations being zero at the separatrix on which the magnetic field lines pass through x-point. The simulation results reveal that the dipole field plays a role of a good magnetic field line curvature to the GAMMA10 A-divertor, and so the flute modes are stabilized without help of electron short circuit effects.

  10. Automated magnetic divertor design for optimal power exhaust

    Energy Technology Data Exchange (ETDEWEB)

    Blommaert, Maarten

    2017-07-01

    The so-called divertor is the standard particle and power exhaust system of nuclear fusion tokamaks. In essence, the magnetic configuration hereby 'diverts' the plasma to a specific divertor structure. The design of this divertor is still a key issue to be resolved to evolve from experimental fusion tokamaks to commercial power plants. The focus of this dissertation is on one particular design requirement: avoiding excessive heat loads on the divertor structure. The divertor design process is assisted by plasma edge transport codes that simulate the plasma and neutral particle transport in the edge of the reactor. These codes are computationally extremely demanding, not in the least due to the complex collisional processes between plasma and neutrals that lead to strong radiation sinks and macroscopic heat convection near the vessel walls. One way of improving the heat exhaust is by modifying the magnetic confinement that governs the plasma flow. In this dissertation, automated design of the magnetic configuration is pursued using adjoint based optimization methods. A simple and fast perturbation model is used to compute the magnetic field in the vacuum vessel. A stable optimal design method of the nested type is then elaborated that strictly accounts for several nonlinear design constraints and code limitations. Using appropriate cost function definitions, the heat is spread more uniformly over the high-heat load plasma-facing components in a practical design example. Furthermore, practical in-parts adjoint sensitivity calculations are presented that provide a way to an efficient optimization procedure. Results are elaborated for a fictituous JET (Joint European Torus) case. The heat load is strongly reduced by exploiting an expansion of the magnetic flux towards the solid divertor structure. Subsequently, shortcomings of the perturbation model for magnetic field calculations are discussed in comparison to a free boundary equilibrium (FBE) simulation

  11. A semi-automatic image-based close range 3D modeling pipeline using a multi-camera configuration

    National Research Council Canada - National Science Library

    Rau, Jiann-Yeou; Yeh, Po-Chia

    2012-01-01

    .... This study proposes an image-based 3D modeling pipeline which takes advantage of a multi-camera configuration and multi-image matching technique that does not require any markers on or around the object...

  12. Impact of the plasma geometry on divertor power exhaust: experimental evidence from TCV and simulations with SolEdge2D and TOKAM3X

    Science.gov (United States)

    Gallo, A.; Fedorczak, N.; Elmore, S.; Maurizio, R.; Reimerdes, H.; Theiler, C.; Tsui, C. K.; Boedo, J. A.; Faitsch, M.; Bufferand, H.; Ciraolo, G.; Galassi, D.; Ghendrih, P.; Valentinuzzi, M.; Tamain, P.; the EUROfusion MST1 team; the TCV team

    2018-01-01

    A deep understanding of plasma transport at the edge of magnetically confined fusion plasmas is needed for the handling and control of heat loads on the machine first wall. Experimental observations collected on a number of tokamaks over the last three decades taught us that heat flux profiles at the divertor targets of X-point configurations can be parametrized by using two scale lengths for the scrape-off layer (SOL) transport, separately characterizing the main SOL ({λ }q) and the divertor SOL (S q ). In this work we challenge the current interpretation of these two scale lengths as well as their dependence on plasma parameters by studying the effect of divertor geometry modifications on heat exhaust in the Tokamak à Configuration Variable. In particular, a significant broadening of the heat flux profiles at the outer divertor target is diagnosed while increasing the length of the outer divertor leg in lower single null, Ohmic, L-mode discharges. Efforts to reproduce this experimental finding with both diffusive (SolEdge2D-EIRENE) and turbulent (TOKAM3X) modelling tools confirm the validity of a diffusive approach for simulating heat flux profiles in more traditional, short leg, configurations while highlighting the need of a turbulent description for modified, long leg, ones in which strongly asymmetric divertor perpendicular transport develops.

  13. Clinical study of internal derangement of the temporomandibular joint with closed lock, 2; Correlation of the disk configuration at MR imaging with clinical parameters

    Energy Technology Data Exchange (ETDEWEB)

    Moriya, Yoshiyuki; Murakami, Ken-ichiro; Fujimura, Kazuma; Yokoyama, Tadaaki; Nose, Masahiro; Miyaki, Katsuaki; Segami, Natsuki; Iizuka, Tadahiko (Kyoto Univ. (Japan). Faculty of Medicine)

    1990-10-01

    Fifty-three closed lock cases of internal derangement of the temporomandibular joint (TMJ) were studied on the correlation between disk configuration at MR imaging and nine clinical parameters composed of opening degree, age, clicking and locking duration, visual analogue scale of pain (VAS), pain score, jaw dysfunction score, life activity limited score, and TMJ X-ray photo findings. Disk configuration and degree of anterior disk displacement were assessed on MR imaging in closed mouth position: the antero-posterior length of disk and the distance from condyle to anterior and posterior portion of disk were measured, respectively. Duration of clicking and locking were not correlated with MR index except that there was a strong correlation between clicking duration and the distance from condyle to anterior portion of the disk at MR imaging. Opening degree was related to the disk deformity and the access of posterior portion of the disk to condyle. Disk configuration and degree of anterior disk displacement were not correlated with TMJ pain, but jaw dysfunction was related to the disk deformity and the distance from posterior portion of the disk to condyle. TMJ X-ray photo findings were not correlated with clicking duration (below 3 years), locking duration (below 30 weeks), opening degree and disk configuration at MR imaging. In evaluation of factors related to opening degree in 53 patients with closed lock by means of multiple regression analysis, age, locking and clicking duration, TMJ pain, life activity limited score were more strongly correlated to opening degree than the others. (author).

  14. Alternative power exhaust studies in an advanced upper divertor in ASDEX Upgrade supported by SOLPS and EMC3-EIRENE simulations

    Science.gov (United States)

    Lunt, Tilmann; Pan, Ou; Herrmann, Albrecht; Coster, David; Dunne, Mike; Feng, Yuehe; Kallenbach, Arne; Wischmeier, Marco; Zohm, Hartmut; ASDEX Upgrade Team

    2017-10-01

    In order to study alternative divertor configurations, currently discussed as a possible solution for the power exhaust problem in a fusion reactor, the installation of a pair of in-vessel poloidal field coils in the upper divertor of ASDEX Upgrade was recently decided. Besides the conventional single- and double null configurations, a series of new configurations ranging from an X- divertor, to a low- (LFS SF-) and finally a high field side snowflake minus will be possible with these coils in a machine with a high P / R ratio. The arangement of these coils was based on the pioneering work of TCV as well as simulations with EMC3-EIRENE, which can rather easily handle topologies with two X-points and which identified a series of heat flux mitigation effects. Due to the lack of drifts and volumetric recombination in the code, however, a clear prediction on the detachment degree and threshold is missing as well as a realistic description of the in-out divertor asymmetries. This limit has now been overcome by creating an adequate computational grid for a LFS SF- configuration for SOLPS. In this contribution we will present the worldwide first simulation on this grid as well as the upgrade plans and discuss the potential different heat flux mitigation mechanisms.

  15. A Comparison of Closed-Loop Performance of Multirotor Configurations Using Non-Linear Dynamic Inversion Control

    Directory of Open Access Journals (Sweden)

    Murray L. Ireland

    2015-06-01

    Full Text Available Multirotor is the umbrella term for the family of unmanned aircraft, which include the quadrotor, hexarotor and other vertical take-off and landing (VTOL aircraft that employ multiple main rotors for lift and control. Development and testing of novel multirotor designs has been aided by the proliferation of 3D printing and inexpensive flight controllers and components. Different multirotor configurations exhibit specific strengths, while presenting unique challenges with regards to design and control. This article highlights the primary differences between three multirotor platforms: a quadrotor; a fully-actuated hexarotor; and an octorotor. Each platform is modelled and then controlled using non-linear dynamic inversion. The differences in dynamics, control and performance are then discussed.

  16. Experience on divertor fuel retention after two ITER-Like Wall campaigns

    Science.gov (United States)

    Heinola, K.; Widdowson, A.; Likonen, J.; Ahlgren, T.; Alves, E.; Ayres, C. F.; Baron-Wiechec, A.; Barradas, N.; Brezinsek, S.; Catarino, N.; Coad, P.; Guillemaut, C.; Jepu, I.; Krat, S.; Lahtinen, A.; Matthews, G. F.; Mayer, M.; Contributors, JET

    2017-12-01

    The JET ITER-Like Wall experiment, with its all-metal plasma-facing components, provides a unique environment for plasma and plasma-wall interaction studies. These studies are of great importance in understanding the underlying phenomena taking place during the operation of a future fusion reactor. Present work summarizes and reports the plasma fuel retention in the divertor resulting from the two first experimental campaigns with the ITER-Like Wall. The deposition pattern in the divertor after the second campaign shows same trend as was observed after the first campaign: highest deposition of 10–15 μm was found on the top part of the inner divertor. Due to the change in plasma magnetic configurations from the first to the second campaign, and the resulted strike point locations, an increase of deposition was observed on the base of the divertor. The deuterium retention was found to be affected by the hydrogen plasma experiments done at the end of second experimental campaign.

  17. Magnetohydrodynamic transport characterization of a field reversed configuration

    Science.gov (United States)

    Onofri, M.; Yushmanov, P.; Dettrick, S.; Barnes, D.; Hubbard, K.; Tajima, T.

    2017-09-01

    The transport phenomenon of a Field Reversed Configuration (FRC) is studied using the newly developed two-dimensional code Q2D, which couples a magnetohydrodynamic code with a Monte Carlo code for the beam component. The simulation by Q2D of the transport parallel to the simple open θ-pinch fields and its associated outflow phenomenon shows an excellent agreement with one of the leading theories, elevating the Q2D validity and simultaneously deepening the theoretical understanding of this fundamental process. We find a sharp distinction between the evolved radial density profiles of the FRC and mirror plasmas as a result of the transport processes, underpinning the crucial role of the closed flux surfaces of the FRC to enhance the confinement over that of the mirror. We characterize the scrap-off layer (SOL) transport by including the mirror trapping effects, and we find a relationship between the confinement time in the SOL and the ion collisional time. The Q2D code further illuminates the basic transport properties of the divertor region and the formation of an electrostatic potential in the divertor.

  18. Poloidal magnetics of a divertor compact ignition tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Strickler, D.J.; Peng, Y.K.M.; Jardin, S.C.

    1987-10-01

    A technique is presented for calculating bounds on the poloidal field (PF) coil currents required to constrain critical plasma shape parameters when plasma pressure and current density profiles are changed. Such considerations are important in the conceptual design of the PF coils for the Compact Ignition Tokamak (CIT) and their electrical power systems in view of the uncertainty in plasma profiles and operating scenarios. Four relatively independent coil groups are sufficient to find a coil current distribution and equilibrium satisfying a prescribed plasma major radius, minor radius, and divertor strike point coordinates. The variation in the coil current distribution with plasma profiles tends to be large for external PF systems and provides a measure by which coil configurations may be compared. 6 refs., 7 figs., 4 tabs.

  19. Plasma recombination and molecular effects in tokamak divertors and divertor simulators

    Energy Technology Data Exchange (ETDEWEB)

    Krasheninnikov, S.I.; Pigarov, A.Y.; Knoll, D.A.; LaBombard, B.; Lipschultz, B.; Sigmar, D.J.; Soboleva, T.K.; Terry, J.L.; Wising, F. [Plasma Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States)]|[Department of Physics, The College of William and Mary, Williamsburg, Virginia 23187 (United States)]|[Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States)]|[Instituto de Ciencias Nucleares, Universidad Nacional Autonoma de Mexico, Mexico D.F. (Mexico)]|[Institute for Electromagnetic Field Theory, Chalmers University of Technology, S-41296 Gothenburg (Sweden)

    1997-05-01

    Analysis of the experimental data from tokamaks and linear divertor simulators leads to the conclusion that plasma recombination is a crucial element of plasma detachment. Different mechanisms of plasma recombination relevant to the experimental conditions of the tokamak scrape-off layer (SOL) and divertor simulators are considered. The physics of Molecular Activated Recombination (MAR) involving vibrationally excited molecular hydrogen are discussed. Although conventional Electron{endash}Ion Recombination (EIR) alone can strongly alter the plasma parameters, MAR impact can be substantial for both tokamak SOL plasma and divertor simulators. Investigation of the effects of EIR on the plasma flow in divertor simulators shows that due to the balances of (a) energy transport and electron cooling, and (b) the plasma flow and recombination, that EIR extinguishes the simulator plasma at an electron temperature about 0.15 eV. {copyright} {ital 1997 American Institute of Physics.}

  20. Conceptual design of CFETR divertor remote handling compatible structure

    Energy Technology Data Exchange (ETDEWEB)

    Dai, Huaichu, E-mail: yaodm@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei (China); Yao, Damao; Cao, Lei; Zhou, Zibo; Li, Lei [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2016-11-15

    Highlights: • Conceptual design for the CFETR divertor have been proposed, especially the divertor remote handling compatible structure. • The degrees of freedom of the divertor are analyzed in order to validate the design the divertor supports structure. • Besides the ITER-like scheme, a new scheme for the divertor remote handling compatible supports is proposed, that is the rack and pinion mechanism. • The installation/removel process is verified through simulation in Delmia in order to check design quality for remote handling requirements. - Abstract: Divertor is one of key components of tokamak fusion reactor. The CFETR is China Fusion Engineering Test Reactor. Its divertor will expose to tritium environment and neutron radiation. Materials of the divertor will be radioactived, and cannot be handled by personnel directly. To develop structure which compatible with robots handle for installation, maintenance and removing is required. This paper introduces a conceptual design of CFETR divertor module which compatible with remote handling end-effectors. The divertor module is confined by inner and outer support. The inner support is only confined divertor module radial, toroidal and vertical moving freedom degrees, but not confined rotating freedom degrees. The outer support is the structure that can confine rotating freedom degrees and should also be compatible with remote handling end-effectors.

  1. Phase-induced transparency-mediated structured-beam generation in a closed-loop tripod configuration

    Science.gov (United States)

    Sharma, Sandeep; Dey, Tarak N.

    2017-09-01

    We present a phase-induced transparency-based scheme to generate structured-beam patterns in a closed four-level atomic system. We employ a phase-structured probe beam and a transverse magnetic field (TMF) to create phase-dependent medium susceptibility. We show that such phase-dependent modulation of absorption holds the key to formation of a structured beam. We use a full density matrix formalism to explain the experiments of Radwell et al. [Phys. Rev. Lett. 114, 123603 (2015), 10.1103/PhysRevLett.114.123603] at weak probe limits. Our numerical results on beam propagation confirms that the phase information present in the absorption profile gets encoded on the spatial probe envelope, which creates petal-like structures even in the strong field limit. The contrast of the formed structured beam can be enhanced by changing the strength of TMF as well as of the probe intensity. In weak field limits an absorption profile is solely responsible for creating a structured beam, whereas in the strong probe regime, both dispersion and absorption profiles facilitate the generation of a high-contrast structured beam. Furthermore, we find the rotation of structured beams owing to strong-field-induced nonlinear magneto-optical rotation.

  2. Radiative detached divertor with acceptable separatrix Zeff

    Science.gov (United States)

    Pigarov, A. Yu.

    2017-10-01

    The feasibility study is performed for the radiative detached divertor (RDD) concept, which characterizes a variety of detached plasmas with impurity amounts providing the highest levels of divertor radiation without X-point MARFE (XPM), based on a set of restrictive criteria for the leading plasma parameters (LPPs) including, e.g., tolerable peak power loads below 1 MW/m2, low separatrix densities, allowable impurity concentrations, and acceptable Zeff values for a DIII-D like tokamak. For this, extensive simulations with the 2-D edge plasma transport code were done scanning the deuterium and impurity inventories practically for all impurity elements from beryllium to neon and the analysis of LPP variations in these scans is presented. It is shown that, for a given D inventory, the total radiation fraction with an increase in the impurity inventory reaches a flat top level, frad = 0.85 ± 0.01, whereas the higher frad corresponds to XPM. This critical fraction is the same for all elements and values of the D inventory. Successful RDD solutions with a flat top radiation meeting all ad hoc LPP criteria are found for some elements. Boron and nitrogen are shown to be the most promising elements for seeding, since they are capable of providing alone the successful RDD at the lowest concentrations. Several important effects on impurity radiation are considered including: cross-field impurity transport in regions with strong temperature gradients, multi-species thermal force, charge-exchange of impurity ions with D atoms originating from recombination, impurity entrainment by parallel flows, flows caused by inner/outer divertor asymmetries, and Mach ˜ 1 flows reached inside radiation-ionization fronts. The impurity radiation profiles of various elements are analyzed suggesting three patterns differing in the radiation front position with respect to the D ionization source. The modeled relocation of D from the pedestal into divertor regions; an enhanced pedestal

  3. Divertor research on the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Hill, D.N.; Allen, S.L. [Lawrence Livermore National Lab., CA (United States); Brooks, N.H. [General Atomics, San Diego, CA (United States)] [and others

    1994-10-01

    In this paper the authors summarize recent progress on DIII-D in developing techniques for divertor power and particle control relevant to next generation tokamaks such as the proposed ITER and TPX devices. Density control and helium removal by divertor pumping have been demonstrated for the first time in high confinement ELMing H-mode discharges ({tau} {approximately} 2 {times} {tau}{sub ITER-89P}) following installation of a divertor cryopumping system. The peak divertor heat flux in similar H-mode discharges has been reduced through production of a radiating mantle with neon or argon puffing (reductions of 3--5). A number of diagnostics have been added to improve the understanding of the physical processes involved. They are now designing modified double-null divertor structures for DIII-D that will provide improved particle control for high-triangularity VH-mode plasmas while at the same time allowing for gas puffing to reduce the divertor heat flux.

  4. Role of molecular effects in divertor plasma recombination

    Directory of Open Access Journals (Sweden)

    A.S. Kukushkin

    2017-08-01

    Full Text Available Molecule-Activated Recombination (MAR effect is re-considered in view of divertor plasma conditions. A strong isotopic effect is demonstrated. In deuterium plasmas, the reaction chain through D2+ formation, usually considered dominant and included in 2D edge plasma models, is negligible. However, in this case the other branch, through D−, usually neglected in modelling, becomes relatively strong. The overall share of MAR in divertor plasma recycling stays within 20%. The operational parameters of the divertor plasmas, such as the peak power loading on the divertor targets or the pressure limit for partial detachment of the divertor plasma, are insensitive to the presence of MAR, although the latter may be important for correct interpretation of the divertor diagnostics.

  5. NSTX Plasma Response to Lithium Coated Divertor

    Energy Technology Data Exchange (ETDEWEB)

    H.W. Kugel, M.G. Bell, J.P. Allain, R.E. Bell, S. Ding, S.P. Gerhardt, M.A. Jaworski, R. Kaita, J. Kallman, S.M. Kaye, B.P. LeBlanc, R. Maingi, R. Majeski, R. Maqueda, D.K. Mansfield, D. Mueller, R. Nygren, S.F. Paul, R. Raman, A.L. Roquemore, S.A. Sabbagh, H. Schneider, C.H. Skinner, V.A. Soukhanovskii, C.N. Taylor, J.R. Timberlak, W.R. Wampler, L.E. Zakharov, S.J. Zweben, and the NSTX Research Team

    2011-01-21

    NSTX experiments have explored lithium evaporated on a graphite divertor and other plasma facing components in both L- and H- mode confinement regimes heated by high-power neutral beams. Improvements in plasma performance have followed these lithium depositions, including a reduction and eventual elimination of the HeGDC time between discharges, reduced edge neutral density, reduced plasma density, particularly in the edge and the SOL, increased pedestal electron and ion temperature, improved energy confinement and the suppression of ELMs in the H-mode. However, with improvements in confinement and suppression of ELMs, there was a significant secular increase in the effective ion charge Zeff and the radiated power in H-mode plasmas as a result of increases in the carbon and medium-Z metallic impurities. Lithium itself remained at a very low level in the plasma core, <0.1%. Initial results are reported from operation with a Liquid Lithium Divertor (LLD) recently installed.

  6. Migration of Artificially Introduced Micron Size Carbon Dust in the DIII-D Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Rudakov, D; West, W; Wong, C; Brooks, N; Evans, T; Fenstermacher, M; Groth, M; Krasheninnikov, S; Lasnier, C; McLean, A; Pigarov, A Y; Solomon, W; Antar, G; Boedo, J; Doerner, R; Hollmann, E; Hyatt, A; Maingi, R; Moyer, R; Nagy, A; Nishino, N; Roquemore, L; Stangeby, P; Watkins, J

    2006-05-15

    Migration of pre-characterized carbon dust in a tokamak environment was studied by introducing about 30 milligrams of dust flakes 5-10 {micro}m in diameter in the lower divertor of DIII-D using the DiMES sample holder. The dust was exposed to high power ELMing Hmode discharges in lower-single-null magnetic configuration with the strike points swept across the divertor floor. When the outer strike point (OSP) passed over the dust holder exposing it to high particle and heat fluxes, part of the dust was injected into the plasma. In about 0.1 sec following the OSP pass over the dust, 1-2% of the total dust carbon content (2-4 x 10{sup 19} carbon atoms, equivalent to a few million dust particles) penetrated the core plasma, raising the core carbon density by a factor of 2-3. When the OSP was inboard of the dust holder, the dust injection continued at a lower rate. Individual dust particles were observed moving at velocities of 10-100 m/s, predominantly in the toroidal direction for deuteron flow to the outer divertor target, consistent with the ion drag force. The observed behavior of the dust is in qualitative agreement with modeling by the 3D DustT code.

  7. Examining Diagnostic Capability for Determining Divertor Neutral Sourcing to the Pedestal on DIII-D

    Science.gov (United States)

    Shafer, Morgan; Briesemeister, Alexis; Canik, John; Park, Jin Myung; Unterberg, Ezekial; Leonard, Anthony; Guo, Houyang; Moser, Auna

    2017-10-01

    Neutral fueling from the divertor plays a key role in setting the density pedestal, but can not yet be predicted via numerical models and thus remains a crucial variable in predictive core-edge coupling. New neutral diagnostics are planned to address this issue by constraining predictions of neutral density from the divertor through the SOL into the pedestal: (a) Lyman-alpha imaging and (b) extended poloidal coverage of neutral pressure gauges. Forward modeling diagnostic responses across expected pedestal neutral fueling rates is used to estimate the diagnostic sensitivity and range of applicability. Modeled neutral source rates are obtained through interpretive modeling with the OEDGE code of experiments performed across the range of DIII-D divertor baffling configurations and gas puffing rates that result in a range of density profiles Additional forward modeling with the core/edge coupling code CESOL will be used and compared against interpretive analysis. Work supported by US DOE under DE-AC05-00OR22725, DE-FC02-04ER54698.

  8. Mitigation of divertor heat loads by strike point sweeping in high power JET discharges

    Science.gov (United States)

    Silburn, S. A.; Matthews, G. F.; Challis, C. D.; Frigione, D.; Graves, J. P.; Mantsinen, M. J.; Belonohy, E.; Hobirk, J.; Iglesias, D.; Keeling, D. L.; King, D.; Kirov, K.; Lennholm, M.; Lomas, P. J.; Moradi, S.; Sips, A. C. C.; Tsalas, M.; Contributors, JET

    2017-12-01

    Deliberate periodic movement (sweeping) of the high heat flux divertor strike lines in tokamak plasmas can be used to manage the heat fluxes experienced by exhaust handling plasma facing components, by spreading the heat loads over a larger surface area. Sweeping has recently been adopted as a routine part of the main high performance plasma configurations used on JET, and has enabled pulses with 30 MW plasma heating power and 10 MW radiation to run for 5 s without overheating the divertor tiles. We present analysis of the effectiveness of sweeping for divertor temperature control on JET, using infrared camera data and comparison with a simple 2D heat diffusion model. Around 50% reduction in tile temperature rise is obtained with 5.4 cm sweeping compared to the un-swept case, and the temperature reduction is found to scale slower than linearly with sweeping amplitude in both experiments and modelling. Compatibility of sweeping with high fusion performance is demonstrated, and effects of sweeping on the edge-localised mode behaviour of the plasma are reported and discussed. The prospects of using sweeping in future JET experiments with up to 40 MW heating power are investigated using a model validated against existing experimental data.

  9. Evidence and modeling of 3D divertor footprint induced by lower hybrid waves on EAST with tungsten divertor operations

    Science.gov (United States)

    Feng, W.; Wang, L.; Rack, M.; Liang, Y.; Guo, H. Y.; Xu, G. S.; Xu, J. C.; Liu, J. B.; Sun, Y. W.; Jia, M. N.; Yang, Q. Q.; Zhang, B.; Zou, X. L.; Liu, H.; Zhang, T.; Ding, F.; Chen, J. B.; Duan, Y. M.; Zheng, X. W.; Dai, S. Y.; Deng, G. Z.; Chen, R.; Hu, G. H.; Yan, N.; Si, H.; Liu, S. C.; Xu, S.; Wang, M.; Li, M. H.; Ding, B. J.; Wingen, A.; Huang, J.; Gao, X.; Luo, G. N.; Gong, X. Z.; Garofalo, A. M.; Li, J.; Wan, B. N.; the EAST team

    2017-12-01

    Three dimensional (3D) divertor particle flux footprints induced by the lower hybrid wave (LHW) have been systematically investigated in the EAST superconducting tokamak during the recent experimental campaign. We find that the striated particle flux (SPF) peaks away from the strike point (SP) closely fit the pitch of the edge magnetic field line for different safety factors q 95, as predicted by a field line tracing code taking into account the helical current filaments (HCFs) in the scrape-off-layer (SOL). As LHW power increases, it requires the fuelling to be increased e.g. by super molecular beam injection (SMBI), to maintain a similar plasma density, which may be attributed to the pump-out effect due to LHW, and may thus be beneficial for EAST steady state operations. The 3D SPF structure is observed with a LHW power threshold (P LHW ~ 0.9 MW). The ratio of the particle fluxes between SPF and outer strike point (OSP), i.e. {{Γ }ion,SPF}/{{Γ }ion,OSP} , increases with the LHW power. Upon transition to divertor detachment, the particle flux at the main OSP decreases, as expected, however, the particle flux at SPF continues increasing, in contrast to the RMP-induced striations that vanish with increasing divertor density. In addition, we also find that the in–out asymmetry of the 3D particle flux footprint pattern exhibits a clear dependence on the toroidal field direction (B  ×    ∇   B  ↓  and B  ×    ∇   B↑). Experiments using neon impurity seeding show a promising capability in 3D particle and heat flux control on EAST. LHW-induced particle and heat flux striations are also present in the H-mode plasmas, reducing the peak heat flux and erosion at the main strike point, thus facilitating long-pulse operation with a new steady-state H-mode over 60 s being recently achieved in EAST.

  10. Survey of coolant options of a monolithic CFC divertor

    Energy Technology Data Exchange (ETDEWEB)

    Merola, M. (Commission of the European Communities, Joint Research Centre, Institute for Advanced Materials, TP 750, 21020 Ispra (Vatican City State, Holy See) (Italy)); Matera, R. (Commission of the European Communities, Joint Research Centre, Institute for Advanced Materials, TP 750, 21020 Ispra (Vatican City State, Holy See) (Italy))

    1994-06-01

    Different coolant options for a monolithic CFC divertor are examined. Helium gas, HB-40 organic liquid and some liquid metals seem to be viable solutions. The thermal performances of the divertor concept are presented as well as a list of possible advantages and a brief cost evaluation. ((orig.))

  11. Advantages and Challenges of Radiative Liquid Lithium Divertor

    Science.gov (United States)

    Ono, Masayuki

    2017-10-01

    Steady-state fusion power plant designs present major divertor technology challenges, including high divertor heat flux both in steady-state and during transients. In addition to these concerns, there are the unresolved technology issues of long term dust accumulation and associated tritium inventory and safety issues. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid Li divertor (RLLD) concept and its variant, the active liquid Li divertor concept (ARLLD), taking advantage of the enhanced Li radiation in relatively poorly confined divertor plasmas. It has been suggested that radiation-based liquid lithium (LL) divertor concepts with a modest Li-loop could provide a possible solution for the outstanding fusion reactor technology issues such as divertor heat flux mitigation and real time dust removal, while potentially improving the reactor plasma performance. Laboratory tests are also planned to investigate the Li-T recover efficiency and other relevant research topics of the RLLD. This work supported by DoE Contract No. DE-AC02-09CH11466.

  12. Study on divertor particle and heat fluxes from electric probe measurements during ELMy H-modes in KSTAR

    Energy Technology Data Exchange (ETDEWEB)

    Bak, Jun-Gyo, E-mail: jgbak@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Kim, Heung-Su [National Fusion Research Institute, Daejeon (Korea, Republic of); Bae, Min-Keun; Chung, Kyu-Sun [Hanyang University, Seoul (Korea, Republic of); Hong, Suk-Ho [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2016-11-01

    Highlights: • The characteristics of the particle and heat fluxes were investigated during ELMs in H-modes under the LSN configuration in the KSTAR tokamak.. • There was relation between the ELM amplitude and the ELM frequency as ΔW{sub ELM}/W{sub TOT} ∝ 1/f{sub ELM} in the range of f{sub ELM} ≤ 200 Hz. • The trends of the peak amplitude of the divertor flux near the OSP during ELMs due to the ELM mitigation and the plasma shaping were investigated. • The ELMs were mitigated by MP field, SMBI and ECH. The ELM mitigations due to the MP field and the SMBI were stronger than one due to the ECH. • Finally, the particle flux, evaluated at the far scrape-off layer (SOL) region, was estimated to less than 1% of the divertor particle flux. - Abstract: The characteristics of the divertor particle and heat fluxes are investigated during ELM bursts in ELMy H-mode plasmas with the lower single null (LSN) configuration in Korea Superconducting Tokamak Advanced Research (KSTAR). The particle and heat fluxes are evaluated from the electric probe measurements at the divertor region. It is found that the peak amplitude of the divertor flux during an ELM burst obtained near the outer strike point (OSP) decreases up to about 20% as the ELM frequency increases by a factor of ∼6.5 due to the ELM mitigation and the plasma shaping, which is similar to the trend of the amplitude versus the frequency of the ELM observed in other tokamaks. The ELMs are mitigated by using several methods as magnetic perturbation (MP) field, supersonic molecular beam injection (SMBI) and electron cyclotron heating (ECH) at the edge region. In addition, the particle flux, evaluated at the far scrape-off layer (SOL) region, is less than 1% of the divertor particle flux. In this work, results from the experimental investigations of particle and heat fluxes during ELM bursts from the electric probe measurements at the divertor and far SOL regions are presented.

  13. Study of the radiation in divertor plasmas; Etude du rayonnement dans les plasmas de divertor

    Energy Technology Data Exchange (ETDEWEB)

    Laugier, F

    2000-10-19

    We have studied the cooling of the edge plasma by radiation in the divertor volume, in order to optimize the extraction of power in tokamaks and to limit the wall erosion. In attached divertor plasmas experiments, the concentration of intrinsic impurities at the edge is related to the response of the wall to the incident energy flow of plasma, depending on a phenomenological law. We carried out an analysis of the radiation according to this law and to the control parameters of the discharges. The largest radiated fraction and best synergy are obtained when the concentration of intrinsic impurities strongly increases with the energy of incident plasma. On the other hand, the erosion of the wall is stronger. In detached plasmas, we proved that the performances in terms of incident plasma energy loss and pressure loss are optimal when the density of the slowest neutrals is strong at the edge and when their radial penetration is small. On Tore Supra, we highlighted the correlations between the maximum Mach number of incident plasma flow, the radiation front and the penetration of the neutrals. A simple diagnostic based on the localization of the maximum Mach number proves that detached mode is not optimal on Tore Supra, because the radial penetration of the slowest neutrals is not sufficiently small. In the last part, we obtained the three-dimensional topology of the radiation in the ergodic divertor using a spectral analysis code and boundary conditions consistent with the temperature distribution on the wall. The radiation is maximum in front of the divertor modules. As a consequence, radiated power is underestimated by standards measurements of Tore Supra that are located between the modules. We finally showed that the profiles of temperature along the field lines are modulated, this is specific to the ergodic divertor. (author)

  14. Impact of the impurity seeding for divertor protection on the performance of fusion reactors

    Science.gov (United States)

    Siccinio, Mattia; Fable, Emiliano; Angioni, Clemente; Saarelma, Samuli; Scarabosio, Andrea; Zohm, Hartmut

    2017-10-01

    A 0D divertor and scrape-off layer (SOL) model has been coupled to the 1.5D core transport code ASTRA. The resulting numerical tool has been employed for various parameter scans in order to identify the most convenient choices for the operation of electricity producing fusion devices with seeded impurities for the divertor protection. In particular, the repercussions of such radiative species on the main plasma through the fuel dilution have been taken into account. The main result we found is that, when the limits on the maximum tolerable divertor heat flux are enforced, the curves at constant electrical power output are closed on themselves in the R-BT plane, i.e. no improvement would descend from a further increase of R or BT once the maximum has been reached. This occurrence appears as an intrinsic physical limit for all devices where a radiative SOL is needed to deal with the power exhaust. Furthermore, the relative importance of the different power loss channels (e.g. hydrogen radiation, charge exchange, perpendicular transport and impurity radiation), through which the power entering the SOL is dissipated before reaching the target plate, is investigated with our model.

  15. Extinguishing ELMs in detached radiative divertor plasmas

    Science.gov (United States)

    Pigarov, Alexander; Krasheninnikov, Sergei; Rognlien, Thomas

    2016-10-01

    In order to avoid deleterious effects of ELMs on PFCs in next-step fusion devices it has been suggested to operate with small-sized ELMs naturally extinguishing in the divertor. Our modeling effort is focusing at extinguishing type-I ELMs: conditions for expelled plasma dissipation; efficiency of ELM power handling by detached radiative divertors; and the ELM impact on detachment state. Here time-dependent modeling of a sequence of many ELMs was performed with 2-D edge plasma transport code UEDGE-MB-W which incorporates the Macro-Blob (MB) approach to simulate non-diffusive filamentary transport and various ``Wall'' (W) models for time-dependent hydrogen wall inventory and recycling. Three cases were modeled, in which extinguishing ELMs are achieved due to: (i) intrinsic impurities via graphite sputtering, (ii) extrinsic impurity gas puff (Ne), and (iii) =(i) +(ii). For each case, we performed a series of UEDGE-MB-W runs scanning the deuterium and impurity inventories, pedestal losses and ELM frequency. Temporal variations of the degree of detachment, ionization front shape, recombination sink strength, radiated fraction, peak power loads, OSP, impurity charge states, and in/out asymmetries were analyzed. We discuss the onset of extinguishing ELMs, conditions for not burning through and enhanced plasma recombination as functions of scanned parameters. Efficiencies of intrinsic and extrinsic impurities in ELM extinguishing are compared.

  16. Quartz micro-balance results of pulse-resolved erosion/deposition in the JET-ILW divertor

    Directory of Open Access Journals (Sweden)

    G. Sergienko

    2017-08-01

    Full Text Available A set of quartz crystal microbalances (QMB was used at JET with full carbon wall to monitor mass erosion/deposition rates in the remote areas of the divertor. After introduction of the ITER- like wall (ILW in JET with beryllium main wall and tungsten divertor, strong reduction of the material deposition and accompanied fuel retention was observed. Therefore the existing QMB electronics have been modified to improve the accuracy of frequency measurements by a factor of ten down to 0.1Hz which corresponds to 1.4ngcm−2. The averaged deposition rates of 1.2–3ngcm−2s−1 and erosion rates of 5.6–8.1ngcm−2s−1 were observed in the inner divertor of JET -ILW with the inner strike point positions close to the bottom edge of vertical tile 3 and at the horizontal tile 4 respectively. The erosion with averaged rates of ≈2.1ngcm−2s−1 and ≈120ngcm−2s−1 were observed in the outer divertor for the outer strike point positions at tile 5 and tile 6 respectively.

  17. Access to high-confinement regimes on Alcator C-Mod and the complex influence of divertor geometry

    Science.gov (United States)

    Hughes, J. W.; Labombard, B.; Brunner, D.; Hubbard, A.; Terry, J.; Rice, J.; Walk, J.; Cziegler, I.; Edlund, E.; Theiler, C.

    2015-11-01

    Placement of X-points and strike points in a diverted tokamak can have a remarkable impact on plasma properties, including thermal and particle confinement. The distinctive divertor of Alcator C-Mod allows substantial variation of divertor leg length, field line attack angle and divertor baffling, allowing us to induce changes in both L-mode confinement and access to both H-mode and I-mode. With the ion ∇B drift directed toward the divertor, scanning the strike point can induce ~ 2 × reductions in H-mode power threshold, and can produce a window for I-mode operation with H98 > 1 . Detailed high-resolution measurements, spanning the last closed flux surface, provide profiles of key quantities (n, T, ϕ) and their gradients, which are of likely importance in determining whether a discharge evolves an edge transport barrier, or remains in an L-mode state. Advances in Langmuir probes have enabled characterization of both radial profiles and fast (power is approached. These data allow new tests of models for H-mode access, especially those attempting to explain the non-monotonic density dependence of the H-mode power threshold through changes in transport and/or turbulence. Supported by U.S. Department of Energy award DE-FC02-99ER54512, using Alcator C-Mod, a DOE Office of Science User Facility.

  18. Parallel Energy Transport in Detached DIII-D Divertor Plasmas

    Science.gov (United States)

    Leonard, A. W.; Lore, J. D.; Canik, J. M.; McLean, A. G.; Makowski, M. A.

    2017-10-01

    A comparison of experiment and modeling of detached divertor plasmas is examined in the context of parallel energy transport. Experimental estimates of power carried by electron thermal conduction versus plasma convection are experimentally inferred from power balance measurements of radiated power and target plate heat flux combined with Thomson scattering measurements of the Te profile along the divertor leg. Experimental profiles of Te exhibit relatively low gradients with Te 3 eV, characteristic of transport dominated by electron conduction through the bulk of the divertor. This discrepancy with experimental transport dominated by convection and modeling by conduction has significant implications for the radiative capacity of divertor plasmas and may explain at least part of the difficulty for fluid modeling to obtain the experimentally observed radiative losses. Comparisons are also made for helium plasmas where the match between experiment and modeling is much better. Work supported by the US DOE under DE-FC02-04ER54698.

  19. Evaluation of helium cooling for fusion divertors

    Energy Technology Data Exchange (ETDEWEB)

    Baxi, C.B.

    1993-09-01

    The divertors of future fusion reactors will have a power throughput of several hundred MW. The peak heat flux on the diverter surface is estimated to be 5 to 15 MW/m{sup 2} at an average heat flux of 2 MW/m{sup 2}. The divertors have a requirement of both minimum temperature (100{degrees}C) and maximum temperature. The minimum temperature is dictated by the requirement to reduce the absorption of plasma, and the maximum temperature is determined by the thermo-mechanical properties of the plasma facing materials. Coolants that have been considered for fusion reactors are water, liquid metals and helium. Helium cooling has been shown to be very attractive from safety and other considerations. Helium is chemically and neutronically inert and is suitable for power conversion. The challenges associated with helium cooling are: (1) Manifold sizes; (2) Pumping power; and (3) Leak prevention. In this paper the first two of the above design issues are addressed. A variety of heat transfer enhancement techniques are considered to demonstrate that the manifold sizes and the pumping power can be reduced to acceptable levels. A helium-cooled diverter module was designed and fabricated by GA for steady-state heat flux of 10 MW/m{sup 2}. This module was recently tested at Sandia National Laboratories. At an inlet pressure of 4 MPa, the module was tested at a steady-state heat flux of 10 MW/m{sup 2}. The pumping power required was less than 1% of the power removed. These results verified the design prediction.

  20. Design of divertor impurity monitoring system for ITER. 2

    Energy Technology Data Exchange (ETDEWEB)

    Sugie, Tatsuo; Ogawa, Hiroaki; Ebisawa, Katsuyuki; Ando, Toshiro; Kasai, Satoshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Katsunuma, Atsushi; Maruo, Mitsumasa; Kita, Yoshio

    1998-11-01

    The divertor impurity monitoring system of ITER has been designed. The main functions of this system are to identify impurity species and to measure the two-dimensional distributions of the particle influxes in the divertor plasmas. The wavelength range is 200 nm to 1000 nm. The viewing fans are realized by molybdenum mirrors located in the divertor cassette. With additional viewing fans seeing through the gap between the divertor cassettes, the region approximately from the divertor leg to the x-point will be observed. The light from the divertor region passes through the quartz windows on the divertor port plug and the cryostat, and goes through the dog-leg optics in the biological shield. Three different type of spectrometers: (i) survey spectrometers for impurity species monitoring, (ii) filter spectrometers for the particle influx measurement with the spatial resolution of 10 mm and the time resolution of 1 ms and (iii) high dispersion spectrometers for high resolution wavelength measurements are designed. These spectrometers are installed just behind the biological shield (for {lambda} < 450 nm) to prevent the transmission loss in fiber and in the diagnostic room (for {lambda} {>=} 450 nm) from the point of view of accessibility and flexibility. The optics have been optimized by a ray trace analysis. As a result, 10-15 mm spatial resolution will be achieved in all regions of the divertor. In addition, the measurable limit, the neutron and {gamma}-ray irradiation effect on windows, a calibration method, an alignment method, a remote handling method and a data acquisition method are considered. (author)

  1. Regimes with recombining plasma in the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Krasheninnikov, S.I. (Massachusetts Inst. of Tech., Cambridge (United States). Plasma Fusion Center Gosudarstvennyj Komitet po Ispol' zovaniyu Atomnoj Ehnergii SSSR, Moscow (Russian Federation). Inst. Atomnoj Ehnergii); Sigmar, D. (Massachusetts Inst. of Tech., Cambridge (United States). Plasma Fusion Center); Soboleva, T.K. (Gosudarstvennyj Komitet po Ispol' zovaniyu Atomnoj Ehnergii SSSR, Moscow (Russian Federation). Inst. Atomnoj Ehnergii); Kukushkin, A.B. (Gosudarstvennyj Komitet po Ispol' zovaniyu Atomnoj Ehnergii SSSR, Moscow (Russian Federation). Inst. Atomnoj Ehnergii); Batischev, O.V. (Kaldysh Inst. for Applied Mathematics, Moscow (Russian Federation)); Sigov, Yu.S. (Kaldysh Inst. for Applied Mathematics, Moscow (Russian Federation))

    1994-01-01

    The possibility to establish regimes with dense recombining hydrogen plasma in ITER divertor is considered. It is shown that due to the large difference between effective heat transmission coefficients of neutral gas and plasma there is a bifurcation of plasma parameters near the target. Due to this bifurcation a neutral gas layer can occur between the divertor plate and the plasma. The criterion of establishing of this gas layer is found. (orig.)

  2. Thermomechanical simulation of WEST actively cooled upper divertor

    Energy Technology Data Exchange (ETDEWEB)

    Batal, T., E-mail: tristan.batal@cea.fr; Richou, M.; Guilhem, D.; Firdaouss, M.; Larroque, S.; Ferlay, F.; Missirlian, M.; Bucalossi, J.

    2016-11-15

    The Tore Supra tokamak is being transformed in an x-point divertor fusion device in the frame of the WEST (W-for tungsten-Environment in Steady-state Tokamak) project, launched in support to the ITER tungsten divertor strategy. The WEST project aims to test ITER-like W monoblock Plasma Facing Units (PFU). This ITER-like divertor will be tested under long plasma discharge up to 1000 s, with high heat flux density up to 20 MW/m{sup 2}. This paper presents the results of ANSYS thermal-structural simulations of the WEST upper divertor. The upper divertor is made of twelve 30° sectors, each one composed of 38 PFU. The PFUs are actively cooled CuCrZr heat sinks and the incidence surface is coated with a thin tungsten layer. The fixing system is made of pins engaged in slotted holes. Besides, the fixing system of the sector assembly is the same as WEST lower divertor, so one upper divertor sector can be used indifferently in upper or Lower position during transitional operation phases in WEST. The total surface of the upper divertor is 8 m{sup 2}, and it has to be able to extract up to 4 MW in steady-state, with peak heat flux values up to 8 MW/m{sup 2}. The fixing system was designed to handle structural loads such as forces and torques resulting from halo and eddy current, respectively, especially during disruptions and Vertical Displacement Event (VDE). The torque resulting from eddy current is first calculated thanks to an internal CEA ANSYS APDL routine. Then the ANSYS structural and thermal-structural simulations of the PFU are presented, and its design is validated thanks to A-level RCC-MRx criteria. Finally, the most conservative load case is determined in order to validate the design of the pins and the support structure.

  3. Impact of carbon and tungsten as divertor materials on the scrape-off layer conditions in JET

    Science.gov (United States)

    Groth, M.; Brezinsek, S.; Belo, P.; Beurskens, M. N. A.; Brix, M.; Clever, M.; Coenen, J. W.; Corrigan, C.; Eich, T.; Flanagan, J.; Guillemaut, C.; Giroud, C.; Harting, D.; Huber, A.; Jachmich, S.; Kruezi, U.; Lawson, K. D.; Lehnen, M.; Lowry, C.; Maggi, C. F.; Marsen, S.; Meigs, A. G.; Pitts, R. A.; Sergienko, G.; Sieglin, B.; Silva, C.; Sirinelli, A.; Stamp, M. F.; van Rooij, G. J.; Wiesen, S.; JET-EFDA Contributors, the

    2013-09-01

    The impact of carbon and beryllium/tungsten as plasma-facing components on plasma radiation, divertor power and particle fluxes, and plasma and neutral conditions in the divertors has been assessed in JET both experimentally and by edge fluid code simulations for plasmas in low-confinement mode. In high-recycling conditions the studies show a 30% reduction in total radiation in the scrape-off (SOL) layer when replacing carbon (JET-C) with beryllium in the main chamber and tungsten in the divertor (JET-ILW). Correspondingly, at the low-field side (LFS) divertor plate a two-fold increase in power conducted to the plate and a two-fold increase in electron temperature at the strike point were measured. In low-recycling conditions the SOL was found to be nearly identical for both materials' configurations. Saturation and rollover of the ion currents to both low- and high-field side (HFS) plates was measured to occur at 30% higher upstream densities and radiated power fraction in JET-ILW. Past saturation, it was possible to reduce the ion currents to the LFS targets by a factor of 2 and to continue operating in stable, detached conditions in JET-ILW; in JET-C the reduction was limited to 50%. These observations are in qualitative agreement with predictions from the fluid edge code package EDGE2D/EIRENE, for which a 30% reduction of the total radiated power is also yielded when switching from C to Be/W. For matching upstream parameters the magnitude of predicted radiation is, however, 50% to 100% lower than measured, independent of the materials' configuration. Inclusion of deuterium molecules and molecular ions, and temperature and density dependent rates in EIRENE reproduced the experimentally observed rollover of the ion current to the LFS plate, via reducing the electron temperature at the plate.

  4. Turbulent Simulations of Divertor Detachment Based On BOUT + + Framework

    Science.gov (United States)

    Chen, Bin; Xu, Xueqiao; Xia, Tianyang; Ye, Minyou

    2015-11-01

    China Fusion Engineering Testing Reactor is under conceptual design, acting as a bridge between ITER and DEMO. The detached divertor operation offers great promise for a reduction of heat flux onto divertor target plates for acceptable erosion. Therefore, a density scan is performed via an increase of D2 gas puffing rates in the range of 0 . 0 ~ 5 . 0 ×1023s-1 by using the B2-Eirene/SOLPS 5.0 code package to study the heat flux control and impurity screening property. As the density increases, it shows a gradually change of the divertor operation status, from low-recycling regime to high-recycling regime and finally to detachment. Significant radiation loss inside the confined plasma in the divertor region during detachment leads to strong parallel density and temperature gradients. Based on the SOLPS simulations, BOUT + + simulations will be presented to investigate the stability and turbulent transport under divertor plasma detachment, particularly the strong parallel gradient driven instabilities and enhanced plasma turbulence to spread heat flux over larger surface areas. The correlation between outer mid-plane and divertor turbulence and the related transport will be analyzed. Prepared by LLNL under Contract DE-AC52-07NA27344. LLNL-ABS-675075.

  5. Divertor research on the DIII-D tokamak

    Science.gov (United States)

    Hill, D. N.; Allen, S. L.; Brooks, N. H.; Buchenauer, D.; Cuthbertson, J. W.; Evans, T. E.; Fenstermacher, M. E.; Ghendrih, Ph.; Hillis, D. L.; Hogan, J. T.

    1994-10-01

    In this paper the authors summarize recent progress on DIII-D in developing techniques for divertor power and particle control relevant to next generation tokamaks such as the proposed ITER and TPX devices. Density control and helium removal by divertor pumping have been demonstrated for the first time in high confinement ELMing H-mode discharges (tau is approximately 2 times tau(sub ITER-89P)) following installation of a divertor cryopumping system. The peak divertor heat flux in similar H-mode discharges has been reduced through production of a radiating mantle with neon or argon puffing (reductions of 3-5). A number of diagnostics have been added to improve the understanding of the physical processes involved. They are now designing modified double-null divertor structures for DIII-D that will provide improved particle control for high-triangularity VH-mode plasmas while at the same time allowing for gas puffing to reduce the divertor heat flux.

  6. Fabrication of divertor mock-up with ODS-Cu and W by the improved brazing technique

    Science.gov (United States)

    Tokitani, M.; Hamaji, Y.; Hiraoka, Y.; Masuzaki, S.; Tamura, H.; Noto, H.; Tanaka, T.; Muroga, T.; Sagara, A.; FFHR Design Group

    2017-07-01

    Copper alloy has been considered as a divertor cooling tube or heat sink not only in the helical reactor FFHR-d1 but also in the tokamak DEMO reactor, because it has a high thermal conductivity. This work focused on applying an oxide dispersion strengthened copper alloy (ODS-Cu), GlidCop® (Cu-0.3 wt%Al2O3) as the divertor heat sink material of FFHR-d1. This alloy has superior high temperature yield strength exceeding 300 MPa at room temperature even after annealing up to ~1000 °C. The change in material properties of Pure-Cu, GlidCop® and CuCrZr by neutron irradiation are summarized in this paper. A primary dose limit is the radiation-induced hardening/softening (~0.2 dpa/1-2 dpa) which has a temperature dependence. According to such an evaluation, the GlidCop® can be selected as the current best candidate material in the commercial base of the divertor heat sink, and its temperature should be maintained as close as possible to 300 °C during operation. Bonding between the W armour and the GlidCop® heat sink was successfully performed by using an improved brazing technique with BNi-6 (Ni-11%P) filler material. The bonding strength was measured by a three-point bending test and reached up to approximately 200 MPa. Surprisingly, several specimens showed an obvious yield point. This means that the BNi-6 brazing (bonding) layer caused relaxation of the applied stress. The small-scale divertor mock-up of the W/BNi-6/GlidCop® was successfully fabricated by using the improved brazing technique. The heat loading test was carried out by the electron beam device ACT2 in NIFS. The mock-up showed an excellent heat removal capability for use in the FFHR-d1 divertor.

  7. Manufacturing W fibre-reinforced Cu composite pipes for application as heat sink in divertor targets of future nuclear fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, Alexander v.; You, Jeong-Ha [Max-Planck-Institut fuer Plasmaphysik, 85748 Garching (Germany); Ewert, Dagmar [Institut fuer Textil- und Verfahrenstechnik Denkendorf, 73770 Denkendorf (Germany); Siefken, Udo [Louis Renner GmbH, 85221 Dachau (Germany)

    2016-07-01

    An important plasma-facing component (PFC) in future nuclear fusion reactors is the so-called divertor which allows power exhaust and removal of impurities from the main plasma. The most highly loaded parts of a divertor are the target plates which have to withstand intense particle bombardment. This intense particle bombardment leads to high heat fluxes onto the target plates which in turn lead to severe thermomechanical loads. With regard to future nuclear fusion reactors, an improvement of the performance of divertor targets is desirable in order to ensure reliable long term operation of such PFCs. The performance of a divertor target is most closely linked to the properties of the materials that are used for its design. W fibre-reinforced Cu (Wf/Cu) composites are regarded as promising heat sink materials in this respect. These materials do not only feature adequate thermophysical and mechanical properties, they do also offer metallurgical flexibility as their microstructure and hence their macroscopic properties can be tailored. The contribution will point out how Wf/Cu composites can be used to realise an advanced design of a divertor target and how these materials can be fabricated by means of liquid Cu infiltration.

  8. Simulation study of detached plasmas by using one-dimensional SOL-divertor fluid code with virtual divertor model

    Energy Technology Data Exchange (ETDEWEB)

    Togo, S.; Lang, T.L.; Ogawa, Y. [Graduate School of Frontier Sciences, University of Tokyo, Kashiwa (Japan); Takizuka, T.; Ibano, K. [Graduate School of Engineering, Osaka University, Suita (Japan); Nakamura, M.; Hoshino, K. [Japan Atomic Energy Agency, Rokkasho, Aomori (Japan)

    2016-08-15

    The detached plasmas due to the volume recombination are studied by using one-dimensional (1D) scrape-off-layer-divertor (SOL-DIV) plasma fluid code with virtual divertor (VD) model. By introducing the anisotropic ion temperature, the parallel momentum transport equation becomes the first-order differential and the Mach number at the sheath entrance is determined self-consistently by the upstream condition. The total particle flux at the divertor plates and the flux amplification factors are shown as functions of the plasma density at the stagnation point and the dependence of these parameters on the heat flux from the core plasma, radial width of the flux tube in the divertor region and the strength of the impurity radiation is investigated. (copyright 2015 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  9. Verification test for helium panel of cryopump for DIII-D advanced divertor

    Energy Technology Data Exchange (ETDEWEB)

    Baxi, C.B.; Laughon, G.J.; Langhorn, A.R.; Schaubel, K.M.; Smith, J.P.; Gootgeld, A.M.; Campbell, G.L. (General Atomics, San Diego, CA (United States)); Menon, M.M. (Oak Ridge National Lab., TN (United States))

    1991-10-01

    It is planned to install a cryogenic pump in the lower divertor portion of the D3-D tokamak with a pumping speed of 50000{ell}/s and an exhaust of 2670 Pa-{ell}/s (20 Torr-{ell}s). A coaxial counter flow configuration has been chosen for the helium panel of this cryogenic pump. This paper evaluates cooldown rates and fluid stability of this configuration. A prototypic test was performed at General Atomics (GA) to increase confidence in the design. It was concluded that the helium panel cooldown rate agreed quite well with analytical prediction and was within acceptable limits. The design flow rate proved stable and two-phase pressure drop can be predicted quite accurately. 8 refs., 5 figs., 1 tab.

  10. Analytic 1D Approximation of the Divertor Broadening S in the Divertor Region for Conductive Heat Transport

    CERN Document Server

    Nille, Dirk; Eich, Thomas

    2016-01-01

    Topic is the divertor broadening $S$, being a result of perpendicular transport in the scrape-off layer and resulting in a better distribution of the power load onto the divertor target. Recent studies show a scaling of the divertor broadening with an inverse power law to the target temperature $T_t$, promising its reduction to be a way of distributing the power entering the divertor volume onto a large surface area. It is shown that for pure conductive transport in the divertor region the suggested inverse power law scaling to $T_t$ is only valid for high target electron temperatures. For decreasing target temperatures ($T_t < 20\\,$eV) the increase of $S$ stagnates and the conductive model results in a finite value of $S$ even for zero target temperature. It is concluded that the target temperature is no valid parameter for a power law scaling, as it is not representative for the entire divertor volume. This is shown in simulations solving the 2D heat diffusion equation, which is used as reference for an ...

  11. 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.

  12. Plasma density control with ergodic divertor on Tore Supra; Controle de la densite du plasma en presence du divertor ergodique dans le tokamak Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Meslin, B

    1998-04-30

    Plasma density control on the tokamak Tore Supra is important for the optimization of every experimental scenario dealing with the improvement of plasma performances. Specific conditions are required both in the plasma bulk and at the edge. Within the framework of the present study, a magnetic configuration is used in the e plasma edge of Tore Supra: the ergodic divertor configuration. A magnetic perturbation which is resonant with the permanent field destroys the plasma confinement locally, opening the field lines onto the material components. They aim of the study is the characterization of the edge density in every relevant scenario for Tore Supra. The first part of this work is dedicated to density and temperature measurements by a series of fixed Langmuir probes located at the very edge of the plasma. Thanks to them, density regimes have been put in evidence during experiments where the volume averaged density , an usual control parameter of the plasma, was varied. The analysis of heat and particle transport through the plasma edge region explains the mechanisms leading to those regimes. The essential factor in our analysis is the dependence of the electron conductivity and ionization depth on temperature. While heat conduction governs the heat transport, the edge density varies linearly according to . Below a critical temperature, reached when the ion flux amplification at constant power density is large enough, a parallel temperature gradient appears leading to a density gradient in the opposite direction in order to maintain the pressure constant along the field lines. A high recycling regime is obtained and the edge density varies like {sup 3}. The pressure conservation is no more satisfied during the detachment of the plasma, which is characterized by a high neutral density at low temperatures leading to a ion momentum loss by friction against the neutrals. The edge density drops in those conditions. These regimes are similar

  13. Comparison of detached and radiative divertor operation in Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Goetz, J.A.; Kurz, C.; LaBombard, B.; Lipschultz, B.; Niemczewski, A.; McCracken, G.M.; Terry, J.L.; Boivin, R.L.; Bombarda, F.; Bonoli, P.; Fiore, C.; Golovato, S.; Granetz, R.; Greenwald, M.; Horne, S.; Hubbard, A.; Hutchinson, I.; Irby, J.; Marmar, E.; Porkolab, M.; Rice, J.; Snipes, J.; Takase, Y.; Watterson, R.; Welch, B.; Wolfe, S.; Christensen, C.; Garnier, D.; Jablonski, D.; Lo, D.; Lumma, D.; May, M.; Mazurenko, A.; Nachtrieb, R.; OShea, P.; Reardon, J.; Rost, J.; Schachter, J.; Sorci, J.; Stek, P.; Umansky, M.; Wang, Y. [Plasma Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States)

    1996-05-01

    The divertor of the Alcator C-Mod tokamak [Phys. Plasmas {bold 1}, 1511 (1994)] routinely radiates a large fraction of the power entering the scrape-off layer. This dissipative divertor operation occurs whether the divertor is detached or not, and large volumetric radiative emissivities, up to 60 MWm{sup {minus}3} in ion cyclotron range of frequency (ICRF) heated discharges, have been measured using bolometer arrays. An analysis of both Ohmic and ICRF-heated discharges has demonstrated some of the relative merits of detached divertor operation versus high-recycling divertor operation. An advantage of detached divertor operation is that the power flux to the divertor plates is decreased even further than its already low value. Some disadvantages are that volumetric losses outside the separatrix in the divertor region are decreased, the neutral compression ratio is decreased, and the penetration efficiency of impurities increases. {copyright} {ital 1996 American Institute of Physics.}

  14. A practical globalization of one-shot optimization for optimal design of tokamak divertors

    Energy Technology Data Exchange (ETDEWEB)

    Blommaert, Maarten, E-mail: maarten.blommaert@kuleuven.be [Institute of Energy and Climate Research (IEK-4), FZ Jülich GmbH, D-52425 Jülich (Germany); Dekeyser, Wouter; Baelmans, Martine [KU Leuven, Department of Mechanical Engineering, 3001 Leuven (Belgium); Gauger, Nicolas R. [TU Kaiserslautern, Chair for Scientific Computing, 67663 Kaiserslautern (Germany); Reiter, Detlev [Institute of Energy and Climate Research (IEK-4), FZ Jülich GmbH, D-52425 Jülich (Germany)

    2017-01-01

    In past studies, nested optimization methods were successfully applied to design of the magnetic divertor configuration in nuclear fusion reactors. In this paper, so-called one-shot optimization methods are pursued. Due to convergence issues, a globalization strategy for the one-shot solver is sought. Whereas Griewank introduced a globalization strategy using a doubly augmented Lagrangian function that includes primal and adjoint residuals, its practical usability is limited by the necessity of second order derivatives and expensive line search iterations. In this paper, a practical alternative is offered that avoids these drawbacks by using a regular augmented Lagrangian merit function that penalizes only state residuals. Additionally, robust rank-two Hessian estimation is achieved by adaptation of Powell's damped BFGS update rule. The application of the novel one-shot approach to magnetic divertor design is considered in detail. For this purpose, the approach is adapted to be complementary with practical in parts adjoint sensitivities. Using the globalization strategy, stable convergence of the one-shot approach is achieved.

