WorldWideScience

Sample records for divertor triple probe

  1. Upgrade of Langmuir probe diagnostic in ITER-like tungsten mono-block divertor on experimental advanced superconducting tokamak

    Science.gov (United States)

    Xu, J. C.; Wang, L.; Xu, G. S.; Luo, G. N.; Yao, D. M.; Li, Q.; Cao, L.; Chen, L.; Zhang, W.; Liu, S. C.; Wang, H. Q.; Jia, M. N.; Feng, W.; Deng, G. Z.; Hu, L. Q.; Wan, B. N.; Li, J.; Sun, Y. W.; Guo, H. Y.

    2016-08-01

    In order to withstand rapid increase in particle and power impact onto the divertor and demonstrate the feasibility of the ITER design under long pulse operation, the upper divertor of the EAST tokamak has been upgraded to actively water-cooled, ITER-like tungsten mono-block structure since the 2014 campaign, which is the first attempt for ITER on the tokamak devices. Therefore, a new divertor Langmuir probe diagnostic system (DivLP) was designed and successfully upgraded on the tungsten divertor to obtain the plasma parameters in the divertor region such as electron temperature, electron density, particle and heat fluxes. More specifically, two identical triple probe arrays have been installed at two ports of different toroidal positions (112.5-deg separated toroidally), which can provide fundamental data to study the toroidal asymmetry of divertor power deposition and related 3-dimension (3D) physics, as induced by resonant magnetic perturbations, lower hybrid wave, and so on. The shape of graphite tip and fixed structure of the probe are designed according to the structure of the upper tungsten divertor. The ceramic support, small graphite tip, and proper connector installed make it possible to be successfully installed in the very narrow interval between the cassette body and tungsten mono-block, i.e., 13.5 mm. It was demonstrated during the 2014 and 2015 commissioning campaigns that the newly upgraded divertor Langmuir probe diagnostic system is successful. Representative experimental data are given and discussed for the DivLP measurements, then proving its availability and reliability.

  2. Ion temperature measurement using an ion sensitive probe in the LHD divertor plasma

    Energy Technology Data Exchange (ETDEWEB)

    Ezumi, N. E-mail: ezumi@ec.nagano-nct.ac.jp; Masuzaki, S.; Ohno, N.; Uesugi, Y.; Takamura, S

    2003-03-01

    The first reliable measurement of ion temperature in the divertor plasma of the Large Helical Device has been done by using an ion sensitive probe. The satisfactory current-voltage characteristics of the ion collector for evaluating the ion temperature were obtained at the outer part of the divertor leg. Furthermore, simultaneous ion and electron temperature measurements were successfully done in this part. The results show that the ion temperature is higher than the electron temperature in the part. There is a possibility that the profiles of the evaluated ion temperature which shows relatively higher than the electron temperature at the outside of divertor leg are qualitatively explained by particle's orbits around the edge and divertor region.

  3. Investigation of SOL parameters and divertor particle flux from electric probe measurements in KSTAR

    Energy Technology Data Exchange (ETDEWEB)

    Bak, J.G., E-mail: jgbak@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Kim, H.S. [National Fusion Research Institute, Daejeon (Korea, Republic of); Bae, M.K. [Hanyang University, Seoul (Korea, Republic of); Juhn, J.W.; Seo, D.C.; Bang, E.N. [National Fusion Research Institute, Daejeon (Korea, Republic of); Shim, S.B. [Pusan National University, Pusan (Korea, Republic of); Chung, K.S. [Hanyang University, Seoul (Korea, Republic of); Lee, H.J. [Pusan National University, Pusan (Korea, Republic of); Hong, S.H. [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-08-15

    The upstream scrape-off layer (SOL) profiles and downstream particle fluxes are measured with a fast reciprocating Langmuir probe assembly (FRLPA) at the outboard mid-plane and a fixed edge Langmuir probe array (ELPA) at divertor region, respectively in the KSTAR. It is found that the SOL has a two-layer structure in the outboard wall-limited (OWL) ohmic and L-mode: a near SOL (∼5 mm zone) with a narrow feature and a far SOL with a broader profile. The near SOL width evaluated from the SOL profiles in the OWL plasmas is comparable to the scaling for the L-mode divertor plasmas in the JET and AUG. In the SOL profiles and the divertor particle flux profile during the ELMy H-modes, the characteristic e-folding lengths of electron temperature, plasma density and particle flux during an ELM phase are about two times larger than ones at the inter ELM.

  4. High heat flux Langmuir probe array for the DIII-D divertor platesa)

    Science.gov (United States)

    Watkins, J. G.; Taussig, D.; Boivin, R. L.; Mahdavi, M. A.; Nygren, R. E.

    2008-10-01

    Two modular arrays of Langmuir probes designed to handle a heat flux of up to 25 MW/m2 for 10 s exposures have been installed in the lower divertor target plates of the DIII-D tokamak. The 20 pyrolytic graphite probe tips have more than three times higher thermal conductivity and 16 times larger mass than the original DIII-D isotropic graphite probes. The probe tips have a fixed 12.5° surface angle to distribute the heat flux more uniformly than the previous 6 mm diameter domed collectors and a symmetric "rooftop" design to allow operation with reversed toroidal magnetic field. A large spring-loaded contact area improves heat conduction from each probe tip through a ceramic insulator into a cooled graphite divertor floor tile. The probe tips, brazed to molybdenum foil to ensure good electrical contact, are mounted in a ceramic tray for electrical isolation and reliable cable connections. The new probes are located 1.5 cm radially apart in a staggered arrangement near the entrance to the lower divertor pumping baffle and are linearly spaced 3 cm apart on the shelf above the in-vessel cryopump. Typical target plate profiles of Jsat, Te, and Vf with 4 mm spatial resolution are shown.

  5. Edge and divertor plasma measurements with ion sensitive and Mach probes in LHD

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Y., E-mail: shihaya_uki884@yahoo.co.jp [Nagano National College of Technology, 716 Tokuma, Nagano 381-8550 (Japan); Ezumi, N. [Nagano National College of Technology, 716 Tokuma, Nagano 381-8550 (Japan); Masuzaki, S.; Tanaka, H.; Kobayashi, M. [National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292 (Japan); Sawada, K. [Shinshu University, Wakasato, Nagano 380-8553 (Japan); Ohno, N. [Nagoya University, Furo-cho Chikusa-ku, Nagoya 464-8603 (Japan)

    2013-07-15

    Spatial profiles of plasma flow and Mach number in the stochastic magnetic boundary layer as well as ion temperature (T{sub i}) and electron temperature (T{sub e}) in the divertor region in Large Helical Device (LHD) have been studied by a movable multiple functions probe, which consists of Mach probes and an ion sensitive probe. The results of ion saturation current measurements indicated plasma flow direction is alternated in the stochastic magnetic boundary. Mach number profiles for different plasma densities have been evaluated experimentally which compared with 3-D transport code. T{sub i} and T{sub e} in the divertor region measured by the ion sensitive probe decreased with increasing line-averaged density. Although T{sub i} was higher than T{sub e} in the low density plasma, both temperatures became almost the same at higher density.

  6. 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.

  7. Real-time control of divertor detachment in H-mode with impurity seeding using Langmuir probe feedback in JET-ITER-like wall

    Science.gov (United States)

    Guillemaut, C.; Lennholm, M.; Harrison, J.; Carvalho, I.; Valcarcel, D.; Felton, R.; Griph, S.; Hogben, C.; Lucock, R.; Matthews, G. F.; Perez Von Thun, C.; Pitts, R. A.; Wiesen, S.; contributors, JET

    2017-04-01

    Burning plasmas with 500 MW of fusion power on ITER will rely on partially detached divertor operation to keep target heat loads at manageable levels. Such divertor regimes will be maintained by a real-time control system using the seeding of radiative impurities like nitrogen (N), neon or argon as actuator and one or more diagnostic signals as sensors. Recently, real-time control of divertor detachment has been successfully achieved in Type I ELMy H-mode JET-ITER-like wall discharges by using saturation current (I sat) measurements from divertor Langmuir probes as feedback signals to control the level of N seeding. The degree of divertor detachment is calculated in real-time by comparing the outer target peak I sat measurements to the peak I sat value at the roll-over in order to control the opening of the N injection valve. Real-time control of detachment has been achieved in both fixed and swept strike point experiments. The system has been progressively improved and can now automatically drive the divertor conditions from attached through high recycling and roll-over down to a user-defined level of detachment. Such a demonstration is a successful proof of principle in the context of future operation on ITER which will be extensively equipped with divertor target probes.

  8. The Dynamic Ergodic Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Lehnen, M.; Adbullaev, S.; Biel, W.; Bock, M. F. M.; Brezinsek, S.; Busch, C.; Classen, I.; Finken, K. H.; Hartin, D.; Hellermann, M. von; Jachmich, S.; Jakubowski, M.; Jaspers, R.; Koslowski, H. R.; Kramer-Flecken, A.; Kikuchi, Y.; Liang, Y.; Loozen, X.; Pospieszczyk, A.; Rompuy, T. van; Reiter, D.; Samm, U.; Schmitz, O.; Sergienko, G.; Tokar, M.; Unterberg, B.; Wolf, R.; Zimmermann, O.

    2005-07-01

    The concept of the Dynamic Ergodic Divertor (DED) is based on plasma edge ergodisation by a resonant perturbation. Such a divertor concept is closely related to helical or island divertors in stellarators. The base mode of the DED perturbation field can be m/n = 12 /4, 6/2 or 3/1. The 3/1 base mode with its deep penetration of the perturbation field provides the excitation of tearing modes. This topic was presented elsewhere. In this contribution we concentrate on the divertor properties of the DED. We report on the characterisation of the topology, transport properties in ergodic fields, divertor regimes, impurity transport and density limit behaviour. The 12/4 base mode where the perturbation is restricted to the plasma edge is suitable for divertor operation. With increasing perturbation field island chains are built up at the resonance layers. Overlapping islands lead to ergodisation. The plasma is guided in the laminar region via open field lines of short connection length to the divertor target. The magnetic topology is not only controlled by the coil current but especially by the edge safety factor. For appropriate edge safety factor we observe a strong temperature drop in the plasma edge, indicating an expanding laminar region, which is necessary to decouple the divertor plasma from the core plasma. This temperature drop is accompanied by a redistribution of the heat and particle flux on the divertor target which is measured by thermography, visible spectroscopy and Langmuir probes. The modifications of the magnetic topology by the DED are reflected in the distribution of the plasma edge density and temperature measured by atomic beams and can be directly seen for example from carbon emission lines. The magnetic structure is calculated by the ATLAS code and shows good agreement with the experimental findings. The particle and energy transport is modelled with the EMC3-EIRENE code package and is in qualitative agreement with the measured densities and

  9. 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.

  10. Turbulent flow mwasurements with a triple-split hot-film probe

    Science.gov (United States)

    Doiron, M. D.; Zingg, D. W.

    1994-09-01

    Complex turbulent shear flows occur in many aerospace applications, such as aerodynamic devices and gas turbine engines. Measurements of mean and fluctuating velocity components can greatly aid our understanding of such flows. Experimental data are particularly useful in assessing and validating turbulence models used in computational fluid dynamics codes. Velocity measurements are generally made using a pitot-static tube, a constant temperature hot-wire anemometer, or a laser Doppler anemometer (LDA). For separated turbulent flows, pitot-static tubes and conventional hot-wire probes are generally inapplicable. Because of the high cost of LDA measurements, modified hot-wire techniques have been developed which are suitable for reversed flows. These include pulsed hot wires and flying hot wires. Disadvantages of these approaches are discussed by Nakayama. Triple-split hot-film probes are a potentially useful alternative for velocity measurements in separated turbulent flows. Such probes typically consist of three separate films deposited on a cylinder. The operating principle is based on the variation of the local heat transfer coefficient on a cylinder with the magnitude and direction of the oncoming flow velocity. Most studies involving split-film anemometry have been with double-split hot-film probes. These operate on the same principle but retain the directional ambiguity of conventional hot wires and, hence, are not applicable to separated turbulent flows. The results of these studies indicate that split-film probes provide comparable accuracy to hot-wire probes for mean velocities but have a more limited frequency response. Despite their potential, especially for measurements of mean velocities, triple-split hot-film probes have received little use. The only example of their use known to the authors is reported by Modera, who used a triple-split probe for low-frequency reversed flow measurements over a 0 - 8 m/s flow speed range. The purpose of this Note is to

  11. 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...

  12. A large divertor manipulator for ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Herrmann, Albrecht, E-mail: albrecht.herrmann@ipp.mpg.de; Jaksic, Nikola; Leitenstern, Peter; Greuner, Henri; Krieger, Karl; Marné, Pascal de; Oberkofler, Martin; Rohde, Volker; Schall, Gerd

    2015-10-15

    Highlights: • A large divertor manipulator for ASDEX Upgrade is developed and tested. • It allows replacing a relevant part of the divertor by dedicated targets and probes. • Modified solid standard targets. • Electrical and mechanical probes for dedicated investigations. • Test of actively cooled component. - Abstract: In 2013 a new bulk tungsten divertor, Div-III, was installed in ASDEX Upgrade (AUG). During the concept and design phase of Div-III the option of adaptable divertor instrumentation and divertor modification as contribution for divertor investigations in preparation of ITER was given a high priority. To gain flexibility for the test of divertor modifications without affecting the operational space of AUG, the large divertor manipulator, DIM-II, was designed and installed. DIM-II allows to retract 2 out of 128 outer divertor target tiles including the water cooled support structure into a target exchange box and to replace these targets without breaking the vacuum of the AUG vessel. DIM-II is based on a carriage-rail system with a driving rod pushing a front-end with the target module into the divertor position for plasma operation. Three types of front-ends are foreseen for physics investigations: (i) modified standard targets clamped to the standard cooling structure, (ii) dedicated front-ends making use of the whole available volume of about 230 × 160 × 80 mm{sup 3} and (iii) actively cooled/heated targets for cooling water temperatures up to 230 °C. This paper presents the DIM-II design including the FEM calculations for the modified divertor support structure and the front-end options, as well as the test procedure and operation mode.

  13. How to interprete triple probe measurements when non of the tips saturates?

    Energy Technology Data Exchange (ETDEWEB)

    Laux, Michael [Max-Planck-Institut fuer Plasmaphysik, Bereich Plasmadiagnostik, Berlin (Germany)

    2004-11-01

    For safety reasons (e.g. to avoid parasitic discharges) there is a tendency to keep the voltage applied to triple probes in fusion plasmas as low as possible. In the case of high electron temperatures this can result in a gradual loss of saturation for the so-called saturated tip rendering the classical data evaluation non-applicable. Nevertheless, plasma parameters can still be extracted on the basis of the analytical discussion and numerical solution of the non-linear current balance equation describing the given situation. (copyright 2004 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  14. Triple probe interrogation of spokes in a HiPIMS discharge

    Science.gov (United States)

    Lockwood Estrin, F.; Karkari, S. K.; Bradley, J. W.

    2017-07-01

    Using a triple probe situated above the racetrack and inside the magnetic trap of a magnetron, rotating spoke-like structures have been clearly identified in a single HiPIMS pulse as periodic modulations of the electron temperature T e, electron density n e, ion saturation current I isat, floating potential V f and plasma potential V p. The spokes rotate in the E  ×  B direction with a velocity of ~8.8 km s-1. Defining the spoke shape from the footprint of the ion current, they deliver to flush-mounted probes embedded in the target, each spoke can be characterised by a dense but cool leading edge (n e ~ 2.0  ×  1019 m-3, T e ~ 2.1 eV) and a relatively hotter but more rarefied trailing edge (n e ~ 1  ×  1019 m-3, T e ~ 3.9 eV). Measurements of V p show a potential hump towards the rear of the spoke, separated from regions of the highest density, with plasma potentials up to 8 V more positive than the inter-spoke regions. Azimuthal electric fields of ~1 kV m-1 associated with these structures are calculated. Transforming the triple probe time-traces to functions of the azimuthal angle θ and assuming a Gaussian radial profile for the plasma parameters, 2D spatial maps of n e, T e and V p have been constructed as well as the target ion current density J p from the embedded probes. The phase relationship between T e, V p and n e can be clearly seen using this representation with n e leading T e and V p with a phase shift between them of ~50°. Regions of maximum ion current to the target, delivered by individual spokes, coincide with the overlap of regions of high n e and T e measured above the target at a height of 15 mm. Ions created at elevated positions above the target in the observed dense region will take several micro-seconds to reach that surface, so contributing to the target ion current in the following spokes.

  15. Study of Scrape-Off-Layer Width in Ohmic and Lower Hybrid Wave Heated Double-Null Divertor Plasma in EAST%Study of Scrape-Off-Layer Width in Ohmic and Lower Hybrid Wave Heated Double-Null Divertor Plasma in EAST

    Institute of Scientific and Technical Information of China (English)

    王亮; 刘鹏; 蒋敏; 熊豪; 万宝年; 高翔; 郭后扬; 胡立群; 吴振伟; 朱思铮; 罗广南; 徐国盛; 常加峰; 张炜; 颜宁; 丁斯晔; 刘少承; 明廷凤; 汪惠乾

    2011-01-01

    Edge profiles in Ohmic and lower hybrid (LH) wave heated discharges in EAST are presented. A comparison of the measured profiles is made with those from the theoretical prediction for the scrape-off-layer (SOL) width. The edge plasma parameters are diagnosed by a triple probe divertor diagnostic system and fast reciprocating probes at the outer mid-plane. The experimental results show that the SOL width of double-null (DN) divertor plasmas in EAST appears to exhibit a negative dependence on the power crossing the separatrix, which is consistent with the collisional SOL scalings of JET and Alcator C-Mod. This will provide useful information for extrapolation to the ITER SOL width scaling for power deposition.

  16. Triple Higgs Coupling as a Probe of the Twin-Peak Scenario

    CERN Document Server

    Ahriche, Amine; Nasri, Salah

    2015-01-01

    In this letter, we investigate the case of a twin peak around the observed 125 GeV scalar resonance, using di-Higgs production processes at both LHC and $e^{+}e^{-}$ Linear Colliders. We show that the triple Higgs couplings play an important role to identify this scenario; and also that this scenario is surely distinguishable from any Standard Model extension by extra massive particles which might modify the triple Higgs coupling.

  17. Synergy between measurements of the gravitational wave and the triple Higgs coupling in probing first order phase transition

    CERN Document Server

    Hashino, Katsura; Kanemura, Shinya; Matsui, Toshinori

    2016-01-01

    Probing the Higgs potential and new physics behind the electroweak symmetry breaking is one of the most important issues of particle physics. In particular, nature of electroweak phase transition is essential for understanding physics at the early Universe, such that the strongly first order phase transition is required for a successful scenario of electroweak baryogenesis. The strongly first order phase transition is expected to be tested by precisely measuring the triple Higgs boson coupling at future colliders like the International Linear Collider. It can also be explored via the spectrum of stochastic gravitational waves to be measured at future space-based interferometers such as eLISA and DECIGO. We discuss complementarity of both the methods in testing the strongly first order phase transition of the electroweak symmetry in models with additional isospin singlet scalar fields with and without classical scale invariance. We find that they are synergetic in identifying specific models of electroweak sym...

  18. 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, effecti

  19. Semi-automated scoring of triple-probe FISH in human sperm using confocal microscopy.

    Science.gov (United States)

    Branch, Francesca; Nguyen, GiaLinh; Porter, Nicholas; Young, Heather A; Martenies, Sheena E; McCray, Nathan; Deloid, Glen; Popratiloff, Anastas; Perry, Melissa J

    2017-07-05

    Structural and numerical sperm chromosomal aberrations result from abnormal meiosis and are directly linked to infertility. Any live births that arise from aneuploid conceptuses can result in syndromes such as Kleinfelter, Turners, XYY and Edwards. Multi-probe fluorescence in situ hybridization (FISH) is commonly used to study sperm aneuploidy, however manual FISH scoring in sperm samples is labor-intensive and introduces errors. Automated scoring methods are continuously evolving. One challenging aspect for optimizing automated sperm FISH scoring has been the overlap in excitation and emission of the fluorescent probes used to enumerate the chromosomes of interest. Our objective was to demonstrate the feasibility of combining confocal microscopy and spectral imaging with high-throughput methods for accurately measuring sperm aneuploidy. Our approach used confocal microscopy to analyze numerical chromosomal abnormalities in human sperm using enhanced slide preparation and rigorous semi-automated scoring methods. FISH for chromosomes X, Y, and 18 was conducted to determine sex chromosome disomy in sperm nuclei. Application of online spectral linear unmixing was used for effective separation of four fluorochromes while decreasing data acquisition time. Semi-automated image processing, segmentation, classification, and scoring were performed on 10 slides using custom image processing and analysis software and results were compared with manual methods. No significant differences in disomy frequencies were seen between the semi automated and manual methods. Samples treated with pepsin were observed to have reduced background autofluorescence and more uniform distribution of cells. These results demonstrate that semi-automated methods using spectral imaging on a confocal platform are a feasible approach for analyzing numerical chromosomal aberrations in sperm, and are comparable to manual methods. © 2017 International Society for Advancement of Cytometry. © 2017

  20. A new approach to scaling of the scrape-off layer and divertor plasma in JET

    Energy Technology Data Exchange (ETDEWEB)

    Harbour, P.J.; Loarte, A.; Clement, S.; De Kock, L.; Jaeckel, H.J.; Lesourd, M.; O' Brien, D.P.; Summers, D.D.R.; Tagle, J.A. (JET Joint Undertaking, Abingdon (United Kingdom))

    1992-12-01

    An analytical model of the SOL/divertor magnetic geometry is applied to JET. Exponential decay lengths, [lambda], are related to differences in magnetic fluxes and are expressed in terms of [lambda] at midplane. Consistent values of [lambda] are usually obtained from Langmuir probes in the SOL or in the divertor, and with Lyman-[alpha] and Balmer-[alpha] profiles in the divertor. Scaling of [lambda] is presented: It is only slightly affected by , by X-point to target distance and by input power (other than the usual changes [Omega][yields]L[yields]H); it increases strongly with B[sub [phi

  1. 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

  2. Theory of Advanced Magnetic Divertors

    Science.gov (United States)

    Kotschenreuther, Michael; Valanju, Prashant; Mahajan, Swadesh; Covele, Brent

    2013-10-01

    The magnetic field structure in the SOL is the most important determinant of divertor physics. A comprehensive analytical and numerical methodology is developed to investigate SOL magnetic fields in the backdrop of two advanced divertor geometries- the X-divertor (XD) proposed and discussed in 2004, and the snowflake divertor (SFD) of 2007-2010. The analysis shows that XD and SFD represent very distinct and readily distinguishable magnetic geometries, epitomized through a differentiating metric, the Divertor Index (DI). In terms of this simple metric, the XD (DI > 1) and the SFD (DI XD flux surfaces are less convergent, in fact, divergent (flaring). These different SOL magnetics imply different physics, particularly with respect to detachment dynamics. It is also shown that some experiments on NSTX and DIII-D match both the prescription and the predictions of the 2004 XD paper. Work supported under US-DOE projects DE-FG02-04ER54742 and DE-FG02-04ER54754.

  3. Design method of divertor in tokamak reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ueda, Noriaki (Mitsubishi Atomic Power Industries, Inc., Tokyo (Japan)); Itoh, Sanae; Tanaka, Masaaki; Itoh, Kimitaka

    1991-03-01

    Computational method to design the efficient divertor configuration in tokamak reactor is presented. The two-dimensional code has been developed to analyze the distributions of the plasma and neutral particles for realistic configurations. Using this code, a method to design the efficient divertor configuration is developed. An example of new divertor, which consists of the baffle and fin plates, is analyzed. (author).

  4. 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.

  5. Actively convected liquid metal divertor

    Science.gov (United States)

    Shimada, Michiya; Hirooka, Yoshi

    2014-12-01

    The use of actively convected liquid metals with j × B force is proposed to facilitate heat handling by the divertor, a challenging issue associated with magnetic fusion experiments such as ITER. This issue will be aggravated even more for DEMO and power reactors because the divertor heat load will be significantly higher and yet the use of copper would not be allowed as the heat sink material. Instead, reduced activation ferritic/martensitic steel alloys with heat conductivities substantially lower than that of copper, will be used as the structural materials. The present proposal is to fill the lower part of the vacuum vessel with liquid metals with relatively low melting points and low chemical activities including Ga and Sn. The divertor modules, equipped with electrodes and cooling tubes, are immersed in the liquid metal. The electrode, placed in the middle of the liquid metal, can be biased positively or negatively with respect to the module. The j × B force due to the current between the electrode and the module provides a rotating motion for the liquid metal around the electrodes. The rise in liquid temperature at the separatrix hit point can be maintained at acceptable levels from the operation point of view. As the rotation speed increases, the current in the liquid metal is expected to decrease due to the v × B electromotive force. This rotating motion in the poloidal plane will reduce the divertor heat load significantly. Another important benefit of the convected liquid metal divertor is the fast recovery from unmitigated disruptions. Also, the liquid metal divertor concept eliminates the erosion problem.

  6. Advanced divertor configurations with large flux expansion

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V.A., E-mail: vlad@llnl.gov [Lawrence Livermore National Laboratory, Livermore, CA (United States); Bell, R.E.; Diallo, A.; Gerhardt, S.; Kaye, S.; Kolemen, E.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); McLean, A. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Menard, J.E.; Paul, S.F.; Podesta, M. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Raman, R. [University of Washington, Seattle, WA (United States); Ryutov, D.D. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Scotti, F.; Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Maingi, R. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mueller, D.M.; Roquemore, A.L. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Reimerdes, H.; Canal, G.P. [Ecole Polytechnique Fédérale de Lausanne, Centre de Recherches en Physique des Plasmas, Association Euratom Confédération Suisse, Lausanne (Switzerland); and others

    2013-07-15

    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, effective connection length and divertor volumetric power loss to increase beyond those in the standard divertor, potentially reducing heat flux and plasma temperature at the target. It also enables higher magnetic shear inside the separatrix, potentially affecting pedestal MHD stability. Experimental results from NSTX and TCV confirm the predicted properties of the snowflake divertor. In the NSTX, a large spherical tokamak with a compact divertor and lithium-coated graphite plasma-facing components (PFCs), the snowflake divertor operation led to reduced core and pedestal impurity concentration, as well as re-appearance of Type I ELMs that were suppressed in standard divertor H-mode discharges. In the divertor, an otherwise inaccessible partial detachment of the outer strike point with an up to 50% increase in divertor radiation and a peak divertor heat flux reduction from 3–7 MW/m{sup 2} to 0.5–1 MW/m{sup 2} was achieved. Impulsive heat fluxes due to Type-I ELMs were significantly dissipated in the high magnetic flux expansion region. In the TCV, a medium-size tokamak with graphite PFCs, several advantageous snowflake divertor features (cf. the standard divertor) have been demonstrated: an unchanged L–H power threshold, enhanced stability of the peeling–ballooning modes in the pedestal region (and generally an extended second stability region), as well as an H-mode pedestal regime with reduced (×2–3) Type I ELM frequency and slightly increased (20–30%) normalized ELM energy, resulting in a favorable average energy loss comparison to the standard divertor. In the divertor, ELM power partitioning between snowflake divertor strike points was demonstrated. The NSTX

  7. Probing triple-Higgs productions via $4b2\\gamma$ at a 100 TeV hadron collider

    CERN Document Server

    Chen, Chien-Yi; Zhao, Xiaoran; Zhong, Yi-Ming; Zhao, Zhijie

    2016-01-01

    The quartic self-coupling of the Standard Model Higgs boson can only be measured by observing the triple-Higgs production process, but it is challenging for the LHC Run 2 or ILC at a few TeV because of its extremely small production rate. In this paper, we present a detailed MC simulation study of the triple-Higgs production through gluon fusion at a 100 TeV hadron collider and explore the feasibility of observing this production mode. We focus on the decay channel $HHH\\rightarrow b\\bar{b}b\\bar{b}\\gamma\\gamma$, investigating detector effects and optimizing the kinematic cuts to discriminate the signal from the backgrounds. Our study shows that in order to observe the Standard Model triple-Higgs signal, the integrated luminosity of a 100 TeV hadron collider should be greater than $1.8\\times 10^4$ ab$^{-1}$. We also explore the dependence of the cross section upon the trilinear ($\\lambda_3$) and quartic ($\\lambda_4$) self-couplings of the Higgs. We find that, through a search in the triple Higgs production, the...

  8. Porphyrinic metal-organic framework as electrochemical probe for DNA sensing via triple-helix molecular switch.

    Science.gov (United States)

    Ling, Pinghua; Lei, Jianping; Ju, Huangxian

    2015-09-15

    An electrochemical DNA sensor was developed based on the electrocatalysis of porphyrinic metal-organic framework (MOF) and triple-helix molecular switch for signal transduction. The streptavidin functionalized zirconium-porphyrin MOF (PCN-222@SA) was prepared as signal nanoprobe via covalent method and demonstrated high electrocatalysis for O2 reduction. Due to the large steric effect, the designed nanoprobe was blocked for the interaction with the biotin labeled triple-helix immobilized on the surface of glassy carbon electrode. In the presence of target DNA, the assistant DNA in triple-helix will hybridize with target DNA, resulting in the disassembly of triple-helix molecular. Consequently, the end biotin away from the electrode was ''activated'' for easy access to the signal nanoprobe, PCN-222@SA, on the basis of biotin-streptavidin biorecognition. The introduction of signal nanoprobe to a sensor surface led to a significantly amplified electrocatalytic current towards oxygen reduction. Integrating with DNA recycling amplification of Exonuclease III, the sensitivity of the biosensor was improved significantly with detection limit of 0.29 fM. Moreover, the present method has been successfully applied to detect DNA in complex serum matrix. This porphyrinic MOF-based strategy has promising application in the determination of various analytes for signal transduction and has great potential in bioassays.

  9. Divertor Plasma Parameters During Radiative Divertor Operation on DIII--D

    Science.gov (United States)

    Allen, S. L.; Fenstermacher, M. E.; Hill, D. N.; Lasnier, C. J.; Meyer, W. H.; Porter, G. D.; Wood, R. D.; Leonard, A. W.; Mahdavi, M. A.; Petrie, T. W.; West, W. P.; Maingi, R.; Wade, M. R.; Whyte, D. G.

    1996-11-01

    A large array of divertor diagnostics has been used to characterize the DIII--D divertor conditions during radiative divertor operation. We have used both D2 and impurities to reduce the divertor heat flux. Several discharge conditions have been obtained, including attached and detached ELMing H-modes. The multi-chord Divertor Thomson Scattering (DTS) system has been used with divertor sweeping to obtain 2-D measurements of ne and Te in the divertor. The Te drops to <= 2 eV with D2 puffing, ne increases, and the electron pressure Pe decreases. The radiation zone, measured by multi-chord bolometry, moves from the inside leg of the divertor to the outside. Comparisons of the 2-D distribution of ne and Te and the radiation distribution will be presented.

  10. Magnetic Geometry and Physics of Advanced Divertors: The X-Divertor and the Snowflake

    CERN Document Server

    Kotschenreuther, Mike; Covele, Brent; Mahajan, Swadesh

    2013-01-01

    Advanced divertors are magnetic geometries where a second X-point is added in the divertor region to address the serious challenges of burning plasma power exhaust. Invoking physical arguments, numerical work, and detailed model magnetic field analysis, we investigate the magnetic field structure of advanced divertors in the physically relevant region for power exhaust - the Scrape-Off Layer (SOL). A primary result of our analysis is the emergence of a physical "metric", the Divertor Index DI, that quantifies the flux expansion increase as one goes from the main X-point to the strike point. It clearly separates three geometries with distinct consequences for divertor physics - the Standard Divertor (SD, DI = 1), and two advanced geometries: the X-Divertor (XD, DI > 1) and the Snowflake (SFD, DI < 1). The XD, therefore, cannot be classified as one variant of the Snowflake. By this measure, recent NSTX and DIIID experiments are X-Divertors, not Snowflakes.

  11. Development of divertor remote maintenance system

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Nobukazu; Oka, Kiyoshi; Akou, Kentaro; Takiguchi, Yuji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-04-01

    The ITER divertor is categorized as a scheduled maintenance component because of extreme heat and particle loads it is exposed to by plasma. It is also highly activated by 14 MeV neutrons. Reliable remote handling equipment and tools are required for divertor maintenance under intense gamma radiation. To facilitate remote maintenance, the divertor is segmented into 60 cassettes, and each cassette weighing about 25 tons and maintained and replaced through four maintenance ports each 90 degrees. Divertor cassettes must be transported toroidally and radially for replacement through maintenance ports. Remote handling involving cassette movers and carriers for toroidal and radial transport has been developed. Under the ITER R and D program, technology critical to divertor cassette maintenance is being developed jointly by Japan, E.U., and U.S. home teams. This paper summarizes divertor remote maintenance design and the status of technology development by the Japan Home Team. (author)

  12. 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

  13. Simulation Analysis of Divertor Performance in EAST

    Institute of Scientific and Technical Information of China (English)

    Zhu Sizheng; Zha Xuejun

    2005-01-01

    A detailed study of the divertor performance in the EAST has been conducted for both its double null and single null configurations. The results of the application of the SOLPS (B2/Eirene) code package to the analysis of the EAST divertor are summarized. Here we concentrate on the effects of the increased geometrical closure and variation in the magnetic topology on the behavior of divertor plasmas. The results of numerical predictions for the EAST divertor's operational window are also described in this paper.

  14. Opto-mechano-electrical tripling in ZnO nanowires probed by photocurrent spectroscopy in a high-resolution transmission electron microscope

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, C.; Golberg, D., E-mail: xuzhi@iphy.ac.cn, E-mail: golberg.dmitri@nims.go.jp [International Center for Materials Nanoarchitectonics (MANA), National Institute for Materials Science (NIMS), Namiki 1-1, Tsukuba, Ibaraki 3050044 (Japan); Graduate School of Pure and Applied Sciences, University of Tsukuba, Tennodai 1, Tsukuba, Ibaraki 3058577 (Japan); Xu, Z., E-mail: xuzhi@iphy.ac.cn, E-mail: golberg.dmitri@nims.go.jp [Beijing National Laboratory for Condensed Matter Physics, Institute of Physics, Chinese Academy of Sciences, Beijing 100190 (China); Kvashnin, D. G. [National University of Science and Technology, MISIS, Leninskiy Prospect 4, Moscow 119049 (Russian Federation); Tang, D.-M.; Xue, Y. M.; Bando, Y. [International Center for Materials Nanoarchitectonics (MANA), National Institute for Materials Science (NIMS), Namiki 1-1, Tsukuba, Ibaraki 3050044 (Japan); Sorokin, P. B. [National University of Science and Technology, MISIS, Leninskiy Prospect 4, Moscow 119049 (Russian Federation); Moscow Institute of Physics and Technology, Institutsky Lane 9, Dolgoprudny 141700 (Russian Federation)

    2015-08-31

    Photocurrent spectroscopy of individual free-standing ZnO nanowires inside a high-resolution transmission electron microscope (TEM) is reported. By using specially designed optical in situ TEM system capable of scanning tunneling microscopy probing paired with light illumination, opto-mechano-electrical tripling phenomenon in ZnO nanowires is demonstrated. Splitting of photocurrent spectra at around 3.3 eV under in situ TEM bending of ZnO nanowires directly corresponds to nanowire deformation and appearance of expanded and compressed nanowire sides. Theoretical simulation of a bent ZnO nanowire has an excellent agreement with the experimental data. The splitting effect could be explained by a change in the valence band structure of ZnO nanowires due to a lattice strain. The strain-induced splitting provides important clues for future flexible piezo-phototronics.

  15. SPIRAL field mapping on NSTX for comparison to divertor RF heat deposition

    Science.gov (United States)

    Hosea, J. C.; Perkins, R.; Jaworski, M. A.; Kramer, G. J.; Ahn, J.-W.; Bertelli, N.; Gerhardt, S.; Gray, T. K.; LeBlanc, B. P.; Maingi, R.; Phillips, C. K.; Roquemore, L.; Ryan, P. M.; Sabbagh, S.; Taylor, G.; Tritz, K.; Wilson, J. R.; NSTX Team

    2014-02-01

    Field-aligned losses of HHFW power in the SOL of NSTX have been studied with IR cameras and probes, but the interpretation of the data depends somewhat on the magnetic equilibrium reconstruction. Both EFIT02 and LRDFIT04 magnetic equilibria have been used with the SPIRAL code to provide field mappings in the scrape off layer (SOL) on NSTX from the midplane SOL in front of the HHFW antenna to the divertor regions, where the heat deposition spirals are measured. The field-line mapping spiral produced at the divertor plate with LRDFIT04 matches the HHFW-produced heat deposition best, in general. An independent method for comparing the field-line strike patterns on the outer divertor for the two equilibria is provided by measuring Langmuir probe characteristics in the vicinity of the outer vessel strike radius (OVSR) and observing the effect on floating potential, saturation current, and zero-probe-voltage current (IV=0) with the crossing of the OVSR over the probe. Interestingly, these comparisons also reveal that LRDFIT04 gives the more accurate location of the predicted OVSR, and confirm that the RF power flow in the SOL is essentially along the magnetic field lines. Also, the probe characteristics and IV=0 data indicate that current flows under the OVSR in the divertor tiles in most cases studied.

  16. Electrochemical DNA probe for Hg(2+) detection based on a triple-helix DNA and Multistage Signal Amplification Strategy.

    Science.gov (United States)

    Wang, Huan; Zhang, Yihe; Ma, Hongmin; Ren, Xiang; Wang, Yaoguang; Zhang, Yong; Wei, Qin

    2016-12-15

    In this work, an ultrasensitive electrochemical sensor was developed for detection of Hg(2+). Gold nanoparticles decorated bovine serum albumin reduction of graphene oxide (AuNP-BSA-rGO) were used as subsurface material for the immobilization of triple-helix DNA. The triple-helix DNA containing a thiol labelled single-stranded DNA (sDNA) and a thymine-rich DNA (T-rich DNA), which could be unwinded in the present of Hg(2+) to form more stable thymine-Hg(2+)-thymine (T-Hg(2+)-T) complex. T-Hg(2+)-T complex was then removed and the sDNA was left on the electrode. At this time, gold nanoparticle carrying thiol labelled cytosine-rich complementary DNA (cDNA-AuNP) could bind with the free sDNA. Meanwhile, the other free cDNA on AuNP could bind with each other in the present of Ag(+) to form the stable cytosine-Ag(+)-cytosine (C-Ag(+)-C) complex and circle amplification. Plenty of C-Ag(+)-C could form silver nanoclusters by electrochemical reduction and the striping signal of Ag could be measured for purpose of the final electrochemical detection of Hg(2+). This sensor could detect Hg(2+) over a wide concentration range from 0.1 to 130nM with a detection limit of 0.03nM.

  17. Engineering conceptual design of CFETR divertor

    Energy Technology Data Exchange (ETDEWEB)

    Peng, Xuebing, E-mail: pengxb@ipp.cas.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Shushanhu Road 350, 230031 Hefei Anhui (China); Ye, Minyou [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); Song, Yuntao [Institute of Plasma Physics, Chinese Academy of Sciences, Shushanhu Road 350, 230031 Hefei Anhui (China); School of Nuclear Science and Technology, University of Science and Technology of China, Jinzhai Road 96, 230026 Hefei Anhui (China); Mao, Xin [Institute of Plasma Physics, Chinese Academy of Sciences, Shushanhu Road 350, 230031 Hefei Anhui (China); Chen, Peiming; Qian, Xinyuan [School of Nuclear Science and Technology, University of Science and Technology of China, Jinzhai Road 96, 230026 Hefei Anhui (China)

    2015-10-15

    Highlights: • Three divertor structures for two plasma configurations, ITER-like and snowflake. • Property of enlarging wet area for all three divertors is analyzed. • The divertor accommodating with both the plasma configurations is unfeasible. • Divertor cooling system is developed. - Abstract: The China Fusion Engineering Test Reactor (CFETR), which is in conceptual design phase, aims at producing fusion power of 50–200 MW with tritium breeding ratio of ∼1.2 and duty cycle time of 0.3–0.5. Its designed main parameters are major/minor radii of 5.7 m/1.6 m and plasma current of 10 MA. Although the fusion power is lower than the one of ITER, the relative smaller machine dimensions and planed much higher auxiliary heating power of 100–140 MW make that the power exhausting for the CFETR divertor is a very critical issue. To solve this issue, the divertor should be better designed with advanced physical operation mode, advanced configuration/geometry or high efficient cooling structure. In the paper, much effort was put on the divertor configuration and geometry. With designed magnet system, three divertor configurations can be realized, ITER-like, snowflake and super-X. However, considering structural design feasibility and remote handling compatibility, only the first two configurations were selected for the first step of engineering design. Three divertors were designed. They have different first wall geometries to accommodate with different plasma configurations, one for the ITER-like, one for the snowflake and the third one for both the configurations. All three divertors employ the same cassette body as the support and the cooling water manifold for the first wall. This feature simplifies the interface of the divertor to other components in the vacuum vessel. Besides, the cooling structure and the remote maintenance concept are also introduced in the paper.

  18. 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.

  19. A Triple-Probe Channel NO{sub 2}S{sub 2}-Macrocycle: Synthesis, Sensing Characteristics and Crystal Structure of Mercury(II) Nitrate Complex

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ji Eun; Seo, Moo Lyong; Lee, Shim Sung [Gyeongsang National University, Jinju (Korea, Republic of); Choi, Kyu Seong [Kyungnam University, Masan (Korea, Republic of)

    2010-07-15

    A triple-probe channel type chemosensor based on an NO{sub 2}S{sub 2}-macrocycle functionalized with phenyltricyanovinyl group was synthesized and its sensing characteristics were examined. The pink-red solution of L changed selectively to pale yellow upon addition of Hg{sup 2+}. The selective fluorometric response of L to all the tested metal ions was studied. The results showed that a large enhancement of the fluorescence of L was observed only in the case of Hg{sup 2+}. In addition, L showed large anodic shift ({approx} 0.3 V) for the addition of excess Hg{sup 2+}. Through above three observed results by the different techniques, we confirmed that the proposed chemosensor acts as the multiple-probe channel sensing material. The crystal structure of mercury(II) nitrate complexes of L which shows a 1-D polymer network with a formula [Hg{sub 2}(L){sub 2}(NO{sub 3}){sub 2}({mu}-NO{sub 3}){sub 2}]{sub n} was also reported.

  20. 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.

  1. Septum assessment of the JET gas box divertor

    NARCIS (Netherlands)

    Rapp, J.; Fundamenski, W.; Ingesson, L. C.; Jachmich, S.; Huber, A.; Matthews, G. F.; Morgan, P.; Stamp, M. F.

    2008-01-01

    The influence of the physical isolation of inner and outer divertor volumes by a septum plate of the Mk-II gas box divertor, thus increasing divertor closure and neutral compression, on the plasma and divertor performance has been studied at the Joint European Torus (JET). The septum plate was insta

  2. First Divertor Operation on the HL-2A Tokamak

    Institute of Scientific and Technical Information of China (English)

    YANG Qing-Wei; CAO Zeng; LI Xiao-Dong; MAO Wei-Cheng; ZHOU Cai-Pin; WANG En-Yao; YAN Jian-Cheng; LIU Yong; HL-2A team; DING Xuan-Tong; YAN Long-Wen; XUAN Wei-Min; LIU De-Quan; CHEN Liao-Yuan; SONG Xian-Ming; YUAN Bao-Shan; ZHANG Jin-Hua

    2004-01-01

    @@ HL-2A device is the first divertor tokamak in China. One of its main subjects is to study the features of the divertor plasma. In the last campaign, the first divertor configuration has been achieved and sustained on the HL-2A tokamak. Here we give a brief description about the HL-2A tokamak, diagnostics arrangements, and the equilibrium analysis results on divertor configuration. The main results of divertor experiments are also presented.

  3. Characterizing the DIII-D divertor conditions during the tungsten ring experiment

    Science.gov (United States)

    Barton, J. L.; Watkins, J. G.; Wang, H. Q.; Nygren, R. E.; McLean, A.; Makowski, M.; Unterberg, E.; Thomas, D. M.; Guo, H. Y.; Guterl, J.; Buchenauer, B.

    2016-10-01

    Tungsten (W) is the leading divertor material in tokamaks, but the core W impurity fraction must be kept below 5 ×10-5 in a reactor. The DIII-D tokamak, having all graphite PFCs, has done a series of experiments with two W-coated molybdenum rings in the lower divertor to track W migration after plasma exposure. We characterize the divertor plasma conditions at the DIII-D target plate in L- and ELMing H-mode, and ELM suppressed plasmas. We will present data from an array of Langmuir probes in the divertor and divertor Thomson-scattering. We also compare the heat flux from fast thermocouples (7.5 mm below the surface of the metal tile inserts) and IRTV heat flux profiles from graphite tiles. The plasma conditions will be used to benchmark ERO modeling to aid in understanding the migration of sputtered W onto other plasma facing surfaces and will be compared to post exposure W distribution measured on the graphite tiles. Supported by US DOE under DE-AC04-94AL85000, DE-FC02-04ER54698, DE-AC05-000R22725, and DE-AC52-07NA27344.

  4. Dust divertor for a tokamak fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tang, X Z [Los Alamos National Laboratory; Delzanno, G L [Los Alamos National Laboratory

    2009-01-01

    Micron-size tungsten particulates find equilibrium position in the magnetized plasma sheath in the normal direction of the divertor surface, but are convected poloidally and toroidally by the sonic-ion-flow drag parallel to the divertor surface. The natural circulation of dust particles in the magnetized plasma sheath can be used to set up a flowing dust shield that absorbs and exhausts most of the tokamak heat flux to the divertor. The size of the particulates and the choice of materials offer substantial room for optimization.

  5. MAST-Upgrade Divertor Facility and Assessing Performance of Long-Legged Divertors

    CERN Document Server

    Fishpool, G; Cunningham, G; Harrison, J; Katramados, I; Kirk, A; Kovari, M; Meyer, H; Scannell, R

    2013-01-01

    A potentially important feature in a divertor design for a high-power tokamak is an extended and expanded divertor leg. The upgrade to MAST will allow a wide range of such divertor leg geometries to be produced, and hence will allow the roles of greatly increased connection length and flux expansion to be experimentally tested. This will include testing the potential of the Super-X configuration [1]. The design process for the upgrade has required analysis of producing and controlling the magnetic configurations, and has included consideration of the roles that divertor closure and increasing magnetic connection length will play.

  6. Synergy between measurements of gravitational waves and the triple-Higgs coupling in probing the first-order electroweak phase transition

    Science.gov (United States)

    Hashino, Katsuya; Kakizaki, Mitsuru; Kanemura, Shinya; Matsui, Toshinori

    2016-07-01

    Probing the Higgs potential and new physics behind the electroweak symmetry breaking is one of the most important issues of particle physics. In particular, the nature of the electroweak phase transition is essential for understanding the physics of the early Universe, such that the strongly first-order phase transition is required for a successful scenario of electroweak baryogenesis. The strongly first-order phase transition is expected to be tested by precisely measuring the triple Higgs boson coupling at future colliders like the International Linear Collider. It can also be explored via the spectrum of stochastic gravitational waves to be measured at future space-based interferometers such as eLISA and DECIGO. We discuss the complementarity of both the methods in testing the strongly first-order phase transition of the electroweak symmetry in models with additional isospin singlet scalar fields with and without classical scale invariance. We find that they are synergetic in identifying specific models of electroweak symmetry breaking in more detail.

  7. Intact Endogenous Metabolite Analysis of Mice Liver by Probe Electrospray Ionization/Triple Quadrupole Tandem Mass Spectrometry and Its Preliminary Application to in Vivo Real-Time Analysis.

    Science.gov (United States)

    Zaitsu, Kei; Hayashi, Yumi; Murata, Tasuku; Ohara, Tomomi; Nakagiri, Kenta; Kusano, Maiko; Nakajima, Hiroki; Nakajima, Tamie; Ishikawa, Tetsuya; Tsuchihashi, Hitoshi; Ishii, Akira

    2016-04-05

    Probe electrospray ionization (PESI) is a recently developed ionization technique that enables the direct detection of endogenous compounds like metabolites without sample preparation. In this study, we have demonstrated the first combination use of PESI with triple quadrupole tandem mass spectrometry (MS/MS), which was then applied to intact endogenous metabolite analysis of mice liver, achieving detection of 26 metabolites including amino acids, organic acids, and sugars. To investigate its practicality, metabolic profiles of control and CCl4-induced acute hepatic injury mouse model were measured by the developed method. Results showed clear separation of the two groups in score plots of principal component analysis and identified taurine as the primary contributor to group separation. The results were further validated by the established gas chromatography/MS/MS method, demonstrating the present method's usefulness. In addition, we preliminarily applied the method to real-time analysis of an intact liver of a living mouse. We successfully achieved monitoring of the real-time changes of two tricarboxylic acid cycle intermediates, α-ketoglutaric acid and fumaric acid, in the liver immediately after pyruvic acid injection via a cannulated tube to the portal vein. The present method achieved an intact analysis of metabolites in liver without sample preparation, and it also demonstrates future possibility to establish in vivo real-time metabolome analysis of living animals by PESI/MS/MS.

  8. Thioamides in the collagen triple helix.

    Science.gov (United States)

    Newberry, Robert W; VanVeller, Brett; Raines, Ronald T

    2015-06-14

    To probe noncovalent interactions within the collagen triple helix, backbone amides were replaced with a thioamide isostere. This subtle substitution is the first in the collagen backbone that does not compromise thermostability. A triple helix with a thioamide as a hydrogen bond donor was found to be more stable than triple helices assembled from isomeric thiopeptides.

  9. ARIES-III divertor engineering design

    Energy Technology Data Exchange (ETDEWEB)

    Wong, C.P.C.; Schultz, K.R. [General Atomics, San Diego, CA (United States); Cheng, E.T. [TSI Research, Solana Beach, CA (United States); Grotz, S.; Hasan, M.A.; Najmabadi, F.; Sharafat, S. [California Univ., Los Angeles, CA (United States). Dept. of Mechanical, Aerospace and Nuclear Engineering; Brooks, J.N.; Ehst, D.A.; Sze, D.K. [Argonne National Lab., IL (United States); Herring, J.S. [EG and G Idaho, Inc., Idaho Falls, ID (United States); Valenti, M.; Steiner, D. [Rensselaer Polytechnic Inst., Troy, NY (United States). Plasma Dynamics Lab.