  15. 3D Numerical Analysis of Radiative Edge Cooling in Wendelstein 7-X Island Divertor Scenarios

    Science.gov (United States)

    Effenberg, Florian; Feng, Y.; Frerichs, H.; Schmitz, O.; Barbui, T.; Geiger, J.; Jakubowski, M.; Köenig, R.; Krychowiak, M.; Niemann, H.; Sunn Pedersen, T.; Suzuki, Y.; Wurden, G. A.; W7-X-Team Team

    2017-10-01

    Radiative edge cooling is a promising method for mitigation of high heat and particle fluxes in the 3D field geometry of Wendelstein 7-X. A new high mirror island configuration is investigated featuring a more uniform distribution of heat and particle fluxes on horizontal and vertical divertor targets. For an upstream density of nup = 2 × 1019m-3 at PECRH=8MW maximum heat loads up to qmax 7.2MWm-2 are calculated with the 3D fluid and kinetic edge transport Monte Carlo Code EMC3-EIRENE. Carbon eroded from the divertor targets is predicted to serve as effective intrinsic radiator enabling detached operational regimes at higher densities (nup > 4 × 1019m-3). The feasibility of active control of heat and particle flux levels by impurity seeding (CxHy, N2, Ne) will be discussed for the new island geometry. Impurity line radiation tends to concentrate in the islands for lower densities and causes a drop of flux levels correlated to the power loss fraction, Δq Prad/PSOL . β-effects are taken into account based on the 3D MHD-equilibrium code HINT. This work was supported by the U.S. Department of Energy (DOE) under Grant DE-SC0014210.

  16. 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.

  17. Non-resonant divertors for stellarators

    Science.gov (United States)

    Boozer, Allen; Punjabi, Alkesh

    2017-10-01

    The outermost confining magnetic surface in optimized stellarators has sharp edges, which resemble tokamak X-points. The plasma cross section has an even number of edges at the beginning but an odd number half way through the period. Magnetic field lines cannot cross sharp edges, but stellarator edges have a finite length and do not determine the rotational transform on the outermost confining surface. Just outside the last confining surface, surfaces formed by magnetic field lines have splits containing two adjacent magnetic flux tubes: one with entering and the other with an equal existing flux to the walls. The splits become wider with distance outside the outermost confining surface. These flux tubes form natural non-resonant stellarator divertors, which we are studying using maps. This work is supported by the US DOE Grants DE-FG02-95ER54333 to Columbia University and DE-FG02-01ER54624 and DE-FG02-04ER54793 to Hampton University and used resources of the NERSC, supported by the Office of Science, US DOE, under Contract No. DE-AC02-.

  18. Toroidal asymmetries in divertor impurity influxes in NSTX

    Directory of Open Access Journals (Sweden)

    F. Scotti

    2017-08-01

    Full Text Available Toroidal asymmetries in divertor carbon and lithium influxes were observed in NSTX, due to toroidal differences in surface composition, tile leading edges, externally-applied three-dimensional (3D fields and toroidally-localized edge plasma modifications due to radio frequency heating. Understanding toroidal asymmetries in impurity influxes is critical for the evaluation of total impurity sources, often inferred from measurements with a limited toroidal coverage. The toroidally-asymmetric lithium deposition induced asymmetries in divertor lithium influxes. Enhanced impurity influxes at the leading edge of divertor tiles were the main cause of carbon toroidal asymmetries and were enhanced during edge localized modes. Externally-applied 3D fields led to strike point splitting and helical lobes observed in divertor impurity emission, but marginal changes to the toroidally-averaged impurity influxes. Power coupled to the scrape-off layer SOL plasma during radio frequency (RF heating of H-mode discharges enhanced impurity influxes along the non-axisymmetric divertor footprint of flux tubes connecting to plasma in front of the RF antenna.

  19. Numerical study of the ITER divertor plasma with the B2-EIRENE code package

    Energy Technology Data Exchange (ETDEWEB)

    Kotov, V.; Reiter, D. [Forschungszentrum Juelich (DE). Inst. fuer Energieforschung (IEF), Plasmaphysik (IEF-4); Kukushkin, A.S. [ITER International Team, Cadarache (France)

    2007-11-15

    The problem of plasma-wall interaction and impurity control is one of the remaining critical issues for development of an industrial energy source based on nuclear fusion of light isotopes. In this field sophisticated integrated numerical tools are widely used both for the analysis of current experiments and for predictions guiding future device design. The present work is dedicated to the numerical modelling of the edge plasma region in divertor configurations of large-scale tokamak fusion devices. A well established software tool for this kind of modelling is the B2-EIRENE code. It was originally developed for a relatively hot (>> 10 eV) ''high recycling divertor''. It did not take into account a number of physical effects which can be potentially important for ''detached conditions'' (cold, - several eV, - high density, - {approx} 10{sup 21} m{sup -3}, - plasma) typical for large tokamak devices. This is especially critical for the modelling of the divertor plasma of ITER: an international project of an experimental tokamak fusion reactor to be built in Cadarache, France by 2016. This present work is devoted to a major upgrade of the B2-EIRENE package, which is routinely used for ITER modelling, essentially with a significantly revised version of EIRENE: the Monte-Carlo neutral transport code. The main part of the thesis address three major groups of the new physical effects which have been added to the model in frame of this work: the neutral-neutral collisions, the up-to date hydrogen molecular reaction kinetics and the line radiation transport. The impact of the each stage of the upgrade on the self-consistent (between plasma, the neutral gas and the radiation field) solution for the reference ITER case is analysed. The strongest effect is found to be due to the revised molecular collision kinetics, in particular due to hitherto neglected elastic collisions of hydrogen molecules with ions. The newly added non

  20. Aberrations in preliminary design of ITER divertor impurity influx monitor

    Energy Technology Data Exchange (ETDEWEB)

    Kitazawa, Sin-iti, E-mail: kitazawa.siniti@jaea.go.jp [Naka Fusion Institute, Japan Atomic Energy Agency, JAEA, Naka 311-0193 (Japan); Ogawa, Hiroaki [Naka Fusion Institute, Japan Atomic Energy Agency, JAEA, Naka 311-0193 (Japan); Katsunuma, Atsushi; Kitazawa, Daisuke [Core Technology Center, Nikon Corporation, Yokohama 244-8533 (Japan); Ohmori, Keisuke [Customized Products Business Unit, Nikon Corporation, Mito 310-0843 (Japan)

    2015-12-15

    Highlights: • Divertor impurity influx monitor for ITER (DIM) is procured by JADA. • DIM is designed to observe light from nuclear fusion plasma directly. • DIM is under preliminary design phase. • The spot diagrams were suppressed within the core of receiving fiber. • The aberration of DIM is suppressed in the preliminary design. - Abstract: Divertor impurity influx monitor for ITER (DIM) is a diagnostic system that observes light from nuclear fusion plasma directly. This system is affected by various aberrations because it observes light from the fan-array chord near the divertor in the ultraviolet–near infrared wavelength range. The aberrations should be suppressed to the extent possible to observe the light with very high spatial resolution. In the preliminary design of DIM, spot diagrams were suppressed within the core of the receiving fiber's cross section, and the resulting spatial resolutions satisfied the design requirements.

  1. The simple map for a single-null divertor tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Punjabi, A.; Verma, A.; Boozer, A. [Hampton Univ. (Vatican City State, Holy See). Center for Fusion Research and Training

    1996-12-01

    We present the simple map for a single-null divertor tokamak. The simple map is an area-preserving map based on the idea that magnetic field lines are a single-degree-of-freedom time-dependent Hamiltonian system, and that the basic features of such systems near the X-point are generic. We obtain the properties of this map and the resulting footprints of field lines on the divertor plate. These include the width of the stochastic layer, the edge safety factor, the area of the footprint and the amount of magnetic flux diverted. We give the safety factor profile, the average and median values of strike angles, lengths and the Liapunov exponents. We describe how the effects of magnetic perturbations can be included in the simple map. We show how the map can be applied to the problem of the determination of heat flux on the divertor plate in tokamaks. (Author).

  2. Engineering design of a toroidal divertor for the EBT-S fusion device. Final report, Phase II. EBT-S divertor project

    Energy Technology Data Exchange (ETDEWEB)

    Mai, L.P.; Malick, F.S.

    1981-07-01

    The mechanical, structural, thermal, electrical, and vacuum design of a magnetic toroidal divertor system for the Elmo Bumpy Torus (EBT-S) is presented. The EBT-S is a toroidal magnetic fusion device located at the ORNL that operates under steady state conditions. The engineering of the divertor was performed during the second of three phases of a program aimed at the selection, design, fabrication, and installation of a magnetic divertor for EBT-S. The magnetic analysis of the toroidal divertor was performed during Phase I of the program and has been reported in a separate document. In addition to the details of the divertor design, the modest modifications that are required to the EBT-S device and facility to accommodate the divertor system are presented.

  3. Thermal Analysis of the Divertor Primary Heat Transfer System Piping During the Gas Baking Process

    Energy Technology Data Exchange (ETDEWEB)

    Yoder Jr, Graydon L [ORNL; Harvey, Karen [ORNL; Ferrada, Juan J [ORNL

    2011-02-01

    A preliminary analysis has been performed examining the temperature distribution in the Divertor Primary Heat Transfer System (PHTS) piping and the divertor itself during the gas baking process. During gas baking, it is required that the divertor reach a temperature of 350 C. Thermal losses in the piping and from the divertor itself require that the gas supply temperature be maintained above that temperature in order to ensure that all of the divertor components reach the required temperature. The analysis described in this report was conducted in order to estimate the required supply temperature from the gas heater.

  4. Characterization of magnetohydrodynamic transport in a Field Reversed Configuration

    Science.gov (United States)

    Onofri, Marco; Yushmanov, Peter; Dettrick, Sean; Barnes, Daniel; Hubbard, Kevin; Tajima, Toshi; TAE Team

    2017-10-01

    Transport in a Field Reversed Configuration (FRC) is studied by using the two-dimensional code Q2D, which couples a magnetohydrodynamic code with a Monte Carlo code for the beam component. The simulation by Q2D of the parallel transport in the simple open θ-pinch fields and its associated outflow shows an excellent agreement with one of the existing theories, providing a benchmark for Q2D and simultaneously deepening the theoretical understanding of this fundamental process. We find a sharp distinction between the evolved radial density profiles of the FRC and mirror plasmas as a result of the transport processes, showing that the closed flux surfaces of an FRC enhance the confinement over that of a mirror. We characterize the scrape-off layer (SOL) transport by including the mirror trapping effects and we find a relation between the confinement time in the SOL and the ion collisional time. The Q2D code is also used to study the formation of the electrostatic potential in the divertor.

  5. Development of Spatial Heterodyne Spectroscopy Measurements for the C-2W Plasma Expansion Divertor

    Science.gov (United States)

    Sheftman, Daniel; Matsumoto, Tadafumi; Thompson, Matthew; Tri Alpha Energy Team

    2017-10-01

    Accurate operation and high performance of the open field line plasma surrounding the Field Reversed Configuration (FRC) is crucial to achieving the goals of successful temperature ramp up and confinement improvement on C-2W. Attributes such as the outflow velocity and temperature of charge exchange or impurity ions can be measured through spectroscopic methods. However, light throughput is severely limited due to the low plasma density inside the divertors where the plasma expands rapidly before terminating on biasing plates. A field widened spatial heterodyne spectrometer was developed in order to address the challenge of making accurate spectroscope measurements on the diffuse plasma. Design of a prototype of this spectrometer, including lab calibration and spectral line measurements performed on a compact toroid injector test stand, will be presented.

  6. Phosphate Tether-Mediated Ring-Closing Metathesis for the Generation of P-Stereogenic, Z-Configured Bicyclo[7.3.1]- and Bicyclo[8.3.1]phosphates.

    Science.gov (United States)

    Markley, Jana L; Maitra, Soma; Hanson, Paul R

    2016-02-05

    A phosphate tether-mediated ring-closing metathesis (RCM) study to the synthesis of Z-configured, P-stereogenic bicyclo[7.3.1]- and bicyclo[8.3.1]phosphates is reported. Investigations suggest that C3-substitution, olefin substitution, and proximity of the forming olefin to the bridgehead carbon of the bicyclic affect the efficiency and stereochemical outcome of the RCM event. This study demonstrates the utility of phosphate tether-mediated desymmetrization of C2-symmetric, 1,3-anti-diol-containing dienes in the generation of macrocyclic phosphates with potential synthetic and biological utility.

  7. Maximizing Heat Dissipation via Target Optimization of the Small-Angle Slot Divertor

    Science.gov (United States)

    Covele, Brent; Halpern, Federico; Casali, Livia; Canik, John; Thomas, Dan; Guo, Houyang

    2017-10-01

    The planned SAS 2 divertor uses a combination of grazing target angles and closure to direct recycling neutrals near the strike point, thus facilitating detachment onset. SAS 2 should also provide adequate pumping efficiency to be consistent with high-power steady-state scenarios on DIII-D. Initial SOLPS results indicate significantly higher neutral densities and lower electron temperatures in the SAS 2 slot, compared to a closed reference divertor model with baseline plasma profiles appropriate for high power. A systematic optimization of the parameterized SAS 2 target shape is performed in SOLPS to further reduce target heat fluxes and temperatures at lowest upstream density. To speed up the target optimization process, target neutral densities calculated by Eirene act as a performance metric by proxy for detachment facilitation. The efficacy of this proxy metric is discussed. Results are also presented from SAS 2 neutral pumping simulations in Eirene with a stationary background plasma. The feasibility of mutually satisfactory particle control and detachment control is discussed. Work supported under USDOE Cooperative Agreements DE-FC02-04ER54698 and DE-AC05-00OR22725.

  8. Mechanical Design of the NSTX Liquid Lithium Divertor

    Energy Technology Data Exchange (ETDEWEB)

    R. Ellis, R. Kaita, H. Kugel, G. Paluzzi, M. Viola and R. Nygren

    2009-02-19

    The Liquid Lithium Divertor (LLD) on NSTX will be the first test of a fully-toroidal liquid lithium divertor in a high-power magnetic confinement device. It will replace part of the lower outboard divertor between a specified inside and outside radius, and ultimately provide a lithium surface exposed to the plasma with enough depth to absorb a significant particle flux. There are numerous technical challenges involved in the design. The lithium layer must be as thin as possible, and maintained at a temperature between 200 and 400 degrees Celsius to minimize lithium evaporation. This requirement leads to the use of a thick copper substrate, with a thin stainless steel layer bonded to the plasma-facing surface. A porous molybdenum layer is then plasma-sprayed onto the stainless steel, to provide a coating that facilitates full wetting of the surface by the liquid lithium. Other challenges include the design of a robust, vacuumcompatible heating and cooling system for the LLD. Replacement graphite tiles that provided the proper interface between the existing outer divertor and the LLD also had to be designed, as well as accommodation for special LLD diagnostics. This paper describes the mechanical design of the LLD, and presents analyses showing the performance limits of the LLD.

  9. High confinement dissipative divertor operation on Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Goetz, J.A.; LaBombard, B.; Lipschultz, B.; Pitcher, C.S.; Terry, J.L.; Boswell, C.; Gangadhara, S.; Pappas, D.; Weaver, J.; Welch, B.; Boivin, R.L.; Bonoli, P.; Fiore, C.; Granetz, R.; Greenwald, M.; Hubbard, A.; Hutchinson, I.; Irby, J.; Marmar, E.; Mossessian, D.; Porkolab, M.; Rice, J.; Rowan, W.L.; Schilling, G.; Snipes, J.; Takase, Y.; Wolfe, S.; Wukitch, S. [Plasma Science Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States)

    1999-05-01

    Alcator C-Mod [I. H. Hutchinson {ital et al.}, Phys. Plasmas {bold 1}, 1511 (1994)] has operated a High-confinement-mode (H-mode) plasma together with a dissipative divertor and low core Z{sub eff}. The initially attached plasma is characterized by steady-state enhancement factor, H{sub ITER89P} [P. N. Yushmanov {ital et al.}, Nucl. Fusion {bold 30}, 1999 (1990)], of 1.9, central Z{sub eff} of 1.1, and a radiative fraction of {approximately}50{percent}. Feedback control of a nitrogen gas puff is used to increase radiative losses in both the core/edge and divertor plasmas in almost equal amounts. Simultaneously, the core plasma maintains H{sub ITER89P} of 1.6 and Z{sub eff} of 1.4 in this nearly 100{percent} radiative state. The power and particle flux to the divertor plates have been reduced to very low levels while the core plasma is relatively unchanged by the dissipative nature of the divertor. {copyright} {ital 1999 American Institute of Physics.}

  10. Spectroscopic study of JT-60U divertor plasma

    Energy Technology Data Exchange (ETDEWEB)

    Kubo, Hirotaka; Higashijima, Satoshi; Takenaga, Hidenobu; Shimizu, Katsuhiro; Sugie, Tatsuo; Suzuki, S.; Sakasai, Akira; Asakura, Nobuyuki [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Kumagai, A. [Plasma Research Center, Tsukuba Univ., Ibaraki (Japan)

    2000-01-01

    Particle behavior in the JT-60U divertor plasmas has been studied spectroscopically. Doppler profiles of the D{alpha} line have been investigated for understanding of atomic and molecular processes in deuterium particle recycling and D{alpha} line emission. Near the divertor plates, dissociative excitation from deuterium molecules and molecular ions plays an important role for the line emission. By investigation of spectral profiles of the He I line (667.8 nm), Doppler broadening due to elastic scattering by protons has been found. It is estimated that the penetration probability of the helium atoms from the divertor plates to the main plasma and the helium atom flux to the gap for pumping increase by 30% due to the elastic scattering. Intensity distribution of the CD band (around 430.5 nm) has been compared between the W-shaped divertor with a dome in the private flux region and the previous open one. The dome prevents the upstream transport of hydrocarbon impurity produced by chemical sputtering. (author)

  11. Edge and divertor physics with reversed toroidal field in JET

    Energy Technology Data Exchange (ETDEWEB)

    Pitts, R.A. [Ecole Polytechnique Federale, Association Euratom-Confederation Suisse, Lausanne (Switzerland). Centre de Recherche en Physique des Plasma (CRPP); Andrew, P.; Corrigan, G.; Erents, S.K.; Fundamenski, W.; Lomas, P.J.; Matthews, G.F.; Stamp, M.F. [Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon, OX (United Kingdom); Bonnin, X.; Corre, Y.; Tsitrone, E. [Association Euratom-CEA, Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Chankin, A.V.; Coster, D.; Eich, T. [Max-Planck-Institut fuer Plasmaphysik, Euratom-Association, Garching (Germany); Duran, I. [Institute of Plasma Physics, Association Euratom-IPP.CR, Prague (Czech Republic); Huber, A.; Lehnen, M.; Rapp, J. [FZJ Julich GmbH/Euratom Institut fur Plasmaphysik, TEC, Julich D (Germany); Jachmich, S. [Association Euratom-Belgian State, LPP, ERM/KMS (Belgium); Kirnev, G. [Moscow Nuclear Fusion Institute, RRC Kurchatov Institute, Moscow (Russian Federation); Loarte, A. [Max-Planck-Institut fur Plasmaphysik, EFDA-CSU, Garching (Germany); Silva, C. [Association Euratom-IST, Lisbon (Portugal); Strachan, J.D. [Princeton Univ., NJ (United States). Plasma Physics Lab

    2004-07-01

    Results from the most recent reversed field campaign at JET in combination with numerical modelling are providing some valuable insights into the pattern of scrape-off layer (SOL) flows and divertor energy and particle asymmetries. This has been made possible by comparing carefully matched discharges in both field directions. Earlier measurements of strong parallel flow at the top of the machine from outer to inner divertor in normal field operation have been confirmed and improved upon. New data in reversed field show an almost stagnant flow throughout most of the SOL except near the separatrix. The forward field flow is almost an order of magnitude larger than be accounted for by EDGE2D code simulations including all classical drifts. Likewise, the model does not reproduce the flow offset (M{sub ||} {approx} 0.2) from outer to inner target seen experimentally for both field directions. A number of avenues are being pursued to increase the predicted EDGE2D forward field flow - the inclusion of anomalous convective pinch terms, ballooning like diffusive particle transport and the perturbing effect of the probe. Divertor energy asymmetries are observed to be strongly dependent on the sign of toroidal field but not its magnitude. This finding is a direct consequence of radial energy transport which is independent of field direction and which scales inversely with B{sub {phi}}. It is strong evidence for drift effects being the main driver for the observed change in in/out asymmetry with field reversal. Divertor tile temperature measurements using infra-red thermography have revealed the build-up of a thermally resistant surface layer on the outer target during reversed field operation, implying that the outer divertor switches from a region of net erosion (the case in forward field) to net redeposition. This new observation is not inconsistent with the rearrangement of the poloidal distribution of parallel SOL flow seen when the field is reversed in EDGE2D simulations

  12. 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.

  13. 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.

  14. Realization of high heat flux tungsten monoblock type target with graded interlayer for application to DEMO divertor

    Science.gov (United States)

    Richou, M.; Gallay, F.; Böswirth, B.; Chu, I.; Lenci, M.; Loewenhoff, Th; Quet, A.; Greuner, H.; Kermouche, G.; Meillot, E.; Pintsuk, G.; Visca, E.; You, J. H.

    2017-12-01

    The divertor is the key in-vessel plasma-facing component being in charge of power exhaust and removal of impurity particles. In DEMO, divertor targets must survive an environment of high heat fluxes (˜up to 20 MW m-2 during slow transients) and neutron irradiation. One advanced concept for components in monoblock configuration concerns the insertion of a compositionally graded layer between tungsten and CuCrZr instead of the soft copper interlayer. As a first step, a thin graded layer (˜25 μm) was developed. As a second step, a thicker graded layer (˜500 μm), which is actually being developed, will also be inserted to study the compliant role of a macroscopic graded layer. This paper reports the results of cyclic high heat flux loading tests up to 20 MW m-2 and to heat flux higher than 25 MW m-2 that mock-ups equipped with thin graded layer survived without visible damage. First feedback on manufacturing steps is also presented. Moreover, the first results obtained on the development of the thick graded layer and its integration in a monoblock configuration are shown.

  15. Neutron diffraction stress determination in W-laminates for structural divertor applications

    Directory of Open Access Journals (Sweden)

    R. Coppola

    2015-07-01

    Full Text Available Neutron diffraction measurements have been carried out to develop a non-destructive experimental tool for characterizing the crystallographic structure and the internal stress field in W foil laminates for structural divertor applications in future fusion reactors. The model sample selected for this study had been prepared by brazing, at 1085 °C, 13 W foils with 12 Cu foils. A complete strain distribution measurement through the brazed multilayered specimen and determination of the corresponding stresses has been obtained, assuming zero stress in the through-thickness direction. The average stress determined from the technique across the specimen (over both ‘phases’ of W and Cu is close to zero at −17 ± 32 MPa, in accordance with the expectations.

  16. Examination of high heat flux components for the ITER divertor after thermal fatigue testing

    Energy Technology Data Exchange (ETDEWEB)

    Missirlian, M., E-mail: marc.missirlian@cea.fr [CEA, IRFM, F-13108 Saint Paul lez Durance (France); Escourbiac, F., E-mail: frederic.escourbiac@cea.fr [CEA, IRFM, F-13108 Saint Paul lez Durance (France); Schmidt, A., E-mail: a.schmidt@fz-juelich.de [Forschungszentrum Juelich, IFE-2 (Germany); Riccardi, B., E-mail: Bruno.Riccardi@f4e.europa.eu [Fusion For Energy, E-08019 Barcelona (Spain); Bobin-Vastra, I., E-mail: isabelle.bobinvastra@areva.com [AREVA-NP, 71200 Le Creusot (France)

    2011-10-01

    An extensive development programme has been carried out in the EU on high heat flux components within the ITER project. In this framework, a full-scale vertical target (VTFS) prototype was manufactured with all the main features of the corresponding ITER divertor design. The fatigue cycling campaign on CFC and W armoured regions, proved the capability of such a component to meet the ITER requirements in terms of heat flux performances for the vertical target. This paper discusses metallographic observations performed on both CFC and W part after this intensive thermal fatigue testing campaign for a better understanding of thermally induced mechanical stress within the component, especially close to the armour-heat sink interface.

  17. Divertor heat and particle control experiments on the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Mahdavi, M.A; Baker, D.R. [General Atomics, San Diego, CA (United States); Allen, S.L. [Lawrence Livermore National Lab., CA (United States)] [and others

    1994-05-01

    In this paper we present a summary of recent DIII-D divertor physics activity and plans for future divertor upgrades. During the past year, DIII-D experimental effort was focused on areas of active heat and particle control and divertor target erosion studies. Using the DIII-D Advanced Divertor system we have succeeded for the first time to control the plasma density and demonstrate helium exhaust in H-mode plasmas. Divertor heat flux control by means of D{sub 2} gas puffing and impurity injection were studied separately and in, both cases up to a factor of five reduction of the divertor peak heat flux was observed. Using the DiMES sample transfer system we have obtained erosion data on various material samples in well diagnosed plasmas and compared the results with predictions of numerical models.

  18. The control of convection by fuelling and pumping in the JET pumped divertor

    Energy Technology Data Exchange (ETDEWEB)

    Harbour, P.J.; Andrew, P.; Campbell, D.; Clement, S.; Davies, S.; Ehrenberg, J.; Erents, S.K.; Gondhalekar, A.; Gadeberg, M.; Gottardi, N.; Von Hellermann, M.; Horton, L.; Loarte, A.; Lowry, C.; Maggi, C.; McCormick, K.; O`Brien, D.; Reichle, R.; Saibene, G.; Simonini, R.; Spence, J.; Stamp, M.; Stork, D.; Taroni, A.; Vlases, G. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking

    1994-07-01

    Convection from the scrape-off layer (SOL) to the divertor will control core impurities, if it retains them in a cold, dense, divertor plasma. This implies a high impurity concentration in the divertor, low at its entrance. Particle flux into the divertor entrance can be varied systematically in JET, using the new fuelling and pumping systems. The convection ratio has been estimated for various conditions of operation. Particle convection into the divertor should increase thermal convection, decreasing thermal conduction, and temperature and density gradients along the magnetic field, hence increasing the frictional force and decreasing the thermal force on impurities. Changes in convection in the SOL, caused by gaseous fuelling, have been studied, both experimentally in the JET Mk I divertor and with EDGE2/NIMBUS. 1 ref., 4 figs., 1 tab.

  19. Hypertext Configurations

    DEFF Research Database (Denmark)

    Finnemann, Niels Ole

    2017-01-01

    The article presents a conceptual framework for distinguishing different sorts of heterogeneous digital materials. The hypothesis is that a wide range of heterogeneous data resources can be characterized and classified due to their particular configurations of hypertext features such as scripts......, links, interactive processes, and time scalings, and that the hypertext configuration is a major but not sole source of the messiness of big data. The notion of hypertext will be revalidated, placed at the center of the interpretation of networked digital media, and used in the analysis of the fast...

  20. Influence of kinetic effects on a sheath potential and divertor plasma parameters in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Krasheninnikov, S.I.; Soboleva, T.K.; Igitkhanov, Yu.L.; Runov, A.M. (Kurchatov Institute, Moscow (Russian Federation))

    1991-01-01

    It was already noted that strong inhomogeneity of ITER divertor plasma parameters may be a reason of a pronounced deviation of a sheath potential U[sub d] at a plasma-divertor plate contact from the local value U[sub d][approx]3.5T[sub d] (T[sub d] is an electron temperature in a vicinity of the divertor plate). This effect may badly influence the divertor plates sputtering resulting in a plasma contamination. (author) 6 refs., 5 figs.

  1. Plasma flow in recycling region of tokamak divertor and plasma recombination

    Energy Technology Data Exchange (ETDEWEB)

    Soboleva, T.K. [Universidad Nacional Autonoma de Mexico, Mexico City (Mexico). Inst. de Ciencias Nucleares; Krasheninnikov, S.I.; Pigarov, A.Yu.

    1997-12-31

    We investigate the effects of hydrogen molecules and plasma recombination on self-consistent plasma-neutral gas interactions in the recycling region of a tokamak divertor. We treat the plasma flow in a fluid approximation retaining the effects of plasma recombination and employing a Knudsen neutral transport model for a `gas box` divertor geometry. For the model of plasma-neutral interactions we employ we find: a) molecular activated recombination is a dominant channel of divertor plasma recombination; and b) plasma recombination is a key element leading to a decrease in the plasma flux onto the target and substantial plasma pressure drop which are the main features of detached divertor regimes. (author)

  2. First annual report of the Divertor Task Force: Progress and plans

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-10-01

    This report describes the work of the Divertor Task Force of the Massachusetts Institute of Technology Plasma Fusion Center, particularly the Task Force`s founding meeting, original research and development needs, organization, and achievements of its first year. The Task Force`s goal is to obtain an increasingly complete physics understanding of existing divertor plasmas, to build analytical and numerical models of the scrape-off-layer divertor plasmas, and to extrapolate them to find design solutions for the high power divertors of ignited tokamak plasmas such as those of ITER and other high performance future tokamaks. 67 refs., 2 figs.

  3. Measurements of tungsten migration in the DIII-D divertor

    Science.gov (United States)

    Wampler, W. R.; Rudakov, D. L.; Watkins, J. G.; McLean, A. G.; Unterberg, E. A.; Stangeby, P. C.

    2017-12-01

    An experimental study of migration of tungsten in the DIII-D divertor is described, in which the outer strike point of L-mode plasmas was positioned on a toroidal ring of tungsten-coated metal inserts. Net deposition of tungsten on the divertor just outside the strike point was measured on graphite samples exposed to various plasma durations using the divertor materials evaluation system. Tungsten coverage, measured by Rutherford backscattering spectroscopy (RBS), was found to be low and nearly independent of both radius and exposure time closer to the strike point, whereas farther from the strike point the W coverage was much larger and increased with exposure time. Depth profiles from RBS show this was due to accumulation of thicker mixed-material deposits farther from the strike point where the plasma temperature is lower. These results are consistent with a low near-surface steady-state coverage on graphite undergoing net erosion, and continuing accumulation in regions of net deposition. This experiment provides data needed to validate, and further improve computational simulations of erosion and deposition of material on plasma-facing components and transport of impurities in magnetic fusion devices. Such simulations are underway and will be reported later.

  4. Divertor extreme ultraviolet (EUV) survey spectroscopy in DIII-D

    Science.gov (United States)

    McLean, Adam; Allen, Steve; Ellis, Ron; Jarvinen, Aaro; Soukhanovskii, Vlad; Boivin, Rejean; Gonzales, Eduardo; Holmes, Ian; Kulchar, James; Leonard, Anthony; Williams, Bob; Taussig, Doug; Thomas, Dan; Marcy, Grant

    2017-10-01

    An extreme ultraviolet spectrograph measuring resonant emissions of D and C in the lower divertor has been added to DIII-D to help resolve an 2X discrepancy between bolometrically measured radiated power and that predicted by boundary codes for DIII-D, JET and ASDEX-U. With 290 and 450 gr/mm gratings, the DivSPRED spectrometer, an 0.3 m flat-field McPherson model 251, measures ground state transitions for D (the Lyman series) and C (e.g., C IV, 155 nm) which account for >75% of radiated power in the divertor. Combined with Thomson scattering and imaging in the DIII-D divertor, measurements of position, temperature and fractional power emission from plasma components are made and compared to UEDGE/SOLPS-ITER. Mechanical, optical, electrical, vacuum, and shielding aspects of DivSPRED are presented. Work supported under USDOE Cooperative Agreement DE-FC02-04ER54698 and DE-AC52-07NA27344, and by the LLNL Laboratory Directed R&D Program, project #17-ERD-020.

  5. Divertor power load studies for attached L-mode single-null plasmas in TCV

    Science.gov (United States)

    Maurizio, R.; Elmore, S.; Fedorczak, N.; Gallo, A.; Reimerdes, H.; Labit, B.; Theiler, C.; Tsui, C. K.; Vijvers, W. A. J.; The TCV Team; The MST1 Team

    2018-01-01

    This paper investigates the power loads at the inner and outer divertor targets of attached, Ohmic L-mode, deuterium plasmas in the TCV tokamak, in various experimental situations using an Infrared thermography system. The study comprises variations of the outer divertor leg length and target flux expansion, the plasma current and a reversal of the magnetic field direction. The direct impact of the divertor magnetic geometry on scrape-off layer (SOL) transport—parameterised by the SOL power fall-off length λq, u , the divertor spreading factor S u and the in-out power asymmetry—is reported for constant core properties. The in-out power asymmetry increases, either with the divertor leg length, or the target flux expansion. The SOL width λq, u scales positively with divertor leg length, with a strength that depends on the field direction and differs between the inner and outer divertor. This implies a parametric dependence of λq, u that is not explicitly included in current multi-machine scaling laws. The divertor spreading factor at the target S = Su fx , where f x is the target flux expansion, appears unaffected by changes in the divertor geometry and in the plasma current, is independent of the magnetic field direction and is similar between inner and outer divertor. Possible interpretations of these observations using an ad-hoc analytical purely conductive model for the SOL, by ion drifts or by asymmetric turbulent cross-field transport in the divertor are presented. The observed values of λq, u are related to existing L-mode and H-mode scaling laws and to similar studies in other tokamaks. Finally, potential implications of these findings for future larger fusion machines are discussed.

  6. A multi-institutional Stellarator Configuration Study

    Science.gov (United States)

    Gates, David

    2017-10-01

    A multi-institutional study aimed at mapping the space of quasi-axisymmetric stellarators has begun. The goal is to gain improved understanding of the dependence of important physics and engineering parameters (e.g. bootstrap current, stability, coil complexity, etc.) on plasma shape (average elongation, aspect ratio, number of periods). In addition, the stellarator optimization code STELLOPT will be upgraded with new capabilities such as improved coil design algorithms such as COILOPT + + and REGCOIL, divertor optimization options, equilibria with islands using the SPEC code, and improved bootstrap current calculations with the SFINCS code. An effort is underway to develop metrics for divertor optimization. STELLOPT has also had numerous improvements to numerical algorithms and parallelization capabilities. Simultaneously, we also are pursuing the optimization of turbulent transport according to the method of proxy functions. Progress made to date includes an elongation scan on quasi-axisymmetric equilibria and an initial comparison between the SFINCS code and the BOOTSJ calculation of bootstrap current currently available in STELLOPT. Further progress on shape scans and subsequent physics analysis will be reported. The status of the STELLOPT upgrades will be described. The eventual goal of this exercise is to identify attractive configurations for future US experimental facilities.. This work is supported by US DoE Contract Number DE-AC02-09CH11466.

  7. Relationship of edge localized mode burst times with divertor flux loop signal phase in JET

    Energy Technology Data Exchange (ETDEWEB)

    Chapman, S. C., E-mail: S.C.Chapman@warwick.ac.uk [Centre for Fusion, Space and Astrophysics, Department of Physics, University of Warwick, Coventry (United Kingdom); Max Planck Institute for the Physics of Complex Systems, Dresden (Germany); Dendy, R. O. [Centre for Fusion, Space and Astrophysics, Department of Physics, University of Warwick, Coventry (United Kingdom); Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxfordshire (United Kingdom); JET-EFDA, Culham Science Centre, Abingdon, Oxfordshire (United Kingdom); Todd, T. N.; Webster, A. J.; Morris, J. [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxfordshire (United Kingdom); JET-EFDA, Culham Science Centre, Abingdon, Oxfordshire (United Kingdom); Watkins, N. W. [Centre for Fusion, Space and Astrophysics, Department of Physics, University of Warwick, Coventry (United Kingdom); Max Planck Institute for the Physics of Complex Systems, Dresden (Germany); Centre for the Analysis of Time Series, London School of Economics, London (United Kingdom); Department of Engineering and Innovation, Open University, Milton Keynes (United Kingdom); Calderon, F. A. [Centre for Fusion, Space and Astrophysics, Department of Physics, University of Warwick, Coventry (United Kingdom); JET-EFDA, Culham Science Centre, Abingdon, Oxfordshire (United Kingdom)

    2014-06-15

    A phase relationship is identified between sequential edge localized modes (ELMs) occurrence times in a set of H-mode tokamak plasmas to the voltage measured in full flux azimuthal loops in the divertor region. We focus on plasmas in the Joint European Torus where a steady H-mode is sustained over several seconds, during which ELMs are observed in the Be II emission at the divertor. The ELMs analysed arise from intrinsic ELMing, in that there is no deliberate intent to control the ELMing process by external means. We use ELM timings derived from the Be II signal to perform direct time domain analysis of the full flux loop VLD2 and VLD3 signals, which provide a high cadence global measurement proportional to the voltage induced by changes in poloidal magnetic flux. Specifically, we examine how the time interval between pairs of successive ELMs is linked to the time-evolving phase of the full flux loop signals. Each ELM produces a clear early pulse in the full flux loop signals, whose peak time is used to condition our analysis. The arrival time of the following ELM, relative to this pulse, is found to fall into one of two categories: (i) prompt ELMs, which are directly paced by the initial response seen in the flux loop signals; and (ii) all other ELMs, which occur after the initial response of the full flux loop signals has decayed in amplitude. The times at which ELMs in category (ii) occur, relative to the first ELM of the pair, are clustered at times when the instantaneous phase of the full flux loop signal is close to its value at the time of the first ELM.

  8. Near-infrared spectroscopy for divertor plasma diagnosis and control in DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V. A., E-mail: vlad@llnl.gov; McLean, A. G.; Allen, S. L. [Lawrence Livermore National Laboratory, Livermore, California 94551 (United States)

    2014-11-15

    New near infrared (NIR) spectroscopic measurements performed in the DIII-D tokamak divertor plasma suggest new viable diagnostic applications: divertor recycling and low-Z impurity flux measurements, a spectral survey for divertor Thomson scattering (DTS) diagnostic, and T{sub e} monitoring for divertor detachment control. A commercial 0.3 m spectrometer coupled to an imaging lens via optical fiber and a InGaAs 1024 pixel array detector enabled deuterium and impurity emission measurements in the range 800–2300 nm. The first full NIR survey identified D, He, B, Li, C, N, O, Ne lines and provided plasma T{sub e}, n{sub e} estimates from deuterium Paschen and Brackett series intensity and Stark line broadening analysis. The range 1.000–1.060 mm was surveyed in high-density and neon seeded divertor plasmas for spectral background emission studies for λ = 1.064 μm laser-based DTS development. The ratio of adjacent deuterium Paschen-α and Brackett Br9 lines in recombining divertor plasmas is studied for divertor T{sub e} monitoring aimed at divertor detachment real-time feedback control.

  9. Near-infrared spectroscopy for divertor plasma diagnosis and control in DIII-D tokamak.

    Science.gov (United States)

    Soukhanovskii, V A; McLean, A G; Allen, S L

    2014-11-01

    New near infrared (NIR) spectroscopic measurements performed in the DIII-D tokamak divertor plasma suggest new viable diagnostic applications: divertor recycling and low-Z impurity flux measurements, a spectral survey for divertor Thomson scattering (DTS) diagnostic, and Te monitoring for divertor detachment control. A commercial 0.3 m spectrometer coupled to an imaging lens via optical fiber and a InGaAs 1024 pixel array detector enabled deuterium and impurity emission measurements in the range 800-2300 nm. The first full NIR survey identified D, He, B, Li, C, N, O, Ne lines and provided plasma Te, ne estimates from deuterium Paschen and Brackett series intensity and Stark line broadening analysis. The range 1.000-1.060 mm was surveyed in high-density and neon seeded divertor plasmas for spectral background emission studies for λ = 1.064 μm laser-based DTS development. The ratio of adjacent deuterium Paschen-α and Brackett Br9 lines in recombining divertor plasmas is studied for divertor Te monitoring aimed at divertor detachment real-time feedback control.

  10. Surface heat flux feedback controlled impurity seeding experiments with Alcator C-Mod’s high-Z vertical target plate divertor: performance, limitations and implications for fusion power reactors

    Science.gov (United States)

    Brunner, D.; Wolfe, S. M.; LaBombard, B.; Kuang, A. Q.; Lipschultz, B.; Reinke, M. L.; Hubbard, A.; Hughes, J.; Mumgaard, R. T.; Terry, J. L.; Umansky, M. V.; The Alcator C-Mod Team

    2017-08-01

    The Alcator C-Mod team has recently developed a feedback system to measure and control surface heat flux in real-time. The system uses real-time measurements of surface heat flux from surface thermocouples and a pulse-width modulated piezo valve to inject low-Z impurities (typically N2) into the private flux region. It has been used in C-Mod to mitigate peak surface heat fluxes  >40 MW m-2 down to    1. While the system works quite well under relatively steady conditions, use of it during transients has revealed important limitations on feedback control of impurity seeding in conventional vertical target plate divertors. In some cases, the system is unable to avoid plasma reattachment to the divertor plate or the formation of a confinement-damaging x-point MARFE. This is due to the small operational window for mitigated heat flux in the parameters of incident plasma heat flux, plasma density, and impurity density as well as the relatively slow response of the impurity gas injection system compared to plasma transients. Given the severe consequences for failure of such a system to operate reliably in a reactor, there is substantial risk that the conventional vertical target plate divertor will not provide an adequately controllable system in reactor-class devices. These considerations motivate the need to develop passively stable, highly compliant divertor configurations and experimental facilities that can test such possible solutions.

  11. First experience with the Dynamic Ergodic Divertor on TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Finken, K.H.; Abdullaev, S.S.; Jakubowski, M.; Kobayshi, M.; Lehnen, M.; Matsunaga, G.; Pospieszczyk, A.; Schweer, B.; Sergienko, G.; Wolf, R. [Inst. fuer Plasmaphysik, Forschungszentrum Juelich GmbH, EURATOM Association, Partner in the Trilateral Euregio Cluster, Juelich (Germany)

    2004-07-01

    The Dynamic Ergodic Divertor, DED, is a new experiment on the TEXTOR tokamak in Juelich. The DED consists of a set of coils with DC or AC (4 phases) electrical currents flowing parallel to the magnetic field lines. This causes a braiding of the magnetic flux tubes which is called ergodization. The strongly deflected field lines at the plasma edge form the laminar zone. The dynamic operation of the DED (AC current operation) should distribute the heat load to a large surface area and possibly induce a rotation of the plasma. First results are discussed. (orig.)

  12. Supply of a prototype component for the ITER divertor baffle

    Energy Technology Data Exchange (ETDEWEB)

    Bobin-Vastra, I. E-mail: isabelle.bobinvastra@framatome-anp.com; Febvre, M. E-mail: max.febvre@framatome-anp.com; Schedler, B. E-mail: bertram.schedler@plansee.at; Ploechl, L.; Bouveret, Y.; Cauvin, D. E-mail: dominique.cauvin@htm-sa.fr; Raisson, G. E-mail: gerard.raisson@htm-sa.fr; Merola, M. E-mail: merolam@ipp.mpg.de

    2001-10-01

    The ITER divertor baffle is one of the Plasma facing components which are developed in the frame of the ITER concept. The supply consisted in the manufacturing of four panels with four First Wall geometries using macroblock or heat sink+armour concepts. DS-Copper, and CuCrZr were the materials for the heat sink, and CFC or Tungsten Plasma spray were the armour. The panels included two Copper-based tubes each. The final purpose is the comparison of the fabricability of each type and the performances of each panel under heat fluxes.

  13. Surface thermocouples for measurement of pulsed heat flux in the divertor of the Alcator C-Mod tokamak.

    Science.gov (United States)

    Brunner, D; LaBombard, B

    2012-03-01

    A novel set of thermocouple sensors has been developed to measure heat fluxes arriving at divertor surfaces in the Alcator C-Mod tokamak, a magnetic confinement fusion experiment. These sensors operate in direct contact with the divertor plasma, which deposits heat fluxes in excess of ~10 MW/m(2) over an ~1 s pulse. Thermoelectric EMF signals are produced across a non-standard bimetallic junction: a 50 μm thick 74% tungsten-26% rhenium ribbon embedded in a 6.35 mm diameter molybdenum cylinder. The unique coaxial geometry of the sensor combined with its single-point electrical ground contact minimizes interference from the plasma/magnetic environment. Incident heat fluxes are inferred from surface temperature evolution via a 1D thermal heat transport model. For an incident heat flux of 10 MW/m(2), surface temperatures rise ~1000 °C/s, corresponding to a heat flux flowing along the local magnetic field of ~200 MW/m(2). Separate calorimeter sensors are used to independently confirm the derived heat fluxes by comparing total energies deposited during a plasma pulse. Langmuir probes in close proximity to the surface thermocouples are used to test plasma-sheath heat transmission theory and to identify potential sources of discrepancies among physical models.

  14. Results and analysis of high heat flux tests on a full scale vertical target prototype of ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Missirlian, M.; Escourbiac, F.; Schlosser, J.; Durocher, A. [Association Euratom-CEA, Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Merola, M. [EFDA Close Support Unit, Garching (Germany); Bobin-Vastra, I. [Framatome, 71 - Le Creusot (France)

    2004-07-01

    After an extensive development program, a Full-Scale Divertor Target prototype (VTFS) manufactured with all the main features of the corresponding ITER divertor, was intensively tested in the high heat flux FE200 facility. The prototype consists of four units having a full mono-block geometry. The lower part (CFC armour) and the upper part (W armour) of each mono-block were joined to the solution annealed, quenched and cold worked CuCrZr tube by HIP technique. The CFC mono-block was successfully tested up to 1000 cycles at 23 MW/m{sup 2} without any indication of failure. This value is well beyond the ITER design target of 300 cycles at 20 MW/m{sup 2}. The W mono-block endured {approx}600 cycles at 10 MW/m{sup 2}. This value of flux is one order of magnitude higher than the ITER design target for the upper part of the vertical target. Fatigue damage is observed when pursuing the cycling up to 15 MW/m{sup 2}. A first stress analysis seems to predict these factual results. However, macro-graphic examinations should bring a better damage valuation. Meanwhile, the fatigue testing will continue on the W healthy part of the VTFS prototype with castellation located on the heated surface (reducing the stresses close to the W-Cu interface). (authors)

  15. Research proposal on : amplitude modulated reflectometry system for JET divertor

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez, J.; Branas, T.; Estrada, T.; Luna, E. de la

    1992-12-31

    Amplitude Modulated reflectrometry is presented here as a tool for density profile measurements in the JET divertor plasmas. One of the main problems which has been presented in most reflectometers during the last years is the need for a coherent tracking of the phase delay: fast density fluctuations and strong modulation on the amplitude of the reflected signal usually bring to fringe jumps` in the phase signal, which are a big problem when the phase values are much larger than 2 pi. The conditions in the JET divertor plasmas: plasma geometry, access and long oversized broad-band waveguide paths makes very difficult the phase measurements at the millimeter wave range. AM reflectrometry is to some extension an intermediate solution between the classical phase delay reflectometry, so far applied to small distances, and the time domain reflectrometry, used for ionospheric studies and recently also proposed for fusion plasma. the main advantage is to allow the use of millimeter wave reflectometry with moderate phase shifts (approx 2 pi). (author)

  16. Research proposal on : amplitude modulated reflectometry system for JET divertor

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez, J.; Branas, T.; Estrada, T.; Luna, E. de la.

    1992-01-01

    Amplitude Modulated reflectrometry is presented here as a tool for density profile measurements in the JET divertor plasmas. One of the main problems which has been presented in most reflectometers during the last years is the need for a coherent tracking of the phase delay: fast density fluctuations and strong modulation on the amplitude of the reflected signal usually bring to fringe jumps' in the phase signal, which are a big problem when the phase values are much larger than 2 pi. The conditions in the JET divertor plasmas: plasma geometry, access and long oversized broad-band waveguide paths makes very difficult the phase measurements at the millimeter wave range. AM reflectrometry is to some extension an intermediate solution between the classical phase delay reflectometry, so far applied to small distances, and the time domain reflectrometry, used for ionospheric studies and recently also proposed for fusion plasma. the main advantage is to allow the use of millimeter wave reflectometry with moderate phase shifts (approx 2 pi). (author)

  17. Implementation of a long leg X-point target divertor in the ARC fusion pilot plant

    Science.gov (United States)

    Kuang, A. Q.; Cao, N. M.; Creely, A. J.; Dennett, C. A.; Hecla, J.; Hoffman, H.; Major, M.; Ruiz Ruiz, J.; Tinguely, R. A.; Tolman, E. A.; Brunner, D.; Labombard, B.; Sorbom, B. N.; Whyte, D. G.; Grover, P.; Laughman, C.

    2017-10-01

    A long leg X-point target divertor geometry in a double null geometry has been implemented in the ARC pilot plant design, exploiting ARC's demountable toroidal field (TF) coils and FLiBe immersion blanket, which allow superconducting poloidal field coils to be located inside the TF coils, adequately shielded from neutrons. This new design maintains the original TF coil size, core plasma shape, and attains a tritium breedin ratio 1.08. The long leg divertor geometry provides significant advantages. Neutron transport computations indicate a factor of 10 reduction in divertor material neutron damage rate compared to the first wall, easing requirements for high heat flux components. Simulations have shown that long legged divertors are able to maintain a passively stable detachment front that stays in the divertor leg over a wide power window, in principle, responding immediately to fast changes in power exhaust. The ARC design exploits this new paradigm for divertor heat flux control: fewer concerns about coping with fast transients and a focus on neutron-tolerant diagnostics to measure and adjust detachment front locations in the outer divertor legs over long timescales.

  18. Enhanced visible and near-infrared capabilities of the JET mirror-linked divertor spectroscopy system.

    Science.gov (United States)

    Lomanowski, B A; Meigs, A G; Conway, N J; Zastrow, K-D; Sharples, R M; Heesterman, P; Kinna, D

    2014-11-01

    The mirror-linked divertor spectroscopy diagnostic on JET has been upgraded with a new visible and near-infrared grating and filtered spectroscopy system. New capabilities include extended near-infrared coverage up to 1875 nm, capturing the hydrogen Paschen series, as well as a 2 kHz frame rate filtered imaging camera system for fast measurements of impurity (Be II) and deuterium Dα, Dβ, Dγ line emission in the outer divertor. The expanded system provides unique capabilities for studying spatially resolved divertor plasma dynamics at near-ELM resolved timescales as well as a test bed for feasibility assessment of near-infrared spectroscopy.

  19. Enhanced visible and near-infrared capabilities of the JET mirror-linked divertor spectroscopy system

    Energy Technology Data Exchange (ETDEWEB)

    Lomanowski, B. A., E-mail: b.a.lomanowski@durham.ac.uk; Sharples, R. M. [Centre for Advanced Instrumentation, Department of Physics, Durham University, Durham DH1 3LE (United Kingdom); Meigs, A. G.; Conway, N. J.; Zastrow, K.-D.; Heesterman, P.; Kinna, D. [EURATOM/CCFE Fusion Association, Culham Science Center, Abingdon OX14 3DB (United Kingdom); Collaboration: JET-EFDA Team

    2014-11-15

    The mirror-linked divertor spectroscopy diagnostic on JET has been upgraded with a new visible and near-infrared grating and filtered spectroscopy system. New capabilities include extended near-infrared coverage up to 1875 nm, capturing the hydrogen Paschen series, as well as a 2 kHz frame rate filtered imaging camera system for fast measurements of impurity (Be II) and deuterium Dα, Dβ, Dγ line emission in the outer divertor. The expanded system provides unique capabilities for studying spatially resolved divertor plasma dynamics at near-ELM resolved timescales as well as a test bed for feasibility assessment of near-infrared spectroscopy.

  20. Kinetic approach to the helium transport in a divertor plasma along the magnetic field

    Energy Technology Data Exchange (ETDEWEB)

    Krasheninnikov, S.I.; Soboleva, T.K. (I.V. Kurchatov Inst. of Atomic Energy, Ploshchad akademika Kurchatova, 123281 Moscos (SU)); Gac, K. (Instytut Fizyki Plazmy i Laserowej Mikrosyntezy, Warsaw (Poland))

    1990-11-01

    This paper considers impurity (helium) ion transport kinetics in a tokamak divertor along magnetic field lines, both analytically and numerically, for the case when the ratio of collisional mean-free-path to the characteristic length of plasma parameter variation is not too small. To obtain the numerical solution of the kinetics equation, the stochastic modeling method is used. For International Thermonuclear Experimental Reactor (ITER) divertor plasma conditions, the influence of thermal force on helium ions is expected to be decreased considerably. As a result, the helium ion flux toward the divertor plates may be significantly enhanced compared to that predicted by the hydrodynamics approach.

  1. Particle and power deposition on divertor targets in EAST H-mode plasmas

    DEFF Research Database (Denmark)

    Wang, L.; Xu, G.S.; Guo, H.Y.

    2012-01-01

    ELMs were chosen for analysis in order to reduce the uncertainty resulting from the influence of fast electrons on Langmuir triple-probe evaluation during ELMs. The power deposition obtained from Langmuir triple probes was consistent with that from the divertor infra-red camera during an ELM-free...... significantly broadening the SOL width and plasma-wetted area on the divertor target in both LHCD and LHCD + ICRH H-modes, thus posing a great challenge for the next-step high-power, long-pulse operation in EAST. Increasing the divertor-wetted area was also observed to reduce the peak heat flux and particle...

  2. Development of heat sink concept for near-term fusion power plant divertor

    Science.gov (United States)

    Rimza, Sandeep; Khirwadkar, Samir; Velusamy, Karupanna

    2017-04-01

    Development of an efficient divertor concept is an important task to meet in the scenario of the future fusion power plant. The divertor, which is a vital part of the reactor has to discharge the considerable fraction of the total fusion thermal power (∼15%). Therefore, it has to survive very high thermal fluxes (∼10 MW/m2). In the present paper, an efficient divertor heat exchanger cooled by helium is proposed for the fusion tokamak. The Plasma facing surface of divertor made-up of several modules to overcome the stresses caused by high heat flux. The thermal hydraulic performance of one such module is numerically investigated in the present work. The result shows that the proposed design is capable of handling target heat flux values of 10 MW/m2. The computational model has been validated against high-heat flux experiments and a satisfactory agreement is noticed between the present simulation and the reported results.

  3. The Influence of Opacity on Hydrogen Line Emission and Ionisation Balance in High Density Divertor Plasmas

    OpenAIRE

    Behringer, K.

    1997-01-01

    The influence of opacity on hydrogen line emission and ionisation balance in high density divertor plasmas. - Garching bei München : Max-Planck-Inst. für Plasmaphysik, 1997. - 21 S. - (IPP-Report ; 10/5)

  4. ATHENA calculation model for the ITER-FEAT divertor cooling system. Final report with updates

    Energy Technology Data Exchange (ETDEWEB)

    Eriksson, John; Sjoeberg, A.; Sponton, L.L

    2001-05-01

    An ATHENA model of the ITER-FEAT divertor cooling system has been developed for the purpose of calculating and evaluating consequences of different thermal-hydraulic accidents as specified in the Accident Analysis Specifications for the ITER-FEAT Generic Site Safety Report. The model is able to assess situations for a variety of conceivable operational transients from small flow disturbances to more critical conditions such as total blackout caused by a loss of offsite and emergency power. The main objective for analyzing this type of scenarios is to determine margins against jeopardizing the integrity of the divertor cooling system components and pipings. The model of the divertor primary heat transport system encompasses the divertor cassettes, the port limiter systems, the pressurizer, the heat exchanger and all feed and return pipes of these components. The development was pursued according to practices and procedures outlined in the ATHENA code manuals using available modelling components such as volumes, junctions, heat structures and process controls.

  5. Initial observations on core transport in W7-X island divertor plasmas

    Science.gov (United States)

    Pablant, Novimir; W7-X Team

    2017-10-01

    The current campaign of the Wendelstein 7-X (W7-X) stellarator, specified as OP1.2a, features the first operation with an island divertor and a completed carbon first wall. With the completion of the divertor, and recent upgrades to the ECRH heating system, higher temperatures and densities are expected than previously available during the first campaign (OP1.1), which featured a limiter plasma. After completion of wall conditioning, plasmas with Te Ti are expected to become accessible, allowing the investigation of plasma performance in the ion-root regime. Initial investigations of core transport in the W7-X island divertor are reported, along with measurements of the radial electric field. Measurements of temperature, density and radial electric field are compared at similar ECRH input powers between the island divertor plasmas from OP1.2a and the limiter plasmas from OP1.1.

  6. A snowflake divertor: a possible solution to the power exhaust problem for tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Ryutov, D. D.; Cohen, R. H.; Rognlien, T. D.; Umansky, M. V.

    2012-11-21

    This paper summarizes recent progress in the theory of a snowflake divertor, a possible path to reduce both steady-state and intermittent heat loads on the divertor plates to an acceptable level. The most important feature of a SF divertor is the presence of a large zone of a very weak poloidal magnetic field around the poloidal field (PF) null. Qualitative explanation of a variety of new features characteristic of a SF divertor is provided based on simple scaling relations. The main part of the paper is focused on the concept of spreading of the heat flux by curvature-driven convection near the PF null. References to experimental results from the NSTX and TCV tokamaks are provided.

  7. Stability and heating of a poloidal divertor tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Biddle, A. P.; Dexter, R. N.; Holly, D. T.; Lipschultz, B.; Osborne, T. H.; Prager, S. C.; Shepard, D.A., Sprott, J.C.; Witherspoon, F. D.

    1980-06-01

    Five experimental studies - two stability and three heating investigations - have been carried out on Tokapole II, a Tokamak with a four node poloidal divertor. First, discharges have been attained with safety factor q as low as 0.6 over most of the column without degradation of confinement, and correlation of helical instability onset with current profile shape is being studied. Second, the axisymmetric instability has been investigated in detail for various noncircular cross-sectional shapes, and results have been compared with a numerical stability code adapted to the Tokapole machine. Third, application of high power fast wave ion cyclotron resonance heating doubles the ion temperature and permits observation of heating as a function of harmonic number and spatial location of the resonance. Fourth, low power shear Alfven wave propagation is underway to test the applicability of this heating method to tokamaks. Fifth, preionization by electron cyclotron heating has been employed to reduce the startup loop voltage by approx. 60%.

  8. Physics conclusions in support of ITER W divertor monoblock shaping

    Directory of Open Access Journals (Sweden)

    R.A. Pitts

    2017-08-01

    Full Text Available The key remaining physics design issue for the ITER tungsten (W divertor is the question of monoblock (MB front surface shaping in the high heat flux target areas of the actively cooled targets. Engineering tolerance specifications impose a challenging maximum radial step between toroidally adjacent MBs of 0.3mm. Assuming optical projection of the parallel heat loads, magnetic shadowing of these edges is required if quasi-steady state melting is to be avoided under certain conditions during burning plasma operation and transiently during edge localized mode (ELM or disruption induced power loading. An experiment on JET in 2013 designed to investigate the consequences of transient W edge melting on ITER, found significant deficits in the edge power loads expected on the basis of simple geometric arguments, throwing doubt on the understanding of edge loading at glancing field line angles. As a result, a coordinated multi-experiment and simulation effort was initiated via the International Tokamak Physics Activity (ITPA and through ITER contracts, aimed at improving the physics basis supporting a MB shaping decision from the point of view both of edge power loading and melt dynamics. This paper reports on the outcome of this activity, concluding first that the geometrical approximation for leading edge power loading on radially misaligned poloidal leading edges is indeed valid. On this basis, the behaviour of shaped and unshaped monoblock surfaces under stationary and transient loads, with and without melting, is compared in order to examine the consequences of melting, or power overload in context of the benefit, or not, of shaping. The paper concludes that MB top surface shaping is recommended to shadow poloidal gap edges in the high heat flux areas of the ITER divertor targets.

  9. Design of ITER divertor VUV spectrometer and prototype test at KSTAR tokamak

    Science.gov (United States)

    Seon, Changrae; Hong, Joohwan; Song, Inwoo; Jang, Juhyeok; Lee, Hyeonyong; An, Younghwa; Kim, Bosung; Jeon, Taemin; Park, Jaesun; Choe, Wonho; Lee, Hyeongon; Pak, Sunil; Cheon, MunSeong; Choi, Jihyeon; Kim, Hyeonseok; Biel, Wolfgang; Bernascolle, Philippe; Barnsley, Robin; O'Mullane, Martin

    2017-12-01

    Design and development of the ITER divertor VUV spectrometer have been performed from the year 1998, and it is planned to be installed in the year 2027. Currently, the design of the ITER divertor VUV spectrometer is in the phase of detail design. It is optimized for monitoring of chord-integrated VUV signals from divertor plasmas, chosen to contain representative lines emission from the tungsten as the divertor material, and other impurities. Impurity emission from overall divertor plasmas is collimated through the relay optics onto the entrance slit of a VUV spectrometer with working wavelength range of 14.6-32 nm. To validate the design of the ITER divertor VUV spectrometer, two sets of VUV spectrometers have been developed and tested at KSTAR tokamak. One set of spectrometer without the field mirror employs a survey spectrometer with the wavelength ranging from 14.6 nm to 32 nm, and it provides the same optical specification as the spectrometer part of the ITER divertor VUV spectrometer system. The other spectrometer with the wavelength range of 5-25 nm consists of a commercial spectrometer with a concave grating, and the relay mirrors with the same geometry as the relay mirrors of the ITER divertor VUV spectrometer. From test of these prototypes, alignment method using backward laser illumination could be verified. To validate the feasibility of tungsten emission measurement, furthermore, the tungsten powder was injected in KSTAR plasmas, and the preliminary result could be obtained successfully with regard to the evaluation of photon throughput. Contribution to the Topical Issue "Atomic and Molecular Data and their Applications", edited by Gordon W.F. Drake, Jung-Sik Yoon, Daiji Kato, Grzegorz Karwasz.

  10. Enhanced visible and near-infrared capabilities of the JET mirror-linked divertor spectroscopy system.

    OpenAIRE

    Lomanowski, B.A.; Meigs, A.G.; Conway, N. J.; Zastrow, K.-D.; Sharples, R. M.; Heesterman, P.; Kinna, D.; JET EFDA Contributors,

    2014-01-01

    The mirror-linked divertor spectroscopy diagnostic on JET has been upgraded with a new visible and near-infrared grating and filtered spectroscopy system. New capabilities include extended near-infrared coverage up to 1875 nm, capturing the hydrogen Paschen series, as well as a 2 kHz frame rate filtered imaging camera system for fast measurements of impurity (Be II) and deuterium Dα, Dβ, Dγ line emission in the outer divertor. The expanded system provides unique capabilities for studying spat...

  11. Calculation of the radial electric field with RF sheath boundary conditions in divertor geometry

    Science.gov (United States)

    Gui, B.; Xia, T. Y.; Xu, X. Q.; Myra, J. R.; Xiao, X. T.

    2018-02-01

    The equilibrium electric field that results from an imposed DC bias potential, such as that driven by a radio frequency (RF) sheath, is calculated using a new minimal two-field model in the BOUT++ framework. Biasing, using an RF-modified sheath boundary condition, is applied to an axisymmetric limiter, and a thermal sheath boundary is applied to the divertor plates. The penetration of the bias potential into the plasma is studied with a minimal self-consistent model that includes the physics of vorticity (charge balance), ion polarization currents, force balance with E× B , ion diamagnetic flow (ion pressure gradient) and parallel electron charge loss to the thermal and biased sheaths. It is found that a positive radial electric field forms in the scrape-off layer and it smoothly connects across the separatrix to the force-balanced radial electric field in the closed flux surface region. The results are in qualitative agreement with the experiments. Plasma convection related to the E× B net flow in front of the limiter is also obtained from the calculation.

  12. High Performance Double-null Plasma Operation Under Radiating Divertor Conditions

    Science.gov (United States)

    Petrie, T. W.; Osborne, T.; Leonard, A. W.; Luce, T. C.; Petty, C. C.; Fenstermacher, M. E.; Lasnier, C. J.; Turco, F.; Watkins, J. G.

    2017-10-01

    We report on heat flux reduction experiments in which deuterium/neon- or deuterium/argon-based radiating mantle/divertor approaches were applied to high performance double-null (DN) plasmas (H98 1.4-1.7,βN 4 , q 95 6) with a combined neutral beam and ECH power input PIN 15 MW. When the radial location of the ECH deposition is close to the magnetic axis (e.g., ρ seeding' with respect to core dilution, energy confinement, and heat flux reduction under these conditions favors argon. Conditions that lead to an improved τE as predicted previously from ELITE code analysis, i.e., very high PIN, proximity to magnetic balance, and higher q95, are largely consistent with this data. Work was supported by the US DOE under DE-FC02-04ER54698, DE-AC52-07NA27344, DE-FG02-04ER54761, and DE-AC04-94AL85000.

  13. Comparative studies of liquid metals for an alternative divertor target in a fusion reactor

    Science.gov (United States)

    Tabarés, F. L.; Oyarzabal, E.; Tafalla, D.; Martin-Rojo, A. B.; Pastor, I.; Ochando, M. A.; Medina, F.; Zurro, B.; McCarthy, K. J.; the TJ-II Team

    2017-12-01

    Two liquid metals (LM), Li and LiSn (20:80 at), presently considered as alternative materials for the divertor target of a fusion reactor, have been exposed to the plasma in a capillary porous system (CPS) arrangement in TJ-II. A negligible perturbation of the plasma has been recorded in both cases, even when stellarator plasmas are particularly sensitive to high Z elements due to the tendency to central impurity accumulation. The surface temperature of the LM CPS samples (made of a tungsten mesh impregnated in SnLi or Li) has been measured during the plasma pulse with ms resolution by pyrometry and the thermal balance during heating and cooling has been used to obtain the thermal parameters of the SnLi and Li CPS arrangements. Temperatures as high as 1150 K during TJ-II plasma exposure were observed for the LiSn solid case. Strong changes in the thermal conductivity of the alloy were recorded in the cooling phase at temperatures close to the nominal melting point. The deduced values for the thermal conductivity of the LiSn alloy/CPS sample were significantly lower than those predicted from their individual components.

  14. Divertor Heat Flux Control with 3D Stochastic Magnetic Fields during ELM Suppression

    Science.gov (United States)

    Orlov, Dm; Moyer, Ra; Bykov, Io; Evans, Te; Wu, W.; Loarte, A.; Teklu, A.; Watkins, Jg; Wang, H.; Lyons, Bc; Trevisan, Gl; Makowski, Ma; Lasnier, C.; Fenstermacher, Me

    2017-10-01

    Experiments in DIII-D have been performed to modify the divertor heat and particle flux pattern during suppression of ELMs with resonant magnetic perturbation (RMP) fields. In this work, we assessed the impact of small current modulations in a subset of DIII-D I-coils on pedestal profiles, transport and stability as well as divertor conditions. Different I-coil subset ramps were performed allowing for a slow transition of the divertor footprints from n =3 to n =2 and n =1 distributions. We obtained long periods of RMP ELM suppression with slow I-coil quartet ramps. Strong divertor particle flux splitting was observed in these discharges as well as modulation of the divertor heat flux due to changes in toroidal spectrum of applied perturbation. Experimental results are compared to the TRIP3D modeling and to linear M3D-C1 simulations to understand the role of the plasma response on quantitative predictions of the divertor flux splitting. Work supported by US DOE under DE-FC02-04ER54698 and DE-FG02-05ER54809.

  15. Critical need for MFE: the Alcator DX advanced divertor test facility

    Science.gov (United States)

    Vieira, R.; Labombard, B.; Marmar, E.; Irby, J.; Wolf, S.; Bonoli, P.; Fiore, C.; Granetz, R.; Greenwald, M.; Hutchinson, I.; Hubbard, A.; Hughes, J.; Lin, Y.; Lipschultz, B.; Parker, R.; Porkolab, M.; Reinke, M.; Rice, J.; Shiraiwa, S.; Terry, J.; Theiler, C.; Wallace, G.; White, A.; Whyte, D.; Wukitch, S.

    2013-10-01

    Three critical challenges must be met before a steady-state, power-producing fusion reactor can be realized: how to (1) safely handle extreme plasma exhaust power, (2) completely suppress material erosion at divertor targets and (3) do this while maintaining a burning plasma core. Advanced divertors such as ``Super X'' and ``X-point target'' may allow a fully detached, low temperature plasma to be produced in the divertor while maintaining a hot boundary layer around a clean plasma core - a potential game-changer for magnetic fusion. No facility currently exists to test these ideas at the required parallel heat flux densities. Alcator DX will be a national facility, employing the high magnetic field technology of Alcator combined with high-power ICRH and LHCD to test advanced divertor concepts at FNSF/DEMO power exhaust densities and plasma pressures. Its extended vacuum vessel contains divertor cassettes with poloidal field coils for conventional, snowflake, super-X and X-point target geometries. Divertor and core plasma performance will be explored in regimes inaccessible in conventional devices. Reactor relevant ICRF and LH drivers will be developed, utilizing high-field side launch platforms for low PMI. Alcator DX will inform the conceptual development and accelerate the readiness-for-deployment of next-step fusion facilities.

  16. Experimental investigation of transport phenomena in the scrape-off layer and divertor

    Energy Technology Data Exchange (ETDEWEB)

    LaBombard, B.; Goetz, J.A.; Hutchinson, I.; Jablonski, D.; Kesner, J.; Kurz, C.; Lipschultz, B.; McCracken, G.M.; Niemczewski, A.; Terry, J.; Allen, A.; Boivin, R.L.; Bombarda, F.; Bonoli, P.; Christensen, C.; Fiore, C.; Garnier, D.; Golovato, S.; Granetz, R.; Greenwald, M.; Horne, S.; Hubbard, A.; Irby, J.; Lo, D.; Lumma, D.; Marmar, E.; May, M.; Mazurenko, A.; Nachtrieb, R.; Ohkawa, H.; O`Shea, P.; Porkolab, M.; Reardon, J.; Rice, J.; Rost, J.; Schachter, J.; Snipes, J.; Sorci, J.; Stek, P.; Takase, Y.; Wang, Y.; Watterson, R.; Weaver, J.; Welch, B.; Wolfe, S. [Massachusetts Inst. of Technol., Cambridge (United States). Plasma Fusion Center]|[Associazione Euratom-ENEA sulla Fusione, Frascati (Italy)]|[Johns Hopkins University, Baltimore, MD (United States)]|[University of Maryland, College Park, MD (United States)

    1997-02-01

    Transport physics in the divertor and scrape-off layer of Alcator C-Mod is investigated for a wide range of plasma conditions. Parallel (parallel) transport topics include: low recycling, high-recycling, and detached regimes, thermoelectric currents, asymmetric heat fluxes driven by thermoelectric currents, and reversed divertor flows. Perpendicular (perpendicular to) transport topics include: expected and measured scalings of perpendicular to gradients with local conditions, estimated {chi} {sub perpendicular} {sub to} profiles and scalings, divertor neutral retention effects, and L-mode/H-mode effects. Key results are: (i) classical parallel transport is obeyed with ion-neutral momentum coupling effects, (ii) perpendicular to heat transport is proportional to local gradients, (iii) {chi} {sub perpendicular} {sub to} {proportional_to}T{sub e}{sup -0.6} n{sup -0.6} L{sup -0.7} in L-mode, insensitive to toroidal field, (iv) {chi} {sub perpendicular} {sub to} is dependent on divertor neutral retention, (v) H-mode transport barrier effects partially extend inside the SOL, (vi) inside/outside divertor asymmetries may be caused by a thermoelectric instability, and (vii) reversed parallel flows depend on divertor asymmetries and their implicit ionization source imbalances. (orig.).

  17. Assessment of the W7-X high heat flux divertor with thermo-mechanical analysis

    Energy Technology Data Exchange (ETDEWEB)

    Qian, Xinyuan [School of Nuclear Science and Technology, University of Science and Technology of China, Jinzhai Road 96, 230026 Hefei,Anhui (China); Institute of Plasma Physics, Chinese Academy of Sciences, Shushanhu Road 350, 230031 Hefei, Anhui (China); Peng, Xuebing, E-mail: pengxb@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Shushanhu Road 350, 230031 Hefei, Anhui (China); Fellinger, Joris [Max Planck Institute for Plasma Physics, Wendelsteinstr. 1, 17491 Greifswald (Germany); Boscary, Jean [Max Planck Institute for Plasma Physics, Boltzmannstr. 2, 85748 Garching (Germany); Bykov, Victor [Max Planck Institute for Plasma Physics, Wendelsteinstr. 1, 17491 Greifswald (Germany); Wang, Zhongwei [Institute of Plasma Physics, Chinese Academy of Sciences, Shushanhu Road 350, 230031 Hefei, Anhui (China); Ye, Minyou; Song, Yuntao [School of Nuclear Science and Technology, University of Science and Technology of China, Jinzhai Road 96, 230026 Hefei,Anhui (China); Institute of Plasma Physics, Chinese Academy of Sciences, Shushanhu Road 350, 230031 Hefei, Anhui (China)

    2016-11-01

    Highlights: • Thermo-mechanical analysis of HHF divertor module, TM2H. • Temperature of all parts is acceptable for long pulse operation. • Stress in different parts is mainly caused by different load. • Radial displacement need to be improved based on FE calculations. - Abstract: The Wendelstein 7-X is an experimental device designed with a stellarator magnetic confinement for stationary plasma operation (up to 30 min). At the first stage, it is scheduled to start with an inertially cooled test divertor unit and a shorter plasma pulse operation up to 10 s. After the completion of this stage, a water-cooled high heat flux (HHF) divertor will be installed for the steady-state operation phase. The divertor consists of individual target modules, which are sets of target elements armored with CFC tiles supported by a stainless steel structure and fed in parallel with manifolds. Detailed thermo-mechanical analysis of the target modules using the finite element method has been performed to validate and/or improve the elected design of the HHF divertor under operation. Different operating conditions have been studied and the effect of the variation of the convective heat flux pattern with localized heating loads as high as 10 MW/m{sup 2} onto the target elements has been computed. The analysis of the thermal response, stress distribution and deformation allowed a better understanding of the behavior of the divertor modules under operation and confirmed the suitability of the design.

  18. Modeling of linear divertor plasma simulator experiments with three-dimensional target structure by using EMC3-eIRENE code

    Energy Technology Data Exchange (ETDEWEB)

    Kuwabara, T. [Institute of Materials and Systems for Sustainability, Nagoya University, Nagoya (Japan); Tanaka, H.; Kobayashi, M. [National Institute for Fusion Science, Toki (Japan); SOKENDAI (The Graduate University For Advanced Studies), Toki (Japan); Kawamura, G. [National Institute for Fusion Science, Toki (Japan); Ohno, N.; Nishikata, H. [Graduate School of Engineering, Nagoya University, Nagoya (Japan); Feng, Y. [Max-Planck-Institut fuer Plasmaphysik, Euratom Association, Garching/Greifswald (Germany)

    2016-08-15

    We have adapted the EMC3-EIRENE code for modeling of a linear divertor plasma simulator in order to demonstrate plasma-wall interactions with three-dimensional (3D) effects. 3D distributions of hydrogen plasma and neutrals can be successfully calculated for four different types of target plates: a V-shaped target, inclined targets with open and closed structures, and a planer target. Hydrogen atoms and molecules are accumulated more effectively in the V-shaped target plate, leading to a higher electron density with lower electron temperature than the planar target plate. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  19. 'Finnova Development Group'. Comb Configurated Costumer-close Network Installations with Underground Service Boxes; 'Finnova' Innovativ Montage och Systemloesning foer Fjaerrvaermeanslutning av Villaomraade. Huvudloesning 'Kamfoerlaeggning med Serviceskaap'

    Energy Technology Data Exchange (ETDEWEB)

    Gudmundson, Tommy [AaF-Process AB, Stockholm (SE)] (and others)

    2006-07-15

    In this report a completely new approach to the distribution of District Heating Services is presented. The general idea of the solution, here presented under the name Comb Configurated Costumer close Network Installations with Underground Service Boxes (in the following referred to as Finnova AF), was given birth during a very early stage of the working process of the FDG. The goal for our 'Vaermegles'- project ('Vaermegles' means for a residential area to have a low heat demand) is to present a technical solution that implies a substantial cost reduction for establishing DH in such low heat demand residential areas, when compared with what is recognized as to-day applied 'best technology'. The aim point was, in the cost level of 2003/2004, VAT not included, a total investment sum of SEK 50,000 (about 7,000 USD). Using flexible piping, delivered to the site as coils of 100 m of pipes each, installed according to the Comb configuration principle, no joints installed directly in the ground are needed. The number of street crossings is strongly reduced, as compared with today's practice. The total of the joint assembly work takes place inside underground service boxes, situated at the building site boundaries. In this way the continuity of the building process is strongly improved. The stages of operation: digging, installing of pipes, backfilling and arranging of service boxes will take place in shortest possible time. As a prototype, the so called service box (e g a service and connection box, let us call it an 'S/C' box) is square, made of concrete and will be installed with its roof 4-8 inches above ground. Inside the box, a fully equipped DH 'substation', designed as an easily replaceable cassette. In the prototype the closing cap is made from a galvanised steel sheet. The same box and cassette designs can be used for almost all single family houses with a heat demand not exceeding 10 kW. This gives a

  20. Development of a Method for Local Electron Temperature and Density Measurements in the Divertor of the JET Tokamak

    Science.gov (United States)

    Jupen, C.; Meigs, A.; Bhatia, A. K.; Brezinsek, S.; OMullane, M.

    2004-01-01

    Plasma volume recombination in the divertor, a process in which charged particles recombine to neutral atoms, contributes to plasma detachment and hence cooling at the divertor target region. Detachment has been observed at JET and other tokamaks and is known to occur at low electron temperatures (T(sub e)10(exp 20)/m(exp 3)). The ability to measure such low temperatures is therefore of interest for modelling the divertor. In present work we report development of a new spectroscopic technique for investigation of local electron density (n(sub e)) and temperature (T,) in the outer divertor at JET.

  1. VH mode accessibility and global H-mode properties in previous and present JET configurations

    Energy Technology Data Exchange (ETDEWEB)

    Jones, T.T.C.; Ali-Arshad, S.; Bures, M.; Christiansen, J.P.; Esch, H.P.L. de; Fishpool, G.; Jarvis, O.N.; Koenig, R.; Lawson, K.D.; Lomas, P.J.; Marcus, F.B.; Sartori, R.; Schunke, B.; Smeulders, P.; Stork, D.; Taroni, A.; Thomas, P.R.; Thomsen, K. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking

    1994-07-01

    In JET VH modes, there is a distinct confinement transition following the cessation of ELMs, observed in a wide variety of tokamak operating conditions, using both NBI and ICRF heating methods. Important factors which influence VH mode accessibility such as magnetic configuration and vessel conditions have been identified. The new JET pumped divertor configuration has much improved plasma shaping control and power and particle exhaust capability and should permit exploitation of plasmas with VH confinement properties over an even wider range of operating regimes, particularly at high plasma current; first H-modes have been obtained in the 1994 JET operating period and initial results are reported. (authors). 7 refs., 6 figs.

  2. 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.

  3. Developing and validating advanced divertor solutions on DIII-D for next-step fusion devices

    Science.gov (United States)

    Guo, H. Y.; Hill, D. N.; Leonard, A. W.; Allen, S. L.; Stangeby, P. C.; Thomas, D.; Unterberg, E. A.; Abrams, T.; Boedo, J.; Briesemeister, A. R.; Buchenauer, D.; Bykov, I.; Canik, J. M.; Chrobak, C.; Covele, B.; Ding, R.; Doerner, R.; Donovan, D.; Du, H.; Elder, D.; Eldon, D.; Lasa, A.; Groth, M.; Guterl, J.; Jarvinen, A.; Hinson, E.; Kolemen, E.; Lasnier, C. J.; Lore, J.; Makowski, M. A.; McLean, A.; Meyer, B.; Moser, A. L.; Nygren, R.; Owen, L.; Petrie, T. W.; Porter, G. D.; Rognlien, T. D.; Rudakov, D.; Sang, C. F.; Samuell, C.; Si, H.; Schmitz, O.; Sontag, A.; Soukhanovskii, V.; Wampler, W.; Wang, H.; Watkins, J. G.

    2016-12-01

    A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, which we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (n e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). This paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.

  4. Geometrical Effects in Plasma Stability and Dynamics of Coherent Structures in the Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Ryutov, D D; Cohen, R H

    2007-05-16

    Plasma dynamics in the divertor region is strongly affected by a variety of phenomena associated with the magnetic field geometry and the shape of the divertor plates. One of the most universal effects is the squeezing of a normal cross-section of a thin magnetic flux-tube on its way from the divertor plate to the main SOL. It leads to decoupling of the most unstable perturbations in the divertor legs from those in the main SOL. For perturbations on either side of the X-point, this effect can be cast as a boundary condition at some 'control surface' situated near the X-point. We discuss several boundary conditions proposed thus far and assess the influence of the magnetic field geometry on them. Another set of geometrical effects is related to the transformation of a flux-tube that occurs when it is displaced in such a way that its central magnetic field line coincides with some other field line, and the magnetic field is not perturbed. These flute-like displacements are of a particular interest for the low-beta edge plasmas. It turns out that this transformation may also lead to a considerable deformation of a flux-tube cross-section; in addition, the distance between plasma particles occupying the flux-tube may change significantly even if there is no parallel plasma motion. We present expressions describing aforementioned transformations for the general tokamak geometry and simplify them for the divertor region (using the proximity of the X-point). We also discuss the effects associated with the shape of the plasma-limiting surfaces, both those designed to intercept the plasma (like divertor plates and limiters) and those that can be hit in some 'abnormal' events, e.g., in the course of a radial motion of an isolated plasma filament. The orientation of the limiting surface with respect to the magnetic field affects the plasma dynamics via the sheath boundary conditions. One can enhance or suppress plasma instabilities in the divertor legs by

  5. Divertor power spreading in DEMO reactor by impurity seeding

    Energy Technology Data Exchange (ETDEWEB)

    Zagórski, Roman, E-mail: Roman.Zagorski@ipplm.pl; Gałązka, Krzysztof; Ivanova-Stanik, Irena

    2016-11-01

    Highlights: • The COREDIV code has been used to simulate DEMO inductive discharges with different impurity seeding (Ne, Ar, Kr) and different sputtering models (with and w/o prompt re-deposition process). • It has been shown that only for Ar and Kr seeding it is possible to achieve H-mode plasma operation with acceptable level of the power to the tungsten target plates. • For neon seeding, such regime of operation seems not to be possible. • Prompt re-deposition model extends the DEMO operation window. - Abstract: Numerical simulation with COREDIV code of DEMO H-mode discharges (tungsten divertor and wall) are performed considering the influence of seeding impurities with different atomic numbers: Ne, Ar and Kr on the DEMO scenarios. The approach is based on integrated numerical modeling using the COREDIV code, which self-consistently solves radial transport equations in the core region and 2D multi-fluid transport in the SOL. In this paper we focus on investigations how the operational domain of DEMO can be influenced by seeding gasses. Simulations with the updated prompt re-deposition model implemented in the code show that only for Ar and Kr, for high enough radial diffusion in the SOL, it is possible to achieve H-mode plasma operation (power to the SOL> L-H transition threshold power) with acceptable level of the power to the target plates. For neon seeding such regime of operation seems not to be possible.

  6. Acceptance criteria for the ITER divertor vertical target

    Energy Technology Data Exchange (ETDEWEB)

    Fouquet, S. [Association Euratom-CEA, CEA/DSM/DRFC, CEA Cadarache, F-13108 Saint Paul Lez Durance, Cedex (France)]. E-mail: fouquet@drfc.cad.cea.fr; Schlosser, J. [Association Euratom-CEA, CEA/DSM/DRFC, CEA Cadarache, F-13108 Saint Paul Lez Durance, Cedex (France); Merola, M. [EFDA CSU Garching, Boltzmannstr. 2, Garching-bei-Munchen, D-85748 (Germany); Durocher, A. [Association Euratom-CEA, CEA/DSM/DRFC, CEA Cadarache, F-13108 Saint Paul Lez Durance, Cedex (France); Escourbiac, F. [Association Euratom-CEA, CEA/DSM/DRFC, CEA Cadarache, F-13108 Saint Paul Lez Durance, Cedex (France); Grosman, A. [Association Euratom-CEA, CEA/DSM/DRFC, CEA Cadarache, F-13108 Saint Paul Lez Durance, Cedex (France); Missirlian, M. [Association Euratom-CEA, CEA/DSM/DRFC, CEA Cadarache, F-13108 Saint Paul Lez Durance, Cedex (France); Portafaix, C. [Association Euratom-CEA, CEA/DSM/DRFC, CEA Cadarache, F-13108 Saint Paul Lez Durance, Cedex (France)

    2006-02-15

    In the frame of the toroidal pump limiter fabrication for Tore Supra, CEA developed a large experience of infrared test for acceptance of high heat flux components armoured with carbon fibre composite flat tiles. The test is based on a thermal transient induced by an alternative hot/cold water flow in the heat sink structure. The tile surface temperature transients are compared with those of a reference element, the maximum difference for each tile leading to a value called {delta}T {sub ref{sub max}}. This method is proposed for the commissioning of plasma facing components for the ITER divertor vertical target. This paper describes the determination of the best acceptance criteria for the 'monoblock' geometry of the carbon part. First, it has been shown that the location and the extension of the defects could reliably be determined by monitoring both top and lateral surfaces during the test. Second, it was possible to fix an acceptance method based on {delta}T {sub ref{sub max}}. Samples with calibrated defects are now under fabrication to validate the results.

  7. Advances in optical thermometry for the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Lott, F. [CEA, IRFM, F-13108 St Paul lez Durance (France)], E-mail: fraser.lott@gmail.com; Netchaieff, A. [Laboratoire National de Metrologie et d' Essais (LNE), ZA de Trappes-Elancourt, 29 avenue Roger Hennequin, 78197 TRAPPES Cedex (France); Escourbiac, F. [CEA, IRFM, F-13108 St Paul lez Durance (France); Jouvelot, J.-L.; Constans, S. [AREVA NP, Centre Technique-FE200, Porte Magenta BP 181, 71205 Le Creusot (France); Hernandez, D. [Procedes, Materiaux et Energie Solaire (PROMES), Centre National de la Recherche Scientifique (CNRS), B.P. 5, 66125 Font-Romeu Cedex (France)

    2010-01-15

    Thermography will be an important diagnostic on the ITER tokamak, but the inclusion of reflective materials such as tungsten in the design for ITER's first wall and divertor region presents problems for optical temperature measurement. The ongoing testing of ITER plasma facing components (PFCs) provides an excellent opportunity to resolve such problems. This has focused on the variation of PFC emissivity with temperature and time, as well as environmental influence on thermography. The sensitivity of these systems to ambient temperature, due primarily to modification of the transmission of the optical path, has been established and minimised. The accuracy of the system is then sufficient to measure the variation of emissivity in heated material samples, by comparing its front-face luminance measured with an infrared camera to the temperature given by an implanted thermocouple. Measurements on both tungsten and carbon fibre composite are in broad agreement with theory, and thus give the material's function of emissivity with temperature at the start of its life. To determine its evolution, a bicolour pyroreflectometer was then installed. This uses two lasers to measure the reflectivity in addition to the luminance at two wavelengths, and thus the true temperature can be calculated. This was validated against the instrumented sample, then used along with the camera to observe an ITER mock-up during {approx}50,000 s of 5 MW/m{sup 2} testing. Emissivity was seen to vary little in the 500 deg. C region. Higher temperature tests are ongoing.

  8. Experimental test campaign on an ITER divertor mock-up

    Energy Technology Data Exchange (ETDEWEB)

    Dell' Orco, G. E-mail: giovanni.dellorco@brasimone.enea.it; Malavasi, A.; Merola, M.; Polazzi, G.; Simoncini, M.; Zito, D

    2002-11-01

    In 1998, in the frame of the European R and D on ITER high heat flux components, the fabrication of a full scale ITER Divertor Outboard mock-up was launched. It comprised a Cassette Body (CB), designed with some mechanical and hydraulic simplifications with respect to the reference body and its actively cooled Dummy Armour Prototype (DAP). This DAP consists of a Vertical Target (VT), a Wing (WI) and a Dump Target (DT), manufactured by European industries, which are integrated to the Gas Box Liner (GBL) supplied by the Russian Federation ITER Home Team. In 1999, in parallel with the manufacturing activity, the ITER European Home Team decided to assign to ENEA a Task for checking the component integration and performing the thermal-hydraulic and thermal mechanical testing of the DAP and CB. In 1999-2000, ENEA performed the experimental campaign at Brasimone Labs. The present work presents the experimental results of the component integration and the thermal-hydraulic and thermo-mechanical fatigue tests.

  9. Effect of Magnetic Islands on Divertors in Tokamaks and Stellarators

    Science.gov (United States)

    Punjabi, Alkesh; Boozer, Allen

    2017-10-01

    Divertors are required for handling the plasma particle and heat exhausts on the walls in fusion plasmas. Relatively simple methods, models, and maps from field line Hamiltonian are developed to better understand the interaction of strong plasma shaping and magnetic islands on the size and behavior of the magnetic flux tubes that go from the plasma edge to the wall in non-axisymmetric system. This approach is applicable not only in tokamaks but also in stellarators. Stellarator diverters in which magnetic islands are dominant are called resonant and when shaping is dominant are called non-resonant. Optimized stellarators generally have sharp edges on their surface, but unlike the case for tokamaks these edges do not encircle the entire plasma, so they do not define an edge value for the rotational transform. The approach is used in the DIII-D tokamak. Computation results are consistent with the predictions of the models. Further simulations are being done to understand why the transition from an effective cubic to a linear increase in loss time and area of footprint occurs and whether this increase is discontinuous or not. This work is supported by the US DOE Grants DE-FG02-01ER54624 and DE-FG02-04ER54793 to Hampton University and DE-FG02-95ER54333 to Columbia University. This research used resources of the NERSC, supported by the Office of Science, US DOE, under Contract No. DE-AC02-05CH11231.

  10. Erosion and deposition in the JET divertor during the second ITER-like wall campaign

    Science.gov (United States)

    Mayer, M.; Krat, S.; Baron-Wiechec, A.; Gasparyan, Yu; Heinola, K.; Koivuranta, S.; Likonen, J.; Ruset, C.; de Saint-Aubin, G.; Widdowson, A.; Contributors, JET

    2017-12-01

    Erosion of plasma-facing materials and successive transport and redeposition of eroded material are crucial processes determining the lifetime of plasma-facing components and the trapped tritium inventory in redeposited material layers. Erosion and deposition in the JET divertor were studied during the second JET ITER-like wall campaign ILW-2 in 2013–2014 by using a poloidal row of specially prepared divertor marker tiles including the tungsten bulk tile 5. The marker tiles were analyzed using elastic backscattering with 3–4.5 MeV incident protons and nuclear reaction analysis using 0.8–4.5 MeV 3He ions before and after the campaign. The erosion/deposition pattern observed during ILW-2 is qualitatively comparable to the first campaign ILW-1 in 2011–2012: deposits consist mainly of beryllium with 5–20 at.% of carbon and oxygen and small amounts of Ni and W. The highest deposition with deposited layer thicknesses up to 30 μm per campaign is still observed on the upper and horizontal parts of the inner divertor. Outer divertor tiles 5, 6, 7 and 8 are net W erosion areas. The observed D inventory is roughly comparable to the inventory observed during ILW-1. The results obtained during ILW-2 therefore confirm the positive results observed in ILW-1 with respect to reduced material deposition and hydrogen isotopes retention in the divertor.

  11. Experimental and calculated basis of the lithium capillary system as divertor material

    Energy Technology Data Exchange (ETDEWEB)

    Antonov, N.V. [Rossijskij Nauchnyj Tsentr ``Kurchatovskij Inst.``, Moscow (Russian Federation); Belan, V.G. [TRINITI, Troitsk, Moscow region (Russian Federation); Evtihin, V.A. [State Enterprise ``Red Star``, Moscow (Russian Federation); Golubchikov, L.G. [RF Ministry of Atomic Energy, Moscow (Russian Federation); Khripunov, V.I. [Rossijskij Nauchnyj Tsentr ``Kurchatovskij Inst.``, Moscow (Russian Federation); Korjavin, V.M. [RF Ministry of Atomic Energy, Moscow (Russian Federation); Lyublinski, I.E. [State Enterprise ``Red Star``, Moscow (Russian Federation); Maynashev, V.S. [TRINITI, Troitsk, Moscow region (Russian Federation); Petrov, V.B. [Rossijskij Nauchnyj Tsentr ``Kurchatovskij Inst.``, Moscow (Russian Federation); Pistunovich, V.I. [Rossijskij Nauchnyj Tsentr ``Kurchatovskij Inst.``, Moscow (Russian Federation); Pozharov, V.A. [Rossijskij Nauchnyj Tsentr ``Kurchatovskij Inst.``, Moscow (Russian Federation); Podkovirnov, V.I. [TRINITI, Troitsk, Moscow region (Russian Federation); Shapkin, V.V. [Rossijskij Nauchnyj Tsentr ``Kurchatovskij Inst.``, Moscow (Russian Federation); Vertkov, A.V. [State Enterprise ``Red Star``, Moscow (Russian Federation)

    1997-02-01

    First results as experimental and calculated basis of a new concept are described in the paper. Experimental models of liquid lithium capillary structure have been tested at long-pulse high heat loads. The power loads on the capillary target up to 50 MW/m{sup 2} were provided by an electron beam with electron energy {<=}9 keV in a longitudinal magnetic field of 0.25 T. Seven experiments were performed with the different capillary targets. The effects of disruption discharges in tokamaks have been simulated by means of magnetized plasma flows with pulse length of 0.2 ms, electron density of 10{sup 22} m{sup 3} and energy density up to 4 MJ/m{sup 2}. The plasma flow was generated by a quasistationary plasma accelerator and interacted with a lithium capillary structure. 2D modelling of the ITER divertor with a lithium target is presented as the first step in the validation of a new divertor concept. A lithium radiative divertor scenario has been examined for the ITER using DDIC95 code. First calculations have shown that thermal loads on the divertor plates are reduced down to 1.3 MW/m{sup 2}. The main power is radiated in the divertor. (orig.).

  12. Survivability of dust in tokamaks: dust transport in the divertor sheath

    CERN Document Server

    Delzanno, Gian Luca

    2014-01-01

    The survivability of dust being transported in the magnetized sheath near the divertor plate of a tokamak and its impact on the mandatory balance of erosion and redeposition for a steady-state reactor are investigated. Two different divertor scenarios are considered. The first is characterized by an energy flux perpendicular to the plate $q_0\\simeq 1$ MW/m$^2$ typical of current short-pulse tokamaks. The second has $q_0\\simeq 10$ MW/m$^2$ and is relevant to long-pulse machines like ITER or DEMO. It is shown that micrometer dust particles can survive rather easily near the plates of a divertor plasma with $q_0\\simeq 1$ MW/m$^2$ because thermal radiation provides adequate cooling for the dust particle. On the other hand, the survivability of micrometer dust particles near the divertor plates is drastically reduced when $q_0\\simeq 10$ MW/m$^2$. Micrometer dust particles redeposit their material non-locally, leading to a net poloidal mass migration across the divertor. Smaller particles (with radius $\\sim 0.1$ $\\...

  13. Plasma-neutral gas interaction in a tokamak divertor: effects of hydrogen molecules and plasma recombination

    Energy Technology Data Exchange (ETDEWEB)

    Krasheninnikov, S.I. [Massachusetts Inst. of Tech., Cambridge, MA (United States). Plasma Fusion Center]|[I.V. Kurchatov Institute of Atomic Energy, 1 Kurchatov Sq., Moscow 123098 (Russian Federation); Pigarov, A.Yu. [Princeton University, Plasma Physics Laboratory, James Forrestal Campus, P.O. Box 451, Princeton, NJ 08543 (United States)]|[I.V. Kurchatov Institute of Atomic Energy, 1 Kurchatov Sq., Moscow 123098 (Russian Federation); Soboleva, T.K. [Instituto de Ciencias Nucleares, Universidad Nacional Autonoma de Mexico, Apdo. Postal 70-543, 04510 Mexico D.F. (Mexico)]|[I.V. Kurchatov Institute of Atomic Energy, 1 Kurchatov Sq., Moscow 123098 (Russian Federation); Sigmar, D.J. [Massachusetts Inst. of Tech., Cambridge, MA (United States). Plasma Fusion Center

    1997-02-01

    We investigate the influence of hydrogen molecules on plasma recombination using a collisional-radiative model for multispecies hydrogen plasmas and tokamak detached divertor parameters. The rate constant found for molecular activated recombination of a plasma can be as high as 2 x 10{sup -10} cm{sup 3}/s, confirming our pervious estimates. We investigate the effects of hydrogen molecules and plasma recombination on self-consistent plasma-neutral gas interactions in the recycling region of a tokamak divertor. We treat the plasma flow in a fluid approximation retaining the effects of plasma recombination and employing a Knudsen neutral transport model for a `gas box` divertor geometry. For the model of plasma-neutral interactions we employ we find: (a) molecular activated recombination is a dominant channel of divertor plasma recombination; and (b) plasma recombination is a key element leading to a decrease in the plasma flux onto the target and substantial plasma pressure drop which are the main features of detached divertor regimes. (orig.).

  14. Scaling and transport analysis of divertor conditions on the Alcator C-Mod tokamak

    Energy Technology Data Exchange (ETDEWEB)

    LaBombard, B.; Goetz, J.; Kurz, C.; Jablonski, D.; Lipschultz, B.; McCracken, G.; Niemczewski, A.; Boivin, R.L.; Bombarda, F.; Christensen, C.; Fairfax, S.; Fiore, C.; Garnier, D.; Graf, M.; Golovato, S.; Granetz, R.; Greenwald, M.; Horne, S.; Hubbard, A.; Hutchinson, I.; Irby, J.; Kesner, J.; Luke, T.; Marmar, E.; May, M.; O`Shea, P.; Porkolab, M.; Reardon, J.; Rice, J.; Schachter, J.; Snipes, J.; Stek, P.; Takase, Y.; Terry, J.; Tinios, G.; Watterson, R.; Welch, B.; Wolfe, S. [Plasma Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States)

    1995-06-01

    Detailed measurements and transport analysis of divertor conditions in Alcator C-Mod [Phys. Plasmas {bold 1}, 1511 (1994)] are presented for a range of line-averaged densities, 0.7{lt}{ital {bar n}}{sub {ital e}}{lt}2.2{times}10{sup 20} m{sup {minus}3}. Three parallel heat transport regimes are evident in the scrape-off layer: sheath-limited conduction, high-recycling divertor, and detached divertor, which can coexist in the same discharge. {ital Local} cross-field pressure gradients are found to scale simply with a {ital local} electron temperature. This scaling is consistent with classical electron parallel conduction being balanced by anomalous cross-field transport ({chi}{sub {perpendicular}}{similar_to}0.2 m{sup 2} s{sup {minus}1}) proportional to the local pressure gradient. A 60%--80% of divertor power is radiated in attached discharges, approaching 100% in detached discharges. Detachment occurs when the heat flux to the plate is low and the plasma pressure is high ({ital T}{sub {ital e}}{similar_to}5 eV). High neutral pressures in the divertor are nearly always present (1--20 mTorr), sufficient to remove parallel momentum via ion--neutral collisions.

  15. Copper matrix composites as heat sink materials for water-cooled divertor target

    Directory of Open Access Journals (Sweden)

    Jeong-Ha You

    2015-12-01

    Full Text Available According to the recent high heat flux (HHF qualification tests of ITER divertor target mock-ups and the preliminary design studies of DEMO divertor target, the performance of CuCrZr alloy, the baseline heat sink material for DEMO divertor, seems to only marginally cover the envisaged operation regime. The structural integrity of the CuCrZr heat sink was shown to be affected by plastic fatigue at 20 MW/m². The relatively high neutron irradiation dose expected for the DEMO divertor target is another serious concern, as it would cause significant embrittlement below 250 °C or irradiation creep above 350 °C. Hence, an advanced design concept of the divertor target needs to be devised for DEMO in order to enhance the HHF performance so that the structural design criteria are fulfilled for full operation scenarios including slow transients. The biggest potential lies in copper-matrix composite materials for the heat sink. In this article, three promising Cu-matrix composite materials are reviewed in terms of thermal, mechanical and HHF performance as structural heat sink materials. The considered candidates are W particle-reinforced, W wire-reinforced and SiC fiber-reinforced Cu matrix composites. The comprehensive results of recent studies on fabrication technology, design concepts, materials properties and the HHF performance of mock-ups are presented. Limitations and challenges are discussed.

  16. Impact of carbon and tungsten as divertor materials on the scrape-off layer conditions in JET

    NARCIS (Netherlands)

    Groth, M.; Brezinsek, S.; Belo, P.; Beurskens, M. N. A.; Brix, M.; Clever, M.; Coenen, J. W.; Corrigan, C.; Eich, T.; Flanagan, J.; Guillemaut, C.; Giroud, C.; Harting, D.; Huber, A.; Jachmich, S.; Kruezi, U.; Lawson, K. D.; Lehnen, M.; Lowry, C.; Maggi, C. F.; Marsen, S.; Meigs, A. G.; Pitts, R.A.; Sergienko, G.; Sieglin, B.; Silva, C.; Sirinelli, A.; Stamp, M. F.; van Rooij, G. J.; Wiesen, S.; JET-EFDA Contributors,

    2013-01-01

    The impact of carbon and beryllium/tungsten as plasma-facing components on plasma radiation, divertor power and particle fluxes, and plasma and neutral conditions in the divertors has been assessed in JET both experimentally and by edge fluid code simulations for plasmas in low-confinement mode. In

  17. 'Finnova Development Group'. Comb Configurated Costumer-close Network Installations with Underground Service Boxes. From project objectives to main solutions; 'Finnova' Innovativ Montage och Systemloesning foer Fjaerrvaermeanslutning av Villaomraade. Fraan projektmaal till huvudloesningar

    Energy Technology Data Exchange (ETDEWEB)

    Gudmundson, Tommy [AaF-Process AB, Stockholm (SE)] (and others)

    2006-07-15

    in the report. For the first one, 'Finnova AF', evidence is given that the overall goal is fulfilled. The goal being investment costs lower than SEK 50,000 per costumer - VAT not included and in the cost level of 2003/2004. The possibilities for the second one, 'Finnova LTH', may in the long run be even more promising, with respect to economy as well as functionally. The fulfilling of this requires, however, investigations and research and the time for these are not to be found within the time schedule given for this project. Other important issues dealt with regarding the two solutions are: estimated technical life length, need for and accessibility at maintenance and heat losses The two main solutions are named Comb Configurated Costumer Close Network (Finnova AF) and Villa Connection with Distribution Chambers (Finnova LTH). Both are in general terms presented in this report and fully and more detailed in reports no 2 and 3. A demo for the Finnova AF approach is right now being built in a villa area in the Granlunda suburb of Trelleborg in south Sweden. The system decisive features of the Finnova AF are three: The conventional DH substation is abandoned and replaced with a service and connection box in the garden at the site boundary, no pipe joints directly in ground are to be found, and almost all needed working moments can be performed by anyone of a working crew of 5 people. Among the identified success factors, especially the following should be mentioned: Maximum continuity. No 'extern specialists' needed, the civil works contractor should be able to perform 'almost' all working procedures with his own crew. One-step finished backfilling must be applied. And to obtain continuity, flexible coiled pipes and cold installation of network are necessary. The DH substation moved out of the customer's house. The DH supplier is supposed to be the owner of the service and maintenance box replacing the FC. This box contains

  18. Analyses of microstructure, composition and retention of hydrogen isotopes in divertor tiles of JET with the ITER-like wall

    Science.gov (United States)

    Masuzaki, S.; Tokitani, M.; Otsuka, T.; Oya, Y.; Hatano, Y.; Miyamoto, M.; Sakamoto, R.; Ashikawa, N.; Sakurada, S.; Uemura, Y.; Azuma, K.; Yumizuru, K.; Oyaizu, M.; Suzuki, T.; Kurotaki, H.; Hamaguchi, D.; Isobe, K.; Asakura, N.; Widdowson, A.; Heinola, K.; Jachmich, S.; Rubel, M.; contributors, JET

    2017-12-01

    Results of the comprehensive surface analyses of divertor tiles and dusts retrieved from JET after the first ITER-like wall campaign (2011–2012) are presented. The samples cored from the divertor tiles were analyzed. Numerous nano-size bubble-like structures were observed in the deposition layer on the apron of the inner divertor tile, and a beryllium dust with the same structures were found in the matter collected from the inner divertor after the campaign. This suggests that the nano-size bubble-like structures can make the deposition layer to become brittle and may lead to cracking followed by dust generation. X-ray photoelectron spectroscopy analyses of chemical states of species in the deposition layers identified the formation of beryllium-tungsten intermetallic compounds on an inner vertical tile. Different tritium retention profiles along the divertor tiles were observed at the top surfaces and at deeper regions of the tiles by using the imaging plate technique.

  19. Increased heat dissipation with the X-divertor geometry facilitating detachment onset at lower density in DIII-D

    Science.gov (United States)

    Covele, B.; Kotschenreuther, M.; Mahajan, S.; Valanju, P.; Leonard, A.; Watkins, J.; Makowski, M.; Fenstermacher, M.; Si, H.

    2017-08-01

    The X-divertor geometry on DIII-D has demonstrated reduced particle and heat fluxes to the target, facilitating detachment onset at 10-20% lower upstream density and higher H-mode pedestal pressure than a standard divertor. SOLPS modeling suggests that this effect cannot be explained by an increase in total connection length alone, but rather by the addition of connection length specifically in the power-dissipating volume near the target, via poloidal flux expansion and flaring. However, poloidal flaring must work synergistically with divertor closure to most effectively reduce the detachment density threshold. The model also points to carbon radiation as the primary driver of power dissipation in divertors on the DIII-D floor, which is consistent with experimental observations. Sustainable divertor detachment at lower density has beneficial consequences for energy confinement and current drive efficiency for core operation, while simultaneously satisfying the exhaust requirements of the plasma-facing components.

  20. Research of the capillary structure heat removal efficiency under divertor conditions

    Energy Technology Data Exchange (ETDEWEB)

    Pistunovich, V.I. [Rossijskij Nauchnyj Tsentr ``Kurchatovskij Inst.``, Moscow (Russian Federation); Vertkov, A.V. [Stock Corp. `Prana`, Moscow (Russian Federation); Evtikhin, V.A. [Stock Corp. `Prana`, Moscow (Russian Federation); Korjavin, V.M. [Fusion Dept. Ministry of Atomic Energy, Moscow (Russian Federation); Lyublinski, I.E. [Stock Corp. `Prana`, Moscow (Russian Federation); Petrov, V.B. [Rossijskij Nauchnyj Tsentr ``Kurchatovskij Inst.``, Moscow (Russian Federation); Khripunov, B.I. [Rossijskij Nauchnyj Tsentr ``Kurchatovskij Inst.``, Moscow (Russian Federation); Shapkin, V.V. [Rossijskij Nauchnyj Tsentr ``Kurchatovskij Inst.``, Moscow (Russian Federation)

    1996-10-01

    Experimental models of capillary structure for liquid metal fusion reactor divertor simulation have been designed, manufactured and tested in order to estimate the behaviour and possibilities of plasma-facing components based on lithium capillary system at long-pulse high heat load. The power load on the capillary target structures up to 50 MW/m{sup 2} was provided by electron beam with electron energy {<=}10 keV. The exposition-time was up to several minutes and was limited by the lithium quantity in the supply vessel. The operation parameters of the models determined in the experiments are in accordance with there design estimations. The tests of various model constructions at the divertor relevant power loads have shown promise for the new concept of a divertor taking into account long life and reliability. (orig.).

  1. Design of a diagnostic residual gas analyzer for the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Klepper, C.C., E-mail: kleppercc@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831-6169 (United States); Biewer, T.M.; Graves, V.B. [Oak Ridge National Laboratory, Oak Ridge, TN 37831-6169 (United States); Andrew, P. [ITER Organisation, Route de Vinon-sur-Verdon, 13067 St. Paul-lez-Durance (France); Lukens, P.C. [US ITER Project Office, 1055 Commerce Park Dr #1, Oak Ridge, TN 37830 (United States); Marcus, C. [Oak Ridge National Laboratory, Oak Ridge, TN 37831-6169 (United States); Shimada, M., E-mail: shimada.michiya@jaea.go.jp [ITER Organisation, Route de Vinon-sur-Verdon, 13067 St. Paul-lez-Durance (France); Hughes, S.; Boussier, B. [ITER Organisation, Route de Vinon-sur-Verdon, 13067 St. Paul-lez-Durance (France); Johnson, D.W. [US ITER Diagnostics Office, Princeton Plasma Physics Laboratory, Princeton, NJ 08540 (United States); Gardner, W.L. [US ITER Project Office, 1055 Commerce Park Dr #1, Oak Ridge, TN 37830 (United States); Hillis, D.L. [Oak Ridge National Laboratory, Oak Ridge, TN 37831-6169 (United States); Vayakis, G.; Walsh, M. [ITER Organisation, Route de Vinon-sur-Verdon, 13067 St. Paul-lez-Durance (France)

    2015-10-15

    Highlights: • The divertor DRGA for ITER will measure neutral gas composition in the pumping ducts during plasma. • System must respond in timescales relevant to compositional changes in the divertor plasma. • It is shown that times can vary from 1 to 6 s for fuel (H2, D2, T2) up to 50 s for He (fusion reaction ash). • It is shown that present design delivers ∼ 1 s response even via an 8m long sampling pipe sampling. • Response time validated with VacTran{sup ®} over anticipated the 0.1–10 Pa pressure range in the ducts. - Abstract: One of the ITER diagnostics having reached an advanced design stage is a diagnostic RGA for the divertor, i.e. residual gas analysis system for the ITER divertor, which is intended to sample the divertor pumping duct region during the plasma pulse and to have a response time compatible with plasma particle and impurity lifetimes in the divertor region. Main emphasis is placed on helium (He) concentration in the ducts, as well as the relative concentration between the hydrogen isotopes (mainly in the form of diatomic molecules of H, D, and T). Measurement of the concentration of radiative gases, such as neon (Ne) and nitrogen (N{sub 2}), is also intended. Numerical modeling of the gas flow from the sampled region to the cluster of analysis sensors, through a long (∼8 m long, ∼110 mm diameter) sampling pipe originating from a pressure reducing orifice, confirm that the desired response time (∼1 s for He or D{sub 2}) is achieved with the present design.

  2. 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.

  3. Innovative design for FAST divertor compatible with remote handling, electromagnetic and mechanical analyses

    Energy Technology Data Exchange (ETDEWEB)

    Di Gironimo, Giuseppe, E-mail: giuseppe.digironimo@unina.it [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Cacace, Maurizio [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Crescenzi, Fabio [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati, Roma (Italy); Labate, Carmelenzo [CREATE, University of Naples Parthenope, Via Acton 38, 80133 Napoli (Italy); Lanzotti, Antonio [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Lucca, Flavio [LT Calcoli srl, Piazza Prinetti 26/B, 23807 Merate, LC (Italy); Marzullo, Domenico; Mozzillo, Rocco [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Pagani, Irene [LT Calcoli srl, Piazza Prinetti 26/B, 23807 Merate, LC (Italy); Ramogida, Giuseppe; Roccella, Selanna [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati, Roma (Italy); Viganò, Fabio [LT Calcoli srl, Piazza Prinetti 26/B, 23807 Merate, LC (Italy)

    2015-10-15

    Highlights: • The conceptual design of FAST divertor has been carried out through a continuous process of requirements refinement and design optimization (V-model approach), in order to achieve a design suited to the needs, RH compatible and ITER-like. • Thermal, structural and electromagnetic analyses have been performed, resulting in requirements refinement. • FAST divertor is now characterized by more realistic, reliable and functional features, satisfying thermo-mechanical capabilities and the remote handling (RH) compatibility. - Abstract: Divertor is a crucial component in Tokamaks, aiming to exhaust the heat power and particles fluxes coming from the plasma during discharges. This paper focuses on the optimization process of FAST divertor, aimed at achieving required thermo-mechanical capabilities and the remote handling (RH) compatibility. Divertor RH system final layout has been chosen between different concept solutions proposed and analyzed within the principles of Theory of Inventive Problem Solving (TRIZ). The design was aided by kinematic simulations performed using Digital Mock-Up capabilities of Catia software. Considerable electromagnetic (EM) analysis efforts and top-down CAD approach enabled the design of a final and consistent concept, starting from a very first dimensioning for EM loads. In the final version here presented, the divertor cassette supports a set of tungsten (W) actively cooled tiles which compose the inner and outer vertical targets, facing the plasma and exhausting the main part of heat flux. W-tiles are assembled together considering a minimum gap tolerance (0.1–0.5 mm) to be mandatorily respected. Cooling channels have been re-dimensioned to optimize the geometry and the layout of coolant volume inside the cassette has been modified as well to enhance the general efficiency.

  4. ATHENA simulations of divertor pump trip and loss of heat sink transients for the GSSR

    Energy Technology Data Exchange (ETDEWEB)

    Sjoeberg, A

    2001-04-01

    The ITER-FEAT Generic Site Safety Report includes evaluations of the consequences of various types of conceivable transients that may occur during operation. The transients that have to be considered in this respect are specified in the Accident Analysis Specifications document of the safety report. For the divertor primary heat transport system the ranges of transients include amongst others a trip of the main circulation pump in the divertor cooling loop as well as a loss of heat sink, both initiated at full fusion power operation. The thermal-hydraulic consequences related to the coolability of the divertor primary heat transport system components for these two transients have been evaluated and summarized in the safety report and in the current report an overview of those efforts and associated outcome is provided. The analyses have been made with the ATHENA thermal-hydraulic code using a separately developed ATHENA model of the ITER-FEAT divertor cooling system. The results from the analyses indicate that for the pump trip transient the margin against overheating of critical highly loaded parts of the divertor cassette is small but seems sufficient. In case of the loss of heat sink transient the conservative analysis reveals that the pressurizer safety valve will be opened for an extended period of time and the long term transient development indicates a risk of completely filling up the pressurizer vessel. Thus the margins against jeopardizing the integrity of the divertor cooling system with the current design are for this case small but can for a long term operation at associate conditions pose a problem.

  5. Design, integration and feasibility studies of the Tore-Supra West divertor structure

    Energy Technology Data Exchange (ETDEWEB)

    Doceul, L., E-mail: louis.doceul@cea.fr [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance Cedex (France); Bucalossi, J.; Larroque, S.; Lipa, M.; Portafaix, C.; Saille, A.; Samaille, F.; Soler, B.; Ferlay, F.; Verger, J.M. [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance Cedex (France)

    2013-10-15

    Actively cooled tungsten plasma facing components will be used in the ITER divertor. In order to fully validate such a technology (industrial manufacturing, operation with long plasma duration), the implementation of a tungsten axis symmetric divertor in the tokamak Tore-Supra is studied. With this major upgrade, so called WEST (Tungsten Environment in Steady state), Tore-Supra will be the only European tokamak able to address the problematic of long plasma discharges with an actively cooled metallic divertor.To do so, it is planned to install two symmetric divertor coils inside the vacuum vessel. This assembly, called divertor structure, is made up of two stainless steel casings containing a copper winding pack cooled by hot pressurized water (200 °C, 4 MPa). These two casings are located at the top and bottom of the vacuum vessel in order to create two magnetic X-point areas, which are protected by W-PFCs (Tungsten Plasma Facing Components) in order to extract the thermal loads. The two casing are robustly maintained together by 18 brackets in order to constitute a rigid assembly attached thanks to 12 legs (one per lower vertical port) outside the Tore{sub S}upra vacuum vessel.The paper will illustrate the technical developments performed during 2011 in order to produce a preliminary design of the Tore-Supra WEST divertor structure with a particular focus on: the mechanical design of this major component and its integration in the Tokamak, the manufacturing issues and the technical results of the feasibility studies done with industry as well as the design of a scale one coil mock up.

  6. Design and test program of a simplified divertor dummy coil structure for the WEST project

    Energy Technology Data Exchange (ETDEWEB)

    Doceul, L., E-mail: louis.doceul@cea.fr [CEA, IRFM, Saint-Paul-Lez-Durance Cedex F-13108 (France); Bucalossi, J.; Dougnac, H.; Ferlay, F.; Gargiulo, L.; Keller, D.; Larroque, S.; Lipa, M.; Pilia, A. [CEA, IRFM, Saint-Paul-Lez-Durance Cedex F-13108 (France); Portafaix, C. [ITER Organization, Route de Vinon-sur-Verdon 13115, St. Paul-lez-Durance (France); Saille, A. [CEA, IRFM, Saint-Paul-Lez-Durance Cedex F-13108 (France); Salami, M. [AVANTIS Engineering Groupe, ZI de l’Aiguille 46100, Figeac (France); Samaille, F.; Soler, B.; Thouvenin, D.; Verger, J.M.; Zago, B. [CEA, IRFM, Saint-Paul-Lez-Durance Cedex F-13108 (France)

    2013-12-15

    Highlights: • The mechanical design and integration of the divertor structure has been performed. • The design of the casing and the winding-pack has been finalized. • The coil assembly process has been validated. • The realization of a coil mock-up scale one is in progress. -- Abstract: In order to fully validate actively cooled tungsten plasma facing components (industrial fabrication, operation with long plasma duration), the implementation of a tungsten axisymmetric divertor structure in the tokamak Tore-Supra is studied. With this major upgrade, so-called WEST (Tungsten Environment in Steady state Tokamak), Tore-Supra will be able to address the problematic of long plasma discharges with a metallic divertor target. To do so, it is planned to install two symmetric divertor coils inside the vacuum vessel. This assembly, called divertor structure, is made up of two stainless steel casings containing a copper winding pack cooled by a pressurized hot water circuit (up to 180 °C, 4 MPa) and is designed to perform steady state plasma operation (up to 1000 s). The divertor structure will be a complex assembly ring of 4 m diameter representing a total weight of around 20 tons. The technical challenge of this component will be the implementation of angular sectors inside the vacuum vessel environment (TIG welding of the coil casing, induction brazing and electrical insulation of the copper winding). Moreover, this complex assembly must sustain harsh environmental conditions in terms of ultra high vacuum conditions, electromagnetical loads and electrical isolation (13 kV ground voltage) under high temperature. In order to fully validate the assembly and the performance of this complex component, the production of a scale one dummy coil is in progress. The paper will illustrate, the technical developments performed in order to finalize the design for the call for tender for fabrication. The progress and the first results of the simplified dummy coils will be also

  7. Nonlinear impact of edge localized modes on carbon erosion in the divertor of the JET tokamak.

    Science.gov (United States)

    Kreter, A; Esser, H G; Brezinsek, S; Coad, J P; Kirschner, A; Fundamenski, W; Philipps, V; Pitts, R A; Widdowson, A

    2009-01-30

    The impact of edge localized modes (ELMs) carrying energies of up to 450 kJ on carbon erosion in the JET inner divertor is assessed by means of time resolved measurements using an in situ quartz microbalance diagnostic. The inner target erosion is strongly nonlinearly dependent on the ELM energy: a single 400 kJ ELM produces the same carbon erosion as ten 150 kJ events. The ELM-induced enhanced erosion is attributed to the presence of codeposited carbon-deuterium layers on the inner divertor target, which are thermally decomposed under the impact of ELMs.

  8. One-dimensional plasma sheath model in front of the divertor plates

    Science.gov (United States)

    Tskhakaya, D.

    2017-11-01

    A new model of the stationary electrostatic plasma sheath in front of divertor plates is developed, which takes into account strong inelastic processes. Using particle-in-cell simulations and analytic estimates it is demonstrated, that the properties of the tokamak divertor plasma sheath can significantly deviate from the properties of the classical sheath model. The most significant deviations are the increased energy flux to the plates and non-monotonic potential and ion velocity profiles in the presheath. Two main reasons for these deviations are identified: strong inelastic collisionality of the plasma presheath and the presence of super-thermal plasma particles originating from the upstream scrape-off layer.

  9. Transport studies in boundary and divertor plasmas of JT-60U

    Energy Technology Data Exchange (ETDEWEB)

    Kumagai, Akira [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1999-03-01

    This thesis describes an investigation on transport of plasma, neutral particle and impurity in the boundary and divertor of the JT-60U tokamak to provide a better understanding of plasma-surface interactions and divertor physics. The asymmetry between the inboard and outboard divertor on plasma parameters (in-out asymmetry) are usually observed in tokamaks with the divertor. In this study, the in-out asymmetry was investigated under various plasma conditions and discharge parameters. The observed results were discussed with several mechanisms that can produce the in-out asymmetry. It was confirmed experimentally that the importance of each mechanism depends on the plasma parameters and discharge conditions. The current flowing in the scrape-off layer (SOL) due to the in-out asymmetry was observed. The SOL currents in the high density plasma with the occurrence of the plasma detachment were investigated for the first time in this study. The ion temperature in the divertor region is one of the most important factors for both generation and transport of impurity. However, the background ion temperature in the divertor region has not been measured in any tokamak so far. The ion temperature in the divertor region has been measured for the first time with the Doppler broading of the C{sup 3+} ion emission line. The measured temperature was analyzed by an impurity particle transport code. The code calculation showed that the measured temperature reflects the low temperature at the outside of the separatrix in the inboard region. The spectral profile of Balmer-{alpha} (D{sub {alpha}}) line emitted from the deuterium atoms reflects the velocity distribution of neutral particles by the Doppler effect and is effective for investigating the detailed neutral behavior and recycling process. The spatial variation of the D{sub {alpha}} line spectral profile in the divertor region has been measured for the first time in this study. The observed results were compared with the

  10. 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.

  11. Concept for spectrally resolved ITER divertor thermography with fibres

    Energy Technology Data Exchange (ETDEWEB)

    Reichle, R.; Thomas, E. [Association EURATOM-CEA, CEA/DSM/DRFC, CEA-Cadarache, 13 - St Paul-lez-Durance (France); Henry, S. [Ecole Internationale des Sciences du Traitement de l' Information (EISTI), 95 - Cergy-Pontoise (France); Migozzi, J.B. [JBM Optique, 92 - Sevres (France); Walker, C. [ITER-IT, p/a Max-Planck IPP, Garching bei Munchen (Germany)

    2005-07-01

    Infrared thermography on tokamak target plates under plasma impact performed at a single wavelength may be misleading because the temperature at the surface of a target is not homogeneous. Since the existing ITER divertor thermography diagnostic proposal did not include the possibility to measure at multiple wavelengths at one place, a study was performed to remedy this with a diagnostic proposal based on a fibre-optics approach. We have found an inverse matrix method to deduce the distribution of the target temperature from the spectral radiance distribution. The method seems to be robust against calibration errors and may allow to discriminate thermal radiation against the Bremsstrahlung from the plasma. Fibres are a natural choice for spectroscopic diagnostics. They minimise movements problems and they offer good possibilities for laser methods for calibration and active measurements as presumed necessary for an environment containing deposited layers and low emissivity, high reflection materials as tungsten and beryllium. Due to the high environmental temperature of 150 Celsius degrees the choice of fibres is limited. An optical study was performed to conceive an all mirror optical front-end design suitable to a fibre solution. The optical resolution of the design is about 3 mm on the targets which fits ITER requirements. About 500 fibres are necessary to exploit this fully. Looking only at the centre of the tiles (20 mm pitch) reduces the number of fibres to 100. The mirrors (and their box) and the fibres should be cooled. A detection system similar to the existing Tore Supra multi-fibre sapphire prism spectrometer coupled to a focal plane array InSb infrared camera is a viable detection solution for such a system. The logical next step is to perform radiation tests of true infrared fibres. (A.C.)