    1992-01-01

    This paper reports the engineering design of the ARIES-III double- null divertor. The divertor coolant tubes are made from W-3Re alloy and cooled by subcooled flow boiling of organic coolant. A coating of 4 mm thick tungsten is plasma sprayed onto the divertor surface. This W layer can withstand the thermal deposition of a few disruptions. At a maximum surface heat flux of 5.4 MW/m{sup 2}, a conventional divertor design can be used. The divertor surface is contoured to have a constant heat flux of 5.4 MW/m{sup 2}. The net erosion of the W-surface was found to be negligible at about 0.1 mm/year. After 3 years of operation, the W-3Re alloy ARIES-III divertor can be disposed of as Class A waste. In order to control the prompt dose release at site boundary to less than 200 Rem, isotopic tailoring of the W-alloy will be needed.

  10. Application of the radiating divertor approach to innovative tokamak divertor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Petrie, T.W., E-mail: petrie@fusion.gat.com [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Allen, S.L.; Fenstermacher, M.E. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Groebner, R.J. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Holcomb, C.T. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Kolemen, E. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543-0451 (United States); La Haye, R.J. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Lasnier, C.J. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Leonard, A.W.; Luce, T.C. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); McLean, A.G. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Maingi, R. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543-0451 (United States); Moyer, R.A. [University of California San Diego, 9500 Gilman Dr., La Jolla, CA 92093-0417 (United States); Solomon, W.M. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543-0451 (United States); Soukhanovskii, V.A. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Turco, F. [Columbia University, 2960 Broadway, New York, NY 10027 (United States); Watkins, J.G. [Sandia National Laboratory, PO Box 5800, Albuquerque, NM 87185 (United States)

    2015-08-15

    We survey the results of recent DIII-D experiments that tested the effectiveness of three innovative tokamak divertor concepts in reducing divertor heat flux while still maintaining acceptable energy confinement under neon/deuterium-based radiating divertor (RD) conditions: (1) magnetically unbalanced high performance double-null divertor (DND) plasmas, (2) high performance double-null “Snowflake” (SF-DN) plasmas, and (3) single-null H-mode plasmas having different isolation from their divertor targets. In general, all three concepts adapt well to RD conditions, achieving significant reduction in divertor heat flux (q{sub ⊥p}) and maintaining high performance metrics, e.g., 50–70% reduction in peak divertor heat flux for DND and SF-DN plasmas that are characterized by β{sub N} ≅ 3.0 and H{sub 98(y,2)} ≈ 1.35. It is also demonstrated that q{sub ⊥p} could be reduced ≈50% by extending the parallel connection length (L{sub ||-XPT}) in the scrape-off layer between the X-point and divertor targets over a variety of the RD and non-RD environments tested.

  11. 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)

  12. Snowflake divertor configuration studies in National Spherical Torus Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V. A.; McLean, A. G.; Rognlien, T. D.; Ryutov, D. D.; Umansky, M. V. [Lawrence Livermore National Laboratory, Livermore, California 94551 (United States); Bell, R. E.; Diallo, A.; Gerhardt, S.; Kaye, S.; Kolemen, E.; LeBlanc, B. P.; Menard, J. E.; Paul, S. F.; Podesta, M.; Roquemore, A. L.; Scotti, F.; Battaglia, D.; Bell, M. G.; Gates, D. A.; Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); and others

    2012-08-15

    Experimental results from NSTX indicate that the snowflake divertor (D. Ryutov, Phys. Plasmas 14, 064502 (2007)) may be a viable solution for outstanding tokamak plasma-material interface issues. Steady-state handling of divertor heat flux and divertor plate erosion remains to be critical issues for ITER and future concept devices based on conventional and spherical tokamak geometry with high power density divertors. Experiments conducted in 4-6 MW NBI-heated H-mode plasmas in NSTX demonstrated that the snowflake divertor is compatible with high-confinement core plasma operation, while being very effective in steady-state divertor heat flux mitigation and impurity reduction. A steady-state snowflake divertor was obtained in recent NSTX experiments for up to 600 ms using three divertor magnetic coils. The high magnetic flux expansion region of the scrape-off layer (SOL) spanning up to 50% of the SOL width {lambda}{sub q} was partially detached in the snowflake divertor. In the detached zone, the heat flux profile flattened and decreased to 0.5-1 MW/m{sup 2} (from 4-7 MW/m{sup 2} in the standard divertor) indicative of radiative heating. An up to 50% increase in divertor, P{sub rad} in the snowflake divertor was accompanied by broadening of the intrinsic C III and C IV radiation zones, and a nearly order of magnitude increase in divertor high-n Balmer line emission indicative of volumetric recombination onset. Magnetic reconstructions showed that the x-point connection length, divertor plasma-wetted area and divertor volume, all critical parameters for geometric reduction of deposited heat flux, and increased volumetric divertor losses were significantly increased in the snowflake divertor, as expected from theory.

  13. Snowflake divertor configuration studies for NSTX-Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V A

    2011-11-12

    Snowflake divertor experiments in NSTX provide basis for PMI development toward NSTX-Upgrade. Snowflake configuration formation was followed by radiative detachment. Significant reduction of steady-state divertor heat flux observed in snowflake divertor. Impulsive heat loads due to Type I ELMs are partially mitigated in snowflake divertor. Magnetic control of snowflake divertor configuration is being developed. Plasma material interface development is critical for NSTX-U success. Four divertor coils should enable flexibility in boundary shaping and control in NSTX-U. Snowflake divertor experiments in NSTX provide good basis for PMI development in NSTX-Upgrade. FY 2009-2010 snowflake divertor experiments in NSTX: (1) Helped understand control of magnetic properties; (2) Core H-mode confinement unchanged; (3) Core and edge carbon concentration reduced; and (4) Divertor heat flux significantly reduced - (a) Steady-state reduction due to geometry and radiative detachment, (b) Encouraging results for transient heat flux handling, (c) Combined with impurity-seeded radiative divertor. Outlook for snowflake divertor in NSTX-Upgrade: (1) 2D fluid modeling of snowflake divertor properties scaling - (a) Edge and divertor transport, radiation, detachment threshold, (b) Compatibility with cryo-pump and lithium conditioning; (2) Magnetic control development; and (3) PFC development - PFC alignment and PFC material choice.

  14. First results from the dynamic ergodic divertor at TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Lehnen, M. [Institut fuer Plasmaphysik, Forschungszentrum Juelich GmbH, EURATOM Association, Trilateral Euregio Cluster, 52425 Juelich (Germany)]. E-mail: m.lehnen@fz-juelich.de; Abdullaev, S.S.; Biel, W.; Brezinsek, S.; Finken, K.H.; Harting, D.; Hellermann, M. von; Jakubowski, M.; Jaspers, R.; Kobayashi, M.; Koslowski, H.R.; Kraemer-Flecken, A.; Matsunaga, G.; Pospieszczyk, A.; Reiter, D.; Van Rompuy, T.; Samm, U.; Schmitz, O.; Sergienko, G.; Unterberg, B.; Wolf, R.; Zimmermann, O. [Institut fuer Plasmaphysik, Forschungszentrum Juelich GmbH, EURATOM Association, Trilateral Euregio Cluster, 52425 Juelich (Germany)

    2005-03-01

    Experimental results from the dynamic ergodic divertor (DED) at TEXTOR are given, describing the complex structure of the edge plasma and the properties of the divertor as well as its influence on the plasma rotation.

  15. Impurity-induced divertor plasma oscillations

    Energy Technology Data Exchange (ETDEWEB)

    Smirnov, R. D., E-mail: rsmirnov@ucsd.edu; Krasheninnikov, S. I.; Pigarov, A. Yu. [University of California, San Diego, La Jolla, California 92093 (United States); Kukushkin, A. S. [NRC “Kurchatov Institute”, Moscow 123182 (Russian Federation); National Research Nuclear University MEPhI, Moscow 115409 (Russian Federation); Rognlien, T. D. [Lawrence Livermore National Laboratory, Livermore, California 94551 (United States)

    2016-01-15

    Two different oscillatory plasma regimes induced by seeding the plasma with high- and low-Z impurities are found for ITER-like divertor plasmas, using computer modeling with the DUSTT/UEDGE and SOLPS4.3 plasma-impurity transport codes. The oscillations are characterized by significant variations of the impurity-radiated power and of the peak heat load on the divertor targets. Qualitative analysis of the divertor plasma oscillations reveals different mechanisms driving the oscillations in the cases of high- and low-Z impurity seeding. The oscillations caused by the high-Z impurities are excited near the X-point by an impurity-related instability of the radiation-condensation type, accompanied by parallel impurity ion transport affected by the thermal and plasma friction forces. The driving mechanism of the oscillations induced by the low-Z impurities is related to the cross-field transport of the impurity atoms, causing alteration between the high and low plasma temperature regimes in the plasma recycling region near the divertor targets. The implications of the impurity-induced plasma oscillations for divertor operation in the next generation tokamaks are also discussed.

  16. Impurity-induced divertor plasma oscillations

    Science.gov (United States)

    Smirnov, R. D.; Kukushkin, A. S.; Krasheninnikov, S. I.; Pigarov, A. Yu.; Rognlien, T. D.

    2016-01-01

    Two different oscillatory plasma regimes induced by seeding the plasma with high- and low-Z impurities are found for ITER-like divertor plasmas, using computer modeling with the DUSTT/UEDGE and SOLPS4.3 plasma-impurity transport codes. The oscillations are characterized by significant variations of the impurity-radiated power and of the peak heat load on the divertor targets. Qualitative analysis of the divertor plasma oscillations reveals different mechanisms driving the oscillations in the cases of high- and low-Z impurity seeding. The oscillations caused by the high-Z impurities are excited near the X-point by an impurity-related instability of the radiation-condensation type, accompanied by parallel impurity ion transport affected by the thermal and plasma friction forces. The driving mechanism of the oscillations induced by the low-Z impurities is related to the cross-field transport of the impurity atoms, causing alteration between the high and low plasma temperature regimes in the plasma recycling region near the divertor targets. The implications of the impurity-induced plasma oscillations for divertor operation in the next generation tokamaks are also discussed.

  17. Impact of divertor geometry on radiative divertor performance in JET H-mode plasmas

    Science.gov (United States)

    Jaervinen, A. E.; Brezinsek, S.; Giroud, C.; Groth, M.; Guillemaut, C.; Belo, P.; Brix, M.; Corrigan, G.; Drewelow, P.; Harting, D.; Huber, A.; Lawson, K. D.; Lipschultz, B.; Maggi, C. F.; Matthews, G. F.; Meigs, A. G.; Moulton, D.; Stamp, M. F.; Wiesen, S.; Contributors, JET

    2016-04-01

    Radiative divertor operation in JET high confinement mode plasmas with the ITER-like wall has been experimentally investigated and simulated with EDGE2D-EIRENE in horizontal and vertical low field side (LFS) divertor configurations. The simulations show that the LFS divertor heat fluxes are reduced with N2-injection in similar fashion in both configurations, qualitatively consistent with experimental observations. The simulations show no substantial difference between the two configurations in the reduction of the peak LFS heat flux as a function of divertor radiation, nitrogen concentration, or pedestal Zeff. Consistently, experiments show similar divertor radiation and nitrogen injection levels for similar LFS peak heat flux reduction in both configurations. Nevertheless, the LFS strike point is predicted to detach at 20% lower separatrix density in the vertical than in the horizontal configuration. However, since the peak LFS heat flux in partial detachment in the vertical configurations is shifted towards the far scrape-off layer (SOL), the simulations predict no benefit in the reduction of LFS peak heat flux for a given upstream density in the vertical configuration relative to a horizontal one. A factor of 2 reduction of deuterium ionization source inside the separatrix is observed in the simulations when changing to the vertical configuration. The simulations capture the experimentally observed particle and heat flux reduction at the LFS divertor plate in both configurations, when adjusting the impurity injection rate to reproduce the measured divertor radiation. However, the divertor D α -emissions are underestimated by a factor of 2-5, indicating a short-fall in radiation by the fuel species. In the vertical configuration, detachment is experimentally measured and predicted to start next to the strike point, extending towards the far SOL with increasing degree of detachment. In contrast, in the horizontal configuration, the entire divertor particle flux

  18. Designing divertor targets for uniform power load

    Science.gov (United States)

    Dekeyser, W.; Reiter, D.; Baelmans, M.

    2015-08-01

    Divertor design for next step fusion reactors heavily relies on 2D edge plasma modeling with codes as e.g. B2-EIRENE. While these codes are typically used in a design-by-analysis approach, in previous work we have shown that divertor design can alternatively be posed as a mathematical optimization problem, and solved very efficiently using adjoint methods adapted from computational aerodynamics. This approach has been applied successfully to divertor target shape design for more uniform power load. In this paper, the concept is further extended to include all contributions to the target power load, with particular focus on radiation. In a simplified test problem, we show the potential benefits of fully including the radiation load in the design cycle as compared to only assessing this load in a post-processing step.

  19. The Influence of Filaments in the Private Flux Region on Divertor Particle and Power Deposition

    CERN Document Server

    Harrison, J R; Thornton, A J; Walkden, N R

    2015-01-01

    The transport of particles via intermittent filamentary structures in the private flux region of plasmas in the MAST tokamak has been investigated using a fast framing camera recording visible light emission from the volume of the lower divertor, as well as Langmuir probes and IR thermography monitoring particle and power fluxes to plasma-facing surfaces in the divertor. The visible camera data suggests that, in the divertor volume, fluctuations in light emission above the X-point are strongest in the scrape-off layer (SOL). Conversely, in the region below the X-point, it is found that these fluctuations are strongest in the private flux region (PFR) of the inner divertor leg. Detailed analysis of the appearance of these filaments in the camera data suggests that they are approximately circular, around 1-2cm in diameter. The most probable toroidal mode number is between 2 and 3. These filaments eject plasma deeper into the private flux region, sometimes by the production of secondary filaments, moving at a sp...

  20. 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

    Divertor asymmetry and its dependence on the ion del B direction has been investigated in the Experimental Advanced Superconducting Tokamak by changing the divertor configuration from lower single null (LSN), via double null (DN), to upper single null (USN) during one single discharge. Divertor p...

  1. Characteristics of divertor detachment for ITER conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kukushkin, A.S., E-mail: andre.kukushkin@iter.org [ITER Organization, St Paul Lez Durance (France); Pacher, H.D. [INRS-EMT, Varennes, Québec (Canada); Pitts, R.A. [ITER Organization, St Paul Lez Durance (France)

    2015-08-15

    The relative role of particle balance vs. momentum balance in the phenomenon of divertor plasma detachment in tokamaks is re-assessed. Ion removal from the plasma flow by volumetric recombination and/or cross-field transport is identified as the key element in the formation of the rollover of the ion saturation current on the targets, whereas “momentum removal” (friction) is responsible for maintaining high plasma pressure upstream. The deterioration of neutral particle confinement in the divertor as particle throughput increases is the primary cause of the solution collapse typically seen when deep detachment is modelled for present day experiments.

  2. 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.

  3. Small angle slot divertor concept for long pulse advanced tokamaks

    Science.gov (United States)

    Guo, H. Y.; Sang, C. F.; Stangeby, P. C.; Lao, L. L.; Taylor, T. S.; Thomas, D. M.

    2017-04-01

    SOLPS-EIRENE edge code analysis shows that a gas-tight slot divertor geometry with a small-angle (glancing-incidence) target, named the small angle slot (SAS) divertor, can achieve cold, dissipative/detached divertor conditions at relatively low values of plasma density at the outside midplane separatrix. SAS exhibits the following key features: (1) strong enhancement of the buildup of neutral density in a localized region near the plasma strike point on the divertor target; (2) spreading of the cooling front across the divertor target with the slot gradually flaring out from the strike point, thus effectively reducing both heat flux and erosion on the entire divertor target surface. Such a divertor may potentially provide a power and particle handling solution for long pulse advanced tokamaks.

  4. Exponential triples

    CERN Document Server

    Sisto, Alessandro

    2011-01-01

    Using ultrafilter techniques we show that in any partition of $\\mathbb{N}$ into 2 cells there is one cell containing infinitely many exponential triples, i.e. triples of the kind $a,b,a^b$ (with $a,b>1$). Also, we will show that any multiplicative $IP^*$ set is an "exponential $IP$ set", the analogue of an $IP$ set with respect to exponentiation.

  5. The snowflake divertor, physics of a new concept for power exhaust of fusion plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Lunt, Tilmann; Feng, Yuehe [Max-Planck-Institut fuer Plasmaphysik, Garching/Greifswald (Germany); Canal, Gustavo; Reimerdes, Holger [Ecole Polytechnique Federale de Lausanne, Lausanne (Switzerland)

    2014-07-01

    Fusion reactors based on the tokamak design will have to deal with very high heat loads on the divertor plates. One of the approaches to solve this heat load problem is the so called 'snowflake divertor', a magnetic configuration with two nearby x-points and two additional divertor legs. In this contribution we report on 'EMC3-Eirene' simulations of the plasma- and neutral particle transport in the scrape-off layer of the swiss tokamak TCV of a series of snowflake equilibria with different values of σ, the distance between the x-points normalized to the minor radius of the plasma. The constant anomalous transport coefficients were chosen such that the power- and particle deposition profiles at the primary inner strike point match the Langmuir probe measurements for the σ=0.1 case. At one of the secondary strike points, however, a significantly larger power flux than that predicted by the simulation was measured by the probes, indicating the presence of an enhanced transport across the primary separatrix. We discuss the possible reason for this enhanced transport as well as its scaling with machine size. Another prediction from the simulation is that the density as well as the radiation maximum are moving from the recycling region in front of the plates upwards to the x-point.

  6. In-out divertor flow asymmetries during ELMs in ASDEX Upgrade H-mode plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Tsalas, M. [NCSR ' Demokritos' , Institute of Nuclear Technology - Radiation Protection, 153 10 Aghia Paraskevi, Attica (Greece) and EDFA-JET CSU, Abingdon OX14 3DB, Oxon (United Kingdom)]. E-mail: maximos@ipta.demokritos.gr; Coster, D. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Garching (Germany); Fuchs, C. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Garching (Germany); Herrmann, A. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Garching (Germany); Kallenbach, A. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Garching (Germany); Mueller, H.W. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Garching (Germany); Neuhauser, J. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Garching (Germany); Rohde, V. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Garching (Germany); Tsois, N. [NCSR ' Demokritos' , Institute of Nuclear Technology - Radiation Protection, 153 10 Aghia Paraskevi, Attica (Greece)

    2007-06-15

    In ASDEX Upgrade, a fast reciprocating probe positioned in the lower divertor, capable of accessing the low-field side (LFS) and high-field side (HFS) scrape-off layer (SOL) just below the x-point, as well as the private flux region, was equipped with a Mach head and used to investigate fast flow fluctuations and in-out divertor flow asymmetries during ELMs. We compare the flow behaviour during ELMs in the three separate regions. Flow enhancement is observed in the HFS SOL, with Mach number values reaching or exceeding M = 2, flow reversal in the LFS SOL, and complex fluctuating behaviour in the private flux region (which includes flow reversal). We discuss the possible mechanisms that could drive these observations.

  7. Development of the NSTX-U Advanced Divertor Control

    Science.gov (United States)

    Vail, Patrick; Kolemen, Egemen

    2016-10-01

    Advanced magnetic divertor configurations such as the snowflake (SF) divertor are being investigated at NSTX-U for reducing the peak heat flux onto plasma-facing components. Initial efforts include development of plasma scenarios incorporating SF configurations using an upgraded set of divertor coils as well as implementation of a feedback control system for real-time detection and manipulation of two closely-spaced magnetic null points. Closed-loop plasma simulations are performed to demonstrate precise control of various SF configurations. The simulations are then used to demonstrate that the controller can be enhanced to regulate additional parameters such as strike point location and divertor flux expansion. The advanced divertor control will be used in the coming years to enable experiments investigating the physics of advanced divertors at NSTX-U. Supported by the US DOE under DE-AC02-09CH11466.

  8. ITER tungsten divertor design development and qualification program

    Energy Technology Data Exchange (ETDEWEB)

    Hirai, T., E-mail: takeshi.hirai@iter.org [ITER Organization, Route de Vinon sur Verdon, F-13115 Saint Paul lez Durance (France); Escourbiac, F.; Carpentier-Chouchana, S.; Fedosov, A.; Ferrand, L.; Jokinen, T.; Komarov, V.; Kukushkin, A.; Merola, M.; Mitteau, R.; Pitts, R.A.; Shu, W.; Sugihara, M. [ITER Organization, Route de Vinon sur Verdon, F-13115 Saint Paul lez Durance (France); Riccardi, B. [F4E, c/ Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Suzuki, S. [JAEA, Fusion Research and Development Directorate JAEA, 801-1 Mukouyama, Naka, Ibaragi 311-0193 (Japan); Villari, R. [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, I-00044 Frascati, Rome (Italy)

    2013-10-15

    Highlights: • Detailed design development plan for the ITER tungsten divertor. • Latest status of the ITER tungsten divertor design. • Brief overview of qualification program for the ITER tungsten divertor and status of R and D activity. -- Abstract: In November 2011, the ITER Council has endorsed the recommendation that a period of up to 2 years be set to develop a full-tungsten divertor design and accelerate technology qualification in view of a possible decision to start operation with a divertor having a full-tungsten plasma-facing surface. To ensure a solid foundation for such a decision, a full tungsten divertor design, together with a demonstration of the necessary high performance tungsten monoblock technology should be completed within the required timescale. The status of both the design and technology R and D activity is summarized in this paper.

  9. Examining Innovative Divertor and Main Chamber Options for a National Divertor Test Tokamak

    Science.gov (United States)

    Labombard, B.; Umansky, M.; Brunner, D.; Kuang, A. Q.; Marmar, E.; Wallace, G.; Whyte, D.; Wukitch, S.

    2016-10-01

    The US fusion community has identified a compelling need for a National Divertor Test Tokamak. The 2015 Community Planning Workshop on PMI called for a national working group to develop options. Important elements of a NDTT, adopted from the ADX concept, include the ability to explore long-leg divertor `solutions for power exhaust and particle control' (Priority Research Direction B) and to employ inside-launch RF actuators combined with double-null topologies as `plasma solution for main chamber wall components, including tools for controllable sustained operation' (PRD-C). Here we examine new information on these ideas. The projected performance of super-X and X-point target long-leg divertors is looking very promising; a stable fully-detached divertor condition handling an order-of-magnitude increase in power handling over conventional divertors may be possible. New experiments on Alcator C-Mod are addressing issues of high-field side versus low-field side heat flux sharing in double-null topologies and the screening of impurities that might originate from RF actuators placed in the high-field side - both with favorable results. Supported by USDoE Awards DE-FC02-99ER54512 and DE-AC52-07NA27344.

  10. Neutral recirculation—the key to control of divertor operation

    Science.gov (United States)

    Kukushkin, A. S.; Pacher, H. D.

    2016-12-01

    Interaction of the plasma with neutral gas in the divertor affects virtually all aspects of divertor functionality (power loading of the targets, pumping and fuelling, sustaining the operational conditions of the core plasma). In the course of ITER design development, this interaction has been the subject of intense modelling analysis, supported by experiments on various tokamaks. Neutral gas puffing is found to be the most effective means of divertor control. The results of those studies are summarized and assessed in the paper.

  11. 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.

  12. Module of lithium divertor for KTM tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lyublinski, I., E-mail: yublinski@yandex.ru [FSUE ' Red Star' , Moscow (Russian Federation); Vertkov, A.; Evtikhin, V.; Balakirev, V.; Ionov, D.; Zharkov, M. [FSUE ' Red Star' , Moscow (Russian Federation); Tazhibayeva, I. [IAE NNC RK, Kurchatov (Kazakhstan); Mirnov, S. [TRINITI, Troitsk, Moscow Region (Russian Federation); Khomiakov, S.; Mitin, D. [OJSC Dollezhal Institute, Moscow (Russian Federation); Mazzitelli, G. [ENEA RC Frascati (Italy); Agostini, P. [ENEA RC Brasimone (Italy)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer Problems of PFE degradation, tritium accumulation and plasma pollution can be overcome by the use of liquid lithium-metal with low Z. Black-Right-Pointing-Pointer Capillary-porous system (CPS) - new material in which liquid lithium fill a solid matrix from porous material. Black-Right-Pointing-Pointer Lithium divertor module for KTM tokamak is under development. Black-Right-Pointing-Pointer Lithium filled tungsten felt is offered as the base plasma facing material of divertor. Black-Right-Pointing-Pointer Results of this project addresses to the progress in the field of fusion neutrons source and fusion energy source creation. - Abstract: Activity on projects of ITER and DEMO reactors has shown that solution of problems of divertor target plates and other plasma facing elements (PFEs) based on the solid plasma facing materials cause serious difficulties. Problems of PFE degradation, tritium accumulation and plasma pollution can be overcome by the use of liquid lithium-metal with low Z. Application of lithium will allow to create a self-renewal and MHD stable liquid metal surface of the in-vessel devices possessing practically unlimited service life; to reduce power flux due to intensive re-irradiation on lithium atoms in plasma periphery that will essentially facilitate a problem of heat removal from PFE; to reduce Z{sub eff} of plasma to minimally possible level close to 1; to exclude tritium accumulation, that is provided with absence of dust products and an opportunity of the active control of the tritium contents in liquid lithium. Realization of these advantages is based on use of so-called lithium capillary-porous system (CPS) - new material in which liquid lithium fill a solid matrix from porous material. The progress in development of lithium technology and also activity in lithium experiments in the tokamaks TFTR, T-11M, T-10, FTU, NSTX, HT-7 and stellarator TJ II permits of solving the problems in development of

  13. 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.

  14. Divertor E X B Plasma Convection in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Boedo, J.A.; Schaffer, M.J.; Maingi, M.; Lasnier, C.J.; Watkins, J.G.

    1999-07-01

    Extensive two-dimensional measurements of plasma potential in the DIII-D tokamak divertor region are reported for standard (ion VB{sub T} drift toward divertor X-point) and reversed B{sub T} directions; for low (L) and high (H) confinement modes; and for partially detached divertor mode. The data are consistent with recent computational modeling identifying E x B{sub T} circulation, due to potentials sustained by plasma gradients, as the main cause of divertor plasma sensitivity to B{sub T} direction.

  15. Magnetic geometry and particle source drive of supersonic divertor regimes

    Science.gov (United States)

    Bufferand, H.; Ciraolo, G.; Dif-Pradalier, G.; Ghendrih, P.; Tamain, Ph; Marandet, Y.; Serre, E.

    2014-12-01

    We present a comprehensive picture of the mechanisms driving the transition from subsonic to supersonic flows in tokamak plasmas. We demonstrate that supersonic parallel flows into the divertor volume are ubiquitous at low density and governed by the divertor magnetic geometry. As the density is increased, subsonic divertor plasmas are recovered. On detachment, we show the change in particle source can also drive the transition to a supersonic regime. The comprehensive theoretical analysis is completed by simulations in ITER geometry. Such results are essential in assessing the divertor performance and when interpreting measurements and experimental evidence.

  16. ADX - Advanced Divertor and RF Tokamak Experiment

    Science.gov (United States)

    Greenwald, Martin; Labombard, Brian; Bonoli, Paul; Irby, Jim; Terry, Jim; Wallace, Greg; Vieira, Rui; Whyte, Dennis; Wolfe, Steve; Wukitch, Steve; Marmar, Earl

    2015-11-01

    The Advanced Divertor and RF Tokamak Experiment (ADX) is a design concept for a compact high-field tokamak that would address boundary plasma and plasma-material interaction physics challenges whose solution is critical for the viability of magnetic fusion energy. This device would have two crucial missions. First, it would serve as a Divertor Test Tokamak, developing divertor geometries, materials and operational scenarios that could meet the stringent requirements imposed in a fusion power plant. By operating at high field, ADX would address this problem at a level of power loading and other plasma conditions that are essentially identical to those expected in a future reactor. Secondly, ADX would investigate the physics and engineering of high-field-side launch of RF waves for current drive and heating. Efficient current drive is an essential element for achieving steady-state in a practical, power producing fusion device and high-field launch offers the prospect of higher efficiency, better control of the current profile and survivability of the launching structures. ADX would carry out this research in integrated scenarios that simultaneously demonstrate the required boundary regimes consistent with efficient current drive and core performance.

  17. Resilient non-resonant divertors for stellarators

    Science.gov (United States)

    Bader, A.; Boozer, A. H.; Hegna, C. C.; Lazerson, S. A.

    2016-10-01

    In this work, we investigate whether resilient non-resonant divertor solutions exist for optimized stellarators. Resiliency is measured by the consistency of performance over a broad range of operational states, such as through bootstrap current and modified plasma pressures. A non-resonant configuration is one where the crucial topological feature is the existence and sharpness of ridges along the last closed flux surface. We develop a modified field-line following method for testing the resiliency of stellarator divertors and apply it to altered HSX configurations generated by varying external coil currents, wall positioning, and internal plasma currents. We compare a magnetic diffusion calculation with a ``zero-diffusion'' calculation that endeavors to measure the first escaping flux tubes. The results from these calculations are corroborated with a more complete edge simulation with EMC3-EIRENE. The EMC3-EIRENE simulations show resilient helical stripes that are consistent with the simpler field line following methods. The goal of the study is to find a metric for edge/divertor optimization of stellarators, a crucial piece that is missing from current optimization schemes. Work supported by DE-SC0006103 and DE-FG02-93ER54222,.

  18. 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

  19. Initial operation of the divertor Thompson scattering diagnostic on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Carlstrom, T.N.; Hsieh, C.L.; Stockdale, R.E. [and others

    1996-05-01

    The first Thomson scattering measurements of n{sub e} and T{sub e} in the divertor region of a tokamak are reported. These data are used as input to boundary physics codes such as UEDGE and DEGAS and to benchmark the predictive capabilities of these codes. These measurements have also contributed to the characterization of tokamak disruptions. A Nd:YAG laser (20 Hz, 1 J, 15 ns, 1064 nm) is directed vertically through the lower divertor region of the DIII-D tokamak. A custom, aspherical collection lens (f /6.8) images the laser beam from 1-21 cm above the target plates into eight spatial channels with 1.5 cm vertical and 0.3 cm radial resolution. 2D mapping of the divertor region is achieved by sweeping the divertor X-point location radially through the fixed laser beam location. Fiber optics carry the light to polychromators whose interference filters have been optimized for low T{sub e} measurements. Silicon avalanche photo diodes measure both the scattered and plasma background light. Temperatures and densities are typically in the range of 5-200 eV and 1 - 10 x 10{sup 19} m{sup -3} respectively. Low temperatures, T{sub e} < 1 eV, and high densities, n{sub e} > 8x10{sup 20} m{sup -3} have been observed in detached plasmas. Background light levels have not been a significant problem. Reduction of the laser stray light permits Rayleigh calibration. Because of access difficulties, no in-vessel vacuum alignment target could be used. Instead, an in situ laser alignment monitor provides alignment information for each laser pulse. Results are compared with Langmuir probe measurements where good agreement is found except for regions of high n{sub e} and low T{sub e} as measured by Thomson scattering.

  20. Power distribution in the snowflake divertor in TCV

    NARCIS (Netherlands)

    Reimerdes, H.; G.P. Canal,; Duval, B. P.; Labit, B.; Lunt, T.; Vijvers, W. A. J.; Coda, S.; De Temmerman, G.; Morgan, T. W.; Nespoli, F.; Tal, B.; the TCV Team,

    2013-01-01

    TCV experiments demonstrate the basic power exhaust properties of the snowflake (SF) plus and SF minus divertor configurations by measuring the heat fluxes at each of their four divertor legs. The measurements indicate an enhanced transport into the private flux region and a reduction of peak heat f

  1. Divertor plasma physics experiments on the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Mahdavi, M.A.; Allen, S.L.; Evans, T.E. [and others

    1996-10-01

    In this paper we present an overview of the results and conclusions of our most recent divertor physics and development work. Using an array of new divertor diagnostics we have measured the plasma parameters over the entire divertor volume and gained new insights into several divertor physics issues. We present direct experimental evidence for momentum loss along the field lines, large heat convection, and copious volume recombination during detachment. These observations are supported by improved UEDGE modeling incorporating impurity radiation. We have demonstrated divertor exhaust enrichment of neon and argon by action of a forced scrape off layer (SOL) flow and demonstrated divertor pumping as a substitute for conventional wall conditioning. We have observed a divertor radiation zone with a parallel extent that is an order of magnitude larger than that estimated from a 1-D conduction limited model of plasma at coronal equilibrium. Using density profile control by divertor pumping and pellet injection we have attained H-mode confinement at densities above the Greenwald limit. Erosion rates of several candidate ITER plasma facing materials are measured and compared with predictions of a numerical model.

  2. Comparison of ELM heat loads in snowflake and standard divertors

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-05-08

    An analysis is given of the impact of the tokamak divertor magnetic structure on the temporal and spatial divertor heat flux from edge localized modes (ELMs). Two configurations are studied: the standard divertor where the poloidal magnetic field (B{sub p}) varies linearly with distance (r) from the magnetic null and the snowflake where B{sub p} varies quadratrically with r. Both one and two-dimensional models are used to analyze the effect of the longer magnetic field length between the midplane and the divertor plate for the snowflake that causes a temporal dilation of the ELM divertor heat flux. A second effect discussed is the appearance of a broad region near the null point where the poloidal plasma beta can substantially exceed unity, especially for the snowflake configuration during the ELM; such a condition is likely to drive additional radial ELM transport.

  3. RELAP5 MODEL OF THE DIVERTOR PRIMARY HEAT TRANSFER SYSTEM

    Energy Technology Data Exchange (ETDEWEB)

    Popov, Emilian L [ORNL; Yoder Jr, Graydon L [ORNL; Kim, Seokho H [ORNL

    2010-08-01

    This report describes the RELAP5 model that has been developed for the divertor primary heat transfer system (PHTS). The model is intended to be used to examine the transient performance of the divertor PHTS and evaluate control schemes necessary to maintain parameters within acceptable limits during transients. Some preliminary results are presented to show the maturity of the model and examine general divertor PHTS transient behavior. The model can be used as a starting point for developing transient modeling capability, including control system modeling, safety evaluations, etc., and is not intended to represent the final divertor PHTS design. Preliminary calculations using the models indicate that during normal pulsed operation, present pressurizer controls may not be sufficient to keep system pressures within their desired range. Additional divertor PHTS and control system design efforts may be required to ensure system pressure fluctuation during normal operation remains within specified limits.

  4. In situ spectral calibration method for the impurity influx monitor (divertor) for ITER using angled physical contact fibers.

    Science.gov (United States)

    Iwamae, A; Ogawa, H; Sugie, T; Kusama, Y

    2011-03-01

    The in situ calibration method for the impurity influx monitor (divertor) is experimentally examined. The total reflectance of the optical path from the focal point of the Cassegrain telescope to the first mirror is derived using a micro retroreflector array. An optical fiber with angled physical contact (APC) connectors reduces the return edge reflection. APC fibers and a multimode coupler increase the signal-to-noise ratio by about one order compared to that of triple-branched fibers and enable measurement of the wavelength dependence of the total reflectance of the optical system even after potential deterioration of mirror surfaces reduces reflectance.

  5. Researches on the Neutral Gas Pressure in the Divertor Chamber of the HL-2A Tokamak

    Institute of Scientific and Technical Information of China (English)

    WANGMingxu; LIBo; YANGZhigang; YANLongwen; HONGWenyu; YUANBaoshan; LIULi; CAOZeng; CUIChenghe; LIUYong; WANGEnyao; ZHANGNianman

    2003-01-01

    The neutral gas pressure in divertor chamber is a very basic and important physics parameter because it determines the temperature of charged particles, the thermal flux density onto divertor plates, the erosion of divertor plates, impurity retaining and exhausting, particle transportation and confinement performance of plasma in tokamaks. Therefore, the pressure measurement in divertor chamber is taken into account in many large tokamaks.

  6. A triple-emission fluorescent probe reveals distinctive amyloid fibrillar polymorphism of wild-type alpha-synuclein and its familial Parkinson's disease mutants.

    Science.gov (United States)

    Celej, M Soledad; Caarls, Wouter; Demchenko, Alexander P; Jovin, Thomas M

    2009-08-11

    Intracytoplasmic neuronal deposits containing amyloid fibrils of the 140-amino acid presynaptic protein alpha-synuclein (AS) are the hallmark of Parkinson's (PD) disease and related neurodegenerative disorders. Three point mutations (A53T, A30P, and E46K) are linked to early onset PD. Compared to the wild-type (WT) protein, the mutants aggregate faster in vitro, but their fibrillar products are quite similar. Using the extrinsic multiple-emission probe 4'-(diethylamino)-3-hydroxyflavone (FE), we demonstrate unique and distinct spectroscopic signatures for the amyloid fibrils formed by the WT and mutant AS, presumably indicative of subtle differences in supramolecular structure. The two well-separated emission bands of the FE probe originate from a proton transfer reaction in the excited state. The ratiometric response constitutes a sensitive, tunable reporter of microenvironmental properties such as polarity and hydrogen bonding. The very distinctive fluorescence spectra of the FE probe bound to the four AS variants reflect different tautomeric equilibria in the excited state and the existence of at least two different binding sites in the fibrils for the dye. Deconvolution of the two-dimensional excitation-emission spectra leads to estimations of different local dielectric constants and extents of hydration characteristic of the proteins. The sensitivity of such a simple external probe to conformational alterations induced by point mutations is unprecedented and provides new insight into key phenomena related to amyloid fibrils: plasticity, polymorphism, propagation of structural features, and structure-function relationships underlying toxicity.

  7. 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)

  8. 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.

  9. Numerical studies on divertor plasmas in helical systems

    Energy Technology Data Exchange (ETDEWEB)

    Ueda, Noriaki (Mitsubishi Atomic Power Industries, Inc., Tokyo (Japan)); Itoh, Kimitaka; Itoh, Sanae

    1989-12-01

    Scrape-off layer and divertor plasmas in helical systems are studied by using the two-dimensional (2D) numerical simulation code. Unified edge divertor analysis code (UEDA code) is applied to the straight helical model of torsatron/helical heliotron configurations. 2D profiles of plasma parameter, neutrals and impurities are obtained. Erosion rate and neutral back flow rate to the core plasma are also evaluated. Various shapes of the buffle plate are examined from the view point of the establishment of 'dense-cold divertor plasma' by which we can avoid the damage of the target plate. (author).

  10. A review of ELMs in divertor tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Hill, D.N.

    1996-05-23

    This paper reviews what is known about edge localized modes (ELMs), with an emphasis on their effect on the scrape-off layer and divertor plasmas. ELM effects have been measured in the ASDEX-U, C-Mod, COMPASS-D, DIII-D, JET, JFT-2M,JT-60U, and TCV tokamaks and are reported here. At least three types of ELMs have been identified and their salient features determined. Type-1 giant ELMs can cause the sudden loss of up to 10-15% of the plasma stored energy but their amplitude ({Delta}W/W) does not increase with increasing power. Type- 3 ELMs are observed near the H-mode power threshold and produce small energy dumps (1-3% of the stored energy). All ELMs increase the scrape- off layer plasma and produce particle fluxes on the divertor targets which are as much as ten times larger that the quiescent phase between ELMs. The divertor heat pulse is largest on the inner target, unlike that of L-Mode or quiescent H-mode; some tokamaks report radial structure in the heat flux profile which is suggestive of islands or helical structures. The power scaling of Type-1 ELM amplitude and frequency have been measured in several tokamaks and has recently been applied to predictions of the ELM Size in ITER. Concern over the expected ELM amplitude has led to a number of experiments aimed at demonstrating active control of ELMs. Impurity gas injection with feedback control on the radiation loss in ASDEX-U suggests that a promising mode of operation (the CDH-mode) with a very small type-3 ELMs can be maintained with heating power sell above the H-mode threshold, where giant type-1 ELMs can be maintained with heating power well above the H-mode threshold, where Giant type-1 ELMs are normally observed. While ELMs have many potential negative effects, the beneficial effect of ELMs in providing density control and limiting the core plasma impurity content in high confinement H- mode discharges should not be overlooked.

  11. Compatibility of detached divertor operation with robust edge pedestal performance

    Energy Technology Data Exchange (ETDEWEB)

    Leonard, A.W., E-mail: leonard@fusion.gat.com [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Makowski, M.A.; McLean, A.G. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Osborne, T.H.; Snyder, P.B. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States)

    2015-08-15

    The compatibility of detached radiative divertor operation with a robust H-mode pedestal is examined in DIII-D. A density scan produced low temperature plasmas at the divertor target, T{sub e} ⩽ 2 eV, with high radiation leading to a factor of ⩾4 drop in peak divertor heat flux. The cold radiative plasma was confined to the divertor and did not extend across the separatrix in X-point region. A robust H-mode pedestal was maintained with a small degradation in pedestal pressure at the highest densities. The response of the pedestal pressure to increasing density is reproduced by the EPED pedestal model. However, agreement of the EPED model with experiment at high density requires an assumption of reduced diamagnetic stabilization of edge Peeling–Ballooning modes.

  12. Thermal Fatigue Study on the Divertor Plate Materials

    Institute of Scientific and Technical Information of China (English)

    吴继红; 张斧; 许增裕; 严建成

    2002-01-01

    Thermal fatigue property of the divertor plate is one of the key issues that governs the lifetime of the divertor plate. Taking tungsten as surface material, a small-mock-up divertor plate was made by hot isostatic press welding (HIP). A thermal cycling experiment for divertor mock-up was carried out in the vacuum, where a high-heat-flux electronic gun was used as the thermal source. A cyclic heat flux of 9 MW/m2 was loaded onto the mock-up, a heating duration of 20 s was selected, the cooling water flow rate was 80 ml/s. After 1000 cycles, the surface and the W/Cu joint of the mock-up did not show any damage. The SEM was used to analyze the microstructure of the welding joint, where no cracks were found also.

  13. Numerical analysis of divertor plasma for demo-CREST

    Energy Technology Data Exchange (ETDEWEB)

    Ishida, M.; Maeki, K.; Hatayama, A. [Graduate School of Fundamental Science and Technology, Keio University, Yokohama (Japan); Hiwatari, R. [Central Research Institute of Electric Power Industry (CRIEPI), Tokyo (Japan); Bonnin, X. [LIMHP-CNRS, Universite Paris 13, Villetaneuse (France); Zhu, S. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Schneider, R. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Greifswald (Germany); Coster, D. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Garching (Germany)

    2010-05-15

    The numerical analysis of the demonstration fusion reactor Demo-CREST has been carried out; this analysis focuses on impurity seeding. Several design activities for DEMO have been carried out; however, its detailed divertor plasma analysis remains to be carried out. Therefore, in this study, we discuss the possibility of neon puffing in demo-CREST to decrease the power load to the divertor plate by using the B2-EIRENE code. It has been shown that the radiation power loss by neon increases with upstream plasma density and that the peak power load to the divertor plate comes close to the allowable level by using the preliminary divertor configuration (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  14. Plasma detachment with molecular processes in divertor plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Ohno, N.; Ezumi, N.; Nishijima, D.; Takamura, S. [Dept. of Energy Engineering and Science, Graduate School of Engineering, Nagoya Univ., Nagoya, Aichi (Japan); Krasheninnikov, S.I.; Pigarov, A.Yu. [MIT Plasma Science and Fusion Center, Cambridge, MA (United States)

    2000-01-01

    Molecular processes in detached recombining plasmas are briefly reviewed. Several reactions with vibrationally excited hydrogen molecule related to recombination processes are described. Experimental evidence of molecular activated recombination observed in a linear divertor plasma simulator is also shown. (author)

  15. Divertor IR thermography on Alcator C-Mod.

    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.

  16. 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.

  17. 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.

  18. Beryllium flux distribution and layer deposition in the ITER divertor

    Science.gov (United States)

    Schmid, K.

    2008-10-01

    The deposition of Be eroded from the main chamber wall on the W surfaces in the ITER divertor could result in the formation of Be rich Be/W mixed layers with a low melting temperature compared with pure W. To predict whether or not these layers form the Be flux distribution in the ITER divertor is required. This paper presents the results of a combination of plasma transport with erosion/deposition simulations that allow one to calculate both the Be flux distribution and the Be layer deposition in the ITER divertor. This model includes the Be source due to Be erosion in the main chamber and the deposition, re-erosion and re-deposition of Be in the ITER divertor. The calculations show that the fraction of Be in the incident particle flux in the divertor ranges from ≈10-3 to ≈5% with a pronounced inner-outer divertor asymmetry. The flux fractions in the inner divertor are on average ten times higher than in the outer divertor. Thick Be layers only form at the inner strike point and the dome baffles. The highest Be layer growth rate is found to be 1.0 nm s-1. Despite the Be deposition the formation of Be rich Be/W mixed layers is not to be expected in ITER. The expected surface temperature at these locations during steady-state operation is too low as to result in Be diffusion into W and thus Be/W mixed layers cannot form. The paper also discusses the influence of off normal events such as ELMs or VDEs on the formation of Be/W mixed layers.

  19. Characterizing the Outer Divertor Leg Transition to Full Detachment

    Science.gov (United States)

    McLean, A. G.; Allen, S. L.; Fenstermacher, M. E.; Lasnier, C. J.; Meyer, W. H.; Porter, G. D.; Soukhanovskii, V. A.; Bray, B. D.; Carlstrom, T. N.; Leonard, A. W.; Liu, C.; Eldon, D.; Groth, M.; Stangeby, P. C.; Tsui, C. K.

    2013-10-01

    Experiments at DIII-D have explored the transition from an attached to fully detached divertor condition in L- and H-mode with an unprecedented level of detail. Improved divertor Thomson scattering capturing Te operation in future devices. This work supported in part by the US Department of Energy under DE-AC52-07NA27344 and DE-FC02-04ER54698.

  20. 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.

  1. X-Divertors on ITER - with no hardware changes

    Science.gov (United States)

    Valanju, Prashant; Covele, Brent; Kotschenreuther, Mike; Mahajan, Swadesh; Kessel, Charles

    2014-10-01

    Using CORSICA, we have discovered that X-Divertor (XD) equilibria are possible on ITER - without any extra PF coils inside the TF coils, and with no changes to ITER's poloidal field (PF) coil set, divertor cassette, strike points, or first wall. Starting from the Standard Divertor (SD), a sequence of XD configurations (with increasing flux expansions at the divertor plate) can be made by reprogramming ITER PF coil currents while keeping them all under their design limits (Lackner and Zohm have shown this to be impossible for Snowflakes). The strike point is held fixed, so no changes in the divertor or pumping hardware will be needed. The main plasma shape is kept very close to the SD case, so no hardware changes to the main chamber will be needed. Time-dependent ITER-XD operational scenarios are being checked using TSC. This opens the possibility that many XDs could be tested and used to assist in high-power operation on ITER. Because of the toroidally segmented ITER divertor plates, strongly detached operation may be critical for making use of the largest XD flux expansion possible. The flux flaring in XDs is expected to increase the stability of detachment, so that H-mode confinement is not affected. Detachment stability is being examined with SOLPS. This work supported by US DOE Grants DE-FG02-04ER54742 and DE-FG02-04ER54754 and by TACC at UT Austin.

  2. 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 ...

  3. Probing the Repulsive Core of the Nucleon-Nucleon Interaction via the He4(e ,e'pN) Triple-Coincidence Reaction

    Science.gov (United States)

    Korover, I.; Muangma, N.; Hen, O.; Shneor, R.; Sulkosky, V.; Kelleher, A.; Gilad, S.; Higinbotham, D. W.; Piasetzky, E.; Watson, J. W.; Wood, S. A.; Aguilera, P.; Ahmed, Z.; Albataineh, H.; Allada, K.; Anderson, B.; Anez, D.; Aniol, K.; Annand, J.; Armstrong, W.; Arrington, J.; Averett, T.; Badman, T.; Baghdasaryan, H.; Bai, X.; Beck, A.; Beck, S.; Bellini, V.; Benmokhtar, F.; Bertozzi, W.; Bittner, J.; Boeglin, W.; Camsonne, A.; Chen, C.; Chen, J.-P.; Chirapatpimol, K.; Cisbani, E.; Dalton, M. M.; Daniel, A.; Day, D.; de Jager, C. W.; De Leo, R.; Deconinck, W.; Defurne, M.; Flay, D.; Fomin, N.; Friend, M.; Frullani, S.; Fuchey, E.; Garibaldi, F.; Gaskell, D.; Gilman, R.; Glamazdin, O.; Gu, C.; Gueye, P.; Hamilton, D.; Hanretty, C.; Hansen, J.-O.; Hashemi Shabestari, M.; Holmstrom, T.; Huang, M.; Iqbal, S.; Jin, G.; Kalantarians, N.; Kang, H.; Khandaker, M.; LeRose, J.; Leckey, J.; Lindgren, R.; Long, E.; Mammei, J.; Margaziotis, D. J.; Markowitz, P.; Marti Jimenez-Arguello, A.; Meekins, D.; Meziani, Z.; Michaels, R.; Mihovilovic, M.; Monaghan, P.; Munoz Camacho, C.; Norum, B.; Nuruzzaman, Pan, K.; Phillips, S.; Pomerantz, I.; Posik, M.; Punjabi, V.; Qian, X.; Qiang, Y.; Qiu, X.; Rakhman, A.; Reimer, P. E.; Riordan, S.; Ron, G.; Rondon-Aramayo, O.; Saha, A.; Schulte, E.; Selvy, L.; Shahinyan, A.; Sirca, S.; Sjoegren, J.; Slifer, K.; Solvignon, P.; Sparveris, N.; Subedi, R.; Tireman, W.; Wang, D.; Weinstein, L. B.; Wojtsekhowski, B.; Yan, W.; Yaron, I.; Ye, Z.; Zhan, X.; Zhang, J.; Zhang, Y.; Zhao, B.; Zhao, Z.; Zheng, X.; Zhu, P.; Zielinski, R.; Jefferson Lab Hall A Collaboration

    2014-07-01

    We studied simultaneously the He4(e ,e'p), He4(e ,e'pp), and He4(e ,e'pn) reactions at Q2=2(GeV/c)2 and xB>1, for an (e ,e'p) missing-momentum range of 400 to 830 MeV/c. The knocked-out proton was detected in coincidence with a proton or neutron recoiling almost back to back to the missing momentum, leaving the residual A =2 system at low excitation energy. These data were used to identify two-nucleon short-range correlated pairs and to deduce their isospin structure as a function of missing momentum, in a region where the nucleon-nucleon (NN) force is expected to change from predominantly tensor to repulsive. The abundance of neutron-proton pairs is reduced as the nucleon momentum increases beyond ˜500 MeV/c. The extracted fraction of proton-proton pairs is small and almost independent of the missing momentum. Our data are compared with calculations of two-nucleon momentum distributions in He4 and discussed in the context of probing the elusive repulsive component of the NN force.