  12. Business Model Process Configurations

    DEFF Research Database (Denmark)

    Taran, Yariv; Nielsen, Christian; Thomsen, Peter

    2015-01-01

    Purpose – The paper aims: 1) To develop systematically a structural list of various business model process configuration and to group (deductively) these selected configurations in a structured typological categorization list. 2) To facilitate companies in the process of BM innovation, by develop......Purpose – The paper aims: 1) To develop systematically a structural list of various business model process configuration and to group (deductively) these selected configurations in a structured typological categorization list. 2) To facilitate companies in the process of BM innovation......, by developing (inductively) an ontological classification framework, in view of the BM process configurations typology developed. Design/methodology/approach – Given the inconsistencies found in the business model studies (e.g. definitions, configurations, classifications) we adopted the analytical induction...... method of data analysis. Findings - A comprehensive literature review and analysis resulted in a list of business model process configurations systematically organized under five classification groups, namely, revenue model; value proposition; value configuration; target customers, and strategic...

  13. Impact of nitrogen seeding on carbon erosion in the JET divertor

    NARCIS (Netherlands)

    Brezinsek, S.; Jachmich, S.; Rapp, J.; Meigs, A. G.; Nicholas, C.; O' Mullane, M.; Pospieszczyk, A.; van Rooij, G. J.

    2011-01-01

    Nitrogen has been introduced in H-mode plasmas in JET in order to study its radiation cooling capability and impact on the erosion of divertor plasma-facing components made of carbon-fiber composites (CFC). Experiments in the ionizing plasma regime with low nitrogen injection show a reduction of the

  14. Spectroscopic diagnostics for liquid lithium divertor studies on National Spherical Torus Experiment.

    Science.gov (United States)

    Soukhanovskii, V A; Roquemore, A L; Bell, R E; Kaita, R; Kugel, H W

    2010-10-01

    The use of lithium-coated plasma facing components for plasma density control is studied in the National Spherical Torus Experiment (NSTX). A recently installed liquid lithium divertor (LLD) module has a porous molybdenum surface, separated by a stainless steel liner from a heated copper substrate. Lithium is deposited on the LLD from two evaporators. Two new spectroscopic diagnostics are installed to study the plasma surface interactions on the LLD: (1) A 20-element absolute extreme ultraviolet (AXUV) diode array with a 6 nm bandpass filter centered at 121.6 nm (the Lyman-α transition) for spatially resolved divertor recycling rate measurements in the highly reflective LLD environment, and (2) an ultraviolet-visible-near infrared R=0.67 m imaging Czerny-Turner spectrometer for spatially resolved divertor D I, Li I-II, C I-IV, Mo I, D(2), LiD, CD emission and ion temperature on and around the LLD module. The use of photometrically calibrated measurements together with atomic physics factors enables studies of recycling and impurity particle fluxes as functions of LLD temperature, ion flux, and divertor geometry.

  15. Divertor tungsten tiles erosion in the region of the castellated gaps

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Wanpeng, E-mail: wangdez@dlut.edu.cn; Sang, Chaofeng; Sun, Zhenyue; Wang, Dezhen

    2016-11-01

    Highlights: • Simulation of the tungsten tiles erosion by different impurities in the divertor gap region is done by using a 2d3v Particle-In-Cell code. • High-Z impurity causes the largest erosion rate on W tile. • The peak physical sputtering erosion rate locates at the plasma-facing corners. - Abstract: Erosion of tungsten (W) is a very important issue for the future fusion device. The castellated divertor makes it more complicated due to complex geometry of the gap between the tiles. In this work, the plasma behaviors and resulting W tile erosion in the divertor tile gap region are studied by using a two dimension-in-space and three dimension-in-velocity (2d3 v) Particle-In-Cell (PIC) code. Deuterium ions (D{sup +}) and electrons are traced self-consistently in the simulation to provide the plasma background. Since there are lots of impurities, which may make a great impact on the tile erosion, in the divertor region to radiate the power, the erosion of W tile by different species are thus considered. The contributions of deuterium and impurities: Li, C, Ne, and Ar, to the W erosion, are studied under EAST conditions to show a straightforward insight. It is observed that the physical sputtering of W tile by impurities is much higher than that by the D ions, and the peak erosion region locates at the plasma-facing corners.

  16. Nuclear analysis of structural damage and nuclear heating on enhanced K-DEMO divertor model

    Science.gov (United States)

    Park, J.; Im, K.; Kwon, S.; Kim, J.; Kim, D.; Woo, M.; Shin, C.

    2017-12-01

    This paper addresses nuclear analysis on the Korean fusion demonstration reactor (K-DEMO) divertor to estimate the overall trend of nuclear heating values and displacement damages. The K-DEMO divertor model was created and converted by the CAD (Pro-Engineer™) and Monte Carlo automatic modeling programs as a 22.5° sector of the tokamak. The Monte Carlo neutron photon transport and ADVANTG codes were used in this calculation with the FENDL-2.1 nuclear data library. The calculation results indicate that the highest values appeared on the upper outboard target (OT) area, which means the OT is exposed to the highest radiation conditions among the three plasma-facing parts (inboard, central and outboard) in the divertor. Especially, much lower nuclear heating values and displacement damages are indicated on the lower part of the OT area than others. These are important results contributing to thermal-hydraulic and thermo-mechanical analyses on the divertor and also it is expected that the copper alloy materials may be partially used as a heat sink only at the lower part of the OT instead of the reduced activation ferritic-martensitic steel due to copper alloy’s high thermal conductivity.

  17. The impact of divertor detachment on carbon sources in JET L-mode discharges

    NARCIS (Netherlands)

    Brezinsek, S.; Meigs, A. G.; Jachmich, S.; Stamp, M. F.; Rapp, J.; Felton, R.; Pitts, R.A.; Philipps, V.; Huber, A.; Pugno, R.; Sergienko, G.; Pospieszczyk, A.

    2009-01-01

    Hydrocarbon injection experiments have been performed to investigate the chemical sputtering yield of carbon-fibre composites at elevated temperatures (T-surface similar or equal to 500 K) and detached plasma conditions in the JET outer divertor. A plasma scenario in L-mode with the outer

  18. Evaluation of copper alloys for fusion reactor divertor and first wall components

    DEFF Research Database (Denmark)

    Fabritsiev, S.A.; Zinkle, S.J.; Singh, B.N.

    1996-01-01

    This paper presents a critical analysis of the main factors of radiation damage limiting the possibility to use copper alloys in the ITER divertor and first wall structure. In copper alloys the most significant types of radiation damage in the proposed temperature-dose operation range are swellin...

  19. Transient thermal hydraulic modeling and analysis of ITER divertor plate system

    Energy Technology Data Exchange (ETDEWEB)

    El-Morshedy, Salah El-Din [Argonne National Laboratory, Argonne, IL (United States); Atomic Energy Authority, Cairo (Egypt)], E-mail: selmorshedy@etrr2-aea.org.eg; Hassanein, Ahmed [Purdue University, West Lafayette, IN (United States)], E-mail: hassanein@purdue.edu

    2009-12-15

    A mathematical model has been developed/updated to simulate the steady state and transient thermal-hydraulics of the International Thermonuclear Experimental Reactor (ITER) divertor module. The model predicts the thermal response of the armour coating, divertor plate structural materials and coolant channels. The selected heat transfer correlations cover all operating conditions of ITER under both normal and off-normal situations. The model also accounts for the melting, vaporization, and solidification of the armour material. The developed model is to provide a quick benchmark of the HEIGHTS multidimensional comprehensive simulation package. The present model divides the coolant channels into a specified axial regions and the divertor plate into a specified radial zones, then a two-dimensional heat conduction calculation is created to predict the temperature distribution for both steady and transient states. The model is benchmarked against experimental data performed at Sandia National Laboratory for both bare and swirl tape coolant channel mockups. The results show very good agreements with the data for steady and transient states. The model is then used to predict the thermal behavior of the ITER plasma facing and structural materials due to plasma instability event where 60 MJ/m{sup 2} plasma energy is deposited over 500 ms. The results for ITER divertor response is analyzed and compared with HEIGHTS results.

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

    Science.gov (United States)

    Domalapally, Phani; Di Caro, Marco

    2017-11-01

    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.

  1. Divertor heat load in ASDEX Upgrade L-mode in presence of external magnetic perturbation

    Science.gov (United States)

    Faitsch, M.; Sieglin, B.; Eich, T.; Herrmann, A.; Suttrop, W.; the ASDEX Upgrade Team

    2017-09-01

    Power exhaust is one of the major challenges for a future fusion device. Applying a non-axisymmetric external magnetic perturbation is one technique that is studied in order to mitigate or suppress large edge localized modes which accompany the high confinement regime in tokamaks. The external magnetic perturbation induces breaking in the axisymmetry of a tokamak and leads to a 2D heat flux pattern on the divertor target. The 2D heat flux pattern at the outer divertor target is studied on ASDEX Upgrade in stationary L-mode discharges. The amplitude of the 2D characteristic of the heat flux depends on the alignment between the field lines at the edge and the vacuum response of the applied magnetic perturbation spectrum. The 2D characteristic reduces with increasing density. The increasing divertor broadening, S, with increasing density is proposed as the main actuator. This is supported by a generic model using field line tracing and the vacuum field approach that is in quantitative agreement with the measured heat flux. The perturbed heat flux, averaged over a full toroidal rotation of the magnetic perturbation, is identical to the non-perturbed heat flux without magnetic perturbation. The transport qualifiers, power fall-off length {λ }q and divertor broadening, S, are the same within the uncertainty compared to the unperturbed reference. No additional cross field transport is observed.

  2. Optical design study for divertor observation at the stellarator W7-X

    NARCIS (Netherlands)

    König, R.; Hildebrandt, D.; Hübner, T.; Klinkhamer, J.F.F.; Moddemeijer, K.; Vliegenthart, W.A.

    2006-01-01

    The stellarator W7-X will be capable of running in a quasicontinuous operating mode with 10 MW of electron cyclotron heating (ECRH) heating for 30 min, the duration only being limited by the capacity of the available cooling reservoir. The integrated ten discrete water cooled divertor modules need

  3. Flow and mixing of gas in cylinder of a stratified charge engine with two intake valves. Effects of late closing valve timing and intake port configurations; Kyuki nibenshiki sojo kyuki engine no cylinder nai gas ryudo to kongo. Osotoji valve timing oyobi port keijo ni yoru eikyo

    Energy Technology Data Exchange (ETDEWEB)

    Charoenphonphanich, C.; Niwa, H.; Ennoji, H.; Iijima, T. [Tokai University, Tokyo (Japan)

    1997-10-01

    A numerical analysis of the flow and mixing of rich mixture and air inducted into the cylinder through each of the two intake ports of a stratified charge engine have been carried out. Numerical calculations were performed by finite volume method for three types of the intake port configurations: inverse V type, parallel type and V type and two types of valve timing; conventional and late closing (Miller cycle). Velocity field, turbulent kinetic energy and distribution of mixture concentration in the cylinder were examined. 3 refs., 10 figs.

  4. Manufacturing monitoring and mock-ups validation of the WEST divertor structure and coils

    Energy Technology Data Exchange (ETDEWEB)

    Doceul, Louis, E-mail: louis.doceul@cea.fr [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Bucalossi, Jérôme; Decool, Patrick; Dougnac, Hubert; Ferlay, Fabien; Gargiulo, Laurent; Keller, Delphine; Larroque, Sébastien; Lipa, Manfred; Martino, Patrick; Pilia, Arnaud; Poli, Serge [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Portafaix, Christophe [ITER Organization, Route de Vinon-sur-Verdon, 13115 Saint-Paul-lez-Durance (France); Saille, Alain [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Salami, Michael [AVANTIS Engineering Groupe, ZI de l’Aiguille, 46100 Figeac (France); Samaille, Frank; Soler, Bernard; Thouvenin, Didier; Verger, Jean-Marc [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Voyard, Olivier [CNIM, ZI de Brégaillon, 83500 La Seyne-sur-Mer (France); and others

    2015-10-15

    Highlights: • The mechanical design and integration of the divertor structure have been performed. • The design of the casing and the winding-pack has been optimized. • The coil assembly process has been assessed. • The realization of a coil mock-up scale one is scheduled. - Abstract: In order to fully validate “ITER-like” actively water cooled tungsten plasma facing units, the implementation of an axisymmetric divertor structure in the Tokamak Tore-Supra has been studied. With this major upgrade, the so-called WEST (Tungsten Environment in Steady state Tokamak), Tore-Supra will be able to address the issues of long plasma discharges using a tungsten divertor based on monoblock targets. The divertor structure and coils assembly are made up of two stainless steel casings containing a copper winding pack cooled by a pressurized hot water circuit (up to 180 °C, 4 MPa) in which a total divertor current of up to 16 × 13 kA is circulating in steady state. The conductor is electrically insulated and wedged inside the casing in order to be mechanically protected. The divertor which is designed to perform steady state plasma operation (up to 1000 s), must sustain harsh environmental conditions in terms of ultra light vacuum conditions, electromagnetical loads and electrical insulation (5 kV ground voltage) under high temperature (180 °C). Therefore, a feasibility study of such a complex structure has been performed. It implied activities on a scale one dummy coil, such as installation, assembly issues and representative tests (electric, thermal and hydraulic). The manufacturing of the divertor structure, which is a large assembly of 4-m diameter representing a total weight of around 20 tonnes, started in the second half of 2013 and is expected to be delivered by the end of 2014. The paper will illustrate the technical developments and tests performed during 2013 and beginning of 2014 in order to fully validate the design concept before the industrial phase

  5. In-pile thermocycling testing and post-test analysis of beryllium divertor mockups

    Energy Technology Data Exchange (ETDEWEB)

    Giniatulin, R.; Mazul, I. [Efremov Inst., St. Petersburg (Russian Federation); Melder, R.; Pokrovsky, A.; Sandakov, V.; Shiuchkin, A.

    1998-01-01

    The main damaging factors which impact the ITER divertor components are neutron irradiation, cyclic surface heat loads and hydrogen environment. One of the important questions in divertor mockups development is the reliability of beryllium/copper joints and the beryllium resistance under neutron irradiation and thermal cycling. This work presents the experiment, where neutron irradiation and thermocyclic heat loads were applied simultaneously for two beryllium/copper divertor mockups in a nuclear reactor channel to simulate divertor operational conditions. Two mockups with different beryllium grades were mounted facing each other with the tantalum heater placed between them. This device was installed in the active zone of the nuclear reactor SM-2 (Dimitrovgrad, Russia) and the tantalum block was heated by neutron irradiation up to a high temperature. The main part of the heat flux from the tantalum surface was transported to the beryllium surface through hydrogen, as a result the heat flux loaded two mockups simultaneously. The mockups were cooled by reactor water. The device was lowered to the active zone so as to obtain the heating regime and to provide cooling lifted. This experiment was performed under the following conditions: tantalum heater temperature - 1950degC; hydrogen environment -1000 Pa; surface heat flux density -3.2 MW/m{sup 2}; number of thermal cycles (lowering and lifting) -101; load time in each cycle - 200-5000 s; dwell time (no heat flux, no neutrons) - 300-2000 s; cooling water parameters: v - 1 m/s, Tin - 50degC, Pin - 5 MPa; neutron fluence -2.5 x 10{sup 20} cm{sup -2} ({approx}8 years of ITER divertor operation from the start up). The metallographic analysis was performed after experiment to investigate the beryllium and beryllium/copper joint structures, the results are presented in the paper. (author)

  6. Preparation of 3D Printed Divertor Mock-up Design and Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Park, Sung Dae; Kim, Dong Jun; Kim, Suk Kwon; Lee, Eo Hwak [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The divertor for fusion reactor is known to be able to remove the extreme heat flux up to 10 MW/m2 and the various type of divertors have been developed for enhancing the heat transfer such as hypervapotron, twisted tape insertion, screwed tube, and so on. In order to overcome this limitation, 3D printing method is considered to be used in the fusion reactor divertor design in present study. With the advantages of the 3D printing, the various shapes of the inner divertor cooling tube are investigated to enhance the turbulence of coolant and to reduce the pressure drop. The metallic powder of the fusion reactor candidate material is produced as the preliminary step for using in 3D printer. The material is a reduced activation ferritic-matensitic steel named as ARAA (Advanced Reduced Activation Alloy) which have been independently developed in Korea. Gas atomization method was used to make the spherical particles with average diameter of 100 μm. Several candidates were presented to achieve the excellent heat removal capacity and the low pressure drop. Thermal-hydraulic analysis was performed to confirm the effects of the inner cooling tube geometry with a conventional CFD code, ANSYS-CFX v14.5. The modified screw type called as a rail type twisted tube was presented through the optimization process. This complicated tube could be made by 3D printing technology. (metallic powder). Thermal-hydraulic analysis was conducted to compare the 3 type geometric divertor. A rail type twisted tube has good heat transfer performance in comparison with a conventional twisted tube. The pressure drop of a rail type twisted tube was reduced about 36% compared with a conventional twisted tube.

  7. Recent progress of divertor simulation research using the GAMMA 10/PDX tandem mirror

    Science.gov (United States)

    Nakashima, Y.; Ichimura, K.; Islam, M. S.; Sakamoto, M.; Ezumi, N.; Hirata, M.; Ichimura, M.; Ikezoe, R.; Imai, T.; Kariya, T.; Katanuma, I.; Kohagura, J.; Minami, R.; Numakura, T.; Yoshikawa, M.; Iijima, T.; Islam, M. M.; Nojiri, K.; Shimizu, K.; Terakado, A.; Togo, S.; Asakura, N.; Fukumoto, M.; Hatayama, A.; Hirooka, Y.; Kado, S.; Kubo, H.; Masuzaki, S.; Matsuura, H.; Nakano, T.; Nagata, S.; Nishino, N.; Ohno, N.; Sagara, A.; Sawada, K.; Shoji, M.; Tonegawa, A.; Ueda, Y.

    2017-11-01

    This paper describes the recent progress in divertor simulation research using the GAMMA 10/PDX tandem mirror towards the development of divertors in fusion reactors. During a plasma flow generation experiment in the end cell of the GAMMA 10/PDX, ICRF heating in the anchor cell successfully extended the particle flux up to 3.3  ×  1023 m2 s-1. Superimposing the short pulse of the ECH also attained a maximum heat flux of ~30 MW m-2. We have succeeded in achieving and characterizing the detachment of the high-temperature plasma, which is equivalent to the SOL plasma of tokamaks, by using the divertor simulation experimental module (D-module) in the GAMMA 10/PDX end cell, in spite of using a linear device with a short magnetic field line connection length. Various gases (Ar, Xe, Ne and N2) are examined to evaluate the effect of radiation cooling against the plasma flow at the MW m-2 level in the divertor simulation region and the following results are obtained: (i) Xe gas was most effective in the reduction of heat and particle fluxes (1%, 3%, respectively) and has a stronger effect on electron cooling (down to ~1.6 eV) in the used gas species. (ii) Ne gas was less effective. On the other hand, (iii) N2 gas showed more favorable effects than Ar in the lower pressure range. These results will contribute to the progress in detached plasma operation and in clarifying the radiation cooling mechanism towards the development of future divertors.

  8. HLT configuration management system

    CERN Document Server

    Daponte, Vincenzo

    2015-01-01

    The CMS High Level Trigger (HLT) is implemented running a streamlined version of the CMS offline reconstruction software running on thousands of CPUs. The CMS software is written mostly in C++, using Python as its configuration language through an embedded CPython interpreter. The configuration of each process is made up of hundreds of modules, organized in sequences and paths. As an example, the HLT configurations used for 2011 data taking comprised over 2200 different modules, organized in more than 400 independent trigger paths. The complexity of the HLT configurations and the large number of configuration produced require the design of a suitable data management system. The present work describes the designed solution to manage the considerable number of configurations developed and to assist the editing of new configurations. The system is required to be remotely accessible and OS-independent as well as easly maintainable easy to use. To meet these requirements a three-layers architecture has been choose...

  9. Conceptualizing Embedded Configuration

    DEFF Research Database (Denmark)

    Oddsson, Gudmundur Valur; Hvam, Lars; Lysgaard, Ole

    2006-01-01

    Installing and servicing complex electromechanical systems is more tedious than is necessary. By putting the product knowledge into the product itself, which then would allow automation in constructing the product from modules, could solve that. It would support personnel in aftersales installation...... and services. The general idea can be named embedded configuration. In this article we intend to conceptualize embedded configuration, what it is and is not. The difference between embedded configuration, sales configuration and embedded software is explained. We will look at what is needed to make embedded...... configuration systems. That will include requirements to product modelling techniques. An example with consumer electronics will illuminate the elements of embedded configuration in settings that most can relate to. The question of where embedded configuration would be relevant is discussed, and the current...

  10. 2010 OFES Joint Research TargetDivertor Heat Flux Profile Width Final Report DIII-D Contribution

    Energy Technology Data Exchange (ETDEWEB)

    Lasnier, C. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Makowski, M. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Boedo, J. A. [Univ. of California, San Diego, CA (United States); Hill, D. N. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Leonard, A. W. [General Atomics, San Diego, CA (United States); Porter, G. D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Rensink, M. E. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Watkins, J. G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2010-09-17

    Conduct experiments on major fusion facilities to improve understanding of the heat transport in the tokamak scrape-off layer (SOL) plasma, strengthening the basis for projecting divertor conditions in ITER. In FY10, FES will measure the divertor heat flux profiles and plasma characteristics in the tokamak scrape-off layer in multiple devices to investigate the underlying thermal transport processes. The unique characteristics of C-Mod, DIII-D, and NSTX will enable collection of data over a broad range of SOL and divertor parameters (e.g., collisionality, beta, parallel heat flux, and divertor geometry). Regimes similar to the ITER operating scenarios will be among those studied and characterized. Coordinated experiments using common analysis methods will generate a data set that will be compared with theory and simulation.

  11. Software configuration management

    CERN Document Server

    Keyes, Jessica

    2004-01-01

    Software Configuration Management discusses the framework from a standards viewpoint, using the original DoD MIL-STD-973 and EIA-649 standards to describe the elements of configuration management within a software engineering perspective. Divided into two parts, the first section is composed of 14 chapters that explain every facet of configuration management related to software engineering. The second section consists of 25 appendices that contain many valuable real world CM templates.

  12. Experimental simulation and numerical modeling of vapor shield formation and divertor material erosion for ITER typical plasma disruptions

    Energy Technology Data Exchange (ETDEWEB)

    Wuerz, H. [Kernforschungszentrum Karlsruhe, INR, Postfach 36 40, D-76021 Karlsruhe (Germany); Arkhipov, N.I. [Troitsk Institute for Innovation and Fusion Research, 142092 Troitsk (Russian Federation); Bakhtin, V.P. [Troitsk Institute for Innovation and Fusion Research, 142092 Troitsk (Russian Federation); Konkashbaev, I. [Troitsk Institute for Innovation and Fusion Research, 142092 Troitsk (Russian Federation); Landman, I. [Troitsk Institute for Innovation and Fusion Research, 142092 Troitsk (Russian Federation); Safronov, V.M. [Troitsk Institute for Innovation and Fusion Research, 142092 Troitsk (Russian Federation); Toporkov, D.A. [Troitsk Institute for Innovation and Fusion Research, 142092 Troitsk (Russian Federation); Zhitlukhin, A.M. [Troitsk Institute for Innovation and Fusion Research, 142092 Troitsk (Russian Federation)

    1995-04-01

    The high divertor heat load during a tokamak plasma disruption results in sudden evaporation of a thin layer of divertor plate material, which acts as vapor shield and protects the target from further excessive evaporation. Formation and effectiveness of the vapor shield are theoretically modeled and are experimentally analyzed at the 2MK-200 facility under conditions simulating the thermal quench phase of ITER tokamak plasma disruptions. ((orig.)).

  13. Airport Configuration Prediction Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Airport configuration is a primary factor in various airport characteristics such as arrival and departure capacities and terminal area traffic patterns. These...

  14. Ansible configuration management

    CERN Document Server

    Hall, Daniel

    2013-01-01

    Ansible Configuration Management"" is a step-by-step tutorial that teaches the use of Ansible for configuring Linux machines.This book is intended for anyone looking to understand the basics of Ansible. It is expected that you will have some experience of how to set up and configure Linux machines. In parts of the book we cover configuration files of BIND, MySQL, and other Linux daemons, therefore a working knowledge of these would be helpful but are certainly not required.

  15. Gyrokinetic projection of the divertor heat-flux width from present tokamaks to ITER

    Science.gov (United States)

    Chang, C. S.; Ku, S.; Loarte, A.; Parail, V.; Köchl, F.; Romanelli, M.; Maingi, R.; Ahn, J.-W.; Gray, T.; Hughes, J.; LaBombard, B.; Leonard, T.; Makowski, M.; Terry, J.

    2017-11-01

    The XGC1 edge gyrokinetic code is used to study the width of the heat-flux to divertor plates in attached plasma condition. The flux-driven simulation is performed until an approximate power balance is achieved between the heat-flux across the steep pedestal pressure gradient and the heat-flux on the divertor plates. The simulation results compare well against the empirical scaling λ q \\propto 1/BPγ obtained from present tokamak devices, where λ q is the divertor heat-flux width mapped to the outboard midplane, γ  =  1.19 as found by Eich et al (2013 Nucl. Fusion 53 093031), and B P is the magnitude of the poloidal magnetic field at the outboard midplane separatrix surface. This empirical scaling predicts λ q  ≲  1 mm when extrapolated to ITER, which would require operation with very high separatrix densities (n sep/n Greenwald  >  0.6) (Kukushkin et al 2013 J. Nucl. Mater. 438 S203) in the Q  =  10 scenario to achieve semi-detached plasma operation and high radiative fractions for acceptable divertor power fluxes. Using the same simulation code and technique, however, the projected λ q for ITER’s model plasma is 5.9 mm, which could be suggesting that operation in the ITER Q  =  10 scenario with acceptable divertor power loads may be obtained over a wider range of plasma separatrix densities and radiative fractions. The physics reason behind this difference is, according to the XGC1 results, that while the ion magnetic drift contribution to the divertor heat-flux width is wider in the present tokamaks, the turbulent electron contribution is wider in ITER. Study will continue to verify further this important projection. A high current C-Mod discharge is found to be in a mixed regime: While the heat-flux width by the ion neoclassical magnetic drift is still wider than the turbulent electron heat-flux width, the heat-flux magnitude is dominated by the narrower electron heat-flux.

  16. A mature industrial solution for ITER divertor plasma facing components: hypervapotron cooling concept adapted to Tore Supra flat tile technology

    Energy Technology Data Exchange (ETDEWEB)

    Escourbiac, F.; Missirlian, M.; Schlosser, J. [Association EURATOM-CEA Cadarache, Departement de Recherches sur la Fusion Controlee, 13 - Saint Paul lez Durance (France); Bobin-Vastra, I. [AREVA Centre Technique de Framatome, 71 - Le Creusot (France); Kuznetsov, V. [Efremov Institute, Doroga na Metallostroy, St. Petersburg (Russian Federation); Schedler, B. [Plansee AG, Reutte (Austria)

    2004-07-01

    The use of flat tile technology to handle heat fluxes in the range of 20 MW/m{sup 2} with components relevant for fusion experiment applications is technically possible with the hypervapotron cooling concept. This paper deals with recent high heat flux performances operated with success on 2 identical mock-ups, based on this concept, that were tested in 2 different electron gun facilities. Each mock-up consisted of a CuCrZr heat sink armored with 25 flat tiles of the 3D carbon fibre composite material SEPcarb NS31 assembled with pure copper by active metal casting (AMC). The AMC tiles were electron beam welded on the CuCrZr bar, fins and slots on the neutral beam JET design were machined into the bar, then the bar was closed with a thick CuCrZr rear plug including hydraulic connections then the bar was electron beam welded onto the sidewalls. The testing results show that full ITER design specifications were achieved with margins, the critical heat flux limit was even higher than 30 MW/m{sup 2}. These results highlight the high potential of this technology for ITER divertor application.

  17. Experimental study of the recombination of a drifting low temperature plasma in the divertor simulator Mistral-B

    Energy Technology Data Exchange (ETDEWEB)

    Brault, C.; Escarguel, A.; Koubiti, M.; Stamm, R.; Pierre, Th.; Quotb, K.; Guyomarc' h, D. [Universite de Provence, Lab. PIIM, CNRS, 13 - Marseille (France)

    2004-07-01

    In a new divertor simulator, an ultra-cold (T{sub e} < 1 eV) high density recombining magnetized laboratory plasma is studied using probes, spectroscopic measurements, and ultra-fast imaging of spontaneous emission. The Mistral-B device consists in a linear high density magnetized plasma column. The ionizing electrons originate from a large cathode array located in the fringing field of the solenoid. The ionizing electrons are focused in a 3 cm diameter hole at the entrance of the solenoid. The typical plasma density on the axis is close to 2.10{sup 18} m{sup -3}. The collector is segmented into two plates and a transverse electric field is applied through a potential difference between the plates. The Lorentz force induces the ejection of a very-low temperature plasma jet in the limiter shadow. The characteristic convection time and decay lengths have been obtained with an ultra-fast camera. The study of the atomic physics of the recombining plasma allows to understand the measured decay time and to explain the emission spectra. (authors)

  18. Improved structural strength and lifetime of monoblock divertor targets by using doped tungsten alloys under cyclic high heat flux loading

    Science.gov (United States)

    Nogami, S.; Guan, W. H.; Hattori, T.; James, K.; Hasegawa, A.

    2017-12-01

    The divertor is one of the most important components of a fusion reactor, which performs the function of the removal of waste material from fusion plasma. Because the divertor is subjected to cyclic high heat flux loading up to about 20 MW m-2 induced by the plasma, the plasma facing material of the divertor should exhibit good thermo-mechanical properties. In this work, the possibility of improving the structural strength and the lifetime of fusion reactor divertors by using K-doped W and K-doped W-3%Re as plasma facing material instead of ordinary pure W was evaluated by thermo-mechanical finite element analysis (FEA). These materials have been developed for divertor applications in Japan and show higher recrystallization temperature and strength than pure W. The results of the present study indicated that K-doped W and K-doped W-3%Re render lower applied strain to the divertor and longer fatigue life of the plasma facing material. The evaluation results regarding the macro-crack formation life based on the FEA analyses indicated the possibility of an extension of the fatigue life by using K-doped W and K-doped W-3%Re.

  19. Determination of volumetric plasma parameters from spectroscopic N II and N III line ratio measurements in the ASDEX Upgrade divertor

    Science.gov (United States)

    Henderson, S. S.; Bernert, M.; Brezinsek, S.; Carr, M.; Cavedon, M.; Dux, R.; Lipschultz, B.; O’Mullane, M. G.; Reimold, F.; Reinke, M. L.; The ASDEX Upgrade Team; The MST1 Team

    2018-01-01

    The diagnosis of tokamak divertor plasmas is limited in the ability to understand the behaviour and role of impurities, central to the overall understanding of how the divertor plasma can be utilised to control the power exhaust. New methods have been developed to extract the N concentration as well as plasma characteristics; the use of three visible N II lines has been shown to provide a unique solution of the background plasma density and temperature. Those techniques are applied to data from two sightlines sampling horizontally across the outer divertor plasma. The plasma densities obtained from the N II line ratios during a scan of the divertor temperature in a partially detached H-mode plasma suggest that, as the temperature drops, the plasma density decreases further up the divertor leg while closer to the strike point the plasma density increases. The former is consistent with the emission zone moving from the private flux region into the scrape-off-layer plasma, and therefore sampling two different density regimes, while the latter is consistent with electron pressure conservation along a field line. With an approximate model of the length of the emission region, the N II divertor concentration is calculated in this discharge to be  ≈5-25% . The single N III line ratio measurement available within the same spectral range is dependent on temperature and density and therefore cannot provide a unique solution of both.

  20. PIV Logon Configuration Guidance

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Glen Alan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-03-04

    This document details the configurations and enhancements implemented to support the usage of federal Personal Identity Verification (PIV) Card for logon on unclassified networks. The guidance is a reference implementation of the configurations and enhancements deployed at the Los Alamos National Laboratory (LANL) by Network and Infrastructure Engineering – Core Services (NIE-CS).

  1. Configuration by Modularisation

    DEFF Research Database (Denmark)

    Riitahuhta, Asko; Andreasen, Mogens Myrup

    1998-01-01

    Management and i Configuration Management the most important means is Modularisation.The goal of this paper is to show Configuration Management as a contribution to the Mass Customisation and Modularisation as a contribution to the industrialisation of the design area [Andreasen 1997]. A basic model...

  2. An Implicit Monte Carlo Method for Simulation of Impurity Transport in Divertor Plasma

    Science.gov (United States)

    Suzuki, Akiko; Takizuka, Tomonori; Shimizu, Katsuhiro; Hayashi, Nobuhiko; Hatayama, Akiyoshi; Ogasawara, Masatada

    1997-02-01

    A new "implicit" Monte Carlo (IMC) method has been developed to simulate ionization and recombination processes of impurity ions in divertor plasmas. The IMC method takes into account many ionization and recombination processes during a time step Δ t. The time step is not limited by a condition, Δ t≪ τ min(τ min; the minimum characteristic time of atomic processes), which is forced to be adopted in conventional Monte Carlo methods. We incorporate this method into a one-dimensional impurity transport model. In this transport calculation, impurity ions are followed with the time step about 10 times larger than that used in conventional methods. The average charge state of impurities, , and the radiative cooling rate, L( Te), are calculated at the electron temperature Tein divertor plasmas. These results are compared with thosed obtained from the simple noncoronal model.

  3. Fatigue strength of tungsten-copper duplex structures for divertor plates

    Science.gov (United States)

    Seki, M.; Horie, T.; Tone, T.; Nagata, K.; Kitamura, K.; Shibutani, Y.; Shibui, M.; Araki, T.

    1988-07-01

    A tungsten-copper duplex structure is specified in a conceptual design of the Japan Fusion Experimental Reactor (FER). The evaluation of the fatigue and creep life of the interface region between tungsten and copper is essential for design of the divertor plate. Fatigue crack initiation life and crack propagation behavior at room temperature and 200°C were measured for fully-annealed OFHC copper and for tungsten-OFHC copper joints brazed with amorphous nickel-base filler metal. The debonding fatigue strength for the brazed joints was relatively high, but less than that of the copper. Fatigue crack growth rates in the braze layer was approximately similar to that of the copper. Fatigue lives were estimated for the divertor plate with small defects, and a method for analyzing the apparent K- values of interface cracks was presented.

  4. Analytic Criteria for Power Exhaust in Divertors due to Impurity Radiation

    CERN Document Server

    Post, D; Perkins, F W; Nevins, W

    1995-01-01

    Present divertor concepts for next step experiments such ITER and TPX rely upon impurity and hydrogen radiation to transfer the energy from the edge plasma to the main chamber and divertor chamber walls. The efficiency of these processes depends strongly on the heat flux, the impurity species, and the connection length. Using a database for impurity radiation rates constructed from the ADPAK code package, we have developed criteria for the required impurity fraction, impurity species, connection length and electron temperature and density at the mid-plane. Consistent with previous work, we find that the impurity radiation from coronal equilibrium rates is, in general, not adequate to exhaust the highest expected heating powers in present and future experiments. As suggested by others, we examine the effects of enhancing the radiation rates with charge exchange recombination and impurity recycling, and develop criteria for the minimum neutral fraction and impurity recycling rate that is required to exhaust a s...

  5. Non-destructive testing of CFC monoblock divertor mock-ups

    Science.gov (United States)

    Ezato, K.; Dairaku, M.; Taniguchi, M.; Sato, K.; Akiba, M.

    2002-12-01

    Non-destructive examination (NDE) methods for joint interfaces between different materials in high heat flux (HHF) components of divertor should be urgently developed to assure quality and reliability of joining techniques. The purpose of this work is to demonstrate the ability of using ultrasonic wave and thermography NDE techniques to detect the defect in the joining interface (joint defect) of divertor mock-ups with carbon-fiber reinforced carbon monoblock armor tiles brazed on a copper cooling tube. The results of both NDEs are benchmarked with HHF tests and cross-sectional observation of the mock-up to correlate the joint defect size detected with NDEs to the thermal response of the mock-up with initial joint defects. From the results of the HHF tests and the cross-sectional observations, it can be concluded that both NDE techniques have sufficient accuracy to predict the surface temperature of the HHF components.

  6. Retention property of deuterium for fuel recovery in divertor by using hydrogen storage material

    Science.gov (United States)

    Mera, Saori; Tonegawa, Akira; Matsumura, Yoshihito; Sato, Kohnosuke; Kawamura, Kazutaka

    2014-10-01

    Magnetic confinement fusion reactor by using Deuterium and Tritium of hydrogen isotope as fuels is suggested as one of the future energy source. Most fuels don't react and are exhausted out of fusion reactor. Especially, Tritium is radioisotope and rarely exists in nature, so fuels recovery is necessary. This poster presentation will explain about research new fuel recovery method by using hydrogen storage materials in divertor simulator TPD-Sheet IV. Samples are tungsten coated with titanium; tungsten of various thickness, and titanium films deposited by ion plating on tungsten substrates. The sample surface temperature is measured by radiation thermometer. Retention property of deuterium after deuterium plasma irradiation was examined with thermal desorption spectroscopy (TDS). As a result, the TDS measurement shows that deuterium is retained in titanium. Therefore, Titanium as a hydrogen storage material expects to be possible to use separating and recovering fuel particles in divertor.

  7. The simulation of the ITER divertor plates erosion in stationary plasma

    Energy Technology Data Exchange (ETDEWEB)

    Antonov, N.V.; Muksunov, A.M.; Nikiforov, V.A.; Petrov, V.B.; Pistunovich, V.I.; Khripunov, B.I.; Shapkin, V.V. (Kurchatov Institute of Atomic Energy, Moscow (Russian Federation))

    1991-01-01

    The problem of the divertor development for the ITER is put by very high heat loads (up to 15 MW/m[sup 2]) and high erosion rates of the plates structures. The plates are considered to be under floating potential (negative). Ions moving to the plates are accelerated by the potential difference U=-3.5T[sub e] thus causing destruction of the material if their energy is higher than the threshold value. One can reduce the energy of ions near the plates by lowering T[sub e]. This is possible in particular conditions of particle recycling when the gas pressure near the plates is increased (gas divertor). In this report is considered the possibility to diminish erosion of the material by applying voltage between the reactor chamber and the plates and some aspects of plasma flow interaction with the wall. (author) 4 refs., 4 figs.

  8. Upgrade of the infrared camera diagnostics for the JET ITER-like wall divertor.

    Science.gov (United States)

    Balboa, I; Arnoux, G; Eich, T; Sieglin, B; Devaux, S; Zeidner, W; Morlock, C; Kruezi, U; Sergienko, G; Kinna, D; Thomas, P D; Rack, M

    2012-10-01

    For the new ITER-like wall at JET, two new infrared diagnostics (KL9B, KL3B) have been installed. These diagnostics can operate between 3.5 and 5 μm and up to sampling frequencies of ∼20 kHz. KL9B and KL3B image the horizontal and vertical tiles of the divertor. The divertor tiles are tungsten coated carbon fiber composite except the central tile which is bulk tungsten and consists of lamella segments. The thermal emission between lamellae affects the surface temperature measurement and therefore KL9A has been upgraded to achieve a higher spatial resolution (by a factor of 2). A technical description of KL9A, KL9B, and KL3B and cross correlation with a near infrared camera and a two-color pyrometer is presented.

  9. End loss analyzer system for measurements of plasma flux at the C-2U divertor electrode

    Energy Technology Data Exchange (ETDEWEB)

    Griswold, M. E., E-mail: mgriswold@trialphaenergy.com; Korepanov, S.; Thompson, M. C. [Tri Alpha Energy, P.O. Box 7010, Rancho Santa Margarita, California 92688 (United States)

    2016-11-15

    An end loss analyzer system consisting of electrostatic, gridded retarding-potential analyzers and pyroelectric crystal bolometers was developed to characterize the plasma loss along open field lines to the divertors of C-2U. The system measures the current and energy distribution of escaping ions as well as the total power flux to enable calculation of the energy lost per escaping electron/ion pair. Special care was taken in the construction of the analyzer elements so that they can be directly mounted to the divertor electrode. An attenuation plate at the entrance to the gridded retarding-potential analyzer reduces plasma density by a factor of 60 to prevent space charge limitations inside the device, without sacrificing its angular acceptance of ions. In addition, all of the electronics for the measurement are isolated from ground so that they can float to the bias potential of the electrode, 2 kV below ground.

  10. 1D fluid regime of plasma-neutral interaction and divertor detachment

    Energy Technology Data Exchange (ETDEWEB)

    Soboleva, T.K. [UNAM, Mexico D.F. (Mexico). Inst. de Ciencias Nucleares]|[Kurchatov Inst. of Atomic Energy, Moscow (Russian Federation); Krasheninnikov, S.I. [Massachusetts Inst. of Tech., Cambridge, MA (United States). Plasma Fusion Center]|[Kurchatov Inst. of Atomic Energy, Moscow (Russian Federation)

    1996-08-01

    We show that self consistent decrease of both plasma flux and neutral ionization in current tokamaks is only possible when neutrals can be treated in a short mean free path approximation. We investigate these fluid regimes of plasma-neutral interaction with 1D fluid equations employing a neutral viscosity term to treat the neutral interaction with the divertor plate. We have found that plasma flux onto the target starts to decrease at a very low heat flux coming into the hydrogen recycling region, when the temperature near the target drops below 1 eV, which seems lower than observed in the experiments. We conclude that the neutral-neutral collisions, 2D effects of plasma-neutral interaction, and plasma recombination processes can play a very important role in divertor plasma detachment. (orig.)

  11. Divertor-localized fluctuations in NSTX-U L-mode discharges

    Science.gov (United States)

    Scotti, Filippo; Soukhanovskii, V. A.; Zweben, S.; Myra, J.; Baver, D.; Sabbagh, S. A.

    2017-10-01

    The 3-D structure of divertor turbulence is characterized in NSTX-U by means of fast camera imaging. Edge and divertor turbulence can be important in determining the heat flux width in fusion devices. Field-aligned filaments are found on the divertor legs via imaging of C III and D- α emission in NBI-heated diverted L-mode discharges, similar to observations in Alcator C-Mod and MAST. These flute-like fluctuations of up to 10-20% in RMS/mean are radially localized around the separatrix and limited to the region below the X-point. Poloidal and parallel correlation lengths are a few cm (10-50ρi) and several meters, respectively. For the outer leg filaments, poloidal correlation lengths decrease along the leg away from the strike point and typical effective toroidal mode numbers are in the range of 10-20. Opposite toroidal rotation is observed for inner (co-current rotation) and outer leg (counter-current rotation) filaments with apparent poloidal propagation of 1 km/s. The poloidal motion of outer leg filaments is opposite to the one typically observed for NSTX upstream blobs in the scrape-off layer. The shape, dynamics and absence of correlation with upstream turbulence suggest that these fluctuations are generated and localized in the divertor region. Supported by US DOE DE-AC52-07NA27344, DE-AC02-09CH11466, DE-FG02- 02ER54678, DE-FG02-99ER54524.

  12. Expanding the role of impurity spectroscopy for investigating the physics of high-Z dissipative divertors

    OpenAIRE

    M.L. Reinke; Meigs, A.; Delabie, E; Mumgaard, R.; Reimold, F.; Potzel, S; Bernert, M.; Brunner, D.; Canik, J.; Cavedon, M.; Coffey, I.; Edlund, E.; J. Harrison; LaBombard, B.; Lawson, K.

    2017-01-01

    New techniques that attempt to more fully exploit spectroscopic diagnostics in the divertor and pedestal region during highly dissipative scenarios are demonstrated using experimental results from recent low-Z seeding experiments on Alcator C-Mod, JET and ASDEX Upgrade. To exhaust power at high parallel heat flux, q∥ > 1 GW/m2, while minimizing erosion, reactors with solid, high-Z plasma facing components (PFCs) are expected to use extrinsic impurity seeding. Due to transport and atomic physi...

  13. Expanding the role of impurity spectroscopy for investigating the physics of high-Z dissipative divertors

    OpenAIRE

    M.L. Reinke; Meigs, A.; Delabie, E; Mumgaard, R.; Reimold, F.; Potzel, S; Bernert, M.; Brunner, D.; Canik, J.; Cavedon, M.; Coffey, I.; Edlund, E.; J. Harrison; LaBombard, B.; Lawson, K.

    2016-01-01

    New techniques that attempt to more fully exploit spectroscopic diagnostics in the divertor and pedestal region during highly dissipative scenarios are demonstrated using experimental results from recent low-Z seeding experiments on Alcator C-Mod, JET and ASDEX Upgrade. To exhaust power at high parallel heat flux, q ∥ > 1 GW/m2, while minimizing erosion, reactors with solid, high-Z plasma facing components (PFCs) are expected to use extrinsic impurity seeding. Due to transport and atomic phys...

  14. Predictions for Non-Solenoidal Startup in Pegasus with Lower Divertor Helicity Injectors

    Science.gov (United States)

    Perry, J. M.; Barr, J. L.; Bongard, M. W.; Fonck, R. J.; Lewicki, B. T.

    2014-10-01

    Non-solenoidal startup in Pegasus has focused on using arrays of local helicity injectors situated on the outboard midplane to leverage PF induction. In contrast, injector assemblies located in the lower divertor region can provide improved performance. Higher toroidal field at the injector increases the helicity injection rate, providing a higher effective loop voltage. Poloidal flux expansion in the divertor region will increase the Taylor relaxation current limit. Radial position control requirements are lessened, as plasma expansion naturally couples to injectors in the divertor region. Advances in cathode design and plasma-facing guard rings allow operation at bias voltages over 1.5 kV, three times higher than previously available. This results in increased effective loop voltage and reduced impurity generation. Operation of helicity injectors in the high field side elevates the current requirements for relaxation to a tokamak-like state, but these are met through the improved injector design and increased control over the poloidal field structure via the addition of new coil sets. These advances, combined with the relocation of the injectors to the divertor region, will allow access to the operational regime where helicity injection current drive, rather the poloidal induction, dominates the discharge--a prerequisite for scaling to larger devices. Initial estimates indicate that plasma currents of 0.25-0.30 MA are attainable at full toroidal field with 4 injectors of 2 cm2 each and 8 kA total injected current. Work supported by US DOE Grant DE-FG02-96ER54375.

  15. Extreme Ultraviolet Spectra of Few-Times Ionized Tungsten for Divertor Plasma Diagnostics

    Directory of Open Access Journals (Sweden)

    Joel Clementson

    2015-09-01

    Full Text Available The extreme ultraviolet (EUV emission from few-times ionized tungsten atoms has been experimentally studied at the Livermore electron beam ion trap facility. The ions were produced and confined during low-energy operations of the EBIT-I electron beam ion trap. By varying the electron-beam energy from around 30–300 eV, tungsten ions in charge states expected to be abundant in tokamak divertor plasmas were excited, and the resulting EUV emission was studied using a survey spectrometer covering 120–320 Å. It is found that the emission strongly depends on the excitation energy; below 150 eV, it is relatively simple, consisting of strong isolated lines from a few charge states, whereas at higher energies, it becomes very complex. For divertor plasmas with tungsten impurity ions, this emission should prove useful for diagnostics of tungsten flux rates and charge balance, as well as for radiative cooling of the divertor volume. Several lines in the 194–223 Å interval belonging to the spectra of five- and seven-times ionized tungsten (Tm-like W VI and Ho-like W VIII were also measured using a high-resolution spectrometer.

  16. Failure mode analysis of preliminary design of ITER divertor impurity monitor

    Energy Technology Data Exchange (ETDEWEB)

    Kitazawa, Sin-iti, E-mail: kitazawa.siniti@qst.go.jp; Ogawa, Hiroaki

    2016-11-15

    Highlights: • Divertor impurity influx monitor for ITER (DIM) is procured by JADA. • DIM is designed to observe light from nuclear fusion plasma directly. • DIM is under preliminary design phase. • Failure mode of DIM was prepared for RAMI analysis. • RAMI analysis on DIM was performed to reduce technical risks. - Abstract: The objective of the divertor impurity influx monitor (DIM) for ITER is to measure the parameters of impurities and hydrogen isotopes (tritium, deuterium, and hydrogen) in divertor plasma using visible and UV spectroscopic techniques in the 200–1000 nm wavelength range. In ITER, special provisions are required to ensure accuracy and full functionality of the diagnostic components under harsh conditions (high temperature, high magnetic field, high vacuum condition, and high radiation field). Japan Domestic Agency is preparing the preliminary design of the ITER DIM system, which will be installed in the upper, equatorial and lower ports. The optical and mechanical designs of the DIM are conducted to fit ITER’s requirements. The optical and mechanical designs meet the requirements of spatial resolution. Some auxiliary systems were examined via prototyping. The preliminary design of the ITER DIM system was evaluated by RAMI analysis. The availability of the designed system is adequately high to satisfy the project requirements. However, some equipment does not have certain designs, and this may cause potential technical risks. The preliminary design should be modified to reduce technical risks and to prepare the final design.

  17. A numerical study of plasma detachment conditions in JET divertor plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Simonini, R.; Corrigan, G.; Radford, G.; Spence, J.; Taroni, A.; Weber, S. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking

    1994-07-01

    Simulation results obtained with the EDGE2D/U code confirm that for a given particle inventory in the SOL (including the divertor), the main parameter determining whether or not particle, momentum and energy detachment occurs, is the residual power P - P{sub lost}, where P is the total power entering the SOL and P{sub lost} is the power lost by transport to walls and by volume losses in the SOL outside the region where detachment takes place. For particle contents leading to reasonable values of the separatrix mid-plane density, detachment is found if the residual power is low enough. Typically the residual power must be inferior to 3 MW for good detachment, with the exact value depending on the geometry of the divertor, the transport assumptions and the neutral recirculation scheme. The results show that divertor plasma conditions relevant for the study of power exhaust and impurity control problems are possible in JET. 9 refs., 2 figs., 1 tab.

  18. Using Divertor Strike Point Splitting to Study Plasma Response and Its Sensitivity to Equilibrium Uncertainties

    Science.gov (United States)

    Teklu, Abraham; Orlov, D. M.; Moyer, R. A.; Bykov, I.; Evans, T. E.; Wu, W.; Trevisan, G. L.; Lyons, B. C.; Abrams, T.; Makowski, M. A.; Lasnier, C. S.; Fenstermacher, M. E.

    2017-10-01

    Resonant magnetic perturbations (RMPs) from 3D coils have been varied to modify the splitting of the divertor strike points in DIII-D. This splitting is imaged in filtered visible and infrared emission from the divertor to determine the particle and heat flux patterns on the target plates. The observed splitting is compared to vacuum and plasma response modeling in discharges where a subset of the RMP coils were ramped to shift the divertor footprints from dominantly n = 3 to n = 2 pattern. These results will be used to determine if the plasma response model can be validated with the measured splitting. We will also study the sensitivity of the modeled splitting to details of the 2D equilibrium. This RMP ramp technique could be used in ITER to spread out the heat flux while avoiding excessive forces on the RMP coils. Work supported by U.S. DOE under the Science Undergraduate Laboratory Internship (SULI) program and DE-FC02-04ER54698, DE-FG02-07ER54917, DE-FG02-05ER54809 and DE-AC52-07NA27344.

  19. Thermal bifurcation of scrape-off layer plasma and divertor detachment

    Energy Technology Data Exchange (ETDEWEB)

    Krasheninnikov, S.I.; Catto, P.J.; Helander, P.; Sigmar, D.J. [Massachusetts Institute of Technology, Plasma Fusion Center, Cambridge, Massachusetts 02139 (United States); Soboleva, T.K. [Instituto de Ciencias Nucleares, Universidad Nacional Autonoma de Mexico, Mexico D.F. (Mexico)

    1995-07-01

    Models to investigate the main features of plasma--neutral interactions in the recycling region of a tokamak divertor are developed for the two opposite extremes of fluid and Knudsen neutrals. Both neutral models show that a reduction of the heat flux into the hydrogen recycling region below a critical value leads to bifurcation (or rapid change) of the plasma parameters near the target. This bifurcation causes behavior in the scrape-off layer, which is in agreement with the following main features of detached divertor regimes in current tokamak experiments: (i) strong decrease of the plasma temperature near the target, (ii) plasma pressure drop in the recycling region, and (iii) strong decrease of the target heat load and plasma flux onto the target. It is also shown that in the Knudsen limit, the neutral density in the divertor region cannot exceed a maximum density, which is of the order of 1--2{times}10{sup 13} cm{sup {minus}3} for current experiments. {copyright} {ital 1995} {ital American} {ital Institute} {ital of} {ital Physics}.

  20. Measurements of non-axisymmetric effects in the DIII-D divertor

    Energy Technology Data Exchange (ETDEWEB)

    Evans, T.E,; Leonard, A.W.; Petrie, T.W.; Schaffer, M.J. [General Atomics, San Diego, CA (United States); Lasnier, C.J.; Hill, D.N.; Fenstermacher, M.E. [Lawrence Livermore National Lab., CA (United States)

    1994-07-01

    Non-stationary toroidal asymmetries are observed in the DIII-D divertor heat flux and scrape-off layer (SOL) currents. Using the present DIII-D diagnostics asymmetries are seen much less frequently in single-null H-modes (<5%) than in double-null H-modes (>50%). Divertor heat flux asymmetries are characterized by toroidal variations in the radial profile (i.e., multiple or bifurcated peaks at some toroidal locations and single peaks at others) while SOL currents sometimes have a strongly bipolar toroidal structure. SOL current asymmetries are particularly large during Edge Localized Modes (ELMs). In some cases heat flux variations of as much as a factor of two are seen. The measurements reported here indicate that these asymmetries are best described by a model in which non-axisymmetric radial magnetic perturbations create magnetic islands in the plasma boundary and scrape-off layer which then cause toroidal variation in the divertor heat flux and the scrape-off layer currents.

  1. ATHENA simulations of divertor loss of heat sink transient for the GSSR - Final report with updates

    Energy Technology Data Exchange (ETDEWEB)

    Sponton, L.L

    2001-05-01

    The ITER-FEAT Generic Site Safety Report includes evaluations of the consequences of various types of conceivable transients that can occur during operation. The transients that have to be considered in this respect are specified in the Accident Analysis Specifications document of the safety report. For the divertor primary heat transport system the ranges of transients include amongst others a loss of heat sink at full fusion power operation. The thermal-hydraulic consequences related to the coolability of the divertor primary heat transport system components for this transient have been evaluated and summarised in the safety report and in the current report an overview of those efforts and associated outcome is provided. The analyses have been made with the ATHENA thermal-hydraulic code using a separately developed ATHENA model of the ITER-FEAT divertor cooling system. In the current report results from calculations with an updated pressurizer model and pressurizer control system are outlined. The results show that the pressurizer safety valve does not open, that the pressurizer level increase is moderate and that no temperature increases jeopardize the structure integrity.

  2. Time-dependent modeling of dust injection in semi-detached ITER divertor plasma

    Science.gov (United States)

    Smirnov, Roman; Krasheninnikov, Sergei

    2017-10-01

    At present, it is generally understood that dust related issues will play important role in operation of the next step fusion devices, i.e. ITER, and in the development of future fusion reactors. Recent progress in research on dust in magnetic fusion devises has outlined several topics of particular concern: a) degradation of fusion plasma performance; b) impairment of in-vessel diagnostic instruments; and c) safety issues related to dust reactivity and tritium retention. In addition, observed dust events in fusion edge plasmas are highly irregular and require consideration of temporal evolution of both the dust and the fusion plasma. In order to address the dust-related fusion performance issues, we have coupled the dust transport code DUSTT and the edge plasma transport code UEDGE in time-dependent manner, allowing modeling of transient dust-induced phenomena in fusion edge plasmas. Using the coupled codes we simulate burst-like injection of tungsten dust into ITER divertor plasma in semi-detached regime, which is considered as preferable ITER divertor operational mode based on the plasma and heat load control restrictions. Analysis of transport of the dust and the dust-produced impurities, and of dynamics of the ITER divertor and edge plasma in response to the dust injection will be presented. This material is based upon work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences, under Award Number DE-FG02-06ER54852.

  3. Understanding of Neutral Gas Transport in the Alcator C-Mod Tokamak Divertor

    Energy Technology Data Exchange (ETDEWEB)

    D.P. Stotler; C.S. Pitcher; C.J. Boswell; B. LaBombard; J.L. Terry; J.D. Elder; S. Lisgo

    2002-05-07

    A series of experiments on the effect of divertor baffling on the Alcator C-Mod tokamak provides stringent tests on models of neutral gas transport in and around the divertor region. One attractive feature of these experiments is that a trial description of the background plasma can be constructed from experimental measurements using a simple model, allowing the neutral gas transport to be studied with a stand-alone code. The neutral-ion and neutral-neutral elastic scattering processes recently added to the DEGAS 2 Monte Carlo neutral transport code permit the neutral gas flow rates between the divertor and main chamber to be simulated more realistically than before. Nonetheless, the simulated neutral pressures are too low and the deuterium Balmer-alpha emission profiles differ qualitatively from those measured, indicating an incomplete understanding of the physical processes involved in the experiment. Some potential explanations are examined and opportunities for future exploration a re highlighted. Improvements to atomic and surface physics data and models will play a role in the latter.

  4. In situ measurement of erosion/deposition in the DIII-D divertor by colorimetry

    Science.gov (United States)

    Weschenfelder, F.; Jackson, G. L.; Wienhold, P.; Winter, J.; Brooks, N. H.; West, W. P.; Lee, R.

    1996-07-01

    Colorimetry was introduced into the DIII-D tokamak to measure in situ the growth and erosion of transparent wall coatings (a-C:H) on the divertor. The colorimetric measurement system consisting of a halogen light source, a set of three filters and a black/white camera is described together with a first erosion measurement. An insertable graphite sample with a diameter of 4.7 cm was precoated with a 300 nm thick amorphous carbon film and was exposed in the divertor for several discharges with its surface coplanar to the surrounding graphite tiles. For each of the discharges the plasma strike point was moved onto the sample for 1 s to erode the coating. Between the discharges a camera signal with each filter was recorded and the film thickness was evaluated along a radial line across the DIMES sample. Thus it has been possible for the first time to measure erosion and deposition of divertor material in situ and shot-by-shot. The average peak heat flux with the strike point on DIMES was about 110 W 0741-3335/38/7/009/img10. The measurement shows a strong decrease in the film thickness almost over the entire sample with an average erosion rate of 0741-3335/38/7/009/img11.

  5. Design and construction of a lithium vapor box divertor similarity experiment

    Science.gov (United States)

    Schwartz, J. A.; Cohen, R. A.; Emdee, E. D.; Jaworski, M. A.; Goldston, R. J.

    2017-10-01

    Future fusion devices will require handling extreme heat fluxes. The lithium vapor box divertor is a concept to manage this heat flux. The divertor plasma impinges on a dense cloud of lithium vapor, leading to volumetric cooling, radiation, and recombination. The vapor is localized by baffles and condensation on the divertor slot walls upstream of the target, limiting the lithium reaching the main chamber. A series of test stand experiments will study vapor confinement and plasma plugging in a simplified baffled-pipe geometry. A first experiment without plasma will validate a DSMC model for evaporation, flow, and condensation of lithium vapor. Three stainless steel cylindrical cans will be heated to 550C, 600C, and 650C respectively inside a vacuum chamber. Lithium flow will be measured by weighing the cans before and after heating and by calorimetry of the latent heat of the vapor. Progress on the experiment will be presented. This work supported by DOE Contract No. DE-AC02-09CH11466.

  6. Global Value Chain Configuration

    DEFF Research Database (Denmark)

    Hernandez, Virginia; Pedersen, Torben

    2017-01-01

    This paper reviews the literature on global value chain configuration, providing an overview of this topic. Specifically, we review the literature focusing on the concept of the global value chain and its activities, the decisions involved in its configuration, such as location, the governance...... modes chosen and the different ways of coordinating them. We also examine the outcomes of a global value chain configuration in terms of performance and upgrading. Our aim is to review the state of the art of these issues, identify research gaps and suggest new lines for future research that would...

  7. Magnetospheric configuration of Neptune.

    Science.gov (United States)

    Schulz, M.; McNab, M. C.; Lepping, R. P.; Voigt, G.-H.

    Voyager 2 encountered Neptune's magnetosphere in a nearly pole-on configuration and proceeded to explore the magnetosphere for about 1.2 Neptunian days. During this time the angle ψ between the planetary dipole moment μ and the solar wind velocity u varied from about 20° to about 114° and back, thereby producing a range of magnetospheric configurations previously attainable only in computer simulations. This chapter provides an overview of observations (with emphasis on the cusp region) made by the spacecraft magnetometer and other instruments during this traversal of Neptune's magnetosphere and places these observational results in the perspective of a global magnetospheric configuration, as provided by quantitative magnetospheric models.

  8. Drupal 8 configuration management

    CERN Document Server

    Borchert, Stefan

    2015-01-01

    Drupal 8 Configuration Management is intended for people who use Drupal 8 to build websites, whether you are a hobbyist using Drupal for the first time, a long-time Drupal site builder, or a professional web developer.

  9. The LHCb configuration database

    CERN Document Server

    Abadie, Lana; Gaspar, Clara; Jacobsson, Richard; Jost, Beat; Neufeld, Niko

    2005-01-01

    The Experiment Control System (ECS) will handle the monitoring, configuration and operation of all the LHCb experimental equipment. All parameters required to configure electronics equipment under the control of the ECS will reside in a configuration database. The database will contain two kinds of information: 1.\tConfiguration properties about devices such as hardware addresses, geographical location, and operational parameters associated with particular running modes (dynamic properties). 2.\tConnectivity between devices : this consists of describing the output and input connections of a device (static properties). The representation of these data using tables must be complete so that it can provide all the required information to the ECS and must cater for all the subsystems. The design should also guarantee a fast response time, even if a query results in a large volume of data being loaded from the database into the ECS. To fulfil these constraints, we apply the following methodology: Determine from the d...

  10. Airport Configuration Prediction Project

    Data.gov (United States)

    National Aeronautics and Space Administration — There is presently poor knowledge throughout the National Airspace System (NAS) of the airport configurations currently in use at each airport. There is even less...

  11. Configuration Management Automation (CMA) -

    Data.gov (United States)

    Department of Transportation — Configuration Management Automation (CMA) will provide an automated, integrated enterprise solution to support CM of FAA NAS and Non-NAS assets and investments. CMA...

  12. Quantification of chemical erosion in the divertor of the DIII-D tokamak

    Science.gov (United States)

    McLean, Adam Gordon

    The International Thermonuclear Experimental Reactor (ITER) is currently designed to use graphite targets in the divertor for power handling and impurity control. Understanding and quantifying chemical sputtering is therefore key to the success of fusion as a clean energy source. The principal goal of this thesis is to design and carry out experiments, then analyze and interpret the results in order to elucidate the role of chemical sputtering in carbon sources in the DIII-D tokamak. A self-contained gas puff system has been designed, constructed, and employed for in-situ study of chemical erosion. The porous plug injector (PPI) releases methane through a porous graphite surface into the divertor plasma at a precisely calibrated rate, minimizing perturbation to local plasma while replicating the immediate environment of methane molecules released from a solid graphite surface more accurately than done previously. For the first time in a tokamak environment, the methane flow rate used in a puffing experiment was the same order of magnitude as that expected from laboratory experiments for intrinsic chemical sputtering. Effective photon efficiencies for CH4 injection are reported; results are found to have significant dependencies on surface conditions and the divertor operating regime. The contribution of sputtering processes to sources of C0 and C+ are assessed through measurement of background and incremental spectroscopic emissions of both physically and chemically-released sputtering products and by CI, 910 nm line profile fitting. Comparison of background and incremental emissions of chemically-released products demonstrate a dramatic drop in production of CH in cold and detached conditions. Finally, the chemical erosion yield is calculated in both attached and cold-divertor conditions and found to be much closer to that measured ex-situ in ion beam experiments than previously determined in DII-D. These observations represent a positive result for ITER which

  13. ADX: a high field, high power density, advanced divertor and RF tokamak

    Science.gov (United States)

    LaBombard, B.; Marmar, E.; Irby, J.; Terry, J. L.; Vieira, R.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.; Baek, S.; Beck, W.; Bonoli, P.; Brunner, D.; Doody, J.; Ellis, R.; Ernst, D.; Fiore, C.; Freidberg, J. P.; Golfinopoulos, T.; Granetz, R.; Greenwald, M.; Hartwig, Z. S.; Hubbard, A.; Hughes, J. W.; Hutchinson, I. H.; Kessel, C.; Kotschenreuther, M.; Leccacorvi, R.; Lin, Y.; Lipschultz, B.; Mahajan, S.; Minervini, J.; Mumgaard, R.; Nygren, R.; Parker, R.; Poli, F.; Porkolab, M.; Reinke, M. L.; Rice, J.; Rognlien, T.; Rowan, W.; Shiraiwa, S.; Terry, D.; Theiler, C.; Titus, P.; Umansky, M.; Valanju, P.; Walk, J.; White, A.; Wilson, J. R.; Wright, G.; Zweben, S. J.

    2015-05-01

    The MIT Plasma Science and Fusion Center and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX)—a tokamak specifically designed to address critical gaps in the world fusion research programme on the pathway to next-step devices: fusion nuclear science facility (FNSF), fusion pilot plant (FPP) and/or demonstration power plant (DEMO). This high-field (⩾6.5 T, 1.5 MA), high power density facility (P/S ˜ 1.5 MW m-2) will test innovative divertor ideas, including an ‘X-point target divertor’ concept, at the required performance parameters—reactor-level boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region—while simultaneously producing high-performance core plasma conditions that are prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fuelling from external heating and current drive systems. Equally important, the experimental platform will test innovative concepts for lower hybrid current drive and ion cyclotron range of frequency actuators with the unprecedented ability to deploy launch structures both on the low-magnetic-field side and the high-magnetic-field side—the latter being a location where energetic plasma-material interactions can be controlled and favourable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination—advanced divertors, advanced RF actuators, reactor-prototypical core plasma conditions—will enable ADX to explore enhanced core confinement physics, such as made possible by reversed central shear, using only the types of external drive systems that are considered viable for a fusion power plant. Such an integrated demonstration of high-performance core-divertor operation with steady-state sustainment would pave the way towards an attractive pilot plant, as envisioned in the ARC concept

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-08-01

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

  15. Electric probe diagnostics for measuring SOL parameters, wall and divertor fluxes in KSTAR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Heung-Su, E-mail: kimhs@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Bak, Jun-Gyo [National Fusion Research Institute, Daejeon (Korea, Republic of); Bae, Min-Keun; Chung, Kyu-Sun [Hanyang University, Seoul (Korea, Republic of); Hong, Suk-Ho [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2016-11-01

    Highlights: • Some components in EPDs were improved to investigate characteristics of the SOL plasmas and to measure wall and divertor fluxes in the KSTAR tokamak plasmas. From the upgrades in the EPDs, the measured error of the elapsed distance for the evaluation of the SOL profiles can be reduced up to 1%. • In the SOL parameter measurement during IWL plasma, the e-folding lengths in the main SOL region lTe and lne were evaluated as 3.5 cm and 2.1 cm, respectively. • From flux measurement at the far SOL during a diverted ELMy H-mode, peaked heat flux toward to outboard wall during ELMs might be less than 1% of the peaked divertor heat flux. • The movement of an OSP during a diverted H-mode can be detected from the divertor probe measurement, and the peaked heat flux near the OSP was estimated as few MW m-2. - Abstract: Some components in electric probe diagnostics (EPDs) are improved in order to investigate characteristics of edge plasmas in the upstream scrape-off-layer (SOL) region and to measure wall and divertor fluxes during L-mode and H-mode plasma discharges in the Korea Superconducting Tokamak Advanced Research (KSTAR). From the upgrades in the EPDs, the measured error of the elapsed distance for the evaluation of the SOL profiles can be reduced up to 1% and the ion saturation current of up to 1.0 A near an outer strike point (OSP) can be measured at the divertor region. In the SOL profile measurements during L-mode and inner wall limited plasma (B{sub T} = 2.0 T, I{sub p} = 0.4 MA), the e-folding lengths in the main SOL region λ{sub Te} and λ{sub ne} are evaluated as 3.5 cm and 2.1 cm, respectively. From particle flux measurement at the far SOL region during a diverted ELMy H-mode discharge (B{sub T} = 1.8 T, I{sub p} = 0.65 MA), peaked heat flux toward to outboard wall during ELM bursts is estimated up to ∼20 k Wm{sup −2}, which may be less than 1% of the peaked divertor heat flux expected for the neutral beam (NB) heating power P{sub NB

  16. Studies of power exhaust and divertor design for a 1.5 GW-level fusion power DEMO

    Science.gov (United States)

    Asakura, N.; Hoshino, K.; Suzuki, S.; Tokunaga, S.; Someya, Y.; Utoh, H.; Kudo, H.; Sakamoto, Y.; Hiwatari, R.; Tobita, K.; Shimizu, K.; Ezato, K.; Seki, Y.; Ohno, N.; Ueda, Y.; Joint Special TeamDEMO Design

    2017-12-01

    Power exhaust to the divertor and the conceptual design have been investigated for a steady-state DEMO in Japan with 1.5 GW-level fusion power and the major radius of 8.5 m, where the plasma parameters were revised appropriate for the impurity seeding scenario. A system code survey for the Ar impurity seeding suggested the volume-averaged density, impurity concentration and exhaust power from the main plasma of {{P}sep ~ }   =  205-285 MW. The divertor plasma simulation (SONIC) was performed in the divertor leg length of 1.6 m with the fixed exhaust power to the edge of {{P}out}   =  250 MW and the total radiation fraction at the edge, SOL and divertor ({{P}rad}/{{P}out}   =  0.8), as a first step to investigate appropriate design of the divertor size and geometry. At the outer target, partial detachment was produced near the strike-point, and the peak heat load ({{q}target} ) at the attached region was reduced to ~5 MW m-2 with appropriate fuel and impurity puff rates. At the inner divertor target, full detachment of ion flux was produced and the peak {{q}target} was less than 10 MW m-2 mostly due to the surface-recombination. These results showed a power exhaust scenario and the divertor design concept. An integrated design of the water-cooling heat sink for the long leg divertor was proposed. Cu-ally (CuCrZr) cooling pipe was applicable as the heat sink to handle the high heat flux near the strike-point, where displacements per atom rate was estimated to be 0.5-1.5 per year by neutronics calculation. An arrangement of the coolant rooting for Cu-alloy and Reduced Activation Ferritic Martensitic (RAFM) steel (F82H) pipes in a divertor cassette was investigated, and the heat transport analysis of the W-monoblock and Cu-alloy pipe under the peak {{q}target} of 10 MWm-2 and nuclear heating was performed. The maximum temperatures on the W-surface and Cu-alloy pipe were 1021 and 331 °C. Heat flux of 16 MW m-2 was distributed in the major part

  17. The Design and Use of Tungsten Coated TZM Molybdenum Tile Inserts in the DIII-D Tokamak Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, Christopher [General Atomics, San Diego; Nygren, R. E. [Sandia National Laboratories (SNL); Chrobak, C P. [General Atomics, San Diego; Buchenauer, Dean [Sandia National Laboratories (SNL); Holtrop, Kurt [General Atomics, San Diego; Unterberg, Ezekial A. [ORNL; Zach, Mike P. [ORNL

    2017-08-01

    Future tokamak devices are envisioned to utilize a high-Z metal divertor with tungsten as theleading candidate. However, tokamak experiments with tungsten divertors have seen significantdetrimental effects on plasma performance. The DIII-D tokamak presently has carbon as theplasma facing surface but to study the effect of tungsten on the plasma and its migration aroundthe vessel, two toroidal rows of carbon tiles in the divertor region were modified with high-Zmetal inserts, composed of a molybdenum alloy (TZM) coated with tungsten. A dedicated twoweek experimental campaign was run with the high-Z metal inserts. One row was coated withtungsten containing naturally occurring levels of isotopes. The second row was coated withtungsten where the isotope 182W was enhanced from the natural level of 26% up to greater than90%. The different isotopic concentrations enabled the experiment to differentiate between thetwo different sources of metal migration from the divertor. Various coating methods wereexplored for the deposition of the tungsten coating, including chemical vapor deposition,electroplating, vacuum plasma spray, and electron beam physical vapor deposition. The coatingswere tested to see if they were robust enough to act as a divertor target for the experiment. Testsincluded cyclic thermal heating using a high power laser and high-fluence deuterium plasmabombardment. The issues associate with the design of the inserts (tile installation, thermal stress,arcing, leading edges, surface preparation, etc.), are reviewed. The results of the tests used toselect the coating method and preliminary experimental observations are presented.

  18. Large Area Divertor Temperature Measurements Using A High-speed Camera With Near-infrared FiIters in NSTX

    Energy Technology Data Exchange (ETDEWEB)

    Lyons, B C; Zweben, S J; Gray, T K; Hosea, J; Kaita, R; Kugel, H W; Maqueda, R J; McLean, A G; Roquemore, A L; Soukhanovskii, V A

    2011-04-05

    Fast cameras already installed on the National Spherical Torus Experiment (NSTX) have be equipped with near-infrared (NIR) filters in order to measure the surface temperature in the lower divertor region. Such a system provides a unique combination of high speed (> 50 kHz) and wide fi eld-of-view (> 50% of the divertor). Benchtop calibrations demonstrated the system's ability to measure thermal emission down to 330 oC. There is also, however, signi cant plasma light background in NSTX. Without improvements in background reduction, the current system is incapable of measuring signals below the background equivalent temperature (600 - 700 oC). Thermal signatures have been detected in cases of extreme divertor heating. It is observed that the divertor can reach temperatures around 800 oC when high harmonic fast wave (HHFW) heating is used. These temperature profiles were fi t using a simple heat diffusion code, providing a measurement of the heat flux to the divertor. Comparisons to other infrared thermography systems on NSTX are made.

  19. The influence of divertor geometry on access to high confinement regimes on the Alcator C-Mod tokamak

    Science.gov (United States)

    Hughes, J. W.; Labombard, B.; Hubbard, A.; Marmar, E.; Terry, J.; Rice, J.; Walk, J.; Whyte, D.; Ma, Y.; Cziegler, I.; Edlund, E.; Theiler, C.

    2014-10-01

    The placement of X-point and strike points in a diverted tokamak can have a remarkable impact on properties of the discharge, including thermal and particle confinement. The distinctive divertor of Alcator C-Mod allows us to demonstrate these effects experimentally, as we vary equilibrium shaping to obtain substantial variation of divertor leg length, field line attack angle and divertor baffling. In response to these changes, we observe differences in both L-mode confinement and access to high-confinement regimes (i.e. ELMy H-mode and I-mode). With the ion grad-B drift directed toward the divertor, scanning the strike point can induce ~2× reductions in H-mode power threshold, and can produce a window for I-mode operation with H98 > 1. Recent experiments seek to explore these effects using improved diagnostics, and to extend them to the case with ion grad-B drift directed away from the divertor. Supported by USDoE award DE-FC02-99ER54512.

  20. Real-time radiative divertor feedback control development for the NSTX-U tokamak using a vacuum ultraviolet spectrometer

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V. A., E-mail: vlad@llnl.gov [Lawrence Livermore National Laboratory, 7000 East Ave., Livermore, California 94550 (United States); Kaita, R.; Stratton, B. [Princeton Plasma Physics Laboratory, 100 Stellarator Rd., Princeton, New Jersey 08543 (United States)

    2016-11-15

    A radiative divertor technique is planned for the NSTX-U tokamak to prevent excessive erosion and thermal damage of divertor plasma-facing components in H-mode plasma discharges with auxiliary heating up to 12 MW. In the radiative (partially detached) divertor, extrinsically seeded deuterium or impurity gases are used to increase plasma volumetric power and momentum losses. A real-time feedback control of the gas seeding rate is planned for discharges of up to 5 s duration. The outer divertor leg plasma electron temperature T{sub e} estimated spectroscopically in real time will be used as a control parameter. A vacuum ultraviolet spectrometer McPherson Model 251 with a fast charged-coupled device detector is developed for temperature monitoring between 5 and 30 eV, based on the Δn = 0, 1 line intensity ratios of carbon, nitrogen, or neon ion lines in the spectral range 300–1600 Å. A collisional-radiative model-based line intensity ratio will be used for relative calibration. A real-time T{sub e}-dependent signal within a characteristic divertor detachment equilibration time of ∼10–15 ms is expected.

  1. Configuration Control Office

    CERN Multimedia

    Beltramello, O

    In order to enable Technical Coordination to manage the detector configuration and to be aware of all changes in this configuration, a baseline of the envelopes has been created in April 2001. Fifteen system and multi-system envelope drawings have been approved and baselined. An EDMS file is associated with each approved envelope, which provides a list of the current known unsolved conflicts related to the envelope and a list of remaining drawing inconsistencies to be corrected. The envelope status with the associated drawings and EDMS file can be found on the web at this adress: http://atlasinfo.cern.ch/Atlas/TCOORD/Activities/Installation/Configuration/ Any modification in the baseline has to be requested via the Engineering Change Requests. The procedure can be found under: http://atlasinfo.cern.ch/Atlas/TCOORD/Activities/TcOffice/Quality/ECR/ TC will review all the systems envelopes in the near future and manage conflict resolution with the collaboration of the systems.

  2. Relevance of collisionality in the transport model assumptions for divertor detachment multi-fluid modelling on JET

    DEFF Research Database (Denmark)

    Wiesen, S.; Fundamenski, W.; Wischmeier, M.

    2011-01-01

    A revised formulation of the perpendicular diffusive transport model in 2D multi-fluid edge codes is proposed. Based on theoretical predictions and experimental observations a dependence on collisionality is introduced into the transport model of EDGE2D–EIRENE. The impact on time-dependent JET gas...... fuelled ramp-up scenario modelling of the full transient from attached divertor into the high-recycling regime, following a target flux roll over into divertor detachment, ultimately ending in a density limit is presented. A strong dependence on divertor geometry is observed which can mask features...... of the new transport model: a smoothly decaying target recycling flux roll over, an asymmetric drop of temperature and pressure along the field lines as well as macroscopic power dependent plasma oscillations near the density limit which had been previously observed also experimentally. The latter effect...

  3. Kinetic modeling of divertor heat load fluxes in the Alcator C-Mod and DIII-D tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Pankin, A. Y. [Tech-X Corporation, Boulder, Colorado 80303 (United States); Rafiq, T.; Kritz, A. H. [Department of Physics, Lehigh University, Bethlehem, Pennsylvania 18015 (United States); Park, G. Y. [National Fusion Research Institute, Daejeon, 305-333 (Korea, Republic of); Chang, C. S.; Ku, S. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States); Brunner, D.; Hughes, J. W.; LaBombard, B.; Terry, J. L. [MIT Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Groebner, R. J. [General Atomics, San Diego, California 92121 (United States)

    2015-09-15

    The guiding-center kinetic neoclassical transport code, XGC0 [Chang et al., Phys. Plasmas 11, 2649 (2004)], is used to compute the heat fluxes and the heat-load width in the outer divertor plates of Alcator C-Mod and DIII-D tokamaks. The dependence of the width of heat-load fluxes on neoclassical effects, neutral collisions, and anomalous transport is investigated using the XGC0 code. The XGC0 code includes realistic X-point geometry, a neutral source model, the effects of collisions, and a diffusion model for anomalous transport. It is observed that the width of the XGC0 neoclassical heat-load is approximately inversely proportional to the total plasma current I{sub p.} The scaling of the width of the divertor heat-load with plasma current is examined for an Alcator C-Mod discharge and four DIII-D discharges. The scaling of the divertor heat-load width with plasma current is found to be weaker in the Alcator C-Mod discharge compared to scaling found in the DIII-D discharges. The effect of neutral collisions on the 1/I{sub p} scaling of heat-load width is shown not to be significant. Although inclusion of poloidally uniform anomalous transport results in a deviation from the 1/I{sub p} scaling, the inclusion of the anomalous transport that is driven by ballooning-type instabilities results in recovering the neoclassical 1/I{sub p} scaling. The Bohm or gyro-Bohm scalings of anomalous transport do not strongly affect the dependence of the heat-load width on plasma current. The inclusion of anomalous transport, in general, results in widening the width of neoclassical divertor heat-load and enhances the neoclassical heat-load fluxes on the divertor plates. Understanding heat transport in the tokamak scrape-off layer plasmas is important for strengthening the basis for predicting divertor conditions in ITER.

  4. 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.

  5. Thermomechanical characterization of joints for blanket and divertor application processed by electrochemical plating

    Energy Technology Data Exchange (ETDEWEB)

    Krauss, Wolfgang; Lorenz, Julia; Konys, Jürgen; Basuki, Widodo; Aktaa, Jarir

    2016-11-01

    Highlights: • Electroplating is a relevant technology for brazing of blanket and divertor parts. • Tungsten, Eurofer and steel joints successfully fabricated. • Reactive interlayers improve adherence and reduce failure risks. • Qualification of joints performed by thermo-mechanical testing and aging. • Shear strength of joints comparable with conventionally brazing of steels. - Abstract: Fusion technology requires in the fields of first wall and divertor development reliable and adjusted joining processes of plasma facing tungsten to heat sinks or blanket structures. The components to be bonded will be fabricated from tungsten, steel or other alloys like copper. The parts have to be joined under functional and structural aspects considering the metallurgical interactions of alloys to be assembled and the filler materials. Application of conventional brazing showed lacks ranging from bad wetting of tungsten up to embrittlement of fillers and brazing zones. Thus, the deposition of reactive interlayers and filler components, e.g. Ni, Pd or Cu was initiated to overcome these metallurgical restrictions and to fabricate joints with aligned mechanical behavior. This paper presents results concerning the joining of tungsten, Eurofer and stainless steel for blanket and divertor application by applying electroplating technology. Metallurgical and mechanical characterization by shear testing were performed to analyze the joints quality and application limits in dependence on testing temperature between room temperature and 873 K and after thermal aging of up to 2000 h. The tested interlayers Ni and Pd enhanced wetting and enabled the processing of reliable joints with a shear strength of more than 200 MPa at RT.

  6. Production and fabrication of vanadium alloys for the radiative divertor program of DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, W.R.; Smith, J.P.; Stambaugh, R.D.

    1996-04-01

    V-4Cr-4-Ti alloy has been recently selected for use in the manufacture of a portion of the DIII-D Radiative Divertor modification, as part of an overall DIII-D vanadium alloy deployment effort developed by General Atomics (GA) in conjunction with the Argonne and Oak Ridge National Laboratories (ANL or ORNL). The goal of this work is to produce a production-scale heat of the alloy and fabricate it into product forms for the manufacture of a portion of the Radiative Divertor (RD) for the DIII-D tokamak, to develop the fabrications technology for manufacture of the vanadium alloy radiative Divertor components, and to determine the effects of typical tokamak environments in the behavior of the vanadium alloy. The production of a {approx}1300-kg heat of V-4Cr-4Ti alloy is currently in progress at Teledyne Wah Chang of Albany, oregon (TWCA) to provide sufficient material for applicable product forms. Two unalloyed vanadium ingots for the alloy have already been produced by electron beam melting of raw processes vanadium. Chemical compositions of one ingot and a portion of the second were acceptable, and Charpy V-Notch (CVN) impact test performed on processed ingot samples indicated ductile behavior. Material from these ingots are currently being blended with chromium and titanium additions, and will be vacuum-arc remelted into a V-4Cr-4Ti alloy ingot and converted into product forms suitable for components of the DIII-D RD structure. Several joining methods selected for specific applications in fabrication of the RD components are being investigated, and preliminary trials have been successful in the joining of V-alloy to itself by both resistance and inertial welding processes and to Inconel 625 by inertial welding.

  7. DNS BIND Server Configuration

    Directory of Open Access Journals (Sweden)

    Radu MARSANU

    2011-01-01

    Full Text Available After a brief presentation of the DNS and BIND standard for Unix platforms, the paper presents an application which has a principal objective, the configuring of the DNS BIND 9 server. The general objectives of the application are presented, follow by the description of the details of designing the program.

  8. Ansible configuration management

    CERN Document Server

    Hall, Daniel

    2015-01-01

    This book is intended for anyone who wants to learn Ansible starting from the basics. Some experience of how to set up and configure Linux machines and a working knowledge of BIND, MySQL, and other Linux daemons is expected.

  9. Evaluating organizational configurations

    NARCIS (Netherlands)

    Penserini, L.; Dignum, F.; Dignum, V.; Aldewereld, H.; Grossi, D.; Baeza-Yates, R.; Lang, J.; Mitra, S.; Parsons, S.; Pasi, G.

    2009-01-01

    A Multi-Agent System is often conceived as an organization of autonomous software agents that participate into social and evolving structures (e.g., organizational configurations) suitable to deal with highly dynamic environments. Nevertheless, systems based on agent technologies rarely capitalize

  10. Reference frame for Product Configuration

    DEFF Research Database (Denmark)

    Ladeby, Klaes Rohde; Oddsson, Gudmundur Valur

    2011-01-01

    on configuration systems in the shape of anecdotal reporting on the development of information systems that perhaps support the configuration task – perhaps not. Consequently, the definition of configuration has become ambiguous as different research groups defines configuration differently. This paper propose......This paper presents a reference frame for configuration. The reference frame is established by review of existing literature, and consequently it is a theoretical frame of reference. The review of literature shows a deterioration of the understanding of configuration. Most recent literature reports...... a reference frame for configuration that permits 1) a more precise understanding of a configuration system, 2) a understanding of how the configuration system relate to other systems, and 3) a definition of the basic concepts in configuration. The total configuration system, together with the definition...

  11. Testing candidate interlayers for an enhanced water-cooled divertor target

    Energy Technology Data Exchange (ETDEWEB)

    Hancock, David, E-mail: david.hancock@ccfe.ac.uk [CCFE, Culham Science Centre, Oxfordshire OX14 3DB (United Kingdom); Barrett, Tom; Foster, James; Fursdon, Mike; Keech, Gregory; McIntosh, Simon; Timmis, William [CCFE, Culham Science Centre, Oxfordshire OX14 3DB (United Kingdom); Rieth, Michael; Reiser, Jens [Karlsruhe Institute of Technology, IAM-AWP, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2015-10-15

    Highlights: • We introduce an optimised divertor target concept: the “Thermal Break”. • We suggest a candidate interlayer material for this concept: FeltMetal. • We describe a bespoke rig for testing the thermal conductivity of this material. • We present preliminary results for a number of samples. - Abstract: The design of a divertor target for DEMO remains one of the most challenging engineering tasks to be overcome on the path to fusion power. Under the European DEMO programme, a promising concept known as Thermal Break has been developed at CCFE. This concept is a variation of the ITER tungsten divertor in which the pure Copper interlayer between Copper Chrome Zirconium coolant pipe and Tungsten monoblock armour is replaced with a low thermal conductivity compliant interlayer, with the aim of reducing the thermal mismatch stress between the armour and structure. One candidate material for this interlayer is FeltMetal™ (Technetics Group, USA). This material consists of an amorphous matrix of fine copper wires which are sintered onto a thin copper foil, creating a sheet of approximately 1 mm thickness. FeltMetal has been successfully used for many years to provide compliant sliding electrical contacts for the MAST TF coils and on ALCATOR C-Mod and extensive material testing has therefore been undertaken to quantify thermal and mechanical properties. These tests, however, have not been performed under vacuum or DEMO-relevant conditions. A bespoke experimental test rig has therefore been designed and constructed with which to measure the interlayer thermal conductance as a function of temperature and pressure under vacuum conditions. The design of this apparatus and the results of experiments on FeltMetal as well as other candidate interlayers are presented here. In parallel, joint mockups using the candidate interlayers have been prepared and Thermal Break divertor target mockups have been manufactured, requiring the development of a dedicated

  12. On the asymmetries of ELM divertor power deposition in JET and ASDEX Upgrade

    DEFF Research Database (Denmark)

    Eich, T.; Kallenbach, A.; Fundamenski, W.

    2009-01-01

    An analytical expression was derived for describing the divertor target power during ELMs based on the model discussed in [W. Fundamenski, R.A. Pitts, Plasma Phys. Control. Fus. 48 (2006) 109] where the power load arises from a Maxwellian distribution of particles released into the SOL region....... The paper discusses a comparable simple extension of the model by introducing a non-zero characteristic velocity of the Maxwellian distributed particles. This way the experimentally observed temporal evolution as well as the in/out energy imbalance can be described. The extended model named free...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-01-01

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

  14. Electron temperature and heat load measurements in the COMPASS divertor using the new system of probes

    Czech Academy of Sciences Publication Activity Database

    Adámek, Jiří; Seidl, Jakub; Horáček, Jan; Komm, Michael; Eich, T.; Pánek, Radomír; Cavalier, J.; Devitre, A.; Peterka, Matěj; Vondráček, Petr; Stöckel, Jan; Šesták, David; Grover, Ondřej; Bílková, Petra; Böhm, Petr; Varju, Jozef; Havránek, Aleš; Weinzettl, Vladimír; Lovell, J.; Dimitrova, Miglena; Mitošinková, Klára; Dejarnac, Renaud; Hron, Martin

    2017-01-01

    Roč. 57, č. 11 (2017), č. článku 116017. ISSN 0029-5515 R&D Projects: GA ČR(CZ) GA15-10723S; GA ČR(CZ) GA16-14228S; GA MŠk(CZ) LM2015045 EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : COMPASS * divertor * heat load * ELM * electron temperature * Ball-pen probe Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.307, year: 2016 http://iopscience.iop.org/article/10.1088/1741-4326/aa7e09

  15. Feasibility study of fast swept divertor strike point suppressing transient heat fluxes in big tokamaks.

    Czech Academy of Sciences Publication Activity Database

    Horáček, Jan; Cunningham, G.; Entler, Slavomír; Dobias, P.; Duban, R.; Imríšek, Martin; Markovič, Tomáš; Havlíček, Josef; Enikeev, R.

    2017-01-01

    Roč. 123, November (2017), s. 646-649 ISSN 0920-3796. [SOFT 2016: Symposium on Fusion Technology /29./. Prague, 05.09.2016-09.09.2016] R&D Projects: GA ČR(CZ) GA16-14228S; GA MŠk(CZ) LM2015045; GA MŠk(CZ) 8D15001; GA MŠk LG14002 Institutional support: RVO:61389021 Keywords : DEMO * ELM * Divertor * Heat flux * Tokamak Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S0920379617300376

  16. Results and analysis of high heat flux tests on a full-scale vertical target prototype of ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Missirlian, M. [Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 Saint Paul Lez Durance, Cedex (France)]. E-mail: missir@drfc.cad.cea.fr; Escourbiac, F. [Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 Saint Paul Lez Durance, Cedex (France); Merola, M. [EFDA Close Support Unit, Garching (Germany); Bobin-Vastra, I. [FRAMATOME, Le Creusot (France); Schlosser, J. [Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 Saint Paul Lez Durance, Cedex (France); Durocher, A. [Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 Saint Paul Lez Durance, Cedex (France)

    2005-11-15

    After an extensive R and D development program, a full-scale divertor target prototype, manufactured with all the main features of the corresponding ITER divertor, was intensively tested in the high heat flux FE200 facility. The prototype consists of four units having a full monoblock geometry. The lower part (CFC armour) and the upper part (W armour) of each monoblock were joined to the solution annealed, quenched and cold worked CuCrZr tube by HIP technique. This paper summarises and analyses the main test results obtained on this prototype.

  17. The ALICE Configuration Tool

    Science.gov (United States)

    Boccioli, M.; Carena, F.; Chapeland, S.; Chibante Barroso, V.; Lechman, M.; Jusko, A.; Pinazza, O.; ALICE Collaboration

    2011-12-01

    ALICE (A Large Ion Collider Experiment) is the heavy-ion detector designed to study the physics of strongly interacting matter and the quark-gluon plasma at the CERN Large Hadron Collider (LHC). It includes 18 different sub-detectors and 5 online systems, each one made of many different components and developed by different teams inside the collaboration. The operation of a large experiment over several years to collect billions of events acquired in well defined conditions requires predictability and repeatability of the experiment configuration. The logistics of the operation is also a major issue and it is mandatory to reduce the size of the shift crew needed to operate the experiment. Appropriate software tools are therefore needed to automate daily operations. This ensures minimizing human errors and maximizing the data taking time. The ALICE Configuration Tool (ACT) is ALICE first step to achieve a high level of automation, implementing automatic configuration and calibration of the sub-detectors and online systems. This presentation describes the goals and architecture of the ACT, the web-based Human Interface and the commissioning performed before the start of the collisions. It also reports on the first experiences with real use in daily operations, and finally it presents the road-map for future developments.

  18. Aquarius Main Structure Configuration

    Science.gov (United States)

    Eremenko, Alexander

    2012-01-01

    The Aquarius/SAC-D Observatory is a joint US-Argentine mission to map the salinity at the ocean surface. This information is critical to improving our understanding of two major components of Earth's climate system - the water cycle and ocean circulation. By measuring ocean salinity from space, the Aquarius/SAC-D Mission will provide new insights into how the massive natural exchange of freshwater between the ocean, atmosphere and sea ice influences ocean circulation, weather and climate. Aquarius is the primary instrument on the SAC-D spacecraft. It consists of a Passive Microwave Radiometer to detect the surface emission that is used to obtain salinity and an Active Scatterometer to measure the ocean waves that affect the precision of the salinity measurement. The Aquarius Primary Structure houses instrument electronics, feed assemblies, and supports a deployable boom with a 2.5 m Reflector, and provides the structural interface to the SAC-D Spacecraft. The key challenge for the Aquarius main structure configuration is to satisfy the needs of component accommodations, ensuring that the instrument can meet all operational, pointing, environmental, and launch vehicle requirements. This paper describes the evolution of the Aquarius main structure configuration, the challenges of balancing the conflicting requirements, and the major configuration driving decisions and compromises.

  19. Advanced Plasma Shape Control to Enable High-Performance Divertor Operation on NSTX-U

    Science.gov (United States)

    Vail, Patrick; Kolemen, Egemen; Boyer, Mark; Welander, Anders

    2017-10-01

    This work presents the development of an advanced framework for control of the global plasma shape and its application to a variety of shape control challenges on NSTX-U. Operations in high-performance plasma scenarios will require highly-accurate and robust control of the plasma poloidal shape to accomplish such tasks as obtaining the strong-shaping required for the avoidance of MHD instabilities and mitigating heat flux through regulation of the divertor magnetic geometry. The new control system employs a high-fidelity model of the toroidal current dynamics in NSTX-U poloidal field coils and conducting structures as well as a first-principles driven calculation of the axisymmetric plasma response. The model-based nature of the control system enables real-time optimization of controller parameters in response to time-varying plasma conditions and control objectives. The new control scheme is shown to enable stable and on-demand plasma operations in complicated magnetic geometries such as the snowflake divertor. A recently-developed code that simulates the nonlinear evolution of the plasma equilibrium is used to demonstrate the capabilities of the designed shape controllers. Plans for future real-time implementations on NSTX-U and elsewhere are also presented. Supported by the US DOE under DE-AC02-09CH11466.

  20. Thermal analysis of an exposed tungsten edge in the JET divertor

    Energy Technology Data Exchange (ETDEWEB)

    Arnoux, G., E-mail: gilles.arnoux@ccfe.ac.uk [CCFE Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Coenen, J. [Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung – Plasmaphysik, 52425 Jülich (Germany); Bazylev, B. [Forshungzentrum Karlsruhe GmbH, P.O.Box 3640, D-76021 Karlsruhe (Germany); Corre, Y. [CEA/DSM/IRFM, CEA Cadarache, 13108 Saint Paul Lez Durance (France); Matthews, G.F.; Balboa, I. [CCFE Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Clever, M. [Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung – Plasmaphysik, 52425 Jülich (Germany); Dejarnac, R. [IPP.CR, Institute of Plasma Physics AS CR, Za Slovankou 3, 182 21 Praha 8 (Czech Republic); Devaux, S.; Eich, T. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Gauthier, E. [CEA/DSM/IRFM, CEA Cadarache, 13108 Saint Paul Lez Durance (France); Frassinetti, L. [Fusion Plasma Physics, EES, KTH, SE-10044 Stockholm (Sweden); Horacek, J. [IPP.CR, Institute of Plasma Physics AS CR, Za Slovankou 3, 182 21 Praha 8 (Czech Republic); Jachmich, S. [Laboratory for Plasma Physics Koninklijke Militaire School – Ecole Royale Militaire, Renaissancelaan, 30 Avenue de la Renaissance, B-1000 Brussels (Belgium); Kinna, D. [CCFE Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Marsen, S. [Max-Planck-Institut für Plasmaphysik, Teilinsitut Greifswald, D-17491 Greifswald (Germany); and others

    2015-08-15

    Highlights: • We provide experimental evidences that melting of the JET tungsten divertor is achieved by transients only. • The measurements show that less than half the parallel heat flux reaches the melted sample. • We propose ideas to investigate to explain the missing heat flux. - Abstract: In the recent melt experiments with the JET tungsten divertor, we observe that the heat flux impacting on a leading edge is 3–10 times lower than a geometrical projection would predict. The surface temperature, tungsten vaporisation rate and melt motion measured during these experiments is consistent with the simulations using the MEMOS code, only if one applies the heat flux reduction. This unexpected observation is the result of our efforts to demonstrate that the tungsten lamella was melted by ELM induced transient heat loads only. This paper describes in details the measurements and data analysis method that led us to this strong conclusion. The reason for the reduced heat flux are yet to be clearly established and we provide some ideas to explore. Explaining the physics of this heat flux reduction would allow to understand whether it can be extrapolated to ITER.

  1. Molecular activated recombination in divertor simulation plasma on GAMMA 10/PDX

    Directory of Open Access Journals (Sweden)

    M. Sakamoto

    2017-08-01

    Full Text Available In the tandem mirror GAMMA 10/PDX, molecular activated recombination (MAR leading to plasma detachment has been observed by additional hydrogen gas injection to the divertor simulation plasma (i.e. end loss plasma which is exposed to the V-shaped target in the divertor simulation experimental module (D-module. The temperature near the corner of the V-shaped target decreased from ∼23eV to ∼2eV as the neutral pressure in the D-module increased. A clear density rollover was observed at ∼2Pa. A position of the density maximum moves to upstream of the plasma with increase in the neutral pressure and the density near the corner of the target decreases to detach the plasma from the target. After the occurrence of the density rollover, the Balmer β intensity decreases as with the density but the Balmer α intensity continues to increase, indicating the dissociative attachment process in MAR is more dominant than the ion conversion process although the rate coefficient of the former process is lower than that of the latter one, which is calculated by using a collisional radiative model. This would be caused by the MAR process related to triatomic hydrogen molecules which significantly contributed to the detachment process.

  2. Divertor simulation experiment and its future research plan making use of a large tandem mirror device

    Energy Technology Data Exchange (ETDEWEB)

    Nakashima, Y., E-mail: nakashma@prc.tsukuba.ac.jp [Plasma Research Center, University of Tsukuba, Tsukuba, Ibaraki 305-8577 (Japan); Takeda, H.; Hosoi, K.; Yonenaga, R.; Katanuma, I.; Ichimura, K.; Ichimura, M.; Imai, T.; Ishii, T.; Kariya, T.; Kiwamoto, Y.; Minami, R.; Miyata, Y.; Ozawa, H.; Shidara, H.; Yamaguchi, Y.; Yoshikawa, M. [Plasma Research Center, University of Tsukuba, Tsukuba, Ibaraki 305-8577 (Japan); Asakura, N. [Japan Atomic Energy Agency, Naka Fusion Institute, 801-1 Mukouyama, Naka 311-0193 (Japan); Hatayama, A. [Faculty of Science and Technology, Keio University, Kanagawa 220-8522 (Japan); Higashizono, Y. [RIAM, Kyushu University, 87, Kasuga, Fukuoka 816-8580 (Japan)

    2011-08-01

    Divertor simulation study has been started as a new research plan, by making best use of a large linear plasma device. The experiment of generating the plasma flow with high heat and particle flux was successfully performed at an end-mirror exit of the GAMMA 10 tandem mirror. In typical hot-ion-mode plasmas, the heat-flux density of 0.6 MW/m{sup 2} and the particle-flux density of 10{sup 22} particles/s m{sup 2} were simultaneously achieved in the case of only ICRF heating and superimposing the 300 kW ECH pulse attained the peak value of the net heat-flux up to 8 MW/m{sup 2} on axis. The above experimental results and the simulation analysis of ICRF heating using the Fokker-Planck code give a clear prospect of generating the required performance for divertor studies by building up the plasma heating systems to the end-mirror cell. Detailed behavior of the plasma flow and the future research plan are also described.

  3. Electron temperature and heat load measurements in the COMPASS divertor using the new system of probes

    Science.gov (United States)

    Adamek, J.; Seidl, J.; Horacek, J.; Komm, M.; Eich, T.; Panek, R.; Cavalier, J.; Devitre, A.; Peterka, M.; Vondracek, P.; Stöckel, J.; Sestak, D.; Grover, O.; Bilkova, P.; Böhm, P.; Varju, J.; Havranek, A.; Weinzettl, V.; Lovell, J.; Dimitrova, M.; Mitosinkova, K.; Dejarnac, R.; Hron, M.; The COMPASS Team; The EUROfusion MST1 Team

    2017-11-01

    A new system of probes was recently installed in the divertor of tokamak COMPASS in order to investigate the ELM energy density with high spatial and temporal resolution. The new system consists of two arrays of rooftop-shaped Langmuir probes (LPs) used to measure the floating potential or the ion saturation current density and one array of Ball-pen probes (BPPs) used to measure the plasma potential with a spatial resolution of ~3.5 mm. The combination of floating BPPs and LPs yields the electron temperature with microsecond temporal resolution. We report on the design of the new divertor probe arrays and first results of electron temperature profile measurements in ELMy H-mode and L-mode. We also present comparative measurements of the parallel heat flux using the new probe arrays and fast infrared termography (IR) data during L-mode with excellent agreement between both techniques using a heat power transmission coefficient γ  =  7. The ELM energy density {{\\varepsilon }\\parallel } was measured during a set of NBI assisted ELMy H-mode discharges. The peak values of {{\\varepsilon }\\parallel } were compared with those predicted by model and with experimental data from JET, AUG and MAST with a good agreement.

  4. Coil Designs for Novel Magnetic Geometries to Cure the Divertor Heat Flux Problem for Reactors

    Science.gov (United States)

    Pekker, M.; Valanju, P.; Kotschenreuther, M.; Wiley, J. C.; Strickler, D.

    2004-11-01

    Coil designs are developed for novel magnetic divertor geometries with a second axi-symmetric x-point and flux expansion region along the separatrix. Adjacent posters describe how these lead to spreading of heat flux and the possibility of stable, complete detachment to overcome serious physics and engineering problems in reactors. The principal feasibility issue is creating, with simple coils, additional X-points on the separatrix without extensively deforming the magnetic field in the main plasma. For the spherical tokamak NSTX, we show that adding one or two poloidal coils suffices to create a divergent flux at the divertor, i.e., a new x-point. The currents and forces for the extra coils are small. We also modify ARIES ST design to show reactor feasibility. Optimized coil designs for PEGASUS, ARIES RS/AT, and a modular ITER retrofit are also being developed. For our calculations we used self consistent code FBEQ, which was used to design NSTX. We also use NCSX tools for optimization of designs with competing physics and engineering constraints.

  5. Research proposal on: amplitude modulated reflectometry system for the JET divertor

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez, J.; Branas, B.; Estrada, T.; Luna, E. de la

    1992-07-01

    Amplitude Modulated reflectometry is presented here as a tool for density profile measurements in the JET divertor plasmas. One of the main problems which has been present in most reflectometers during the last years is the need for a coherent tracking of the phase delay: fast density fluctuations and strong modulation on the amplitude of the reflected signal usually bring to fringe jumps in the phase signal, which are a big problem when the phase values are much larger than 2{pi} The conditions in the JET divertor plasmas: plasma geometry, access and long oversized broad- band waveguide paths makes very difficult the phase measurements at the millimeter wave range. AM reflectometry is to some extension an intermediate solution between the classical phase delay reflectometry, so far applied to small distances, and the time domain reflectometry, used for onospheric studies and recently also proposed for fusion plasmas. The main advantage is to allow the use of millimeter wave reflectometry with moderate phase shifts ( {approx} 2{pi} ). (Author) 2 refs.

  6. Exfoliation of the tungsten fibreform nanostructure by unipolar arcing in the LHD divertor plasma

    Science.gov (United States)

    Tokitani, M.; Kajita, S.; Masuzaki, S.; Hirahata, Y.; Ohno, N.; Tanabe, T.; LHD Experiment Group

    2011-10-01

    The tungsten nanostructure (W-fuzz) created in the linear divertor simulator (NAGDIS) was exposed to the Large Helical Device (LHD) divertor plasma for only 2 s (1 shot) to study exfoliation/erosion and microscopic modifications due to the high heat/particle loading under high magnetic field conditions. Very fine and randomly moved unipolar arc trails were clearly observed on about half of the W-fuzz area (6 × 10 mm2). The fuzzy surface was exfoliated by continuously moving arc spots even for the very short exposure time. This is the first observation of unipolar arcing and exfoliation of some areas of the W-fuzz structure itself in a large plasma confinement device with a high magnetic field. The typical width and depth of each arc trail were about 8 µm and 1 µm, respectively, and the arc spots moved randomly on the micrometre scale. The fractality of the arc trails was analysed using a box-counting method, and the fractal dimension (D) of the arc trails was estimated to be D ≈ 1.922. This value indicated that the arc spots moved in Brownian motion, and were scarcely influenced by the magnetic field. One should note that such a large scale exfoliation due to unipolar arcing may enhance the surface erosion of the tungsten armour and act as a serious impurity source for fusion plasmas.

  7. Improving concept design of divertor support system for FAST tokamak using TRIZ theory and AHP approach

    Energy Technology Data Exchange (ETDEWEB)

    Di Gironimo, G., E-mail: giuseppe.digironimo@unina.it [Association Euratom/ENEA/CREATE, Università di Napoli Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Carfora, D.; Esposito, G.; Labate, C.; Mozzillo, R.; Renno, F.; Lanzotti, A. [Association Euratom/ENEA/CREATE, Università di Napoli Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Siuko, M. [VTT Systems Engineering, Tekniikankatu 1, 33720 Tampere (Finland)

    2013-11-15

    Highlights: • Optimization of the RH system for the FAST divertor using TRIZ. • Participative design approach using virtual reality. • Comparison of product alternatives in an immersive virtual reality environment. • Prioritization of concept alternatives based on AHP. -- Abstract: The paper focuses on the application of the Theory of Inventive Problem Solving (TRIZ) to divertor Remote Handling (RH) issues in Fusion Advanced Studies Torus (FAST), a satellite tokamak acting as a test bed for the study and the development of innovative technologies oriented to ITER and DEMO programs. The objective of this study consists in generating concepts or solutions able to overcome design and technical weak points in the current maintenance procedure. Two different concepts are designed with the help of a parametric CAD software, CATIA V5, using a top-down modeling approach; kinematic simulations of the remote handling system are performed using Digital Mock-Up (DMU) capabilities of the software. The evaluation of the concepts is carried out involving a group of experts in a participative design approach using virtual reality, classifying the concepts with the help of the Analytical Hierarchy Process (AHP)

  8. Dynamics and stability of divertor detachment in H-mode plasmas on JET

    Science.gov (United States)

    Field, A. R.; Balboa, I.; Drewelow, P.; Flanagan, J.; Guillemaut, C.; Harrison, J. R.; Huber, A.; Huber, V.; Lipschultz, B.; Matthews, G.; Meigs, A.; Schmitz, J.; Stamp, M.; Walkden, N.; contributors, JET

    2017-09-01

    The dynamics and stability of divertor detachment in {{{N}}}2 seeded, type-I, ELMy H-mode plasmas with dominant NBI heating in the JET ITER-like wall device is studied by means of an integrated analysis of diagnostic data from several systems, classifying data relative to the ELM times. It is thereby possible to study the response of the detachment evolution to the control parameters (SOL input power, upstream density and impurity fraction) prevailing during the inter-ELM periods and the effect of ELMs on the detached divertor. A relatively comprehensive overview is achieved, including the interaction with the targets at various stages of the ELM cycle, the role of ELMs in affecting the detachment process and the overall performance of the scenario. The results are consistent with previous studies in devices with an ITER-like, metal wall, with the important advance of distinguishing data from intra- and inter-ELM periods. Operation without significant degradation of the core confinement can be sustained in the presence of strong radiation from the x-point region (MARFE).

  9. Quantitative thermal imperfection definition using non-destructive infrared thermography on an advanced DEMO divertor concept

    Science.gov (United States)

    Gallay, F.; Richou, M.; Vignal, N.; Lenci, M.; Roccella, S.; Kermouche, G.; Visca, E.; You, J. H.

    2017-12-01

    The future DEMO divertor is currently under conceptual design within the European Consortium. In this regard, several concepts have been proposed and mock-ups have been fabricated to investigate their thermo-mechanical behaviour. Indeed, as a key plasma facing component, the divertor will have to withstand extreme thermal loads (up to 20 MW m-2 during slow transient events) and will have to be able to exhaust a large amount of heat. The presence of structural defects in the component may significantly affect the thermal response and must therefore be considered. A non-destructive technique based on infrared thermography is proposed here to detect defects in mock-ups where graded material was used as an interlayer between the heatsink material and the armor material. Two methods to characterize the size and location of such defects are presented. It was shown that finite element analysis combined with experimental data from infrared thermography, provides accurate means to assess quantitatively the size and position of thermal imperfections.

  10. Pre-irradiation testing of actively cooled Be-Cu divertor modules

    Energy Technology Data Exchange (ETDEWEB)

    Linke, J.; Duwe, R.; Kuehnlein, W. [Forschungszentrum Juelich GmbH (Germany)] [and others

    1995-09-01

    A set of neutron irradiation tests is prepared on different plasma facing materials (PFM) candidates and miniaturized components for ITER. Beside beryllium the irradiation program which will be performed in the High Flux Reactor (HFR) in Petten, includes different carbon fiber composites (CFQ) and tungsten alloys. The target values for the neutron irradiation will be 0.5 dpa at temperatures of 350{degrees}C and 700{degrees}C, resp.. The post irradiation examination (PIE) will cover a wide range of mechanical tests; in addition the degradation of thermal conductivity will be investigated. To determine the high heat flux (HHF) performance of actively cooled divertor modules, electron beam tests which simulate the expected heat loads during the operation of ITER, are scheduled in the hot cell electron beam facility JUDITH. These tests on a selection of different actively cooled beryllium-copper and CFC-copper divertor modules are performed before and after neutron irradiation; the pre-irradiation testing is an essential part of the program to quantify the zero-fluence high heat flux performance and to detect defects in the modules, in particular in the brazed joints.

  11. Impurity re-distribution in the corner regions of the JET divertor

    Science.gov (United States)

    Widdowson, A.; Coad, J. P.; Alves, E.; Baron-Wiechec, A.; Barradas, N. P.; Catarino, N.; Corregidor, V.; Heinola, K.; Krat, S.; Likonen, J.; Matthews, G. F.; Mayer, M.; Petersson, P.; Rubel, M.; Contributors, JET

    2017-12-01

    The International Thermonuclear Experimental Reactor (ITER) will use a mixture of deuterium (D) and tritium (T) as the fuel to generate power. Since T is both radioactive and expensive the Joint European Torus (JET) has been at the forefront of research to discover how much T is used and where it may be retained within the main reaction chamber. Until the year 2010 the JET plasma facing components were constructed of carbon fibre composites. During the JET carbon (C) phases impurities accumulated at the corners of the divertor located towards the bottom of the chamber in regions shadowed from the plasma where they are very difficult to reach and remove. This build-up of C and the associated H-isotope (including T) retention were of particular concern for future fusion reactors therefore, in 2010 JET changed the wall protection to (mainly) Be and the divertor to tungsten (W)—the JET ITER-like wall (ILW)—the choice of materials for ITER. This paper reveals that with the JET ILW impurities are still accumulating in the shadowed regions, with Be being the majority element, though the overall quantities are very much reduced from those in the C phases. Material will be transported into the shadowed regions principally when the plasma strike points are on the corner tiles, but particles typically have about a 75% probability of reflection from line-of sight surfaces, and multiple reflection/scattering results in deposition over all surfaces.

  12. High heat flux testing of EU tungsten monoblock mock-ups for the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Gavila, P., E-mail: pierre.gavila@f4e.europa.eu [Fusion for Energy, 08019 Barcelona (Spain); Riccardi, B. [Fusion for Energy, 08019 Barcelona (Spain); Pintsuk, G. [Forschungszentrum Juelich, 52425 Juelich (Germany); Ritz, G. [AREVA NP, Centre Technique France, 71205 Le Creusot (France); Kuznetsov, V. [JCS “Efremov Institute”, Doroga na Metallostroy 3, Metallostroy, Saint-Petersburg 196641 (Russian Federation); Durocher, A. [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 Saint Paul-lez-Durance (France)

    2015-10-15

    Highlights: • All the tested items sustained the ITER Full W divertor qualification program requirements. This confirms that the technology for the manufacturing of the first set of the ITER Divertor is available in Europe. • The surface roughening and local melting of the W surface under high heat flux was proven to be significantly reduced for an armour thickness lower or equal to 6 mm. • However, this campaign highlighted some specific areas of improvement to be implemented ideally before the upcoming ITER Divertor IVT serial production. • The issue of the self-castellation of the W monoblocks, which typically appears after a few tenths of cycles at 20 MW/m{sup 2}, is critical because it generates some uncontrolled defects at the amour to heat sink joints. Besides, they create a gap which exposure is almost perpendicular to the magnetic field lines and which might lead to local W melting in the strike point region. • This campaign also evidenced that the minimum IO requirements on the CuCrZr ductility could be revised to avoid the occurrence of rather early fatigue failures. Although the W material characterization program has been set up by the IO, the strategy on the CuCrZr still needs to be defined. - Abstract: With the aim to assess the option to start the ITER operation with a full tungsten divertor, an R&D program was launched in order to evaluate the performances of tungsten (W) armoured plasma facing components (PFCs) under high heat flux. The F4E program consisted in the manufacturing and high heat flux (HHF) testing of W monoblock mock-ups and medium scale prototypes up to 20 MW/m{sup 2}. During the test campaign, 26 W mock-ups and two medium scale prototypes manufactured by Plansee SE (Austria) and by Ansaldo Nucleare (Italy) have been tested at the FE200 (AREVA, Le Creusot, France) and ITER Divertor Test Facility (IDTF) (Efremov Institute Saint Petersburg, Russian Federation) electron beam test facilities. The high heat flux (HHF) testing

  13. Effects of carbon impurity on deuterium retention in VPS-tungsten coatings exposed to JT-60U divertor plasmas

    Science.gov (United States)

    Fukumoto, M.; Nakano, T.; Itami, K.; Wada, T.; Ueda, Y.; Tanabe, T.

    2011-08-01

    Carbon eroded from carbon armor tiles during plasma discharge was implanted into and accumulated in tungsten coating exposed to JT-60U divertor plasmas. The D/C ratio of 0.06 ± 0.02 evaluated in the tungsten coating was half to one-quarter that in carbon codeposits formed at similar temperature of the tungsten coating. These results suggest that simultaneous use of carbon and tungsten coating would enhance tritium retention in the tungsten coating in future deuterium-tritium fusion devices. To investigate the carbon diffusion mechanism in the tungsten coating exposed to JT-60U divertor plasmas, the carbon diffusion coefficient in tungsten coating was measured by tracer methods. Using the apparent carbon diffusion coefficient obtained in this study (˜8 × 10-19 m2/s), the carbon diffusion length in the tungsten coating exposed to JT-60U divertor plasmas was evaluated to ˜100 nm. This diffusion length was quite shorter than that observed in the tungsten coating exposed to JT-60U divertor plasmas. Therefore, it remains possible that diffusion of implanted carbon in tungsten coating would be enhanced by other diffusion mechanisms which did not arise in the diffusion experiments or heat loads to the tungsten coating during transient events and plasma discharges with a strike point positioned on the tungsten-coated tiles.

  14. 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.

  15. Investigation of the influence of divertor recycling on global plasma confinement in JET ITER-like wall

    NARCIS (Netherlands)

    Tamain, P.; Joffrin, E.; Bufferand, H.; Jarvinen, A.; Brezinsek, S.; Ciraolo, G.; Delabie, E.; Frassinetti, L.; Giroud, C.; Groth, M.; Lipschultz, B.; Lomas, P.; Marsen, S.; Menmuir, S.; Oberkofler, M.; Stamp, M.; Wiesen, S.; JET-EFDA Contributors,

    2015-01-01

    Abstract The impact of the divertor geometry on global plasma confinement in type I ELMy H-mode has been investigated in the JET tokamak equipped with ITER-Like Wall. Discharges have been performed in which the position of the strike-points was changed while keeping the bulk plasma equilibrium

  16. Configuration and Heating Power Dependence of Edge Parameters and H-mode Dynamics in National Spherical Torus Experiment (NSTX)

    Energy Technology Data Exchange (ETDEWEB)

    C.E. Bush; M.G. Bell; R.E. Bell; J. Boedo; E.D. Fredrickson; S.M. Kaye; S. Kubota; B.P. LeBlanc; R. Maingi; R.J. Maqueda; S.A. Sabbagh; V.A. Soukhanovskii; D. Stutman; D.W. Swain; J.B. Wilgen; S.J. Zweben; W.M. Davis; D.A. Gates; D.W. Johnson; R. Kaita; H.W. Kugel; D. Mastrovito; S. Medley; J.E. Menard; D. Mueller; M. Ono; F. Paoletti; S.J. Paul; Y-K.M. Peng; R. Raman; P.G. Roney; A.L. Roquemore; C.H. Skinner; E.J. Synakowski; G. Taylor; the NSTX Team

    2003-01-09

    Edge parameters play a critical role in H-mode (high-confinement mode) access, which is a key component of plasma discharge optimization in present-day toroidal confinement experiments and the design of next-generation devices. Because the edge magnetic topology of a spherical torus (ST) differs from a conventional aspect ratio tokamak, H-modes in STs exhibit important differences compared with tokamaks. The dependence of the NSTX (National Spherical Torus Experiment) edge plasma on heating power, including the L-H transition requirements and the occurrence of edge-localized modes (ELMs), and on divertor configuration is quantified. Comparisons between good L-modes (low-confinement modes) and H-modes show greater differences in the ion channel than the electron channel. The threshold power for the H-mode transition in NSTX is generally above the predictions of a recent ITER (International Thermonuclear Experimental Reactor) scaling. Correlations of transition and ELM phenomena with turbulent fluctuations revealed by Gas Puff Imaging (GPI) and reflectometry are observed. In both single-null and double-null divertor discharges, the density peaks off-axis, sometimes developing prominent ''ears'' which can be sustained for many energy confinement times, tau subscript ''E'', in the absence of ELMs. A wide variety of ELM behavior is observed, and ELM characteristics depend on configuration and fueling.

  17. Configuration of Spontaneous Flows in a Plasma Coherent Structure

    Science.gov (United States)

    Hasegawa, Hiroki; Ishiguro, Seiji

    2013-10-01

    Spontaneous particle flows in a plasma coherent structure (blob) have been studied with a three dimensional electrostatic plasma particle simulation code. In the particle simulation, the particle absorbing boundaries are placed on the both ends in the z direction (corresponding to the toroidal direction) and one end in the x direction (corresponding to the counter radial direction). The former boundaries and the latter one corresponds to end plates (divertor plates) and the first wall. A coherent structure is initially set as a column along the external magnetic field and propagates to the first wall because the magnetic field is set as ∂B / ∂x ≠ 0 . In this study, we have investigated the configuration of spontaneous particle flows in a plasma coherent structure. We have found that a spiral current system in a plasma blob and characteristic features in the velocity distribution of plasma particles in a blob. Supported by NIFS Collaboration Research programs (NIFS13KNSS038 and NIFS13KNXN258) and a Grant-in-Aid for Scientific Research from Japan Society for the Promotion of Science (KAKENHI 23740411).

  18. Configurational Entropy Revisited

    Science.gov (United States)

    Lambert, Frank L.

    2007-09-01

    Entropy change is categorized in some prominent general chemistry textbooks as being either positional (configurational) or thermal. In those texts, the accompanying emphasis on the dispersal of matter—independent of energy considerations and thus in discord with kinetic molecular theory—is most troubling. This article shows that the variants of entropy can be treated from a unified viewpoint and argues that to decrease students' confusion about the nature of entropy change these variants of entropy should be merged. Molecular energy dispersal in space is implicit but unfortunately tacit in the cell models of statistical mechanics that develop the configurational entropy change in gas expansion, fluids mixing, or the addition of a non-volatile solute to a solvent. Two factors are necessary for entropy change in chemistry. An increase in thermodynamic entropy is enabled in a process by the motional energy of molecules (that, in chemical reactions, can arise from the energy released from a bond energy change). However, entropy increase is only actualized if the process results in a larger number of arrangements for the system's energy, that is, a final state that involves the most probable distribution for that energy under the new constraints. Positional entropy should be eliminated from general chemistry instruction and, especially benefiting "concrete minded" students, it should be replaced by emphasis on the motional energy of molecules as enabling entropy change.

  19. Ames Optimized TCA Configuration

    Science.gov (United States)

    Cliff, Susan E.; Reuther, James J.; Hicks, Raymond M.