  4. 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

  5. Spectroscopic investigations of divertor detachment in TCV

    CERN Document Server

    Verhaegh, K; Duval, B P; Harrison, J R; Reimerdes, H; Theiler, C; Labit, B; Maurizio, R; Marini, C; Nespoli, F; Sheikh, U; Tsui, C K; Vianello, N; Vijvers, W A J

    2016-01-01

    The aim of this work is to provide an understanding of detachment at TCV with emphasis on analysis of the Balmer line emission. A new Divertor Spectroscopy System has been developed for this purpose. Further development of Balmer line analysis techniques has allowed detailed information to be extracted on free-free and three-body recombination. During density ramps, the plasma at the target detaches as inferred from a drop in density at, and ion current to, the target. At the same time the Balmer $6\\rightarrow2$ and $7\\rightarrow2$ line emission near the target is dominated by recombination, indicating that the ionization region has also detached from the target to be replaced by a recombining region with densities more than a factor 2 higher than at the target. As the core density increases further, the density and recombination rate are rising all along the outer leg to the x-point while remaining highest at the target. Even at the highest core densities accessed (Greenwald fraction 0.7) the peaks in recomb...

  6. 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.

  7. Effects of 2D and 3D Error Fields on the SAS Divertor Magnetic Topology

    Science.gov (United States)

    Trevisan, G. L.; Lao, L. L.; Strait, E. J.; Guo, H. Y.; Wu, W.; Evans, T. E.

    2016-10-01

    The successful design of plasma-facing components in fusion experiments is of paramount importance in both the operation of future reactors and in the modification of operating machines. Indeed, the Small Angle Slot (SAS) divertor concept, proposed for application on the DIII-D experiment, combines a small incident angle at the plasma strike point with a progressively opening slot, so as to better control heat flux and erosion in high-performance tokamak plasmas. Uncertainty quantification of the error fields expected around the striking point provides additional useful information in both the design and the modeling phases of the new divertor, in part due to the particular geometric requirement of the striking flux surfaces. The presented work involves both 2D and 3D magnetic error field analysis on the SAS strike point carried out using the EFIT code for 2D equilibrium reconstruction, V3POST for vacuum 3D computations and the OMFIT integrated modeling framework for data analysis. An uncertainty in the magnetic probes' signals is found to propagate non-linearly as an uncertainty in the striking point and angle, which can be quantified through statistical analysis to yield robust estimates. Work supported by contracts DE-FG02-95ER54309 and DE-FC02-04ER54698.

  8. Upgraded divertor Thomson scattering system on DIII-D

    Science.gov (United States)

    Glass, F.; Carlstrom, T. N.; Du, D.; McLean, A. G.; Taussig, D. A.; Boivin, R. L.

    2016-11-01

    A design to extend the unique divertor Thomson scattering system on DIII-D to allow measurements of electron temperature and density in high triangularity plasmas is presented. Access to this region is selectable on a shot-by-shot basis by redirecting the laser beam of the existing divertor Thomson system inboard — beneath the lower floor using a moveable, high-damage threshold, in-vacuum mirror — and then redirecting again vertically. The currently measured divertor region remains available with this mirror retracted. Scattered light is collected from viewchords near the divertor floor using in-vacuum, high temperature optical elements and relayed through the port window, before being coupled into optical fiber bundles. At higher elevations from the floor, measurements are made by dynamically re-focusing the existing divertor system collection optics. Nd:YAG laser timing, analysis of the scattered light spectrum via polychromators, data acquisition, and calibration are all handled by existing systems or methods of the current multi-pulse Thomson scattering system. Existing filtered polychromators with 7 spectral channels are employed to provide maximum measurement breadth (Te in the range of 0.5 eV-2 keV, ne in the range of 5 × 1018-1 × 1021 m3) for both low Te in detachment and high Te measurement up beyond the separatrix.

  9. Initial Development of the NSTX-U Snowflake Divertor Control

    Science.gov (United States)

    Vail, Patrick; Kolemen, Egemen; Welander, Anders; Lanctot, Matthew

    2015-11-01

    A feedback control system has been implemented at NSTX-U for real-time detection and manipulation of snowflake divertor (SFD) magnetic configurations. The SFD is an alternative magnetic divertor concept that is characterized by a second-order null formed by two x-points in close proximity. The SFD is an attractive option for heat flux mitigation for NSTX-U in which unmitigated peak heat fluxes in standard divertor operation near 20 MW/m2 may compromise plasma-facing components. The real-time control system at NSTX-U is capable of simultaneous control of multiple SFD parameters, such as the separation between the two x-points in the divertor region and their orientation. Control of SFD configurations in NSTX-U has been simulated in TOKSYS using the upgraded sets of poloidal field coils in both the upper and lower divertor regions. Performance of the real-time control system and its effect on plasma performance will be assessed experimentally as an initial step toward the development of the SFD concept at NSTX-U. Supported by the US DOE under DE-AC02-09CH11466.

  10. Upgraded divertor Thomson scattering system on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Glass, F., E-mail: glassf@fusion.gat.com; Carlstrom, T. N.; Du, D.; Taussig, D. A.; Boivin, R. L. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); McLean, A. G. [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, California 94550 (United States)

    2016-11-15

    A design to extend the unique divertor Thomson scattering system on DIII-D to allow measurements of electron temperature and density in high triangularity plasmas is presented. Access to this region is selectable on a shot-by-shot basis by redirecting the laser beam of the existing divertor Thomson system inboard — beneath the lower floor using a moveable, high-damage threshold, in-vacuum mirror — and then redirecting again vertically. The currently measured divertor region remains available with this mirror retracted. Scattered light is collected from viewchords near the divertor floor using in-vacuum, high temperature optical elements and relayed through the port window, before being coupled into optical fiber bundles. At higher elevations from the floor, measurements are made by dynamically re-focusing the existing divertor system collection optics. Nd:YAG laser timing, analysis of the scattered light spectrum via polychromators, data acquisition, and calibration are all handled by existing systems or methods of the current multi-pulse Thomson scattering system. Existing filtered polychromators with 7 spectral channels are employed to provide maximum measurement breadth (T{sub e} in the range of 0.5 eV–2 keV, n{sub e} in the range of 5 × 10{sup 18}–1 × 10{sup 21} m{sup 3}) for both low T{sub e} in detachment and high T{sub e} measurement up beyond the separatrix.

  11. 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.

  12. Radiative divertor plasmas with convection in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Leornard, A.W. [General Atomics, San Diego, CA (United States); Porter, G.D.; Wood, R.D. [Lawrence Livermore National Lab., CA (United States)] [and others

    1998-01-01

    The radiation of divertor heat flux on DIII-D is shown to greatly exceed the limits imposed by assumptions of energy transport dominated by electron thermal conduction parallel to the magnetic field. Approximately 90% of the power flowing into the divertor is dissipated through low Z radiation and plasma recombination. The dissipation is made possible by an extended region of low electron temperature in the divertor. A one-dimensional analysis of the parallel heat flux finds that the electron temperature profile is incompatible with conduction dominated parallel transport. Plasma flow at up to the ion acoustic speed, produced by upstream ionization, can account for the parallel heat flux. Modeling with the two-dimensional fluid code UEDGE has reproduced many of the observed experimental features.

  13. A novel approach to magnetic divertor configuration design

    Science.gov (United States)

    Blommaert, M.; Baelmans, M.; Dekeyser, W.; Gauger, N. R.; Reiter, D.

    2015-08-01

    Divertor exhaust system design and analysis tools are crucial to evolve from experimental fusion reactors towards commercial power plants. In addition to material research and dedicated vessel geometry design, improved magnetic configurations can contribute to sustaining the diverted heat loads. Yet, computational design of the magnetic divertor is a challenging process involving a magnetic equilibrium solver, a plasma edge grid generator and a computationally demanding plasma edge simulation. In this paper, an integrated approach to efficient sensitivity calculations is discussed and applied to a set of slightly reduced divertor models. Sensitivities of target heat load performance to the shaping coil currents are directly evaluated. Using adjoint methods, the cost for a sensitivity evaluation is reduced to about two times the simulation cost of one specific configuration. Further, the use of these sensitivities in an optimal design framework is illustrated by a case with realistic Joint European Torus (JET) configurational parameters.

  14. TECXY study of a liquid lithium divertor for DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Pelka, G.; Chmielewski, P.; Zagorski, R. [Institute of Plasma Physics and Laser Microfusion, Warsaw (Poland); Pericoli-Ridolfini, V.; Viola, B. [ENEA C.R. Frascati, Roma (Italy)

    2016-08-15

    Divertor targets made out of a capillary porous system (CPS) filled with liquid lithium, have been proposed as an alternative to standard, solid state plates. In the current work we simulate the DEMO edge plasma in either a standard single-null or snowflake divertor configuration. Our tool is the 2D code TECXY. Lithium ablated from the target plate surface and released into the plasma is shown here to partially screen the incoming heat flux. Lithium's moderate SOL radiation levels suggest additional seeding to be beneficial. Very high heat fluxes to the divertor need to be avoided, as intensive lithium evaporation might unacceptably pollute the plasma. (copyright 2016 The Authors. Contributions to Plasma Physics published by Wiley-VCH Verlag GmbH and Co. KGaA, Weinheim)

  15. Turbulence during ergodic divertor experiments in Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Payan, J.; Garbet, X.; Chatenet, J.H.; Clairet, F.; De Michelis, C.; Devynck, P.; Ghendrih, P.; Gil, C.; Grosman, A.; Guirlet, R. [and others

    1994-12-01

    The level of density fluctuations is shown to decrease during ergodic divertor operation in Tore Supra. This decrease of the turbulence is correlated with the onset of a temperature pedestal and a local improvement of the confinement. This pedestal is located close to the electric shear layer, i.e., within a narrow region between the plasma core and the ergodic layer. The onset of such a pedestal explains why the central electron temperature is not changed when the ergodic divertor is switched on, in spite of an ergodic zone where the temperature is low. (author). 30 refs., 14 figs.

  16. Resonant magnetic perturbations and divertor footprints in poloidally diverted tokamaks

    CERN Document Server

    Cahyna, Pavel

    2010-01-01

    General formula describing both the divertor strike point splitting and width of magnetic islands created by resonant magnetic perturbations (RMPs) in a poloidally diverted tokamak equilibrium is derived. Under the assumption that the RMP is produced by coils at the low-field side such as those used to control edge localized modes (ELMs) it is demonstrated that the width of islands on different magnetic surfaces at the edge and the amount of divertor splitting are related to each other. Explanation is provided of aligned maxima of the perturbation spectra with the safety factor profile - an effect empirically observed in models of many perturbation coil designs.

  17. 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.

  18. Invariance of divertor retention on external particle flow in detached ASDEX upgrade discharges

    Energy Technology Data Exchange (ETDEWEB)

    Bosch, H.; Dux, R.; Haas, G.; Kallenbach, A.; Kaufmann, M.; Lackner, K.; Mertens, V.; Murmann, H.; Poschenrieder, W.; Salzmann, H.; Schweinzer, J.; Suttrop, W.; Weinlich, M. [Max-Planck-Institut fuer Plasmaphysik, IPP-EURATOM Association, D-85748 Garching (Germany); ASDEX Upgrade team% NI team

    1996-04-01

    Divertor plasmas with strong external gas puffing in ASDEX Upgrade have shown very efficient impurity retention, increasing with the divertor neutral gas density. The experiments presented here use feedback-controlled gas puffs in discharges with different pumping speed to keep the divertor neutral gas flux density the same. This allows for the first time a decoupling of the divertor neutral gas flux density and the external gas flow. The resulting plasmas are almost identical and show identical impurity retention, clearly demonstrating the importance of the divertor neutral gas density over the externally induced flow. {copyright} {ital 1996 The American Physical Society.}

  19. The contribution of radio-frequency rectification to field-aligned losses of high-harmonic fast wave power to the divertor in the National Spherical Torus eXperiment

    Energy Technology Data Exchange (ETDEWEB)

    Perkins, R. J., E-mail: rperkins@pppl.gov; Hosea, J. C.; Jaworski, M. A.; Diallo, A.; Bell, R. E.; Bertelli, N.; Gerhardt, S.; Kramer, G. J.; LeBlanc, B. P.; Phillips, C. K.; Podestà, M.; Roquemore, L.; Taylor, G.; Wilson, J. R. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Ahn, J.-W.; Gray, T. K. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); McLean, A. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Sabbagh, S. [Department of Applied Physics and Applied Mathematics, Columbia University, New York, New York 10027 (United States)

    2015-04-15

    The National Spherical Torus eXperiment (NSTX) can exhibit a major loss of high-harmonic fast wave (HHFW) power along scrape-off layer (SOL) field lines passing in front of the antenna, resulting in bright and hot spirals on both the upper and lower divertor regions. One possible mechanism for this loss is RF sheaths forming at the divertors. Here, we demonstrate that swept-voltage Langmuir probe characteristics for probes under the spiral are shifted relative to those not under the spiral in a manner consistent with RF rectification. We estimate both the magnitude of the RF voltage across the sheath and the sheath heat flux transmission coefficient in the presence of the RF field. Although precise comparison between the computed heat flux and infrared (IR) thermography cannot yet be made, the computed heat deposition compares favorably with the projections from IR camera measurements. The RF sheath losses are significant and contribute substantially to the total SOL losses of HHFW power to the divertor for the cases studied. This work will guide future experimentation on NSTX-U, where a wide-angle IR camera and a dedicated set of coaxial Langmuir probes for measuring the RF sheath voltage directly will quantify the contribution of RF sheath rectification to the heat deposition from the SOL to the divertor.

  20. 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.

  1. Overview of experiments with the dynamic ergodic divertor on TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Finken, K.H.; Abdullaev, S.; Biel, W.; Brezinsek, S.; Busch, C.; Harting, D.; Jakubowski, M.; Koslowski, H.R.; Kraemer-Flecken, A.; Kikuchi, Y.; Lehnen, M.; Liang, Y.; Nicolai, A.; Pospieszczyk, A. [Institut fuer Plasmaphysik, Forschungszentrum Juelich GmbH, EURATOM Association, Trilateral Euregio Cluster, D-52425 Juelich (Germany); Bock, M.F.M. de; Classen, I.; Hellermann, M. von; Jaspers, R. [FOM-Institute for Plasma Physics Rijnhuizen, Association EURATOM-FOM, Trilateral Euregio Cluster, P.O. Box: 1207, NL-3430 BE Nieuwegein (Netherlands); Jachmich, S. [Laboratory for Plasma Physics, Association EURATOM - Belgian State, KMS - ERM, Trilateral Euregio Cluster, B-1000 Brussels (Belgium); Kobayashi, M. [National Institute for Fusion Science, 322-6 Oroshi-cho, Toki-shi 509-52 Toki (Japan); Reiter, D.; Rompuy, T. van; Samm, U.; Schmitz, O.; Sergienko, G.; Unterberg, B.; Westerhof, E.; Wolf, R.C.; Zimmermann, O.

    2006-09-15

    The Dynamic Ergodic Divertor (DED) has recently been taken into operation on TEXTOR. The device is rather flexible and allows the investigation of very different questions. In the present context we concentrate on the divertor aspect and on results of the m/n=12/4 base mode. The DED-field generates the proper ergodic zone and an area of open magnetic field lines, the laminar zone and the tangle structure. The properties of the laminar zone resemble the divertor region of a poloidal divertor. However, the distribution of the density and temperature is highly 3D and strongly related to the structure of the laminar and ergodic zones. The structures of the heat and particle fluxes to the wall agree well with the predicted patterns. A prominent feature of the ergodization is the creation of an edge electric field which results in a rotation of the plasma. (copyright 2006 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  2. Enhancing the DEMO divertor target by interlayer engineering

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  3. Divertor performance on carbon and beryllium targets in JET

    Energy Technology Data Exchange (ETDEWEB)

    Janeschitz, G.; Koenig, R.; Lauro-Taroni, L.; Lingertat, J.; Matthews, G.; Stamp, M.; Vlases, G.; Campbell, D.; Clement, S.; De Kock, L.; Ehrenberg, J.; Gottardi, N.; Harbour, P.; Horton, L.; Jaeckel, H.; Lesourd, M.; Loarte, A.; Lowry, C.; Saibene, G.; Summers, D.; Tagle, J.A.; Thomas, P.R.; Von Hellerman, M. (JET Joint Undertaking, Abingdon (United Kingdom)); Eckstein, W.; Roth, J. (Max Planck Inst. fuer Plasmaphysik, Garching (Germany))

    1992-12-01

    The dependence of impurity production and retention on the divertor density, on the power flow into this region as well as on the X-point to target distance are investigated. Model predictions suggest a good impurity retention above a certain divertor (scrape-off) density threshold, which is dependent on heating power. In our experiments pre-programmed midplane or X-point gas puffs were used to scan the density, as well as to avoid the depletion of particles from the divertor and the scrape-off during H-models. The gas puffs reduce T[sub e] and increase N[sub e] in particular at the outer strike zone. In general the Be as well as the C influx increases with density, which is understood from the T[sub e] (T[sub i]) dependence of the sputtering yields. The impurity retention shows the expected improvement with increasing scrape-off (divertor) density as well as with increasing X-point to target distance (connection length). (orig.).

  4. Diagnostic options for radiative divertor feedback control on NSTX-U.

    Science.gov (United States)

    Soukhanovskii, V A; Gerhardt, S P; Kaita, R; McLean, A G; Raman, R

    2012-10-01

    A radiative divertor technique is used in present tokamak experiments and planned for ITER to mitigate high heat loads on divertor plasma-facing components (PFCs) to prevent excessive material erosion and thermal damage. In NSTX, a large spherical tokamak with lithium-coated graphite PFCs and high divertor heat flux (q(peak) ≤ 15 MW/m(2)), radiative divertor experiments have demonstrated a significant reduction of divertor peak heat flux simultaneously with good core H-mode confinement using pre-programmed D(2) or CD(4) gas injections. In this work diagnostic options for a new real-time feedback control system for active radiative divertor detachment control in NSTX-U, where steady-state peak divertor heat fluxes are projected to reach 20-30 MW/m(2), are discussed. Based on the NSTX divertor detachment measurements and analysis, the control diagnostic signals available for NSTX-U include divertor radiated power, neutral pressure, spectroscopic deuterium recombination signatures, infrared thermography of PFC surfaces, and thermoelectric scrape-off layer current. In addition, spectroscopic "security" monitoring of possible confinement or pedestal degradation is recommended. These signals would be implemented in a digital plasma control system to manage the divertor detachment process via an actuator (impurity gas seeding rate).

  5. 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.

  6. Diagnostic options for radiative divertor feedback control on NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V. A.; Gerhardt, S. P.; Kaita, R.; McLean, A. G.; Raman, R.

    2012-10-01

    A radiative divertor technique is used in present tokamak experiments and planned for ITER to mitigate high heat loads on divertor plasma-facing components (PFCs) to prevent excessive material erosion and thermal damage. In NSTX, a large spherical tokamak with lithium-coated graphite PFCs and high divertor heat flux (qpeak ≤ 15 MW/m2), radiative divertor experiments have demonstrated a significant reduction of divertor peak heat flux simultaneously with good core H-mode confinement using pre-programmed D2 or CD4 gas injections. In this work diagnostic options for a new real-time feedback control system for active radiative divertor detachment control in NSTX-U, where steady-state peak divertor heat fluxes are projected to reach 20–30 MW/m2, are discussed. Based on the NSTX divertor detachment measurements and analysis, the control diagnostic signals available for NSTX-U include divertor radiated power, neutral pressure, spectroscopic deuterium recombination signatures, infrared thermography of PFC surfaces, and thermoelectric scrape-off layer current. In addition, spectroscopic “security” monitoring of possible confinement or pedestal degradation is recommended. These signals would be implemented in a digital plasma control system to manage the divertor detachment process via an actuator (impurity gas seeding rate).

  7. 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.

  8. 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.

  9. 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.

  10. Kinetic effects in edge plasma: kinetic modeling for edge plasma and detached divertor

    Science.gov (United States)

    Takizuka, T.

    2017-03-01

    Detached divertor is considered a solution for the heat control in magnetic-confinement fusion reactors. Numerical simulations using the comprehensive divertor codes based on the plasma fluid modeling are indispensable for the design of the detached divertor in future reactors. Since the agreement in the results between detached-divertor experiments and simulations has been rather fair but not satisfactory, further improvement of the modeling is required. The kinetic effect is one of key issues for improving the modeling. Complete kinetic behaviors are able to be simulated by the kinetic modeling. In this paper at first, major kinetic effects in edge plasma and detached divertor are listed. One of the most powerful kinetic models, particle-in-cell (PIC) model, is described in detail. Several results of PIC simulations of edge-plasma kinetic natures are presented. Future works on PIC modeling and simulation for the deeper understanding of edge plasma and detached divertor are discussed.

  11. Studies of short-range tungsten migration in DIII-D divertor

    Science.gov (United States)

    Rudakov, D. L.; Stangeby, P. C.; Elder, J. D.; Ding, R.; Abrams, T.; Unterberg, E. A.; Briesemeister, A.; Donovan, D.; McLean, A. G.; Guo, H. Y.; Thomas, D. M.; Hinson, E.; Wampler, W. R.; Watkins, J. G.

    2016-10-01

    Two toroidal rings of 5 cm wide W-coated TZM inserts were installed in the lower divertor of DIII-D. Migration of W on the graphite tile surfaces 1-6 cm radially outwards from the outermost ring was studied in a series of 23 reproducible lower single null L-mode discharges with the Outer Strike Point (OSP) placed on the ring. The discharges used 3.2 MW of NBI heating power; plasma density and electron temperature at the OSP were about 1x1020m-3 and 30 eV. W gross erosion rates were measured via monitoring 400.9 nm WI line and applying S/XB coefficient. W deposition was measured on a graphite DiMES sample used as a divertor collector probe. The sample featured two 1 mm wide radial inserts; one was exposed for the whole experiment, the other was exchanged every 4-8 plasma discharges. Measurements of the areal density of W on the inserts by post-mortem RBS analysis show that W deposition is largest in the area of net carbon deposition, possibly due to W re-erosion suppression by C deposits. Measured W coverage in the area of net C erosion is comparable to ERO modeling predictions. Supported by US DOE under DE-FG02-07ER54917, DE-AC04-94AL85000, DE-AC05-00OR22725, DE-AC52-07NA27344, DE-FC02-04ER54698.

  12. An automated approach to magnetic divertor configuration design

    Science.gov (United States)

    Blommaert, M.; Dekeyser, W.; Baelmans, M.; Gauger, N. R.; Reiter, D.

    2015-01-01

    Automated methods based on optimization can greatly assist computational engineering design in many areas. In this paper an optimization approach to the magnetic design of a nuclear fusion reactor divertor is proposed and applied to a tokamak edge magnetic configuration in a first feasibility study. The approach is based on reduced models for magnetic field and plasma edge, which are integrated with a grid generator into one sensitivity code. The design objective chosen here for demonstrative purposes is to spread the divertor target heat load as much as possible over the entire target area. Constraints on the separatrix position are introduced to eliminate physically irrelevant magnetic field configurations during the optimization cycle. A gradient projection method is used to ensure stable cost function evaluations during optimization. The concept is applied to a configuration with typical Joint European Torus (JET) parameters and it automatically provides plausible configurations with reduced heat load.

  13. Preparation of the liquid lithium divertor plates for NSTX

    Science.gov (United States)

    Nygren, R. E.; McKee, G. R.; Fordham, J. A.; Lewis, S. A.; Kugel, H.; Ellis, R. A.; Viola, M. E.; O'Dell, J. S.

    2011-10-01

    Each of the four toroidal panels of the liquid lithium divertor being installed in NSTX for operation in the 2010 campaign is a conical section inclined at 22° like the previous graphite divertor tiles. Each panel is a copper plate clad with stainless steel and a surface layer of porous plasma sprayed molybdenum (Mo) that will host lithium deposited from an evaporator. This paper describes the processes in fabrication; these include cutting to rough shape, die pressing into conical sections, machining to near final shape with holes for electrical heaters, thermocouples and a groove for a cooling tube, brazing of the 0.25-mm cladding and vacuum plasma spraying of the Mo coating.

  14. Spectroscopic Characterizations of the DIII--D Divertor

    Science.gov (United States)

    Isler, R. C.; Klepper, C. C.; Wood, R. D.; Fenstermacher, M. E.; Leonard, A. W.

    1996-11-01

    Radiative losses from the DIII--D divertor have been characterized for various types of discharges by making extensive use of vacuum ultraviolet spectral lines in conjunction with a collisional-radiative model. Carbon and hydrogen account for essentially all the emission with the carbon fraction usually between 50% and 80% of the total. Ion densities are estimated from a simplified approach to modeling using a one-dimensional transport code. The concentrations range from 2%--6% of the electron density in partially detached plasmas, but it appears that carbon may supply most of the electrons in the divertor in attached plasmas. Ion temperatures are measured from Doppler broadening of spectral lines after accounting precisely for the Zeeman/Paschen-Back effect. In general, the ion temperatures agree well with the electron temperatures at the location of the radiating ions as deduced from spectral line ratio measurements and from the modeling.

  15. Progress of ITER full tungsten divertor technology qualification in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Ezato, K., E-mail: ezato.koichiro@jaea.go.jp [Japan Atomic Energy Agency, 801-1, Mukoyma, Naka-shi, Ibaraki (Japan); Suzuki, S.; Seki, Y.; Mohri, K.; Yokoyama, K. [Japan Atomic Energy Agency, 801-1, Mukoyma, Naka-shi, Ibaraki (Japan); Escourbiac, F.; Hirai, T. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Kuznetcov, V. [NIIEFA, 3 doroga na Metallostroy, Metallostroy, St. Petersburg 196641 (Russian Federation)

    2015-10-15

    Highlights: • JAEA has demonstrated tungsten monoblock technology for ITER divertor that needs to withstand the repetitive heat load as high as 20 MW/m{sup 2}. This includes as follows; • Bonding technologies between W and Cu interlayer, and between Cu interlayer and CuCrZr tube. • Non-destructive examination techniques, especially, ultrasonic testing method, and. • Load carrying capability of W monoblock attachment to support structure of ITER divertor. - Abstract: Japan Atomic Energy Agency (JAEA) is in progress for technology qualification toward full-tungsten (W) ITER divertor outer vertical target (OVT), especially, tungsten monoblock technology that needs to withstand the repetitive heat load as high as 20 MW/m{sup 2}. To demonstrate the armor heat sink bonding technology and heat removal capability, 6 small-scale W monoblock mock-ups manufactured by different bonding technologies using different W materials in addition to 4 full-scale prototype plasma-facing units (PFUs). After non-destructive test, the W components were tested under high heat flux (HHF) in ITER Divertor Test Facility (IDTF) at NIIEFA. Consequently, all of the W monoblocks endured the repetitive heat load at 20 MW/m{sup 2} for 1000 cycles (requirements 20 MW/m{sup 2} for 300 cycles) without any failure. In addition to the armor to heat sink joints, the load carrying capability test on the W monoblock with a leg attachment was carried out. In uniaxial tensile test, all of the W monoblock attachments with different bonding technologies such as brazing and HIPping withstand the tensile load exceeding 20 kN that is the value more than twice the design value. The failures occurred at the leg attachments or the W monoblocks, rather than the bonding interface of the W monoblocks to the leg attachment.

  16. 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.

  17. The effect of the magnetic topology on particle recycling in the ergodic divertor of TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Lehnen, M. [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ (Germany)]. E-mail: m.lehnen@fz-juelich.de; Abdullaev, S.S. [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ (Germany); Brezinsek, S. [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ (Germany); Finken, K.H. [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ (Germany); Harting, D. [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ (Germany); Hellermann, M. von [FOM-Rijnhuizen, Association EURATOM-FOM (Netherlands); Jakubowski, M.W. [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ (Germany); Jaspers, R. [FOM-Rijnhuizen, Association EURATOM-FOM (Netherlands); Kirschner, A. [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ (Germany); Pospieszczyk, A. [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ (Germany); Reiter, D. [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ (Germany); Samm, U. [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ (Germany); Schmitz, O. [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ (Germany); Sergienko, G. [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ (Germany); Unterberg, B. [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ (Germany); Wolf, R. [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM-FZJ (Germany)

    2007-06-15

    The influence of the divertor geometry of the dynamic ergodic divertor (DED) in TEXTOR on particle recycling is discussed. The geometry can be varied by the choice of the base mode, the edge safety factor and the divertor coil current. The divertor volume is split into the upstream and the downstream area. Strong plasma flows in the downstream area, essential for high screening efficiency, are predicted. The source strength of deuterium and carbon in the downstream area is estimated by using the two-dimensional distribution of D{sub {alpha}} and CIII emission in front of the target. The results are compared to EMC3 and ERO-code calculations.

  18. 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......; the power deposition increases slightly at the outer targets while shows no obvious variation at the inner targets in double null configuration. The radiated power measured by the extreme ultraviolet arrays increases significantly due to helium gas injection, especially in the outer divertor. The edge...

  19. 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...... 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...

  20. Hybrid formulation of radiation transport in optically thick divertor plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Rosato, J.; Marandet, Y.; Bufferand, H.; Stamm, R. [PIIM, UMR 7345 Aix-Marseille Universite / CNRS, Centre de St-Jerome, Marseille (France); Reiter, D. [IEK-4 Plasmaphysik, Forschungszentrum Juelich GmbH, Juelich (Germany)

    2016-08-15

    Kinetic Monte Carlo simulations of coupled atom-radiation transport in optically thick divertor plasmas can be computationally very demanding, in particular in ITER relevant conditions or even larger devices, e.g. for power plant divertor studies. At high (∝ 10{sup 15} cm{sup -3}) atomic densities, it can be shown that sufficiently large divertors behave in certain areas like a black body near the first resonance line of hydrogen (Lyman α). This suggests that, at least in part, the use of continuum model (radiation hydrodynamics) can be sufficiently accurate, while being less time consuming. In this work, we report on the development of a hybrid model devoted to switch automatically between a kinetic and a continuum description according to the plasma conditions. Calculations of the photo-excitation rate in a homogeneous slab are performed as an illustration. The outlined hybrid concept might be also applicable to neutral atom transport, due to mathematical analogy of transport equations for neutrals and radiation. (copyright 2016 The Authors. Contributions to Plasma Physics published by Wiley-VCH Verlag GmbH and Co. KGaA Weinheim. This)

  1. An Exploration of Advanced X-Divertors on ITER

    CERN Document Server

    Covele, Brent; Kotschenreuther, Mike; Mahajan, Swadesh

    2013-01-01

    It is found that the X-Divertor (XD) configuration [1-3] can be made with the conventional PF coil set on ITER[4], where all PF coils are outside the TF coils. Desirable configurations are possible where the PF currents are below the present maximum design limits on ITER, and where the baseline divertor cassette is used. It is possible that the XD could be used to assist in high-power operation on ITER, but some further issues need examination. Note that the increased major radius of the Super X-Divertor (SXD) [5-8] is not a feature of the XD geometry. In addition, we present an XD configuration for K-DEMO [9], to demonstrate that it is also possible to attain the XD configuration in advanced tokamak reactors with all PF coils outside the TF coils. The results given here for the XD are far more encouraging than recent calculations by Lackner and Zohm [10] for the Snowflake [11,12], where the required high PF currents represent a major technological challenge. The magnetic field structure in the outboard diver...

  2. Divertor plasma conditions and neutral dynamics in horizontal and vertical divertor configurations in JET-ILW low confinement mode plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Groth, M., E-mail: mathias.groth@aalto.fi [Aalto University, Association EURATOM-Tekes, Otakaari 4, Espoo (Finland); Brezinsek, S. [Forschungszentrum Jülich, IEK4 – Plasma Physik, Jülich (Germany); Belo, P. [Institute of Plasmas and Nuclear Fusion, Association EURATOM/IST, Lisbon (Portugal); Culham Centre for Fusion Energy, Association EURATOM-CCFE, Abingdon (United Kingdom); Brix, M. [Culham Centre for Fusion Energy, Association EURATOM-CCFE, Abingdon (United Kingdom); Calabro, G. [Association EURATOM-ENEA, Frascati (Italy); Chankin, A. [Max-Planck Institute for Plasma Physics, EURATOM Association, Garching (Germany); Clever, M.; Coenen, J.W. [Forschungszentrum Jülich, IEK4 – Plasma Physik, Jülich (Germany); Corrigan, G. [Culham Centre for Fusion Energy, Association EURATOM-CCFE, Abingdon (United Kingdom); Drewelow, P. [Max-Planck-Institute for Plasma Physics, EURATOM Association, Greifswald (Germany); Guillemaut, C. [Association EURATOM CEA, CEA/DSM/IRFM, Cadarache (France); Harting, D. [Culham Centre for Fusion Energy, Association EURATOM-CCFE, Abingdon (United Kingdom); Huber, A. [Forschungszentrum Jülich, IEK4 – Plasma Physik, Jülich (Germany); Jachmich, S. [Association ‘Euratom-Belgian state’, Ecole Royale Militaire, Brussels (Belgium); Järvinen, A. [Aalto University, Association EURATOM-Tekes, Otakaari 4, Espoo (Finland); Kruezi, U.; Lawson, K.D. [Culham Centre for Fusion Energy, Association EURATOM-CCFE, Abingdon (United Kingdom); Lehnen, M. [Forschungszentrum Jülich, IEK4 – Plasma Physik, Jülich (Germany); ITER Organisation, 13115 Saint-Paul-Lez-Durance (France); and others

    2015-08-15

    Measurements of the plasma conditions at the low field side target plate in JET ITER-like wall ohmic and low confinement mode plasmas show minor differences in divertor plasma configurations with horizontally and vertically inclined targets. Both the reduction of the electron temperature in the vicinity of the strike points and the rollover of the ion current to the plates follow the same functional dependence on the density at the low field side midplane. Configurations with vertically inclined target plates, however, produce twice as high sub-divertor pressures for the same upstream density. Simulations with the EDGE2D-EIRENE code package predict significantly lower plasma temperatures at the low field side target in vertical than in horizontal target configurations. Including cross-field drifts and imposing a pumping by-pass leak at the low-field side plate can still not recover the experimental observations.

  3. Complications with flush-mounted probe analysis beyond sheath-expansion

    Science.gov (United States)

    Kuang, A. Q.; Labombard, B.; Brunner, D.

    2016-10-01

    In a reactor relevant divertor, the heat-flux onto the target plate would be too large and traditional proud Langmuir probes will melt. By making the probes flush with the surface of the target plate they become nearly as robust as the divertor plates themselves. However, without a theoretically rigorous derivation of the sheath thickness, sheath expansion has been a primary concern for the interpretation of flush mounted probe data. Following the installation of a flush-mounted Langmuir probe system at Alcator C-Mod (toroidally-elongated and field-aligned to give it a `rail' geometry) that effectively mitigates the effects of sheath expansion down to incident field line angles of 0.5 degree, further complications have arisen that cannot be explained by sheath-expansion. The `rail' probes systematically measure lower densities and higher temperatures but have the same pressure. The evolution of the scrape-off layer profiles measured on the divertor target plate from sheath-limited to detached regimes is also different. These are indicative of important physics, perhaps unique to conditions in a vertical-target plate divertor with small field-line attack angles, that affects the I-V characteristics and is not currently included in probe data analyses. Controlled experiments performed at Alcator C-Mod mapped out this discrepancy and the results will be presented. Supported by USDoE Awards DE-FC02-99ER54512.

  4. 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

  5. Investigations on the heat flux and impurity for the HL-2M divertor

    Science.gov (United States)

    Zheng, G. Y.; Cai, L. Z.; Duan, X. R.; Xu, X. Q.; Ryutov, D. D.; Cai, L. J.; Liu, X.; Li, J. X.; Pan, Y. D.

    2016-12-01

    The controllability of the heat load and impurity in the divertor is very important, which could be one of the critical problems to be solved in order to ensure the success for a steady state tokamak. HL-2M has the advantage of the poloidal field (PF) coils placed inside the demountable toroidal field (TF) coils and close to the main plasma. As a result, it is possible to make highly accurate configuration control of the advanced divertor for HL-2M. The divertor target geometry of HL-2M has been designed to be compatible with different divertor configurations to study the divertor physics and support the high performance plasma operations. In this paper, the heat loads and impurities with different divertor configurations, including the standard X-point divertor, the snowflake-minus divertor and two tripod divertor configurations for HL-2M, are investigated by numerical simulations with the SOLPS5.0 code under the current design of the HL-2M divertor geometry. The plasmas with different conditions, such as the low discharge parameters with {{I}\\text{p}}   =  0.5 MA at the first stage of HL-2M and the high parameters with {{I}\\text{p}}   =  2.0 MA during the normal operations, are simulated. The heat load profiles and the impurity distributions are obtained, and the control of the peak heat load and the effect of impurity on the core plasma are discussed. The compatibility of different divertor configurations for HL-2M is also evaluated. It is seen that the excellent compatibility of different divertor configurations with the current divertor geometry has been verified. The results show that the snowflake-minus divertor and the tripod divertor with {{d}x}=30 \\text{cm} present good performance in terms of the heat load profiles and the impurity distributions under different conditions, which may not have a big effect on the core plasma. In addition, it is possible to optimize the distance between the two X-points, {{d}x} , to achieve a better

  6. 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.

  7. Design study of JT-60SA divertor for high heat and particle controllability

    Energy Technology Data Exchange (ETDEWEB)

    Kawashima, H. [Japan Atomic Energy Agency, 801-1, Mukoyama, Naka-shi, Ibaraki-ken 311-0193 (Japan)], E-mail: kawashima.hisato@jaea.go.jp; Shimizu, K.; Takizuka, T.; Asakura, N.; Sakurai, S.; Matsukawa, M.; Fujita, T. [Japan Atomic Energy Agency, 801-1, Mukoyama, Naka-shi, Ibaraki-ken 311-0193 (Japan)

    2008-12-15

    The modification of JT-60 to a fully superconducting coil tokamak, JT-60SA (JT-60 Super Advanced) device, has been programmed to contribute and supplement ITER toward to DEMO. Lower divertor design with the ITER-like lower single null divertor configuration is studied to obtain high heat and particle controllability using SOLDOR/NEUT2D code. With anticipated total power flux into SOL of 37 MW (90% of input power), the peak heat load on outer divertor target can be reduced to 5.8 MW/m{sup 2} at the detached condition by gas puffing in the vertical divertor target with the 'V-shaped corner'. It is {approx}2/5 of the allowable level of 15 MW/m{sup 2}. On the other hand, particle controllability such as control of detached to attached condition by divertor pumping is improved by increase the strike point distance from 20 to 120 mm with above divertor geometry, suggesting that recover from severe detachment at the small distance case can be achieving by elevation of the strike point locations. Optimization of upper divertor design is in progress for high {beta} steady-state operation using upper single null divertor configuration.

  8. Investigation of scrape-off layer and divertor heat transport in ASDEX Upgrade L-mode

    Science.gov (United States)

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

    2016-05-01

    Power exhaust is one of the major challenges for the development of a fusion power plant. Predictions based upon a multimachine database give a scrape-off layer power fall-off length {λq}≤slant 1 mm for large fusion devices such as ITER. The power deposition profile on the target is broadened in the divertor by heat transport perpendicular to the magnetic field lines. This profile broadening is described by the power spreading S. Hence both {λq} and S need to be understood in order to estimate the expected divertor heat load for future fusion devices. For the investigation of S and {λq} L-Mode discharges with stable divertor conditions in hydrogen and deuterium were conducted in ASDEX Upgrade. A strong dependence of S on the divertor electron temperature and density is found which is the result of the competition between parallel electron heat conductivity and perpendicular diffusion in the divertor region. For high divertor temperatures it is found that the ion gyro radius at the divertor target needs to be considered. The dependence of the in/out asymmetry of the divertor power load on the electron density is investigated. The influence of the main ion species on the asymmetric behaviour is shown for hydrogen, deuterium and helium. A possible explanation for the observed asymmetry behaviour based on vertical drifts is proposed.

  9. 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

    The effects of edge-localized modes (ELMs) on divertor particle and heat fluxes were investigated for the first time in the Experimental Advanced Superconducting Tokamak (EAST). The experiments were carried out with both double null and lower single null divertor configurations, and comparisons w...

  10. Geometry and expected performance of the solid tungsten outer divertor row in JET

    NARCIS (Netherlands)

    Rapp, J.; Pintsuk, G.; Mertens, P.; Altmann, H.; Lomas, P. J.; Riccardo, V.

    2010-01-01

    At JET new plasma-facing components for the main chamber wall and the divertor are being designed and built to mimic the expected ITER plasma wall conditions in the deuterium-tritium operation phase. The main wall elements at JET will be made of beryllium and the divertor plasma-facing surface will

  11. Temporal Lorentzian spectral triples

    Science.gov (United States)

    Franco, Nicolas

    2014-09-01

    We present the notion of temporal Lorentzian spectral triple which is an extension of the notion of pseudo-Riemannian spectral triple with a way to ensure that the signature of the metric is Lorentzian. A temporal Lorentzian spectral triple corresponds to a specific 3 + 1 decomposition of a possibly noncommutative Lorentzian space. This structure introduces a notion of global time in noncommutative geometry. As an example, we construct a temporal Lorentzian spectral triple over a Moyal-Minkowski spacetime. We show that, when time is commutative, the algebra can be extended to unbounded elements. Using such an extension, it is possible to define a Lorentzian distance formula between pure states with a well-defined noncommutative formulation.

  12. The dynamical mechanical properties of tungsten under compression at working temperature range of divertors

    Science.gov (United States)

    Zhu, C. C.; Song, Y. T.; Peng, X. B.; Wei, Y. P.; Mao, X.; Li, W. X.; Qian, X. Y.

    2016-02-01

    In the divertor structure of ITER and EAST with mono-block module, tungsten plays not only a role of armor material but also a role of structural material, because electromagnetic (EM) impact will be exerted on tungsten components in VDEs or CQ. The EM loads can reach to 100 MN, which would cause high strain rates. In addition, directly exposed to high-temperature plasma, the temperature regime of divertor components is complex. Aiming at studying dynamical response of tungsten divertors under EM loads, an experiment on tungsten employed in EAST divertors was performed using a Kolsky bar system. The testing strain rates and temperatures is derived from actual working conditions, which makes the constitutive equation concluded by using John-Cook model and testing data very accurate and practical. The work would give a guidance to estimate the dynamical response, fatigue life and damage evolution of tungsten divertor components under EM impact loads.

  13. Simulation of divertor targets shielding during transients in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Pestchanyi, Sergey, E-mail: serguei.pestchanyi@kit.edu [KIT, Hermann-von-Helmholtz-Platz 1, Eggenstein-Leopoldshafen (Germany); Pitts, Richard; Lehnen, Michael [ITER Organization,Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

    2016-11-01

    Highlights: • We simulated plasma shielding effect during disruption in ITER using the TOKES code. • It has been found that vaporization is unavoidable under action of ITER transients, but plasma shielding drastically reduces the divertor target damage: the melt pool and the vaporization region widths reduced 10–15 times. • A simplified 1D model describing the melt pool depth and the shielded heat flux to the divertor targets have been developed. • The results of the TOKES simulations have been compared with the analytic model when the model is valid. - Abstract: Direct extrapolation of the disruptive heat flux on ITER conditions predicts severe melting and vaporization of the divertor targets causing their intolerable damage. However, tungsten vaporized from the target at initial stage of the disruption can create plasma shield in front of the target, which effectively protects the target surface from the rest of the heat flux. Estimation of this shielding efficiency has been performed using the TOKES code. The shielding effect under ITER conditions is found to be very strong: the maximal depth of the melt layer reduced 4 times, the melt layer width—more than 10 times and vaporization region shrinks 10–15 times due to shielding for unmitigated disruption of 350 MJ discharge. The simulation results show complex, 2D plasma dynamics of the shield under ITER conditions. However, a simplified analytic model, valid for rough estimation of the maximum value for the shielded flux to the target and for the melt depth at the target surface has been developed.

  14. Designs of Langmuir probes for W7-X

    Energy Technology Data Exchange (ETDEWEB)

    Laube, Ralph, E-mail: ralph.laube@ipp.mpg.de [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Teilinstitut Greifswald, Wendelsteinstr. 1, D-17491 Greifswald (Germany); Laux, Michael; Ye, Min You [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Teilinstitut Greifswald, Wendelsteinstr. 1, D-17491 Greifswald (Germany); Greuner, Henri; Lindig, Stefan [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Boltzmannstr. 2, D-85748 Garching (Germany)

    2011-10-15

    Several designs of Langmuir probes for the stellarator Wendelstein 7-X (W7-X) are described. Different types of probes are proposed for the different divertors to be used during different operational phases of W7-X. Comb-like arrays of stiff probes, arrays of flexible probes, and fixed inlay probes are reviewed. For the initial phase of W7-X it was decided to install arrays of fixed inlay probes. Two mockups were manufactured and one of them was tested with success in the high heat flux test facility GLADIS. For long-pulse operation of W7-X different conceptual designs are proposed and are still developed further. This paper summarizes the different design constrains for the Langmuir probes in the different divertor surroundings, describes the design of the array of inlay probes for the initial phase and the result of the GLADIS test, and gives a preview of the conceptual designs of probes for the long-pulse operational phase of W7-X.

  15. SOLPS modeling of the effect on plasma detachment of closing the lower divertor in DIII-D

    Science.gov (United States)

    Sang, C. F.; Stangeby, P. C.; Guo, H. Y.; Leonard, A. W.; Covele, B.; Lao, L. L.; Moser, A. L.; Thomas, D. M.

    2017-02-01

    Scrape-off layer plasma simulation modeling has been carried out to assess the effect of tightly closing the lower divertor in DIII-D, which at present is almost fully open, on the achievement of cold dissipative/detached divertor conditions. To isolate the impact of other factors on the divertor plasma solution and to make direct comparisons, most of the parameters including the meshes were kept as similar as possible. Only the neutral baffling was modified to compare a fully open divertor with a tightly closed one. The modeling shows that the tightly closed divertor greatly improves trapping of recycling neutrals, thereby increasing radiative and charge exchange losses in the divertor and reducing the electron temperature T et and deposited power density q dep at the target plate. Furthermore, the closed structure enables the divertor plasma to enter into highly dissipative and detached divertor conditions at a significantly lower upstream density. The effects of divertor closure on the neutral density and pressure, and their correlation with the divertor plasma conditions are also demonstrated. The effect of molecular D2-ion D+ elastic collisions and neutral-neutral collisions on the divertor plasma solution are assessed.

  16. Experiments and computational modeling focused on divertor and SOL optimization for advanced tokamak operation on DIII-D

    Science.gov (United States)

    Allen, S. L.; Boedo, J. A.; Bozek, A. S.; Brooks, N. H.; Carlstrom, T. N.; Casper, T. A.; Colchin, R. J.; Evans, T. E.; Fenstermacher, M. E.; Friend, M. E.; Isler, R. C.; Jayakumar, R.; Lasnier, C. J.; Leonard, A. W.; Mahdavi, M. A.; Maingi, R.; McKee, G. R.; Moyer, R. A.; Murakami, M.; Osborne, T. H.; O'Neill, R. C.; Petrie, T. W.; Porter, G. D.; Ramsey, A. T.; Schaffer, M. J.; Stangeby, P. C.; Stambaugh, R. D.; Wade, M. R.; Watking, J. G.; West, W. P.; Whyte, D. G.; Wolf, N. S.

    2001-03-01

    We present the results from DIII-D experiments and modeling focused on the divertor issues of an `Advanced Tokamak' (AT). Operation at high plasma pressure β with good energy confinement H requires core and divertor plasma shaping and current profile J( r) control with ECH current drive. Transport modeling indicates that the available DIII-D ECH power determines a density and temperature regime for sustained DIII-D AT experiments. We demonstrate that a high-δ, unbalanced double null divertor with cryopumping (D-2000) is a flexible AT divertor. Impurity levels in AT experiments have been reduced by careful alignment of the divertor tiles; this, in turn has changed the time evolution of the core J( r) profiles. New physics has been observed near the X-point and private flux regions, including flow reversal and recombination, that is important in understanding and controlling the flows and thereby the radiation in the divertor region, which reduces the divertor heat flux.

  17. Triple Pulsar Tests Mass Triplets

    CERN Document Server

    Shao, Lijing

    2016-01-01

    Three conceptually different masses appear in equations of motion for objects under gravity, namely, inertial mass, $m_{\\cal I}$, passive gravitational mass, $m_{\\cal P}$, and active gravitational mass, $m_{\\cal A}$. It is assumed that, for any objects, $m_{\\cal I} = m_{\\cal P} = m_{\\cal A}$ in Newtonian gravity, and $m_{\\cal I} = m_{\\cal P}$ in Einsteinian gravity, oblivious to objects' sophisticated internal structure. Empirical examination of the equivalence probes deep into gravity theories. We propose new tests based on pulsar timing of the stellar triple system, PSR J0337+1715. Various machine-precision three-body simulations are performed, from which, equivalence-violating parameters are extracted with Markov chain Monte Carlo sampling that takes full correlations into account. We show that the difference in masses can be probed to $3\\times10^{-8}$, improving the post-Newtonian constraints of strong equivalence principle by $10^3$--$10^6$. The test of $m_{\\cal P}=m_{\\cal A}$ presents the first test of ...

  18. Surface heat loads on the ITER divertor vertical targets

    Science.gov (United States)

    Gunn, J. P.; Carpentier-Chouchana, S.; Escourbiac, F.; Hirai, T.; Panayotis, S.; Pitts, R. A.; Corre, Y.; Dejarnac, R.; Firdaouss, M.; Kočan, M.; Komm, M.; Kukushkin, A.; Languille, P.; Missirlian, M.; Zhao, W.; Zhong, G.