    1999-01-01

    Configuration design at Ames was carried out with the SYN87-SB (single block) Euler code using a 193 x 49 x 65 C-H grid. The Euler solver is coupled to the constrained (NPSOL) and the unconstrained (QNMDIF) optimization packages. Since the single block grid is able to model only wing-body configurations, the nacelle/diverter effects were included in the optimization process by SYN87's option to superimpose the nacelle/diverter interference pressures on the wing. These interference pressures were calculated using the AIRPLANE code. AIRPLANE is an Euler solver that uses a unstructured tetrahedral mesh and is capable of computations about arbitrary complete configurations. In addition, the buoyancy effects of the nacelle/diverters were also included in the design process by imposing the pressure field obtained during the design process onto the triangulated surfaces of the nacelle/diverter mesh generated by AIRPLANE. The interference pressures and nacelle buoyancy effects are added to the final forces after each flow field calculation. Full details of the (recently enhanced) ghost nacelle capability are given in a related talk. The pseudo nacelle corrections were greatly improved during this design cycle. During the Ref H and Cycle 1 design activities, the nacelles were only translated and pitched. In the cycle 2 design effort the nacelles can translate vertically, and pitch to accommodate the changes in the lower surface geometry. The diverter heights (between their leading and trailing edges) were modified during design as the shape of the lower wing changed, with the drag of the diverter changing accordingly. Both adjoint and finite difference gradients were used during optimization. The adjoint-based gradients were found to give good direction in the design space for configurations near the starting point, but as the design approached a minimum, the finite difference gradients were found to be more accurate. Use of finite difference gradients was limited by the

  20. Development of remote pipe welding tool for divertor cassettes in JT-60SA

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Takao, E-mail: hayashi.takao@jaea.go.jp [Fusion Research and Development Directorate, Japan Atomic Energy Agency, Naka (Japan); Sakurai, Shinji; Sakasai, Akira; Shibanuma, Kiyoshi [Fusion Research and Development Directorate, Japan Atomic Energy Agency, Naka (Japan); Kono, Wataru; Ohnawa, Toshio; Matsukage, Takeshi [Toshiba Corporation, Yokohama, Kanagawa (Japan)

    2015-12-15

    Highlights: • Remote pipe welding tool accessing from inside of the pipe has been newly developed. • Cooling pipe with a jut on the edge expands the acceptable welding gap up to 0.5 mm. • Positioning accuracy of the laser beam is realized to be less than ±0.1 mm. • We have achieved robust welding for an angular misalignment of 0.5°. - Abstract: Remote pipe welding tool accessing from inside of the pipe has been newly developed for JT-60SA. Remote handling (RH) system is necessary for the maintenance and repair of the divertor cassette in JT-60SA. Because the space around the cooling pipe connected with the divertor cassette is very limited, the cooling pipe is to be remotely cut and welded from inside for the maintenance. A laser welding method was employed to perform circumferential welding by rotating the focusing mirror inside the pipe. However, the grooves of connection pipes are not precisely aligned for welding. The most critical issue is therefore to develop a reliable welding tool for pipe connection without a defect such as undercut weld due to a gap caused by angular and axial misalignments of grooves. In addition, an angular misalignment between two pipes due to inclination of pipe has to be taken into account for the positioning of the laser beam during welding. In this paper, the followings are proposed to solve the above issues: (1) Cooling pipe connected with the divertor is machined to have a jut on the edge so as to expand the acceptable welding gap up to 0.5 mm by filling the gap with welded jut. (2) Positioning accuracy of the laser beam for reliable welding is realized to be less than ±0.1 mm along the circumferential target for welding by a position control mechanism provided in the tool even in the case of up to angular misalignment of 0.5° between connection pipes. Based on the above proposals, we have achieved robust welding for a large gap up to 0.5 mm as well as the maximum angular misalignment of 0.5° between connection pipes

  1. TWRS Configuration management program plan

    Energy Technology Data Exchange (ETDEWEB)

    Vann, J.M.

    1996-06-03

    The TWRS Configuration Management Program Plan (CMPP) integrates technical and administrative controls to establish and maintain consistency among requirements, product configuration, and product information for TWRS products during all life cycle phases. This CMPP will be used by TWRS management and configuration management personnel to establish and manage the technical and integrated baselines and controls and status changes to those baselines.

  2. A novel plasma position and shape controller for advanced configuration development on the TCV tokamak

    Science.gov (United States)

    Anand, H.; Coda, S.; Felici, F.; Galperti, C.; Moret, J.-M.

    2017-12-01

    A novel plasma position and shape controller has been developed for the highly flexible shaping poloidal-field coil set of the TCV tokamak, to aid in the precise control of advanced configurations such as negative-triangularity plasmas, snowflake and super-X divertors, and doublets. This work follows and relies on the deployment of a new, sub-ms, real-time magnetic equilibrium-reconstruction algorithm. The controller formulation ensures flexibility through an ordering of controlled variables from the most easily to the least easily controlled, while respecting the hardware limits on the poloidal-field coil currents. A rigid, linearised plasma response model for the TCV tokamak is used for the verification and determination of the control parameters. The controller has been applied successfully to a variety of TCV plasma discharges.

  3. Advanced Biasing Experiments on the C-2 Field-Reversed Configuration Device

    Science.gov (United States)

    Thompson, Matthew; Korepanov, Sergey; Garate, Eusebio; Yang, Xiaokang; Gota, Hiroshi; Douglass, Jon; Allfrey, Ian; Valentine, Travis; Uchizono, Nolan; TAE Team

    2014-10-01

    The C-2 experiment seeks to study the evolution, heating and sustainment effects of neutral beam injection on field-reversed configuration (FRC) plasmas. Recently, substantial improvements in plasma performance were achieved through the application of edge biasing with coaxial plasma guns located in the divertors. Edge biasing provides rotation control that reduces instabilities and E × B shear that improves confinement. Typically, the plasma gun arcs are run at ~ 10 MW for the entire shot duration (~ 5 ms), which will become unsustainable as the plasma duration increases. We have conducted several advanced biasing experiments with reduced-average-power plasma gun operating modes and alternative biasing cathodes in an effort to develop an effective biasing scenario applicable to steady state FRC plasmas. Early results show that several techniques can potentially provide effective, long-duration edge biasing.

  4. SASSI system software configuration

    Energy Technology Data Exchange (ETDEWEB)

    Weiner, E.O.

    1994-08-01

    The SASSI (System for Analysis for Soil-Structure Interaction) computer program was obtained by WHC from the University of California at Berkeley for seismic structural analysis of complex embedded building configurations. SASSI was developed in the 1980`s by a team of doctoral students under the direction of Prof. J. Lysmer. The program treats three-dimensional soil-structure interaction problems with the flexible volume substructuring method. In the 1970`s, the same organization developed the FLUSH program, which has achieved widespread international usage in the seismic analysis of structures. SASSI consists of nine modules, each of which are to be run as a separate execution. The SASSI source code, dated 1989 and identified as a Cray version, was put up on the RL Cray XM/232 Unicos system in 1991. That system was removed at the end of 1993, and SASSI is now installed on the LANL Cray YMP systems.

  5. Configuring the autism epidemic

    DEFF Research Database (Denmark)

    Seeberg, Jens; Christensen, Fie Lund Lindegaard

    2017-01-01

    Autism has been described as an epidemic, but this claim is contested and may point to an awareness epidemic, i.e. changes in the definition of what autism is and more attention being invested in diagnosis leading to a rise in registered cases. The sex ratio of children diagnosed with autism...... is skewed in favour of boys, and girls with autism tend to be diagnosed much later than boys. Building and further developing the notion of ‘configuration’ of epidemics, this article explores the configuration of autism in Denmark, with a particular focus on the health system and social support to families...... with children diagnosed with autism, seen from a parental perspective. The article points to diagnostic dynamics that contribute to explaining why girls with autism are not diagnosed as easily as boys. We unfold these dynamics through the analysis of a case of a Danish family with autism....

  6. Configuring the autism epidemic

    DEFF Research Database (Denmark)

    Christensen, Fie Lund Lindegaard; Seeberg, Jens

    2017-01-01

    is skewed in favour of boys, and girls with autism tend to be diagnosed much later than boys. Building and further developing the notion of ‘configuration’ of epidemics, this article explores the configuration of autism in Denmark, with a particular focus on the health system and social support to families...... with children diagnosed with autism, seen from a parental perspective. The article points to diagnostic dynamics that contribute to explaining why girls with autism are not diagnosed as easily as boys. We unfold these dynamics through the analysis of a case of a Danish family with autism.......Autism has been described as an epidemic, but this claim is contested and may point to an awareness epidemic, i.e. changes in the definition of what autism is and more attention being invested in diagnosis leading to a rise in registered cases. The sex ratio of children diagnosed with autism...

  7. Configurable Aperture Space Telescope

    Science.gov (United States)

    Ennico, Kimberly; Vassigh, Kenny; Bendek, Selman; Young, Zion W; Lynch, Dana H.

    2015-01-01

    In December 2014, we were awarded Center Innovation Fund to evaluate an optical and mechanical concept for a novel implementation of a segmented telescope based on modular, interconnected small sats (satlets). The concept is called CAST, a Configurable Aperture Space Telescope. With a current TRL is 2 we will aim to reach TLR 3 in Sept 2015 by demonstrating a 2x2 mirror system to validate our optical model and error budget, provide strawman mechanical architecture and structural damping analyses, and derive future satlet-based observatory performance requirements. CAST provides an alternative access to visible andor UV wavelength space telescope with 1-meter or larger aperture for NASA SMD Astrophysics and Planetary Science community after the retirement of HST.

  8. Power converter connection configuration

    Science.gov (United States)

    Beihoff, Bruce C.; Kehl, Dennis L.; Gettelfinger, Lee A.; Kaishian, Steven C.; Phillips, Mark G.; Radosevich, Lawrence D.

    2008-11-11

    EMI shielding is provided for power electronics circuits and the like via a direct-mount reference plane support and shielding structure. The thermal support may receive one or more power electronic circuits. The support may aid in removing heat from the circuits through fluid circulating through the support. The support forms a shield from both external EMI/RFI and from interference generated by operation of the power electronic circuits. Features may be provided to permit and enhance connection of the circuitry to external circuitry, such as improved terminal configurations. Modular units may be assembled that may be coupled to electronic circuitry via plug-in arrangements or through interface with a backplane or similar mounting and interconnecting structures.

  9. Calculations of Energy Losses due to Atomic Processes in Tokamaks with Applications to the ITER Divertor

    CERN Document Server

    Post, D; Clark, R E H; Putvinskaya, N

    1995-01-01

    Reduction of the peak heat loads on the plasma facing components is essential for the success of the next generation of high fusion power tokamaks such as the International Thermonuclear Experimental Reactor (ITER) 1 . Many present concepts for accomplishing this involve the use of atomic processes to transfer the heat from the plasma to the main chamber and divertor chamber walls and much of the experimental and theoretical physics research in the fusion program is directed toward this issue. The results of these experiments and calculations are the result of a complex interplay of many processes. In order to identify the key features of these experiments and calculations and the relative role of the primary atomic processes, simple quasi-analytic models and the latest atomic physics rate coefficients and cross sections have been used to assess the relative roles of central radiation losses through bremsstrahlung, impurity radiation losses from the plasma edge, charge exchange and hydrogen radiation losses f...

  10. High heat flux test of tungsten brazed mock-ups developed for KSTAR divertor

    Energy Technology Data Exchange (ETDEWEB)

    Song, J.H. [National Fusion Research Institute, Daejeon (Korea, Republic of); Kim, K.M., E-mail: kyungmin@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Hong, S.H.; Kim, H.T.; Park, S.H.; Park, H.K.; Ahn, H.J. [National Fusion Research Institute, Daejeon (Korea, Republic of); Kim, S.K.; Lee, D.W. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-11-01

    The tungsten (W) brazed flat type mock-up which consists of W, OFHC-Cu (oxygen-free high conductive copper) and CuCrZr alloy has been designed for KSTAR divertor in preparation for KSTAR upgrade with 17 MW heating power. For verification of the W brazed mock-up, the high heat flux test is performed at KoHLT-EB (Korea High Heat Load Test Facility-Electron Beam) in KAERI (Korea Atomic Energy Research Institute). Three mock-ups are tested for several thousand thermal cycles with absorbed heat flux up to 5 MW/m{sup 2} for 20 s duration. There is no evidence of the failure at the bonding joints of all mock-ups after HHF test. Finite element analysis (FEA) is performed to interpret the result of the test. As a result, it is considered that the local area in the water is in the subcooled boiling regime.

  11. Ex-vessel break in ITER divertor cooling loop analysis with the ECART code

    CERN Document Server

    Cambi, G; Parozzi, F; Porfiri, MT

    2003-01-01

    A hypothetical double-ended pipe rupture in the ex-vessel section of the International Thermonuclear Experimental Reactor (ITER) divertor primary heat transfer system during pulse operation has been assessed using the nuclear source term ECART code. That code was originally designed and validated for traditional nuclear power plant safety analyses, and has been internationally recognized as a relevant nuclear source term codes for nuclear fission plants. It permits the simulation of chemical reactions and transport of radioactive gases and aerosols under two-phase flow transients in generic flow systems, using a built-in thermal-hydraulic model. A comparison with the results given in ITER Generic Site Safety Report, obtained using a thermal-hydraulic system code (ATHENA), a containment code (INTRA) and an aerosol transportation code (NAUA), in a sequential way, is also presented and discussed.

  12. Damage evaluation under thermal fatigue of a vertical target full scale component for the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Missirlian, M. [Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 Saint Paul Lez Durance cedex (France)]. E-mail: missir@drfc.cad.cea.fr; Escourbiac, F. [Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 Saint Paul Lez Durance cedex (France); Merola, M. [EFDA Close Support Unit, Garching (Germany); Durocher, A. [Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 Saint Paul Lez Durance cedex (France); Bobin-Vastra, I. [FRAMATOME, Le Creusot (France); Schedler, B. [PLANSEE , Aktiengesellschaft-A-6600 Reutte (Austria)

    2007-08-01

    An extensive development programme has been carried out in the EU on high heat flux components within the ITER project. In this framework, a Full Scale Vertical Target (VTFS) prototype was manufactured with all the main features of the corresponding ITER divertor design. The fatigue cycling campaign on CFC and W armoured regions, proved the capability of such a component to meet the ITER requirements in terms of heat flux performances for the vertical target. This paper discusses thermographic examination and thermal fatigue testing results obtained on this component. The study includes thermal analysis, with a tentative proposal to evaluate with finite element approach the location/size of defects and the possible propagation during fatigue cycling.

  13. Evaluation of Nb-base alloys for the divertor structure in fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Purdy, I.M. [Argonne National Laboratory, Upton, IL (United States)

    1996-04-01

    Niobium-base alloys are candidate materials for the divertor structure in fusion reactors. For this application, an alloy should resist aqueous corrosion, hydrogen embrittlement, and radiation damage and should have high thermal conductivity and low thermal expansion. Results of corrosion and embrittlement screening tests of several binary and ternary Nb alloys in high-temperature water indicated the Mb-1Zr, Nb-5MO-1Zr, and Nb-5V-1Z4 (wt %) showed sufficient promise for further investigation. These alloys, together with pure Nb and Zircaloy-4 have been exposed to high purity water containing a low concentration of dissolved oxygen (<12 ppb) at 170, 230, and 300{degrees}C for up to {approx}3200 h. Weight-change data, microstructural observations, and qualitative mechanical-property evaluation reveal that Nb-5V-1Zr is the most promising alloy at higher temperatures. Below {approx}200{degrees}C, the alloys exhibit similiar corrosion behavior.

  14. An FPGA-based bolometer for the MAST-U Super-X divertor.

    Science.gov (United States)

    Lovell, Jack; Naylor, Graham; Field, Anthony; Drewelow, Peter; Sharples, Ray

    2016-11-01

    A new resistive bolometer system has been developed for MAST-Upgrade. It will measure radiated power in the new Super-X divertor, with millisecond time resolution, along 16 vertical and 16 horizontal lines of sight. The system uses a Xilinx Zynq-7000 series Field-Programmable Gate Array (FPGA) in the D-TACQ ACQ2106 carrier to perform real time data acquisition and signal processing. The FPGA enables AC-synchronous detection using high performance digital filtering to achieve a high signal-to-noise ratio and will be able to output processed data in real time with millisecond latency. The system has been installed on 8 previously unused channels of the JET vertical bolometer system. Initial results suggest good agreement with data from existing vertical channels but with higher bandwidth and signal-to-noise ratio.

  15. Macroscopic erosion of divertor and first wall armour in future tokamaks

    Science.gov (United States)

    Würz, H.; Bazylev, B.; Landman, I.; Pestchanyi, S.; Safronov, V.

    2002-12-01

    Sputtering, evaporation and macroscopic erosion determine the lifetime of the 'in vessel' armour materials CFC, tungsten and beryllium presently under discussion for future tokamaks. For CFC armour macroscopic erosion means brittle destruction and dust formation whereas for metallic armour melt layer erosion by melt motion and droplet splashing. Available results on macroscopic erosion from hot plasma and e-beam simulation experiments and from tokamaks are critically evaluated and a comprehensive discussion of experimental and numerical macroscopic erosion and its extrapolation to future tokamaks is given. Shielding of divertor armour materials by their own vapor exists during plasma disruptions. The evolving plasma shield protects the armour from high heat loads, absorbs the incoming energy and reradiates it volumetrically thus reducing drastically the deposited energy. As a result, vertical target erosion by vaporization turns out to be of the order of a few microns per disruption event and macroscopic erosion becomes the dominant erosion source.

  16. Design and characterization of a prototype divertor viewing infrared video bolometer for NSTX-U

    Science.gov (United States)

    van Eden, G. G.; Reinke, M. L.; Peterson, B. J.; Gray, T. K.; Delgado-Aparicio, L. F.; Jaworski, M. A.; Lore, J.; Mukai, K.; Sano, R.; Pandya, S. N.; Morgan, T. W.

    2016-11-01

    The InfraRed Video Bolometer (IRVB) is a powerful tool to measure radiated power in magnetically confined plasmas due to its ability to obtain 2D images of plasma emission using a technique that is compatible with the fusion nuclear environment. A prototype IRVB has been developed and installed on NSTX-U to view the lower divertor. The IRVB is a pinhole camera which images radiation from the plasma onto a 2.5 μm thick, 9 × 7 cm2 Pt foil and monitors the resulting spatio-temporal temperature evolution using an IR camera. The power flux incident on the foil is calculated by solving the 2D+time heat diffusion equation, using the foil's calibrated thermal properties. An optimized, high frame rate IRVB, is quantitatively compared to results from a resistive bolometer on the bench using a modulated 405 nm laser beam with variable power density and square wave modulation from 0.2 Hz to 250 Hz. The design of the NSTX-U system and benchtop characterization are presented where signal-to-noise ratios are assessed using three different IR cameras: FLIR A655sc, FLIR A6751sc, and SBF-161. The sensitivity of the IRVB equipped with the SBF-161 camera is found to be high enough to measure radiation features in the NSTX-U lower divertor as estimated using SOLPS modeling. The optimized IRVB has a frame rate up to 50 Hz, high enough to distinguish radiation during edge-localized-modes (ELMs) from that between ELMs.

  17. Design and characterization of a prototype divertor viewing infrared video bolometer for NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    Eden, G. G. van; Morgan, T. W. [Dutch Institute for Fundamental Energy Research, 5612 AJ Eindhoven (Netherlands); Reinke, M. L.; Gray, T. K.; Lore, J. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Peterson, B. J.; Mukai, K. [National Institute for Fusion Science, Toki 509-5292 (Japan); Delgado-Aparicio, L. F.; Jaworski, M. A. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, New Jersey 08543 (United States); Sano, R. [National Institutes for Quantum and Radiological Science and Technology, Naka 311-0193 (Japan); Pandya, S. N. [Institute for Plasma Research, Bhat Village, Gandhinagar, 382428 Gujarat (India)

    2016-11-15

    The InfraRed Video Bolometer (IRVB) is a powerful tool to measure radiated power in magnetically confined plasmas due to its ability to obtain 2D images of plasma emission using a technique that is compatible with the fusion nuclear environment. A prototype IRVB has been developed and installed on NSTX-U to view the lower divertor. The IRVB is a pinhole camera which images radiation from the plasma onto a 2.5 μm thick, 9 × 7 cm{sup 2} Pt foil and monitors the resulting spatio-temporal temperature evolution using an IR camera. The power flux incident on the foil is calculated by solving the 2D+time heat diffusion equation, using the foil’s calibrated thermal properties. An optimized, high frame rate IRVB, is quantitatively compared to results from a resistive bolometer on the bench using a modulated 405 nm laser beam with variable power density and square wave modulation from 0.2 Hz to 250 Hz. The design of the NSTX-U system and benchtop characterization are presented where signal-to-noise ratios are assessed using three different IR cameras: FLIR A655sc, FLIR A6751sc, and SBF-161. The sensitivity of the IRVB equipped with the SBF-161 camera is found to be high enough to measure radiation features in the NSTX-U lower divertor as estimated using SOLPS modeling. The optimized IRVB has a frame rate up to 50 Hz, high enough to distinguish radiation during edge-localized-modes (ELMs) from that between ELMs.

  18. Particle exhaust with vented structures: application to the ergodic divertor of Tore Supra; Pompage des particules dans les tokamaks au moyen d'une structure a events: le divertor ergodique de Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Azeroual, A

    2000-04-04

    In a thermonuclear reactor, one must continuously fuel the discharge and extract the ashes resulting from fusion reactions. To avoid the risk of discharge poisoning, {alpha}-particle concentration is limited to {approx} 10 %. To allow for steady-state conditions requires then to extract {>=}2 % of the helium out flux. In Tore Supra, the ergodic divertor is the main component managing the heat and particle fluxes at the edge. Its principle consists in generating a resonant perturbation able to destroy magnetic surfaces at the plasma periphery. In this region, the field lines are open and connected at both ends to neutralizers which are wetted by the major part of the heat and particle fluxes and are the structures through which a part of the plasma out flux is pumped for maintaining the discharge in steady-state conditions. This work describes the neutral recirculation around the ergodic divertor and is based on a data base of 56 discharges. One discuss the two processes allowing for particle exhaust: the ballistic collection of ions and that of neutrals backscattered by atomic reactions. These two processes are modelled accounting for a realistic description of the divertor geometry. A comparison between simulations and experiments is presented for measurements characterising the three main actors of plasma-wall interaction: the edge plasma, the D{sub {alpha}} light emission and the neutral pressure in the divertor plenum. Last, one question how such a system can be extrapolated to next step machines, for which one must account for technical constraints linked to the presence of the shield protecting the coils from the high neutron flux. (author)

  19. Experimental investigation of heat transport and divertor loads of fusion plasmas in all metal ASDEX upgrade and JET

    Energy Technology Data Exchange (ETDEWEB)

    Sieglin, Bernhard A.

    2014-04-28

    This work presents divertor heat load studies conducted at two of the largest tokamaks currently in operation, ASDEX Upgrade and the Joint European Torus (JET). A commonly agreed empirical scaling for the power fall-off length in H-mode obtained in carbon devices is validated in JET with the ILW. Bohm and Gyro-Bohm like models are identified as possible candidates describing the divertor broadening. Quantities for the assessment of the thermal load induced by transient heat loads are defined. JET with the ILW exhibits an on average longer ELM duration as compared to the carbon wall. For identical pedestal conditions the ELM durations in both cases are found to be the same within error bars. The energy fluency is found to depend mainly on the pedestal pressure with a weak dependence on the relative loss in stored energy. This is noteworthy since the current extrapolation to ITER assumes a linear dependence on the relative ELM size.

  20. High Heat Flux Test Simulation of Tungsten Macro-brush Mock-ups for the KSTAR Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Song, J. H.; Kim, K. M.; Kim, H. T.; Kim, H. K.; Bang, E. N.; Lee, K. S.; Park, S. H.; Yang, H. L.; Oh, Y. K. [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The divertor has been an important part of PFC because of its intrinsic function achieving effective particle control to keep high quality plasma with enough shaping flexibility. The divertor will be exposed to heat loads during operation of KSTAR-like plasma fusion device. Therefore, it is important to withstand high heat loads. In this paper, the hydraulic thermo-mechanical analysis was performed by ANSYS WORKBENCH 15.0 in order to predict the fatigue lifetime of the mock-ups for the high heat flux (HHF) test under the KSTAR base mode operating conditions. Under KSTAR base mode operating conditions, the finite element analysis was performed to predict the fatigue lifetime of the mock-ups for the HHF test. The results of analyses showed that the mock-up's temperatures were within the recommended operational range, and its fatigue lifetime was about 1,513 cycles.

  1. EMC3-EIRENE modeling of toroidally-localized divertor gas injection experiments on Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Lore, J.D., E-mail: lorejd@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Reinke, M.L. [York Plasma Institute, Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); LaBombard, B. [Plasma Science and Fusion Center, MIT, Cambridge, MA 02139 (United States); Lipschultz, B. [York Plasma Institute, Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); Churchill, R.M. [Plasma Science and Fusion Center, MIT, Cambridge, MA 02139 (United States); Pitts, R.A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Feng, Y. [Max Planck Institute for Plasma Physics, Greifswald (Germany)

    2015-08-15

    Experiments on Alcator C-Mod with toroidally and poloidally localized divertor nitrogen injection have been modeled using the three-dimensional edge transport code EMC3-EIRENE to elucidate the mechanisms driving measured toroidal asymmetries. In these experiments five toroidally distributed gas injectors in the private flux region were sequentially activated in separate discharges resulting in clear evidence of toroidal asymmetries in radiated power and nitrogen line emission as well as a ∼50% toroidal modulation in electron pressure at the divertor target. The pressure modulation is qualitatively reproduced by the modeling, with the simulation yielding a toroidal asymmetry in the heat flow to the outer strike point. Toroidal variation in impurity line emission is qualitatively matched in the scrape-off layer above the strike point, however kinetic corrections and cross-field drifts are likely required to quantitatively reproduce impurity behavior in the private flux region and electron temperatures and densities directly in front of the target.

  2. Strong correlation between D 2 density and electron temperature at the target of divertors found in SOLPS analysis

    Science.gov (United States)

    Stangeby, P. C.; Sang, Chaofeng

    2017-05-01

    A companion paper (Sang et al 2016 Nucl. Fusion (https://doi.org/10.1088/1741-4326/aa6548)) reports an assessment, using the SOLPS5.0 (B2-EIRENE) code, of the relative importance of two key aspects of divertor-baffle geometry: (i) divertor closure, and (ii) field-target angle. A wide range of the degree of divertor closure and field-target angle were modeled. An unexpectedly strong and simple correlation has been discovered in these data (and is reported here) between the electron temperature, T et, and the D 2 density, n{{D2}t}{} at the target, for T et  grid spanning two power decay widths outward from the separatrix. This may imply that achievement of low T et reduces, essentially, to identifying the divertor-baffle geometry which achieves the highest gas density near the target. To try to identify the controlling physics involved, two-point model formatting (2PMF) has been applied to the code output; it finds an equally strong and simple correlation between the 2PMF volumetric power-loss factor, {{f}\\text{vol-\\text{pwr}-\\text{loss}}} , and n{{D2}t}{} for each flux tube: {{f}\\text{vol-\\text{pwr}-\\text{loss}}}=1.2× {{10}29}n{{D2}t}-1.54~ with R 2 = 0.93. While these trends are broadly as would be expected, the simplicity, tightness and span of the correlations are not understood at present. Additionally, since more of the volumetric power loss is due to impurities than to deuterium, and as the impurities do not radiate just at the target, it is not evident why {{f}\\text{vol-\\text{pwr}-\\text{loss}}} is so strongly correlated with n{{D2}t}{} . To address these questions, in future work 2PMF analysis will be extended to compute the individual contributions to {{f}\\text{vol-\\text{pwr}-\\text{loss}}} .

  3. A mature industrial solution for ITER divertor plasma facing components: Hypervapotron cooling concept adapted to Tore Supra flat tile technology

    Energy Technology Data Exchange (ETDEWEB)

    Escourbiac, F. [Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 Saint Paul Lez Durance Cedex (France)]. E-mail: frederic.escourbiac@cea.fr; Bobin-Vastra, I. [AREVA Centre Technique de Framatome, Porte Magenta, BP181, 71200 Le Creusot (France); Kuznetsov, V. [Efremov Institute, Doroga na Metallostroy, St. Petersburg 196641 (Russian Federation); Missirlian, M. [Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 Saint Paul Lez Durance Cedex (France); Schedler, B. [Plansee AG, 6600 Reutte (Austria); Schlosser, J. [Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 Saint Paul Lez Durance Cedex (France)

    2005-11-15

    The use of flat tile technology to handle heat fluxes in the range of 20 MW/m{sup 2} with components relevant for fusion experiment applications is technically possible with the hypervapotron cooling concept. This paper deals with recent high heat flux performances operated with success in two different electron gun facilities and highlights the high potential of this technology for ITER divertor application.

  4. Typology of Product Configuration Systems

    DEFF Research Database (Denmark)

    Jensen, Klaes Ladeby; Edwards, Kasper; Haug, Anders

    2007-01-01

    Many organisations are moving from mass production to mass customization. Product configuration systems (PCS) are increasingly seen as an interesting option for firms who wish to pursue a strategy with a high degree of product variance while retaining a low cost of specifying the product. To become...... more specific in relation to how product configuration systems can support mass customization, it is necessary to understand how different product configuration systems can be classified, and how these differ. This paper presents a typology of product configuration systems based on the five kinds...

  5. Predictions of VRF on a Langmuir Probe under the RF Heating Spiral on the Divertor Floor on NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    Hosea, J C [PPPL; Perkins, R J [PPPL; Jaworski, M A [PPPL; Kramer, G J [PPPL; Ahn, J-W [ORNL

    2014-07-01

    RF heating deposition spirals are observed on the divertor plates on NSTX as shown in for a NB plus RF heating case. It has been shown that the RF spiral is tracked quite well by the spiral mapping of the strike points on the divertor plate of magnetic field lines passing in front of the high harmonic fast wave (HHFW) antenna on NSTX. Indeed, both current instrumented tiles and Langmuir probes respond to the spiral when it is positioned over them. In particular, a positive increment in tile current (collection of electrons) is obtained when the spiral is over the tile. This current can be due to RF rectification and/or RF heating of the scrape off layer (SOL) plasma along the magnetic field lines passing in front of the the HHFW antenna. It is important to determine quantitatively the relative contributions of these processes. Here we explore the properties of the characteristics of probes on the lower divertor plate to determine the likelyhood that the primary cause of the RF heat deposition is RF rectification.

  6. Characterizations of power loads on divertor targets for type-I, compound and small ELMs in the EAST superconducting tokamak

    DEFF Research Database (Denmark)

    Wang, L.; Xu, G.S.; Guo, H.Y.

    2013-01-01

    -III ELMy H-modes. The energy loss and divertor power load are systematically characterized for these different ELMy H-modes to provide a physics basis for the next-step high-power long-pulse operations in EAST. Both type-I and compound ELMs exhibit good confinement (H98(y,2) ∼ 1). A significant loss...... is about 10 MW m−2, as determined from the divertor-embedded triple Langmuir probe system with high time resolution. As expected, type-III ELMs lead to much smaller divertor power loads with a peak heat flux of about 2 MW m−2. Peak power loads for compound ELMs are between those for type-I and type......-III ELMs. It is remarkable that the new very small ELMy H-modes exhibit even lower target power deposition than type-III ELMs, with the peak heat flux generally below 1 MW m−2. These very small ELMs are usually accompanied by broadband fluctuations with frequencies ranging from 20 to 50 kHz, which may...

  7. 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.

  8. 'Finnova Development Group'. Comb Configurated Costumer-close Network Installations with Underground Service Boxes. Main solution 'Connection of Houses from a Distribution Chamber'; 'Finnova' Innovativ Montage och Systemloesning foer Fjaerrvaermeanslutning av Villaomraade. Huvudloesning 'Villanslutning fraan Foerdelningskammare'

    Energy Technology Data Exchange (ETDEWEB)

    Gudmundson, Tommy [AaF-Process AB, Stockholm (SE)] (and others)

    2006-07-15

    This report describes a technical solution for distribution of District Heating that has been given the name Finnova LTH. The general concept of Finnova LTH is based on a new kind of network configuration applied to residential areas with single family houses. In such areas, normally having a low heat load density, the houses are divided into groups of about 10, each group of 10 houses will be provided with a substation, in Finnova LTH called a distribution chamber. From this distribution chamber, groups of pipes, forming a kind of bundle, are leaving; each one of those groups supplying DH to one house. Each pipe group can be built up of 1, 2 or 4 pipes. In a case with 4 pipes, we in fact have a variation of the familiar '4 pipe design' principle. According to this principle, water for space heating as well as for tap water, is produced in a common station and distributed to a group of consumers. To be able to maintain the intended hot water temperature, such a system always includes a special piping for circulation of the hot water. The bundle of pipes leaving the distribution chamber will normally be installed under the sidewalk. For each building site (with an intended costumer) one pipe group leaves the bundle and heads toward the house in question. This means that the branches of conventional systems are replaced with such 'diversions'. In this way we will have one pipe, with the same diameter, and without joints or branching, all the way from the distribution chamber to the each costumers house. The number of houses served from one such distribution chamber may vary, depending on the geographical structure of the area in question and how the houses are distributed with respect to one another. Typically, one chamber may serve 10 houses. Two approaches are described in this report. They are referred to as 'generation 2' and 'generation 3', respectively. This is in line with the fact that the Finnova AF approach, described

  9. Configurational entropy of glueball states

    Energy Technology Data Exchange (ETDEWEB)

    Bernardini, Alex E., E-mail: alexeb@ufscar.br [Departamento de Física, Universidade Federal de São Carlos, PO Box 676, 13565-905, São Carlos, SP (Brazil); Braga, Nelson R.F., E-mail: braga@if.ufrj.br [Instituto de Física, Universidade Federal do Rio de Janeiro, Caixa Postal 68528, RJ 21941-972 (Brazil); Rocha, Roldão da, E-mail: roldao.rocha@ufabc.edu.br [CMCC, Universidade Federal do ABC, UFABC, 09210-580, Santo André (Brazil)

    2017-02-10

    The configurational entropy of glueball states is calculated using a holographic description. Glueball states are represented by a supergravity dual picture, consisting of a 5-dimensional graviton–dilaton action of a dynamical holographic AdS/QCD model. The configurational entropy is studied as a function of the glueball spin and of the mass, providing information about the stability of the glueball states.

  10. Viscous Design of TCA Configuration

    Science.gov (United States)

    Krist, Steven E.; Bauer, Steven X. S.; Campbell, Richard L.

    1999-01-01

    The goal in this effort is to redesign the baseline TCA configuration for improved performance at both supersonic and transonic cruise. Viscous analyses are conducted with OVERFLOW, a Navier-Stokes code for overset grids, using PEGSUS to compute the interpolations between overset grids. Viscous designs are conducted with OVERDISC, a script which couples OVERFLOW with the Constrained Direct Iterative Surface Curvature (CDISC) inverse design method. The successful execution of any computational fluid dynamics (CFD) based aerodynamic design method for complex configurations requires an efficient method for regenerating the computational grids to account for modifications to the configuration shape. The first section of this presentation deals with the automated regridding procedure used to generate overset grids for the fuselage/wing/diverter/nacelle configurations analysed in this effort. The second section outlines the procedures utilized to conduct OVERDISC inverse designs. The third section briefly covers the work conducted by Dick Campbell, in which a dual-point design at Mach 2.4 and 0.9 was attempted using OVERDISC; the initial configuration from which this design effort was started is an early version of the optimized shape for the TCA configuration developed by the Boeing Commercial Airplane Group (BCAG), which eventually evolved into the NCV design. The final section presents results from application of the Natural Flow Wing design philosophy to the TCA configuration.

  11. Characterizing Low-Z erosion and deposition in the DIII-D divertor using aluminum

    Directory of Open Access Journals (Sweden)

    C.P. Chrobak

    2017-08-01

    Full Text Available We present measurements and modeling of aluminum erosion and redeposition experiments in separate helium and deuterium low power, low density L-mode plasmas at the outer divertor strike point of DIII-D to provide a low-Z material benchmark dataset for tokamak erosion-deposition modeling codes. Coatings of Al ∼100nm thick were applied to ideal (smooth and realistic (rough surfaces and exposed to repeat plasma discharges using the DiMES probe. Redeposition in all cases was primarily in the downstream toroidal field direction, evident from both in-situ spectroscopic and post-mortem non-spectroscopic measurements. The gross Al erosion yield was estimated from film thickness change measurements of small area samples, and was found to be ∼40–70% of the expected erosion yield based on theoretical physical sputtering yields after including sputtering by a 1–3% carbon impurity. The multi-step redeposition and re-erosion process, and hence the measured net erosion yield and material migration patterns, were found to be influenced by the surface roughness and/or porosity. A time-dependent model of material migration accounting for deposit accumulation in hidden areas was developed to reproduce the measurements in these experiments and determine a redeposition probability distribution function for sputtered atoms.

  12. Fundamental physics behind the divertor heat-flux width in the present tokamaks and ITER

    Science.gov (United States)

    Chang, C. S.; Ku, S.; Churchill, R. M.; Hager, R.; Parker, Scott; Myra, Jim

    2017-10-01

    Electrostatic gyrokinetic simulation using XGC1 recovers the empirical scaling for the divertor heat-load width λq in the present tokamaks (λq 1 /Bpγ , with γ 1). λq is dominated by the neoclassical magnetic drift of ions. However, XGC1 predicts that λq in ITER is much larger than the value predicted by the empirical scaling. An in-depth study shows that the edge turbulence characteristics in ITER is highly different from that in the present tokamaks. In the present tokamaks, the edge turbulence in an H-mode plasma is ``blobby,'' with most of the convective blob motion in the poloidal direction yielding little radial transport. Blobby electron radial transport is passive, only keeping the quasi-neutrality with ion magnetic drift. However, in ITER, the edge turbulence is found to be `streamer-like,' giving rise to active radial particle and thermal transport. There appears to be a bifurcation of the edge turbulence characteristics from blobs to streamers between JET and ITER, most likely due to the size effect, in the XGC simulation. Fundamental physics behind this turbulence bifurcation will be discussed, in relation to the sheared ExB flow, and the Kelvin-Helmholtz, TEM and ITG turbulence. Funded by US DOE FES and ASCR. Computing resources provided by ALCC and INCITE programs on Titan.

  13. High-Z material erosion and its control in DIII-D carbon divertor

    Directory of Open Access Journals (Sweden)

    R. Ding

    2017-08-01

    Full Text Available As High-Z materials will likely be used as plasma-facing components (PFCs in future fusion devices, the erosion of high-Z materials is a key issue for high-power, long pulse operation. High-Z material erosion and redeposition have been studied using tungsten and molybdenum coated samples exposed in well-diagnosed DIII-D divertor plasma discharges. By coupling dedicated experiments and modelling using the 3D Monte Carlo code ERO, the roles of sheath potential and background carbon impurities in determining high-Z material erosion are identified. Different methods suggested by modelling have been investigated to control high-Z material erosion in DIII-D experiments. The erosion of Mo and W is found to be strongly suppressed by local injection of methane and deuterium gases. The 13C deposition resulting from local 13CH4 injection also provides information on radial transport due to E ×B drifts and cross field diffusion. Finally, D2 gas puffing is found to cause local plasma perturbation, suppressing W erosion because of the lower effective sputtering yield of W at lower plasma temperature and for higher carbon concentration in the mixed surface layer.

  14. Microanalysis of deposited layers in the inner divertor of JET with ITER-like wall

    Directory of Open Access Journals (Sweden)

    Y. Zhou

    2017-08-01

    Full Text Available In JET with ITER-like wall, beryllium eroded in the main chamber is transported to the divertor and deposited mainly at the horizontal surfaces of tiles 1 and 0 (high field gap closure, HFGC. These surfaces are tungsten coated carbon fibre composite (CFC. Surface sampleswere collected following the plasma operations in 2011–2012 and 2013–2014 respectively. The surfaces, as well as polished cross sections of the deposited layers at the surfaces have been studied with micro ion beam analysis methods (µ-IBA.Deposition of Beand other impurities, and retention of D is microscopically inhomogeneous. Impurities and trapped deuterium accumulate preferentially in cracks, pits and depressed regions, and at the sides of large pits in the substrate (e.g. arc tracks where the W coating has been removed. With careful overlaying of µ-NRA elemental maps with optical microscopy images, it is possible to separate surface roughness effects from depth profiles at microscopically flat surface regions.

  15. Divertor heat flux simulations in ELMy H-mode discharges of EAST

    Science.gov (United States)

    Xia, T. Y.; Xu, X. Q.; Wu, Y. B.; Huang, Y. Q.; Wang, L.; Zheng, Z.; Liu, J. B.; Zang, Q.; Li, Y. Y.; Zhao, D.; EAST Team

    2017-11-01

    This paper presents heat flux simulations for the ELMy H-mode on the Experimental Advanced Superconducting Tokamak (EAST) using a six-field two-fluid model in BOUT++. Three EAST ELMy H-mode discharges with different plasma currents I p and geometries are studied. The trend of the scrape-off layer width λq with I p is reproduced by the simulation. The simulated width is only half of that derived from the EAST scaling law, but agrees well with the international multi-machine scaling law. Note that there is no radio-frequency (RF) heating scheme in the simulations, and RF heating can change the boundary topology and increase the flux expansion. Anomalous electron transport is found to contribute to the divertor heat fluxes. A coherent mode is found in the edge region in simulations. The frequency and poloidal wave number kθ are in the range of the edge coherent mode in EAST. The magnetic fluctuations of the mode are smaller than the electric field fluctuations. Statistical analysis of the type of turbulence shows that the turbulence transport type (blobby or turbulent) does not influence the heat flux width scaling. The two-point model differs from the simulation results but the drift-based model shows good agreement with simulations.

  16. Imitation of deuterium plasma interaction with the surface of carbon materials in gaseous divertor conditions

    Energy Technology Data Exchange (ETDEWEB)

    Korshunov, S.N. E-mail: sinet@nfi.kiae.ru; Guseva, M.I.; Gureev, V.M.; Danelyan, L.S.; Khripunov, B.I.; Kolbasov, B.N.; Kulikauskas, V.S.; Litnovsky, A.M.; Martynenko, Yu.V.; Petrov, V.B.; Zatekin, V.V

    2003-03-01

    The experiments on simulation of gas divertor conditions were done in the LENTA facility under interaction of a plasma flow with neutral gas. The samples of carbon materials were exposed in a steady-state deuterium plasma (ion energy 5 eV, ion flux 5x10{sup 21} m{sup -2} s{sup -1}, fluence 10{sup 26} m{sup -2}) at 1470 K (MPG-8) and at 1320 K (SEP NB31). Heavy deuterocarbon molecules (C{sub 2}D{sub 2}, C{sub 2}D{sub 4}, C{sub 2}D{sub 6}) were observed in mass spectra of the discharge. This fact and high erosion yields show the presence of chemical erosion. Deuterium accumulation in carbon materials was studied by elastic recoil detection analysis. The integral deuterium content is 6x10{sup 18} m{sup -2} in SEP NB31 and 1.95x10{sup 19} m{sup -2} in MPG-8. The profiles of C and Mo atom distributions in deposited layer on Mo collector is 'X'-like. Carbon atoms distribution in deposited layer on Si is uniform. The integral deuterium content in co-deposited layers is 1.4x10{sup 21} m{sup -2} on Si and 4.8x10{sup 20} m{sup -2} on Mo. A globular structure of co-deposited layer on Mo collector was found.

  17. Energy removal and MHD performance of lithium capillary-pore systems for divertor target application

    Energy Technology Data Exchange (ETDEWEB)

    Evtikhin, V.A. E-mail: evtikhin@protein.bio.msu.ru; Lyublinski, I.E.; Vertkov, A.V.; Yezhov, N.I.; Khripunov, B.I.; Sotnikov, S.M.; Mirnov, S.V.; Petrov, V.B

    2000-11-01

    Experimental results of complex studies of lithium capillary-pore systems (CPS) for application as a plasma facing structure in divertor and on the first wall of a fusion reactor are reported. The ability of CPS to accept and to remove high heat fluxes (up to 30 MW m{sup -2}) in steady-state conditions (tens of minutes) has been evaluated on target plate imitator mock-ups supplied with cooling and lithium feed systems under electron beam power load in a linear plasma facility. Experimental study of lithium flow up to 2.5 m s{sup -1} in CPS made of material with final conductivity for various mesh sizes and of the effect of cross magnetic field up to 1.6 T on its parameters has been made. The results of successful experiments on the T-11M tokamak helium and hydrogen plasma interaction with a CPS-based lithium limiter and lithium puff influence on the plasma performances are presented and analysed.

  18. High heat flux testing of mock-ups for a full tungsten ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Gavila, P., E-mail: pierre.gavila@f4e.europa.eu [Fusion for Energy, 08019 Barcelona (Spain); Riccardi, B. [Fusion for Energy, 08019 Barcelona (Spain); Constans, S.; Jouvelot, J.L.; Vastra, I. Bobin [AREVA NP, Centre Technique, Fusion, 71200, Le Creusot (France); Missirlian, M.; Richou, M. [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France)

    2011-10-15

    In order to evaluate the option to start the ITER operation with a full tungsten (W) divertor, the EU-DA launched an extensive R and D program. It consisted in its initial phase in the high heat flux (HHF) testing of W mock-ups and medium scale prototypes up to 20 MW/m{sup 2} in the AREVA FE 200 facility (F). Critical heat flux (CHF) experiments were carried out on the items which survived the above thermal fatigue testing. After 1000 cycles at 10 MW/m{sup 2}, the full W Plasma Facing Components (PFCs) mock-ups successfully sustained either 1000 cycles at 15 MW/m{sup 2} or 500 cycles at 20 MW/m{sup 2}. However, some significant surface melting, as well as the complete melting of a few monoblocks, occurred during the HHF thermal fatigue testing program representative of the present ITER requirements for the strike point region, namely 1000 cycles at 10 MW/m{sup 2} followed by 1000 cycles at 20 MW/m{sup 2}. The results of the CHF experiments were also rather encouraging, since the tested items sustained heat fluxes in the range of 30 MW/m{sup 2} in steady-state conditions.

  19. Experimental activity on the definition of acceptance criteria for the ITER divertor plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Escourbiac, F. [CEA, IRFM, F-13108 St Paul Lez Durance (France)], E-mail: frederic.escourbiac@cea.fr; Constans, S. [SOM Ortec, Marseille (France); Vignal, N.; Cantone, V.; Richou, M.; Durocher, A. [CEA, IRFM, F-13108 St Paul Lez Durance (France); Riccardi, B. [Fusion For Energy, Barcelona (Spain); Bobin, I.; Jouvelot, J.L. [AREVA-NP, Le Creusot (France); Merola, M. [ITER Organization, Cadarache (France)

    2009-06-15

    Tens of thousands of armor/heat sink joints will be produced by the industry during the manufacturing of ITER divertor PFC, statistically, there is a probability that joints with defects be delivered. The purpose of this paper is to study the detection and evolution during operation of calibrated defects artificially implemented on samples, as an experimental basis for the definition of acceptance criteria for the bond armor/heat sink in the frame of industrial manufacturing conditions.It was found that current CFC monoblock design option was compatible with the heat loads specified at the lower part of the vertical target (up to 20 MW/m{sup 2}), including the presence of armor/heat sink defects (up to 50 deg. extension for a location at 0 deg. or 45 deg.) detectable with NDE techniques developed in Europe (US, SATIR). The current W monoblock design appeared suitable for the upper part of the vertical target with defects extension up to 50 deg. but is not adapted for heat flux of 20 MW/m{sup 2}. The studied W flat tile design proved to be compatible with fluxes of 5 MW/m{sup 2} but unable to sustain cycling fluxes of 10 MW/m{sup 2}.

  20. The heat removal capability of actively cooled plasma-facing components for the ITER divertor

    Science.gov (United States)

    Missirlian, M.; Richou, M.; Riccardi, B.; Gavila, P.; Loarer, T.; Constans, S.

    2011-12-01

    Non-destructive examination followed by high-heat-flux testing was performed for different small- and medium-scale mock-ups; this included the most recent developments related to actively cooled tungsten (W) or carbon fibre composite (CFC) armoured plasma-facing components. In particular, the heat-removal capability of these mock-ups manufactured by European companies with all the main features of the ITER divertor design was investigated both after manufacturing and after thermal cycling up to 20 MW m-2. Compliance with ITER requirements was explored in terms of bonding quality, heat flux performances and operational compatibility. The main results show an overall good heat-removal capability after the manufacturing process independent of the armour-to-heat sink bonding technology and promising behaviour with respect to thermal fatigue lifetime under heat flux up to 20 MW m-2 for the CFC-armoured tiles and 15 MW m-2 for the W-armoured tiles, respectively.

  1. Measurements of gross erosion of Al in the DIII-D divertor

    Energy Technology Data Exchange (ETDEWEB)

    Chrobak, C., E-mail: chrobak@fusion.gat.com [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Stangeby, P.C. [University of Toronto Institute for Aerospace Studies, Toronto M3H 5T6 (Canada); Leonard, A.W. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Rudakov, D.L. [University of California San Diego, 9500 Gilman Dr., La Jolla, CA 92093-0417 (United States); Wong, C.P.C. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); McLean, A.G. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Wright, G.M. [Massachusetts Institute of Technology, 77 Massachusetts Ave, Cambridge, MA 02139 (United States); Buchenauer, D.A.; Watkins, J.G.; Wampler, W.R. [Sandia National Laboratory, P.O. Box 5800, Albuquerque, NM 87185 (United States); Elder, J.D. [University of Toronto Institute for Aerospace Studies, Toronto M3H 5T6 (Canada); Doerner, R.P.; Nishijima, D.; Tynan, G.R. [University of California San Diego, 9500 Gilman Dr., La Jolla, CA 92093-0417 (United States)

    2015-08-15

    Aluminum (Al) is a convenient proxy for beryllium (Be) plasma material interaction studies since they have a number of physical and chemical similarities. Al samples were exposed at the lower outer strike point of an L-mode divertor plasma in DIII-D (conditions 7–11 × 10{sup 18} D-ions cm{sup −2} s{sup −1}, T{sub e} = 12–47 eV). The gross erosion rate was directly measured using post-mortem ion beam analysis of small 1 mm-sized samples where local re-deposition was determined to be negligible. The gross erosion rate was also calculated using spectroscopic methods, but these rates greatly underestimate the direct (i.e. non-spectroscopic) measurement. The direct measured erosion yields were within the range of published D{sup +} → Al ion beam sputtering yields. The ionizations per photon (S/XB) coefficients used in the spectroscopic analysis were determined in separate experiments using He plasmas at the PISCES-B linear plasma facility at UCSD. The measured S/XB coefficients were on average ∼6× higher than the theoretically calculated values.

  2. The influence of the dynamic ergodic divertor on the radial electric field at the Tokamak TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Coenen, Jan Willem

    2009-11-06

    In this work the influence of external Resonant Magnetic Perturbations (RMPs) on the radial electric field Er in magnetically confined plasmas is investigated by Charge Exchange Recombination Spectroscopy (CXRS) at the Tokamak TEXTOR. Here, the RMPs are produced with the Dynamic Ergodic Divertor (DED), a set of 16 helical perturbation coils located at the high field side of TEXTOR. Within this work, the base mode number of perturbations has been m/n=6/2. We have first investigated the influence of external torque from neutral heating beams on plasma rotation and E{sub r}. The ergodic zone causes an electron loss, and subsequently a (vector)j x (vector)B force driven by the compensating ion return current. In addition, the DED changes the global confinement properties. Depending on the edge safety factor (''field line twist'') q{sub a}, either increased or decreased particle confinement is observed. In case of the increased particle confinement (IPC) the increase in density (40%) and particle confinement time {tau}{sub p} (30%) is correlated to the connection of field lines at the q=5/2 surface to the DED target, locally changing the transport properties and the E{sub r}. Transport is reduced and the E{sub r} shear is increased locally at q=5/2 up to 1.5 . 10{sup 5}s{sup -1}, while the E{sub r} becomes more positive. (orig.)

  3. R and D issues of W/Cu divertor for EAST

    Energy Technology Data Exchange (ETDEWEB)

    Li, Q.; Qi, P.; Zhou, H.S.; Zhang, Y.Y.; Yang, Z.S.; Wu, J. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Luo, G.-N., E-mail: gnluo@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Li, J.G. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)

    2010-12-15

    To improve plasma performance of EAST device, the SiC-coated doped C tiles in the divertor region will be replaced by tungsten coated CrZrCu heat sink in 3-5 years. Vacuum plasma spraying (VPS) tungsten coatings on Cu substrate are being developed in collaboration with Guangzhou Research Institute of Nonferrous Metals (GZRINM) and Shanghai Institute of Ceramics, Chinese Academy of Sciences (ASSIC). In order to reduce thermal stresses, several kinds of interlayer structures have been considered to be coated on the heat sink before preparing the pure tungsten layer to the final thickness, and the castellation concepts also taken into account. An actively cooled VPS-W/Cu movable limiter (ML) has been tested in HT-7 and a Material and Plasma Evaluation System (MAPES) will be built soon on EAST. High heat flux testing and non-destructive testing are also being studied. Plasma-wall interaction (PWI) issues like recycling and retention of H isotopes in the coatings, surface and bulk modification of the coatings and service life of the plasma-facing material and component (PFMC) are being studied not only on tokamaks but also in laboratories.

  4. A Software Configuration Management Course

    DEFF Research Database (Denmark)

    Asklund, U.; Bendix, Lars Gotfred

    2003-01-01

    Software Configuration Management has been a big success in research and creation of tools. There are also many vendors in the market of selling courses to companies. However, in the education sector Software Configuration Management has still not quite made it - at least not into the university...... curriculum. It is either not taught at all or is just a minor part of a general course in software engineering. In this paper, we report on our experience with giving a full course entirely dedicated to Software Configuration Management topics and start a discussion of what ideally should be the goal...

  5. Moderator Configuration Options for ESS

    DEFF Research Database (Denmark)

    Zanini, L.; Batkov, K.; Klinkby, Esben Bryndt

    2016-01-01

    The current, still evolving status of the design and the optimization work for the moderator configuration for the European Spallation Source is described. The moderator design has been strongly driven by the low-dimensional moderator concept recently proposed for use in spallation neutron sources...... conventional, principles were also considered,such as the importance of moderator positioning, of the premoderator, and beam extraction considerations. Different design and configuration options are evaluated and compared with the reference volume moderator configuration described in the ESS Technical Design...

  6. Device configuration-management system

    Energy Technology Data Exchange (ETDEWEB)

    Nowell, D.M.

    1981-01-01

    The Fusion Chamber System, a major component of the Magnetic Fusion Test Facility, contains several hundred devices which report status to the Supervisory Control and Diagnostic System for control and monitoring purposes. To manage the large number of diversity of devices represented, a device configuration management system was required and developed. Key components of this software tool include the MFTF Data Base; a configuration editor; and a tree structure defining the relationships between the subsystem devices. This paper will describe how the configuration system easily accomodates recognizing new devices, restructuring existing devices, and modifying device profile information.

  7. Configurational entropy of glueball states

    Directory of Open Access Journals (Sweden)

    Alex E. Bernardini

    2017-02-01

    Full Text Available The configurational entropy of glueball states is calculated using a holographic description. Glueball states are represented by a supergravity dual picture, consisting of a 5-dimensional graviton–dilaton action of a dynamical holographic AdS/QCD model. The configurational entropy is studied as a function of the glueball spin and of the mass, providing information about the stability of the glueball states.

  8. A study on tokamak fusion reactor - Numerical analyses of MHD equilibrium= and edge plasma transport in tokamak fusion reactor with divertor configurations

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sang Hee; Lim, Ki Hang; Kang, Kyung Doo; Ryu, Ji Myung; Kim, Duk Kyu [Seoul National University, Seoul (Korea, Republic of); Cho, Soo Won [Kyungki Unviersity, Suwon (Korea, Republic of)

    1995-08-01

    In the present project for developing the numerical codes of 2-DMHD equilibrium, edge plasma transport and neutral particle transport for the tokamak plasmas, we compute the plasma equilibrium of double null type and calculate the external coil currents and the plasma parameters used for operation and control data. Also the numerical algorithm is developed to analyse the behavior of edge plasmas in poloidal and radial directions and the programming and debugging of a 2-D transport code are completed. Furthermore, a neutral particle transport code for the edge region is developed and then used for the analysis of the neutral transport phenomena giving the sources in the fluid equations, and expected to supply the input parameters for the edge plasma transport code. 34 refs., 5 tabs., 28 figs. (author)

  9. Study of non-axisymmetric divertor footprints using 2-D IR and visible cameras and a 3-D heat conduction solver in NSTX

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, J-W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gan, K. F. [Chinese Academy of Sciences (CAS), Hefei (China). Inst. of Plasma Physics; Scotti, F. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Lore, J. D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Maingi, R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Canik, J. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gray, T. K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McLean, A. G. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Roquemore, A. L. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Soukhanovskii, V. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2013-01-12

    Toroidally non-axisymmetric divertor profiles during the 3-D field application and for ELMs are studied with simultaneous observation by a new wide angle visible camera and a high speed IR camera. A newly implemented 3-D heat conduction code, TACO, is used to obtain divertor heat flux. The wide angle camera data confirmed the previously reported result on the validity of vacuum field line tracing on the prediction of split strike point pattern by 3-D fields as well as the phase locking of ELM heat flux to the 3-D fields. TACO calculates the 2- D heat flux distribution allowing assessment of toroidal asymmetry of peak heat flux and heat flux width. Lastly, the degree of asymmetry (εDA) is defined to quantify the asymmetric heat deposition on the divertor surface and is found to have a strong positive dependence on peak heat flux.

  10. Dynamic behavior of the intensified alternative configurations for quaternary distillation

    DEFF Research Database (Denmark)

    Ramirez-Marquez, Cesar; Cabrera-Ruiz, Julián; Juan Gabriel Segovia-Hernandez, Juan Gabriel

    2016-01-01

    Process intensification emerges as an important tool in the synthesis of multicomponent distillation configurations aimed at the reduction of the energy use and capital costs. Operational and fixed costs savings coupled with simplicity and controllability design configurations appear as an essent......Process intensification emerges as an important tool in the synthesis of multicomponent distillation configurations aimed at the reduction of the energy use and capital costs. Operational and fixed costs savings coupled with simplicity and controllability design configurations appear...... value decomposition technique in all frequency domain. In order to complete the control study, the distillation schemes were subjected to closed-loop dynamic simulations. The results show that there are cases in which the intensified sequences do not only provide energy savings, but also may offer...

  11. Progress of ITER full tungsten divertor technology qualification in Japan: Manufacturing full-scale plasma-facing unit prototypes

    Energy Technology Data Exchange (ETDEWEB)

    Ezato, Koichiro, E-mail: ezato.koichiro@jaea.go.jp [Department of ITER Project, Naka Fusion Institute, Sector of Fusion Research and Development, Japan Atomic Energy Agency (Japan); Suzuki, Satoshi; Seki, Yohji; Yamada, Hirokazu; Hirayama, Tomoyuki; Yokoyama, Kenji [Department of ITER Project, Naka Fusion Institute, Sector of Fusion Research and Development, Japan Atomic Energy Agency (Japan); Escourbiac, Frederic; Hirai, Takeshi [ITER Organization, route de vinon sur Verdon, 13067 St Paul lez Durance (France)

    2016-11-01

    Highlights: • JADA has demonstrated the feasibility of manufacturing the full-W plasma-facing units (W-PFU). • The surface profiles of the W monoblocks of the W-PFU prototypes on the test frame to mimic the support structure of the ITER OVT were examined by using an optical three-dimensional measurement system. The results show the most W monoblock surface in the target part locates within + 0.25 mm from the CAD data. • The strict profile control with the profile tolerance of ±0.3 mm is imposed on the OVT to prevent the leading edges of the W monoblocks from over-heating. • The present full-scale prototyping demonstrates to satisfy this requirement on the surface profile. • It can be concluded that the technical maturities of JADA and its suppliers are as high as to start series manufacturing the ITER divertor components. - Abstract: Japan Atomic Energy Agency (JAEA) is in progress for technology demonstration toward Full-tungsten (W) ITER divertor outer vertical target (OVT), especially, W monoblock technology that needs to withstand the repetitive heat load as high as 20 MW/m{sup 2} for 10 s. Under the framework of the W divertor qualification program developed ITER organization, JAEA as Japanese Domestic Agency (JADA) manufactured seven full-scale plasma-facing unit (PFU) prototypes with the Japanese industries. Four prototypes that have 146 W monoblock joint with casted copper (Cu) interlayer passed successfully the ultrasonic testing. In the other three prototypes that have the different W/Cu interlayer joint, joint defects were found. The dimension measurements reveal the requirements of the gap between W monoblocks and the surface profile of PFU are feasible.

  12. Structural impact of armor monoblock dimensions on the failure behavior of ITER-type divertor target components: Size matters

    Energy Technology Data Exchange (ETDEWEB)

    Li, Muyuan; You, Jeong-Ha, E-mail: you@ipp.mpg.de

    2016-12-15

    Highlights: • Quantitative assessment of size effects was conducted numerically for W monoblock. • Decreasing the width of W monoblock leads to a lower risk of failure. • The Cu interlayer was not affected significantly by varying armor thickness. • The predicted trends were in line with the experimental observations. - Abstract: Plenty of high-heat-flux tests conducted on tungsten monoblock type divertor target mock-ups showed that the threshold heat flux density for cracking and fracture of tungsten armor seems to be related to the dimension of the monoblocks. Thus, quantitative assessment of such size effects is of practical importance for divertor target design. In this paper, a computational study about the thermal and structural impact of monoblock size on the plastic fatigue and fracture behavior of an ITER-type tungsten divertor target is reported. As dimensional parameters, the width and thickness of monoblock, the thickness of sacrificial armor, and the inner diameter of cooling tube were varied. Plastic fatigue lifetime was estimated for the loading surface of tungsten armor and the copper interlayer by use of a cyclic-plastic constitutive model. The driving force of brittle crack growth through the tungsten armor was assessed in terms of J-integral at the crack tip. Decrease of the monoblock width effectively reduced accumulation of plastic strain at the armor surface and the driving force of brittle cracking. Decrease of sacrificial armor thickness led to decrease of plastic deformation at the loading surface due to lower surface temperature, but the thermal and mechanical response of the copper interlayer was not affected by the variation of armor thickness. Monoblock with a smaller tube diameter but with the same armor thickness and shoulder thickness experienced lower fatigue load. The predicted trends were in line with the experimental observations.