    2017-04-01

    The heating of tungsten monoblocks at the ITER divertor vertical targets is calculated using the heat flux predicted by three-dimensional ion orbit modelling. The monoblocks are beveled to a depth of 0.5 mm in the toroidal direction to provide magnetic shadowing of the poloidal leading edges within the range of specified assembly tolerances, but this increases the magnetic field incidence angle resulting in a reduction of toroidal wetted fraction and concentration of the local heat flux to the unshadowed surfaces. This shaping solution successfully protects the leading edges from inter-ELM heat loads, but at the expense of (1) temperatures on the main loaded surface that could exceed the tungsten recrystallization temperature in the nominal partially detached regime, and (2) melting and loss of margin against critical heat flux during transient loss of detachment control. During ELMs, the risk of monoblock edge melting is found to be greater than the risk of full surface melting on the plasma-wetted zone. Full surface and edge melting will be triggered by uncontrolled ELMs in the burning plasma phase of ITER operation if current models of the likely ELM ion impact energies at the divertor targets are correct. During uncontrolled ELMs in pre-nuclear deuterium or helium plasmas at half the nominal plasma current and magnetic field, full surface melting should be avoided, but edge melting is predicted.

  19. On the W7-X divertor performance under detached conditions

    Science.gov (United States)

    Feng, Y.; Beidler, C. D.; Geiger, J.; Helander, P.; Hölbe, H.; Maassberg, H.; Turkin, Y.; Reiter, D.; W7-X Team

    2016-12-01

    We present a theoretical/numerical predictive analysis of the performance of the W7-X island divertor under conditions of detachment characterized by intensive radiation. The analysis is based on EMC3-Eirene simulations and the earlier W7-AS experimental and numerical experience. Carbon is employed as a representative radiator. The associated drawbacks, i.e. core contamination and recycling degradation (reduced recycling flux), are evaluated by determining the carbon density at the last closed flux surface (LCFS) and the neutral pressure in the divertor chamber. Optimum conditions are explored in both configuration and plasma parameter space. This study aims to identify the key geometric/magnetic and plasma parameters that affect the performance of detached plasmas in W7-X. Emphasis is placed on what occurs when the islands are enlarged far beyond the maximum size available in W7-AS and whether an island size limit for optimal detachment operation exists, and why. Further issues addressed are the power removal ability of the W7-X edge islands, potentially limiting factors, compatibility between particle and power exhaust, and particle refueling capability of the recycling neutrals.

  20. Spectroscopic characterization of the DIII-D divertor

    Science.gov (United States)

    Isler, R. C.; Wood, R. W.; Klepper, C. C.; Brooks, N. H.; Fenstermacher, M. E.; Leonard, A. W.

    1997-02-01

    Radiative losses along a fixed view into the divertor chamber of the DIII-D tokamak [Plasma Physics Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol I, p. 159] have been characterized for attached and partially detached discharges by analyzing line-integrated vacuum ultraviolet (VUV) signals. Essentially all the emission can be ascribed to carbon and deuterium. Because the majority of the most intense lines, which lie at wavelengths above 1100 Å, are not accessible to the present instrumentation, extensive use has been made of collisional-radiative (CR) calculations for level populations of the important ions in order to relate the total radiated power to shorter wavelength transitions. In beam-heated plasmas, the fraction of radiation detected from carbon along the VUV spectrometer view is usually between 50% and 80% of the total. Carbon densities are estimated from a simplified approach to modelling the emission using a one-dimensional transport code. For partially detached plasmas the concentrations range from 2%-6% of the electron density; but in attached plasmas it appears that carbon may supply most of the electrons in the divertor region just below the X point. Ion temperatures are measured from Doppler broadening of spectral lines by fitting measured profiles to theoretical lineshapes, which account precisely for atomic sublevel splitting caused by the Zeeman/Paschen-Back effect in the tokamak magnetic field.

  1. An exploration of advanced X-divertor scenarios on ITER

    Science.gov (United States)

    Covele, B.; Valanju, P.; Kotschenreuther, M.; Mahajan, S.

    2014-07-01

    It is found that the X-divertor (XD) configuration (Kotschenreuther et al 2004 Proc. 20th Int. Conf. on Fusion Energy (Vilamoura, Portugal, 2004) (Vienna: IAEA) CD-ROM file [IC/P6-43] www-naweb.iaea.org/napc/physics/fec/fec2004/datasets/index.html, Kotschenreuther et al 2006 Proc. 21st Int. Conf. on Fusion Energy 2006 (Chengdu, China, 2006) (Vienna: IAEA), CD-ROM file [IC/P7-12] www-naweb.iaea.org/napc/physics/FEC/FEC2006/html/index.htm, Kotschenreuther et al 2007 Phys. Plasmas 14 072502) can be made with the conventional poloidal field (PF) coil set on ITER (Tomabechi et al and Team 1991 Nucl. Fusion 31 1135), where all PF coils are outside the TF coils. Starting from the standard divertor, a sequence of desirable XD configurations are possible where the PF currents are below the present maximum design limits on ITER, and where the baseline divertor cassette is used. This opens the possibility that the XD could be tested and used to assist in high-power operation on ITER, but some further issues need examination. Note that the increased major radius of the super-X-divertor (Kotschenreuther et al 2007 Bull. Am. Phys. Soc. 53 11, Valanju et al 2009 Phys. Plasmas 16 5, Kotschenreuther et al 2010 Nucl. Fusion 50 035003, Valanju et al 2010 Fusion Eng. Des. 85 46) is not a feature of the XD geometry. In addition, we present an XD configuration for K-DEMO (Kim et al 2013 Fusion Eng. Des. 88 123) to demonstrate that it is also possible to attain the XD configuration in advanced tokamak reactors with all PF coils outside the TF coils. The results given here for the XD are far more encouraging than recent calculations by Lackner and Zohm (2012 Fusion Sci. Technol. 63 43) for the Snowflake (Ryutov 2007 Phys. Plasmas 14 064502, Ryutov et al 2008 Phys. Plasmas 15 092501), where the required high PF currents represent a major technological challenge. The magnetic field structure in the outboard divertor SOL (Kotschenreuther 2013 Phys. Plasmas 20 102507) in the recently created

  2. Improvement of the divertor bolometer diagnostic in the ASDEX Upgrade tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Sehmer, Till; Meister, Hans; Bernert, Matthias; Koll, Juergen; Reimold, Felix; Wischmeier, Marco; Fantz, Ursel [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Collaboration: ASDEX Upgrade Team

    2015-05-01

    For future fusion devices such as ITER, the radiation balance in the divertor region will have a significant impact on the power exhaust balance. Therefore, scenarios with strongly localized radiation, like radiation in the high field side high density (HFSHD) region, X-Point radiation or radiation in the divertor legs during detachment, will be investigated in the next ASDEX Upgrade (AUG) operation campaign 2015. To obtain accurately the absolute divertor radiation out of these measurements, the AUG foil bolometer diagnostic system in the divertor region has been enhanced; two new cameras have been designed and manufactured. One will be mounted below the roof baffle and contains 28 lines of sight (LOS), which will observe the mentioned regions of particular physical interest. The second camera consists of 4 LOS and will be mounted at the high field side above the inner divertor nose. It will observe radiation arising from the X-Point region and from the outer divertor. The data will be analysed with a tomographic reconstruction algorithm to localize and quantify the divertor radiation.

  3. Motivation and goals of the new heated outer divertor for Alcator C-Mod

    Science.gov (United States)

    Lipschultz, B.; Doody, J.; Ellis, R.; Granetz, R.; Harrison, S.; Labombard, B.; Vieira, R.; Zhang, H.; Zhou, L.

    2012-10-01

    A precision-aligned, high-temperature outer divertor is being developed for Alcator C-Mod to enhance heatflux handling and to advance our knowledge and experience with high-Z Plasma Facing Components (PFCs) in a reactor-level power density environment. Several departures from the design of the current divertor will be implemented: Instead of 10 toroidal divertor segments that expand toroidally as they heat up, the divertor plate will be toroidally continuous, with no openings or leading edges in the high-heat flux region. It will expand in the radial direction when heated while maintaining good alignment with shallow field line angles (˜ 2 degrees), a requirement for future divertors. Those characteristics will reduce both impurity sources and disruption forces. A second design goal is to be able to control the divertor temperature up to 600^oC by installing heaters in the structure. Given the Arrhenius relation between hydrogen diffusivity and temperature in tungsten (and molybdenum) this will open up a new area of study for tokamaks - exploration of the effect of PFC temperature on fuel retention. Temperature control may also open up a new area of study into the effect of changes in divertor recycling on fueling and core confinement.

  4. Simulation of tokamak SOL and divertor region including heat flux mitigation by gas puffing

    Science.gov (United States)

    Park, Jin-Woo; Na, Yong-Su; Hong, Sang Hee; Ahn, Joon-Wook; Kim, Deok-Kyu; Han, Hyunsun; Shim, Seong Bo; Lee, Hae June

    2012-08-01

    Two-dimensional (2D), scrape-off layer (SOL)-divertor transport simulations are performed using the integrated plasma-neutral-impurity code KTRAN developed at Seoul National University. Firstly, the code is applied to reproduce a National Spherical Torus eXperiment (NSTX) discharge by using the prescribed transport coefficients and the boundary conditions obtained from the experiment. The plasma density, the heat flux on the divertor plate, and the D α emission rate profiles from the numerical simulation are found to follow experimental trends qualitatively. Secondly, predictive simulations are carried out for the baseline operation mode in Korea Superconducting Tokamak Advanced Research (KSTAR) to predict the heat flux on the divertor target plates. The stationary peak heat flux in the KSTAR baseline operation mode is expected to be 6.5 MW/m2 in the case of an orthogonal divertor. To study the mitigation of the heat flux, we investigated the puffing effects of deuterium and argon gases. The puffing position is assumed to be in front of the strike point at the outer lower divertor plate. In the simulations, mitigation of the peak heat flux at the divertor target plates is found to occur when the gas puffing rate exceeds certain values, ˜1.0 × 1020 /s and ˜5.0 × 1018 /s for deuterium and argon, respectively. Multi-charged impurity transport is also investigated for both NSTX and KSTAR SOL and divertor regions.

  5. The WEST project: Current status of the ITER-like tungsten divertor

    Energy Technology Data Exchange (ETDEWEB)

    Missirlian, M., E-mail: marc.missirlian@cea.fr; Bucalossi, J.; Corre, Y.; Ferlay, F.; Firdaouss, M.; Garin, P.; Grosman, A.; Guilhem, D.; Gunn, J.; Languille, P.; Lipa, M.; Richou, M.; Tsitrone, E.

    2014-10-15

    Highlights: • We presented the ITER-like W components occurred for the WEST divertor. • The main features including key elements of the design were detailed. • The main results of studies investigating the integration constraints or issues were reported. • The WEST ITER-like divertor design reached a mature stage to enable the launching of the procurement phase. - Abstract: The WEST (W – for tungsten – Environment in Steady-state Tokamak) project is an upgrade of Tore Supra from a limiter based tokamak with carbon PFCs into an X-point divertor tokamak with full-tungsten armour while keeping its long discharge capability. The WEST project will primarily offer the key capability of testing for the first time the ITER technology in real plasma environment. In particular, the main divertor (i.e. the lower divertor) of the WEST project will be based on actively cooled tungsten monoblock components and will follow as closely as possible the design and the assembling technology, foreseen for the ITER divertor units. The current design of WEST ITER-like tungsten divertor has now reached a mature stage following the 2013 WEST Final Design Review. This paper presents the key elements of the design, reports the technological requirements and reviews the main design and integration issues.

  6. Waardecreatie in triple helix : Recepten voor triple helix samenwerking

    NARCIS (Netherlands)

    Vos, P.M.; Vries, F. de

    2016-01-01

    Om innovaties in het veiligheidsdomein te realiseren worden triple helix samenwerkingen gezien als een belangrijke motor. Een triple helix samenwerking is een tijdelijk samenwerkingsverband tussen drie of meer organisaties die middelen, risico’s en opbrengsten delen om individuele organisatiedoelen,

  7. Triple Point Topological Metals

    Directory of Open Access Journals (Sweden)

    Ziming Zhu

    2016-07-01

    Full Text Available Topologically protected fermionic quasiparticles appear in metals, where band degeneracies occur at the Fermi level, dictated by the band structure topology. While in some metals these quasiparticles are direct analogues of elementary fermionic particles of the relativistic quantum field theory, other metals can have symmetries that give rise to quasiparticles, fundamentally different from those known in high-energy physics. Here, we report on a new type of topological quasiparticles—triple point fermions—realized in metals with symmorphic crystal structure, which host crossings of three bands in the vicinity of the Fermi level protected by point group symmetries. We find two topologically different types of triple point fermions, both distinct from any other topological quasiparticles reported to date. We provide examples of existing materials that host triple point fermions of both types and discuss a variety of physical phenomena associated with these quasiparticles, such as the occurrence of topological surface Fermi arcs, transport anomalies, and topological Lifshitz transitions.

  8. Diagnostic tools for studying divertor detachment: bolometry, spectroscopy, and thermography for surface heat-flux

    Science.gov (United States)

    Terry, J. L.; Reinke, M. L.

    2017-04-01

    Some of the key aspects of divertor detachment that are addressed by bolometry, impurity spectroscopy, hydrogen spectroscopy, and measurements of divertor target heat-flux are reviewed. Measurement requirements for these diagnostic areas are defined, and brief descriptions of the techniques used for these diagnostics are given. Examples from the literature of measurements using these tools applied to detachment are presented. Feedback control of detachment using some of these diagnostics as the ‘sensors’ is reviewed. Challenges and some future directions for these diagnostics in the context of studying divertor detachment are described.

  9. Modeling of neutral pressure and pumping in the Tore Supra ergodic divertor and outboard pump limiter

    Energy Technology Data Exchange (ETDEWEB)

    Owen, L.W. [Oak Ridge National Lab., TN (United States); Loarer, T. [Association CEA-Euratom, CEN/Cadarache, F-13108 St. Paul-les-Durance Cedex (France); Grosman, A. [Association CEA-Euratom, CEN/Cadarache, F-13108 St. Paul-les-Durance Cedex (France); Meslin, B. [Association CEA-Euratom, CEN/Cadarache, F-13108 St. Paul-les-Durance Cedex (France); Klepper, C.C. [Oak Ridge National Lab., TN (United States); Mioduszewski, P.K. [Oak Ridge National Lab., TN (United States); Uckan, T. [Oak Ridge National Lab., TN (United States)

    1997-02-01

    Active control of the core plasma density and partial depletion of the wall particle content have been achieved in experiments on Tore Supra with the plasma leaning on either the ergodic divertor (ED) or the pump limiters. Measurements of neutral pressures in the ED and outboard pump limiter (OPL) are modeled with 1D parallel transport equations (continuity and momentum balance) for the SOL plasma coupled to 2D neutral particle transport simulations. SOL density and temperature profiles from reciprocating Langmuir probe measurements for a range of volume-averaged densities are renormalized, where necessary, to agree with Langmuir probe measurements in the OPL throat and constitute the upstream boundary conditions for the 1D calculations. Good agreement with measured pressures and exhaust rates are obtained for both the ED and OPL in scans that span a factor of 2-3 in volume-averaged density. The importance of a self-consistent treatment of the plasma and neutral particle transport in the neighborhood of the neutralizer plate is demonstrated, particularly in the stronger recycling regimes characteristic of densities at the high end of the scans. Plasma flow reversal near the plasma/plenum interface is predicted to occur at the higher densities due to the large local ionization source. Predictions of pressure buildup in the plenum behind the prototype vented neutralizer plate agree with experiment if it is assumed that both the tops and partially the sides of the needles comprising the plate are wetted by the plasma. A discharge in which the ED pumps are active is analyzed; the calculated pressure and exhaust rate agree with experiment. The core fueling rate is the same as without pumping, suggesting, as is seen in the experiment, a small density decay rate and significant wall particle depletion. (orig.).

  10. Pythagorean Triples from Harmonic Sequences.

    Science.gov (United States)

    DiDomenico, Angelo S.; Tanner, Randy J.

    2001-01-01

    Shows how all primitive Pythagorean triples can be generated from harmonic sequences. Use inductive and deductive reasoning to explore how Pythagorean triples are connected with another area of mathematics. (KHR)

  11. 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%.

  12. 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.

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

    Science.gov (United States)

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

    2012-12-01

    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.

  14. One dimensional simulation on stability of detached plasma in a tokamak divertor

    Energy Technology Data Exchange (ETDEWEB)

    Nakazawa, Shinji; Nakajima, Noriyoshi; Okamoto, Masao; Ohyabu, Nobuyoshi [National Inst. for Fusion Science, Toki, Gifu (Japan)

    1999-06-01

    The stability of radiation front in the Scrape-Off-Layer (SOL) of a tokamak is studied with a one dimensional fluid code; the time-dependent transport equations are solved in the direction parallel to a magnetic field line. The simulation results show that stable detached solutions exist, where the plasma temperature near the divertor target is {approx}2 eV. It is found that whenever such stable detached states are attained, the strong radiation front is contact with or at a small distance from the divertor target. When the energy externally injected into the SOL is decreased below a critical value, the radiation front starts to move towards the X-point, cooling the SOL plasma. In such cases, no stationary solutions such that the radiation front rests in the divertor channel are observed in our parameter space. This qualitatively corresponds to the results of tokamak divertor experiments which show the movement of radiation front. (author)

  15. Comparison of JET main chamber erosion with dust collected in the divertor

    CERN Document Server

    Widdowson, A; Booth, S; Coad, J P; Hakola, A; Heinola, K; Ivanova, S; Koivuranta, S; Likonen, J; Mayer, M; Stamp, M; Contributors, JET-EFDA

    2013-01-01

    A complete global balance for carbon in JET requires knowledge of the net erosion in the main chamber, net deposition in the divertor and the amount of dust and flakes collecting in the divertor region. This paper describes a number of measurements on aspects of this global picture. Profiler measurements and cross section microscopy on tiles that were removed in the 2009 JET intervention are used to evaluate the net erosion in the main chamber and net deposition in the divertor. In addition the mass of dust and flakes collected from the JET divertor during the same intervention is also reported and included as part of the balance. Spectroscopic measurements of carbon erosion from the main chamber are presented and compared with the erosion measurements for the main chamber.

  16. 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.

  17. 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.

  18. Numerical optimization of tungsten monoblock tile in EAST divertor

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Xiahua [Harbin Engineering University, Harbin 150001 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Ding, Fang, E-mail: fding@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Mao, Hongmin; Luo, Guangnan; Hu, Zhenhua; Xu, Feng; Niu, Guojian [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2016-10-15

    Highlights: • A method based on Kriging model and Uniform Design is developed and applied to the geometry optimization of EAST W tile. • An optimized chamfering geometry is obtained and significantly reduces the maximum temperature on W monoblock. • The incident angle of plasma flux has a strong impact on the optimized chamfering geometry. - Abstract: The ITER-like tungsten divertor with toroidally symmetric 1 mm × 1 mm chamfers on monoblock tiles has been installed in EAST in 2014. Hot spots were experimentally observed mostly along the toridial facing gaps between two columns of W/Cu monoblock units, which are often aggravated by installation misalignment. These hot spots can significantly degrade the power handling capability of W divertor and need to be alleviated. A numerical optimization model for tile chamfering design is built based on the finite element method (FEM), in which the numerical experiments are designed by the uniform table. The calculation results in ANSYS for these experiments are then processed employing the code Design and Analysis of Computer Experiments (DACE) in which the Kriging method is adopted to reconstruct a response surface. The optimum geometry can be derived from the minimum point on the surface. The results show that, under 200 MW/m{sup 2} parallel heat flux with an inclination angle of 3° with respect to tile surface, the maximum temperature on W tile with a 0.5 mm misalignment can be decreased to 2084 °C by adopting an optimized single-sided chamfer, 1.8 times lower than 1 mm × 1 mm symmetrically chamfered tile. The optimum chamfering geometry has a strong dependence on the inclination angle of plasma flux to tile surface. As a result, the monoblock tiles in a flat cassette module need to be chamfered differently to adapt to the varied inclination angles.

  19. Design study of JT-60SA divertor for high heat and particle controllability

    Energy Technology Data Exchange (ETDEWEB)

    Kawashima, H.; Shimizu, K.; Takizuka, T.; Asakura, N.; Sakurai, S.; Matsukawa, M.; Fujita, T. [Japan Atomic Energy Agency (Japan)

    2007-07-01

    In steady-state high performance plasma over 41 MW/100 s in the JT-60SA tokamak, the heat and particle flux density on the divertor targets are considerably higher than those of existing devices such as JT-60U. A divertor modeling code, SOLDOR/NEUT2D, has been applied in order to optimiz the JT-60SA divertor design in such conditions. The heat load q{sub heat} on divertor target is estimated for a conceptual divertor design as the first step. Simulation of SOL/divertor plasmas is carried out at lower single null divertor (LSN) configuration with I{sub p}/B{sub t}=3.5 MA/2.5 T. For the present calculation, anticipated SOL power flux of Q{sub total}=35 MW and particle fuelling flux of G{sub ion}=5.10{sup 21}/s (n{sub e-dege}=3.10{sup 19}/m) are applied. The pumping speed (S{sub pump}=50 m{sup 3}/s) is specified by an albedo for neutrals in front of the cryopump set bottom of exhaust chamber. The recycling of deuterium is assumed to be 100% at the first wall. For the first simulation, the carbon contamination in SOL/divertor regions is set to 2% of electron density uniformly. Gas puff flux G{sub puff}=0.5.10{sup 21}/s is introduced from outside midplane. We assume particle diffusion coefficient D=0.3 m{sup 2}/s and thermal diffusivity of electron and ion X{sub e}=X{sub i}=1 m{sup 2}/s. As a result, attached and detached plasma conditions are simulated on outer and inner divertor regions, respectively. The heat load around the outer strike point reaches 31 MW/m{sup 2}, which largely exceeds the allowable range of 15 MW/m{sup 2} for CFC materials. Reduction of heat load must be achieved somehow. An effect of the radiation cooling is simulated to reduce such a large heat load as the second step. To enlarge the radiative cooling, we increased the gas puff flux by a factor of ten and the carbon contamination partly in the outer divertor region from 2% to 4%. It gives a favorable result that the peak heat load is reduced to 12 MW/m{sup 2} with radiation enhancement by a

  20. Experimental study of electroinsulating coatings in gallium coolant related to the divertor cooling loop

    Science.gov (United States)

    Beznosov, A. V.; Sherbakov, R. V.; Karatushina, I. V.; Romanov, P. V.

    1996-10-01

    Experimental investigation of electroinsulating coatings stability on the samples made of stainless stell, vanadium alloy and beryllium has been conducted at 80-350°C. The impact of gas pressure upon the liquid gallium open surface was studied. The stability of electroinsulating film parameters on divertor structure materials was confirmed for the divertor with open liquid metal coolant surface in the vacuum chamber.

  1. Favorable effects of turbulent plasma mixing on the performance of innovative tokamak divertors

    Science.gov (United States)

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

    2013-10-01

    The problem of reducing the heat load on plasma-facing components is one of the most demanding issues for MFE devices. The general approach to the solution of this problem is the use of a specially configured poloidal magnetic field, so called magnetic divertors. In recent years, novel divertors possessing the 2-nd and 3-rd order nulls of the poloidal field (PF) have been proposed. They are called a ``snowflake'' (SF) and a ``cloverleaf'' (CL) divertor, respectively, due to characteristic shape of the magnetic separatrix. Among several beneficial features of such divertors is an effect of strong turbulent plasma mixing that is intrinsic to the zone of weak PF near the null-point. The turbulence spreads the heat flux between multiple divertor exhaust channels and increases the heat flux width within each channel. Among physical processes affecting the onset of convection the curvature-driven mode of axisymmetric rolls is most prominent. The effect is quite significant for the SF and is even stronger for the CL divertor. Projections to future ITER-scale facilities are discussed. Work performed for U.S. DoE by LLNL under Contract DE-AC52-07NA27344.

  2. 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.

  3. 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)

  4. Evaluation of heat and particle controllability on the JT-60SA divertor

    Energy Technology Data Exchange (ETDEWEB)

    Kawashima, H., E-mail: kawashima.hisato@jaea.go.jp [Japan Atomic Energy Agency, 801-1, Mukoyama, Naka, Ibaraki 311-0193 (Japan); Hoshino, K.; Shimizu, K.; Takizuka, T.; Ide, S.; Sakurai, S.; Asakura, N. [Japan Atomic Energy Agency, 801-1, Mukoyama, Naka, Ibaraki 311-0193 (Japan)

    2011-08-01

    The JT-60SA divertor design has been established on the basis of engineering requirements and physics analysis. Heat and particle fluxes under the full input power of 41 MW can give severe heat loads on the divertor targets, while the allowable heat load is limited below 15 MW/m{sup 2}. Dependence of the heat flux mitigation on a D{sub 2} gas-puff is evaluated by SONIC simulations for high density (n{sub e{_}ave} {approx} 1 x 10{sup 20} m{sup -3}) high current plasmas. It is found that the peak heat load 10 MW/m{sup 2} with dense (n{sub ed} > 4 x 10{sup 20} m{sup -3}) and cold (T{sub ed}, T{sub id} {<=} 1 eV) divertor plasmas are obtained at a moderate gas-puff of {Gamma}{sub puff} = 15 x 10{sup 21} s{sup -1}. Divertor plasmas are controlled from attached to detached condition using the divertor pump with pumping-speed below 100 m{sup 3}/s. In full non-inductive current drive plasmas with low density (n{sub e{_}ave} {approx} 5 x 10{sup 19} m{sup -3}), the reduction of divertor heat load is achieved with the Ar injection.

  5. Energy deposition onto HL-2A divertor plates in ELMy H-mode discharges using infrared thermography

    Energy Technology Data Exchange (ETDEWEB)

    Gao, J.M., E-mail: gaojm@swip.ac.cn; Li, W.; Liu, Y.; Ji, X.Q.; Cheng, J.; Dong, Y.B.; Chen, C.Y.; Feng, B.B.; Lu, J.; Yi, P.; Yang, Q.W.

    2015-08-15

    Using infrared (IR) thermography, power loads onto the divertor plates have been investigated in ELMy H-mode plasmas on HL-2A. In the ELMy H-mode discharges, ELMs are the largest contributors to the divertor target energy load. Analysis of energy balance shows that up to 45% of the energy losses are deposited onto the divertor targets during ELMs and about 30% are found as plasma radiation. Moreover, divertor heat flux mitigation has been achieved during an ELMy H-mode phase by using Supersonic Molecular Beam Injection (SMBI), characterized by a sharp increase of ELM frequency and a reduction in peak heat flux. The increased plasma radiation energy losses, especially the doubled plasma radiation in the divertor region, should be responsible for the reduction of integrated energy deposition onto divertor targets.

  6. Non-resonant triple alpha reaction rate at low temperature

    Energy Technology Data Exchange (ETDEWEB)

    Itoh, T.; Tamii, A.; Aoi, N.; Fujita, H.; Hashimoto, T.; Miki, K.; Ogata, K. [Research Center for Nuclear Physics, Osaka University, Ibaraki, Osaka 567-0047 (Japan); Carter, J.; Donaldson, L.; Sideras-Haddad, E. [Schools of Physics, University of Witwatersrand, Johannesburg 2050 (South Africa); Furuno, T.; Kawabata, T. [Departments of Physics, Kyoto University, Sakyo, Kyoto, 606-8502 (Japan); Kamimura, M. [RIKEN Nishina Center, Wako, Saitama, 351-0198 (Japan); Nemulodi, F.; Neveling, R.; Smit, F. D.; Swarts, C. [iThemba Laboratory for Accelerator Based Sciences Somerset, West, 7129 (South Africa)

    2014-05-02

    Our experimental goal is to study the non-resonant triple alpha reaction rate at low temperture (T < 10{sup 8} K). The {sup 13}C(p,d) reaction at 66 MeV has been used to probe the alpha-unbound continuum state in {sup 12}C just below the 2{sup nd} 0{sup +} state at 7.65 MeV. The transition strength to the continuum state is predicted to be sensitive to the non-resonant triple alpha reaction rate. The experiment has been performed at iThemba LABS. We report the present status of the experiment.

  7. Mobile Probing and Probes

    DEFF Research Database (Denmark)

    2013-01-01

    Mobile probing is a method, developed for learning about digital work situations, as an approach to discover new grounds. The method can be used when there is a need to know more about users and their work with certain tasks, but where users at the same time are distributed (in time and space......). Mobile probing was inspired by the cultural probe method, and was influenced by qualitative interview and inquiry approaches. The method has been used in two subsequent projects, involving school children (young adults at 15-17 years old) and employees (adults) in a consultancy company. Findings point...... to mobile probing being a flexible method for uncovering the unknowns, as a way of getting rich data to the analysis and design phases. On the other hand it is difficult to engage users to give in depth explanations, which seem easier in synchronous dialogs (whether online or face2face). The development...

  8. Mobile Probing and Probes

    DEFF Research Database (Denmark)

    2012-01-01

    Mobile probing is a method, which has been developed for learning about digital work situations, as an approach to discover new grounds. The method can be used when there is a need to know more about users and their work with certain tasks, but where users at the same time are distributed (in time...... and space). Mobile probing was inspired by the cultural probe method, and was influenced by qualitative interview and inquiry approaches. The method has been used in two subsequent projects, involving school children (young adults at 15-17 years old) and employees (adults) in a consultancy company. Findings...... point to mobile probing being a flexible method for uncovering the unknowns, as a way of getting rich data to the analysis and design phases. On the other hand it is difficult to engage users to give in depth explanations, which seem easier in synchronous dialogs (whether online or face2face...

  9. Mobile Probing and Probes

    DEFF Research Database (Denmark)

    2012-01-01

    Mobile probing is a method, which has been developed for learning about digital work situations, as an approach to discover new grounds. The method can be used when there is a need to know more about users and their work with certain tasks, but where users at the same time are distributed (in time...... and space). Mobile probing was inspired by the cultural probe method, and was influenced by qualitative interview and inquiry approaches. The method has been used in two subsequent projects, involving school children (young adults at 15-17 years old) and employees (adults) in a consultancy company. Findings...... point to mobile probing being a flexible method for uncovering the unknowns, as a way of getting rich data to the analysis and design phases. On the other hand it is difficult to engage users to give in depth explanations, which seem easier in synchronous dialogs (whether online or face2face...

  10. 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.

  11. The dynamical mechanical properties of tungsten under compression at working temperature range of divertors

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, C.C. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China); 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); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China); Peng, X.B., E-mail: pengxb@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Wei, Y.P. [Key Laboratory of Mechanics in Fluid Solid Coupling Systems, Institute of Mechanics, Chinese Academy of Sciences, Beijing 100190 (China); Mao, X. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Li, W.X.; Qian, X.Y. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China)

    2016-02-15

    In the divertor structure of ITER and EAST with mono-block module, tungsten plays not only a role of armor material but also a role of structural material, because electromagnetic (EM) impact will be exerted on tungsten components in VDEs or CQ. The EM loads can reach to 100 MN, which would cause high strain rates. In addition, directly exposed to high-temperature plasma, the temperature regime of divertor components is complex. Aiming at studying dynamical response of tungsten divertors under EM loads, an experiment on tungsten employed in EAST divertors was performed using a Kolsky bar system. The testing strain rates and temperatures is derived from actual working conditions, which makes the constitutive equation concluded by using John-Cook model and testing data very accurate and practical. The work would give a guidance to estimate the dynamical response, fatigue life and damage evolution of tungsten divertor components under EM impact loads. - Graphical abstract: From the comparison between the experimental curves and the predicted curves calculated by adopting the corrected m, it is very clear that the new model is of great capability to explain the deformation behavior of the tungsten material under dynamic compression at high temperatures. (EC, PC and PCM refers to experimental curve, predicted curve and predicted curve with a corrected m. Different colors represent different scenarios.). - Highlights: • Test research on dynamic properties of tungsten at working temperature range and strain rate range of divertors. • Constitutive equation descrbing strain hardening, strain rate hardening and temperature softening. • A guidance to estimate dynamical response and damage evolution of tungsten divertor components under impact.

  12. Pulse plasma sintering of a tungsten/steel divertor module

    Energy Technology Data Exchange (ETDEWEB)

    Kruszewski, Mirosław J., E-mail: m.kruszewski@inmat.pw.edu.pl; Ciupiński, Łukasz; Rosiński, Marcin; Michalski, Andrzej; Kurzydłowski, Krzysztof J.

    2013-10-15

    Highlights: • W/WL10 and WL10/steel joints were fabricated via pulse plasma sintering. • Fe interlayer successfully compensated thermal stresses at the WL10/steel joint. • Maximum temperature of a single stage sintering of the module was established. • Better accuracy in machining of W and WL10 elements is needed. -- Abstract: The paper presents the preliminary evaluation of the potential of a pulse plasma sintering (PPS) technique for the fabrication of a He-cooled modular divertor with a multiple-jet cooling module. In this work the W and WL10 elements were directly bonded by PPS. Examination of the microstructure revealed some minor defects at the interface, but the overall quality of the joint was good with no cracks or delamination being detected. To reduce the thermal stress gradient a thin transition layer of iron was used at the WL10/steel interface. In addition an attempt was made to fabricate the complete module by a single sintering process. The microstructures of the fabricated modules were examined and the findings were reported.

  13. Response of NSTX liquid lithium divertor to high heat loads

    Energy Technology Data Exchange (ETDEWEB)

    Abrams, T., E-mail: tabrams@pppl.gov [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Jaworski, M.A. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Kallman, J. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Foley, E.L. [Nova Photonics, Inc., Princeton, NJ 08543 (United States); Gray, T.K. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Kugel, H. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Levinton, F. [Nova Photonics, Inc., Princeton, NJ 08543 (United States); McLean, A.G. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Skinner, C.H. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States)

    2013-07-15

    Samples of the NSTX Liquid Lithium Divertor (LLD) with and without an evaporative Li coating were directly exposed to a neutral beam ex-situ at a power of ∼1.5 MW/m{sup 2} for 1–3 s. Measurements of front face and bulk sample temperature were obtained. Predictions of temperature evolution were derived from a 1D heat flux model. No macroscopic damage occurred when the “bare” sample was exposed to the beam but microscopic changes to the surface were observed. The Li-coated sample developed a lithium hydroxide (LiOH) coating, which did not change even when the front face temperature exceeded the pure Li melting point. These results are consistent with the lack of damage to the LLD surface and imply that heating alone may not expose pure liquid Li if the melting point of surface impurities is not exceeded. This suggests that flow and heat are needed for future PFCs requiring a liquid Li surface.

  14. 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.

  15. Effects of ELMs on ITER divertor armour materials

    Energy Technology Data Exchange (ETDEWEB)

    Zhitlukhin, A. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation)]. E-mail: zhitlukh@triniti.ru; Klimov, N. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation); Landman, I. [Forschungszentrum Karlsruhe, P.O. Box 3640, 76021 Karlsruhe (Germany); Linke, J. [Forschungszentrum Juelich, EURATOM-Association, Juelich (Germany)]. E-mail: j.linke@fz-juelich.de; Loarte, A. [EFDA, Boltzmannstr. 2, 85748 Garching (Germany); Merola, M. [EFDA, Boltzmannstr. 2, 85748 Garching (Germany); Podkovyrov, V. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation); Federici, G. [ITER JWS Garching, Boltzmannstr. 2, 85748 Garching (Germany); Bazylev, B. [Forschungszentrum Karlsruhe, P.O. Box 3640, 76021 Karlsruhe (Germany); Pestchanyi, S. [Forschungszentrum Karlsruhe, P.O. Box 3640, 76021 Karlsruhe (Germany); Safronov, V. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation); Hirai, T. [Forschungszentrum Juelich, EURATOM-Association, Juelich (Germany); Maynashev, V. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation); Levashov, V. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation); Muzichenko, A. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation)

    2007-06-15

    This paper is concerned with investigation of an erosion of the ITER-like divertor plasma facing components under plasma heat loads expected during the Type I ELMs in ITER. These experiments were carried out on plasma accelerator QSPA at the SRC RF TRINITI under EU/RF collaboration. Targets were exposed by series repeated plasma pulses with heat loads in a range of 0.5-1.5 MJ/m{sup 2} and pulse duration 0.5 ms. Erosion of CFC macrobrushes was determined mainly by sublimation of PAN-fibres that was less than 2.5 {mu}m per pulse. The CFC erosion was negligible at the energy density less than 0.5 MJ/m{sup 2} and was increased to the average value 0.3 {mu}m per pulse at 1.5 MJ/m{sup 2}. The pure tungsten macrobrushes erosion was small in the energy range of 0.5-1.3 MJ/m{sup 2}. The sharp growth of tungsten erosion and the intense droplet ejection were observed at the energy density of 1.5 MJ/m{sup 2}.

  16. Current understanding of divertor detachment: experiments and modelling

    Energy Technology Data Exchange (ETDEWEB)

    Wischmeier, W; Groth, M; Kallenbach, A; Chankin, A; Coster, D; Dux, R; Herrmann, A; Muller, H; Pugno, R; Reiter, D; Scarabosio, A; Watkins, J; Team, T D; Team, A U

    2008-05-23

    A qualitative as well as quantitative evaluation of experimentally observed plasma parameters in the detached regime proves to be difficult for several tokamaks. A series of ohmic discharges have been performed in ASDEX Upgrade and DIII-D at similar as possible plasma parameters and at different line averaged densities, {bar n}{sub e}. The experimental data represent a set of well diagnosed discharges against which numerical simulations are compared. For the numerical modeling the fluid-code B2.5 coupled to the Monte Carlo neutrals transport code EIRENE is used. Only the combined enhancement of effects, such as geometry, drift terms, neutral conductance, increased radial transport and divertor target composition, explains a significant fraction of the experimentally observed asymmetries of the ion fluxes as a function of {bar n}{sub e} to the inner and outer target plates in ASDEX Upgrade. The relative importance of the mechanisms leading to detachment are different in DIII-D and ASDEX Upgrade.

  17. Tokamak edge Er studies by turbulence and divertor simulations

    Science.gov (United States)

    Nishimura, Y.; Coster, D.; Scott, B.

    2002-11-01

    Numerical coupling of the divertor code B2(B. J. Braams, Next European Torus Technical Report 68 (1987).) and the turbulence code DALF(B. D. Scott, Phys. Fluids B 4), 2468 (1992). is pursued. Within this model, space and time dependent transport coefficients (D and i) respond to the dynamics of drift wave turbulence. The Braginskii transport model of the B2 code incorporates guiding-center plasma drifts self-consistently and generate Er shear in the presence of steep pressure gradients. This Braginskii type Er can enter the turbulence model as a background E × B shear flow which suppresses the radial flux together with Reynolds stress induced electric fields. As an example of L-H transition, influx at the core boundary is controlled to produce steepening of the edge gradients. ( Y.Hamada et al.), in Proceedings of the 17th IAEA Fusion Energy Conference (IAEA-F1-CN-69/PD, 1998) reveals heat pulse induced L-H transitions after sawtooth events.

  18. Plasma flow interaction with ITER divertor related surfaces

    Science.gov (United States)

    Dojčinović, Ivan P.

    2010-11-01

    It has been found that the plasma flow generated by quasistationary plasma accelerators can be used for simulation of high energy plasma interaction with different materials of interest for fusion experiments. It is especially important for the studies of the processes such as ELMs (edge localized modes), plasma disruptions and VDEs (vertical displacement events), during which a significant part of the confined hot plasma is lost from the core to the SOL (scrape off layer) enveloping the core region. Experiments using plasma guns have been used to assess erosion from disruptions and ELMs. Namely, in this experiment modification of different targets, like tungsten, molybdenum, CFC and silicon single crystal surface by the action of hydrogen and nitrogen quasistationary compression plasma flow (CPF) generated by magnetoplasma compressor (MPC) has been studied. MPC plasma flow with standard parameters (1 MJ/m2 in 0.1 ms) can be used for simulation of transient peak thermal loads during Type I ELMs and disruptions. Analysis of the targets erosion, brittle destruction, melting processes, and dust formation has been performed. These surface phenomena are results of specific conditions during CPF interaction with target surface. The investigations are related to the fundamental aspects of high energy plasma flow interaction with different material of interest for fusion. One of the purposes is a study of competition between melting and cleavage of treated solid surface. The other is investigation of plasma interaction with first wall and divertor component materials related to the ITER experiment.

  19. Effects of ELMs on ITER divertor armour materials

    Science.gov (United States)

    Zhitlukhin, A.; Klimov, N.; Landman, I.; Linke, J.; Loarte, A.; Merola, M.; Podkovyrov, V.; Federici, G.; Bazylev, B.; Pestchanyi, S.; Safronov, V.; Hirai, T.; Maynashev, V.; Levashov, V.; Muzichenko, A.

    2007-06-01

    This paper is concerned with investigation of an erosion of the ITER-like divertor plasma facing components under plasma heat loads expected during the Type I ELMs in ITER. These experiments were carried out on plasma accelerator QSPA at the SRC RF TRINITI under EU/RF collaboration. Targets were exposed by series repeated plasma pulses with heat loads in a range of 0.5-1.5 MJ/m2 and pulse duration 0.5 ms. Erosion of CFC macrobrushes was determined mainly by sublimation of PAN-fibres that was less than 2.5 μm per pulse. The CFC erosion was negligible at the energy density less than 0.5 MJ/m2 and was increased to the average value 0.3 μm per pulse at 1.5 MJ/m2. The pure tungsten macrobrushes erosion was small in the energy range of 0.5-1.3 MJ/m2. The sharp growth of tungsten erosion and the intense droplet ejection were observed at the energy density of 1.5 MJ/m2.

  20. Deuterium Balmer/Stark spectroscopy and impurity profiles: first results from mirror-link divertor spectroscopy system on the JET ITER-like wall

    CERN Document Server

    Meigs, A G; Clever, M; Huber, A; Marsen, S; Nicholas, C; Stamp, M; Zastrow, K-D; Contributors, JET EFDA

    2013-01-01

    For the ITER-like wall, the JET mirror link divertor spectroscopy system was redesigned to fully cover the tungsten horizontal strike plate with faster time resolution and improved near-UV performance. Since the ITER-like wall project involves a change in JET from a carbon dominated machine to a beryllium and tungsten dominated machine with residual carbon, the aim of the system is to provide the recycling flux, equivalent, to the impinging deuterium ion flux, the impurity fluxes (C, Be, O) and tungsten sputtering fluxes and hence give information on the tungsten divertor source. In order to do this self-consistently, the system also needs to provide plasma characterization through the deuterium Balmer spectra measurements of electron density and temperature during high density. L-Mode results at the density limit from Stark broadening/line ratio analysis will be presented and compared to Langmuir probe profiles and 2D-tomography of low-n Balmer emission [1]. Comparison with other diagnostics will be vital fo...

  1. Deuterium Balmer/Stark spectroscopy and impurity profiles: First results from mirror-link divertor spectroscopy system on the JET ITER-like wall

    Energy Technology Data Exchange (ETDEWEB)

    Meigs, A.G., E-mail: Andrew.Meigs@ccfe.ac.uk [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Brezinsek, S.; Clever, M.; Huber, A. [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich (Germany); Marsen, S. [Max-Planck-Institut for Plasma Physics, EURATOM Association, Greifswald (Germany); Nicholas, C. [Dept. of Physics, University of Strathclyde, Glasgow G4 0NG (United Kingdom); Stamp, M.; Zastrow, K.-D. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom)

    2013-07-15

    For the ITER-like wall, the JET mirror link divertor spectroscopy system was redesigned to fully cover the tungsten horizontal strike plate with faster time resolution and improved near-UV performance. Since the ITER-like wall project involves a change in JET from a carbon dominated machine to a beryllium and tungsten machine with residual carbon, the aim of the system is to provide the recycling flux, equivalent to the impinging deuterium ion flux, the impurity fluxes (C, Be, O) and tungsten sputtering fluxes and hence give information on the tungsten divertor source. In order to do this self-consistently, the system provides plasma characterization through the deuterium Balmer spectra measurements of electron density and temperature during high density. L-mode results at the density limit from Stark broadening/line ratio analysis will be presented and compared to Langmuir probe profiles and 2D-tomography of low-n Balmer emission [1]. Comparison with other diagnostics will be vital for modeling attempts with the EDGE2D-EIRENE code [2] as the best possible data sets need to be provided to study detachment.

  2. Deuterium Balmer/Stark spectroscopy and impurity profiles: First results from mirror-link divertor spectroscopy system on the JET ITER-like wall

    Science.gov (United States)

    Meigs, A. G.; Brezinsek, S.; Clever, M.; Huber, A.; Marsen, S.; Nicholas, C.; Stamp, M.; Zastrow, K.-D.; JET EFDA Contributors

    2013-07-01

    For the ITER-like wall, the JET mirror link divertor spectroscopy system was redesigned to fully cover the tungsten horizontal strike plate with faster time resolution and improved near-UV performance. Since the ITER-like wall project involves a change in JET from a carbon dominated machine to a beryllium and tungsten machine with residual carbon, the aim of the system is to provide the recycling flux, equivalent to the impinging deuterium ion flux, the impurity fluxes (C, Be, O) and tungsten sputtering fluxes and hence give information on the tungsten divertor source. In order to do this self-consistently, the system provides plasma characterization through the deuterium Balmer spectra measurements of electron density and temperature during high density. L-mode results at the density limit from Stark broadening/line ratio analysis will be presented and compared to Langmuir probe profiles and 2D-tomography of low-n Balmer emission [1]. Comparison with other diagnostics will be vital for modeling attempts with the EDGE2D-EIRENE code [2] as the best possible data sets need to be provided to study detachment.

  3. 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.

  4. Design of the Wendelstein 7-X inertially cooled Test Divertor Unit Scraper Element

    Energy Technology Data Exchange (ETDEWEB)

    Lumsdaine, Arnold, E-mail: lumsdainea@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Boscary, Jean [Max Planck Institute for Plasma Physics, Garching (Germany); Fellinger, Joris [Max Planck Institute for Plasma Physics, Greifswald (Germany); Harris, Jeff [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Hölbe, Hauke; König, Ralf [Max Planck Institute for Plasma Physics, Greifswald (Germany); Lore, Jeremy; McGinnis, Dean [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Neilson, Hutch; Titus, Peter [Princeton Plasma Physics Lab, Princeton, NJ (United States); Tretter, Jörg [Max Planck Institute for Plasma Physics, Garching (Germany)

    2015-10-15

    Highlights: • The justification for the installation of the Test Divertor Unit Scraper Element is given. • Specially designed operational scenarios for the component are presented. • Plans for the design of the component are detailed. - Abstract: The Wendelstein 7-X stellarator is scheduled to begin operation in 2015, and to achieve full power steady-state operation in 2019. Computational simulations have indicated that for certain plasma configurations in the steady-state operation, the ends of the divertor targets may receive heat fluxes beyond their qualified technological limit. To address this issue, a high heat-flux “scraper element” (HHF-SE) has been designed that can protect the sensitive divertor target region. The surface profile of the HHF-SE has been carefully designed to meet challenging engineering requirements and severe spatial limitations through an iterative process involving physics simulations, engineering analysis, and computer aided design rendering. The desire to examine how the scraper element interacts with the plasma, both in terms of how it protects the divertor, and how it affects the neutral pumping efficiency, has led to the consideration of installing an inertially cooled version during the short pulse operation phase. This Test Divertor Unit Scraper Element (TDU-SE) would replicate the surface profile of the HHF-SE. The design and instrumentation of this component must be completed carefully in order to satisfy the requirements of the machine operation, as well as to support the possible installation of the HHF-SE for steady-state operation.

  5. Intermittent Divertor Filaments in the National Spherical Torus Experiment and Their Relation to Midplane Blobs

    Energy Technology Data Exchange (ETDEWEB)

    R.J. Maqueda, D.P. Stotler and the NSTX Team.

    2010-05-19

    While intermittent filamentary structures, also known as blobs, are routinely seen in the low-field-side scrape-off layer of the National Spherical Torus Experiment (NSTX) (Ono et al 2000 Nucl. Fusion 40 557), fine structured filaments are also seen on the lower divertor target plates of NSTX. These filaments, not associated with edge localized modes, correspond to the interaction of the turbulent blobs seen near the midplane with the divertor plasma facing components. The fluctuation level of the neutral lithium light observed at the divertor, and the skewness and kurtosis of its probability distribution function, is similar to that of midplane blobs seen in Dα; e.g. increasing with increasing radii outside the outer strike point (OSP) (separatrix). In addition, their toroidal and radial movement agrees with the typical movement of midplane blobs. Furthermore, with the appropriate magnetic topology, i.e. mapping between the portion of the target plates being observed into the field of view of the midplane gas puff imaging diagnostic, very good correlation is observed between the blobs and the divertor filaments. The correlation between divertor plate filaments and midplane blobs is lost close to the OSP. This latter observation is consistent with the existence of ‘magnetic shear disconnection’ due to the lower X-point, as proposed by Cohen and Ryutov (1997 Nucl. Fusion 37 621).

  6. Investigation of power spreading in a tokamak divertor using numerical tools

    Energy Technology Data Exchange (ETDEWEB)

    Hoppe, Felix; Scarabosio, Andrea; Wischmeier, Marco [Max-Planck Institut fuer Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Collaboration: ASDEX Upgrade Team

    2013-07-01

    Divertors are widely used in today's fusion devices in order to reduce plasma core impurities and improve energy confinement. As the divertor targets are exposed to the largest part of the particle and heat loads reaching the wall, these loads must be reduced to prevent material damage. An enhancement of the plasma-wetted area on the targets is one approach. In low density plasmas, the plasma-wetted area is mainly given by the width of the scrap-off-layer (SOL) plasma at the divertor entrance, modified by heat diffusion into the private flux region (PFR) in the divertor. The heat diffusion broadens the heat flux profile at the targets. This can be approximated by a convolution of the upstream profile with a Gaussian of width S. The SOLPS5.0 code package is used to study the influence of divertor geometry and neutral pressure on S. The code is then validated by comparing the numerical results to the experimental findings in the ASDEX Upgrade tokamak.

  7. ADX: A high Power Density, Advanced RF-Driven Divertor Test Tokamak for PMI studies

    Science.gov (United States)

    Whyte, Dennis; ADX Team

    2015-11-01

    The MIT PSFC and collaborators are proposing an advanced divertor experiment, ADX; a divertor test tokamak dedicated to address critical gaps in plasma-material interactions (PMI) science, and the world fusion research program, on the pathway to FNSF/DEMO. Basic ADX design features are motivated and discussed. In order to assess the widest range of advanced divertor concepts, a large fraction (>50%) of the toroidal field volume is purpose-built with innovative magnetic topology control and flexibility for assessing different surfaces, including liquids. ADX features high B-field (>6 Tesla) and high global power density (P/S ~ 1.5 MW/m2) in order to access the full range of parallel heat flux and divertor plasma pressures foreseen for reactors, while simultaneously assessing the effect of highly dissipative divertors on core plasma/pedestal. Various options for efficiently achieving high field are being assessed including the use of Alcator technology (cryogenic cooled copper) and high-temperature superconductors. The experimental platform would also explore advanced lower hybrid current drive and ion-cyclotron range of frequency actuators located at the high-field side; a location which is predicted to greatly reduce the PMI effects on the launcher while minimally perturbing the core plasma. The synergistic effects of high-field launchers with high total B on current and flow drive can thus be studied in reactor-relevant boundary plasmas.

  8. Divertor ExB and Parallel Flows on the DIII-D Tokamak

    Science.gov (United States)

    Boedo, J.; Rudakov, D.

    2016-10-01

    E ×B convection is an important particle transport mechanism responsible for up to 50 % of the total particle flux into the divertor, changing direction with B, and playing a role in divertor asymmetries. The gradient of the plasma potential, Vp =Vf + 2.5Te , reaches 5 kV/m across the SOL-private boundary, causing a poloidal particle flux, calculated as, Γθ = 2 πRne (Vp 1 -Vp 2) /BT , (along flux surfaces) of about 1022 s-1 , comparable to the target flow of 2 ×1022 s-1 , and consistent with previous work. Floating potential Vf, temperature Te, density Ne, and D+ flow were measured in the DIII-D divertor. The data will be compared to simulations by SOLPS and UEDGE. The D+ parallel flow velocity, V ∥ , calculated by multiplying the Mach number by the local sound speed cs =(γ ZkTe /mi) 1 / 2 show increasing velocity towards the plate in attached conditions and bulk sonic flows over the whole detached region in detached conditions. We compare measurements in the divertor to similar measurements made at the midplane to show how divertor conditions reflect upstream. Supported under USDOE Grant DE-FC02-04ER54698.

  9. Numerical analyses of JT-60SA tokamak with tungsten divertor by COREDIV code

    Science.gov (United States)

    Gałązka, K.; Ivanova-Stanik, I.; Stępniewski, W.; Zagórski, R.; Neu, R.; Romanelli, M.; Nakano, T.

    2017-04-01

    An analysis of radiative power exhaust for the JT-60SA tokamak with a tungsten divertor is performed with the help of the self-consistent, core-edge integrated COREDIV code. Two scenarios of operation (low and high density) were investigated in the scope of different parameters (electron density at the separatrix and the perpendicular transport in the scrape-off layer) with impurity seeding (Ne and Kr). The calculations show that in the case of the tungsten divertor the power load to the divertor plate is mitigated and the central plasma dilution is smaller compared to the carbon divertor. In the most cases the energy flux through the separatrix is above the L–H transition threshold. For the high density case with neon seeding operation in full detachment mode is observed. Changing the diffusion coefficient in the SOL has a strong influence on the result of the calculations as increased radial transport causes stronger screening effect. Also by changing the electron density on the separatrix the influx of heavy impurities (W, Kr) into the core region can be reduced. The results demonstrate that it is easier to achieve sustainable conditions in the divertor region for the high density scenario, whereas for the low density one reducing the auxiliary heating power seems unavoidable to prevent damaging of the target plate, even for strong seeding gas influx.

  10. Scrape-off layer ion temperature measurements at the divertor target in MAST by retarding field energy analyser

    CERN Document Server

    Elmore, S; Kirk, A; Thornton, A J; Harrison, J R; Tamain, P; Kocan, M; Bradley, J W

    2013-01-01

    Knowledge of the ion temperature (Ti) is of key importance for determining heat fluxes to the divertor and plasma facing components, however data regarding this is limited compared to electron temperature (Te) data. Ti measurements at the divertor target, between edge-localised modes (inter-ELM) H-mode, have been made using a novel retarding field energy analyser (RFEA).

  11. Mobile Probing and Probes

    DEFF Research Database (Denmark)

    2013-01-01

    to mobile probing being a flexible method for uncovering the unknowns, as a way of getting rich data to the analysis and design phases. On the other hand it is difficult to engage users to give in depth explanations, which seem easier in synchronous dialogs (whether online or face2face). The development...

  12. Numerical simulation study on density dependence of plasma detachment in simulated gas divertor experiments of the TPD-I device

    Energy Technology Data Exchange (ETDEWEB)

    Ohno, N. [Nagoya Univ. (Japan). Dept. of Energy Eng. and Sci.; Mori, S. [Nagoya Univ. (Japan). Dept. of Energy Eng. and Sci.; Ezumi, N. [Nagoya Univ. (Japan). Dept. of Energy Eng. and Sci.; Takagi, M. [Nagoya Univ. (Japan). Dept. of Energy Eng. and Sci.; Takamura, S. [Nagoya Univ. (Japan). Dept. of Energy Eng. and Sci.; Suzuki, H. [National Inst. for Fusion Science, Nagoya (Japan)

    1996-08-01

    It is one of the most critical requirements to reduce the heat load to the divertor plate in the next generation fusion devices such as ITER, intended to have a long pulse or a steady state operation. Dynamic gas target divertor as well as high recycling divertor is one of the most important candidates for ITER. Recently a detached plasma has been observed in experimental fusion devices. Knowledge of the basic physics of the plasma detachment is required for any application of the gas target and high recycling divertor to the next generation experimental reactors. Linear plasma divertor simulators with high heat flux plasmas are used to investigate the plasma detachment because its good accessibility for comprehensive measurements and simple geometry leads a deeper understanding of the plasma detachment by comparing between simulation predictions and the experimental results. (orig.)

  13. Studies of Plasma Detachment Using a One Dimensional Model for Divertor Operation

    CERN Document Server

    Vesey, R A; Bateman, G

    1995-01-01

    To characterize the conditions required to reach advanced divertor regimes, a one-dimensional computational model has been developed based on a coordinate transformation to incorporate two-dimensional effects. This model includes transport of ions, two species each of atoms and molecules, momentum, and ion and electron energy both within and across the flux surfaces. Impurity radiation is calculated using a coronal equilibrium model which includes the effects of charge-exchange recombination. Numerical results indicate that impurity radiation acts to facilitate plasma detachment and enhances the power lost from the divertor channel in escaping neutral atoms by cooling the electrons and suppressing ionization. As divertor particle densities increase, cold and thermal molecules become increasingly important in cooling the plasma, with molecular densities dominating electron and atomic densities under some conditions.

  14. Towards the development of workable acceptance criteria for the divertor CFC monoblock armour

    Energy Technology Data Exchange (ETDEWEB)

    D' Agata, E. [ITER International Team, ITER Joint Work Site, Boltzmannstr. 2, D-85748 Garching (Germany)]. E-mail: dagatae@itereu.de; Tivey, R. [ITER International Team, ITER Joint Work Site, Boltzmannstr. 2, D-85748 Garching (Germany)

    2005-11-15

    The plasma-facing components (PFCs) of the ITER divertor will be subjected to high heat flux (HHF). Carbon-fibre composite (CFC) is selected as the armour for the region of highest heat flux where the scrape-off layer of the plasma intercepts the vertical targets (VT). Failure of the armour to heat sink joints will compromise the performance of the divertor and could ultimately result in its failure and the shut down of the ITER machine. There are tens of thousands of CFCs to CuCrZr joints. The aim of the PFC design is to ensure that the divertor can continue to function even with the failure of a few joints. In preparation for writing the procurement specification for the ITER vertical target PFCs, a programme of work is underway with the objective of defining workable acceptance criteria for the PFC armour joints.

  15. Definition of acceptance criteria for the ITER divertor plasma-facing components through systematic experimental analysis

    Energy Technology Data Exchange (ETDEWEB)

    Escourbiac, F; Richou, M; Guigon, R; Durocher, A; Schlosser, J; Grosman, A [CEA/IRFM, F-13108, Saint-Paul-lez-Durance (France); Constans, S [AREVA-NP, Le Creusot (France); Merola, M [ITER Organization, Cadarache (France); Riccardi, B [Fusion For Energy, Barcelona (Spain)], E-mail: frederic.escourbiac@cea.fr

    2009-12-15

    Experience has shown that a critical part of the high-heat flux (HHF) plasma-facing component (PFC) is the armour to heat sink bond. An experimental study was performed in order to define acceptance criteria with regards to thermal hydraulics and fatigue performance of the International Thermonuclear Experimental Reactor (ITER) divertor PFCs. This study, which includes the manufacturing of samples with calibrated artificial defects relevant to the divertor design, is reported in this paper. In particular, it was concluded that defects detectable with non-destructive examination (NDE) techniques appeared to be acceptable during HHF experiments relevant to heat fluxes expected in the ITER divertor. On the basis of these results, a set of acceptance criteria was proposed and applied to the European vertical target medium-size qualification prototype: 98% of the inspected carbon fibre composite (CFC) monoblocks and 100% of tungsten (W) monoblock and flat tiles elements (i.e. 80% of the full units) were declared acceptable.

  16. Divertor with a third-order null of the poloidal field

    Energy Technology Data Exchange (ETDEWEB)

    Ryutov, D. D.; Umansky, M. V. [Lawrence Livermore National Laboratory, Livermore, California 94551 (United States)

    2013-09-15

    A concept and preliminary feasibility analysis of a divertor with the third-order poloidal field null is presented. The third-order null is the point where not only the field itself but also its first and second spatial derivatives are zero. In this case, the separatrix near the null-point has eight branches, and the number of strike-points increases from 2 (as in the standard divertor) to six. It is shown that this magnetic configuration can be created by a proper adjustment of the currents in a set of three divertor coils. If the currents are somewhat different from the required values, the configuration becomes that of three closely spaced first-order nulls. Analytic approach, suitable for a quick orientation in the problem, is used. Potential advantages and disadvantages of this configuration are briefly discussed.

  17. Analysis of FAST snowflake divertor by EDGE2D/EIRENE

    Energy Technology Data Exchange (ETDEWEB)

    Viola, B., E-mail: bruno.viola@enea.it [ENEA Unità Tecnica Fusione, C.R. Frascati, Via E. Fermi 45, 00044 Frascati, Rome (Italy); Pericoli Ridolfini, V. [Consorzio CREATE, Università di Napoli Federico II, Via Claudio 21, 80125 Napoli (Italy); Visona, N. [Consorzio RFX, C.so Stati Uniti 4, Padova 35127 (Italy); Corrigan, G.; Harting, D. [Culham Centre of Fusion Energy, OX14 3DB Abingdon (United Kingdom); Maddaluno, G. [ENEA Unità Tecnica Fusione, C.R. Frascati, Via E. Fermi 45, 00044 Frascati, Rome (Italy); Zagórski, R. [Institute of Plasma Physics and Laser Microfusion, 01-497 Warsaw (Poland)

    2015-08-15

    The snowflake [1,2] divertor is a proposal for solving the heat and particle exhaust problem in fusion grade plasmas. Turning the X-point into a second order null gives the possibility of radially expanding the poloidal flux in the divertor region much more than in a SD, increasing the connection length, redistributing the power load on a larger area and enhancing radiative losses. Since the efforts associated to the design of reactor-relevant configurations, like the snowflake, are large, ENEA is studying this configuration using efficient and flexible numerical tools to design and optimise tokamak equilibrium configurations. Such studies are applied to the Divertor Test Tokamak FAST, a satellite tokamak proposed for the European roadmap towards fusion.

  18. Design and optimization of W/Cu divertor mock-ups

    Institute of Scientific and Technical Information of China (English)

    Qiong Li; Weiping Shen

    2007-01-01

    Tungsten is a promising candidate for plasma-facing materials to cover the surface of the divertor plate in the design of an international thermonuclear experimental reactor (ITER). Copper as a heat sink material serves to transfer heat excellently. Divertor mock-ups with W/Cu graded interlayers were designed to reduce thermal stresses. Thermally induced stresses and temperature in a W/Cu divertor mock-up were analyzed using the finite element method. The graded structures with different exponents p and thicknesses were designed and discussed. The conclusions drawn from these analyses are that thermal stresses reach the minimum and the temperature is suitable when exponent p is 1.5 and the thickness of five graded interlayers is 5 mm.

  19. Fast Identification of Recycling Properties of Wall-Released Hydrogenic Neutrals in Divertor Plasma

    Institute of Scientific and Technical Information of China (English)

    LI Chengyue; DENG Baiquan; YAN Jiancheng; G. A. EMMERT

    2007-01-01

    A new bipartition neutral transport model was developed for quick identification of the recycling properties of the wall-released hydrogenic neutral particles in the vicinity of the divertor target plate. Based on this model, the numerical calculation results are fairly consistent with the results obtained with the 'multi-generation method'. This model can not only be utilized to provide a source term from neutral transport calculations for the B2 edge plasma transport code, which has been used to simulate edge plasma transport of an HL-2A divertor configuration, but can also be specifically applied for fast classification of the divertor plasma as high recycling or low recycling. Our results also show that the transmissivity is lower in the high-recycling regime.

  20. New bipartition model of neutral particle transport in the HL-2A divertor region

    Institute of Scientific and Technical Information of China (English)

    DENG Bai-quan; YAN Jian-cheng; PENG Li-lin

    2005-01-01

    A new bipartition neutral transport model has been developed for simulation of the hydrogenic neutral particle transport in the vicinity of HL-2A divertor target plate. The numerical calculation results on the basis of this model are fairly consistent with the results obtained with the "multi-generation method". One possible application of this model is to provide a source term originating from neutral transport calculation for any other edge plasma transport code, for instance, B-2 code, which has been used to simulate edge plasma transport of the HL-2A divertor configuration. Especially it can be utilized to quickly classify the plasma in divertor region as high or low recycling regime.

  1. Calculation of Divertor Thermal Response as a Function of Material Composition for NSTX

    Science.gov (United States)

    Chaffin, Michael; Maingi, Rajesh

    2007-11-01

    Present tokamak designs use a magnetic divertor to deposit heat from the edge plasma onto Plasma Facing Components (PFCs) designed to remove the heat. Studying how this heat is distributed under various discharge conditions gives insight into how heat deposition can be optimized, and how different materials respond to plasma heating. In the National Spherical Torus eXperiment (NSTX), infrared cameras are used to measure divertor surface temperature, from which heat flux is computed using a 1D semi-infinite slab model with constant thermal conductivity. Here, a 1D simulation of the PFCs incorporating temperature-dependent thermal properties is used to compute heat flux profiles resolved across time and tile thickness. The PFC response to a given heat flux is also computed, and comparisons of resulting temperature profiles are made for a variety of materials including ATJ graphite (presently in the NSTX divertor), pyrolytic graphite, molybdenum, and tungsten.

  2. Collagenolytic Matrix Metalloproteinase Activities toward Peptomeric Triple-Helical Substrates.

    Science.gov (United States)

    Stawikowski, Maciej J; Stawikowska, Roma; Fields, Gregg B

    2015-05-19

    Although collagenolytic matrix metalloproteinases (MMPs) possess common domain organizations, there are subtle differences in their processing of collagenous triple-helical substrates. In this study, we have incorporated peptoid residues into collagen model triple-helical peptides and examined MMP activities toward these peptomeric chimeras. Several different peptoid residues were incorporated into triple-helical substrates at subsites P3, P1, P1', and P10' individually or in combination, and the effects of the peptoid residues were evaluated on the activities of full-length MMP-1, MMP-8, MMP-13, and MMP-14/MT1-MMP. Most peptomers showed little discrimination between MMPs. However, a peptomer containing N-methyl Gly (sarcosine) in the P1' subsite and N-isobutyl Gly (NLeu) in the P10' subsite was hydrolyzed efficiently only by MMP-13 [nomenclature relative to the α1(I)772-786 sequence]. Cleavage site analysis showed hydrolysis at the Gly-Gln bond, indicating a shifted binding of the triple helix compared to the parent sequence. Favorable hydrolysis by MMP-13 was not due to sequence specificity or instability of the substrate triple helix but rather was based on the specific interactions of the P7' peptoid residue with the MMP-13 hemopexin-like domain. A fluorescence resonance energy transfer triple-helical peptomer was constructed and found to be readily processed by MMP-13, not cleaved by MMP-1 and MMP-8, and weakly hydrolyzed by MT1-MMP. The influence of the triple-helical structure containing peptoid residues on the interaction between MMP subsites and individual substrate residues may provide additional information about the mechanism of collagenolysis, the understanding of collagen specificity, and the design of selective MMP probes.

  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. 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.

  5. A multichannel visible spectroscopy system for the ITER-like W divertor on EAST

    Science.gov (United States)

    Mao, Hongmin; Ding, Fang; Luo, Guang-Nan; Hu, Zhenhua; Chen, Xiahua; Xu, Feng; Yang, Zhongshi; Chen, Jingbo; Wang, Liang; Ding, Rui; Zhang, Ling; Gao, Wei; Xu, Jichan; Wu, Chengrui

    2017-04-01

    To facilitate long-pulse high power operation, an ITER-like actively cooled tungsten (W) divertor was installed in Experimental Advanced Superconducting Tokamak (EAST) to replace the original upper graphite divertor in 2014. A dedicated multichannel visible spectroscopic diagnostic system has been accordingly developed for the characterization of the plasma and impurities in the W divertor. An array of 22 lines-of-sight (LOSs) provides a profile measurement of the light emitted from the plasma along upper outer divertor, and the other 17 vertical LOSs view the upper inner divertor, achieving a 13 mm poloidal resolution in both regions. The light emitted from the plasma is collected by a specially designed optical lens assembly and then transferred to a Czerny-Turner spectrometer via 40 m quartz fibers. At the end, the spectra dispersed by the spectrometer are recorded with an Electron-Multiplying Charge Coupled Device (EMCCD). The optical throughput and quantum efficiency of the system are optimized in the wavelength range 350-700 nm. The spectral resolution/coverage can be adjusted from 0.01 nm/3 nm to 0.41 nm/140 nm by switching the grating with suitable groove density. The frame rate depends on the setting of LOS number in EMCCD and can reach nearly 2 kHz for single LOS detection. The light collected by the front optical lens can also be divided and partly transferred to a photomultiplier tube array with specified bandpass filter, which can provide faster sampling rates by up to 200 kHz. The spectroscopic diagnostic is routinely operated in EAST discharges with absolute optical calibrations applied before and after each campaign, monitoring photon fluxes from impurities and H recycling in the upper divertor. This paper presents the technical details of the diagnostic and typical measurements during EAST discharges.

  6. 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.

  7. Changes in divertor conditions in response to changing core density with RMPs

    Science.gov (United States)

    Briesemeister, A. R.; Ahn, J.-W.; Canik, J. M.; Fenstermacher, M. E.; Frerichs, H.; Lasnier, C. J.; Lore, J. D.; Leonard, A. W.; Makowski, M. A.; McLean, A. G.; Meyer, W. H.; Schmitz, O.; Shafer, M. W.; Unterberg, E. A.; Wang, H. Q.; Watkins, J. G.

    2017-07-01

    The effects of changes in core density on divertor electron temperature, density and heat flux when resonant magnetic perturbations (RMPs) are applied are presented, notably a reduction in RMP induced secondary radial peaks in the electron temperature profile at the target plate is observed when the core density is increased, which is consistent with modeling. RMPs is used here to indicate non-axisymmetric magnetic field perturbations, created using in-vessel control coils, which have at least one but typically many resonances with the rotational transform of the plasma (Evans et al 2006 Phys. Plasmas 13 056121). RMPs are found to alter inter-ELM heat flux to the divertor by modifying the core plasma density. It is shown that applying RMPs reduces the core density and increases the inter-ELM heat flux to both the inner and outer targets. Using gas puffing to return the core density to the pre-RMP levels more than eliminates the increase in inter-ELM heat flux, but a broadening of the heat flux to the outer target remains. These measurements were made at a single toroidal location, but the peak in the heat flux profile was found near the outer strike point where simulations indicate little toroidal variation should exist and tangentially viewing diagnostics showed no evidence of strong asymmetries. In experiments where divertor Thomson scattering measurements were available it is shown that local secondary peaks in the divertor electron temperature profile near the target plate are reduced as the core density is increased, while peaks in the divertor electron density profile near the target are increased. These trends observed in the divertor electron temperature and density are qualitatively reproduced by scanning the upstream density in EMC3-Eirene modeling. Measurements are presented showing that higher densities are needed to induce detachment of the outer strike point in a case where an increase in electron temperature, likely due to a change in MHD activity

  8. Status of design and experimental activity on module of lithium divertor for KTM tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lyublinski, Igor E., E-mail: lyublinski@yandex.ru [JSC “Red Star”, Moscow (Russian Federation); Vertkov, Alexey V.; Zharkov, Mikhail Yu.; Semenov, Vladimir V. [JSC “Red Star”, Moscow (Russian Federation); Mirnov, Sergey V.; Lazarev, Vladimir B. [GSC RF TRINITI, Troitsk, Moscow Region (Russian Federation); Tazhibayeva, Irina L.; Shapovalov, Gennadiy V.; Kulsartov, Timur V.; D’yachenko, Alexandr V. [IAE of National Nuclear Center, Kurchatov (Kazakhstan); Mazzitelli, Giuseppe [Associazione EURATOM-ENEA sulla Fusione, C.R. ENEA Frascati, Rome (Italy); Agostini, Pietro [ENEA Brasimone, Camugnano, BO (Italy)

    2013-10-15

    Highlights: • Lithium divertor module based on capillary-porous system is created for KTM tokamak. • The hydraulic tests of lithium divertor module were conducted. • The results were compared with the calculation data. • The analysis of results’ discrepancies was conducted. • The lithium divertor module is ready for tests on KTM tokamak. -- Abstract: The projects of ITER and DEMO reactors showed that there are serious difficulties with solving the issues of plasma facing elements (PFE) based on the solid materials. Problems of PFE can be overcome by the use of liquid lithium. Application of lithium will allow to create a self-renewal and MHD stable liquid metal surface of the in-vessel devices possessing practically unlimited service life. Realization of these advantages is based on use of so-called lithium capillary-porous system (CPS) – new material, in which liquid lithium fills a solid matrix from porous material. The progress in development of lithium technology and also lithium experiments in the tokamaks TFTR, T-11M, T-10, FTU, NSTX, LTX, HT-7 and stellarator TJ II is a good basis for development of the project of steady-state operating lithium divertor module for Kazakhstan tokamak. At present the lithium divertor module for KTM tokamak is development and manufacturing. The paper describes main design features of the module of lithium divertor (MLD). The first step of the hydraulic tests of MLD with fully assembled external thermo-stabilization system, which was connected to in-vessel lithium unit, were performed using ethanol as a model heat transfer media. Test results of MLD have shown that operating parameters of designed and manufactured system for thermo-stabilization are sufficient for proper operation; basic hydraulic characteristics of the system are close to expected values.

  9. 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.

  10. Optimization of tungsten castellated structures for the ITER divertor

    Science.gov (United States)

    Litnovsky, A.; Hellwig, M.; Matveev, D.; Komm, M.; van den Berg, M.; De Temmerman, G.; Rudakov, D.; Ding, F.; Luo, G.-N.; Krieger, K.; Sugiyama, K.; Pitts, R. A.; Petersson, P.

    2015-08-01

    In ITER, the plasma-facing components (PFCs) of the first wall and the divertor armor will be castellated to improve their thermo-mechanical stability and to limit forces due to induced currents. The fuel accumulation in the gaps may significantly contribute to the in-vessel fuel inventory. Castellation shaping may be the most straightforward way to minimize the fuel inventory and to alleviate the thermal loads onto castellations. A new castellation shape was proposed and comparative modeling of conventional (rectangular) and shaped castellation was performed for ITER conditions. Shaped castellation was predicted to be capable to operate under stationary heat load of 20 MW/m2. An 11-fold decrease of beryllium (Be) content in the gaps of the shaped cells alone with a 7-fold decrease of carbon content was predicted. In order to validate the predictive capabilities of modeling tools used for ITER conditions, the dedicated modeling with the same codes was made for existing tokamaks and benchmarked with the results of multi-machine experiments. For the castellations exposed in TEXTOR and DIII-D, the carbon amount in the gaps of shaped cells was 1.9-2.3 times smaller than that of rectangular ones. Modeling for TEXTOR conditions yielded to 1.5-fold decrease of carbon content in the gaps of shaped castellation outlining fair agreement with the experiment. At the same time, a number of processes, like enhanced erosion of molten layer yet need to be implemented in the codes in order to increase the accuracy of predictions for ITER.

  11. Triple-resonant transducers.

    Science.gov (United States)

    Butler, Stephen C

    2012-06-01

    A detailed analysis is presented of two novel multiple-resonant transducers which produce a wider transmit response than that of a conventional Tonpilz-type transducer. These multi-resonant transducers are Tonpilz-type longitudinal vibrators that produce three coupled resonances and are referred to as triple-resonant transducers (TRTs). One of these designs is a mechanical series arrangement of a tail mass, piezoelectric ceramic stack, central mass, compliant spring, second central mass, second compliant spring, and a piston-radiating head mass. The other TRT design is a mechanical series arrangement of a tail mass, piezoelectric ceramic stack, central mass, compliant spring, and head mass with a quarter-wave matching layer of poly(methyl methacrylate) on the head mass. Several prototype transducer element designs were fabricated that demonstrated proof-of-concept.

  12. Achievements and challenges in automated parameter, shape and topology optimization for divertor design

    Science.gov (United States)

    Baelmans, M.; Blommaert, M.; Dekeyser, W.; Van Oevelen, T.

    2017-03-01

    Plasma edge transport codes play a key role in the design of future divertor concepts. Their long simulation times in combination with a large number of control parameters turn the design into a challenging task. In aerodynamics and structural mechanics, adjoint-based optimization techniques have proven successful to tackle similar design challenges. This paper provides an overview of achievements and remaining challenges with these techniques for complex divertor design. It is shown how these developments pave the way for fast sensitivity analysis and improved design from different perspectives.

  13. 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

  14. Spectroscopic measurements of impurity temperatures and parallel ion flows in the DIII-D divertor

    Energy Technology Data Exchange (ETDEWEB)

    Isler, R.C. [Oak Ridge National Lab., TN (United States); Brooks, N.H.; West, W.P.; Leonard, A.W. [General Atomics, San Diego, CA (United States); McKee, G.R. [Univ. of Wisconsin, Madison, WI (United States); Porter, G.D. [Lawrence Livermore National Lab., CA (United States)

    1998-06-01

    Impurity ion temperatures and parallel flow velocities in the DIII-D divertor have been measured from the shapes and shifts of visible spectral lines of C II, C III, and B II. Spectral multiplet patterns are analyzed by fitting them to theoretical profiles that incorporate exact calculations for the Zeeman/Paschen-Back effect. Ion temperatures range from 4--20 eV. Both normal flows toward the target plate and reversed flows away from the target plate are observed in the outer divertor leg; only flows toward the plate are detected in the inner leg.

  15. Spectroscopic measurements of impurity temperatures and parallel ion flows in the DIII-D divertor

    Science.gov (United States)

    Isler, R. C.; Brooks, N. H.; West, W. P.; Leonard, A. W.; McKee, G. R.; Porter, G. D.

    Impurity ion temperatures and parallel flow velocities in the DIII-D divertor have been measured from the shapes and shifts of visible spectral lines of C II, C III, and B II. Spectral multiplet patterns are analyzed by fitting them to theoretical profiles that incorporate exact calculations for the Zeeman/Paschen-Back effect. Both normal flows toward the target plate and reversed flows away from the target plate are observed in the outer divertor leg; only flows toward the plate are detected in the inner leg.

  16. Development of divertor tungsten coatings for the JET ITER-like wall

    Science.gov (United States)

    Matthews, G. F.; Coad, P.; Greuner, H.; Hill, M.; Hirai, T.; Likonen, J.; Maier, H.; Mayer, M.; Neu, R.; Philipps, V.; Pitts, R.; Riccardo, V.; JET EFDA Contributors

    2009-06-01

    The main objectives of the JET ITER-like Wall Project are to provide a beryllium main wall and tungsten divertor with at least a 4 year lifetime to allow full evaluation of the materials and related plasma scenarios for ITER. Tungsten coatings will be used over most of the divertor area and this paper describes the latest developments in the coating technology and an analysis of the implications for the coating lifetime and machine operation. Both steady state and transient heat loads are assessed.

  17. Harmonic analysis on triple spaces

    DEFF Research Database (Denmark)

    Danielsen, Thomas Hjortgaard

    In this thesis we study examples of triple spaces, both their structure theory, their invariant differential operators as well as analysis on them. The first major results provide us with some examples of triple spaces which are strongly spherical, i.e. satisfy some conditions reminiscent...... of properties of symmetric spaces. The algebras of invariant differential operators for these spaces are studied and the conclusion is that most of them are non-commutative. Finally, we restrict our attention to a single triple space, giving a specific polar decomposition and corresponding integration formula......, and studying the relations between open orbits of parabolic subgroups, multiplicities and distribution vectors....

  18. Nanostructural evolution of Cr-rich precipitates in a Cu-Cr-Zr alloy during heat treatment studied by 3 dimensional atom probe

    DEFF Research Database (Denmark)

    Hatakeyama, Masahiko; Toyama, Takeshi; Nagai, Yasuyoshi

    2008-01-01

    Nanostructural evolution of Cr (Cr-rich) precipitates in a Cu-0.78%Cr-0.13%Zr alloy has been studied after aging and overaging (reaging) by laser assisted local electrode 3 dimensional atom probe (Laser-LEAP). This material is a candidate for the first wall and divertor components of future fusion...

  19. 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)

  20. Evidence for the importance of radial transport in plasma detachment in the Nagoya University Divertor Simulator (NAGDIS-II)

    Energy Technology Data Exchange (ETDEWEB)

    Hollmann, E. M.; Whyte, D. G.; Nishijima, D.; Ohno, N.; Uesugi, Y.; Ezumi, N.

    2001-07-01

    Measurements of ion and electron temperatures have been performed in detaching helium--hydrogen plasmas in a linear divertor simulator experiment using spectroscopy, a Langmuir probe, and an omegatron mass spectrometer. Detachment in these plasmas is characterized by a significant ({approx}20 x) reduction in the central plasma flux at the target plate as the target region neutral pressure is increased from 2 to 12 mTorr. The data indicate that partially detached gas-target plasmas consist of a hot (T{sub e}{approx}5 eV) core region along the axis of the plasma column, surrounded by a cold (T{sub e}{approx}0.1 eV) halo region of recombining plasma. At T{sub e}=5 eV, plasma recombination is negligible compared with ionization; these experiments therefore provide evidence that detachment is primarily caused by radial transport and by a gradual drop in the ionization source as the temperature of the core region drops below 5 eV.

  1. Monte Carlo simulations of tungsten redeposition at the divertor target

    Science.gov (United States)

    Chankin, A. V.; Coster, D. P.; Dux, R.

    2014-02-01

    Recent modeling of controlled edge-localized modes (ELMs) in ITER with tungsten (W) divertor target plates by the SOLPS code package predicted high electron temperatures (>100 eV) and densities (>1 × 1021 m-3) at the outer target. Under certain scenarios W sputtered during ELMs can penetrate into the core in quantities large enough to cause deterioration of the discharge performance, as was shown by coupled SOLPS5.0/STRAHL/ASTRA runs. The net sputtering yield, however, was expected to be dramatically reduced by the ‘prompt redeposition’ during the first Larmor gyration of W1+ (Fussman et al 1995 Proc. 15th Int. Conf. on Plasma Physics and Controlled Nuclear Fusion Research (Vienna: IAEA) vol 2, p 143). Under high ne/Te conditions at the target during ITER ELMs, prompt redeposition would reduce W sputtering by factor p-2 ˜ 104 (with p ≡ τionωgyro ˜ 0.01). However, this relation does not include the effects of multiple ionizations of sputtered W atoms and the electric field in the magnetic pre-sheath (MPS, or ‘Chodura sheath’) and Debye sheath (DS). Monte Carlo simulations of W redeposition with the inclusion of these effects are described in the paper. It is shown that for p ≪ 1, the inclusion of multiple W ionizations and the electric field in the MPS and DS changes the physics of W redeposition from geometrical effects of circular gyro-orbits hitting the target surface, to mainly energy considerations; the key effect is the electric potential barrier for ions trying to escape into the main plasma. The overwhelming majority of ions are drawn back to the target by a strong attracting electric field. It is also shown that the possibility of a W self-sputtering avalanche by ions circulating in the MPS can be ruled out due to the smallness of the sputtered W neutral energies, which means that they do not penetrate very far into the MPS before ionizing; thus the W ions do not gain a large kinetic energy as they are accelerated back to the surface by the

  2. On Split Lie Triple Systems

    Indian Academy of Sciences (India)

    Antonio J Calderón Martín

    2009-04-01

    We begin the study of arbitrary split Lie triple systems by focussing on those with a coherent 0-root space. We show that any such triple systems with a symmetric root system is of the form $T=\\mathcal{U}+\\sum_j I_j$ with $\\mathcal{U}$ a subspace of the 0-root space $T_0$ and any $I_j$ a well described ideal of , satisfying $[I_j,T,I_k]=0$ if $j≠ k$. Under certain conditions, it is shown that is the direct sum of the family of its minimal ideals, each one being a simple split Lie triple system, and the simplicity of is characterized. The key tool in this job is the notion of connection of roots in the framework of split Lie triple systems.

  3. Toothbrush probe for instantaneous measurement of radial profile in tokamak boundary plasma

    Energy Technology Data Exchange (ETDEWEB)

    Uehara, Kazuya; Sengoku, Seio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Amemiya, Hiroshi

    1997-04-01

    A new probe for the instantaneous measurement of radial profiles of the boundary scrape-off layer (SOL) plasma has been developed in a tokamak. Five asymmetric double-probe chips are aligned in parallel to a strong magnetic field in the boundary plasma in a tokamak. This probe is named the `toothbrush probe` and can measure the ion temperature as well as the electron temperature and the plasma density in the SOL plasma within only one tokamak plasma shot. First, only one asymmetric probe is mounted on the divertor plate and it is tried to determine the ion temperature. Then, a manufactured toothbrush probe is mounted in the SOL plasma and the radial plasma profiles are simultaneously obtained. Data on the e-folding length of the plasma profile obtained by the toothbrush probe can determine the information on the transport properties such as the diffusion coefficient and the thermal conductivity of electrons and ions. (author)

  4. 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...

  5. 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 t

  6. Enhanced -->E*-->B drift effects in the TCV snowflake divertor

    NARCIS (Netherlands)

    G.P. Canal,; Lunt, T.; Reimerdes, H.; Duval, B. P.; Labit, B.; Vijvers, W. A. J.; TCV team,

    2015-01-01

    Measurements of various plasma parameters at the divertor targets of snowflake (SF) and conventional single-null configurations indicate an enhanced effect of the -->E*-->B drift in the scrape-off layer of plasmas in the SF configuration. Plasma boundary transport simulations using the EMC3-Ei

  7. Upstream Density for Plasma Detachment with Conventional and Lithium Vapor-Box Divertors

    Science.gov (United States)

    Goldston, Rj; Schwartz, Ja

    2016-10-01

    Fusion power plants are likely to require detachment of the divertor plasma from material targets. The lithium vapor box divertor is designed to achieve this, while limiting the flux of lithium vapor to the main plasma. We develop a simple model of near-detachment to evaluate the required upstream plasma density, for both conventional and lithium vapor-box divertors, based on particle and dynamic pressure balance between up- and down-stream, at near-detachment conditions. A remarkable general result is found, not just for lithium-induced detachment, that the upstream density divided by the Greenwald-limit density scales as (P 5 / 8 /B 3 / 8) Tdet1 / 2 / (ɛcool + γTdet) , with no explicit size scaling. Tdet is the temperature just before strong pressure loss, 1/2 of the ionization potential of the dominant recycling species, ɛcool is the average plasma energy lost per injected hydrogenic and impurity atom, and γ is the sheath heat transmission factor. A recent 1-D calculation agrees well with this scaling. The implication is that the plasma exhaust problem cannot be solved by increasing R. Instead significant innovation, such as the lithium vapor box divertor, will be required. This work supported by DOE Contract No. DE-AC02-09CH11466.

  8. 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

  9. Melt damage to the JET ITER-like Wall and divertor

    Science.gov (United States)

    Matthews, G. F.; Bazylev, B.; Baron-Wiechec, A.; Coenen, J.; Heinola, K.; Kiptily, V.; Maier, H.; Reux, C.; Riccardo, V.; Rimini, F.; Sergienko, G.; Thompson, V.; Widdowson, A.; Contributors, JET

    2016-02-01

    In October 2014, JET completed a scoping study involving high power scenario development in preparation for DT along with other experiments critical for ITER. These experiments have involved intentional and unintentional melt damage both to bulk beryllium main chamber tiles and to divertor tiles. This paper provides an overview of the findings of concern for machine protection in JET and ITER, illustrating each case with high resolution images taken by remote handling or after removal from the machine. The bulk beryllium upper dump plate tiles and some other protection tiles have been repeatedly flash melted by what we believe to be mainly fast unmitigated disruptions. The flash melting produced in this way is seen at all toroidal locations and the melt layer is driven by j × B forces radially outward and upwards against gravity. In contrast, the melt pools caused while attempting to use MGI to mitigate deliberately generated runaway electron beams are localized to several limiters and the ejected material appears less influenced by j × B forces and shows signs of boiling. In the divertor, transient melting of bulk tungsten by ELMs was studied in support of the ITER divertor material decision using a specially prepared divertor module containing an exposed edge. Removal of the module from the machine in 2015 has provided improved imaging of the melt and this confirms that the melt layers are driven by ELMs. No other melt damage to the other 9215 bulk tungsten lamellas has yet been observed.

  10. Lesson from Tungsten Leading Edge Heat Load Analysis in KSTAR Divertor

    Science.gov (United States)

    Hong, Suk-Ho; Pitts, Richard Anthony; Lee, Hyeong-Ho; Bang, Eunnam; Kang, Chan-Soo; Kim, Kyung-Min; Kim, Hong-Tack; ITER Organization Collaboration; Kstar Team Team

    2016-10-01

    An important design issue for the ITER tungsten (W) divertor and in fact for all such components using metallic plasma-facing elements and which are exposed to high parallel power fluxes, is the question of surface shaping to avoid melting of leading edges. We have fabricated a series of tungsten blocks with a variety of leading edge heights (0.3, 0.6, 1.0, and 2.0 mm), from the ITER worst case to heights even beyond the extreme value tested on JET. They are mounted into adjacent, inertially cooled graphite tile installed in the central divertor region of KSTAR, within the field of view of an infra-red (IR) thermography system with a spatial resolution to 0.4 mm/pixel. Adjustment of the outer divertor strike point position is used to deposit power on the different blocks in different discharges. The measured power flux density on flat regions of the surrounding graphite tiles is used to obtain the parallel power flux, q|| impinging on the various W blocks. Experiments have been performed in Type I ELMing H-mode with Ip = 600 kA, BT = 2 T, PNBI = 3.5 MW, leading to a hot attached divertor with typical pulse lengths of 10 s. Three dimensional ANSYS simulations using q|| and assuming geometric projection of the heat flux are found to be consistent with the observed edge loading. This research was partially supported by Ministry of Science, ICT, and Future Planning under KSTAR project.

  11. 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.

  12. 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

  13. Design and concept validation of the new solid tungsten divertor for ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Herrmann, A., E-mail: albrecht.herrmann@ipp.mpg.de; Greuner, H.; Jaksic, N.; Böswirth, B.; Reimold, F.; Scarabosio, A.; Vorbrugg, S.; Wischmeier, M.

    2013-10-15

    Div-III, a divertor with solid tungsten target tiles for ASDEX Upgrade is designed and tested and will be installed in 2013. It is a further step in exploring tungsten as material for plasma facing components. It avoids the restrictions of tungsten coatings on graphite and realizes an operation range up to 50 MJ energy removing capability in the outer divertor. In addition, it allows physics investigation such as erosion and deuterium retention as well as effects of castellation and target tilting. The design of the target itself and the attachment was optimized with FE-analysis and was intensively high heat tested up to a double overload. Cyclic tests reveal that the target and the attachment can be operated with the design load of 50 MJ without any damage. Even a twofold overload results in local recrystallization and minor cracks but the targets did not fail during operation. The redesign of the divertor structure was used to increase the conductance between the cryo-pump and the divertor region. The impact of the changed pumping efficiency was investigated with SOLPS/Eirene modeling. The modeling results are an indication for an easier access to lower SOL densities as expected for a higher pumping efficiency in the main chamber.

  14. 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.

  15. 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 strike-poi

  16. Flow Field and Thermal Analysis of the Divertor Target Plate for HL-2A Tokamak

    Institute of Scientific and Technical Information of China (English)

    2005-01-01

    In the initial phase of the physics experiment, the double-null divertor plates used consist of graphite armor tiles, Mo-alloy intermediate layers and Cu-alloy coolant tubes. In the later operating phase, tungsten will be used as armor tiles.A multi-physical field numerical analysis method is used in this paper. Its analysis model reflects more realistically the real divertor structure than other models. Two-dimensional (2D)and three-dimensional (3D) fluid flow field, temperature distribution and thermal stress analyses of the divertor plates are carried out by the ANSYS code. During the physics experimental phase with a heat flux of 1 MW/m2, a coolant velocity of 5.48 m/s, and a thermal stress of 750 kg/cm2,the graphite armor tiles successfully meet the requirements of temperature, thermal stress and sputtering erosion. The tungsten armor will be considered as a second candidate. The result of simulation can be used for upgrading the design parameters of the HL-2A poloidal divertor.

  17. 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...

  18. Orbit Alignment in Triple Stars

    Science.gov (United States)

    Tokovinin, Andrei

    2017-08-01

    The statistics of the angle Φ between orbital angular momenta in hierarchical triple systems with known inner visual or astrometric orbits are studied. A correlation between apparent revolution directions proves the partial orbit alignment known from earlier works. The alignment is strong in triples with outer projected separation less than ∼50 au, where the average Φ is about 20^\\circ . In contrast, outer orbits wider than 1000 au are not aligned with the inner orbits. It is established that the orbit alignment decreases with the increasing mass of the primary component. The average eccentricity of inner orbits in well-aligned triples is smaller than in randomly aligned ones. These findings highlight the role of dissipative interactions with gas in defining the orbital architecture of low-mass triple systems. On the other hand, chaotic dynamics apparently played a role in shaping more massive hierarchies. The analysis of projected configurations and triples with known inner and outer orbits indicates that the distribution of Φ is likely bimodal, where 80% of triples have {{Φ }}< 70^\\circ and the remaining ones are randomly aligned.

  19. Zero Triple Product Determined Matrix Algebras

    Directory of Open Access Journals (Sweden)

    Hongmei Yao

    2012-01-01

    triple product in the aforementioned definition is replaced by Jordan triple product, then A is called zero Jordan triple product determined. This paper mainly shows that matrix algebra Mn(B, n≥3, where B is any commutative unital algebra even different from the above mentioned commutative unital algebra C, is always zero triple product determined, and Mn(F, n≥3, where F is any field with chF≠2, is also zero Jordan triple product determined.

  20. 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)

  1. 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.

  2. Lifetime measurements probing triple shape coexistence in 175Au

    Science.gov (United States)

    Watkins, H.; Joss, D. T.; Grahn, T.; Page, R. D.; Carroll, R. J.; Dewald, A.; Greenlees, P. T.; Hackstein, M.; Herzberg, R.-D.; Jakobsson, U.; Jones, P. M.; Julin, R.; Juutinen, S.; Ketelhut, S.; Kröll, T.; Krücken, R.; Labiche, M.; Leino, M.; Lumley, N.; Maierbeck, P.; Nyman, M.; Nieminen, P.; O'Donnell, D.; Ollier, J.; Pakarinen, J.; Peura, P.; Pissulla, T.; Rahkila, P.; Revill, J. P.; Rother, W.; Ruotsalainen, P.; Rigby, S. V.; Sarén, J.; Sapple, P. J.; Scheck, M.; Scholey, C.; Simpson, J.; Sorri, J.; Uusitalo, J.; Venhart, M.

    2011-11-01

    Lifetimes of the low-lying excited states in the very neutron-deficient nucleus 175Au have been measured by the recoil-distance Doppler-shift method using γ-ray spectra obtained with the recoil-decay tagging technique. Transition quadrupole moments and reduced transition probabilities extracted for this odd-Z nucleus indicate the existence of three different shapes and the competition between collective and noncollective structures.

  3. The WEST project: Testing ITER divertor high heat flux component technology in a steady state tokamak environment

    Energy Technology Data Exchange (ETDEWEB)

    Bucalossi, J., E-mail: jerome.bucalossi@cea.fr; Missirlian, M.; Moreau, P.; Samaille, F.; Tsitrone, E.; Houtte, D. van; Batal, T.; Bourdelle, C.; Chantant, M.; Corre, Y.; Courtois, X.; Delpech, L.; Doceul, L.; Douai, D.; Dougnac, H.; Faïsse, F.; Fenzi, C.; Ferlay, F.; Firdaouss, M.; Gargiulo, L.; and others

    2014-10-15

    The WEST project recently launched at Cadarache consists in transforming Tore Supra in an X-point divertor configuration while extending its long pulse capability, in order to test the ITER divertor technology. The implementation of a full tungsten actively cooled divertor with plasma facing unit representative of ITER divertor targets will allow addressing risks both in terms of industrial-scale manufacturing and operation of such components. Relevant plasma scenarios are foreseen for extensive testing under high heat load in the 10–20 MW/m{sup 2} range and ITER-like fluences (1000 s pulses). Plasma facing unit monitoring and development of protection strategies will be key elements of the WEST program. WEST is scheduled to enter into operation in 2016, and will provide a key facility to prepare and be prepared for ITER.

  4. The first results of divertor discharge and supersonic molecular beam injection on the HL-2A tokamak

    Institute of Scientific and Technical Information of China (English)

    Yao Liang-Hua; Yuan Bau-Shan; Feng Bei-Bin; Chen Cheng-Yuan; Hong Wen-Yu; Li Ying-Liang

    2007-01-01

    HL-2A tokamak is the first tokamak with divertors in China. The plasma boundary and the position of the striking point on the target plates of the HL-2A closed divertor were simulated by the current filament code and they were in agreement with the diagnostic results in the divertor. Supersonic molecular beam injection (SMBI) system was first installed and tested on the HL-2A tokamak in 2004. In the present experiment low pressure SMBI fuelling on the HL-2A and during the period of SMB pulse injection into the HL-2A plasma the power density convected at the target plate surfaces was 0.4 times of that before or after the beam injection. It is a useful fuelling method for decreasing the heat load on the neutralizer plates of the divertor.

  5. 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€.

  6. Effects of discharge operation regimes and magnetic field geometry on the in-out divertor asymmetry in EAST

    Energy Technology Data Exchange (ETDEWEB)

    Du, Hailong; Sang, Chaofeng [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams (Ministry of Education), School of Physics and Optoelectronic Technology, Dalian University of Technology, Dalian 116024 (China); Wang, Liang [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams (Ministry of Education), School of Physics and Optoelectronic Technology, Dalian University of Technology, Dalian 116024 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Bonnin, Xavier [LSPM-CNRS, Université Paris 13, Sorbonne Paris Cité, F-93430 Villetaneuse (France); Sun, Jizhong [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams (Ministry of Education), School of Physics and Optoelectronic Technology, Dalian University of Technology, Dalian 116024 (China); Wang, Dezhen, E-mail: wangdez@dlut.edu.cn [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams (Ministry of Education), School of Physics and Optoelectronic Technology, Dalian University of Technology, Dalian 116024 (China)

    2016-11-01

    Highlights: • The in-out divertor asymmetry is studied using SOLPS. • The discharge operation and the magnetic filed have a great influence on the divertor asymmetry. • The asymmetry is not obvious in low recycling regime as that in high recycling regime. - Abstract: This paper aims to investigate the reason why the divertor in-out asymmetry was not obvious as experimentally observed in EAST only considering the classical drifts from previous simulations (Guo et al., J. Nucl. Mater. 438 (2013) 280; Du et al., J. Nucl. Mater. 463 (2015) 485). With consideration of the classical drifts, a series of different typical discharge scenarios in EAST in different magnetic field geometries were simulated by using the SOLPS5.2 code package. The simulated results reveal that the classical drifts make a major contribution to the in-out divertor asymmetry in the high recycling regime (HRR) and partial detachment (one divertor target begins to detach, while the other divertor remains attached) regime. In comparison, in low recycling regime the classical drifts play a much smaller role in the contributions to the in-out divertor asymmetry, which can explain reasonably the reason for it in Guo et al. (J. Nucl. Mater. 438 (2013) 280). In addition, the magnetic field geometry also has a great impact on the classical drifts inducing the asymmetry; it is found that for lower single-null, upper single-null and connected double-null topologies, in HRR the classical drifts play an dominant role in the contribution to the in-out divertor asymmetry, while for a disconnected double null magnetic field configuration, they play a minor role, which is the reason why the in-out asymmetry was unobvious by considering the drifts in Du et al. (J. Nucl. Mater. 463 (2015) 485).