  13. Modeling of complex gas distribution systems operating under any vacuum conditions: Simulations of the ITER divertor pumping system

    Energy Technology Data Exchange (ETDEWEB)

    Vasileiadis, N.; Tatsios, G.; Misdanitis, S.; Valougeorgis, D., E-mail: diva@mie.uth.gr

    2016-02-15

    Highlights: • An integrated s/w for modeling complex rarefied gas distribution systems is presented. • Analysis is based on kinetic theory of gases. • Code effectiveness is demonstrated by simulating the ITER divertor pumping system. • The present s/w has the potential to support design work in large vacuum systems. - Abstract: An integrated software tool for modeling and simulation of complex gas distribution systems operating under any vacuum conditions is presented and validated. The algorithm structure includes (a) the input geometrical and operational data of the network, (b) the definition of the fundamental set of network loops and pseudoloops, (c) the formulation and solution of the mass and energy conservation equations, (d) the kinetic data base of the flow rates for channels of any length in the whole range of the Knudsen number, supporting, in an explicit manner, the solution of the conservation equations and (e) the network output data (mainly node pressures and channel flow rates/conductance). The code validity is benchmarked under rough vacuum conditions by comparison with hydrodynamic solutions in the slip regime. Then, its feasibility, effectiveness and potential are demonstrated by simulating the ITER torus vacuum system with the six direct pumps based on the 2012 design of the ITER divertor. Detailed results of the flow patterns and paths in the cassettes, in the gaps between the cassettes and along the divertor ring, as well as of the total throughput for various pumping scenarios and dome pressures are provided. A comparison with previous results available in the literature is included.

  14. 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.

  15. Time-resolved deposition in the remote region of the JET-ILW divertor: measurements and modelling

    Science.gov (United States)

    Catarino, N.; Widdowson, A.; Baron-Wiechec, A.; Coad, J. P.; Heinola, K.; Rubel, M.; Alves, E.; Contributors, JET

    2017-12-01

    One crucial requirement for the development of fusion power is to know where, and how much, impurities collect in the machine, and how much of the fuelling isotope tritium will be trapped therein. The most relevant information on this issue comes from the operation of the Joint European Tokamak (JET), which is the world’s largest operating tokamak and has the same interior plasma-facing materials as the next step machine, ITER. Much of the information gained so far has been from post-mortem analysis of samples collected after whole campaigns involving varied types of operation. This paper describes time-resolved measurements of the deposition rate using rotating collectors (RC) placed in remote areas of the JET divertor during the 2013–2014 campaign with the ITER-like Wall (ILW). These techniques allow the effects of different types of operation to be distinguished. Rotating collectors made of silicon discs housed behind an aperture are exposed to the plasma. Each time the magnetic field coils are ramped up for a discharge the disc rotates, providing a linear relationship between the exposed region and the discharge number. Post-mortem ion beam analyses provide information on the deposit composition as a function of the discharge number. The results show that the Be deposition average for the RC in the corners of the inner and outer divertor are 4.9 × 1016 cm‑2 and 1.8 × 1017 cm‑2, respectively, accumulated over an average of ∼25 pulses. Data from the rotating collector below Tile 5 in the central region of divertor indicate a Be deposition rate of 9.3 × 1015 cm‑2, per ∼25 pulses.

  16. Knowledge Based Product Configuration - a documentatio tool for configuration projects

    DEFF Research Database (Denmark)

    Hvam, Lars; Malis, Martin

    2003-01-01

    . A lot of knowledge isput into these systems and many domain experts are involved. This calls for an effective documentation system in order to structure this knowledge in a way that fits to the systems. Standard configuration systems do not support this kind of documentation. The chapter deals...... with the development of a Lotus Notes application that serves as a knowledge based documentation tool for configuration projects. A prototype has been developed and tested empirically in an industrial case-company. It has proved to be a succes....

  17. Study of near SOL decay lengths in ASDEX Upgrade under attached and detached divertor conditions

    Science.gov (United States)

    Sun, H. J.; Wolfrum, E.; Kurzan, B.; Eich, T.; Lackner, K.; Scarabosio, A.; Paradela Pérez, I.; Kardaun, O.; Faitsch, M.; Potzel, S.; Stroth, U.; the ASDEX Upgrade Team

    2017-10-01

    A database with attached, partially detached and completely detached divertors has been constructed in ASDEX Upgrade discharges in both H-mode and L-mode plasmas with Thomson Scattering data suitable for the analysis of the upstream SOL electron profiles. By comparing upstream temperature decay width, {λ }{Te,u}, with the scaling of the SOL power decay width, {λ }{q\\parallel e}, based on the downstream IR measurements, it is found that a simple relation based on classical electron conduction can relate {λ }{Te,u} and {λ }{q\\parallel e} well. The combined dataset can be described by both a single scaling and a separate scaling for H-modes and L-modes. For the single scaling, a strong inverse dependence of, {λ }{Te,u} on the separatrix temperature, {T}e,u, is found, suggesting the classical parallel Spitzer-Harm conductivity as dominant mechanism controlling the SOL width in both L-mode and H-mode over a large set of plasma parameters. This dependence on {T}e,u explains why, for the same global plasma parameters, {λ }{q\\parallel e} in L-mode is approximately twice that in H-mode and under detached conditions, the SOL upstream electron profile broadens when the density reaches a critical value. Comparing the derived scaling from experimental data with power balance, gives the cross-field thermal diffusivity as {χ }\\perp \\propto {T}e{1/2}/{n}e, consistent with earlier studies on Compass-D, JET and Alcator C-Mod. However, the possibility of the separate scalings for different regimes cannot be excluded, which gives results similar to those previously reported for the H-mode, but here the wider SOL width for L-mode plasmas is explained simply by the larger premultiplying coefficient. The relative merits of the two scalings in representing the data and their theoretical implications are discussed.

  18. Expanding the role of impurity spectroscopy for investigating the physics of high-Z dissipative divertors

    Directory of Open Access Journals (Sweden)

    M.L. Reinke

    2017-08-01

    Full Text Available New techniques that attempt to more fully exploit spectroscopic diagnostics in the divertor and pedestal region during highly dissipative scenarios are demonstrated using experimental results from recent low-Z seeding experiments on Alcator C-Mod, JET and ASDEX Upgrade. To exhaust power at high parallel heat flux, q∥ > 1 GW/m2, while minimizing erosion, reactors with solid, high-Z plasma facing components (PFCs are expected to use extrinsic impurity seeding. Due to transport and atomic physics processes which impact impurity ionization balance, so-called ‘non-coronal’ effects, we do not accurately know and have yet to demonstrate the maximum q∥ which can be mitigated in a tokamak. Radiation enhancement for nitrogen is shown to arise primarily from changes in Li- and Be-like charge states on open field lines, but also through transport-driven enhancement of H- and He-like charge states in the pedestal region. Measurements are presented from nitrogen seeded H-mode and L-mode plasmas where emission from N1+ through N6+ are observed. Active charge exchange spectroscopy of partially ionized low-Z impurities in the plasma edge is explored to measure N5+ and N6+ within the confined plasma, while passive UV and visible spectroscopy is used to measure N1+-N4+ in the boundary. Examples from recent JET and Alcator C-Mod experiments which employ nitrogen seeding highlight how improving spectroscopic coverage can be used to gain empirical insight and provide more data to validate boundary simulations.

  19. Stochastic layer scaling in the two-wire model for divertor tokamaks

    Science.gov (United States)

    Ali, Halima; Punjabi, Alkesh; Boozer, Allen

    2009-06-01

    The question of magnetic field structure in the vicinity of the separatrix in divertor tokamaks is studied. The authors have investigated this problem earlier in a series of papers, using various mathematical techniques. In the present paper, the two-wire model (TWM) [Reiman, A. 1996 Phys. Plasmas 3, 906] is considered. It is noted that, in the TWM, it is useful to consider an extra equation expressing magnetic flux conservation. This equation does not add any more information to the TWM, since the equation is derived from the TWM. This equation is useful for controlling the step size in the numerical integration of the TWM equations. The TWM with the extra equation is called the flux-preserving TWM. Nevertheless, the technique is apparently still plagued by numerical inaccuracies when the perturbation level is low, resulting in an incorrect scaling of the stochastic layer width. The stochastic broadening of the separatrix in the flux-preserving TWM is compared with that in the low mn (poloidal mode number m and toroidal mode number n) map (LMN) [Ali, H., Punjabi, A., Boozer, A. and Evans, T. 2004 Phys. Plasmas 11, 1908]. The flux-preserving TWM and LMN both give Boozer-Rechester 0.5 power scaling of the stochastic layer width with the amplitude of magnetic perturbation when the perturbation is sufficiently large [Boozer, A. and Rechester, A. 1978, Phys. Fluids 21, 682]. The flux-preserving TWM gives a larger stochastic layer width when the perturbation is low, while the LMN gives correct scaling in the low perturbation region. Area-preserving maps such as the LMN respect the Hamiltonian structure of field line trajectories, and have the added advantage of computational efficiency. Also, for a $1\\frac12$ degree of freedom Hamiltonian system such as field lines, maps do not give Arnold diffusion.

  20. Hydrogenic retention of high-Z refractory metals exposed to ITER divertor-relevant plasma conditions

    Science.gov (United States)

    Wright, G. M.; Alves, E.; Alves, L. C.; Barradas, N. P.; Carvalho, P. A.; Mateus, R.; Rapp, J.

    2010-05-01

    Tungsten (W) and molybdenum (Mo) targets are exposed to the plasma conditions expected at the strike point of a detached ITER divertor (ne ~ 1020 m-3, Te ~ 2 eV) in the linear plasma device Pilot-PSI. The peak surface temperatures of the targets are ~1600 K for W and ~1100 K for Mo. The surface temperatures and plasma flux densities decrease radially towards the edges of the target due to the Gaussian distribution of electron density (ne) and temperature (Te) in the plasma column. A 2D spatial scan of the W and Mo targets using nuclear reaction analysis (NRA) shows D retention is strongly influenced by surface temperature in the range 800-1600 K and this dependence dominates over any plasma flux dependence. NRA and thermal desorption spectroscopy (TDS) show no clear dependence of retention on incident plasma fluence for the W targets with retained fractions ranging from 10-8-10-5 Dretained/Dincident. NRA and TDS for the Mo targets show retention rates a factor of 4-5 higher than the W targets and this is likely due to the lower surface temperatures for the Mo plasma exposures. NRA also reveals a thin boron layer on the Mo targets but the presence of boron does not correspond to a significant increase in D retention. Overall hydrogenic retention in W and Mo is shown to be low (Dretained = 1019-1020 D m-2) despite exposure to high plasma flux densities (~1024 D m-2 s-1). This is likely due to the elevated surface temperature due to plasma thermal loading during exposure.

  1. Inflation in a closed universe

    Science.gov (United States)

    Ratra, Bharat

    2017-11-01

    To derive a power spectrum for energy density inhomogeneities in a closed universe, we study a spatially-closed inflation-modified hot big bang model whose evolutionary history is divided into three epochs: an early slowly-rolling scalar field inflation epoch and the usual radiation and nonrelativistic matter epochs. (For our purposes it is not necessary to consider a final dark energy dominated epoch.) We derive general solutions of the relativistic linear perturbation equations in each epoch. The constants of integration in the inflation epoch solutions are determined from de Sitter invariant quantum-mechanical initial conditions in the Lorentzian section of the inflating closed de Sitter space derived from Hawking's prescription that the quantum state of the universe only include field configurations that are regular on the Euclidean (de Sitter) sphere section. The constants of integration in the radiation and matter epoch solutions are determined from joining conditions derived by requiring that the linear perturbation equations remain nonsingular at the transitions between epochs. The matter epoch power spectrum of gauge-invariant energy density inhomogeneities is not a power law, and depends on spatial wave number in the way expected for a generalization to the closed model of the standard flat-space scale-invariant power spectrum. The power spectrum we derive appears to differ from a number of other closed inflation model power spectra derived assuming different (presumably non de Sitter invariant) initial conditions.

  2. BAYESIAN IMAGE RESTORATION, USING CONFIGURATIONS

    Directory of Open Access Journals (Sweden)

    Thordis Linda Thorarinsdottir

    2011-05-01

    Full Text Available In this paper, we develop a Bayesian procedure for removing noise from images that can be viewed as noisy realisations of random sets in the plane. The procedure utilises recent advances in configuration theory for noise free random sets, where the probabilities of observing the different boundary configurations are expressed in terms of the mean normal measure of the random set. These probabilities are used as prior probabilities in a Bayesian image restoration approach. Estimation of the remaining parameters in the model is outlined for salt and pepper noise. The inference in the model is discussed in detail for 3 X 3 and 5 X 5 configurations and examples of the performance of the procedure are given.

  3. Configuration of distributed message converter systems

    NARCIS (Netherlands)

    Risse, Thomas; Aberer, Karl; Wombacher, Andreas; Surridge, Mike; Taylor, Stephen

    2004-01-01

    Finding a configuration of a distributed system satisfying performance goals is a complex search problem that involves many design parameters, like hardware selection, job distribution and process configuration. Performance models are a powerful tool to analyze potential system configurations,

  4. Modularisation of Software Configuration Management

    DEFF Research Database (Denmark)

    Christensen, Henrik Bærbak

    2000-01-01

    management, and outline how modularisation is natural and powerful also in this context. The analysis is partly based on experiences from case studies where small- to medium-sized development projects are using a prototype tool that supports modular software configuration management.......The principle of modularisation is one of the main techniques that software designers use to tame the complexity of programming. A software project, however, is complex in many other areas than just programming. In this paper, we focus on one of these complex areas, namely software configuration...

  5. Instance-specific algorithm configuration

    CERN Document Server

    Malitsky, Yuri

    2014-01-01

    This book presents a modular and expandable technique in the rapidly emerging research area of automatic configuration and selection of the best algorithm for the instance at hand. The author presents the basic model behind ISAC and then details a number of modifications and practical applications. In particular, he addresses automated feature generation, offline algorithm configuration for portfolio generation, algorithm selection, adaptive solvers, online tuning, and parallelization.    The author's related thesis was honorably mentioned (runner-up) for the ACP Dissertation Award in 2014,

  6. Numerical simulation in a subcooled water flow boiling for one-sided high heat flux in reactor divertor

    Energy Technology Data Exchange (ETDEWEB)

    Liu, P., E-mail: pinliu@aust.edu.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230026 (China); School of Mechanical Engineering, Anhui University of Science and Technology, Huainan 232001 (China); Peng, X.B., E-mail: pengxb@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Song, Y.T. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230026 (China); Fang, X.D. [Institute of Air Conditioning and Refrigeration, Nanjing University of Aeronautics and Astronautics, Nanjing 210016 (China); Huang, S.H. [University of Science and Technology of China, Hefei 230026 (China); Mao, X. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2016-11-15

    Highlights: • The Eulerian multiphase models coupled with Non-equilibrium Boiling model can effectively simulate the subcooled water flow boiling. • ONB and FDB appear earlier and earlier with the increase of heat fluxes. • The void fraction increases gradually along the flow direction. • The inner CuCrZr tube deteriorates earlier than the outer tungsten layer and the middle OFHC copper layer. - Abstract: In order to remove high heat fluxes for plasma facing components in International Thermonuclear Experimental Reactor (ITER) divertor, a numerical simulation of subcooled water flow boiling heat transfer in a vertically upward smooth tube was conducted in this paper on the condition of one-sided high heat fluxes. The Eulerian multiphase model coupled with Non-equilibrium Boiling model was adopted in numerical simulation of the subcooled boiling two-phase flow. The heat transfer regions, thermodynamic vapor quality (x{sub th}), void fraction and temperatures of three components on the condition of the different heat fluxes were analyzed. Numerical results indicate that the onset of nucleate boiling (ONB) and fully developed boiling (FDB) appear earlier and earlier with increasing heat flux. With the increase of heat fluxes, the inner CuCrZr tube will deteriorate earlier than the outer tungsten layer and the middle oxygen-free high-conductivity (OFHC) copper layer. These results provide a valuable reference for the thermal-hydraulic design of a water-cooled W/Cu divertor.

  7. Optimization of the Water-Cooled Structure for the Divertor Plates in EAST Based on an Orthogonal Theory

    Science.gov (United States)

    Li, Lei; Yao, Damao; Liu, Changle; Zhou, Zibo; Cao, Lei; Liang, Chao

    2015-05-01

    An orthogonal experimental scheme was designed for optimizing a water-cooled structure of the divertor plate. There were three influencing factors: the radius R of the water-cooled pipe, and the pipe spacing L1 and L3. The influence rule of different factors on the cooling effect and thermal stress of the plate were studied, for which the influence rank was respectively R > L1 > L3 and L3 > R > L1. The highest temperature value decreased when R and L1 increased, and the maximum thermal stress value dropped when R, L1 and L3 increased. The final optimized results can be summarized as: R equals 6 mm or 7 mm, L1 equals 19 mm, and L3 equals 20 mm. Compared with the initial design, the highest temperature value had a small decline, and the maximum thermal stress value dropped by 19% to 24%. So it was not ideal to improve the cooling effect by optimizing the geometry sizes of the water-cooled structure, even worse than increasing the flow speed, but it was very effective for dropping the maximum thermal stress value. The orthogonal experimental method reduces the number of experiments by 80%, and thus it is feasible and effective to optimize the water-cooled structure of the divertor plate with the orthogonal theory. supported by National Basic Research Program of China (973 Program) (No. 2013GB102000)

  8. Production and fabrication of vanadium alloys for the radiative divertor program of DIII-D - Annual report input for 1996

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, W.R.; Smith, J.P.; Stambaugh, R.D.

    1996-10-01

    V-4Cr-4Ti alloy has been selected for use in the manufacture of a portion of the DIII-D Radiative Divertor (RD) upgrade. The production of a 1200-kg ingot of V-4Cr-4Ti alloy has been completed at Teledyne Wah Chang of Albany, Oregon (TWCA) to provide {approximately}800-kg of applicable product forms, and two billets have been extruded from the ingot. Chemical compositions of the ingot and both extruded billets were acceptable. Material from these billets will be converted into product forms suitable for components of the DIII-D Radiative Divertor structure. Joining of V-4Cr-4Ti alloy has been identified as the most critical fabrication issue for its use in the RD Program, and research into several joining methods for fabrication of the RD components, including resistance seam, friction, and electron beam welding, is continuing. Preliminary trials have been successful in the joining of V-alloy to itself by electron beam, resistance, and friction welding processes and to Inconel 625 by friction welding.

  9. Compatibility of separatrix density scaling for divertor detachment with H-mode pedestal operation in DIII-D

    Science.gov (United States)

    Leonard, A. W.; McLean, A. G.; Makowski, M. A.; Stangeby, P. C.

    2017-08-01

    The midplane separatrix density is characterized in response to variations in upstream parallel heat flux density and central density through deuterium gas injection. The midplane density is determined from a high spatial resolution Thomson scattering diagnostic at the midplane with power balance analysis to determine the separatrix location. The heat flux density is varied by scans of three parameters, auxiliary heating, toroidal field with fixed plasma current, and plasma current with fixed safety factor, q 95. The separatrix density just before divertor detachment onset is found to scale consistent with the two-point model when radiative dissipation is taken into account. The ratio of separatrix to pedestal density, n e,sep/n e,ped varies from  ⩽30% to  ⩾60% over the dataset, helping to resolve the conflicting scaling of core plasma density limit and divertor detachment onset. The scaling of the separatrix density at detachment onset is combined with H-mode power threshold scaling to obtain a scaling ratio of minimum n e,sep/n e,ped expected in future devices.

  10. Results of high heat flux tests of tungsten divertor targets under plasma heat loads expected in ITER and tokamaks (review)

    Energy Technology Data Exchange (ETDEWEB)

    Budaev, V. P., E-mail: budaev@mail.ru [National Research Centre Kurchatov Institute (Russian Federation)

    2016-12-15

    Heat loads on the tungsten divertor targets in the ITER and the tokamak power reactors reach ~10MW m{sup −2} in the steady state of DT discharges, increasing to ~0.6–3.5 GW m{sup −2} under disruptions and ELMs. The results of high heat flux tests (HHFTs) of tungsten under such transient plasma heat loads are reviewed in the paper. The main attention is paid to description of the surface microstructure, recrystallization, and the morphology of the cracks on the target. Effects of melting, cracking of tungsten, drop erosion of the surface, and formation of corrugated and porous layers are observed. Production of submicron-sized tungsten dust and the effects of the inhomogeneous surface of tungsten on the plasma–wall interaction are discussed. In conclusion, the necessity of further HHFTs and investigations of the durability of tungsten under high pulsed plasma loads on the ITER divertor plates, including disruptions and ELMs, is stressed.

  11. The global build-up to intrinsic ELM bursts seen in divertor full flux loops in Jet

    CERN Document Server

    Chapman, S C; Todd, T N; Watkins, N W; Calderon, F A; Morris, J; Contributors, JET

    2015-01-01

    A global signature of the build-up to an intrinsic ELM is found in the phase of signals measured in full flux azimuthal loops in the divertor region of JET. Full flux loop signals provide a global measurement proportional to the voltage induced by changes in poloidal magnetic flux; they are electromagnetically induced by the dynamics of spatially integrated current density. We perform direct time-domain analysis of the high time-resolution full flux loop signals VLD2 and VLD3. We analyze plasmas where a steady H-mode is sustained over several seconds, during which all the observed ELMs are intrinsic; there is no deliberate intent to pace the ELMing process by external means. ELM occurrence times are determined from the Be II emission at the divertor. We previously found that the occurrence times of intrinsic ELMs correlate with specific phases of the VLD2 and VLD3 signals. Here, we investigate how the VLD2 and VLD3 phases vary with time in advance of the ELM occurrence time. We identify a build-up to the ELM ...

  12. Parallel Robots with Configurable Platforms

    NARCIS (Netherlands)

    Lambert, P.

    2013-01-01

    This thesis explores the fundamentals of a new class of parallel mechanisms called parallel mechanisms with configurable platforms as well as the design and analysis of parallel robots that are based on those mechanisms. Pure parallel robots are formed by two rigid links, the base and the

  13. Product Configuration Systems and Productivity

    DEFF Research Database (Denmark)

    Pedersen, Jørgen Lindgaard; Edwards, Kasper

    2004-01-01

    Twelve companies have been interviewed with the purpose to get information about technical, economic and organisational matters in respect of Product Configuration Systems (PCS).Combinations of qualitative interviews and quantitative scoring have been used in ranking expected and realized results...

  14. Knowledge Engineering for Embedded Configuration

    DEFF Research Database (Denmark)

    Oddsson, Gudmundur Valur

    2008-01-01

    This thesis presents a way to simplify setup of complex product systems with the help of embedded configuration. To achieve this, one has to focus on what subsystems need to communicate between themselves. The required internal knowledge is then structured at three abstraction levels. Simplificat......This thesis presents a way to simplify setup of complex product systems with the help of embedded configuration. To achieve this, one has to focus on what subsystems need to communicate between themselves. The required internal knowledge is then structured at three abstraction levels....... Simplifications of the internal workings are both due to hardware- and application-induced configuration taking place both within the overall system and in each subsystem. By relating parameters in such a way, the number of user inputs or decision variables should decrease drastically, thus increasing the overall...... usability of the installation. In our case, we have rationalized that this should be done with embedded configuration, and the expected result is enhanced usability. The suggested method is deeply rooted in system theory. It draws on the emergent properties expected from the system, and tries to embed...

  15. NCCDS configuration management process improvement

    Science.gov (United States)

    Shay, Kathy

    1993-01-01

    By concentrating on defining and improving specific Configuration Management (CM) functions, processes, procedures, personnel selection/development, and tools, internal and external customers received improved CM services. Job performance within the section increased in both satisfaction and output. Participation in achieving major improvements has led to the delivery of consistent quality CM products as well as significant decreases in every measured CM metrics category.

  16. Bayesian image restoration, using configurations

    DEFF Research Database (Denmark)

    Thorarinsdottir, Thordis

    configurations are expressed in terms of the mean normal measure of the random set. These probabilities are used as prior probabilities in a Bayesian image restoration approach. Estimation of the remaining parameters in the model is outlined for salt and pepper noise. The inference in the model is discussed...

  17. Bayesian image restoration, using configurations

    DEFF Research Database (Denmark)

    Thorarinsdottir, Thordis Linda

    2006-01-01

    configurations are expressed in terms of the mean normal measure of the random set. These probabilities are used as prior probabilities in a Bayesian image restoration approach. Estimation of the remaining parameters in the model is outlined for the salt and pepper noise. The inference in the model is discussed...

  18. Spatial Configuration of Macromolecular Chains

    Indian Academy of Sciences (India)

    highlight in this lecture give promise of far-reaching advances in our understanding of macromolecular ... in articles of commerce. The chemical bonds in macromolecules differ in no discernible respect from those in ... The number and variety of configurations (or conformations in the language of organic chemistry) that.

  19. Integrated core–SOL–divertor modelling for ITER including impurity: effect of tungsten on fusion performance in H-mode and hybrid scenario

    NARCIS (Netherlands)

    Zagorski, R.; Voitsekhovitch, I.; Ivanova-Stanik, I.; Kochl, F.; Belo, da Silva Ares; Fable, E.; Garcia, J.; Garzotti, L.; Hobirk, J.; Hogeweij, G. M. D.; Joffrin, E.; Litaudon, X.; Polevoi, A. R.; Telesca, G.; JET Contributors,

    2015-01-01

    The compatibility of two operational constraints—operation above the L–H power threshold and at low power to divertor—is examined for ITER long pulse H-mode and hybrid scenarios in integrated core–scrape off layer (SOL)–divertor modelling including impurities (intrinsic Be, He, W and seeded Ne). The

  20. Feedback system for divertor impurity seeding based on real-time measurements of surface heat flux in the Alcator C-Mod tokamak

    Science.gov (United States)

    Brunner, D.; Burke, W.; Kuang, A. Q.; LaBombard, B.; Lipschultz, B.; Wolfe, S.

    2016-02-01

    Mitigation of the intense heat flux to the divertor is one of the outstanding problems in fusion energy. One technique that has shown promise is impurity seeding, i.e., the injection of low-Z gaseous impurities (typically N2 or Ne) to radiate and dissipate the power before it arrives to the divertor target plate. To this end, the Alcator C-Mod team has created a first-of-its-kind feedback system to control the injection of seed gas based on real-time surface heat flux measurements. Surface thermocouples provide real-time measurements of the surface temperature response to the plasma heat flux. The surface temperature measurements are inputted into an analog computer that "solves" the 1-D heat transport equation to deliver accurate, real-time signals of the surface heat flux. The surface heat flux signals are sent to the C-Mod digital plasma control system, which uses a proportional-integral-derivative (PID) algorithm to control the duty cycle demand to a pulse width modulated piezo valve, which in turn controls the injection of gas into the private flux region of the C-Mod divertor. This paper presents the design and implementation of this new feedback system as well as initial results using it to control divertor heat flux.

  1. Configuring Symantec AntiVirus

    CERN Document Server

    Shimonski, Robert

    2003-01-01

    This is the only book that will teach system administrators how to configure, deploy, and troubleshoot Symantec Enterprise Edition in an enterprise network. The book will reflect Symantec''s philosophy of "Centralized Antivirus Management." For the same reasons that Symantec bundled together these previously separate products, the book will provide system administrators with a holistic approach to defending their networks from malicious viruses. This book will also serve as a Study Guide for those pursuing Symantec Product Specialist Certifications.Configuring Symantec AntiVirus Enterprise Edition contains step-by-step instructions on how to Design, implement and leverage the Symantec Suite of products in the enterprise.ØFirst book published on market leading product and fast-growing certification. Despite the popularity of Symantec''s products and Symantec Product Specialist certifications, there are no other books published or announced.ØLess expensive substitute for costly on-sight training. Symantec off...

  2. Microsoft System Center Configuration Manager

    CERN Document Server

    Sandbu, Marius

    2013-01-01

    This book is a step-by-step tutorial that guides you through the key steps in implementing best solutions for high availability and performance tuning. It is split into two distinct approaches: client and site side HA and optimization.Microsoft SCCM High Availability and Performance Tuning is for IT professionals and consultants working with Configuration Manager who wish to learn the skills to deploy a redundant and scalable solution.

  3. Vacuum configuration for inflationary superstring

    Energy Technology Data Exchange (ETDEWEB)

    Baadhio, R.A. (Theoretical Physics Group, Physics Division, Lawrence Berkeley Laboratory and Department of Physics, University of California at Berkeley, Berkeley, California 94720 (United States))

    1993-02-01

    The vacuum configuration for the inflationary superstring theory is established. It is argued that the basic physical contents of the inflationary universe are characterized by the Novikov higher signature. Finally it is shown, with respect to the splitting of Paper II, that the index of the Dirac operator defined in our inflated universe, and in the parallel shadow one, is indeed [ital h]-cobordant.

  4. Drupal 7 Multi Sites Configuration

    CERN Document Server

    Butcher, Matt

    2012-01-01

    Follow the creation of a multi-site instance with Drupal. The practical examples and accompanying screenshots will help you to get multiple Drupal sites set up in no time. This book is for Drupal site builders. It is assumed that readers are familiar with Drupal already, with a basic grasp of its concepts and components. System administration concepts, such as configuring Apache, MySQL, and Vagrant are covered but no previous knowledge of these tools is required.

  5. Subsonic/Transonic Configuration Aerodynamics.

    Science.gov (United States)

    1980-09-01

    d’origine dite Ni et mat I - avec nacelle ddfinitive dite N5 (position et forme modifiges) et mat 8. 1 CARACTERISATION DES PROBLEMES TR.ANSSONIQUES...aspect ratio delta wing section found on the Space Shuttle Orbiter . CONSERVATION OF COMPUTING RESOURCES An Inexperienced analyst applying some of the...shuttle orbite ~r geometry is quite blunt. The transport configuration shock waves are comparable to those typically encountered at cruise conditions

  6. Configuration Management Process Assessment Strategy

    Science.gov (United States)

    Henry, Thad

    2014-01-01

    Purpose: To propose a strategy for assessing the development and effectiveness of configuration management systems within Programs, Projects, and Design Activities performed by technical organizations and their supporting development contractors. Scope: Various entities CM Systems will be assessed dependent on Project Scope (DDT&E), Support Services and Acquisition Agreements. Approach: Model based structured against assessing organizations CM requirements including best practices maturity criteria. The model is tailored to the entity being assessed dependent on their CM system. The assessment approach provides objective feedback to Engineering and Project Management of the observed CM system maturity state versus the ideal state of the configuration management processes and outcomes(system). center dot Identifies strengths and risks versus audit gotcha's (findings/observations). center dot Used "recursively and iteratively" throughout program lifecycle at select points of need. (Typical assessments timing is Post PDR/Post CDR) center dot Ideal state criteria and maturity targets are reviewed with the assessed entity prior to an assessment (Tailoring) and is dependent on the assessed phase of the CM system. center dot Supports exit success criteria for Preliminary and Critical Design Reviews. center dot Gives a comprehensive CM system assessment which ultimately supports configuration verification activities.*

  7. Component configuration control system development at EBR-II

    Energy Technology Data Exchange (ETDEWEB)

    Monson, L.R.; Stratton, R.C.

    1984-01-01

    One ofthe major programs being pursued by the EBR-II Division of Argonne National Laboratory is to improve the reliability of plant control and protection systems. This effort involves looking closely at the present state of the art and needs associated with plant diagnostic, control and protection systems. One of the areas of development at EBR-II involves a component configuration control system (CCCS). This system is a computerized control and planning aid for the nuclear power operator.

  8. ELM-free and inter-ELM divertor heat flux broadening induced by edge harmonics oscillation in NSTX

    Science.gov (United States)

    Gan, K. F.; Ahn, J.-W.; Gray, T. K.; Zweben, S. J.; Fredrickson, E. D.; Scotti, F.; Maingi, R.; Park, J.-K.; Canal, G. P.; Soukhanovskii, V. A.; Mclean, A. G.; Wirth, B. D.

    2017-12-01

    A new n  =  1 dominated edge harmonic oscillation (EHO) has been found in NSTX. The new EHO, rotating toroidally in the counter-current direction and the opposite direction of the neutral beam, was observed during certain inter-ELM and ELM-free periods of H-mode operation. This EHO is associated with a significant broadening of the integral heat flux width ({λ\\operatorname{int}} ) by up to 150%, and a decrease in the divertor peak heat flux by  >60%. An EHO induced filament was also observed by the gas puff imaging diagnostic. The toroidal rotating filaments could change the edge magnetic topology resulting in toroidal rotating strike point splitting and heat flux broadening. Experimental result of the counter current rotation of strike points splitting is consistent with the counter-current EHO.

  9. Modelling of steady state erosion of CFC actively water-cooled mock-up for the ITER divertor

    Science.gov (United States)

    Ogorodnikova, O. V.

    2008-04-01

    Calculations of the physical and chemical erosion of CFC (carbon fibre composite) monoblocks as outer vertical target of the ITER divertor during normal operation regimes have been done. Off-normal events and ELM's are not considered here. For a set of components under thermal and particles loads at glancing incident angle, variations in the material properties and/or assembly of defects could result in different erosion of actively-cooled components and, thus, in temperature instabilities. Operation regimes where the temperature instability takes place are investigated. It is shown that the temperature and erosion instabilities, probably, are not a critical point for the present design of ITER vertical target if a realistic variation of material properties is assumed, namely, the difference in the thermal conductivities of the neighbouring monoblocks is 20% and the maximum allowable size of a defect between CFC armour and cooling tube is +/-90° in circumferential direction from the apex.

  10. Stopped nucleons in configuration space

    Energy Technology Data Exchange (ETDEWEB)

    Bialas, Andrzej [Jagellonian Univ., Krakow (Poland); Bzdak, Adam [AGH - Univ. of Science and Technology, Krakow (Poland); Koch, Volker [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

    2017-05-09

    In this note, using the colour string model, we study the configuration space distribution of stopped nucleons in heavy-ion collisions. We find that the stopped nucleons from the target and the projectile end up separated from each other by the distance increasing with the collision energy. In consequence, for the center of mass energies larger than 6 or 10 GeV (depending on the details of the model) it appears that the system created is not in thermal and chemical equilibrium, and the net baryon density reached is likely not much higher than that already present in the colliding nuclei.

  11. Numerical Calculation of the Peaking Factor of a Water-Cooled W/Cu Monoblock for a Divertor

    Science.gov (United States)

    Han, Le; Chang, Haiping; Zhang, Jingyang; Xu, Tiejun

    2015-09-01

    In order to accurately predict the incident critical heat flux (ICHF, the heat flux at the heated surface when CHF occurs) of a water-cooled W/Cu monoblock for a divertor, the exact knowledge of its peaking factors (fp) under one-sided heating conditions with different design parameters is a key issue. In this paper, the heat conduction in the solid domain of a water-cooled W/Cu monoblock is calculated numerically by assuming the local heat transfer coefficients (HTC) of the cooling wall to be functions of the local wall temperature, so as to obtain fp. The reliability of the calculation method is validated by an experimental example result, with the maximum error of 2.1% only. The effects of geometric and flow parameters on the fp of a water-cooled W/Cu monoblock are investigated. Within the scope of this study, it is shown that the fp increases with increasing dimensionless W/Cu monoblock width and armour thickness (the shortest distance between the heated surface and Cu layer), and the maximum increases are 43.8% and 22.4% respectively. The dimensionless W/Cu monoblock height and Cu thickness have little effect on fp. The increase of Reynolds number and Jakob number causes the increase of fp, and the maximum increases are 6.8% and 9.6% respectively. Based on the calculated results, an empirical correlation on peaking factor is obtained via regression. These results provide a valuable reference for the thermal-hydraulic design of water-cooled divertors. supported by National Magnetic Confinement Fusion Science Program of China (No. 2010GB104005) and Funding of Jiangsu Innovation Program for Graduate Education, China (CXLX12_0170), the Fundamental Research Funds for the Central Universities of China

  12. On the use of CX atom analyzer for study characteristics of ion component in a LHD divertor plasma

    Energy Technology Data Exchange (ETDEWEB)

    Voitsenya, V.S.; Masuzaki, S.; Motojima, O.; Noda, N.; Ohyabu, N.

    1996-12-01

    In this paper the analysis was provided for the possibility to use the charge exchange atom analyzer using the ion reflection phenomena on solid surfaces for measuring the characteristics of the ion component of a divertor plasma in LHD. As an ion-atom converter the target plate made of refractory metal (Ta or W) is proposed to be used. This target plate can withstand the energy flux transported by the divertor plasma during LHD pulse: {approx}10 MW/m{sup 2} for {approx}2 s with water cooling. The particle brightness of such target is much higher than the one of a gas target with a reasonable value of molecular hydrogen density (10{sup 14} cm{sup -3}). The efficiency of W-made ion reflected atom converter is rather high, 40-65% in the incident ion energy range 50-1000 eV, however the energy reflection coefficient is lower for these energies: {approx}20-40 eV. Beside, the appearing the carbon or boron film on the target surface can lead to the decrease of ion-atom conversion efficiency. In such conditions the use of a time-of-flight (t-o-f) atom analyzer has some advantages as compare to the device with the gas stripping cell for charge exchange atom-ion conversion and electrostatic analysis of the ion energy distribution. In this paper we give the short description of energy component of the scheme with t-o-f atom analyzer in use, and the estimation of atom fluxes into the direction of atom analyzer with metal and gas targets. (author)

  13. Measurements of Plasma Power Losses in the C-2 Field-Reversed Configuration Experiment

    Science.gov (United States)

    Korepanov, Sergey; Smirnov, Artem; Garate, Eusebio; Donin, Alexandr; Kondakov, Alexey; Singatulin, Shavkat

    2013-10-01

    A high-confinement operating regime with plasma lifetimes significantly exceeding past empirical scaling laws was recently obtained by combining plasma gun edge biasing and tangential Neutral Beam Injection in the C-2 field-reversed configuration (FRC) experiment. To analyze the power balance in C-2, two new diagnostic instruments - the pyroelectric (PE) and infrared (IR) bolometers - were developed. The PE bolometer, designed to operate in the incident power density range from 0.1-100 W/cm2, is used to measure the radial power loss, which is dominated by charge-exchange neutrals and radiation. The IR bolometer, which measures power irradiated onto a thin metal foil inserted in the plasma, is designed for the power density range from 0.5-5 kW/cm2. The IR bolometer is used to measure the axial power loss from the plasma near the end divertors. The maximum measurable pulse duration of ~ 10 ms is limited by the heat capacitance of the IR detector. Both detectors have time resolution of about 10-100 μs and were calibrated in absolute units using a high power neutral beam. We present the results of first direct measurements of axial and radial plasma power losses in C-2.

  14. Overview of C-2W Field-Reversed Configuration Experimental Program

    Science.gov (United States)

    Gota, H.; Binderbauer, M. W.; Tajima, T.; Putvinski, S.; Tuszewski, M.; Dettrick, S.; Korepanov, S.; Romero, J.; Smirnov, A.; Song, Y.; Thompson, M. C.; van Drie, A.; Yang, X.; Ivanov, A. A.; TAE Team

    2017-10-01

    Tri Alpha Energy's research has been devoted to producing a high temperature, stable, long-lived field-reversed configuration (FRC) plasma state by neutral-beam injection (NBI) and edge biasing/control. C-2U experiments have demonstrated drastic improvements in particle and energy confinement properties of FRC's, and the plasma performance obtained via 10 MW NBI has achieved plasma sustainment of up to 5 ms and plasma (diamagnetism) lifetimes of 10 + ms. The emerging confinement scaling, whereby electron energy confinement time is proportional to a positive power of the electron temperature, is very attractive for higher energy plasma confinement; accordingly, verification of the observed Te scaling law will be a key future research objective. The new experimental device, C-2W (now also called ``Norman''), has the following key subsystem upgrades from C-2U: (i) higher injected power, optimum energies, and extended pulse duration of the NBI system; (ii) installation of inner divertors with upgraded edge-biasing systems; (iii) fast external equilibrium/mirror-coil current ramp-up capability; and (iv) installation of trim/saddle coils for active feedback control of the FRC plasma. This paper will review highlights of the C-2W program.

  15. Industrial requirements for interactive product configurators

    DEFF Research Database (Denmark)

    Queva, Matthieu Stéphane Benoit; Probst, Christian W.; Vikkelsøe, Per

    2009-01-01

    The demand for highly customized products at low cost is driving the industry towards Mass Customization. Interactive product configurators play an essential role in this new trend, and must be able to support more and more complex features. The purpose of this paper is, firstly, to identify...... requirements for modern interactive configurators. Existing modeling and solving technologies for configuration are then reviewed and their limitations discussed. Finally, a proposition for a future product configuration system is described....

  16. Offshore Vendors' Software Development Team Configuration

    DEFF Research Database (Denmark)

    Chakraborty, Suranjan; Sarker, Saonee; Rai, Sudhanshu

    2011-01-01

    This research uses configuration theory and data collected from a major IT vendor organization to examine primary configurations of distributed teams in a global off-shoring context. The study indicates that off-shoring vendor organizations typically deploy three different types of configurations...

  17. Offshore Vendors’ Software Development Team Configurations

    DEFF Research Database (Denmark)

    Chakraborty, Suranjan; Sarker, Saonee; Rai, Sudhanshu

    2012-01-01

    This research uses configuration theory and data collected from a major IT vendor organization to examine primary configurations of distributed teams in a global off-shoring context. The study indicates that off-shoring vendor organizations typically deploy three different types of configurations...

  18. Test of divertor materials under simulated ITER plasma disruption conditions using the hot plasma stream of the 2MK-200 facility

    Energy Technology Data Exchange (ETDEWEB)

    Arkhipov, N.I.; Bakhtin, V.; Konkashbaev, I. [Troitsk Inst. for Innovation and Fusion Research (Russian Federation)] [and others

    1994-12-31

    The high divertor heat load during tokamak plasma disruption results in sudden evaporation of a thin layer of divertor plate material, which acts as vapor shield and protects the target from further excessive evaporation. Formation and effectiveness of the vapor shield and material erosion are experimentally investigated at the 2MK-200 facility under conditions simulating the thermal quench phase of ITER tokamak plasma disruptions. Material samples from graphite, tungsten, boron nitrite and quartz were exposed to deuterium plasma streams with the following parameters: density < 10{sup 16}cm{sup {minus}3}, temperature T{sub e}+T{sub i} < 0.8 keV, plasma beta 0.25, plasma flow width 2 cm, power density 10 MW/cm{sup 2} and time duration of the pulse 20 {mu}s.

  19. Supply chain configuration concepts, solutions, and applications

    CERN Document Server

    Chandra, Charu

    2016-01-01

    This book discusses the models and tools available for solving configuration problems, emphasizes the value of model integration to obtain comprehensive and robust configuration decisions, proposes solutions for supply chain configuration in the presence of stochastic and dynamic factors, and illustrates application of the techniques discussed in applied studies. It is divided into four parts, which are devoted to defining the supply chain configuration problem and identifying key issues, describing solutions to various problems identified, proposing technologies for enabling supply chain confirmations, and discussing applied supply chain configuration problems. Its distinguishing features are: an explicit focus on the configuration problem an in-depth coverage of configuration models an emphasis on model integration and application of information modeling techniques in decision-making New to this edition is Part II: Technologies, which introduces readers to various technologies being utilized for supply chai...

  20. Close-ups

    DEFF Research Database (Denmark)

    Fausing, Bent

    2013-01-01

    Investigations in the close-up and its meaning regarding nearness, abstraction and transparency. Face, facelike and animism are also major key-words in thsi article... . I have always been fascinated with the close-up, not as an end, but a filter of opportunities to open up for and nearness...... of transcendence, associations and memories. The close-up is not the end, it rather in my view to be regarded as a beginning of different perceptions...

  1. Micro-/nano-characterization of the surface structures on the divertor tiles from JET ITER-like wall

    Energy Technology Data Exchange (ETDEWEB)

    Tokitani, M., E-mail: tokitani.masayuki@LHD.nifs.ac.jp [National Institute for Fusion Science, Oroshi, Toki, Gifu 509-5292 (Japan); Miyamoto, M. [Shimane University, Matsue, Shimane 690-8504 (Japan); Masuzaki, S. [National Institute for Fusion Science, Oroshi, Toki, Gifu 509-5292 (Japan); Fujii, Y. [Shimane University, Matsue, Shimane 690-8504 (Japan); Sakamoto, R. [National Institute for Fusion Science, Oroshi, Toki, Gifu 509-5292 (Japan); Oya, Y. [Shizuoka University, Shizuoka 422-8529 (Japan); Hatano, Y. [University of Toyama, Toyama 930-8555 (Japan); Otsuka, T. [Kindai University, Higashi-Osaka, Osaka, 577-8502 (Japan); Oyaidzu, M.; Kurotaki, H.; Suzuki, T.; Hamaguchi, D.; Isobe, K.; Asakura, N. [National Institute for Quantum and Radiological Science and Technology (QST), Rokkasho Aomori 039-3212 (Japan); Widdowson, A. [EUROfusion Consortium, JET, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Rubel, M. [Royal Institute of Technology (KTH), 100 44 Stockholm (Sweden)

    2017-03-15

    Highlights: • Micro-/nano-characterization of the surface structures on the divertor tiles from JET ITER-like wall were studied. • The stratified mixed-material deposition layer composed by W, C, O, Mo and Be with the thickness of ∼1.5 μm was formed on the apron of Tile 1. • The study revealed the micro- and nano-scale modification of the inner tile surface of the JET ILW. - Abstract: Micro-/nano-characterization of the surface structures on the divertor tiles used in the first campaign (2011–2012) of the JET tokamak with the ITER-like wall (JET ILW) were studied. The analyzed tiles were a single poloidal section of the tile numbers of 1, 3 and 4, i.e., upper, vertical and horizontal targets, respectively. A sample from the apron of Tile 1 was deposition-dominated. Stratified mixed-material layers composed of Be, W, Ni, O and C were deposited on the original W-coating. Their total thickness was ∼1.5 μm. By means of transmission electron microscopy, nano-size bubble-like structures with a size of more than 100 nm were identified in that layer. They could be related to deuterium retention in the layer dominated by Be. The surface microstructure of the sample from Tile 4 also showed deposition: a stratified mixed-material layer with the total thickness of 200–300 nm. The electron diffraction pattern obtained with transmission electron microscope indicated Be was included in the layer. No bubble-like structures have been identified. The surface of Tile 3, originally coated by Mo, was identified as the erosion zone. This is consistent with the fact that the strike point was often located on that tile during the plasma operation. The study revealed the micro- and nano-scale modification of the inner tile surface of the JET ILW. In particular, a complex mixed-material deposition layer could affect hydrogen isotope retention and dust formation.

  2. Progress towards modeling tokamak boundary plasma turbulence and understanding its role in setting divertor heat flux widths

    Science.gov (United States)

    Chen, Bin

    2017-10-01

    QCMs (quasi-coherent modes) are well characterized in the edge of Alcator C-Mod, when operating in the Enhanced Dα (EDA) H-mode, a promising alternative regime for ELM (edge localized modes) suppressed operation. To improve the understanding of the physics behind the QCMs, three typical C-Mod EDA H-Mode discharges are simulated by BOUT + + using a six-field two-fluid model (based on the Braginskii equations). The simulated characteristics of the frequency versus wave number spectra of the modes is in reasonable agreement with phase contrast imaging data. The key simulation results are: 1) Linear spectrum analysis and the nonlinear phase relationship indicate the dominance of resistive-ballooning modes and drift-Alfven wave instabilities; 2) QCMs originate inside the separatrix; (3) magnetic flutter causes the mode spreading into the SOL; 4) the boundary electric field Er changes the turbulent characteristics of the QCMs and controls edge transport and the divertor heat flux width; 5) the magnitude of the divertor heat flux depends on the physics models, such as sources and sinks, sheath boundary conditions, and parallel heat flux limiting coefficient. The BOUT + + simulations have also been performed for inter-ELM periods of DIII-D and EAST discharges, and similar quasi-coherent modes have been found. The parallel electron heat fluxes projected onto the target from these BOUT + + simulations follow the experimental heat flux width scaling, in particular the inverse dependence of the width on the poloidal magnetic field with an outlier. Further turbulence statistics analysis shows that the blobs are generated near the pedestal peak gradient region inside the separatrix and contribute to the transport of the particle and heat in the SOL region. To understand the Goldston heuristic drift-based model, results will also be presented from self-consistent transport simulations with the electric and magnetic drifts in BOUT + + and with the sheath potential included in the

  3. Comparison between FEM and high heat flux thermal fatigue testing results of ITER divertor plasma facing mock-ups

    Energy Technology Data Exchange (ETDEWEB)

    Crescenzi, F., E-mail: fabio.crescenzi@enea.it; Roccella, S.; Visca, E.; Moriani, A.

    2014-10-15

    Highlights: • Divertor is an important part of the ITER machine. • Finite element analysis allows designers to explore multiple design options, reducing physical prototypes and optimizing design performance. • The hydraulic thermal-mechanical analysis performed by ANSYS and the test results on small-scale mock-ups manufactured by HRP were compared. • FEA results confirmed many experimental data, then it could be very useful for next design optimization. - Abstract: The divertor is one of the most challenging components of “DEMO” the next step ITER machine, so many tasks regarding modeling and experiments have been made in the past years to assess manufacturing processes, materials and thus the life-time of the components. In this context the finite element analysis (FEA) allows designers to explore multiple design options, to reduce physical prototypes and to optimize design performance. The comparison between the hydraulic thermal-mechanical analysis performed by ANSYS WORKBENCH 14.5 and the test results [1] on small-scale mock-ups manufactured with the Hot Radial Pressing (HRP) [2] technology is presented in this paper. During the thermal fatigue testing in the Efremov TSEFEY facility to assess the heat flux load-carrying capability of the mock-ups, only the surface temperature was measured, so the FEA was important because it allowed to know any other information (temperature inside the materials, local water temperature, local stress, etc.). FEA was performed coupling the thermal-hydraulic analysis, that calculated the temperature distributions on the components and the heat transfer coefficient (HTC) between water and heat sink tube, with the mechanical analysis. The comparison between analysis and testing results was based on the temperature maps of the loaded surface and on number of the cycles supported during the testing and those predicted by the mechanical analysis using the experimental fatigue curves for CuCrZr-IG, that is the structural

  4. Close Air Support versus Close Combat Attack

    Science.gov (United States)

    2012-12-06

    with parallel Air Force and Army structures and incorporating tactical air control parties ( TACP ) down to the battalion level. This system greatly...Operations Center (ASOC) and the Tactical Air Control Party ( TACP ). Each of these agencies works in concert to facilitate close air support platforms...needed to maintain combat readiness.93 At the corps level and below is a Tactical Air Control Party ( TACP ) organized under the Army Fires Cell (FC

  5. In-memory interconnect protocol configuration registers

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, Kevin Y.; Roberts, David A.

    2017-09-19

    Systems, apparatuses, and methods for moving the interconnect protocol configuration registers into the main memory space of a node. The region of memory used for storing the interconnect protocol configuration registers may also be made cacheable to reduce the latency of accesses to the interconnect protocol configuration registers. Interconnect protocol configuration registers which are used during a startup routine may be prefetched into the host's cache to make the startup routine more efficient. The interconnect protocol configuration registers for various interconnect protocols may include one or more of device capability tables, memory-side statistics (e.g., to support two-level memory data mapping decisions), advanced memory and interconnect features such as repair resources and routing tables, prefetching hints, error correcting code (ECC) bits, lists of device capabilities, set and store base address, capability, device ID, status, configuration, capabilities, and other settings.

  6. Metrics for measuring distances in configuration spaces.

    Science.gov (United States)

    Sadeghi, Ali; Ghasemi, S Alireza; Schaefer, Bastian; Mohr, Stephan; Lill, Markus A; Goedecker, Stefan

    2013-11-14

    In order to characterize molecular structures we introduce configurational fingerprint vectors which are counterparts of quantities used experimentally to identify structures. The Euclidean distance between the configurational fingerprint vectors satisfies the properties of a metric and can therefore safely be used to measure dissimilarities between configurations in the high dimensional configuration space. In particular we show that these metrics are a perfect and computationally cheap replacement for the root-mean-square distance (RMSD) when one has to decide whether two noise contaminated configurations are identical or not. We introduce a Monte Carlo approach to obtain the global minimum of the RMSD between configurations, which is obtained from a global minimization over all translations, rotations, and permutations of atomic indices.

  7. In-memory interconnect protocol configuration registers

    Science.gov (United States)

    Cheng, Kevin Y.; Roberts, David A.

    2017-09-19

    Systems, apparatuses, and methods for moving the interconnect protocol configuration registers into the main memory space of a node. The region of memory used for storing the interconnect protocol configuration registers may also be made cacheable to reduce the latency of accesses to the interconnect protocol configuration registers. Interconnect protocol configuration registers which are used during a startup routine may be prefetched into the host's cache to make the startup routine more efficient. The interconnect protocol configuration registers for various interconnect protocols may include one or more of device capability tables, memory-side statistics (e.g., to support two-level memory data mapping decisions), advanced memory and interconnect features such as repair resources and routing tables, prefetching hints, error correcting code (ECC) bits, lists of device capabilities, set and store base address, capability, device ID, status, configuration, capabilities, and other settings.

  8. Visualization of the CMS Python Configuration System

    CERN Document Server

    Erdmann, M; Hegner, B; Hinzmann, A; Klimkovich, T; Muller, G; Steggemann, J

    2010-01-01

    The job configuration system of the CMS experiment is based on the Python programming language. Software modules and their order of execution are both represented by Python objects. In order to investigate and verify configuration parameters and dependencies naturally appearing in modular software, CMS employs a graphical tool. This tool visualizes the configuration objects, their dependencies, and the information flow. Furthermore it can be used for documentation purposes. The underlying software concepts as well as the visualization are presented.

  9. A configurable CDS for the production laboratory

    CERN Document Server

    Meek, Irish

    2003-01-01

    Various aspects of a configurable chromatography data system (CDS) for the production laboratory are discussed. The Atlas CDS can be configured extensively to fit the production laboratory work flow and meet the needs of analysts. The CDS can also be configured to automatically create a sample sequence with the required number of injections and download methods to the dedicated instrument. The Atlas Quick Start wizard offers uses quick way of generating a sequence from a predefined template and starting a run. (Edited abstract).

  10. Commercial technology for aviation configuration management.

    OpenAIRE

    White, P. Scott

    1997-01-01

    Approved for public release; distribution is unlimited This thesis examines the current policy and procedures used to manage naval aviation configuration control. It recommends that the Navy consult with SABRE Decision Technologies, or a company with a similar background, to re- engineer the process for approving configuration changes and create an information technology system to manage the process. During this study, I have identified two major challenges to naval aviation configuration ...

  11. Restaurants closed over Christmas

    CERN Multimedia

    2011-01-01

    The restaurants will be closed during the Christmas holiday period : please note that all three CERN Restaurants will be closed from 5 p.m. on Wednesday, 21 December until Wednesday, 4 January inclusive. The Restaurants will reopen on Thursday, 5 January 2012.

  12. The Change of Expression Configuration Affects Identity-Dependent Expression Aftereffect but Not Identity-Independent Expression Aftereffect

    OpenAIRE

    Song, Miao; Shinomori, Keizo; Qian, Qian; Yin, Jun; Zeng, Weiming

    2015-01-01

    The present study examined the influence of expression configuration on cross-identity expression aftereffect. The expression configuration refers to the spatial arrangement of facial features in a face for conveying an emotion, e.g., an open-mouth smile vs. a closed-mouth smile. In the first of two experiments, the expression aftereffect is measured using a cross-identity/cross-expression configuration factorial design. The facial identities of test faces were the same or different from the ...

  13. Dynamic Airspace Configuration Tool (DACT) Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Metron Aviation will develop optimization algorithms and an automated tool for performing dynamic airspace configuration under different operational scenarios. The...

  14. International Space Station Configuration Analysis and Integration

    Science.gov (United States)

    Anchondo, Rebekah

    2016-01-01

    Ambitious engineering projects, such as NASA's International Space Station (ISS), require dependable modeling, analysis, visualization, and robotics to ensure that complex mission strategies are carried out cost effectively, sustainably, and safely. Learn how Booz Allen Hamilton's Modeling, Analysis, Visualization, and Robotics Integration Center (MAVRIC) team performs engineering analysis of the ISS Configuration based primarily on the use of 3D CAD models. To support mission planning and execution, the team tracks the configuration of ISS and maintains configuration requirements to ensure operational goals are met. The MAVRIC team performs multi-disciplinary integration and trade studies to ensure future configurations meet stakeholder needs.

  15. The Ragnarok Architectural Software Configuration Management Model

    DEFF Research Database (Denmark)

    Christensen, Henrik Bærbak

    1999-01-01

    The architecture is the fundamental framework for designing and implementing large scale software, and the ability to trace and control its evolution is essential. However, many traditional software configuration management tools view 'software' merely as a set of files, not as an architecture....... This introduces an unfortunate impedance mismatch between the design domain (architecture level) and configuration management domain (file level.) This paper presents a software configuration management model that allows tight version control and configuration management of the architecture of a software system...

  16. Comparison between four dissimilar solar panel configurations

    Science.gov (United States)

    Suleiman, K.; Ali, U. A.; Yusuf, Ibrahim; Koko, A. D.; Bala, S. I.

    2017-03-01

    Several studies on photovoltaic systems focused on how it operates and energy required in operating it. Little attention is paid on its configurations, modeling of mean time to system failure, availability, cost benefit and comparisons of parallel and series-parallel designs. In this research work, four system configurations were studied. Configuration I consists of two sub-components arranged in parallel with 24 V each, configuration II consists of four sub-components arranged logically in parallel with 12 V each, configuration III consists of four sub-components arranged in series-parallel with 8 V each, and configuration IV has six sub-components with 6 V each arranged in series-parallel. Comparative analysis was made using Chapman Kolmogorov's method. The derivation for explicit expression of mean time to system failure, steady state availability and cost benefit analysis were performed, based on the comparison. Ranking method was used to determine the optimal configuration of the systems. The results of analytical and numerical solutions of system availability and mean time to system failure were determined and it was found that configuration I is the optimal configuration.

  17. Configuration of the ATLAS Trigger System

    CERN Document Server

    Elsing, M; Armstrong, S; Baines, J T M; Bee, C P; Biglietti, M; Bogaerts, A; Boisvert, V; Bosman, M; Brandt, S; Caron, B; Casado, M P; Cataldi, G; Cavalli, D; Cervetto, M; Comune, G; Corso-Radu, A; Di Mattia, A; Díaz-Gómez, M; Dos Anjos, A; Drohan, J; Ellis, Nick; Epp, B; Etienne, F; Falciano, S; Farilla, A; George, S; Ghete, V M; González, S; Grothe, M; Kaczmarska, A; Karr, K M; Khomich, A; Konstantinidis, N P; Krasny, W; Li, W; Lowe, A; Luminari, L; Ma, H; Meessen, C; Mello, A G; Merino, G; Morettini, P; Moyse, E; Nairz, A; Negri, A; Nikitin, N V; Nisati, A; Padilla, C; Parodi, F; Pérez-Réale, V; Pinfold, J L; Pinto, P; Polesello, G; Qian, Z; Rajagopalan, S; Resconi, S; Rosati, S; Scannicchio, D A; Schiavi, C; Segura, E; De Seixas, J M; Shears, T G; Sivoklokov, S Yu; Smizanska, M; Soluk, R A; Stanescu, C; Tapprogge, Stefan; Touchard, F; Vercesi, V; Watson, A; Wengler, T; Werner, P; Wheeler, S; Wickens, F J; Wiedenmann, W; Wielers, M; Zobernig, G; CHEP 2003 Computing in High Energy Physics

    2003-01-01

    In this paper a conceptual overview is given of the software foreseen to configure the ATLAS trigger system. Two functional software prototypes have been developed to configure the ATLAS Level-1 emulation and the High-Level Trigger software. Emphasis has been put so far on following a consistent approach between the two trigger systems and on addressing their requirements, taking into account the specific use-case of the `Region-of-Interest' mechanism for the ATLAS Level-2 trigger. In the future the configuration of the two systems will be combined to ensure a consistent selection configuration for the entire ATLAS trigger system.