  7. Optimal design of divertor heat sink with different geometric configurations of sectorial extended surfaces

    Energy Technology Data Exchange (ETDEWEB)

    Rimza, Sandeep, E-mail: sandeepr@ipr.res.in [Divertor and First Wall Technology Development Division, Institute for Plasma Research (IPR), Bhat – 382428, Gandhinagar, Gujarat (India); Satpathy, Kamalakanta, E-mail: satpathy@ipr.res.in [Divertor and First Wall Technology Development Division, Institute for Plasma Research (IPR), Bhat – 382428, Gandhinagar, Gujarat (India); Khirwadkar, Samir, E-mail: sameer@ipr.res.in [Divertor and First Wall Technology Development Division, Institute for Plasma Research (IPR), Bhat – 382428, Gandhinagar, Gujarat (India); Velusamy, Karupanna, E-mail: kvelu@igcar.gov.in [Mechanics and Hydraulics Division, Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam 603102 (India)

    2015-11-15

    Highlights: • Effect of design variables in enhancing heat removal potential with pumping power assessed. • The optimization objective is to minimize the thimble temperature. • Investigation of optimum design parameters for various Reynolds number. • Practicability of the optimum designs is verified through structural analysis. • Benchmark validation of divertor finger mock-up against in-house experiment and good agreement is achieved. - Abstract: Cooling of fusion reactor divertor by helium is widely accepted due to its chemical and neutronic inertness and superior safety aspect. However, its poor thermo physical characteristics need high pressure to remove large heat flux encountered in fusion power plant (DEMO). In the perspective of DEMO, it is desirable to explore efficient cooling technology for divertor that can handle high heat flux. Toward this, a novel sectorial extended surface (SES) was proposed by the authors Rimza et al. (2014) [2]. The present work focuses on design optimization of divertor finger mock-up with SES to enhance the thermal hydraulic performance. The maximum thimble temperature is considered as the vital design constraint. Various non-dimensional design variables, viz., relative pitch, thickness, jet diameter, the ratio of height of SES to jet diameter and circumferential position of the SES are considered for the present optimization study. The effects of design variables on thermal performance of the divertor are evaluated in the Reynolds number (Re) range of 7.5 × 10{sup 4}–1.2 × 10{sup 5}. The analysis reveals that, the heat transfer performance of divertor finger mock-up with SES is improved for two optimum designs having relative pitch and thickness of 0.30 and 0.56, respectively. Also, it is observed that finger mock-up heat sink with SES performs better, when the ratio of SES height to jet diameter, reduces to 0.75 at the cost of marginally higher pumping power. The effects of jet diameter and circumferential

  8. Triple gastric peptic ulcer perforation.

    Science.gov (United States)

    Radojkovic, Milan; Mihajlovic, Suncica; Stojanovic, Miroslav; Stanojevic, Goran; Damnjanovic, Zoran

    2016-03-01

    Patients with advanced or metastatic cancer have compromised nutritional, metabolic, and immune conditions. Nevertheless, little is known about gastroduodenal perforation in cancer patients. Described in the present report is the case of a 41-year old woman with stage IV recurrent laryngeal cancer, who used homeopathic anticancer therapy and who had triple peptic ulcer perforation (PUP) that required surgical repair. Triple gastric PUP is a rare complication. Self-administration of homeopathic anticancer medication should be strongly discouraged when evidence-based data regarding efficacy and toxicity is lacking.

  9. 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.

  10. The development of in-situ calibration method for divertor IR thermography in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Takeuchi, M.; Sugie, T.; Ogawa, H.; Takeyama, S.; Itami, K. [Japan Atomic Energy Agency (Japan)

    2014-08-21

    For the development of the calibration method of the emissivity in IR light on the divertor plate in ITER divertor IR thermography system, the laboratory experiments have been performed by using IR instruments. The calibration of the IR camera was performed by the plane black body in the temperature of 100–600 degC. The radiances of the tungsten heated by 280 degC were measured by the IR camera without filter (2.5–5.1 μm) and with filter (2.95 μm, 4.67 μm). The preliminary data of the scattered light of the laser of 3.34 μm that injected into the tungsten were acquired.

  11. Design feasibility study of a divertor component reinforced with fibrous metal matrix composite laminate

    Energy Technology Data Exchange (ETDEWEB)

    You, J.-H. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Boltzmannstr. 2, D-85748 Garching (Germany)]. E-mail: j.h.you@ipp.mpg.de

    2005-01-01

    Fibrous metal matrix composites possess advanced mechanical properties compared to conventional alloys. It is expected that the application of these composites to a divertor component will enhance the structural reliability. A possible design concept would be a system consisting of tungsten armour, copper composite interlayer and copper heat sink where the composite interlayer is locally inserted into the highly stressed domain near the bond interface. For assessment of the design feasibility of the composite divertor concept, a non-linear multi-scale finite element analysis was performed. To this end, a micro-mechanics algorithm was implemented into a finite element code. A reactor-relevant heat flux load was assumed. Focus was placed on the evolution of stress state, plastic deformation and ductile damage on both macro- and microscopic scales. The structural response of the component and the micro-scale stress evolution of the composite laminate were investigated.

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

    Science.gov (United States)

    Griswold, M. E.; Korepanov, S.; Thompson, M. C.

    2016-11-01

    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.

  13. Melt damage simulation of W-macrobrush and divertor gaps after multiple transient events in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Bazylev, B.N. [Forschungszentrum Karlsruhe, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany)]. E-mail: bazylev@ihm.fzk.de; Janeschitz, G. [Forschungszentrum Karlsruhe, Fusion, P.O. Box 3640, 76021 Karlsruhe (Germany); Landman, I.S. [Forschungszentrum Karlsruhe, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany); Loarte, A. [EFDA-CSU, Max-Planck-Institut fuer Plasmaphysik, D-85748 Garching (Germany); Pestchanyi, S.E. [Forschungszentrum Karlsruhe, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2007-06-15

    Tungsten in the form of macrobrush structure is foreseen as one of two candidate materials for the ITER divertor and dome. In ITER, even for moderate and weak ELMs when a thin shielding layer does not protect the armour surface from the dumped plasma, the main mechanisms of metallic target damage remain surface melting and melt motion erosion, which determines the lifetime of the plasma facing components. The melt erosion of W-macrobrush targets with different geometry of brush surface under the heat loads caused by weak ELMs is numerically investigated using the modified code MEMOS. The optimal angle of brush surface inclination that provides a minimum of surface roughness is estimated for given inclination angles of impacting plasma stream and given parameters of the macrobrush target. For multiple disruptions the damage of the dome gaps and the gaps between divertor cassettes caused by the radiation impact is estimated.

  14. Melt damage simulation of W-macrobrush and divertor gaps after multiple transient events in ITER

    Science.gov (United States)

    Bazylev, B. N.; Janeschitz, G.; Landman, I. S.; Loarte, A.; Pestchanyi, S. E.

    2007-06-01

    Tungsten in the form of macrobrush structure is foreseen as one of two candidate materials for the ITER divertor and dome. In ITER, even for moderate and weak ELMs when a thin shielding layer does not protect the armour surface from the dumped plasma, the main mechanisms of metallic target damage remain surface melting and melt motion erosion, which determines the lifetime of the plasma facing components. The melt erosion of W-macrobrush targets with different geometry of brush surface under the heat loads caused by weak ELMs is numerically investigated using the modified code MEMOS. The optimal angle of brush surface inclination that provides a minimum of surface roughness is estimated for given inclination angles of impacting plasma stream and given parameters of the macrobrush target. For multiple disruptions the damage of the dome gaps and the gaps between divertor cassettes caused by the radiation impact is estimated.

  15. Free-boundary ideal MHD stability of W7-X divertor equilibria

    Science.gov (United States)

    Nührenberg, C.

    2016-07-01

    Plasma configurations describing the stellarator experiment Wendelstein 7-X (W7-X) are computationally established taking into account the geometry of the test-divertor unit and the high-heat-flux divertor which will be installed in the vacuum chamber of the device (Gasparotto et al 2014 Fusion Eng. Des. 89 2121). These plasma equilibria are computationally studied for their global ideal magnetohydrodynamic (MHD) stability properties. Results from the ideal MHD stability code cas3d (Nührenberg 1996 Phys. Plasmas 3 2401), stability limits, spatial structures and growth rates are presented for free-boundary perturbations. The work focusses on the exploration of MHD unstable regions of the W7-X configuration space, thereby providing information for future experiments in W7-X aiming at an assessment of the role of ideal MHD in stellarator confinement.

  16. Divertor target profiles and recycling studies in TCV single null lower standard discharges

    Energy Technology Data Exchange (ETDEWEB)

    Pitts, R.A.; Nieswand, C.; Weisen, H.; Anton, M.; Behn, R.; Chavan, R.F.; Dutch, M.J.; Duval, B.P.; Franke, S.; Hofmann, F.; Joye, B.; Lister, J.B.; Llobet, X.; Martin, Y.; Moret, J.-M.; Petrzilka, J.; Pietrzyk, Z.A.; Piffl, V.; Reinke, P.; Rensink, M.E.; Smith, G.R.; Van Toledo, W. [Ecole Polytech. Federale, Lausanne (Switzerland). Centre de Recherches en Physique des Plasmas]|[Institute of Plasma Physics, Czech Academy of Sciences, Prague (Czech Republic)]|[Universitaet Basel, Institut fuer Physik, Klingelbergstr. 82, CH-4056 Basel (Switzerland)]|[Lawrence Livermore National Laboratory, University of California, Berkeley, CA 94551 (United States)

    1997-02-01

    A `standard`, single null lower diverted discharge has been developed to enable continuous monitoring of the first wall conditions and to characterise the effectiveness and influence of wall conditioning in the TCV tokamak. Measurements over a period encompassing nearly 2000 ohmic discharges of varying configuration and input power show the global confinement time and main plasma impurity concentrations to be good general indicators of the first wall condition, whilst divertor target profiles demonstrate strikingly the short term beneficial effects of He glow. Good agreement, consistent with a reduction in recycling at the plates is found between the predictions of the fluid code UEDGE and the observed outer divertor profiles of T{sub e} and n{sub e} before and after He glow. (orig.).

  17. Impurity profiles at the JET divertor targets compared with the DIVIMP code

    Energy Technology Data Exchange (ETDEWEB)

    Matthews, G.F.; Gottardi, N.A.C.; Harbour, P.J.; Horton, L.D.; Jackel, H.J.; De Kock, L.; Loarte, A.; Maggi, C.F.; O' Brien, D.P.J.; Simonini, R.; Spence, J.; Stamp, M.F.; Stott, P.E.; Summers, H.P.; Tagle, J.; Von Hellerman, M. (JET Joint Undertaking, Abingdon (United Kingdom)); Stangeby, P.C.; Elder, J.D. (Univ. Toronto, Inst. for Aerospace Studies, Downsview, Ontario (Canada))

    1992-12-01

    In this paper we describe the simulation of edge diagnostics in JET using the DIVIMP (divertor impurity) Monte Carlo code. We concentrate on two ohmic pulses and show how the results are influenced by a variety of modeling assumptions. Our results show that a wall source must be included to explain the diagnostic signals. The wall source is shown to be a significant source of impurity in the discharges studied and more generally. (orig.).

  18. HRP facility for fabrication of ITER vertical target divertor full scale plasma facing units

    Energy Technology Data Exchange (ETDEWEB)

    Visca, Eliseo, E-mail: eliseo.visca@enea.it [Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, IT-00044 Frascati (Roma) (Italy); Roccella, S. [Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, IT-00044 Frascati (Roma) (Italy); Candura, D.; Palermo, M. [Ansaldo Nucleare S.p.A., Corso Perrone 25, IT-16152 Genova (Italy); Rossi, P.; Pizzuto, A. [Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, IT-00044 Frascati (Roma) (Italy); Sanguinetti, G.P. [Ansaldo Nucleare S.p.A., Corso Perrone 25, IT-16152 Genova (Italy); Mancini, A.; Verdini, L.; Cacciotti, E.; Cerri, V.; Mugnaini, G.; Reale, A.; Giacomi, G. [Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, IT-00044 Frascati (Roma) (Italy)

    2015-10-15

    Highlights: • R&D activities for the manufacturing of ITER divertor high heat flux plasma-facing components (HHFC). • ENEA and Ansaldo have jointly manufactured several actively cooled monoblock mock-ups and prototypical components. • ENEA and ANSALDO NUCLEARE jointly participate to the European program for the qualification of the manufacturing technology for the ITER divertor IVT. • Successful manufacturing by HRP (Hot Radial Pressing) of first full-scale full-W armored IVT qualification prototype. - Abstract: ENEA and Ansaldo Nucleare S.p.A. (ANN) have being deeply involved in the European development activities for the manufacturing of the ITER Divertor Inner Vertical Target (IVT) plasma-facing components. During normal operation the heat flux deposited on the bottom segment of divertor is 5–10 MW/m{sup 2} but the capability to remove up to 20 MW/m{sup 2} during transient events of 10 s must also be demonstrated. In order to fulfill ITER requirements, ENEA has set up and widely tested a manufacturing process, named Hot Radial Pressing (HRP). The last challenge is now to fabricate full-scale prototypes of the IVT, aimed to be qualified for the next step, i.e. the series production. On the basis of the experience of manufacturing hundreds of small mock-ups, ENEA designed and installed a new suitable HRP facility. The objective of getting a final shaped plasma facing unit (PFU) that satisfies these requirements is an ambitious target because tolerances set by ITER/F4E are very tight. The setting-up of the equipment started with the fabrication of full scale and representative ‘dummies’ in which stainless steel instead of CFC or W was used for monoblocks. The results confirmed that dimensions were compliant with the required tolerances. The paper reports a brief description of the innovative HRP equipment and the dimensional check results after HRP of the first full-scale full-W PFU.

  19. Infrared thermography inspection for monoblock divertor target in JT-60SA

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Shigetoshi, E-mail: nakamura.shigetoshi@jaea.go.jp; Sakurai, Shinji; Ozaki, Hidetsugu; Seki, Yohji; Yokoyama, Kenji; Sakasai, Akira; Tsuru, Daigo

    2014-10-15

    Highlights: • Infrared thermography inspection is modified to inspect JT-60SA divertor targets. • Infrared thermography inspection is effective to detect joining defects of targets. • Numerical analysis is in good agreement with inspection results of mock-up targets. • Database for setting screening criteria has been constructed by numerical analysis. - Abstract: Carbon fiber composite (CFC) monoblock divertor target is required for power handling in JT-60SA. Quality of the targets depends on a joining technology in manufacturing process. To inspect the quality of more than 900 target pieces, efficient non-destructive inspection is needed. An infrared thermography inspection (IR inspection), has been proposed by ITER and IRFM, where the quality between CFC and a cooling tube is examined by a use of transient thermal response at a rapid switch from hot to cold water flow. In JT-60SA divertor target, a screw tube will be employed to obtain high heat transfer efficiency with simple structure. Since the time response of the screw tube is much faster than that of smooth tube, it is required to confirm the feasibility of this IR inspection. Thus, the effect of joining defects on transient thermal response of the targets has been investigated experimentally by using the mock-up targets containing defects which are artificially made. It was found that the IR inspection can detect the defects. Moreover, screening criteria of IR inspection for acceptable monoblock target is discussed.

  20. Plasma convection near the magnetic null of a snowflake divertor during an ELM event

    Energy Technology Data Exchange (ETDEWEB)

    Ryutov, D.D.; Cohen, R.H.; Rognlien, T.D.; Umansky, M.V. [Lawrence Livermore National Laboratory, Livermore, CA (United States)

    2012-06-15

    A snowflake magnetic configuration is created in a tokamak when the poloidal magnetic field and its first spatial derivatives become zero at a certain point. The separatrix then acquires a characteristic hexagonal shape reminiscent of a snowflake. We study new features of the plasma macroscopic equilibrium and stability in the vicinity of the snowflake null. We note that, compared to the standard X-point divertor, the zone of weak poloidal magnetic field is much larger. The weak poloidal field leads to development of intense plasma convection over the expanded area around the null-point during the ejection phase of an edge localized mode (ELM) event when the plasma pressure in the scrape-off layer increases compared to its inter-ELM value. Intense convection may lead to a roughly-equal splitting of the heat flux between the 4 snowflake divertor legs and to a broadening of the plasma wetted area in each leg, thereby mitigating damage to divertor plates (copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  1. 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.

  2. Study of power load pattern on EAST divertor using PFCFlux code

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Bin, E-mail: binzhang@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Firdaouss, Mehdi [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Gong, Xianzu [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Ekedahl, Annika [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Peng, Xuebing [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Zhang, Xiaodong, E-mail: xdzhang@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2016-06-15

    Highlights: • This paper demonstrates the modeling result of power load pattern on EAST graphite divertor by using the PFCFlux code. • The grazing angle varies both poloidally and toroidally, changing by half a degree over the distance of 50 mm away from the strike point. • The correlation between both grazing angle and flux expansion and the magnetic equilibrium parameters are found by using the linear regression method. • The modeling result indicates that the edges of graphite tiles of EAST divertor are perfectly shadowed. - Abstract: The power load pattern on an EAST divertor component, spanning six tiles in the poloidal direction, has been studied with the PFCFlux code. A total of 49 different EAST plasma equilibria in lower single null configuration are used in the study. It is found that the incidence angle, or grazing angle, varies both toroidally and poloidally on the target, changing by approximately half a degree over a distance of 50 mm from the strike point. Strong correlations between the triangularity of the magnetic equilibrium and both the grazing angle and the flux expansion are found by using linear regression. A smaller value of triangularity gives wider plasma-wetted region on the target in lower-outer configuration, and a narrower plasma-wetted region in lower-inner configuration.

  3. Investigation of momentum loss mechanisms in the divertor region of ASDEX Upgrade with EMC3-Eirene

    Energy Technology Data Exchange (ETDEWEB)

    Brida, Dominik [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Physik-Department E28, Technische Universitaet Muenchen, Garching (Germany); Lunt, Tilmann; Wischmeier, Marco [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Collaboration: the ASDEX Upgrade Team

    2015-05-01

    In future fusion devices, such as ITER and DEMO, it will be necessary to prevent direct contact between the hot confined plasma and the vessel wall. This will be achieved by employing divertors, which offer a number of desirable advantages, such as screening of impurities from the core plasma, improved energy confinement (H-mode) and effective pumping of helium ash and hydrogen. However, due to material limits, the power and particle flux as well as the temperature at the target must be reduced to acceptable levels. This can be attained by operating the divertor in a (partially) detached regime, which requires considerable volumetric energy and momentum losses in the divertor region. Previous studies identified ion-neutral friction as the principal momentum loss factor. For this contribution the fluid code EMC3-Eirene was applied to simulate ASDEX Upgrade discharges with increasing plasma densities and investigate the role of different momentum loss mechanisms by switching on and off respective terms in the simulation. Interestingly, even without the charge-exchange collisions a strong momentum loss is observed in the simulation.

  4. Experience gained with the 3D machining of the W7-X HHF divertor target elements

    Energy Technology Data Exchange (ETDEWEB)

    Junghanns, P. [Max Planck Institute for Plasma Physics, Greifswald (Germany); Boscary, J., E-mail: jean.boscary@ipp.mpg.de [Max Planck Institute for Plasma Physics, Garching (Germany); Peacock, A. [Max Planck Institute for Plasma Physics, Garching (Germany)

    2015-10-15

    Highlights: • The Wendelstein 7-X surface of the actively cooled divertor is built up of 890 individually 3D machined target elements. • To date 300 target elements have been 3D machined with an accuracy of ±0.015 mm. • Copper discovered on the surface of few elements is no risk to operation. - Abstract: The high heat flux (HHF) divertor of W7-X consists of 100 target modules assembled from 890 actively water-cooled target elements protected with CFC tiles. The divertor surface will be built up of individually 3D machined target elements with 89 individual element types. To date 300 of the 890 target elements have been 3D machined with a very good accuracy. To achieve this successful result, a prototyping phase has been conducted to qualify the manufacturing route and to define the acceptance criteria with measures taken to minimize the risk of unacceptable damage during the manufacturing. After the 3D-machining, during the incoming inspection, copper infiltration from the interface between the CFC tiles and the CuCrZr heat sink to the plasma facing surface was detected in a small number of elements.

  5. HSX as an example of a resilient non-resonant divertor

    Science.gov (United States)

    Bader, A.; Boozer, A. H.; Hegna, C. C.; Lazerson, S. A.; Schmitt, J. C.

    2017-03-01

    This paper describes an initial description of the resilient divertor properties of quasi-symmetric (QS) stellarators using the HSX (Helically Symmetric eXperiment) configuration as a test-case. Divertors in high-performance QS stellarators will need to be resilient to changes in plasma configuration that arise due to evolution of plasma pressure profiles and bootstrap currents for divertor design. Resiliency is tested by examining the changes in strike point patterns from the field line following, which arise due to configurational changes. A low strike point variation with high configuration changes corresponds to high resiliency. The HSX edge displays resilient properties with configuration changes arising from the (1) wall position, (2) plasma current, and (3) external coils. The resilient behavior is lost if large edge islands intersect the wall structure. The resilient edge properties are corroborated by heat flux calculations from the fully 3-D plasma simulations using EMC3-EIRENE. Additionally, the strike point patterns are found to correspond to high curvature regions of magnetic flux surfaces.

  6. Engineering challenges and development of the ITER Blanket System and Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Merola, Mario, E-mail: mario.merola@iter.org; Escourbiac, Frederic; Raffray, Alphonse Rene; Chappuis, Philippe; Hirai, Takeshi; Gicquel, Stefan

    2015-10-15

    The ITER Blanket System and the Divertor are the main components which directly face the plasma. Being the first physical barrier to the plasma, they have very demanding design requirements, which include accommodating: (1) surface heat flux and neutronic volumetric heating, (2) electromagnetic loads, (3) nuclear shielding function, (4) capability of being assembled and remote-handled, (5) interfaces with other in-vessel components, and (6) high heat flux technologies and complex welded structures in the design. The main functions of the Blanket System have been substantially expanded and it has now also to provide limiting surfaces that define the plasma boundary during startup and shutdown. As regards the Divertor, the ITER Council decided in November 2013 to start the ITER operation with a full-tungsten armour in order to minimize costs and already gain operational experience with tungsten during the non-active phase of the machine. This paper gives an overview of the design and technology qualification of the Blanket System and the Divertor.

  7. 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.

  8. Coupled Kinetic-MHD Simulations of Divertor Heat Load with ELM Perturbations

    Science.gov (United States)

    Cummings, Julian; Chang, C. S.; Park, Gunyoung; Sugiyama, Linda; Pankin, Alexei; Klasky, Scott; Podhorszki, Norbert; Docan, Ciprian; Parashar, Manish

    2010-11-01

    The effect of Type-I ELM activity on divertor plate heat load is a key component of the DOE OFES Joint Research Target milestones for this year. In this talk, we present simulations of kinetic edge physics, ELM activity, and the associated divertor heat loads in which we couple the discrete guiding-center neoclassical transport code XGC0 with the nonlinear extended MHD code M3D using the End-to-end Framework for Fusion Integrated Simulations, or EFFIS. In these coupled simulations, the kinetic code and the MHD code run concurrently on the same massively parallel platform and periodic data exchanges are performed using a memory-to-memory coupling technology provided by EFFIS. The M3D code models the fast ELM event and sends frequent updates of the magnetic field perturbations and electrostatic potential to XGC0, which in turn tracks particle dynamics under the influence of these perturbations and collects divertor particle and energy flux statistics. We describe here how EFFIS technologies facilitate these coupled simulations and discuss results for DIII-D, NSTX and Alcator C-Mod tokamak discharges.

  9. Solution Patterns Predicting Pythagorean Triples

    Science.gov (United States)

    Ezenweani, Ugwunna Louis

    2013-01-01

    Pythagoras Theorem is an old mathematical treatise that has traversed the school curricula from secondary to tertiary levels. The patterns it produced are quite interesting that many researchers have tried to generate a kind of predictive approach to identifying triples. Two attempts, namely Diophantine equation and Brahmagupta trapezium presented…

  10. Local derivations on Jordan triples

    OpenAIRE

    Mackey, Michael

    2013-01-01

    R. V. Kadison (J. Algebra 130 (1990) 494–509) defined the notion of local derivation on an algebra and proved that every continuous local derivation on a von Neumann algebra is a derivation. We provide the analogous result in the setting of Jordan triples.

  11. Classifying Two-dimensional Hyporeductive Triple Algebras

    CERN Document Server

    Issa, A Nourou

    2010-01-01

    Two-dimensional real hyporeductive triple algebras (h.t.a.) are investigated. A classification of such algebras is presented. As a consequence, a classification of two-dimensional real Lie triple algebras (i.e. generalized Lie triple systems) and two-dimensional real Bol algebras is given.

  12. Probe Storage

    NARCIS (Netherlands)

    Gemelli, Marcellino; Abelmann, Leon; Engelen, Johan B.C.; Khatib, Mohammed G.; Koelmans, Wabe W.; Zaboronski, Olog; Campardo, Giovanni; Tiziani, Federico; Laculo, Massimo

    2011-01-01

    This chapter gives an overview of probe-based data storage research over the last three decades, encompassing all aspects of a probe recording system. Following the division found in all mechanically addressed storage systems, the different subsystems (media, read/write heads, positioning, data chan

  13. Cultural probes

    DEFF Research Database (Denmark)

    Madsen, Jacob Østergaard

    2016-01-01

    The aim of this study was thus to explore cultural probes (Gaver, Boucher et al. 2004), as a possible methodical approach, supporting knowledge production on situated and contextual aspects of occupation.......The aim of this study was thus to explore cultural probes (Gaver, Boucher et al. 2004), as a possible methodical approach, supporting knowledge production on situated and contextual aspects of occupation....

  14. Thermo-mechanical and damage analyses of EAST carbon divertor under type-I ELMy H-mode operation

    Energy Technology Data Exchange (ETDEWEB)

    Li, W.X. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China); 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); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China); Ye, M.Y. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Peng, X.B., E-mail: pengxb@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Wu, S.T. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Qian, X.Y.; Zhu, C.C. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China)

    2016-04-15

    Highlights: • Type-I ELMy H-mode is one of the most severe operating environment in tokamak. • An actual time-history heat load has been used in thermo-mechanical analysis. • The analysis results are time-dependent during the whole discharge process. • The analysis could be very useful in evaluating the operational capability of the divertor. - Abstract: The lower carbon divertor has been used since 2008 in EAST, and many significant physical results, like the 410 s long pulse discharge and the 32 s H-mode operation, have been achieved. As the carbon divertor will still be used in the next few years while the injected auxiliary heating power would be increased gradually, it’s necessary to evaluate the operational capability of the carbon divertor under the heat loads during future operation. In this paper, an actual time-history heat load during type-I ELMy H-mode from EAST experiment, as one of the most severe operating environment in tokamak, has been used in the calculation and analysis. The finite element (FE) thermal and mechanical calculations have been carried out to analysis the stress and deformation of the carbon divertor during the heat loads. According to the results, the main impact on the overall temperature comes from the relative stable phase before and after the type-I ELMs and local peak load, and the transient thermal load such as type-I ELMy only has a significant effect on the surface temperature of the graphite tiles. The carbon divertor would work with high stress near the screw bolts in the current operational conditions, because of high preload and conservative frictional coefficient between the bolts and heatsink. For the future operation, new plasma facing materials (PFM) and divertor technology should be developed.

  15. Two-dimensional numerical study of ELMs-induced erosion of tungsten divertor target tiles with different edge shapes

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Yan [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams (Ministry of Education), School of Physics and Optoelectronic Technology, Dalian University of Technology, Dalian 116024 (China); School of Information Science and Engineering, Dalian Polytechnic University, Dalian 116034 (China); Sun, Jizhong, E-mail: jsun@dlut.edu.cn [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams (Ministry of Education), School of Physics and Optoelectronic Technology, Dalian University of Technology, Dalian 116024 (China); Hu, Wanpeng; Sang, Chaofeng [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams (Ministry of Education), School of Physics and Optoelectronic Technology, Dalian University of Technology, Dalian 116024 (China); Wang, Dezhen, E-mail: wangdez@dlut.edu.cn [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams (Ministry of Education), School of Physics and Optoelectronic Technology, Dalian University of Technology, Dalian 116024 (China)

    2016-01-15

    Highlights: • Thermal performance of three edge-shaped divertor tiles was assessed numerically. • All the divertor tiles exposed to type-I ELMs like ITER's will melt. • The rounded edge tile thermally performs the best in all tiles of interest. • The incident energy flux density was evaluated with structural effects considered. - Abstract: Thermal performance of the divertor tile with different edge shapes was assessed numerically along the poloidal direction by a two-dimensional heat conduction model with considering the geometrical effects of castellated divertor tiles on the properties of its adjacent plasma. The energy flux density distribution arriving at the castellated divertor tile surface was evaluated by a two-dimension-in-space and three-dimension-in-velocity particle-in-cell plus Monte Carlo Collisions code and then the obtained energy flux distribution was used as input for the heat conduction model. The simulation results showed that the divertor tiles with any edge shape of interest (rectangular edge, slanted edge, and rounded edge) would melt, especially, in the edge surface region of facing plasma poloidally under typical heat flux density of a transient event of type-I ELMs for ITER, deposition energy of 1 MJ/m{sup 2} in a duration of 600 μs. In comparison with uniform energy deposition, the vaporizing erosion was reduced greatly but the melting erosion was aggravated noticeably in the edge area of plasma facing diveror tile. Of three studied edge shapes, the simulation results indicated that the divertor plate with rounded edge was the most resistant to the thermal erosion.

  16. An operational non destructive examination for ITER divertor plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Durocher, A.; Escourbiac, F.; Farjon, J.L.; Vignal, N.; Cismondi, F. [Association Euratom-CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Merola, M. [ITER International Team, Cadarache, 13 - St Paul Lez Durance (France); Riccardi, B. [CEFDA CSU-Garching, Garching bei Munchen (Germany)

    2007-07-01

    Full text of publication follows: To meet the power exhaust - heat flux of 20 MW/m{sup 2} - requirements of Plasma Facing Components (PFCs) during plasma operation requires control of their thermal and mechanical integrity. As heat exhaust capability and lifetime of PFCs during in-situ operation are linked to the manufacturing quality, it is an absolute requirement to develop reliable nondestructive examination methods, in particular of the CFC-CuCrZr joint, throughout the manufacturing process. Within the framework of Tokamak Tore Supra upgrade, a pioneering activity has been developed to evaluate the capability of the PFC to be efficiently cooled. In 1998 a test bed - so called SATIR - based on the heat transient method was developed by the CEA and is used today as an inspection tool in order to guarantee the PFCs performances. The technical procurement plan of ITER Divertor targets stated that all Cu cast layers on CFC armour should be subjected to 100% thermographic examination. Each ITER Party should demonstrate its technical capability to carry out the PFC with the required cooling efficiently. The ITER Divertor PFCs pose new challenges especially for the mono-block CFC thickness, and the number of full scale units to be tested which is higher than on any existing or under construction fusion machine. The SATIR method as functional inspection has been identified as the basis test to decide upon the final acceptance of the Divertor PFCs. In order to increase the detection sensitivity of SATIR test bed, several possibilities have been assessed i) the increase of the convective heat transfer coefficient, which improved in a significant way the sensitivity of SATIR diagnostic on ITER components. ii) the installation of a digital infrared camera and the improvement of the thermal signal processing, has led to a considerable increase of performances iii) an innovative process based on spatial image autocorrelation will allow to localize the interlayer defect

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-08-01

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

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

    Science.gov (United States)

    Soukhanovskii, V. A.; Kaita, R.; Stratton, B.

    2016-11-01

    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 Te 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 Te-dependent signal within a characteristic divertor detachment equilibration time of ˜10-15 ms is expected.

  19. Triple-Negative Breast Carcinoma.

    Science.gov (United States)

    Livasy, Chad A

    2009-06-01

    Triple-negative breast carcinomas (TNBCs) comprise approximately 15% to 20% of breast cancers. Accurate assessment of tumor estrogen receptor, progesterone receptor, and human epidermal growth factor receptor 2 (HER2) status is an essential part of classifying tumors into this group. As a group, these tumors are associated with poor clinical outcomes and have been shown to exhibit an increased propensity for hematogenous metastasis to the brain and lungs. Many TNBCs, particularly ductal, not otherwise specified (NOS), and metaplastic carcinomas, show an overlapping characteristic gene expression pattern when evaluated by cDNA microarrays. This group has been termed basal-like because of the similarity with normal breast basal/myoepithelial cells including basal cytokeratin expression and lack of hormone receptor and HER2 expression. The array data have been used to develop multiple immunohistochemical surrogates to identify basal-like tumors in formalin-fixed, paraffin-embedded tissues, most employing basal cytokeratins and epidermal growth factor receptor. Currently, there is no international consensus on biomarkers used to identify tumors as basal-like, and the routine use of the term basal-like in surgical pathology reports is premature. Tumor morphologic features associated with triple-negative status include Nottingham grade 3 with high mitotic rate, pushing border of invasion, geographic tumor necrosis, solid/sheet-like growth pattern, lymphocytic infiltrate, and large central acellular zone. Most breast cancers arising in patients who have a germ-line BRCA1 mutation show similar histologic features and a triple-negative phenotype. Not all TNBCs are associated with an unfavorable prognosis, drawing attention to the heterogeneity of this tumor group and the continued need to link tumor morphology and grade with triple-negative status. This article focuses on histopathology, molecular characterization, carcinogenesis, clinical behavior, and treatment of these

  20. Simulation study of power load with impurity seeding in advanced divertor “short super-X divertor” for a tokamak reactor

    Energy Technology Data Exchange (ETDEWEB)

    Asakura, N., E-mail: asakura.nobuyuki@jaea.go.jp [Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212 (Japan); Hoshino, K. [Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212 (Japan); Shimizu, K. [Japan Atomic Energy Agency, Naka, Ibaraki 311-0193 (Japan); Shinya, K. [Toshiba Nuclear Engineering Services Co., Isogo, Yokohama 25-8523 (Japan); Utoh, H.; Tokunaga, S.; Tobita, K. [Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212 (Japan); Ohno, N. [Graduate School of Engineering, Nagoya Univ., Nagoya 464-8603 (Japan)

    2015-08-15

    A short super-X divertor (SXD) is proposed as an option for the Demo divertor, where the field line length from the divertor null to the outer target was largely increased compared to a similar-size conventional divertor. Physics and engineering design studies for a 3 GW-level fusion power Demo reactor (SlimCS) (Tobita et al., 2009) have recently progressed. Minimal number of the divertor coils were installed inside the toroidal field coil, i.e. interlink-winding. Arrangement of the poloidal field coils and their currents were determined, taking into account of the engineering design such as vacuum vessel and the neutron shield structures, and the divertor maintenance scenario. Divertor plasma simulation showed that significant radiation region is produced between the super-X null and the target. Radiation loss in the divertor was increased, producing fully detached plasmas efficiently. Advantages of the short SXD were demonstrated, but the total peak heat load was a marginal level (10 MW m{sup −2}) for the engineering design.

  1. The evolution of triple-star systems

    Science.gov (United States)

    Toonen, Silvia; Hamers, Adrian; Portegies Zwart, Simon

    2017-01-01

    While the principles of stellar and binary evolution theory have been accepted for a long time, our understanding of triple-star evolution is lagging behind. It is important to understand these systems, as triples are common in the field. About 15% of low-mass stellar systems are triples, but for high-mass stars the fraction increases to over 50%. At the same time, triple evolution is often invoked to explain exotic systems which cannot be explained easily by binary evolution. Examples are low-mass X-ray binaries, supernova type Ia progenitors and blue stragglers.Modeling triple evolution, however, is challenging as it is a combination of three-body dynamics and stellar evolution. In the past, most studies of three-body systems have focused on purely dynamical aspects without taking stellar evolution into account. However, in recent years, the first interdisciplinary studies have taken place which demonstrate the richness of the interacting regime. Here, I will show the first results of our new code TRES for simulating the evolution of stellar triples, which combines stellar evolution and interactions with three-body dynamics. In this talk, I will give an overview of the evolution of realistic (stellar) triples and I will discuss how triple evolution differs from binary evolution. What are the common evolutionary pathways that triple systems evolve through? Are there any evolutionary pathways open to triples, which are not open to isolated binaries? These are some of the important questions we want to answer.

  2. Control of transient gain absorption via tunneling and incoherent pumping in triple quantum dots

    Science.gov (United States)

    Tian, Si-Cong; Zhang, Xiao-Jun; Wan, Ren-Gang; Wang, Li-Jie; Shu, Shi-Li; Wang, Tao; Lu, Ze-Feng; Sun, Fang-Yuan; Tong, Cun-Zhu

    2017-01-01

    The transient gain-absorption properties of the probe field in vertical triple quantum dots assisted by double tunneling and incoherent pumping are investigated. With a proper intensity value and detuning of the second tunneling, the transient gain in triple quantum dots with incoherent pumping can be completely eliminated. In addition, the incoherent pumping affects both the amplitude of the transient absorption and the steady-state value. The dependence of transient behaviors on other parameters, such as the radiative decay rate and the pure dephasing decay rate of the quantum dots, is also discussed. The scheme may have important applications in quantum information networks and communication.

  3. Triple Higgs boson production at a 100 TeV proton-proton collider

    CERN Document Server

    Papaefstathiou, Andreas

    2016-01-01

    We consider triple Higgs boson production at a future 100 TeV proton-proton collider. We perform a survey of viable final states and compare and contrast triple production to Higgs boson pair production. Focussing on the $hhh \\rightarrow (b\\bar{b}) (b\\bar{b}) (\\gamma \\gamma)$ final state, we construct a baseline analysis for the Standard Model scenario and simple deformations, demonstrating that the process merits investigation in the high-luminosity phase of the future collider as a new probe of the self-coupling sector of the Higgs boson.

  4. Impurity ion flow and temperature measured in a detached divertor with externally applied non-axisymmetric fields on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Briesemeister, A.R., E-mail: briesemeister@fusion.gat.com [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Isler, R.C. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Allen, S.L. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Ahn, J.-W. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); McLean, A.G. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Unterberg, E.A.; Hillis, D.L. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Fenstermacher, M.E.; Meyer, W.H. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States)

    2015-08-15

    Externally applied non-axisymmetric magnetic fields are shown to have little effect on the impurity ion flow velocity and temperature as measured by the multichord divertor spectrometer in the DIII-D divertor for both attached and detached conditions. These experiments were performed in H-mode plasmas with the grad-B drift toward the target plates, with and without n = 3 resonant magnetic perturbations (RMPs). The flow velocity in the divertor is shown to change by as much as 30% when deuterium gas puffing is used to create detachment of the divertor plasma. No measurable changes in the C III flow were observed in response to the RMP fields for the conditions used in this work. Images of the C III emission are used along with divertor Thomson scattering to show that the local electron and C III temperatures are equilibrated for the conditions shown.

  5. Divertor heat fluxes and profiles during mitigated and unmitigated Edge Localised Modes (ELMs) on the Mega Amp Spherical Tokamak (MAST)

    CERN Document Server

    Thornton, A J; Chapman, I T; Harrison, J R

    2013-01-01

    Edge localised modes (ELMs) are a concern for future devices as they can limit the operational lifetime of the divertor. The mitigation of ELMs can be performed by the application of resonant magnetic perturbations (RMPs) which act to degrade the pressure gradient in the edge of the plasma. Investigations of the effect of RMPs on MAST have been performed in a range of plasmas using perturbations with toroidal mode numbers of n=3, 4 and 6. It has been seen that the RMPs increase the ELM frequency, which gives rise to a corresponding decrease in the ELM energy. The reduced ELM energy decreases the peak heat flux to the divertor, with a three fold reduction in the ELM energy, generating a 1.5 fold reduction in the peak heat flux. Measurements of the divertor heat flux profile show evidence of strike point splitting consistent with modelling using the vacuum code ERGOS.

  6. 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

  7. 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.

  8. 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.

  9. A convective divertor utilizing a 2nd-order magnetic field null

    Science.gov (United States)

    Rognlien, Thomas

    2014-10-01

    New results motivate a detailed study of a magnetic divertor concept characterized by strong plasma convection near a poloidal magnetic field (Bp) null region. The configuration is that of a near-2nd-order Bp null (Bp ~ Δ r2) , as in a snowflake divertor. The concept has 2 key features: (A) Convection spreads the heat flux between multiple divertor legs and further broadens the heat-flux profile within each leg, thereby greatly reducing target-plate heat loads. (B) The heat flux is further reduced by line radiation in each leg in detachment-like ionization zones. Theory indicates that convective turbulence arises when the poloidal plasma beta, βp = 2μ0nT/B p 2 >> 1 . Measurements in TCV now more fully quantify earlier NSTX and TCV observations of plasma mixing, and related modeling of TCV indicates that strongly enhanced null-region transport is present. Convective mixing provides a stabilizing mechanism to prevent the ionization fronts (hydrogenic and impurity) from collapsing to a highly radiating core MARFE. Also, the radiating zone maps to a very small region at the midplane owing to the very weak Bp in the convective region, thus minimizing its impact on the core plasma. Detailed calculations are reported that combine features A and B noted above. The plasma mixing mechanisms are described together with the corresponding transport model implemented in the 2D UEDGE edge transport code. UEDGE calculations are presented that quantify the roles of mixing, impurity radiation, and detachment stability for a realistic snowflake configuration. Work in collaboration with D.D. Ryutov, S.I. Krasheninnikov, and M.V. Umansky. Performed for the U.S. DoE by LLNS, LLC, LLNL, under Contract DE-AC52-07NA27344.

  10. Energy Efficient Triple IG Automation EEE (Triple-E)

    Energy Technology Data Exchange (ETDEWEB)

    McGlinchy, Timothy B

    2013-02-28

    GED Integrated Solutions collaborated with US window and door manufactures to investigate, design and verify technical and cost feasibility for producing high performance, high volume, low material and labor cost window, utilizing a modified window design containing a triple insulating glass unit (IGU). This window design approach when combined with a high volume IGU manufacturing system, can produce R5 rated windows for an approximate additional consumer cost of only $4 per square foot when compared to conventional Low-E argon dual pane IG windows, resulting in a verify practical, reliable and affordable high performance window for public use.

  11. Mobile probes

    DEFF Research Database (Denmark)

    2016-01-01

    A project investigating the effectiveness of a collection of online resources for teachers' professional development used mobile probes as a data collection method. Teachers received questions and tasks on their mobile in a dialogic manner while in their everyday context as opposed to in an inter......A project investigating the effectiveness of a collection of online resources for teachers' professional development used mobile probes as a data collection method. Teachers received questions and tasks on their mobile in a dialogic manner while in their everyday context as opposed...... to in an interview. This method provided valuable insight into the contextual use, i.e. how did the online resource transfer to the work practice. However, the research team also found that mobile probes may provide the scaffolding necessary for individual and peer learning at a very local (intra-school) community...... level. This paper is an initial investigation of how the mobile probes process proved to engage teachers in their efforts to improve teaching. It also highlights some of the barriers emerging when applying mobile probes as a scaffold for learning....

  12. 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.

  13. A new radiation-hard endoscope for divertor spectroscopy on JET

    Energy Technology Data Exchange (ETDEWEB)

    Huber, A., E-mail: A.Huber@fz-juelich.de [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich, EURATOM Association, Trilateral Euregio Cluster, D-52425 Jülich (Germany); Brezinsek, S.; Mertens, Ph.; Schweer, B.; Sergienko, G.; Terra, A. [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich, EURATOM Association, Trilateral Euregio Cluster, D-52425 Jülich (Germany); Arnoux, G.; Balshaw, N. [Euratom-CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Clever, M. [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich, EURATOM Association, Trilateral Euregio Cluster, D-52425 Jülich (Germany); Edlingdon, T. [Euratom-CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Egner, S. [Kayser-Threde GmbH, D-81379 Munich (Germany); Farthing, J. [Euratom-CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Hartl, M. [Kayser-Threde GmbH, D-81379 Munich (Germany); Horton, L. [EFDA-JET Close Support Unit, Culham Science Centre, Culham OX14 3DB (United Kingdom); Kampf, D. [Kayser-Threde GmbH, D-81379 Munich (Germany); Klammer, J. [KRP-Mechatec Engineering GbR, D-85748 Garching b. Muenchen (Germany); Lambertz, H.T. [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich, EURATOM Association, Trilateral Euregio Cluster, D-52425 Jülich (Germany); Matthews, G.F. [Euratom-CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Morlock, C.; Murari, A. [EFDA-JET Close Support Unit, Culham Science Centre, Culham OX14 3DB (United Kingdom); and others

    2013-10-15

    Highlights: ► A new radiation-hard endoscope with optimised divertor view has been developed on JET. ► A high optical transmittance (≥30%) in the operating wavelength range from 390 nm to 2500 nm has been achieved. ► The endoscope delivers high spatial resolution ≤2 mm at the object plane and ≤3 mm over the whole depth of field (±0.7 m). ► The new optical design includes options for the in situ calibration of the endoscope transmittance. ► A new type of shutter based on pneumatic techniques has been developed in view of ITER and integrated into the endoscope. -- Abstract: In preparation for ITER, JET has been upgraded with a new ITER-like wall (ILW) whereby the main plasma-facing components, previously made of carbon, have been replaced by Be in the main chamber and W in the divertor. A new endoscope with optimised divertor view has been developed in order to survey and monitor the emission of specific impurities such as tungsten, beryllium and the possibly remaining carbon in the tungsten divertor of the JET-ILW. It operates in the wavelength range from 390 nm to 2500 nm with high optical transmittance (≥30%) as well as high spatial resolution, that is ≤2 mm at the object plane and ≤3 mm over the whole depth of field (±0.7 m). The endoscope is a prototype for testing an ITER relevant design concept based on reflective optics only. The endoscope has an optimised observation in the near ultraviolet and in the blue spectral region to ensure the detection of the W I-emission line at 400.8 nm. In parallel to the new optical design, a new type of ITER-like shutter system based on pneumatic techniques has been developed and integrated in the endoscope head. The new optical design includes options for an in situ calibration of the endoscope transmittance during the experimental campaign.

  14. 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

    . 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......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...

  15. Edge turbulence and transport studies with ergodic divertor, on Tore Supra ohmic discharges

    Energy Technology Data Exchange (ETDEWEB)

    Payan, J.; Garbet, X.; Clairet, F.; Devynck, P.; Laviron, C.; Chatenet, J.H.; Ghendrih, P.N.; Grosman, A. [Association Euratom-CEA, Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Gervais, F.; Hennequin, P.; Quemeneur, A.; Truc, A. [Ecole Polytechnique, 91 - Palaiseau (France). Lab. de Physique des Milieux Ionises

    1995-12-31

    Edge turbulence and transport studies have been performed when the ergodic divertor is applied on Tore Supra ohmic discharges. A modification of radial electric field profiles is expected. Such a change could influence edge transport and turbulence. A CO{sub 2} laser scattering diagnostic, ALTAIR, has been used to study the turbulence changes at the plasma edge. Reflectometry (used at fixed frequency) gives also access to localized turbulence measurements. Preliminary results from reflectometry are presented and compared to ALTAIR results. (K.A.) 6 refs.; 4 figs.

  16. 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

  17. High heat flux plasma generator for new divertor plasma simulator in Nagoya University

    Energy Technology Data Exchange (ETDEWEB)

    Narita, S.; Ezumi, N.; Ohno, N.; Uesugi, Y.; Takamura, S. [Nagoya Univ. (Japan)

    1997-12-31

    A new divertor simulator called NAGDIS-II has been constructed in order to investigate edge plasma physics in fusion devices. Improved TP-D type plasma source, which consists of LaB{sub 6} cathode with a Mo hollow shield and external heating system, water-cooled intermediate electrode and anode was employed to make a high density plasma in the NAGDIS-II. The performance and reliability of the discharge system was confirmed by quantitatively measuring neutral pressure, heating efficiency and plasma parameters. (author)

  18. The relation of edge confinement to global confinement in ASDEX Upgrade (Axially Symmetric Divertor Experiment)

    Energy Technology Data Exchange (ETDEWEB)

    Pitcher, C.S.; Boozer, A.H.; Murmann, H.; Schweinzer, J.; Suttrop, W.; Salzmann, H. [Max-Planck-Institut fuer Plasmaphysik, IPP-EURATOM Association, D--85748 Garching (Germany); ASDEX Upgrade Team% NBI Group

    1997-07-01

    Experimental evidence is presented from the ASDEX Upgrade (Axially Symmetric Divertor Experiment) tokamak [{ital Plasma Physics and Controlled Nuclear Fusion Research 1993} (International Atomic Energy Agency, Vienna, 1994), Vol. I, p. 127] of a robust relation between the edge radial pressure gradient and the global confinement of the plasma. This relation transcends the power flowing across flux surfaces near the edge and thus suggests that the usual model of cross-field heat transport, where local gradients increase with increasing local power flow, is not appropriate. {copyright} {ital 1997 American Institute of Physics.}

  19. The role of ''momentum removal'' in divertor detachment

    Energy Technology Data Exchange (ETDEWEB)

    Kukushkin, A.S. [Kurchatov Institute, Moscow (Russian Federation); NRNU MEPhI, Moscow (Russian Federation); Pacher, H.D. [INRS-EMT, Varennes, Quebec (Canada)

    2016-08-15

    The role of ''momentum removal'' (the drag force on the plasma ion flow) in divertor detachment is considered and analysed in detail. This analysis of the 2D modelling results shows that the drag force cannot reduce the power and particle flux to the target directly. However, it is essential for creating the conditions for efficient radiation and volumetric plasma recombination, which in turn do the job. (copyright 2015 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  20. Simulation of radiative divertor plasmas by Ar seeding with the full W-wall in JT-60SA

    Energy Technology Data Exchange (ETDEWEB)

    Kawashima, H.; Shimizu, K.; Nakano, T.; Asakura, N. [Japan Atomic Energy Agency, Naka, Ibaraki (Japan); Hoshino, K. [Japan Atomic Energy Agency, Rokkasho, Aomori (Japan)

    2016-08-15

    Radiative divertor plasmas for JT-60SA with a full tungsten (W) wall, which is one of options in future, have been simulated with a SOL/divertor integrated code, SONIC. A conventional modified-coronal radiation (MCR) model with a finite confinement time is used for both Ar and W for the purpose of wide-range parameter surveys for the divertor plasma to obtain the required conditions (q{sub t} ≤ 10 MW/m{sup 2}, n{sup Sep}{sub e-mid} = 3∝8 x 10{sup 19} m{sup -3}, P{sub rad} < ∝30 MW), saving the calculation time. At low W density ratio (n{sub W}/n{sub i} = 1 x 10{sup -5}), due to low radiative power from W ions, Ar density ratio (n{sub Ar}/n{sub i} ≥ 1.0 x 10{sup -3}) and a strong gas puff (Γ{sub p} ≥ 3.0 x 10{sup 22} s{sup -1}) are inevitable to suppress the divertor heat flux down to 10 MW/m{sup 2}. Increasing n{sub W}/n{sub i} to 1 x 10{sup -3} in the divertor region, the divertor heat load becomes low and the operative regions are expanded. While, the W production shall be suppressed since the W radiation is increased with replacement of Ar radiation and the particle recycling decreased. A Monte-Carlo module (IMPMC) implemented in SONIC for Ar seeding reveals that the spatial distribution of Ar ions is predominantly determined by shell structures of the Ar ions. The consistency between IMPMC and MCR calculations is demonstrated for the averaged n{sub Ar}/n{sub i} ratio, the electron density and temperature profiles on the divertor target and typical parameter such as the divertor heat load. It shows that the detailed analysis with IMPMC model can be speedily obtained, using a steady state solution obtained by MCR model as an initial state. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  1. An Analysis of Training, Generalization, and Maintenance Effects of Primary Care Triple P for Parents of Preschool-Aged Children with Disruptive Behavior

    Science.gov (United States)

    Boyle, Cynthia L.; Sanders, Matthew R.; Lutzker, John R.; Prinz, Ronald J.; Shapiro, Cheri; Whitaker, Daniel J.

    2010-01-01

    A brief primary care intervention for parents of preschool-aged children with disruptive behavior was assessed using a multiple probe design. Primary Care Triple P, a four session behavioral intervention was sequentially introduced within a multiple probe format to each of 9 families to a total of 10 children aged between 3 and 7 years (males = 4,…

  2. Generalized derivations of Lie triple systems

    Directory of Open Access Journals (Sweden)

    Zhou Jia

    2016-01-01

    Full Text Available In this paper, we present some basic properties concerning the derivation algebra Der (T, the quasiderivation algebra QDer (T and the generalized derivation algebra GDer (T of a Lie triple system T, with the relationship Der (T ⊆ QDer (T ⊆ GDer (T ⊆ End (T. Furthermore, we completely determine those Lie triple systems T with condition QDer (T = End (T. We also show that the quasiderivations of T can be embedded as derivations in a larger Lie triple system.

  3. Conductivity Probe

    Science.gov (United States)

    2008-01-01

    The Thermal and Electrical Conductivity Probe (TECP) for NASA's Phoenix Mars Lander took measurements in Martian soil and in the air. The needles on the end of the instrument were inserted into the Martian soil, allowing TECP to measure the propagation of both thermal and electrical energy. TECP also measured the humidity in the surrounding air. The needles on the probe are 15 millimeters (0.6 inch) long. The Phoenix Mission is led by the University of Arizona, Tucson, on behalf of NASA. Project management of the mission is by NASA's Jet Propulsion Laboratory, Pasadena, Calif. Spacecraft development is by Lockheed Martin Space Systems, Denver.

  4. Pollution Probe.

    Science.gov (United States)

    Chant, Donald A.

    This book is written as a statement of concern about pollution by members of Pollution Probe, a citizens' anti-pollution group in Canada. Its purpose is to create public awareness and pressure for the eventual solution to pollution problems. The need for effective government policies to control the population explosion, conserve natural resources,…

  5. 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.

  6. Calculation of fractal dimension of magnetic footprint in double-null divertor tokamaks

    Science.gov (United States)

    Crank, Willie; Punjabi, Alkesh; Ali, Halima

    2010-11-01

    The simplest symplectic map that represents the magnetic topology of double-null divertor tokamaks is the double-null map, given by the map equations: x1=x0-ky0(1-y0^2 ), y1=y0+kx1. k is the map parameter. The map parameter k represents the generic topological effects of toroidal asymmetries. The O-point is at (0,0). The X-points are at (0,±1). We set k=0.51763, and Np=12. Np is the number of iterations of map that are equivalent to a single toroidal circuit of the tokamak. The width of stochastic layer near the upper and the lower X-points is exactly the same and equals 1.69 mm. We start 100,000 filed lines in the stochastic layer near the X-points and advance them for at most 10,000 toroidal circuits. We use the continuous analog of the map to calculate the magnetic footprints in the double-null divertor tokamaks. We calculate the area of the footprints and their fractal dimension. The area is A=0.0024 m^2, and fractal dimension is dfrac=1.0266. This work is supported by US Department of Energy grants DE-FG02-07ER54937, DE-FG02-01ER54624 and DE-FG02-04ER54793.

  7. Thermal fatigue characterization of CFC divertor modules using a one step brazing process

    Energy Technology Data Exchange (ETDEWEB)

    Pintsuk, G., E-mail: g.pintsuk@fz-juelich.de [Forschungszentrum Juelich, EURATOM Association, 52425 Juelich (Germany); Casalegno, V.; Ferraris, M. [Department of Applied Science and Technology, Politecnico di Torino, C.so Duca degli Abruzzi 24, I-10129 Torino (Italy); Koppitz, T. [Forschungszentrum Juelich, EURATOM Association, 52425 Juelich (Germany); Salvo, M. [Department of Applied Science and Technology, Politecnico di Torino, C.so Duca degli Abruzzi 24, I-10129 Torino (Italy)

    2012-07-15

    From the European side, three directional carbon fiber composites (CFCs) are foreseen to be used as plasma facing material for the strike point region of the initial ITER divertor installed for the non-tritium operational phase. For such divertor components two designs, the flat tile and the monoblock concept, are feasible, comprising a joint of the CFC with a Cu/Cu-alloy heat sink. This paper deals with the qualification of a reliable and cheap joining technology for such components, i.e. the simultaneous joining of the CuCrZr heat sink to a compliant Cu layer for the accommodation of thermal stresses and of the Cu layer and the CFC using a non-active Cu-Ge brazing material. For this purpose flat tile and monoblock mock-ups were manufactured, microstructurally analyzed, and subsequently exposed to cyclic high heat flux tests in the electron beam facility JUDITH. Applying hundreds of cycles at up to 20 MW/m{sup 2} the tested mock-ups underwent partial damaging, which was characterized in post-mortem microstructural investigations to analyze occurring degradation mechanisms, e.g. partial delamination at the CFC/Cu-interface.

  8. 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.

  9. Critical heat flux performance of hypervapotrons proposed for use in the ITER divertor vertical target

    Energy Technology Data Exchange (ETDEWEB)

    Youchison, D.L.; Marshall, T.D.; McDonald, J.M.; Lutz, T.J.; Watson, R.D. [Sandia National Labs., Albuquerque, NM (United States); Driemeyer, D.E. Kubik, D.L.; Slattery, K.T.; Hellwig, T.H. [McDonnell Douglas Aerospace, St. Louis, MO (United States)

    1997-09-01

    Task T-222 of the International Thermonuclear Experimental Reactor (ITER) program addresses the manufacturing and testing of permanent components for use in the ITER divertor. Thermalhydraulic and critical heat flux performance of the heat sinks proposed for use in the divertor vertical target are part of subtask T-222.4. As part of this effort, two single channel, medium scale, bare copper alloy, hypervapotron mockups were designed, fabricated, and tested using the EB-1200 electron beam system. The objectives of the effort were to develop the design and manufacturing procedures required for construction of robust high heat flux (HHF) components, verify thermalhydraulic, thermomechanical and critical heat flux (CHF) performance under ITER relevant conditions, and perform analyses of HHF data to identify design guidelines and failure criteria and possibly modify any applicable CHF correlations. The design, fabrication, and finite element modeling of two types of hypervapotrons are described; a common version already in use at the Joint European Torus (JET) and a new attached fin design. HHF test data on the attached fin hypervapotron will be used to compare the CHF performance under uniform heating profiles on long heated lengths with that of localized, highly peaked, off nominal profiles.

  10. 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.

    1997-08-01

    V-4Cr-4Ti alloy has been selected for use in the manufacture of a portion of the DIII-D Radiative Divertor upgrade. The production of a 1200-kg ingot of V-4Cr-4Ti alloy, and processing into final sheet and rod product forms suitable for components of the DIII-D Radiative Divertor Program (RDP), has been completed by Wah Chang (formerly Teledyne Wah Chang) of Albany, Oregon (WCA). CVN impact tests on sheet material indicate that the material has properties comparable to other previously-processed V-4Cr-4Ti and V-5Cr-5Ti alloys. Joining of V-4Cr-4Ti alloy has been identified as the most critical fabrication issue for its use in the RDP, and research into several joining methods for fabrication of the RDP components, including resistance seam, friction, and electron beam welding, and explosive bonding is being pursued. Preliminary trials have been successful in the joining of V-alloy to itself by resistance, friction, and electron beam welding processes, and to Inconel 625 by friction welding. In addition, an effort to investigate the explosive bonding of V-4Cr-4Ti alloy to Inconel 625, in both tube-to-bar and sheet-to-sheet configurations, has been initiated, and results have been encouraging.

  11. A program to evaluate the erosion on the CFC tiles of the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    D' Agata, E. [ITER International Team, ITER Joint Work Site, Boltzmannstr 2, 85748 Garching (Germany)], E-mail: elio.dagata@iter.org; Ogorodnikova, O.V. [Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 Saint-Paul-Lez-Durance (France); Tivey, R. [ITER International Team, ITER Joint Work Site, Boltzmannstr 2, 85748 Garching (Germany); Lowry, C.; Schlosser, J. [Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 Saint-Paul-Lez-Durance (France)

    2007-10-15

    The plasma-facing surfaces of the ITER divertor are armoured with tungsten in the upper part of the inner and outer vertical targets, and carbon fibre composite (CFC) in the lower part, the region where the scrape-off layer intercepts the divertor. The CFC in the form of a monoblock in the vertical target is the most loaded part of the plasma-facing surfaces, and hence it is subjected to high erosion and has a significant risk of failure. A program has been developed with the aim of understanding the impact on the erosion lifetime due to a combination of two main effects: the material property variations (particularly pronounced in CFC) and the presence of joining defects. The software allows the evolution of the surface profile of the armour to be predicted and the margin on critical heat flux at the heat-sink-to-coolant interface to be estimated for a range of postulated defects, from start-of-life through to end-of-life of the component. In assessing erosion, the code takes account of geometry and sublimation, and physical and chemical erosion of the CFC armour. The incident angle (a glancing angle of a few degrees) of the particle and heat flux onto the target is taken into account. The program has been validated by comparison with analytical approximations very well validated against experimental data. The code has been developed in the APDL language to operate inside a commercial and certificated finite element program such as ANSYS.

  12. Impact of ELM filaments on divertor heat flux dynamics in NSTX

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, J.-W., E-mail: jahn@pppl.gov [Oak Ridge National Laboratory, Oak Ridge (United States); Maingi, R. [Princeton Plasma Physics Laboratory, Princeton (United States); Canik, J.M. [Oak Ridge National Laboratory, Oak Ridge (United States); Gan, K.F. [Institute of Plasma Physics, Chinese Academy of Science, Hefei (China); Gray, T.K. [Oak Ridge National Laboratory, Oak Ridge (United States); McLean, A.G. [Lawrence Livermore National Laboratory, Livermore (United States)

    2015-08-15

    The ELM induced change in wetted area (A{sub wet}) and peak heat flux (q{sub peak}) of divertor heat flux is investigated as a function of the number of striations, which represent ELM filaments, observed in the heat flux profile in NSTX. More striations are found to lead to larger A{sub wet} and lower q{sub peak}. The typical number of striations observed in NSTX is 0–9, while 10–15 striations are normally observed in other machines such as JET, and the ELM contracts heat flux profile when the number of striations is less than 3–4 but broadens it with more of them. The smaller number of striations in NSTX is attributed to the fact that NSTX ELMs are against kink/peeling boundary with lower toroidal mode number (n = 1–5), while typical peeling–ballooning ELMs have higher mode number of n = 10–20. For ELMs with smaller number of striations, relative A{sub wet} change is rather constant and q{sub peak} change rapidly increases with increasing ELM size, while A{sub wet} change slightly increases leading to a weaker increase of q{sub peak} change for ELMs with larger number of striations, both of which are unfavourable trend for the material integrity of divertor tiles.

  13. 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)

  14. ASDEX Upgrade Edge Transport Studies by Turbulence and Braginskii Divertor Transport Codes

    Energy Technology Data Exchange (ETDEWEB)

    Nishimura, Y.; Coster, D.P.; Kim, J.W.; Scott, B.D. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Garching (Germany)

    2001-07-01

    The equilibration time for diverter transport simulations is in the range of milliseconds to seconds. There, perpendicular transport is given empirically and usually assumed to be constant in time and space. In this work, we aim at describing edge plasma profiles in both the H-mode and the L-mode confinement regimes using a model that couples the transport scale to the underlying turbulence scale. There are 2d and 3d variants of DALF, which is a turbulence code that describes short time scale nonlinear phenomena based on first principles of plasma physics. B2 employs an implicit method which is suitable for describing long time scale, quasi-steady state behavior, while fluctuation/intermittency is inherent in turbulence and typically gives rise to short time scale variations of the radial flux. We coarse rained the information from the 2d version of DALF within the order of turbulence auto correlation time and iterated over the divertor simulation (and thus passed plasma parameters to the turbulence code). Numerical algorithm and criteria for convergence in bridging the physics of two different scales is discussed. The generation mechanism of radial electric field in steep gradient regimes is revisited in the ASDEX Upgrade divertor geometry with realistic parameters. Inclusion of turbulent suppression effects by E x B shear flow is considered. (orig.)

  15. 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).

  16. Divertor load footprint of ELMs in pellet triggering and pacing experiments at JET

    Energy Technology Data Exchange (ETDEWEB)

    Frigione, D., E-mail: domenico.frigione@frascati.enea.it [Unità Tecnica Fusione, ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); Garzotti, L. [CCFE, Culham Science Centre, OX14 3DB (United Kingdom); Lennholm, M. [EFDA CSU, Culham Science Centre, OX14 3DB (United Kingdom); Alper, B. [CCFE, Culham Science Centre, OX14 3DB (United Kingdom); Artaserse, G. [Unità Tecnica Fusione, ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); Bennett, P. [CCFE, Culham Science Centre, OX14 3DB (United Kingdom); Giovannozzi, E. [Unità Tecnica Fusione, ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); Eich, T. [Max Planck Institute for Plasma Physics, Garching (Germany); Kocsis, G. [WIGNER RCP RMI, POB 49, 1525 Budapest (Hungary); Lang, P.T. [Max Planck Institute for Plasma Physics, Garching (Germany); Maddaluno, G. [Unità Tecnica Fusione, ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); Mooney, R. [CCFE, Culham Science Centre, OX14 3DB (United Kingdom); Rack, M. [Institut für Energieforschung – Plasmaphysik, Forschungszentrum Jülich, 52425 Jülich (Germany); Sips, G. [EFDA CSU, Culham Science Centre, OX14 3DB (United Kingdom); Tvalashvili, G. [CCFE, Culham Science Centre, OX14 3DB (United Kingdom); Viola, B. [Unità Tecnica Fusione, ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); Wilkes, D. [CCFE, Culham Science Centre, OX14 3DB (United Kingdom)

    2015-08-15

    An investigation of pellet pacing and triggering of Edge Localized Modes (ELMs) was carried out in the frame of ELM mitigation studies aimed at reducing their damaging effects on the plasma-facing components (PFCs). The divertor power load footprint of triggered ELMs was compared with gas puffing controlled ELMs. Small pellets, corresponding to a few per cent of the target plasma particle inventory, were used to minimize the fueling effect and the total particle throughput. There is no evidence that pellets can reduce the divertor power load with respect to gas fueling when operating at the same ELM frequency. The line average density and the energy confinement time remained constant when the gas was progressively substituted by pellets. The launch from the Vertical High Field Side (VHFS) confirmed to be more efficient in ELM triggering than from the Low Field Side (LFS) while the power load footprint remained the same both in time evolution and in spatial distribution when changing the injection geometry.

  17. Numerical Simulation of the Neutralized α Particle Transport near the Divertor Plate Region

    Institute of Scientific and Technical Information of China (English)

    李承跃

    2012-01-01

    The statistical random sample technique has been utilized to develop a new Monte-Carlo algorithm MCHET code recently. A large amount of comparative simulation calculation work relating to the neutralized alpha-particle transport has been performed. As a result, we have found the beneficial optimizing plasma density and temperature profiles in the divertor region, with the great resulting improvement of helium ash removal efficiency by the simultaneously externally applied proper RF ponderomotive force potential energy in the vicinity of the divertor plate region. In this work the dominant atomic processes of electron impact ionization and elastic scattering by plasma ions are included. The thermal and streaming motion of the ions along the magnetic field is taken into consideration. Important conclusions are obtained that the probability of neutral helium turning back to the target plate will increase at least by 50% for the optimized combination of the beneficial density, temperature profiles and proper RF perpendicular electric field. For FEB (Fusion Experimental Breeder) reactor design parameters, the RF ponderomotive potential enhancement from 0.5 to 0.9 of ash removal efficiency can be obviously obtained. In the meantime, the tritium inventory may also be reduced to some extent.

  18. Critical heat flux performance of hypervapotrons proposed for use in the ITER divertor vertical target

    Science.gov (United States)

    Youchison, Dennis L.; Marshall, Theron D.; McDonald, Jimmie M.; Lutz, Thomas J.; Watson, Robert D.; Driemeyer, Daniel E.; Kubik, David L.; Slattery, Kevin T.; Hellwig, Theodore H.

    1997-12-01

    Task T-222 of the International Thermonuclear Experimental Reactor (ITER) program addresses the manufacturing and testing of permanent components for use in the ITER divertor. Thermal-hydraulic and critical heat flux performance of the heat sinks proposed for use in the divertor vertical target are part of subtask T-222.4. As part of this effort, two single channel, medium-scale, bare copper alloy, hypervapotron mock-ups were designed by Sandia National Laboratories and McDonnell Douglas Aerospace (MDA), fabricated at MDA and tested at Sandia' Plasma Materials Test Facility using the EB-1200 electron beam system. The objectives of our effort were to develop the design and manufacturing procedures required for construction of robust HHF components, verify thermal-hydraulic, thermomechanical and CHF performance under ITER relevant conditions, and perform analyses of HHF data to identify design guidelines, failure criteria and possibly modify any applicable CHF correlations. This paper describes the design, fabrication and finite elements modeling of two types of hypervapotrons, a common version already in use at JET and a new attached- fin design. HHF test data on the attached-fin hypervapotron will be used to compare the CHF performance under uniform heating profiles on long heated lengths to that of localized, highly peaked, off-nominal profiles.

  19. Operational limits on WEST inertial divertor sector during the early phase experiment

    Science.gov (United States)

    Firdaouss, M.; Corre, Y.; Languille, P.; Greuner, H.; Autissier, E.; Desgranges, C.; Guilhem, D.; Gunn, J. P.; Lipa, M.; Missirlian, M.; Pascal, J.-Y.; Pocheau, C.; Richou, M.; Tsitrone, E.

    2016-02-01

    The primary goal of the WEST project is to be a test bed to characterize the fatigue and lifetime of ITER-like W divertor components subjected to relevant thermal loads. During the first phase of exploitation (S2 2016), these components (W monoblock plasma facing unit—W-PFU) will be installed in conjunction with graphite components (G-PFU). Since the G-PFU will not be actively cooled, it is necessary to ensure the expected pulse duration allows the W-PFU to reach its steady state without overheating the G-PFU assembly structure or the embedded stainless-steel diagnostics. High heat flux tests were performed at the GLADIS facility to assess the thermal behavior of the G-PFU. Some operational limits based on plasma parameters were determined. It was found that it is possible to operate at an injected power such that the maximal incident heat flux on the lower divertor is 10 MW m-2 for the required pulse length.

  20. Simulation of disruptions on C-Mod in support of the new outer divertor project

    Science.gov (United States)

    Poli, F.; Kessel, C.; Titus, P.; Zhang, H.; Doody, J.; Granetz, R.; Lipschultz, B.

    2011-10-01

    Disruptions in C-Mod lead to large forces on structures inside the vacuum vessel and can be grouped in two classes depending on whether they begin with a thermal quench (midplane disruptions) or not (VDEs). VDEs induce the largest currents in the lower divertor, which is being re-designed to be toroidally continuous and allow operation at high temperatures (< 600C). Both types of disruptions have been simulated with TSC and the vector potential has been integrated in the ANSYS code (ANSYS® Multiphysics, Release 12.1) to calculate magnetic fields, induced currents in the structures of interest and forces. These forces are then used to calculate stress and deformation in the part. The TSC simulations are adjusted (thermal quench time, halo temperature and width, etc) to match the plasma characteristics as close as possible to experiments. The results of these simulations will be shown and the dependence of disruption time scales and characteristics on these plasma parameters and the new outer divertor structures will be discussed. This work is supported by the US Department of Energy under DE-AC02-CH0911466 and DE-FC02-99ER54512.

  1. Numerical investigation of disruption characteristics for the snowflake divertor configuration in HL-2M

    Science.gov (United States)

    Xue, L.; Duan, X. R.; Zheng, G. Y.; Liu, Y. Q.; Pan, Y. D.; Yan, S. L.; Dokuka, V. N.; Lukash, V. E.; Khayrutdinov, R. R.

    2016-05-01

    Cold and hot vertical displacement events (VDEs) are frequently related to the disruption of vertically-elongated tokamaks. The weak poloidal magnetic field around the null-points of a snowflake divertor configuration may influence the vertical displacement process. In this paper, the major disruption with a cold VDE and the vertical disruption in the HL-2M tokamak are investigated by the DINA code. In order to better illustrate the effect from the weak poloidal field, a double-null snowflake configuration is compared with the standard divertor (SD) configuration under the same plasma parameters. Computational results show that the weak poloidal magnetic field can be partly beneficial for mitigating the vertical instability of the plasma under small perturbations. For major disruption, the peak poloidal halo current fraction is almost the same between the snowflake and the SD configurations. However, this fraction becomes much larger for the snowflake in the event of a hot VDE. Furthermore, during the disruption for a snowflake configuration, the distribution of electromagnetic force on a vacuum vessel gets more non-uniform during the current quench.

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

    Science.gov (United States)

    Blommaert, Maarten; Dekeyser, Wouter; Baelmans, Martine; Gauger, Nicolas R.; Reiter, Detlev

    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.

  3. 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.

  4. ADX: a high field, high power density, Advanced Divertor test eXperiment

    Science.gov (United States)

    Vieira, R.; Labombard, B.; Marmar, E.; Irby, J.; Shiraiwa, S.; Terry, J.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.; ADX Team

    2014-10-01

    The MIT PSFC and collaborators are proposing an advanced divertor experiment (ADX) - a tokamak specifically designed to address critical gaps in the world fusion research program on the pathway to FNSF/DEMO. This high field (6.5 tesla, 1.5 MA), high power density (P/S ~ 1.5 MW/m2) facility would utilize Alcator magnet technology to test innovative divertor concepts for next-step DT fusion devices (FNSF, DEMO) at reactor-level boundary plasma pressures and parallel heat flux densities while producing high performance core plasma conditions. The experimental platform would also test advanced lower hybrid current drive (LHCD) and ion-cyclotron range of frequency (ICRF) actuators and wave physics at the plasma densities and magnetic field strengths of a DEMO, with the unique ability to deploy launcher structures both on the low-magnetic-field side and the high-field side - a location where energetic plasma-material interactions can be controlled and wave physics is most favorable for efficient current drive, heating and flow drive. This innovative experiment would perform plasma science and technology R&D necessary to inform the conceptual development and accelerate the readiness-for-deployment of FNSF/DEMO - in a timely manner, on a cost-effective research platform. Supported by DE-FC02-99ER54512.

  5. 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.

  6. Modeling Detached Divertor Plasma Characteristics in the DIII-D Tokamak

    Science.gov (United States)

    Rognlien, T. D.; Joseph, I.; McLean, A. G.; Porter, G. D.; Rensink, M. E.; Umansky, M.; Groth, M.; Pigarov, A. Y.

    2015-11-01

    Detached divertor-plasma operation, where a large fraction of the core exhaust power is radiated before striking the target plates, is attractive for limiting the peak target heat flux. Such plasmas have electron temperature ~ 1 eV near the target. Changing the position of the separatrix strike points on the geometrically varied DIII-D target plates is allowing a systematic study of how plate shape impacts accessibility to detached operation. Reported here are 2D plasma/neutral transport simulations of these configurations using the UEDGE code including cross-field drifts and impurities. Results are given on how the onset of detachment scales with strike-point location, wall pumping of neutrals, separatrix density, and core power. Different initial conditions sometimes yield different steady-state solutions for identical input parameters, one being an attached plasma and the other detached. Comparisons are made of simulation results and experimental measurements, especially divertor Thomson scattering data. Work supported by US DOE, DE-AC52-07NA27344, DE-FC02-04ER54698, DE-FG02-07ER54917.

  7. Studies of high- δ (baffled) and low- δ (open) pumped divertor operation on DIII-D

    Science.gov (United States)

    Allen, S. L.; Fenstermacher, M. E.; Greenfield, C. M.; Hyatt, A. W.; Maingi, R.; Porter, G. D.; Wade, M. R.; Bozek, A. S.; Ellis, R.; Hill, D. N.; Hollerbach, M. A.; Lasnier, C. J.; Leonard, A. W.; Mahdavi, M. A.; Nilson, D. G.; Petrie, T. W.; Schaffer, M. J.; Smith, J. P.; Stambaugh, R. D.; Thomas, D. M.; Watkins, J. G.; West, W. P.; Whyte, D. G.; Wood, R. D.

    We report new experimental results with the Radiative Divertor Project-outer baffle (RDP-OB) and cryopump in both upper single-null (USN) and double-null (DN) ELMing H-mode discharges. The baffled divertor reduced the core ionization (˜2-2.5×), in reasonable agreement with predictions from UEDGE/DEGAS modeling (˜3.75×). The upper cryopump achieved density control of ne/ ngw ˜ 0.22 (line density/Greenwald density) with Zeff ˜ 2 in high- δ plasmas. The measured exhaust is comparable to the lower pump, except at lower core electron densities ( ne < 5 × 10 19 m -3). Efficient impurity exhaust was obtained with deuterium SOL flow. Preliminary experiments with DN operation has shown that the particle exhaust to the upper pump depends on the up/down magnetic balance. Preliminary experiments indicate that the DN exhaust is roughly 40-50% of the USN exhaust at ne ˜ 4 × 10 19 m -3.

  8. Thermal fatigue characterization of CFC divertor modules using a one step brazing process

    Science.gov (United States)

    Pintsuk, G.; Casalegno, V.; Ferraris, M.; Koppitz, T.; Salvo, M.

    2012-07-01

    From the European side, three directional carbon fiber composites (CFCs) are foreseen to be used as plasma facing material for the strike point region of the initial ITER divertor installed for the non-tritium operational phase. For such divertor components two designs, the flat tile and the monoblock concept, are feasible, comprising a joint of the CFC with a Cu/Cu-alloy heat sink. This paper deals with the qualification of a reliable and cheap joining technology for such components, i.e. the simultaneous joining of the CuCrZr heat sink to a compliant Cu layer for the accommodation of thermal stresses and of the Cu layer and the CFC using a non-active Cu-Ge brazing material. For this purpose flat tile and monoblock mock-ups were manufactured, microstructurally analyzed, and subsequently exposed to cyclic high heat flux tests in the electron beam facility JUDITH. Applying hundreds of cycles at up to 20 MW/m2 the tested mock-ups underwent partial damaging, which was characterized in post-mortem microstructural investigations to analyze occurring degradation mechanisms, e.g. partial delamination at the CFC/Cu-interface.

  9. 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.; Trester, P.W.

    1997-04-01

    V-4Cr-4Ti alloy has been selected for use in the manufacture of a portion of the DIII-D Radiative Divertor upgrade. The production of a 1200-kg ingot of V-4Cr-4Ti alloy, and processing into final sheet and rod product forms suitable for components of the DIII-D Radiative Divertor structure, has been completed at Wah Chang (formerly Teledyne Wah Chang) of Albany, Oregon (WCA). 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. An effort to investigate the explosive bonding of V-4Cr-4Ti alloy to Inconel 625 has also been initiated, and results have been encouraging. In addition, preliminary tests have been completed to evaluate the susceptibility of V-4Cr-4Ti alloy to stress corrosion cracking in DIII-D cooling water, and the effects of exposure to DIII-D bakeout conditions on the tensile and fracture behavior of V-4Cr-4Ti alloy.

  10. 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

  11. Holonomy loops, spectral triples and quantum gravity

    DEFF Research Database (Denmark)

    Johannes, Aastrup; Grimstrup, Jesper Møller; Nest, Ryszard

    2009-01-01

    We review the motivation, construction and physical interpretation of a semi-finite spectral triple obtained through a rearrangement of central elements of loop quantum gravity. The triple is based on a countable set of oriented graphs and the algebra consists of generalized holonomy loops...

  12. Spectral triples and the geometry of fractals

    DEFF Research Database (Denmark)

    Christensen, Erik; Ivan, Cristina; Schroe, Elmar

    2012-01-01

    It is shown that one can construct a spectral triple for the Sierpinski gasket such that it represents any given K-homology class, On the other hand if the geodesic distance and the dimension has to be part of the data from the triple, there are certain restriction....

  13. Existence Regions of Shock Wave Triple Configurations

    Science.gov (United States)

    Bulat, Pavel V.; Chernyshev, Mikhail V.

    2016-01-01

    The aim of the research is to create the classification for shock wave triple configurations and their existence regions of various types: type 1, type 2, type 3. Analytical solutions for limit Mach numbers and passing shock intensity that define existence region of every type of triple configuration have been acquired. The ratios that conjugate…

  14. The Killing Forms of Lie Triple Systems

    Institute of Scientific and Technical Information of China (English)

    ZHANG Zhi Xue; GAO Rui

    2009-01-01

    For Lie triple systems in the characteristic zero setting, we obtain by means of the Killing forms two criterions for semisimplicity and for solvability respectively, and then investigate the relationship among the Killing forms of a real Lie triple system To, the complexification T of To, and the realification of T.

  15. Nonunital Spectral Triples Associated to Degenerate Metrics

    Science.gov (United States)

    Rennie, A.

    We show that one can define (p,∞)-summable spectral triples using degenerate metrics on smooth manifolds. Furthermore, these triples satisfy Connes-Moscovici's discrete and finite dimension spectrum hypothesis, allowing one to use the Local Index Theorem [1] to compute the pairing with K-theory. We demonstrate this with a concrete example.

  16. Determination of divertor stray light in high-resolution main chamber H α spectroscopy in JET-ILW

    Science.gov (United States)

    Neverov, V. S.; Kukushkin, A. B.; Stamp, M. F.; Alekseev, A. G.; Brezinsek, S.; von Hellermann, M.; Contributors, JET

    2017-01-01

    The theoretical model suggested for ITER main chamber H α spectroscopy is applied to the high-resolution spectroscopy (HRS) data of recent JET ITER-like wall (ILW) experiments. The model is aimed at reconstructing the neutral hydrogen isotope density in the SOL, as well as the isotope ratio, by solving a multi-parametric inverse problem with allowance for (i) the strong divertor stray light (DSL) on the main-chamber lines of sight (LoS), (ii) substantial deviation of the neutral atom velocity distribution function (VDF) from a Maxwellian in the SOL, and (iii) data for the direct observation of the divertor. The JET-ILW HRS data on resolving the power at the deuterium and hydrogen spectral lines of the Balmer-alpha series is analysed, with direct observation of the divertor from the top and with observation of the inner wall along the tangential and radial LoS from the equatorial ports. This data allows the spectrum of the DSL and the signal-to-background ratio for the Balmer-alpha light emitted from the far SOL and divertor in the JET-ILW to be evaluated. The results support the expectation of the strong impact of the DSL upon the ITER main chamber H α (and visible light) spectroscopy diagnostics.

  17. 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.

  18. A review on two previous divertor target concepts for DEMO: mutual impact between structural design requirements and materials performance

    Science.gov (United States)

    You, Jeong-Ha

    2015-09-01

    Development of a diverter target with a sufficient capability of power exhaust is a crucial prerequisite for the realization of a fusion power plant. While the design and technology for divertor target has been successfully developed for ITER, the applicability of this concept is not necessarily assured yet for DEMO mainly because the neutron irradiation dose expected for the DEMO divertor will be an order of magnitude higher than that of the ITER divertor. The possible embrittlement of structural heat sink materials due to irradiation is likely to restrict the structural performance and the operational flexibility of a target component to a considerable extent. For judgment of design feasibility of a target concept a quantitative evaluation of the thermal and structure mechanical performance is needed. In this article, a review on two representative target design concepts considered for the DEMO divertor is presented. Emphasis is put on the mutual impact between the design requirements and the performance of structural materials. Water-cooled and helium-cooled concepts are discussed considering two baseline heat sink materials, CuCrZr alloy and tungsten, respectively. Conclusions are derived from the critical features of the heat sink performance in terms of structural reliability, design/material interface and further R&D needs.

  19. Modeling of divertor power footprint widths on EAST by SOLPS5.0/B2.5-Eirene

    Science.gov (United States)

    Guozhong, DENG; Xiaoju, LIU; Liang, WANG; Shaocheng, LIU; Jichan, XU; Wei, FENG; Jianbin, LIU; Huan, LIU; Xiang, GAO

    2017-04-01

    The edge plasma code package SOLPS5.0 is employed to simulate the divertor power footprint widths of the experimental advanced superconducting tokamak (EAST) L-mode and ELM-free H-mode plasmas. The divertor power footprint widths, which consist of the scrape-off layer (SOL) width λ q and heat spreading S, are important physical parameters for edge plasmas. In this work, a plasma current scan is implemented in the simulation to obtain the dependence of the divertor power footprint width on the plasma current I p. Strong inverse scaling of the SOL width with I p has been achieved for both L-mode and H-mode plasmas in the forms of {λ }q,{{L}\\text-\\text{mode}}=4.98× {I}{{p}}-0.68 and {λ }q,{{H}\\text-\\text{mode}}=1.86× {I}{{p}}-1.08. Similar trends have also been demonstrated in the study of heat spreading with {S}{{L}\\text-\\text{mode}}=1.95× {I}{{p}}-0.542 and {S}{{H}\\text-\\text{mode}}=0.756× {I}{{p}}-0.872. In addition, studies on divertor peak heat load and the magnetic flux expansion factor show that both of them are proportional to plasma current. The simulation work here can act as a way to explore the power footprint widths of future tokamak fusion devices such as ITER and the China Fusion Engineering Test Reactor (CFETR).

  20. Numerical Study of High Heat Flux Performances of Flat-Tile Divertor Mock-ups with Hypervapotron Cooling Concept

    Science.gov (United States)

    Chen, Lei; Liu, Xiang; Lian, Youyun; Cai, Laizhong

    2015-09-01

    The hypervapotron (HV), as an enhanced heat transfer technique, will be used for ITER divertor components in the dome region as well as the enhanced heat flux first wall panels. W-Cu brazing technology has been developed at SWIP (Southwestern Institute of Physics), and one W/CuCrZr/316LN component of 450 mm×52 mm×166 mm with HV cooling channels will be fabricated for high heat flux (HHF) tests. Before that a relevant analysis was carried out to optimize the structure of divertor component elements. ANSYS-CFX was used in CFD analysis and ABAQUS was adopted for thermal-mechanical calculations. Commercial code FE-SAFE was adopted to compute the fatigue life of the component. The tile size, thickness of tungsten tiles and the slit width among tungsten tiles were optimized and its HHF performances under International Thermonuclear Experimental Reactor (ITER) loading conditions were simulated. One brand new tokamak HL-2M with advanced divertor configuration is under construction in SWIP, where ITER-like flat-tile divertor components are adopted. This optimized design is expected to supply valuable data for HL-2M tokamak. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2011GB110001 and 2011GB110004)

  1. Multi-pin Langmuir probe measurement for identification of blob propagation characteristics in the Large Helical Device

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, H., E-mail: tanaka.hirohiko@lhd.nifs.ac.jp [National Institute for Fusion Science, Toki 509-5292 (Japan); Masuzaki, S. [National Institute for Fusion Science, Toki 509-5292 (Japan); Ohno, N. [Graduate School of Engineering, Nagoya University, Nagoya, Aichi 464-8603 (Japan); Morisaki, T. [National Institute for Fusion Science, Toki 509-5292 (Japan); Tsuji, Y. [Graduate School of Engineering, Nagoya University, Nagoya, Aichi 464-8603 (Japan)

    2015-08-15

    In order to investigate the blobby plasma transport in the Large Helical Device, we have measured electrostatic fluctuations around the divertor leg by using a newly-designed multi-pin reciprocating Langmuir probe. Near the low-field side edge of the divertor leg, positive spikes of ion saturation current fluctuation were observed. In addition, the electric field which correlates with the ion saturation current fluctuation was firstly evaluated with the neighboring floating potential measurement. Considering the positional relationship with the magnetic geometry, the identified direction of the electric field inside the blobs is consistent with the theoretically predicted E × B motion. By applying the conditional averaging method, a quantitative speed of the blobs was preliminary estimated.

  2. Contractive spectral triples for crossed products

    CERN Document Server

    Paterson, Alan L T

    2012-01-01

    Connes showed that spectral triples encode (noncommutative) metric information. Further, Connes and Moscovici in their metric bundle construction showed that, as with the Takesaki duality theorem, forming a crossed product spectral triple can substantially simplify the structure. In a recent paper, Bellissard, Marcolli and Reihani (among other things) studied in depth metric notions for spectral triples and crossed product spectral triples for $Z$-actions, with applications in number theory and coding theory. In the work of Connes and Moscovici, crossed products involving groups of diffeomorphisms and even of \\'{e}tale groupoids are required. With this motivation, the present paper develops part of the Bellissard-Marcolli-Reihani theory for a general discrete group action, and in particular, introduces coaction spectral triples and their associated metric notions. The isometric condition is replaced by the contractive condition.

  3. Three-dimensional modeling of plasma edge transport and divertor fluxes during application of resonant magnetic perturbations on ITER

    Science.gov (United States)

    Schmitz, O.; Becoulet, M.; Cahyna, P.; Evans, T. E.; Feng, Y.; Frerichs, H.; Loarte, A.; Pitts, R. A.; Reiser, D.; Fenstermacher, M. E.; Harting, D.; Kirschner, A.; Kukushkin, A.; Lunt, T.; Saibene, G.; Reiter, D.; Samm, U.; Wiesen, S.

    2016-06-01

    Results from three-dimensional modeling of plasma edge transport and plasma-wall interactions during application of resonant magnetic perturbation (RMP) fields for control of edge-localized modes in the ITER standard 15 MA Q  =  10 H-mode are presented. The full 3D plasma fluid and kinetic neutral transport code EMC3-EIRENE is used for the modeling. Four characteristic perturbed magnetic topologies are considered and discussed with reference to the axisymmetric case without RMP fields. Two perturbation field amplitudes at full and half of the ITER ELM control coil current capability using the vacuum approximation are compared to a case including a strongly screening plasma response. In addition, a vacuum field case at high q 95  =  4.2 featuring increased magnetic shear has been modeled. Formation of a three-dimensional plasma boundary is seen for all four perturbed magnetic topologies. The resonant field amplitudes and the effective radial magnetic field at the separatrix define the shape and extension of the 3D plasma boundary. Opening of the magnetic field lines from inside the separatrix establishes scrape-off layer-like channels of direct parallel particle and heat flux towards the divertor yielding a reduction of the main plasma thermal and particle confinement. This impact on confinement is most accentuated at full RMP current and is strongly reduced when screened RMP fields are considered, as well as for the reduced coil current cases. The divertor fluxes are redirected into a three-dimensional pattern of helical magnetic footprints on the divertor target tiles. At maximum perturbation strength, these fingers stretch out as far as 60 cm across the divertor targets, yielding heat flux spreading and the reduction of peak heat fluxes by 30%. However, at the same time substantial and highly localized heat fluxes reach divertor areas well outside of the axisymmetric heat flux decay profile. Reduced RMP amplitudes due to screening or reduced RMP

  4. Benchmarking of a 1D Scrape-off layer code SOLF1D with SOLPS and its use in modelling long-legged divertors

    CERN Document Server

    Havlickova, E; Subba, F; Coster, D; Wischmeier, M; Fishpool, G

    2013-01-01

    A 1D code modelling SOL transport parallel to the magnetic field (SOLF1D) is benchmarked with 2D simulations of MAST-U SOL performed via the SOLPS code for two different collisionalities. Based on this comparison, SOLF1D is then used to model the effects of divertor leg stretching in 1D, in support of the planned Super-X divertor on MAST. The aim is to separate magnetic flux expansion from volumetric power losses due to recycling neutrals by stretching the divertor leg either vertically or radially.

  5. Switchable proline derivatives: tuning the conformational stability of the collagen triple helix by pH changes.

    Science.gov (United States)

    Siebler, Christiane; Erdmann, Roman S; Wennemers, Helma

    2014-09-22

    (4S)-Aminoproline is introduced as a pH-sensitive probe for tuning the conformational properties of peptides and proteins. The pH-triggered flip of the ring puckering and the formation/release of a transannular H bond were used to switch the formation of collagen triple helices on and off reversibly.

  6. 2D divertor heat flux distribution using a 3D heat conduction solver in National Spherical Torus Experiment.

    Science.gov (United States)

    Gan, K F; Ahn, J-W; Park, J-W; Maingi, R; McLean, A G; Gray, T K; Gong, X; Zhang, X D

    2013-02-01

    The divertor heat flux footprint in tokamaks is often observed to be non-axisymmetric due to intrinsic error fields, applied 3D magnetic fields or during transients such as edge localized modes. Typically, only 1D radial heat flux profiles are analyzed; however, analysis of the full 2D divertor measurements provides opportunities to study the asymmetric nature of the deposited heat flux. To accomplish this an improved 3D Fourier analysis method has been successfully applied in a heat conduction solver (TACO) to determine the 2D heat flux distribution at the lower divertor surface in the National Spherical Torus Experiment (NSTX) tokamak. This advance enables study of helical heat deposition onto the divertor. In order to account for heat transmission through poorly adhered surface layers on the divertor plate, a heat transmission coefficient, defined as the surface layer thermal conductivity divided by the thickness of the layer, was introduced to the solution of heat conduction equation. This coefficient is denoted as α and a range of values were tested in the model to ensure a reliable heat flux calculation until a specific value of α led to the constant total deposited energy in the numerical solution after the end of discharge. A comparison between 1D heat flux profiles from TACO and from a 2D heat flux calculation code, THEODOR, shows good agreement. Advantages of 2D heat flux distribution over the conventional 1D heat flux profile are also discussed, and examples of 2D data analysis in the study of striated heat deposition pattern as well as the toroidal degree of asymmetry of peak heat flux and heat flux width are demonstrated.

  7. Modeling of carbon transport in the divertor and SOL of DIII-D during high performance plasma operation

    Science.gov (United States)

    West, W. P.; Porter, G. D.; Evans, T. E.; Stangeby, P.; Brooks, N. H.; Fenstermacher, M. E.; Isler, R. C.; Rognlien, T. D.; Wade, M. R.; Whyte, D. G.; Wolf, N. S.

    2001-03-01

    The UEDGE modeling code has been used to study the effect of varying the carbon yield from the plasma facing surfaces on the core plasma carbon contamination in DIII-D. The model of the lower single-null, ELMing H-mode plasma shows a remarkably weak dependence of the core carbon concentration over an approximate factor of two variation in the source. This weak dependence is in agreement with the analysis of spectroscopic data from DIII-D [1]. Examination of the carbon transport shows a general flow pattern of carbon as follows: (1) parallel flow from the divertors to the near scrape off layer (SOL) near the separatrix, (2) cross field diffusion from the near SOL to the far SOL (near the wall), and (3) parallel flow from the far SOL to the far region of the inner divertor. The carbon flux from the divertors to the near SOL drops as the sputtering rate is reduced. In the far SOL, background plasma parameters adjust in small ways to produce an increasing carbon density with decreasing sputtering yield. This increasing density of carbon in the far SOL is consistent with a reduction in the parallel velocity of carbon ions flowing from the far SOL back to the inner divertor. Since the carbon density near the separatrix is constant as the sputtering yield is reduced, the increasing density in the far SOL reduces the radial gradient and therefore the diffusive radial flow. A balance in the outward radial diffusive flow from the near SOL and the flow from the divertor into the near SOL maintains the carbon density in the near SOL nearly constant, even though the carbon throughput changes.

  8. Divertor Heat Flux Mitigation in High-Performance H-mode Discharges in the National Spherical Torus Experiment.

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V A; Maingi, R; Gates, D; Menard, J

    2008-12-31

    Experiments conducted in high-performance 1.0 MA and 1.2 MA 6 MW NBI-heated H-mode discharges with a high magnetic flux expansion radiative divertor in NSTX demonstrate that significant divertor peak heat flux reduction and access to detachment may be facilitated naturally in a highly-shaped spherical torus (ST) configuration. Improved plasma performance with high {beta}{sub t} = 15-25%, a high bootstrap current fraction f{sub BS} = 45-50%, longer plasma pulses, and an H-mode regime with smaller ELMs has been achieved in the strongly-shaped lower single null configuration with elongation {kappa} = 2.2-2.4 and triangularity {delta} = 0.6-0.8. Divertor peak heat fluxes were reduced from 6-12 MW/m{sup 2} to 0.5-2 MW/m{sup 2} in ELMy H-mode discharges using the inherently high magnetic flux expansion f{sub m} = 16-25 and the partial detachment of the outer strike point at several D{sub 2} injection rates. A good core confinement and pedestal characteristics were maintained, while the core carbon concentration and the associated Z{sub eff} were reduced. The partially detached divertor regime was characterized by an increase in divertor radiated power, a reduction of ion flux to the plate, and a large neutral compression ratio. Spectroscopic measurements indicated a formation of a high-density, low temperature region adjacent to the outer strike point, where substantial increases in the volume recombination rate and CII, CIII emission rates was measured.

  9. 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.

  10. Collisions in young triple systems

    CERN Document Server

    Rawiraswattana, Krisada; Goodwin, Simon P

    2011-01-01

    We perform N-body simulations of young triple systems consisting of two low-mass companions orbiting around a significantly more massive primary. We find that, when the orbits of the companions are coplanar and not too widely separated, the chance of a collision between the two companions can be as high as 20 per cent. Collisions between one of the companions (always the less massive) and the primary can also occur in systems with unequal-mass companions. The chance of collisions is a few per cent in systems with more realistic initial conditions, such as with slightly non-coplanar orbits and unequal-mass companions. If the companions start widely separated then collision are very rare except in some cases when the total mass of the companions is large. We suggest that collisions between members of young multiple systems may explain some unusual young multiple systems such as apparently non-coeval companions.

  11. The triple system Zeta Aquarii

    CERN Document Server

    Tokovinin, Andrei

    2016-01-01

    Zeta Aquarii is a bright and nearby (28 pc) triple star with a 26-year astrometric subsystem. Almost a half of the outer 540-year visual orbit has been covered in 238 years of its observations. Both inner and outer orbits are revised here taking into account recent direct resolution of the inner pair Aa,Ab. The inner orbit has a high eccentricity of 0.87 and is inclined to the outer orbit by 140+-10 degrees, suggesting that Kozai-Lidov cycles take place. The masses of the stars Aa, B, and Ab are 1.4, 1.4, and 0.6 solar. The age of the system is about 3 Gyr, and the two main components have just left the main sequence. Hypothetically, this system could have formed by a dynamical capture of the small star Ab in the twin binary Aa,B.

  12. Simulation of triple coincidences in PET.

    Science.gov (United States)

    Cal-González, J; Lage, E; Herranz, E; Vicente, E; Udias, J M; Moore, S C; Park, M-A; Dave, S R; Parot, V; Herraiz, J L

    2015-01-07

    Although current PET scanners are designed and optimized to detect double coincidence events, there is a significant amount of triple coincidences in any PET acquisition. Triple coincidences may arise from causes such as: inter-detector scatter (IDS), random triple interactions (RT), or the detection of prompt gamma rays in coincidence with annihilation photons when non-pure positron-emitting radionuclides are used (β(+)γ events). Depending on the data acquisition settings of the PET scanner, these triple events are discarded or processed as a set of double coincidences if the energy of the three detected events is within the scanner's energy window. This latter option introduces noise in the data, as at most, only one of the possible lines-of-response defined by triple interactions corresponds to the line along which the decay occurred. Several novel works have pointed out the possibility of using triple events to increase the sensitivity of PET scanners or to expand PET imaging capabilities by allowing differentiation between radiotracers labeled with non-pure and pure positron-emitting radionuclides. In this work, we extended the Monte Carlo simulator PeneloPET to assess the proportion of triple coincidences in PET acquisitions and to evaluate their possible applications. We validated the results of the simulator against experimental data acquired with a modified version of a commercial preclinical PET/CT scanner, which was enabled to acquire and process triple-coincidence events. We used as figures of merit the energy spectra for double and triple coincidences and the triples-to-doubles ratio for different energy windows and radionuclides. After validation, the simulator was used to predict the relative quantity of triple-coincidence events in two clinical scanners assuming different acquisition settings. Good agreement between simulations and preclinical experiments was found, with differences below 10% for most of the observables considered. For clinical

  13. Molecular structure of the collagen triple helix.

    Science.gov (United States)

    Brodsky, Barbara; Persikov, Anton V

    2005-01-01

    The molecular conformation of the collagen triple helix confers strict amino acid sequence constraints, requiring a (Gly-X-Y)(n) repeating pattern and a high content of imino acids. The increasing family of collagens and proteins with collagenous domains shows the collagen triple helix to be a basic motif adaptable to a range of proteins and functions. Its rodlike domain has the potential for various modes of self-association and the capacity to bind receptors, other proteins, GAGs, and nucleic acids. High-resolution crystal structures obtained for collagen model peptides confirm the supercoiled triple helix conformation, and provide new information on hydrogen bonding patterns, hydration, sidechain interactions, and ligand binding. For several peptides, the helix twist was found to be sequence dependent, and such variation in helix twist may serve as recognition features or to orient the triple helix for binding. Mutations in the collagen triple-helix domain lead to a variety of human disorders. The most common mutations are single-base substitutions that lead to the replacement of one Gly residue, breaking the Gly-X-Y repeating pattern. A single Gly substitution destabilizes the triple helix through a local disruption in hydrogen bonding and produces a discontinuity in the register of the helix. Molecular information about the collagen triple helix and the effect of mutations will lead to a better understanding of function and pathology.

  14. C, P, and CP asymmetry observables based on triple product asymmetries

    CERN Document Server

    Bevan, Adrian J

    2015-01-01

    The discrete symmetries C, P and CP are known to be violated by the weak interaction. It is possible to probe the breaking of these symmetries using asymmetries constructed from triple products based on the decay of some particle M to a four body final state. These proceedings discuss the full set of possible asymmetries that can be probed and applications to various measurement scenarios, focusing mostly on charm mesons and baryons. The ramifications of what can be learned from such measurements are also discussed.

  15. Protein-free parallel triple-stranded DNA complex formation

    Science.gov (United States)

    Shchyolkina, A. K.; Timofeev, E. N.; Lysov, Yu. P.; Florentiev, V. L.; Jovin, T. M.; Arndt-Jovin, D. J.