  18. Modification of SOL profiles and fluctuations with line-average density and divertor flux expansion in TCV

    Science.gov (United States)

    Vianello, N.; Tsui, C.; Theiler, C.; Allan, S.; Boedo, J.; Labit, B.; Reimerdes, H.; Verhaegh, K.; Vijvers, W. A. J.; Walkden, N.; Costea, S.; Kovacic, J.; Ionita, C.; Naulin, V.; Nielsen, A. H.; Rasmussen, J. Juul; Schneider, B.; Schrittwieser, R.; Spolaore, M.; Carralero, D.; Madsen, J.; Lipschultz, B.; Militello, F.; The TCV Team; The EUROfusion MST1 Team

    2017-11-01

    A set of Ohmic density ramp experiments addressing the role of parallel connection length in modifying scrape off layer (SOL) properties has been performed on the TCV tokamak. The parallel connection length has been modified by varying the poloidal flux expansion f x . It will be shown that this modification does not influence neither the detachment density threshold, nor the development of a flat SOL density profile which instead depends strongly on the increase of the core line average density. The modification of the SOL upstream profile, with the appearance of what is generally called a density shoulder, has been related to the properties of filamentary blobs. Blob size increases with density, without any dependence on the parallel connection length both in the near and far SOL. The increase of the density decay length, corresponding to a profile flattening, has been related to the variation of the divertor normalized collisionality Λ_div (Myra et al 2006 Phys. Plasmas 13 112502, Carralero et al, ASDEX Upgrade Team, JET Contributors and EUROfusion MST1 Team 2015 Phys. Rev. Let. 115 215002), showing that in TCV the increase of Λ_div is not sufficient to guarantee the SOL upstream profile flattening.

  19. Progress of divertor simulation research toward the realization of detached plasma using a large tandem mirror device

    Energy Technology Data Exchange (ETDEWEB)

    Nakashima, Y., E-mail: nakashma@prc.tsukuba.ac.jp [Plasma Research Center, University of Tsukuba, Tsukuba, Ibaraki 305-8577 (Japan); Takeda, H.; Ichimura, K.; Hosoi, K.; Oki, K.; Sakamoto, M.; Hirata, M.; Ichimura, M.; Ikezoe, R.; Imai, T.; Iwamoto, M.; Hosoda, Y.; Katanuma, I.; Kariya, T.; Kigure, S.; Kohagura, J.; Minami, R.; Numakura, T.; Takahashi, S.; Yoshikawa, M. [Plasma Research Center, University of Tsukuba, Tsukuba, Ibaraki 305-8577 (Japan); and others

    2015-08-15

    This paper describes the results of the experiments performed on Tandem Mirror device GAMMA 10/PDX mainly using a new “divertor simulation experimental module (D-module)” installed on one of the end mirror exits which is specially designed to investigate the physics of plasma detachment. The additional ICRF heating in the anchor-cells, connected to both ends of the central-cell, significantly increases the density in the both cells, which attained the generation of the highest particle flux up to 10{sup 23} particles/s m{sup 2} at the end-mirror exit. H{sub 2} and noble gas injection to enhance the radiation cooling in D-module was performed and a remarkable reduction of the electron temperature (from few tens eV to <3 eV) on the target plate were successfully achieved associated with the strong reduction of particle and heat flux. A significant effect of simultaneous injection with hydrogen and noble gases for detached plasma formation was recognized for the first time.

  20. Particle-in-cell simulations of the plasma interaction with poloidal gaps in the ITER divertor outer vertical target

    Science.gov (United States)

    Komm, M.; Gunn, J. P.; Dejarnac, R.; Pánek, R.; Pitts, R. A.; Podolník, A.

    2017-12-01

    Predictive modelling of the heat flux distribution on ITER tungsten divertor monoblocks is a critical input to the design choice for component front surface shaping and for the understanding of power loading in the case of small-scale exposed edges. This paper presents results of particle-in-cell (PIC) simulations of plasma interaction in the vicinity of poloidal gaps between monoblocks in the high heat flux areas of the ITER outer vertical target. The main objective of the simulations is to assess the role of local electric fields which are accounted for in a related study using the ion orbit approach including only the Lorentz force (Gunn et al 2017 Nucl. Fusion 57 046025). Results of the PIC simulations demonstrate that even if in some cases the electric field plays a distinct role in determining the precise heat flux distribution, when heat diffusion into the bulk material is taken into account, the thermal responses calculated using the PIC or ion orbit approaches are very similar. This is a consequence of the small spatial scales over which the ion orbits distribute the power. The key result of this study is that the computationally much less intensive ion orbit approximation can be used with confidence in monoblock shaping design studies, thus validating the approach used in Gunn et al (2017 Nucl. Fusion 57 046025).

  1. Fractional pressure measurements inside of the divertor baffling at W7-X with a spectroscopically assisted Penning gauge

    Science.gov (United States)

    Kremeyer, Thierry; Schmitz, Oliver; Wenzel, Uwe; Flesch, Kurt; W7-X Team

    2017-10-01

    Studies of helium exhaust from stellarator divertors is important to qualify sufficient helium exhaust for future reactors. Penning gauges assisted by spectroscopy were used to measure total neutral pressure and to resolve the D and He partial pressures. A generic feasibility test at W7-X gave successful measurements of the total as well as the fractional neutral pressures of He and H. A first prototype of a new Penning gauge probe head has been tested at UW Madison at 240 mT as well as at the PAX magnet at IPP Greifswald, Germany at 3 T and shows a near linear power law scaling between current and pressure: I = C *Pn with n = 1.0 - 1.2 for the 240 mT case and 2.3 - 2.8 for the 3 T case. Pressure measurements were achieved starting at 10-2 mbar and down to 10-6 mbar. With the new probe head, it was possible to increase the time resolution of the spectroscopically assisted fractional neutral pressure measurements to up to 1MHz. This system is now implemented at three poloidal positions at one toroidal location in W7-X and is ready for measurements. This work was funded in part by the Department of Energy under Grants DE-SC0012315 and DE-SC0014210 and from EUROfusion under Grant No 633053.

  2. Pressure Profiles and Pressure-Driven Equilibrium Currents near Small Magnetic Islands and near Divertor Separatrices: Resonance and Symmetry Effects

    Science.gov (United States)

    Radhakrishnan, Dhanush; Reiman, Allan

    2017-10-01

    A magnetic island whose width is well below a threshold value, determined by the ratio of perpendicular to parallel transport, has only a small effect on the ambient pressure gradient. We calculate the pressure gradient, and the associated pressure driven current in the neighborhood of such an island, assuming that the pressure is determined by a diffusion equation.We similarly calculate the pressure gradient and pressure driven current in the neighborhood of a divertor separatrix. For the small magnetic island, we consider a cylindrical magnetic field with perturbed circular flux surfaces. The perturbation consists of two components, one that modulates the toroidal magnetic field strength without breaking up the flux surfaces, and a second that introduces a resonant radial component of the magnetic field at the rational surface but has little effect on the toroidal field. The relative phase between the two perturbations is varied. The Pfirsch-Schluter current near the X-line is found to be much larger when both perturbations are present and the relative phase between them breaks the stellarator symmetry than it is when these conditions are not satisfied. The calculations are consistent with previous analytical work predicting a logarithmic singularity at the X-line. This work was supported by DOE Contracts Nos. DEAC02-76CH03073 and DE-AC02-09CH1146.

  3. Configurable Multi-Purpose Processor

    Science.gov (United States)

    Valencia, J. Emilio; Forney, Chirstopher; Morrison, Robert; Birr, Richard

    2010-01-01

    Advancements in technology have allowed the miniaturization of systems used in aerospace vehicles. This technology is driven by the need for next-generation systems that provide reliable, responsive, and cost-effective range operations while providing increased capabilities such as simultaneous mission support, increased launch trajectories, improved launch, and landing opportunities, etc. Leveraging the newest technologies, the command and telemetry processor (CTP) concept provides for a compact, flexible, and integrated solution for flight command and telemetry systems and range systems. The CTP is a relatively small circuit board that serves as a processing platform for high dynamic, high vibration environments. The CTP can be reconfigured and reprogrammed, allowing it to be adapted for many different applications. The design is centered around a configurable field-programmable gate array (FPGA) device that contains numerous logic cells that can be used to implement traditional integrated circuits. The FPGA contains two PowerPC processors running the Vx-Works real-time operating system and are used to execute software programs specific to each application. The CTP was designed and developed specifically to provide telemetry functions; namely, the command processing, telemetry processing, and GPS metric tracking of a flight vehicle. However, it can be used as a general-purpose processor board to perform numerous functions implemented in either hardware or software using the FPGA s processors and/or logic cells. Functionally, the CTP was designed for range safety applications where it would ultimately become part of a vehicle s flight termination system. Consequently, the major functions of the CTP are to perform the forward link command processing, GPS metric tracking, return link telemetry data processing, error detection and correction, data encryption/ decryption, and initiate flight termination action commands. Also, the CTP had to be designed to survive and

  4. Geometrical ambiguity of pair statistics: point configurations.

    Science.gov (United States)

    Jiao, Y; Stillinger, F H; Torquato, S

    2010-01-01

    Point configurations have been widely used as model systems in condensed-matter physics, materials science, and biology. Statistical descriptors, such as the n -body distribution function g(n), are usually employed to characterize point configurations, among which the most extensively used is the pair distribution function g(2). An intriguing inverse problem of practical importance that has been receiving considerable attention is the degree to which a point configuration can be reconstructed from the pair distribution function of a target configuration. Although it is known that the pair-distance information contained in g(2) is, in general, insufficient to uniquely determine a point configuration, this concept does not seem to be widely appreciated and general claims of uniqueness of the reconstructions using pair information have been made based on numerical studies. In this paper, we present the idea of the distance space called the D space. The pair distances of a specific point configuration are then represented by a single point in the D space. We derive the conditions on the pair distances that can be associated with a point configuration, which are equivalent to the realizability conditions of the pair distribution function g(2). Moreover, we derive the conditions on the pair distances that can be assembled into distinct configurations, i.e., with structural degeneracy. These conditions define a bounded region in the D space. By explicitly constructing a variety of degenerate point configurations using the D space, we show that pair information is indeed insufficient to uniquely determine the configuration in general. We also discuss several important problems in statistical physics based on the D space, including the reconstruction of atomic structures from experimentally obtained g(2) and a recently proposed "decorrelation" principle. The degenerate configurations have relevance to open questions involving the famous traveling salesman problem.

  5. Experiment Configurations for the DAST

    Science.gov (United States)

    1978-01-01

    This image shows three vehicle configurations considered for the Drones for Aerodynamic and Structural Testing (DAST) program, conducted at NASA's Dryden Flight Research Center between 1977 and 1983. The DAST project planned for three wing configurations. These were the Instrumented Standard Wing (ISW), the Aeroelastic Research Wing-1 (ARW-1), and the ARW-2. After the DAST-1 crash, project personnel fitted a second Firebee II with a rebuilt ARW-1 wing. Due to the project's ending, it never flew the ARW-2 wing. These are the image contact sheets for each image resolution of the NASA Dryden Drones for Aerodynamic and Structural Testing (DAST) Photo Gallery. From 1977 to 1983, the Dryden Flight Research Center, Edwards, California, (under two different names) conducted the DAST Program as a high-risk flight experiment using a ground-controlled, pilotless aircraft. Described by NASA engineers as a 'wind tunnel in the sky,' the DAST was a specially modified Teledyne-Ryan BQM-34E/F Firebee II supersonic target drone that was flown to validate theoretical predictions under actual flight conditions in a joint project with the Langley Research Center, Hampton, Virginia. The DAST Program merged advances in electronic remote control systems with advances in airplane design. Drones (remotely controlled, missile-like vehicles initially developed to serve as gunnery targets) had been deployed successfully during the Vietnamese conflict as reconnaissance aircraft. After the war, the energy crisis of the 1970s led NASA to seek new ways to cut fuel use and improve airplane efficiency. The DAST Program's drones provided an economical, fuel-conscious method for conducting in-flight experiments from a remote ground site. DAST explored the technology required to build wing structures with less than normal stiffness. This was done because stiffness requires structural weight but ensures freedom from flutter-an uncontrolled, divergent oscillation of the structure, driven by aerodynamic

  6. High Flux FRC Facility for the Stability, Confinement and ITER Divertor Studies

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, Alan L. [Univ. of Washington, Seattle, WA (United States). Aerospace and Energetics Research Program. Redmond Plasma Physics Lab.; Milroy, Richard D. [Univ. of Washington, Seattle, WA (United States). Aerospace and Energetics Research Program. Redmond Plasma Physics Lab.

    2014-01-31

    The TCS (Translation, Confinement, & Sustainment) program was begun on 7 August, 1996 to renew basic studies of the Field Reversed Configuration (FRC). The program made use of the old LSX (Large s Experiment) device, which was constructed at STI during the period from 1986 to 1990, but only operated for one year due to a DOE decision at the time to focus exclusively on the tokamak configuration. LSX was transferred to the University of Washington in 1992 and modified (LSX/mod) to perform Tokamak Refueling by Accelerated Plasmoids (TRAP) experiments. The TRAP program was funded from 7 August, 1992 until 6 August, 1996, but was utilized for an additional year while TCS was being constructed. During the first TCS funding period TCS was completed and initial experiments were begun. A large multi-megawatt RF power supply was built by Los Alamos National Laboratory (LANL) for use with a Rotating Magnetic Field (RMF) system, and LANL has been a continuing participant in our experimental program. A smaller prototype facility, called the Star Thrust Experiment (STX) was also built and operated in this period, partly with NASA funding, before TCS came on-line. A final report for this construction period was submitted in September 2000. A first renewal period (2.5 years) provided operating funds for the period between July 7, 2000 and January 6, 2003. A great deal of progress was made in understanding the use of RMF to both form and sustain FRCs during this period. The principal result of the experimental program was the formation of quasi steady-state (as long as RMF power was available) FRCs with densities in the 1-3x1019 m-3 range. However, the plasma temperature (Te or Ti) was limited to sub-25 eV, except transiently during start-up, by the rapid accumulation of impurities. This is not surprising since TCS was only designed to demonstrate RMF flux build-up and was not provided with either fueling capabilities or modern vacuum

  7. Microsoft System Center Configuration Manager advanced deployment

    CERN Document Server

    Coupland, Martyn

    2014-01-01

    If you are an experienced Configuration Manager administrator looking to advance your career or get more from your current environment, then this book is ideal for you. Prior experience of deploying and managing a Configuration Manager site would be helpful in following the examples throughout this book.

  8. Subsets of configurations and canonical partition functions

    DEFF Research Database (Denmark)

    Bloch, J.; Bruckmann, F.; Kieburg, M.

    2013-01-01

    We explain the physical nature of the subset solution to the sign problem in chiral random matrix theory: the subset sum over configurations is shown to project out the canonical determinant with zero quark charge from a given configuration. As the grand canonical chiral random matrix partition...

  9. Modelling Configuration Knowledge in Heterogeneous Product Families

    DEFF Research Database (Denmark)

    Queva, Matthieu Stéphane Benoit; Männistö, Tomi; Ricci, Laurent

    2011-01-01

    Product configuration systems play an important role in the development of Mass Customisation. The configuration of complex product families may nowadays involve multiple design disciplines, e.g. hardware, software and services. In this paper, we present a conceptual approach for modelling the va...

  10. Overview of offshore wind farm configurations

    Science.gov (United States)

    Wei, Q.; Wu, B.; Xu, D.; Zargari, N. R.

    2017-11-01

    Offshore wind energy has been attracting great attention. Compared with onshore wind power systems, offshore wind power applications present significantly greater economic challenges mainly due to the required bulky and costly offshore substation. To lower the cost of offshore wind power systems, various configurations are proposed in both industry and academia. The present work investigates existing offshore wind farm configurations.

  11. Measures on two-component configuration spaces

    Directory of Open Access Journals (Sweden)

    D.L. Finkelshtein

    2009-01-01

    Full Text Available We study the measures on the configuration spaces of particles of two types. Gibbs measures on such spaces are described. Main properties of corresponding relative energy densities and correlation functions are considered. In particular, we show that a support set for such Gibbs measure is the set of pairs of non-intersected configurations.

  12. Exercise in Configurable Products using Creo parametric

    DEFF Research Database (Denmark)

    Christensen, Georg Kronborg

    2017-01-01

    Family tables is a long know method with ProEngineer/Creo parametric to make families of products – like families of bolts and roller bearings. Configurable Products expand these possibilities in two major ways: First it makes configurable assemblies possible where one topologically different...

  13. Observed benefits from product configuration systems

    DEFF Research Database (Denmark)

    Hvam, Lars; Haug, Anders; Mortensen, Niels Henrik

    2013-01-01

    This article presents a study of the benefits obtained from applying product configuration systems based on a case study in four industry companies. The impacts are described according to main objectives in literature for imple-menting product configuration systems: lead time in the specification...

  14. Static Equilibrium Configurations of Charged Metallic Bodies ...

    African Journals Online (AJOL)

    When charged particles are placed on an uncharged metallic body, the charged particles redistribute themselves along the surface of the body until they reach a point or a configuration that no net tangential force is experienced on each particle. That point is referred to as electrostatic equilibrium configuration or simply as ...

  15. Configuration of Web services as parametric design

    NARCIS (Netherlands)

    Ten Teije, Annette; Van Harmelen, Frank; Wielinga, Bob

    2004-01-01

    The configuration of Web services is particularly hard given the heterogeneous, unreliable and open nature of the Web. Furthermore, such composite Web services are likely to be complex services, that will require adaptation for each specific use. Current approaches to Web service configuration are

  16. Optimal sensor configuration for complex systems

    DEFF Research Database (Denmark)

    Sadegh, Payman; Spall, J. C.

    1998-01-01

    configuration is based on maximizing the overall sensor response while minimizing the correlation among the sensor outputs. The procedure for sensor configuration is based on simultaneous perturbation stochastic approximation (SPSA). SPSA avoids the need for detailed modeling of the sensor response by simply...

  17. Aerodynamics of missiles with slotted fin configurations

    Energy Technology Data Exchange (ETDEWEB)

    Abate, G.L.; Winchenbach, G.L. (USAF, Armament Laboratory, Eglin AFB, FL (USA))

    1991-01-01

    Subsonic and transonic aerodynamic data for missiles with solid and slotted wrap around fin configurations are presented. Free-flight aeroballistic tests to obtain this data were conducted at atmospheric pressure over a Mach number range of 0.8 to 1.6. The aerodynamic coefficients and derivatives presented were extracted from the position-attitude-time histories of the experimentally measured trajectories using non-linear numerical integration data reduction routines. Results of this testing and analysis show the static and dynamic stability variations for solid and slotted wrap around fin configurations. The presence of a side moment dependent on pitch angle, inherent to wrap around fin configurations, is measured for both configurations. Results indicate a reduction in the magnitude of this side-moment for missiles with slotted fins. Also, roll dependence with Mach number effects are not present with the slotted fin configurations. Designers should consider these factors whenever wrap around fins are utilized. 14 refs.

  18. Defective Vertex Configurations in Quasicrystalline Structures

    Science.gov (United States)

    Ben-Abraham, S. I.

    Defective vertex configurations are important for the whole range of models for quasicrystalline structures from quasiperiodic tilings through random tilings to polyhedral glasses. The combinatorially possible vertex configurations are enumerated for the 1D Fibonacci chain, for the 2D Penrose pattern with its generalizations, as well as for the Beenker pattern and the triangle pattern, and for the 3D simple icosahedral tiling. The methods for quantifying the deviation of vertex configurations from perfection are reviewed. The simple method of partial dual overlap provides a means to estimate the abundancy of vertex configurations within random tilings. More sophisticated is the method of the defectivity functional; it is particularly suitable to deal with nearly perfect tilings. Local configurations are formally classified by characteristic integers: degree, rank and order. Some possible applications are hinted at.

  19. Speeding up Derivative Configuration from Product Platforms

    Directory of Open Access Journals (Sweden)

    Ruben Heradio

    2014-06-01

    Full Text Available To compete in the global marketplace, manufacturers try to differentiate their products by focusing on individual customer needs. Fulfilling this goal requires that companies shift from mass production to mass customization. Under this approach, a generic architecture, named product platform, is designed to support the derivation of customized products through a configuration process that determines which components the product comprises. When a customer configures a derivative, typically not every combination of available components is valid. To guarantee that all dependencies and incompatibilities among the derivative constituent components are satisfied, automated configurators are used. Flexible product platforms provide a big number of interrelated components, and so, the configuration of all, but trivial, derivatives involves considerable effort to select which components the derivative should include. Our approach alleviates that effort by speeding up the derivative configuration using a heuristic based on the information theory concept of entropy.

  20. Closed Claim Query File

    Data.gov (United States)

    Social Security Administration — This file is used to hold information about disability claims that have been closed and have been selected for sampling.Sampling is the process whereby OQR reviews...

  1. Surgical wound care -- closed

    Science.gov (United States)

    ... this page: //medlineplus.gov/ency/patientinstructions/000738.htm Surgical wound care - closed To use the sharing features on ... made during surgery. It is also called a "surgical wound." Some incisions are small. Others are very long. ...

  2. Density dependence of SOL power width in ASDEX upgrade L-Mode

    Directory of Open Access Journals (Sweden)

    B. Sieglin

    2017-08-01

    A recent study [4] with an open divertor configuration found an asymmetry of the power fall-off length between inner and outer target with a smaller power fall-off length λq,i on the inner divertor target. Measurements with a closed divertor configuration find a similar asymmetry for low recycling divertor conditions. It is found, in the experiment, that the in/out asymmetry λq,i/λq,o is strongly increasing with increasing density. Most notably the heat flux density at the inner divertor target is reducing with increasing λq,i whilst the total power onto each divertor target stays constant. It is found that λq,o exhibits no significant density dependence for hydrogen and deuterium but increases with about the square root of the electron density for helium. The difference between H,D and He could be due to the different recycling behaviour in the divertor. These findings may help current modelling attempts to parametrize the density dependence of the widening of the power channel and thus allow for detailed comparison to both divertor effects like recycling or increased upstream SOL cross field transport.

  3. The behaviour of magnetic field lines and drifts in 3D configurations

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, O. [Ecole Polytechnique Federale de Lausanne, Centre de Recherches en Physique des Plasmas (CRPP), CH-1015 Lausanne (Switzerland)

    2001-04-01

    The magnetic topology and the particles drift orbits in 3D configurations are analyzed with numerical tools developed during this thesis (the MFLT3D code and the VENUS code) or with existing codes (the VMEC code and the TERPSICHORE code). We will focus our study on the effect of a magnetic perturbation in a MHD equilibrium and on the neoclassical transport in new 3D reactor designs. Firstly, the magnetic structure and particle drift orbits are studied in a monotonic q-profile and in a reversed shear TEXTOR equilibrium that is subject to a magnetic perturbation driven by the Dynamic Ergodic Divertor (DED). The main results prove that there exists a transport barrier for the magnetic field lines and for circulating particles in the reversed shear case when the DED is applied. This transport barrier occurs near the surface of minimum q-value where the KAM theory may be invalid. Moreover, we have remarked that trapped particles are lost due to the presence of the ripple and the DED does not affect their trajectories. Then, we have observed that a magnetic perturbation produced by saddle coils, for example, can control internal instabilities like tearing modes in the JET tokamak. We have shown that depending on the n mode number, the saddle coils have beneficial effects on the island width of internal instabilities. Finally, the study of neoclassical transport and Q-particles confinement are analyzed in 3D reactor designs like the QAS3, the ST/sphellamak hybrid and the sphellamak. We have observed that neither the QAS3 nor the ST/sphellamak are quasiaxisymmetric configurations. Thus the transport process is governed by the helical deformation of the magnetic field strength and these configurations do not confine the trapped Q-particles. On the other hand, the sphellamak is a nearly isodynamic structure in the plasma core which leads to good Q-particle confinement and the neoclassical transport is very similar to that obtained in a 2D equivalent tokamak. (author)

  4. Scattering by closely spaced infinite cylinders in an absorbing medium

    Directory of Open Access Journals (Sweden)

    S.-C. Lee

    2011-09-01

    Full Text Available Scattering by closely spaced parallel infinite cylinders in an absorbing medium is considered in this paper. The source wave is arbitrarily polarized and propagates in a general direction at the cylinders. The formulation utilizes the Hertz potential approach, and the scattering cross section and intensity distribution in the far-field are developed. Numerical results are presented to illustrate the influence of the absorbing medium on the scattering properties of two configurations of closely-spaced cylinders.

  5. Sectorization and Configuration Transition in Airspace Design

    Directory of Open Access Journals (Sweden)

    Xiang Zou

    2016-01-01

    Full Text Available Current airspace is sectorized according to some predefined rules that are not flexible. To facilitate utilizing the airspace more efficiently, methods to design sectors need to be promoted. In this paper, we propose an undirected graph cut-based approach that employs a memetic local search-embedded constrained evolution algorithm, NSGA-II, to generate nondominated airspace configurations. We also propose a new concave hull-based method to automatically depict sector boundaries. In addition, we also study the configuration transition problem. We define the similarity of the two different configurations and calculate their similarity with a bisection diagram and a minimum cost flow algorithm. We build a forward network to represent configuration transitions across several consecutive time periods and use multiobjective dynamic programming to determine a series of nondominated configuration links from the first period to the end. We test our approaches by simulation in high-altitude airspace controlled by Beijing Area Control Center. The results show that our sectorization method outperforms the current configuration in practice, providing a lower sector number, lower intersector flow, more balanced workload distribution among the different sectors, and no constraint violations, so that the proposed approach shows its significant potential as practical applications for dynamic airspace configuration.

  6. Improving motorcycle conspicuity through innovative headlight configurations.

    Science.gov (United States)

    Ranchet, Maud; Cavallo, Viola; Dang, Nguyen-Thong; Vienne, Fabrice

    2016-09-01

    Most motorcycle crashes involve another vehicle that violated the motorcycle's right-of-way at an intersection. Two kinds of perceptual failures of other road users are often the cause of such accidents: motorcycle-detection failures and motion-perception errors. The aim of this study is to investigate the effect of different headlight configurations on motorcycle detectability when the motorcycle is in visual competition with cars. Three innovative headlight configurations were tested: (1) standard yellow (central yellow headlight), (2) vertical white (one white light on the motorcyclist's helmet and two white lights on the fork in addition to the central white headlight), and (3) vertical yellow (same configuration as (2) with yellow lights instead of white). These three headlight configurations were evaluated in comparison to the standard configuration (central white headlight) in three environments containing visual distractors formed by car lights: (1) daytime running lights (DRLs), (2) low beams, or (3) DRLs and low beams. Video clips of computer-generated traffic situations were displayed briefly (250ms) to 57 drivers. The results revealed a beneficial effect of standard yellow configuration and the vertical yellow configuration on motorcycle detectability. However, this effect was modulated by the car-DRL environment. Findings and practical recommendations are discussed with regard to possible applications for motorcycles. Copyright © 2016 Elsevier Ltd. All rights reserved.

  7. Configurational Molecular Glue: One Optically Active Polymer Attracts Two Oppositely Configured Optically Active Polymers

    Science.gov (United States)

    Tsuji, Hideto; Noda, Soma; Kimura, Takayuki; Sobue, Tadashi; Arakawa, Yuki

    2017-03-01

    D-configured poly(D-lactic acid) (D-PLA) and poly(D-2-hydroxy-3-methylbutanoic acid) (D-P2H3MB) crystallized separately into their homo-crystallites when crystallized by precipitation or solvent evaporation, whereas incorporation of L-configured poly(L-2-hydroxybutanoic acid) (L-P2HB) in D-configured D-PLA and D-P2H3MB induced co-crystallization or ternary stereocomplex formation between D-configured D-PLA and D-P2H3MB and L-configured L-P2HB. However, incorporation of D-configured poly(D-2-hydroxybutanoic acid) (D-P2HB) in D-configured D-PLA and D-P2H3MB did not cause co-crystallization between D-configured D-PLA and D-P2H3MB and D-configured D-P2HB but separate crystallization of each polymer occurred. These findings strongly suggest that an optically active polymer (L-configured or D-configured polymer) like unsubstituted or substituted optically active poly(lactic acid)s can act as “a configurational or helical molecular glue” for two oppositely configured optically active polymers (two D-configured polymers or two L-configured polymers) to allow their co-crystallization. The increased degree of freedom in polymer combination is expected to assist to pave the way for designing polymeric composites having a wide variety of physical properties, biodegradation rate and behavior in the case of biodegradable polymers.

  8. Configurational Molecular Glue: One Optically Active Polymer Attracts Two Oppositely Configured Optically Active Polymers.

    Science.gov (United States)

    Tsuji, Hideto; Noda, Soma; Kimura, Takayuki; Sobue, Tadashi; Arakawa, Yuki

    2017-03-24

    D-configured poly(D-lactic acid) (D-PLA) and poly(D-2-hydroxy-3-methylbutanoic acid) (D-P2H3MB) crystallized separately into their homo-crystallites when crystallized by precipitation or solvent evaporation, whereas incorporation of L-configured poly(L-2-hydroxybutanoic acid) (L-P2HB) in D-configured D-PLA and D-P2H3MB induced co-crystallization or ternary stereocomplex formation between D-configured D-PLA and D-P2H3MB and L-configured L-P2HB. However, incorporation of D-configured poly(D-2-hydroxybutanoic acid) (D-P2HB) in D-configured D-PLA and D-P2H3MB did not cause co-crystallization between D-configured D-PLA and D-P2H3MB and D-configured D-P2HB but separate crystallization of each polymer occurred. These findings strongly suggest that an optically active polymer (L-configured or D-configured polymer) like unsubstituted or substituted optically active poly(lactic acid)s can act as "a configurational or helical molecular glue" for two oppositely configured optically active polymers (two D-configured polymers or two L-configured polymers) to allow their co-crystallization. The increased degree of freedom in polymer combination is expected to assist to pave the way for designing polymeric composites having a wide variety of physical properties, biodegradation rate and behavior in the case of biodegradable polymers.

  9. Private Cloud Configuration with MetaConfig

    DEFF Research Database (Denmark)

    Nielsen, Thomas; Iversen, Christian; Bonnet, Philippe

    2011-01-01

    With the advent of private clouds, the challenge of configuring a mix of physical and virtual machines is no longer reserved to a few system administrator gurus. How to assign virtual machines onto physical machines to leverage the available resources? How to maintain the virtual machine configur......With the advent of private clouds, the challenge of configuring a mix of physical and virtual machines is no longer reserved to a few system administrator gurus. How to assign virtual machines onto physical machines to leverage the available resources? How to maintain the virtual machine...

  10. Tritium retention measurements by accelerator mass spectrometry and full combustion of W-coated and uncoated CFC tiles from the JET divertor

    Science.gov (United States)

    Stan-Sion, C.; Bekris, N.; Kizane, G.; Enachescu, M.; Likonen, J.; Halitovs, M.; Petre, A.; contributors, JET

    2016-04-01

    Accelerator mass spectrometry (AMS) and the full combustion method (FCM) followed by liquid scintillation counting were applied to quantitatively determine the tritium retention in the tungsten-coated carbon fibre composites (CFC), in comparison to uncoated CFC tiles from the JET divertor. The tiles were adjacent and exposed to plasma operations between 2007 and 2009. The tritium depth profiles are showing that the tritium retention on the W-coated tile was reduced by a factor of 13.5 in comparison to the uncoated tile whereas the bulk tritium concentration is approximately the same for both tiles.

  11. Statistical analysis of particle flux flowing into the end-target in between attached and detached states in the linear divertor plasma simulator NAGDIS-II

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, H. [National Institute for Fusion Science, Toki (Japan); Department of Fusion Science, SOKENDAI, Toki (Japan); Ohno, N.; Onda, T.; Takeyama, K.; Tsuji, Y. [Graduate School of Engineering, Nagoya University, Nagoya (Japan); Kajita, S.; Kuwabara, T. [Institute of Materials and Systems for Sustainability, Nagoya University (Japan)

    2016-08-15

    We have investigated the particle flux flowing into the axisymmetric end-target in the transient state from attached to detached divertor conditions in the linear plasma device NAGDIS-II. In the transient state, a dramatic decrease of the mean particle flux and a large-amplitude fluctuation with negative and positive spikes were observed. We have analyzed the fluctuation with a newly suggested analysis technique: pre-multiplied cubic spectrum with the wavelet transform. Analysis result indicates that these spikes consist of a few kilohertz components. (copyright 2015 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  12. Particle and impurity transport in the Axial Symmetric Divertor Experiment Upgrade and the Joint European Torus, experimental observations and theoretical understanding

    DEFF Research Database (Denmark)

    Angioni, C.; Carraro, L.; Dannert, T.

    2007-01-01

    Experimental observations on core particle and impurity transport from the Axial Symmetric Divertor Experiment Upgrade [O. Gruber, H.-S. Bosch, S. Gunter , Nucl Fusion 39, 1321 (1999)] and the Joint European Torus [J. Pamela, E. R. Solano, and JET EFDA Contributors, Nucl. Fusion 43, 1540 (2003......)] tokamaks are reviewed and compared. Robust general experimental behaviors observed in both the devices and related parametric dependences are identified. The experimental observations are compared with the most recent theoretical results in the field of core particle transport. (C) 2007 American Institute...

  13. Chemically deposited tungsten fibre-reinforced tungsten – The way to a mock-up for divertor applications

    Directory of Open Access Journals (Sweden)

    J. Riesch

    2016-12-01

    Full Text Available The development of advanced materials is essential for sophisticated energy systems like a future fusion reactor. Tungsten fibre-reinforced tungsten composites (Wf/W utilize extrinsic toughening mechanisms and therefore overcome the intrinsic brittleness of tungsten at low temperature and its sensitivity to operational embrittlement. This material has been successfully produced and tested during the last years and the focus is now put on the technological realisation for the use in plasma facing components of fusion devices. In this contribution, we present a way to utilize Wf/W composites for divertor applications by a fabrication route based on the chemical vapour deposition (CVD of tungsten. Mock-ups based on the ITER typical design can be realized by the implementation of Wf/W tiles. A concept based on a layered deposition approach allows the production of such tiles in the required geometry. One fibre layer after the other is positioned and ingrown into the W-matrix until the final sample size is reached. Charpy impact tests on these samples showed an increased fracture energy mainly due to the ductile deformation of the tungsten fibres. The use of Wf/W could broaden the operation temperature window of tungsten significantly and mitigate problems of deep cracking occurring typically in cyclic high heat flux loading. Textile techniques are utilized to optimise the tungsten wire positioning and process speed of preform production. A new device dedicated to the chemical deposition of W enhances significantly, the available machine time for processing and optimisation. Modelling shows that good deposition results are achievable by the use of a convectional flow and a directed temperature profile in an infiltration process.

  14. A configural dominant account of contextual cueing: Configural cues are stronger than colour cues.

    Science.gov (United States)

    Kunar, Melina A; John, Rebecca; Sweetman, Hollie

    2014-01-01

    Previous work has shown that reaction times to find a target in displays that have been repeated are faster than those for displays that have never been seen before. This learning effect, termed "contextual cueing" (CC), has been shown using contexts such as the configuration of the distractors in the display and the background colour. However, it is not clear how these two contexts interact to facilitate search. We investigated this here by comparing the strengths of these two cues when they appeared together. In Experiment 1, participants searched for a target that was cued by both colour and distractor configural cues, compared with when the target was only predicted by configural information. The results showed that the addition of a colour cue did not increase contextual cueing. In Experiment 2, participants searched for a target that was cued by both colour and distractor configuration compared with when the target was only cued by colour. The results showed that adding a predictive configural cue led to a stronger CC benefit. Experiments 3 and 4 tested the disruptive effects of removing either a learned colour cue or a learned configural cue and whether there was cue competition when colour and configural cues were presented together. Removing the configural cue was more disruptive to CC than removing colour, and configural learning was shown to overshadow the learning of colour cues. The data support a configural dominant account of CC, where configural cues act as the stronger cue in comparison to colour when they are presented together.

  15. Closing global material loops

    DEFF Research Database (Denmark)

    Prosman, Ernst-Jan; Wæhrens, Brian Vejrum; Liotta, Giacomo

    2017-01-01

    Replacing virgin materials with waste materials, a practice known as Industrial Symbiosis (IS), has been identified as a key strategy for closing material loops. This article adopts a critical view on geographic proximity and external coordinators – two key enablers of IS. By ‘uncovering’ a case...... for geographic proximity and external coordinators. In doing so, our insights into firm-level challenges of long-distance IS exchanges contribute to closing global material loops by increasing the number of potential circular pathways....

  16. The Southwest Configuration for the Next Generation Very Large Array

    Science.gov (United States)

    Irwin Kellermann, Kenneth; Carilli, Chris; Condon, James; Cotton, William; Murphy, Eric Joseph; Nyland, Kristina

    2018-01-01

    We discuss the planned array configuration for the Next Generation Very Large Array (ngVLA). The configuration, termed the "Southwest Array," consists of 214 antennas each 18 m in diameter, distributed over the Southwest United States and Northern Mexico. The antenna locations have been set applying rough real-world constraints, such as road, fiber, and power access. The antenna locations will be fixed, with roughly 50% of the antennas in a "core" of 2 km diameter, located at the site of the JVLA. Another 30% of the antennas will be distributed over the Plains of San Augustin to a diameter of 30 km, possibly along, or near, the current JVLA arms. The remaining 20% of the antennas will be distributed in a rough two-arm spiral pattern to the South and East, out to a maximum distance of 500 km, into Texas, Arizona, and Chihuahua. Years of experience with the VLA up to 50 GHz, plus intensive antenna testing up to 250 GHz for the ALMA prototype antennas, verify the VLA site as having very good observing conditions (opacity, phase stability), up to 115 GHz (ngVLA Memo No. 1). Using a suite of tools implemented in CASA, we have made extensive imaging simulations with this configuration. We find that good imaging performance can be obtained through appropriate weighting of the visibilities, for resolutions ranging from that of the core of the array (1" at 30 GHz), out to the longest baselines (10 mas at 30 GHz), with a loss of roughly a factor of two in sensitivity relative to natural weighting (ngVLA Memo No. 16). The off-set core, located on the northern edge of the long baseline configuration, provides excellent sensitivity even on the longest baselines. We are considering, in addition, a compact configuration of 16 close-packed 6 m antennas to obtain uv-coverage down to baselines ~ 10 m for imaging large scale structure, as well as a configuration including 9 stations distributed to continental scales.

  17. Configuring the development space for conceptualization

    DEFF Research Database (Denmark)

    Brønnum, Louise; Clausen, Christian

    2013-01-01

    meet and interact. Based on a case study from an industrial medical company, the paper addresses and analyses the configuration of the development space in a number of projects aiming to take up user oriented perspectives in their activities. It presents insights on how the FEI was orchestrated......This paper addresses issues of conceptualization in the early stages of concept development noted as the Front End of Innovation [FEI]. We examine this particular development space as a socio technical space where a diversity of technological knowledge, user perspectives and organizational agendas...... and staged and how different elements and objects contributed to the configuration of the space in order to make it perform in a certain way. The analysis points at the importance of the configuration processes and indicate how these configurations often may act as more or less hidden limitations on concept...

  18. Microsoft Windows 2000 Router Configuration Guide

    National Research Council Canada - National Science Library

    Richburg, Florence

    2001-01-01

    The purpose of this guide is to provide technical guidance to network administrators of small to medium size networks in the configuration and integration of Microsoft Windows 2000 Server Router features...

  19. Configurable Web Warehouses construction through BPM Systems

    Directory of Open Access Journals (Sweden)

    Andrea Delgado

    2016-08-01

    Full Text Available The process of building Data Warehouses (DW is well known with well defined stages but at the same time, mostly carried out manually by IT people in conjunction with business people. Web Warehouses (WW are DW whose data sources are taken from the web. We define a flexible WW, which can be configured accordingly to different domains, through the selection of the web sources and the definition of data processing characteristics. A Business Process Management (BPM System allows modeling and executing Business Processes (BPs providing support for the automation of processes. To support the process of building flexible WW we propose a two BPs level: a configuration process to support the selection of web sources and the definition of schemas and mappings, and a feeding process which takes the defined configuration and loads the data into the WW. In this paper we present a proof of concept of both processes, with focus on the configuration process and the defined data.

  20. Motion planning algorithms for Configuration Spaces

    OpenAIRE

    Mas-Ku, Hugo; Torres-Giese, Enrique

    2014-01-01

    We provide explicit motion planners for Euiclidean configuration spaces. This allows us to recover some known values of the topological complexity and the Lusternik-Schinirelman category of these spaces.

  1. Control Configuration Selection for Multivariable Nonlinear Systems

    DEFF Research Database (Denmark)

    Shaker, Hamid Reza; Komareji, Mohammad

    2012-01-01

    Control configuration selection is the procedure of choosing the appropriate input and output pairs for the design of SISO (or block) controllers. This step is an important prerequisite for a successful industrial control strategy. In industrial practices, it is often the case that systems, which...... are needed to be controlled, are nonlinear, and linear models are insufficient to describe the behavior of the processes. The focus of this work is on the problem of control configuration selection for such systems. A gramian-based interaction measure for control configuration selection of MIMO nonlinear...... processes is described. In general, most of the results on the control configuration selection, which have been proposed so far, can only support linear systems. The proposed gramian-based interaction measure not only supports nonlinear processes but also can be used to propose a richer sparse or block...

  2. Status Configurations, Military Service and Higher Education.

    Science.gov (United States)

    Wang, Lin; Elder, Glen H; Spence, Naomi J

    2012-12-01

    The U.S. Armed Forces offer educational and training benefits as incentives for service. This study investigates the influence of status configurations on military enlistment and their link to greater educational opportunity. Three statuses (socioeconomic status of origin, cognitive ability and academic performance) have particular relevance for life course options. We hypothesize that young men with inconsistent statuses are more likely to enlist than men with consistent status profiles, and that military service improves access to college for certain configurations. Analyses of the National Longitudinal Study of Adolescent Health (Add Health) show (1. that several status configurations markedly increased the likelihood of military enlistment and (2. within status configurations, recruits were generally more likely to enroll in higher education than nonveterans, with associate degrees being more likely.

  3. An intelligent sales assistant for configurable products

    OpenAIRE

    Molina, Martin

    2001-01-01

    Some of the recent proposals of web-based applications are oriented to provide advanced search services through virtual shops. Within this context, this paper proposes an advanced type of software application that simulates how a sales assistant dialogues with a consumer to dynamically configure a product according to particular needs. The paper presents the general knowl- edge model that uses artificial intelligence and knowledge-based techniques to simulate the configuration process. Finall...

  4. Generation Favorable Institutional Configuration Regional Business Environment

    Directory of Open Access Journals (Sweden)

    Natalia Zinovievna Solodilova

    2014-12-01

    Full Text Available This article discusses the theoretical issues of creating an enabling business environment, which is the base platform for the successful development of entrepreneurship in the regions. Provides A definition of a favorable institutional configuration of the regional business environment, which refers to forms of implementing the basic institutions and other regional institutions, taking into account existing regional system of formal and informal interaction between economic actors. States that despite the measures taken, the landscape of the Russian business community in terms of regions, remains uneven, with different indices of investment and business attractiveness, there is differentiation in business conditions in the regions with similar natural and geographical conditions and resource potential, which is primarily determined by , differences in the institutional configuration of the regional business environment and quality of interaction among the business community of the region. Hypothesis about the impossibility of creating a favorable business environment, institutional configurations at the same time in all regions of the country, as well as its limited duration. Conducted theoretical and probabilistic analysis of the parameters of creating an enabling institutional configuration of the business environment in the Russian regions. Grounded approach whereby institutional configuration of regional business environment, may be subject to management and control actions through targeted by the regional authorities can accept the specified (favorable to the business community parameters. The necessity of planning and effective management of a favorable institutional configuration of the business environment by regional authorities to increase the period of its existence.

  5. Intellectual Model-Based Configuration Management Conception

    Directory of Open Access Journals (Sweden)

    Bartusevics Arturs

    2014-07-01

    Full Text Available Software configuration management is one of the most important disciplines within the software development project, which helps control the software evolution process and allows including into the end project only tested and validated changes. To achieve this, software management completes certain tasks. Concrete tools are used for technical implementation of tasks, such as version control systems, servers of continuous integration, compilers, etc. A correct configuration management process usually requires several tools, which mutually exchange information by generating various kinds of transfers. When it comes to introducing the configuration management process, often there are situations when tool installation is started, yet at that given moment there is no general picture of the total process. The article offers a model-based configuration management concept, which foresees the development of an abstract model for the configuration management process that later is transformed to lower abstraction level models and tools are indicated to support the technical process. A solution of this kind allows a more rational introduction and configuration of tools

  6. Dynamic airspace configuration by genetic algorithm

    Directory of Open Access Journals (Sweden)

    Marina Sergeeva

    2017-06-01

    Full Text Available With the continuous air traffic growth and limits of resources, there is a need for reducing the congestion of the airspace systems. Nowadays, several projects are launched, aimed at modernizing the global air transportation system and air traffic management. In recent years, special interest has been paid to the solution of the dynamic airspace configuration problem. Airspace sector configurations need to be dynamically adjusted to provide maximum efficiency and flexibility in response to changing weather and traffic conditions. The main objective of this work is to automatically adapt the airspace configurations according to the evolution of traffic. In order to reach this objective, the airspace is considered to be divided into predefined 3D airspace blocks which have to be grouped or ungrouped depending on the traffic situation. The airspace structure is represented as a graph and each airspace configuration is created using a graph partitioning technique. We optimize airspace configurations using a genetic algorithm. The developed algorithm generates a sequence of sector configurations for one day of operation with the minimized controller workload. The overall methodology is implemented and successfully tested with air traffic data taken for one day and for several different airspace control areas of Europe.

  7. Tungsten foil laminate for structural divertor applications - Analyses and characterisation of tungsten foil

    Energy Technology Data Exchange (ETDEWEB)

    Reiser, Jens, E-mail: jens.reiser@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials (IAM-AWP) (Germany); Rieth, Michael; Dafferner, Bernhard [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials (IAM-AWP) (Germany); Hoffmann, Andreas [PLANSEE SE, Reutte (Austria); Yi Xiaoou; Armstrong, David E.J. [University of Oxford, Department of Materials (United Kingdom)

    2012-05-15

    It has been attempted for several years to synthesise a tungsten material with a low brittle-to-ductile transition temperature and a high fracture toughness that can be used for structural parts. It was shown in our previous work that tungsten foil is ductile at room temperature and that this ductility can be transformed to bulk by synthesising a tungsten laminate. In this work we want to focus on tungsten foil and assess the microstructure as well as the mechanical properties of the foil. The assessment of the microstructure of 0.1 mm tungsten foil will be performed using electron microscopy. It will be shown that the grains of the tungsten foil have a dimension of 0.5 {mu}m Multiplication-Sign 3 {mu}m Multiplication-Sign 15 {mu}m and a clear texture in (1 0 0) Left-Pointing-Angle-Bracket 0 1 1 Right-Pointing-Angle-Bracket . This texture becomes even more pronounced by annealing. Three-point-bending tests with tungsten foil, as-received, will define the barriers: ductile at room temperature and brittle in liquid nitrogen (-196 Degree-Sign C). This shows that the ductility is a thermally activated process. Recrystallised tungsten foil (annealed for 1 h/2700 Degree-Sign C) shows ductile material behaviour at 200 Degree-Sign C. The paper closes with a discussion on the reasons of the ductility of 0.1 mm tungsten foil. These might be the ultra fine grained (UFG) microstructure or, in other words, a nano microstructure (see tungsten foil as-received), the high amount of mobile edge dislocations, and/or the foil effect, which means that dislocations can move to the surface and are annihilated (see tungsten foil recrystallised).

  8. 40 CFR 63.693 - Standards: Closed-vent systems and control devices.

    Science.gov (United States)

    2010-07-01

    ...) Whenever gases or vapors containing HAP are vented through a closed-vent system connected to a control... of such devices include, but are not limited to, a car-seal or a lock-and-key configuration valve. (d...

  9. The Effects of Governing Board Configuration on Profound Organizational Change in Hospitals

    Science.gov (United States)

    Alexander, Jeffrey A.; Ye, Yining; Lee, Shoou-Yih D.; Weiner, Bryan J.

    2006-01-01

    This study extends the literature on governing boards and organizational change by examining how governing board configurations have influenced profound organizational change in U.S. hospitals, and the conditions under which such change occurs. Hospitals governed by boards that more closely resembled a corporate governance model were more likely…

  10. Investigation on the erosion/deposition processes in the ITER-like wall divertor at JET using glow discharge optical emission spectrometry technique

    Science.gov (United States)

    Ruset, C.; Grigore, E.; Luculescu, C.; Tiseanu, I.; Likonen, J.; Mayer, M.; Rubel, M.; Matthews, G. F.; contributors, JET

    2016-02-01

    As a complementary method to Rutherford back scattering (RBS), glow discharge optical emission spectrometry (GDOES) was used to investigate the depth profiles of W, Mo, Be, O and C concentrations into marker coatings (CFC/Mo/W/Mo/W) and the substrate of divertor tiles up to a depth of about 100 μm. A number of 10 samples cored from particular areas of the divertor tiles were analyzed. The results presented in this paper are valid only for those areas and they cannot be extrapolated to the entire tile. Significant deposition of Be was measured on Tile 3 (near to the top), Tile 6 (at about 40 mm from the innermost edge) and especially on Tile 0 (HFGC). Preliminary experiments seem to indicate a penetration of Be through the pores and imperfections of CFC material up to a depth of 100 μm in some cases. No erosion and a thin layer of Be (<1 μm) was detected on Tiles 4, 7 and 8. On Tile 1 no erosion was found at about 1/3 from bottom.

  11. Effects of two-dimensional magnetic uncertainties and three-dimensional error and perturbation fields on the Small Angle Slot divertor geometry and topology

    Science.gov (United States)

    Trevisan, G. L.; Lao, L. L.; Evans, T. E.; Guo, H. Y.; Orlov, D. M.; Strait, E. J.; Wingen, A.; Wu, W.

    2018-02-01

    The Small Angle Slot (SAS) was recently installed on DIII-D as an advanced divertor, promising easier plasma detachment and lower temperatures across the whole target. A twofold study of the SAS magnetic geometry and topology is presented in this paper. On one hand, a two-dimensional uncertainty quantification analysis is carried out through a Monte Carlo approach in order to understand the level of accuracy of two-dimensional equilibrium computations in reconstructing the strike point and angle onto the divertor. Under typical experimental conditions, the uncertainties are found to be roughly 6.8 mm and 0.56 deg, respectively. On the other hand, a three-dimensional ‘vacuum’ analysis is carried out to understand the effects of typical external perturbation fields on the scrape-off layer topology. When the non-axisymmetric I-coils are switched on, poloidally-localized lobes are found to appear, grow, and hit the SAS target, although barely, even for 5 kA; at the same time, the strike point modulation is found to be roughly 1.8 mm and thus negligible for most purposes. Such results complement previous two-dimensional analyses in characterizing typical SAS equilibria and provide useful background information for planning and interpreting SAS experiments.

  12. The investigation of structure, chemical composition, hydrogen isotope trapping and release processes in deposition layers on surfaces exposed to DIII-D divertor plasma

    Energy Technology Data Exchange (ETDEWEB)

    Buzhinskij, O.I.; Opimach, I.V.; Barsuk, V.A. [TRINITI, Troitsk (Russian Federation); Arkhipov, I.I. [Russian Academy of Science, Moscow (Russian Federation). Inst. of Physical Chemistry; West, W.P.; Wong, C.P.C. [General Atomics, San Diego, CA (United States); Whyte, D. [Univ. of California, San Diego, CA (United States); Wampler, W.R. [Sandia National Labs., Albuquerque, NM (United States)

    1998-05-01

    The exposure of ATG graphite sample to DIII-D divertor plasma was provided by the DiMES (Divertor Material Evaluation System) mechanism. The graphite sample arranged to receive the parallel heat flux on a small region of the surface was exposed to 600ms of outer strike point plasma. The sample was constructed to collect the eroded material directed downward into a trapping zone onto s Si disk collector. The average heat flux onto the graphite sample during the exposure was about 200W/cm{sup 2}, and the parallel heat flux was about 10 KW/cm{sup 2}. After the exposure the graphite sample and Si collector disk were analyzed using SEM, NRA, RBS, Auger spectroscopy. IR and Raman spectroscopy. The thermal desorption was studied also. The deposited coating on graphite sample is amorphous carbon layer. Just upstream of the high heat flux zone the redeposition layer has a globular structure. The deposition layer on Si disk is composed also from carbon but has a diamond-like structure. The areal density of C and D in the deposited layer on Si disk varied in poloidal and toroidal directions. The maximum D/C areal density ratio is about 0.23, maximum carbon density is about 3.8 {times} 10{sup 18}cm{sup {minus}2}, maximum D area density is about 3 {times} 10{sup 17}cm{sup 2}. The thermal desorption spectrum had a peak at 1,250K.

  13. On the application of He I collisional-radiative model to the He–H{sub 2} mixture plasmas in MAP-II divertor simulator

    Energy Technology Data Exchange (ETDEWEB)

    Iida, Y., E-mail: iida@flanker.n.t.u-tokyo.ac.jp [School of Engineering, The University of Tokyo, 7-3-1 Hongo Bunkyo-ku, Tokyo 113-8656 (Japan); Kado, S., E-mail: kado@iae.kyoto-u.ac.jp [School of Engineering, The University of Tokyo, 7-3-1 Hongo Bunkyo-ku, Tokyo 113-8656 (Japan); Tanaka, S. [School of Engineering, The University of Tokyo, 7-3-1 Hongo Bunkyo-ku, Tokyo 113-8656 (Japan)

    2013-07-15

    Atomic helium (He I) optical emission spectroscopy was carried out in a MAP-II divertor simulator to investigate the applicability of He I collisional-radiative (CR) model for the measurement of electron temperature (T{sub e}) and density (n{sub e}) in He–H{sub 2} mixture plasmas. The result was compared with Langmuir probe measurements. The excited population distribution of the He–H{sub 2} mixture plasma was not reproduced by the He I CR model, although it was reproduced well for pure He and He–Ar mixture plasma. Specifically, the deviation of the population ratio of singlet to triplet states became more evident as the concentration of H{sub 2} increased. This fact implies that the measurement of T{sub e} for He–H{sub 2} plasma using the He I CR model could lead to a significant error under a certain condition, such that T{sub e} < 10 eV as in the case of divertor plasmas.

  14. Edge turbulence and divertor heat flux width simulations of Alcator C-Mod discharges using an electromagnetic two-fluid model

    Science.gov (United States)

    Chen, B.; Xu, X. Q.; Xia, T. Y.; Porkolab, M.; Edlund, E.; LaBombard, B.; Terry, J.; Hughes, J. W.; Mao, S. F.; Ye, M. Y.; Wan, Y. X.

    2017-11-01

    The BOUT++ code has been exploited in order to improve the understanding of the role of turbulent modes in controlling edge transport and resulting scaling of the scrape-off layer (SOL) heat flux width. For the C-Mod enhanced D_α (EDA) H-mode discharges, BOUT++ six-field two-fluid nonlinear simulations show a reasonable agreement of upstream turbulence and divertor target heat flux behavior: (a) the simulated quasi-coherent modes show consistent characteristics of the frequency versus poloidal wave number spectra of the electromagnetic fluctuations when compared with experimental measurements: frequencies are around 60-120 kHz (experiment: about 70-110 kHz), k_θ are around 2.0 cm-1 which is similar to the phase contrast imaging data; (b) linear spectrum analysis is consistent with the nonlinear phase relationship calculation which indicates the dominance of resistive-ballooning modes and drift-Alfven wave instabilities; (c) the SOL heat flux width λq versus current I p scaling is reproduced by turbulent transport: the simulations yield similar λq to experimental measurements within a factor of 2. However the magnitudes of divertor heat fluxes can be varied, depending on the physics models, sources and sinks, sheath boundary conditions, or flux limiting coefficient; (d) Simple estimate by the ‘2-point model’ for λq is consistent with simulation. Moreover, blobby turbulent spreading is confirmed for these relatively high B p shots.

  15. Close pairs of relative equilibria for identical point vortices

    DEFF Research Database (Denmark)

    Dirksen, Tobias; Aref, Hassan

    2011-01-01

    Numerical solution of the classical problem of relative equilibria for identical point vortices on the unbounded plane reveals configurations that are very close to the analytically known, centered, symmetrically arranged, nested equilateral triangles. New numerical solutions of this kind are found...... also has this property, and new relative equilibria close to the nested, symmetrically arranged, regular heptagons have been found. The centered regular nonagon is also marginally stable. Again, a new family of close relative equilibria has been found. The closest relative equilibrium pairs occur...

  16. Closed timelike curves

    CERN Document Server

    Thorne, K S

    1993-01-01

    This lecture reviews recent research on closed timelike curves (CTCS), including these questions: Do the laws of physics prevent CTCs from ever forming in classical spacetime? If so, by what physical mechanism are C'I‘Cs prevented? Can the laws of physics be adapted in any reasonable way to a. spacetime that contains C'I‘Cs, or do they necessarily give nonsense? What insights into quantum gravity can one gain by asking questions such as these?

  17. Configuring Airspace Sectors with Approximate Dynamic Programming

    Science.gov (United States)

    Bloem, Michael; Gupta, Pramod

    2010-01-01

    In response to changing traffic and staffing conditions, supervisors dynamically configure airspace sectors by assigning them to control positions. A finite horizon airspace sector configuration problem models this supervisor decision. The problem is to select an airspace configuration at each time step while considering a workload cost, a reconfiguration cost, and a constraint on the number of control positions at each time step. Three algorithms for this problem are proposed and evaluated: a myopic heuristic, an exact dynamic programming algorithm, and a rollouts approximate dynamic programming algorithm. On problem instances from current operations with only dozens of possible configurations, an exact dynamic programming solution gives the optimal cost value. The rollouts algorithm achieves costs within 2% of optimal for these instances, on average. For larger problem instances that are representative of future operations and have thousands of possible configurations, excessive computation time prohibits the use of exact dynamic programming. On such problem instances, the rollouts algorithm reduces the cost achieved by the heuristic by more than 15% on average with an acceptable computation time.

  18. Power Optimization Configuration for Piezoelectric Cantilever Arrays

    Directory of Open Access Journals (Sweden)

    Jing Bong Yu

    2017-01-01

    Full Text Available This paper investigates the changes in the output of the piezoelectric cantilever arrays when connected in different configurations. In this research matching load resistance determined and optimum output was measured by connecting the piezoelectric cantilever arrays to resistance ranging from 10 Ω to 1 MΩ while excited by constant vibration source at frequency of 300 Hz and acceleration of 1-g level. The result shows that matching load resistance for one single piezoelectric cantilever is 13 KΩ. When two, three and four cantilevers are connected in series, the matching load resistance is 26 kΩ, 39 kΩ and 52 kΩ respectively. While in parallel connection, matching load resistance reduced to 6.5 kΩ, 4.5 kΩ and 3.5 kΩ for two, three, and four connected cantilevers respectively. In series configuration, the voltage output produced is much higher as compared to the piezoelectric cantilever arrays that are connected in parallel connection. The voltage output of the piezoelectric cantilever increased from 3.41V to 6.09V when it is connected in series configuration with same polarity. Whereas in term of power output, piezoelectric cantilever arrays in parallel configuration produce higher power output as compared to piezoelectric cantilever arrays in series connection. The maximum power increased from 272μW to 521μW when two cantilevers are connected in parallel configuration with same polarity.

  19. Control Configuration Selection for Multivariable Descriptor Systems

    DEFF Research Database (Denmark)

    Shaker, Hamid Reza; Stoustrup, Jakob

    2012-01-01

    Control configuration selection is the procedure of choosing the appropriate input and output pairs for the design of SISO (or block) controllers. This step is an important prerequisite for a successful industrial control strategy. In industrial practices it is often the case that the system, whi...... is that it can be used to propose a richer sparse or block diagonal controller structure. The interaction measure is used for control configuration selection of the linearized CSTR model with descriptor from.......Control configuration selection is the procedure of choosing the appropriate input and output pairs for the design of SISO (or block) controllers. This step is an important prerequisite for a successful industrial control strategy. In industrial practices it is often the case that the system, which...... systems, hydraulic systems to heat transfer, and chemical processes. The focus of this paper is on the problem of control configuration selection for multivariable descriptor systems. A gramian-based interaction measure for control configuration selection of such processes is described in this paper...

  20. A delta configured auxiliary resonant snubber inverter

    Energy Technology Data Exchange (ETDEWEB)

    Lai, J.S.; Young, R.W.; Ott, G.W. Jr.; McKeever, J.W. [Oak Ridge National Lab., TN (United States). Engineering Technology Div.; Peng, F.Z. [Univ. of Tennessee, Knoxville, TN (United States)]|[Oak Ridge National Lab., TN (United States)

    1995-09-01

    A delta ({Delta}) configured auxiliary resonant snubber inverter is developed to overcome the voltage floating problem in a wye (Y) configured resonant snubber inverter. The proposed inverter is to connect auxiliary resonant branches between phase outputs to avoid a floating point voltage which may cause over-voltage failure of the auxiliary switches. Each auxiliary branch consists of a resonant inductor and a reverse blocking auxiliary switch. Instead of using an anti-paralleled diode to allow resonant current to flow in the reverse direction, as in the Y-configured version, the resonant branch in the {Delta}-configured version must block the negative voltage, typically done by a series diode. This paper shows single-phase and three-phase versions of {Delta}-configured resonant snubber inverters and describes in detail the operating principle of a single-phase version. The extended three-phase version is proposed with non-adjacent state space vector modulation. For hardware implementation, a single-phase 1-kW unit and a three-phase 100-kW unit were built to prove the concept. Experimental results show the superiority of the proposed topology.