    2001-01-01

    A 14 nt DNA sequence 5′-AGAATGTGGCAAAG-3′ from the zinc finger repeat of the human KRAB zinc finger protein gene ZNF91 bearing the intercalator 2-methoxy,6-chloro,9-amino acridine (Acr) attached to the sugar–phosphate backbone in various positions has been shown to form a specific triple helix (triplex) with a 16 bp hairpin (intramolecular) or a two-stranded (intermolecular) duplex having the identical sequence in the same (parallel) orientation. Intramolecular targets with the identical sequence in the antiparallel orientation and a non-specific target sequence were tested as controls. Apparent binding constants for formation of the triplex were determined by quantitating electrophoretic band shifts. Binding of the single-stranded oligonucleotide probe sequence to the target led to an increase in the fluorescence anisotropy of acridine. The parallel orientation of the two identical sequence segments was confirmed by measurement of fluorescence resonance energy transfer between the acridine on the 5′-end of the probe strand as donor and BODIPY-Texas Red on the 3′-amino group of either strand of the target duplex as acceptor. There was full protection from OsO4-bipyridine modification of thymines in the probe strand of the triplex, in accordance with the presumed triplex formation, which excluded displacement of the homologous duplex strand by the probe–intercalator conjugate. The implications of these results for the existence of protein-independent parallel triplexes are discussed. PMID:11160932

  16. Methyltransferase-like protein 16 binds the 3'-terminal triple helix of MALAT1 long noncoding RNA.

    Science.gov (United States)

    Brown, Jessica A; Kinzig, Charles G; DeGregorio, Suzanne J; Steitz, Joan A

    2016-12-06

    Metastasis-associated lung adenocarcinoma transcript 1 (MALAT1), a cancer-promoting long noncoding RNA, accumulates in cells by using a 3'-triple-helical RNA stability element for nuclear expression (ENE). The ENE, a stem-loop structure containing a U-rich internal loop, interacts with a downstream A-rich tract (ENE+A) to form a blunt-ended triple helix composed of nine U•A-U triples interrupted by a C•G-C triple and C-G doublet. This unique structure prompted us to explore the possibility of protein binding. Native gel-shift assays revealed a shift in radiolabeled MALAT1 ENE+A RNA upon addition of HEK293T cell lysate. Competitive gel-shift assays suggested that protein binding depends not only on the triple-helical structure but also its nucleotide composition. Selection from the lysate using a biotinylated-RNA probe followed by mass spectrometry identified methyltransferase-like protein 16 (METTL16), a putative RNA methyltransferase, as an interacting protein of the MALAT1 ENE+A. Gel-shift assays confirmed the METTL16-MALAT1 ENE+A interaction in vitro: Binding was observed with recombinant METTL16, but diminished in lysate depleted of METTL16, and a supershift was detected after adding anti-METTL16 antibody. Importantly, RNA immunoprecipitation after in vivo UV cross-linking and an in situ proximity ligation assay for RNA-protein interactions confirmed an association between METTL16 and MALAT1 in cells. METTL16 is an abundant (∼5 × 10(5) molecules per cell) nuclear protein in HeLa cells. Its identification as a triple-stranded RNA binding protein supports the formation of RNA triple helices inside cells and suggests the existence of a class of triple-stranded RNA binding proteins, which may enable the discovery of additional cellular RNA triple helices.

  17. Reverse triple I method of fuzzy reasoning

    Institute of Scientific and Technical Information of China (English)

    宋士吉; 吴澄

    2002-01-01

    A theory of reverse triple I method with sustention degree is presented by using the implication operator R0 in every step of the fuzzy reasoning. Its computation formulas of supremum for fuzzy modus ponens and infimum for fuzzy modus tollens are given respectively. Moreover, through the generalization of this problem, the corresponding formulas of ?-reverse triple I method with sustention degree are also obtained. In addition, the theory of reverse triple I method with restriction degree is proposed as well by using the operator R0, and the computation formulas of infimum for fuzzy modus ponens and supremum for fuzzy modus tollens are shown.

  18. A Fusion Chamber Design with a Liquid First Wall and Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Nygren, R; Sze, D; Nelson, B; Fogarty, P; Eberle, C; Rognlien, T; Rensink, M; Smolentsev, S; Youssef, M; Sawan, M; Merrill, B; Majeski, R

    2003-11-11

    The APEX study is investigating the use of free flowing liquid surfaces to form the inner surface of the chamber around a fusion plasma. We present a design for the chamber of a 3840MW fusion reactor based on the configuration for the chamber and magnets from ARIESRS but with a fast flowing molten salt of mixed Be, Li and Na fluorides for the first wall and divertor and molten salt blanket with a ferritic steel structure. Our design analysis includes strong radiation from the core and edge plasma, (liquid) MHD effects on the weakly conducting molten salt, a recycling first wall stream that enables a high efficiency thermal conversion, and evaluations of breeding, neutronics, tritium recovery and safety.

  19. 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.

  20. 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.

  1. 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.

  2. 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...

  3. Numerical simulation of CFC and tungsten target erosion in ITER-FEAT divertor

    Energy Technology Data Exchange (ETDEWEB)

    Filatov, V. E-mail: filatovv@niiefa.spb.su

    2003-03-01

    Physical, chemical and thermal surface erosion for water-cooled target armoured by CFC and tungsten is simulated by numerical code ERosion OF Immolated Layer (EROFIL-1). Some calculation results on the CFC and tungsten vertical target (VT) erosion in the ITER-FEAT divertor are presented for various operation modes (normal operations, slow transients, ELMs and disruptions). The main erosion mechanisms of CFC armour are the chemical and sublimation ones. Maximum erosion depth per 3000 cycles during normal operations and slow transients is of 2.7 mm at H phase and of 13.5 mm at DT phase. An evaluation of VT tungsten armour erosion per 3000 cycles of H and DT operations shows that no physical or chemical erosion as well as no melting are expected for tungsten armour at normal operations and slow transients. The tungsten armour melting at 2x10{sup 6} ELMs is not allowable. The 300 disruptions are not dangerous in view of evaporation.

  4. Thermo-mechanical tests of a CFC divertor mock-up

    Science.gov (United States)

    Cardella, A.; Akiba, M.; Duwe, R.; Di Pietro, E.; Suzuki, S.; Satoh, K.; Reheis, N.

    1994-04-01

    Thermo-mechanical tests have been performed on a divertor mock-up consisting of a metallic tube armoured with five carbon fibre composite tiles. The tube is inserted inside the tiles and brazed with TiCuSil braze (monoblock concept). The tube material is TZM, a molybdenum alloy, and the armour material is SEP CARB N112, a high conductivity carbon-carbon composite. Using special surface preparation consisting of laser drilling, small (˜- 500 μm) holes in the composite have been made to increase the surface wetted by the braze and the resistance. The mock-up has been tested at the JAERI 400 kW electron beam test facility JEBIS. The aim of the test was to assess the performance of the mock-up in screening and thermal fatigue tests with particular attention to the behaviour of the armour to heat sink joint.

  5. Surface modifications of W divertor components for EAST during exposure to high heat loads with He

    Energy Technology Data Exchange (ETDEWEB)

    Li, C., E-mail: lichun10@mails.tsinghua.edu.cn [Laboratory of Advanced Materials, School of Materials Science and Engineering, Tsinghua University, Beijing 100084 (China); Greuner, H. [Max Planck Institute for Plasma Physics, Boltzmannstr. 2, 85748 Garching (Germany); Yuan, Y. [School of Physics and Nuclear Energy Engineering, Beihang University, Beijing 100191 (China); Zhao, S.X.; Luo, G.N. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Böswirth, B. [Max Planck Institute for Plasma Physics, Boltzmannstr. 2, 85748 Garching (Germany); Fu, B.Q.; Jia, Y.Z. [Laboratory of Advanced Materials, School of Materials Science and Engineering, Tsinghua University, Beijing 100084 (China); Liu, X. [Southwestern Institute of Physics, Chengdu, Sichuan 610041 (China); Liu, W., E-mail: liuw@mail.tsinghua.edu.cn [Laboratory of Advanced Materials, School of Materials Science and Engineering, Tsinghua University, Beijing 100084 (China)

    2015-08-15

    Flat-type W/Cu plasma-facing components have been developed for the new generation divertor of the Chinese Experimental Advanced Superconducting Tokamak. Surface modifications of such actively water-cooled W components following short and long pulse high heat loading coupled with He particle loads with fluence of 3 × 10{sup 22} m{sup −2} have been investigated. An adiabatically loaded W block was investigated as a comparison and exposed to short pulse loads. Blistering was observed on all sample surfaces, but was less pronounced on the components than on the W block, due to the significant lower surface temperature caused by active cooling. For components, longer pulse loads gave rise to a rougher surface. Furthermore, most blisters on components are found to be less than 1 μm in diameter, with just a very few blisters larger than 1 μm, observed only in some near 〈1 1 1〉 grains.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. Assessment of erosion of the ITER divertor targets during type I ELMs

    Energy Technology Data Exchange (ETDEWEB)

    Federici, G [ITER JWS Garching Co-center, Boltzmannstrasse 2, 85748 Garching (Germany); Loarte, A [EFDA Close Support Unit, Boltzmannstrasse 2, 85748 Garching (Germany); Strohmayer, G [ITER JWS Garching Co-center, Boltzmannstrasse 2, 85748 Garching (Germany)

    2003-09-01

    This paper presents the results of a preliminary assessment conducted to estimate the thermal response and erosion lifetime of the ITER divertor targets clad either with carbon-fibre composite or tungsten during type I ELMs. The one-dimensional thermal/erosion model, used for the analyses, is briefly described. It includes all the key surface heat transfer processes such as evaporation, melting, and radiation, and their interaction with the bulk thermal response, and it is based on an implicit finite-difference scheme, which allows for temperature-dependent material properties. The cases analysed clarify the influence of several ELM parameters on the heat transfer and erosion processes at the target (i.e. characteristic plasma ELM energy loss from the pedestal, fraction of the energy reaching the divertor, broadening of the strike-points during ELMs, duration and waveform of the ELM heat load) and design/material parameters (i.e. inclination of the target, type and thickness of the armour material, and for tungsten only, fraction of the melt layer loss). Comparison is made between cases where all ELMs are characterized by the same fixed averaged parameters, and cases where instead the characteristic parameters of each ELM are evaluated in a random fashion by using a standard Monte Carlo technique, based on distributions of some of the variables of interest derived from experiments in today's machines. Although uncertainties rule out providing firm quantitative predictions, the results of this study are useful to illustrate trends. Based on the results, the implications on the design and operation are discussed and priorities are determined for the R and D needed to reduce the remaining uncertainties.

  11. Manufacturing and testing of a Be/OFHCCu divertor module

    Science.gov (United States)

    Araki, M.; Youchison, D. L.; Akiba, M.; Watson, R. D.; Sato, K.; Suzuki, S.

    1996-10-01

    Beryllium, carbon-based materials and tungsten are considered as plasma facing materials for the next generation of fusion machines such as the international thermonuclear experimental reactor (ITER). Beryllium is one of the primary candidate materials because of its low atomic number and lack of tritium codeposition. However, joining of a beryllium armor to a copper heat sink remains a critical problem due to the formation of brittle intermetallics at the interface. To address this concern, the Japan Atomic Energy Research Institute manufactured a beryllium/Cu divertor module with Cr and Ni diffusion barriers. This Be/Cu module was tested in the electron beam test system of Sandia National Laboratories in the framework of the US—Japan Fusion Collaboration. The divertor module consisted of four beryllium tiles, 25 mm × 25 mm, and a square copper heat sink with convolutions like a screw nut inside the coolant channel. To evaluate the integrity of the brazed bonds under various heat fluxes, beryllium tiles of two different thicknesses, 2 and 10 mm, were bonded to the copper heat sink. Cooling conditions of 10 m/s water flow velocity at 1 MPa, and a water inlet temperature of 20°C were selected based on the thermal analysis. During high heat flux testing the 10 mm thick Be tiles detached at an absorbed heat flux around 5 MW/m 2 for several shots due to flaws at the braze joint confirmed by optical observation after manufacturing. One of the 2 mm thick Be tiles failed after 550 cycles at the steady state heat flux of 6.5 MW/m 2. Most likely the failure was caused by brittleness at the interface caused by the presence of BeCu intermetallics.

  12. A Fusion Reactor Design with a Liquid First Wall and Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Nygren, R E; Rognlien, T D; Rensink, M E; Smolentsev, S S; Youssef, M E; Sawan, M Z; Merrill, B J; Eberle, C; Fogarty, P J; Nelson, B E; Sze, D K; Majeski, R

    2003-11-13

    Within the magnetic fusion energy program in the US, a program called APEX is investigating the use of free flowing liquid surfaces to form the inner surface of the chamber around the plasma. As part of this work, the APEX Team has investigated several possible design implementations and developed a specific engineering concept for a fusion reactor with liquid walls. Our approach has been to utilize an already established design for a future fusion reactor, the ARIES-RS, for the basic chamber geometry and magnetic configuration and to replace the chamber technology in this design with liquid wall technology for a first wall and divertor and a blanket with adequate tritium breeding. This paper gives an overview of one design with a molten salt (a mixture of lithium, beryllium and sodium fluorides) forming the liquid surfaces and a ferritic steel for the structural material of the blanket. The design point is a reactor with 3840MW of fusion power of which 767MW is in the form of energetic particles (alpha power) and 3073MW is in the form of neutrons. The alpha plus auxiliary power total 909MW of which 430MW is radiated from the core mostly onto the first wall and the balance flows into the edge plasma and is distributed between the first wall and the divertor. In pursuing the application of liquid surfaces in APEX, the team has developed analytical tools that are significant achievements themselves and also pursued experiments on flowing liquids. This work is covered elsewhere, but the paper will also note several such areas to indicate the supporting science behind the design presented. Significant new work in modeling the plasma edge to understand the interaction of the plasma with the liquid walls is one example. Another is the incorporation of magneto-hydrodynamic (MHD) effects in fluid modeling and heat transfer.

  13. 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.

  14. Application of carbon-aluminum nanostructures in divertor coatings from fusion reactor

    Science.gov (United States)

    Ciupina, V.; Lungu, C. P.; Vladoiu, R.; Epure, T. D.; Prodan, G.; Porosnicu, C.; Prodan, M.; Stanescu, I. M.; Contulov, M.; Mandes, A.; Dinca, V.; Zarovschi, V.

    2012-10-01

    Nanostructured carbon materials have increasingly attracted the interest of the scientific community, because of their fascinating physical properties and potential applications in high-tech devices. In the current ITER design, the tiles made of carbon fiber composites (CFCs) are foreseen for the strike point zone and tungsten (W) for other parts of the divertor region. This choice is a compromise based mainly on experience with individual materials in many different tokamaks. Also Carbon-Aluminum composites are the candidate material for the First Wall in ITER. In order to prepare nanostructured carbon-aluminum nanocomposite for the divertor part in fusion applications, the original method thermionic vacuum arc (TVA) was used in two electronic guns configuration. One of the main advantages of this technology is the bombardment of the growing thin film just by the ions of the depositing film. Moreover, the energy of ions can be controlled. Thermo-electrons emitted by an externally heated cathode and focused by a Wehnelt focusing cylinder are strongly accelerated towards the anode whose material is evaporated and bright plasma is ignited by a high voltage DC supply. The nanostructured C-Al films were characterized by Scanning Electron Microscopy (SEM), Transmission Electron Microscopy (TEM). Tribological properties in dry sliding were evaluated using a CSM ball-on-disc tribometer. The carbon - aluminum films were identified as a nanocrystals complex (from 2nm to 50 nm diameters) surrounded by amorphous structures with a strong graphitization tendency, allowing the creating of adherent and wear resistant films. The friction coefficients (0.1 - 0.2, 0.5) of the C-Al coatings was decreased more than 2-5 times in comparison with the uncoated substrates proving excellent tribological properties. C-Al nanocomposites coatings were designed to have excellent tribological properties while the structure is composed by nanocrystals complex surrounded by amorphous structures

  15. Electron microscopy characterization of some carbon based nanostructures with application in divertors coatings from fusion reactor

    Science.gov (United States)

    Ciupina, V.; Morjan, I.; Lungu, C. P.; Vladoiu, R.; Prodan, G.; Prodan, M.; Zarovschi, V.; Porosnicu, C.; Stanescu, I. M.; Contulov, M.; Mandes, A.; Dinca, V.; Sugiyama, K.

    2011-10-01

    Nanostructured carbon materials have increasingly attracted the interest of the scientific community, because of their fascinating physical properties and potential applications in high-tech devices. In the current ITER design, the tiles made of carbon fiber composites (CFCs) are foreseen for the strike point zone and tungsten (W) for other parts of the divertor region. This choice is a compromise based mainly on experience with individual materials in many different tokamaks. Also Beryllium is the candidate material for the First Wall in ITER. In order to prepare nanostructured carbon-tungsten nanocomposite for the divertor part in fusion applications, the original method thermionic vacuum arc (TVA) was used in two electronic guns configuration. One of the main advantages of this technology is the bombardment of the growing thin film just by the ions of the depositing film. The nanostructured C-W and C-Be films were characterized by Scanning Electron Microscopy (SEM), Transmission Electron Microscopy (TEM) and Atomic Force Microscopy (AFM). The C-W films were identified as a nanocrystals complex (5 nm average diameter) surrounded by amorphous structures with a strong graphitization tendency, allowing the creating of adherent and wear resistant films. The C-Be films are polycrystalline with mean grain size about 15 nm. The friction coefficients (0.15 - 0.35) of the C-W coatings was decreased more than 3-5 times in comparison with the uncoated substrates proving excellent tribological properties. C-W nanocomposites coatings were designed to have excellent tribological properties while the structure is composed by nanocrystals complex surrounded by amorphous structures with a strong graphitization tendency, allowing the creating of adherent and wear resistant films.&updat

  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.

    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 cm(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. R and D and maintenance on graphite tile of divertor region at EAST

    Energy Technology Data Exchange (ETDEWEB)

    Ji, X., E-mail: jixiang@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, P.O. BOX 1126, Hefei, Anhui 230031 (China); Song, Y.T.; Wu, S.T.; Hao, J.; Du, S.; Peng, Y.; Cao, L.; Wang, S. [Institute of Plasma Physics, Chinese Academy of Sciences, P.O. BOX 1126, Hefei, Anhui 230031 (China)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer Find out the reason of damage of graphite tile. Black-Right-Pointing-Pointer Simulation the halo current. Black-Right-Pointing-Pointer Stress analysis of graphite tile by ANSYS. Black-Right-Pointing-Pointer Do the experiments to test the strength of graphite tile. Black-Right-Pointing-Pointer Do the optimization and maintenance of graphite tile. - Abstract: EAST, with full superconducting magnetic coils, had been designed and constructed to address the scientific and engineering issues under steady state operation. The in-vessel components are full graphite tiles as first wall had been operated successfully. In the experiment campaign of 2010, the H mode operation and 1 MA operation have been gotten on EAST. However, in some case, some of the graphite tiles of divertor region are damaged with the plasma parameter enhanced. As most of the damaged graphite tiles are in the divertor region, they are probably damaged by the electro-magnetic force of the halo current when the VDEs occur. The force of the halo current is re-estimated. The structure analysis has been done by the ANSYS software. From the analysis result. It can be obtained that the stress is larger than the allowable stress when the halo current on the graphite tile is larger than 2.7 kA. The tensile testing of the graphite also has been done. As the result, the graphite tiles are damaged when the forces are up to 2400 N. To deal with the problem, two proposes are accepted. In the one hand, the new type graphite material is used, whose tensile strength is up to 45 MPa. In the other hand, the structure of the graphite tiles is optimized.

  19. Using SOLPS to confirm the importance of total flux expansion in Super-X divertors

    Science.gov (United States)

    Moulton, D.; Harrison, J.; Lipschultz, B.; Coster, D.

    2017-06-01

    We show that a central characteristic of Super-X divertors, total flux expansion f R (defined as the ratio of the elementary area normal to the magnetic field at the target to that at the X-point), significantly changes the characteristics of the target plasma for fixed upstream conditions. To isolate the effect of total flux expansion from other effects, we utilise SOLPS-5.0 simulations of an isolated slot divertor leg in a minimally complex, rectangular geometry. The grid is rotated outwards about a fixed X-point in order to perform a scan in which only the total flux expansion increases, by means of a decrease in the target magnetic field at higher major radius. We find that if the SOL remains in the attached, conduction-limited regime throughout the scan, the target electron density scales approximately as {f}{{R}}2, while the target electron temperature scales approximately as 1/{f}{{R}}2, in good agreement with the modified two-point model presented in Petrie et al (2013 Nucl. Fusion 53 113024). If, however, the SOL transitions from the sheath-limited regime to the conduction-limited regime during the scan, the simulated scalings of target electron temperature and density are weaker than predicted by the modified two-point model. The upstream density for transition from sheath- to conduction-limited regimes is found to scale approximately with 1/{f}{{R}}, in agreement with the modified two-point model. Assessing upstream-density-driven detachment onset, we find that the target electron temperature at which target density rollover occurs (∼0.6 eV) is independent of f R. Given this, the modified two-point model predicts a halving of the upstream (and target) densities at which rollover occurs when f R is doubled, in good agreement with the simulation results.

  20. Overview of co-deposition and fuel inventory in castellated divertor structures at JET

    Science.gov (United States)

    Rubel, M. J.; Coad, J. P.; Pitts, R. A.; JET-EFDA Work Programme

    2007-08-01

    The main focus of this work is fuel retention in plasma components of the JET water-cooled Mk-I divertors operated with small tiles, first with carbon fibre composite (CFC) and then with castellated beryllium. Until recently these have been the only large-scale structures of this type used in fusion experiments. Three issues regarding fuel retention and material migration are addressed: (i) accumulation in gaps separating tiles and in the grooves of castellation; (ii) comparison of deposition on carbon and beryllium; (iii) in-depth migration of deuterium into the bulk of CFC. The essential results are summarised as follows: (i) co-deposition occurs up to a few cm deep in the gaps between the Mk-I tiles; (ii) fuel inventory in the CFC tile gaps exceeds that on plasma-facing surfaces by up to a factor of 2; (iii) in gaps between the beryllium tiles from the inner divertor corner the fuel content reaches 30% of that on plasma-facing surfaces, whereas in the grooves of castellation in Be the fuel content is less than 3.0% of that found on the top surface; (iv) fuel inventory on the Be tiles is strongly associated with the carbon co-deposition; (v) the D content measured in the bulk (1.5 mm below the surface) on cleaved CFC tiles exceeds 1 × 10 15 cm -2. Implications of these results for a next-step device are addressed and the transport mechanism into the gaps is briefly discussed. The results presented here suggest that in a machine with non-carbon walls in the main chamber (as foreseen for ITER) the material transport and subsequent fuel inventory in the castellation would be reduced.

  1. Influence of plasma resistance and fluctuation on probe characteristics in detached recombining plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Ohno, N.; Tanaka, N.; Nishijima, D.; Takamura, S. [Nagoya Univ. (Japan). Dept. of Energy Engineering and Science; Ezumi, N. [Dept. of Electronics and Control Engineering, Nagano National Coll. of Technology, Tokuma (Japan)

    2001-07-01

    In order to find the causes of the strong anomaly of current-voltage characteristics of Langmuir probe observed in detached recombining plasmas in a linear divertor plasma simulator, NAGDIS-II, we have investigated plasma resistance along a magnetic field and potential fluctuations in the detached recombining plasmas. Simple calculation on the ratio between the plasma length, at which plasma resistance and resistance of ion sheath formed around a probe tip become equal, and an electron collection length indicates that the evaluation of electron temperature T{sub e} becomes inaccurate at T{sub e} of less than 0.6 eV when plasma density and neutral pressure are 1.0 x 10{sup 18} m{sup -3} and 10 mtorr, respectively. The potential fluctuation in detached recombining plasmas was found to be so large compared to T{sub e}/e, which can also modify the probe characteristics. (orig.)

  2. The triple helix perspective of innovation systems

    NARCIS (Netherlands)

    Leydesdorff, L.; Zawdie, G.

    2010-01-01

    Alongside the neo-institutional model of networked relations among universities, industries, and governments, the triple helix can be provided with a neo-evolutionary interpretation as three selection environments operating upon one another: markets, organisations and technological opportunities. Ho

  3. On triple factorisations of finite groups

    CERN Document Server

    Alavi, S Hassan

    2009-01-01

    This paper introduces and develops a general framework for studying triple factorisations of the form $G=ABA$ of finite groups $G$, with $A$ and $B$ subgroups of $G$. We call such a factorisation nondegenerate if $G\

  4. Triple click reaction strategy for macromolecular diversity.

    Science.gov (United States)

    Tunca, Umit

    2013-01-11

    This Feature Article focuses on the rapidly emerging concept of the "triple click reactions" towards the design and synthesis of macromolecules with well-defined topology and chemical composition, and also precise molecular weight and narrow molecular weight distribution. The term "triple click reaction" used in this feature article is based on the utilization of three chemically and mechanistically different click reactions for polymer-polymer conjugation and post-modification of the polymers. Three sequential click reactions of which two are identical should not be considered to be triple click reactions. The triple click reaction strategy for polymer conjugation and post-modification of polymers is classified in this article based on the resultant architectures: linear and non-linear structures. Copyright © 2013 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  5. Study of Divertor Heat Patterns Induced by LHCD L-Mode Plasmas Using an Infra-Red Camera System on EAST

    Science.gov (United States)

    Zhang, Bin; Gan, Kaifu; Gong, Xianzu; Zhang, Xiaodong; Wang, Fumin; Yang, Zhendong; Chen, Meiwen; Wang, Xiaoqiong

    2015-10-01

    Divertor heat patterns induced by Lower Hybrid Current Drive (LHCD) L-mode plasmas are investigated using an infra-red (IR) camera system on an Experimental Advanced Superconducting Tokamak (EAST). A two-dimensional finite element analysis code DFlux is used to compute heat flux along the poloidal divertor target and corresponding quantities. Outside the Origin Strike Zone (OSZ), a Second Peak Heat Flux (SPHF) zone, where the heat flux is even stronger than that at the OSZ, appears on the lower-outer (LO) divertor plates with LHCD and disappears immediately after switching off the LHCD. The main heat-flux shifts from the SPHF zone towards the OSZ when the divertor configuration converts from double null to lower single null, indicating that the growth of the SPHF zone is apparently affected by a plasma magnetic configuration. The heat patterns on the LO divertor plates are observed to be different from that on the lower-inner (LI) targets as the SPHF zone appears only on the LO divertor target. It is also found that the heat flux at the SPHF zone was obviously enhanced after the Supersonic Molecule Beam Injection (SMBI) pulse. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2014GB101001 and 2014GB101002)

  6. 3D plasma turbulence and neutral simulations using the Hermes model in BOUT + + : a study of linear devices and the tokamak edge and divertor region

    Science.gov (United States)

    Leddy, Jarrod; Dudson, Ben

    2016-10-01

    Understanding the transport processes in the low temperature plasma at the boundary region of magnetic confinement fusion (MCF) devices is crucial to the design and operation of future fusion reactor devices. It influences the divertor heat load, and probably the core confinement as well. The dominant source of this transport is turbulence, which serves to mix the high and low temperature regions of the plasma. The nature of this plasma turbulence is affected by not only the plasma parameters, but also the neutral species that also exist in these low temperature regions. The interaction of neutrals with the plasma turbulence is studied in linear device geometry (for its simplicity, yet similarity in plasma parameters), and the result is a strong interaction that impacts the local plasma and neutral densities, momenta and energies. The neutral gas is found to affect plasma edge turbulence primarily through momentum exchange, reducing the radial electric field and enhancing cross-field transport, with consequent implications for the SOL width and divertor heat loads. Therefore, turbulent plasma and fluid simulations have been performed in multiple tokamak geometries to more closely examine the effects of this interaction. These cases were chosen for the variety in configuration with ISTOK having a toroidal limiter (ie. no divertor), DIII-D having a standard divertor configuration, and MAST-U having a super-X divertor with extended outer divertor legs. Progress towards the characterization of neutral impact on detachment and edge behavior will be presented.

  7. Double Planet Meets Triple Star

    Science.gov (United States)

    2002-08-01

    atmosphere, a large campaign involving more than twenty scientists and engineers from the Paris Observatory and associated institutions [1] was organized to observe the July 20, 2002, event involving an occultation of a star of visual magnitude 11 (i.e., about 100 times fainter than what can be perceived with then unaided eye), referred to as "P126" in McDonald and Elliot's catalogue. In May 2002, preparatory observations showed that star to be double, with the brighter component of the system ( "P126 A" ) being likely to be occulted by Pluto, as seen from South America. However, because of the duplicity, the predictions of exactly where the shadow of Pluto would sweep the ground were uncertain by about 0.1 arcsec in the sky, corresponding to more than 2000 km on the ground. The NACO images ESO PR Photo 21b/02 ESO PR Photo 21b/02 [Preview - JPEG: 400 x 469 pix - 47k] [Normal - JPEG: 800 x 937 pix - 208k] ESO PR Photo 21c/02 ESO PR Photo 21c/02 [Preview - JPEG: 400 x 467 pix - 53k] [Normal - JPEG: 800 x 933 pix - 232k] Caption : PR Photo 21b/02 shows one of the images obtained with the NAOS-CONICA (NACO) adaptive optics (AO) camera mounted on the ESO VLT 8.2-m YEPUN telescope at the Paranal Observatory in connection with a stellar occultation by Pluto on July 20, 2002. The star was found to be triple - the three components (A, B and C), as well as Pluto and its moon, Charon, are indicated in PR Photo 21c/02 for easy orientation. The images are based on data available from the NACO data webpage. See the text for details. In the end, the close approach (an "appulse" in astronomical terminology) of Pluto and P126 A was indeed observed from various sites in South America, with several mobile telescopes and also including major facilities at the ESO La Silla and Paranal Observatories. In particular, unique and very sharp images were obtained with the NAOS-CONICA (NACO) adaptive optics (AO) camera mounted on the ESO VLT 8.2-m YEPUN telescope . One of the NACO images is shown in PR

  8. Close Binaries, Triples, and Eclipses

    Science.gov (United States)

    Sanborn, Jason; Zavala, R. T.

    2013-01-01

    Observations of the variable radio source b Per (HR1324) are part of an ongoing survey of close binary systems using the Navy Precision Optical Interferometer. Historical observations of b Per include sparse photometric and spectroscopic observations dating back to 1923, clearly showing this object to be a non-eclipsing, single-lined ellipsoidal variable. This is where the story for b Per stopped until recent inclusion of optical interferometric data which led to the detection of a third, long-period component. As the interferometric observations continue to build up so to is the understanding of this binary system, with the modeled orbital parameters pointing to an edge-on orientation that may allow for the detection of an eclipse by the long-period component. These types of eclipse events are quite rare for long-period binaries due to the nearly edge-on orientation required for their detection, leaving open the opportunity for more traditional methods of observation to add to the body of knowledge concerning this understudied system. Here we present the latest observational data of the b Per system along with an introduction to the best fit orbital parameters governing the eclipsing nature of this complex triple-system.

  9. The collagen triple-helix structure.

    Science.gov (United States)

    Brodsky, B; Ramshaw, J A

    1997-03-01

    Recent advances, principally through the study of peptide models, have led to an enhanced understanding of the structure and function of the collagen triple helix. In particular, the first crystal structure has clearly shown the highly ordered hydration network critical for stabilizing both the molecular conformation and the interactions between triple helices. The sequence dependent nature of the conformational features is also under active investigation by NMR and other techniques. The triple-helix motif has now been identified in proteins other than collagens, and it has been established as being important in many specific biological interactions as well as being a structural element. The nature of recognition and the degree of specificity for interactions involving triple helices may differ from globular proteins. Triple-helix binding domains consist of linear sequences along the helix, making them amenable to characterization by simple model peptides. The application of structural techniques to such model peptides can serve to clarify the interactions involved in triple-helix recognition and binding and can help explain the varying impact of different structural alterations found in mutant collagens in diseased states.

  10. 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)

  11. 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.

  12. Deuterium retention in tungsten exposed to mixed D + N plasma at divertor relevant fluxes in Magnum-PSI

    Science.gov (United States)

    Lee, H. T.; De Temmerman, G.; Gao, L.; Schwarz-Selinger, T.; Meisl, G.; Höschen, T.; Ueda, Y.

    2015-08-01

    Nitrogen (N) has been proposed as an extrinsic impurity species in the divertor to reduce the local power load onto tungsten (W) plasma-facing components. Laboratory studies at low incident fluxes have indicated N increases deuterium (D) retention in tungsten. Here we show that W exposed to D + N Magnum-PSI plasmas under divertor relevant particle fluxes (∼1024 D/m2 s), also results in an increase in D retention by enhanced near surface trapping up to 1100 K due either to N or Mo impurities, and increased retention in the bulk at T > 700 K. These results demonstrate that N or Mo surface impurities have the potential to alter the tritium inventory in tungsten plasma-facing components under diverter relevant particle fluxes by affecting surface and bulk retention processes.

  13. Deuterium retention in tungsten exposed to mixed D + N plasma at divertor relevant fluxes in Magnum-PSI

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H.T., E-mail: heunlee@wakate.frc.eng.osaka-u.ac.jp [Graduate School of Engineering, Osaka University, Suita, Osaka 565-0871 (Japan); De Temmerman, G. [Dutch Institute for Fundamental Energy Research, Nieuwegein, 3439 MN (Netherlands); Gao, L.; Schwarz-Selinger, T.; Meisl, G.; Höschen, T. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Ueda, Y. [Graduate School of Engineering, Osaka University, Suita, Osaka 565-0871 (Japan)

    2015-08-15

    Nitrogen (N) has been proposed as an extrinsic impurity species in the divertor to reduce the local power load onto tungsten (W) plasma-facing components. Laboratory studies at low incident fluxes have indicated N increases deuterium (D) retention in tungsten. Here we show that W exposed to D + N Magnum-PSI plasmas under divertor relevant particle fluxes (∼10{sup 24} D/m{sup 2} s), also results in an increase in D retention by enhanced near surface trapping up to 1100 K due either to N or Mo impurities, and increased retention in the bulk at T > 700 K. These results demonstrate that N or Mo surface impurities have the potential to alter the tritium inventory in tungsten plasma-facing components under diverter relevant particle fluxes by affecting surface and bulk retention processes.

  14. Drive of parallel flows by turbulence and large-scale E × B transverse transport in divertor geometry

    Science.gov (United States)

    Galassi, D.; Tamain, P.; Bufferand, H.; Ciraolo, G.; Ghendrih, Ph.; Baudoin, C.; Colin, C.; Fedorczak, N.; Nace, N.; Serre, E.

    2017-03-01

    The poloidal asymmetries of parallel flows in edge plasmas are investigated by the 3D fluid turbulence code TOKAM3X. A diverted COMPASS-like magnetic equilibrium is used for the simulations. The measurements and simulations of parallel Mach numbers are compared, and exhibit good qualitative agreement. Small-scale turbulent transport is observed to dominate near the low field side midplane, even though it co-exists with significant large-scale cross-field fluxes. Despite the turbulent nature of the plasma in the divertor region, simulations show the low effectiveness of turbulence for the cross-field transport towards the private flux region. Nevertheless, a complex pattern of fluxes associated with the average field components are found to cross the separatrix in the divertor region. Large-scale and small-scale turbulent E× B transport, along with the \

  15. Experimental and numerical evaluation of IR thermography method for Final Acceptance Tests of the ITER divertor dome

    Energy Technology Data Exchange (ETDEWEB)

    Tanchuk, Victor, E-mail: Victor.Tanchuk@sintez.niiefa.spb.su [Scientific Technical Center SINTEZ, D.V. Efremov Institute, 196641 St. Petersburg (Russian Federation); Grigoriev, Sergey; Makhankov, Alexey; Senik, Konstantin; Yablokov, Nikolay [Scientific Technical Center SINTEZ, D.V. Efremov Institute, 196641 St. Petersburg (Russian Federation); Belenky, Mikhail; Blinov, Mikhail; Lebedev, Mikhail; Fokin, Boris [I.I. Polzunov Scientific and Development Association on Research and Design of Power Equipment, 191167 St. Petersburg (Russian Federation)

    2014-10-15

    Highlights: • The experiments on the assembly of the ¼ ITER divertor dome consisting of three groups of hypervapotrons with aim to prove applicability of the thermography method for detection of defective channels are performed. • Numerical simulation of the FAT procedure on the calculation model of ½ dome is carried out. • It is not only the flow rate difference in parallel channels caused by defective hypervapotrons but also the flow history that affects essentially the dynamics of the temperature field of the dome surface. - Abstract: The divertor dome (DO), being part of the ITER divertor, is designed to extract the major part of the plasma thermal energy. As a plasma-facing component (PFC), the DO experiences high heat fluxes (up to 5.0 MW/m{sup 2}). Such severe operation conditions of the DO imply stringent requirements for the DO design and its cooling system to ensure the required temperature operation regime of the dome. Hence, Final Acceptance Tests (FAT) shall be performed on each DO final assembled component with the aim to demonstrate that none of parallel coolant channels are completely or partially blocked. The paper presents the results of the analytical and experimental testing of the thermography method capability to perform the FAT. The aim is to determine defective hypervapotrons of the divertor dome. The method consists in contactless measurement of the dynamic temperature field of the PFC surface at a step-like increase (from zero to constant value) in the coolant flow rate with a temperature higher than that of the hypervapotron.

  16. Dependence of the L-H transition on X-point geometry and divertor recycling on NSTX

    Science.gov (United States)

    Battaglia, D. J.; Chang, C. S.; Kaye, S. M.; Kim, K.; Ku, S.; Maingi, R.; Bell, R. E.; Diallo, A.; Gerhardt, S.; LeBlanc, B. P.; Menard, J.; Podesta, M.; the NSTX Team

    2013-11-01

    The edge electron (Te) and ion temperature (Ti) at the time of the L-H transition increase when the X-point radius (RX) is reduced to a high-triangularity shape while maintaining constant edge density. Consequently the L-H power threshold (PLH) is larger for the high-triangularity shape. This supports the prediction that a single-particle loss hole, whose properties are strongly linked to RX and Ti, influences the edge radial electric field (Er) and Er × B flow-shearing rate available for turbulence suppression. Simulations using XGC0, a full-f drift-kinetic neoclassical code, indicate that maintaining a constant Er × B flow-shearing rate does require a larger heat flux and edge Ti as RX decreases. NSTX also observes a decrease in PLH when the divertor recycling is decreased using lithium coatings. However, the edge Te and Ti at the L-H transition appear independent of the divertor recycling for a constant shape. XGC0 calculations demonstrate that more heat flux is needed to maintain the edge Ti and the Er × B flow-shearing rate as the contribution of divertor recycling to the overall neutral fuelling rate increases.

  17. Development of residual thermal stress-relieving structure of CFC monoblock target for JT-60SA divertor

    Energy Technology Data Exchange (ETDEWEB)

    Tsuru, Daigo, E-mail: tsuru.daigo@jaea.go.jp; Sakurai, Shinji; Nakamura, Shigetoshi; Ozaki, Hidetsugu; Seki, Yohji; Yokoyama, Kenji; Suzuki, Satoshi

    2015-10-15

    Highlights: • We carried out numerical simulations on residual thermal stress of targets for the JT-60SA divertor. • We developed three measures to reduce residual thermal stress. • We proposed two structures of CFC monoblock target for the JT-60SA divertor. • We confirmed the effectiveness of the structure by infrared thermography inspection and high heat flux test. - Abstract: Carbon fibre-reinforced carbon composite (CFC) monoblock target for JT-60SA divertor is under development towards the mass-production. CFC monoblocks, WCu interlayers and a CuCrZr cooling tube at the centre of the monoblocks were bonded by vacuum brazing in a high temperature, to a target. If residual thermal stress due to difference of thermal expansions between CFC and CuCrZr exceeds the maximum allowable stress of the CFC after the bonding, cracks are generated in the CFC monoblock and heat removal capacity of the target degrades. In this paper, new structures of the targets were proposed, to reduce residual thermal stress and to mitigate the degradation of heat removal capacity of the targets. Some measures, including slitting of the CFC monoblock aside of the cooling tube, replacement of the interlayer material and shifting the position of the cooling tube, were implemented. The effectiveness of the measures was evaluated by numerical simulations. Target mock-ups with the proposed structures were manufactured. Infrared thermography inspection and high heat flux test were carried out on the mock-ups in order to evaluate the heat removal capacity.

  18. Development of high poloidal beta, steady-state scenario with ITER-like tungsten divertor on EAST

    Science.gov (United States)

    Garofalo, A. M.; Gong, X. Z.; Qian, J.; Chen, J.; Li, G.; Li, K.; Li, M. H.; Zhai, X.; Bonoli, P.; Brower, D.; Cao, L.; Cui, L.; Ding, S.; Ding, W. X.; Guo, W.; Holcomb, C.; Huang, J.; Hyatt, A.; Lanctot, M.; Lao, L. L.; Liu, H.; Lyu, B.; McClenaghan, J.; Peysson, Y.; Ren, Q.; Shiraiwa, S.; Solomon, W.; Zang, Q.; Wan, B.

    2017-07-01

    Recent experiments on EAST have achieved the first long pulse H-mode (61 s) with zero loop voltage and an ITER-like tungsten divertor, and have demonstrated access to broad plasma current profiles by increasing the density in fully-noninductive lower hybrid current-driven discharges. These long pulse discharges reach wall thermal and particle balance, exhibit stationary good confinement (H 98y2 ~ 1.1) with low core electron transport, and are only possible with optimal active cooling of the tungsten armors. In separate experiments, the electron density was systematically varied in order to study its effect on the deposition profile of the external lower hybrid current drive (LHCD), while keeping the plasma in fully-noninductive conditions and with divertor strike points on the tungsten divertor. A broadening of the current profile is found, as indicated by lower values of the internal inductance at higher density. A broad current profile is attractive because, among other reasons, it enables internal transport barriers at large minor radius, leading to improved confinement as shown in companion DIII-D experiments. These experiments strengthen the physics basis for achieving high performance, steady state discharges in future burning plasmas.

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

    CERN Document Server

    Pankin, A Y; Kritz, A H; Park, G Y; Chang, C S; Brunner, D; Groebner, R J; Hughes, J W; LaBombard, B; Terry, J L; Ku, S

    2015-01-01

    The guiding-center kinetic neoclassical transport code, XGC0, [C.S. 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 width of the XGC0 neoclassical heat-load is approximately inversely proportional to the total plasma current $I_{\\rm p}$. 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 collisio...

  20. Optimizing stability, transport, and divertor operation through plasma shaping for steady-state scenario development in DIII-Da)

    Science.gov (United States)

    Holcomb, C. T.; Ferron, J. R.; Luce, T. C.; Petrie, T. W.; Politzer, P. A.; Challis, C.; DeBoo, J. C.; Doyle, E. J.; Greenfield, C. M.; Groebner, R. J.; Groth, M.; Hyatt, A. W.; Jackson, G. L.; Kessel, C.; La Haye, R. J.; Makowski, M. A.; McKee, G. R.; Murakami, M.; Osborne, T. H.; Park, J.-M.; Prater, R.; Porter, G. D.; Reimerdes, H.; Rhodes, T. L.; Shafer, M. W.; Snyder, P. B.; Turnbull, A. D.; West, W. P.

    2009-05-01

    Recent studies on the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] have elucidated key aspects of the dependence of stability, confinement, and density control on the plasma magnetic configuration, leading to the demonstration of nearly noninductive operation for >1 s with pressure 30% above the ideal no-wall stability limit. Achieving fully noninductive tokamak operation requires high pressure, good confinement, and density control through divertor pumping. Plasma geometry affects all of these. Ideal magnetohydrodynamics modeling of external kink stability suggests that it may be optimized by adjusting the shape parameter known as squareness (ζ). Optimizing kink stability leads to an increase in the maximum stable pressure. Experiments confirm that stability varies strongly with ζ, in agreement with the modeling. Optimization of kink stability via ζ is concurrent with an increase in the H-mode edge pressure pedestal stability. Global energy confinement is optimized at the lowest ζ tested, with increased pedestal pressure and lower core transport. Adjusting the magnetic divertor balance about a double-null configuration optimizes density control for improved noninductive auxiliary current drive. The best density control is obtained with a slight imbalance toward the divertor opposite the ion grad(B) drift direction, consistent with modeling of these effects. These optimizations have been combined to achieve noninductive current fractions near unity for over 1 s with normalized pressure of 3.565%, and a normalized confinement factor of H98(y ,2)≈1.5.

  1. 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.

  2. Application of LBIC measurements for characterisation of triple junction solar cells

    Energy Technology Data Exchange (ETDEWEB)

    Kwarikunda, N., E-mail: Nicholas.kwarikunda@live.nmmu.ac.za [Nelson Mandela Metropolitan University, P.O. BOX 77000, Port Elizabeth, 6031 (South Africa); Makerere University, P.O. BOX 7062, Kampala (Uganda); Dyk, E.E. van; Vorster, F.J. [Nelson Mandela Metropolitan University, P.O. BOX 77000, Port Elizabeth, 6031 (South Africa); Okullo, W. [Makerere University, P.O. BOX 7062, Kampala (Uganda); Munji, M.K. [Kenyatta University, P.O. BOX 43844-00100, Nairobi (Kenya)

    2014-04-15

    In this study the Light Beam Induced Current (LBIC) imaging technique was used to characterise InGaP/InGaAs/Ge triple junction solar cells. The study focused on the use of monochromatic and solar light as beam probes to obtain photocurrent response maps from which the presence of any current reducing features on the solar cell were identified. Point illuminated current voltage (I–V) curves were obtained simultaneously while LBIC scanning measurements were being made. Curve fitting using an interval division algorithm based on the single diode model was performed to extract basic point device and performance parameters to give a rough indication of the functioning of the triple junction device. Using red and blue lasers as beam probes, reverse voltage breakdown was observed on the I–V curves which could be attributed to the Ge bottom subcell not being fully activated. The extracted parameters obtained when using monochromatic and solar light beam probes showed a large variation, indicating the dependence of I–V parameters on the spectral content of the beam probe.

  3. Probing CP violation systematically in differential distributions

    CERN Document Server

    Durieux, Gauthier

    2015-01-01

    We revisit the topic of triple-product asymmetries which probe CP violation through differential distributions. We construct distributions with well-defined discrete symmetry properties and characterize the asymmetries formed upon them. It is stressed that the simplest asymmetries may not be optimal. We explore systematic generalizations having limited reliance on the process dynamics and phase-space parametrization. They exploit larger fractions of the information contained in differential distributions and may lead to increased sensitivities to CP violation. Our detailed treatment of the case of spinless four-body decays paves the way for further experimental studies.

  4. Building a Distributed Infrastructure for Scalable Triple Stores

    Institute of Scientific and Technical Information of China (English)

    Jing Zhou; Wendy Hall; David De Roure

    2009-01-01

    Built specifically for the Semantic Web, triple stores are required to accommodate a large number of RDF triples and remain primarily centralized. As triple stores grow and evolve with time, there is a demanding need for scalable techniques to remove resource and performance bottlenecks in such systems. To this end, we propose a fully decentralized peer-to-peer architecture for large scale triple stores in which triples are maintained by individual stakeholders, and a semantics-directed search protocol, mediated by topology reorganization, for locating triples of interest. We test our design through simulations and the results show anticipated improvements over existing techniques for distributed triple stores. In addition to engineering future large scale triple stores, our work will in particular benefit the federation of stand-alone triple stores of today to achieve desired scalability.

  5. Nano-analysis of grain boundary and triple junction transport in nanocrystalline Ni/Cu.

    Science.gov (United States)

    Reda Chellali, Mohammed; Balogh, Zoltan; Schmitz, Guido

    2013-09-01

    Nanocrystalline materials are distinguished by a high density of structural defects and grain boundaries. Due to the small grain size, a particular defect of the grain boundary topology, the so-called triple junction takes a dominant role for grain growth and atomic transport. We demonstrate by atom probe tomography that triple junctions in nanocrystalline Cu have 100-300 times higher diffusivity of Ni than standard high angle grain boundaries. Also, a previously unexpected systematic variation of the grain boundary width with temperature is detected. The impurity segregation layer at the grain boundaries grows from the 0.7 nm at 563 K to 2.5 nm at 643 K. This variation is clearly not controlled by simple bulk diffusion. Taking this effect into consideration, the activation energies for Ni diffusion in triple junctions and grain boundaries in Cu can be determined to be (83 ± 10) and (120 ± 15) kJ/mol, respectively. Thus, triple junctions are distinguished by considerably lower activation energy with respect to grain boundaries.

  6. Coherent spectroscopy of a strongly driven triple quantum dot molecule

    Institute of Scientific and Technical Information of China (English)

    Xie Yan; Duan Su-Qing; Chu Wei-Dong; Yang Ning

    2010-01-01

    Based on a calculation model, we study the interference phenomena of serially coupled V-type and A-type triple quantum dots (CTQDs) driven simultaneously by a strong driving field and a weak probe field. Strongly depending on the configuration of the three-level CTQD, the probe absorption spectra, which are shown in the tunneling current,exhibit various quantum coherence properties. In the case where the two pairs of transitions of the CTQD have a small eigenfrequency difference , the double-coupling effect of the driving field results in two Autler-Townes doublets and one weak Mollow triplet in one spectrum. With the value of increasing, only one Autler-Townes splitting remains due to the single-coupling of the field. We also find that the effect of spontaneous emission of phonons may lead to an obvious background current, which can be used to distinguish which transition is driven by the driving field in experiment. The interesting quantum property of a CTQD revealed in our results suggests its potential applications in quantum modulators and quantum logic devices.

  7. Vapor shielding models and the energy absorbed by divertor targets during transient events

    Energy Technology Data Exchange (ETDEWEB)

    Skovorodin, D. I., E-mail: dskovorodin@gmail.com; Arakcheev, A. S. [Budker Institute of Nuclear Physics, Novosibirsk 630090 (Russian Federation); Pshenov, A. A.; Eksaeva, E. A.; Marenkov, E. D.; Krasheninnikov, S. I. [National Research Nuclear University MEPhI, Moscow 115409 (Russian Federation)

    2016-02-15

    The erosion of divertor targets caused by high heat fluxes during transients is a serious threat to ITER operation, as it is going to be the main factor determining the divertor lifetime. Under the influence of extreme heat fluxes, the surface temperature of plasma facing components can reach some certain threshold, leading to an onset of intense material evaporation. The latter results in formation of cold dense vapor and secondary plasma cloud. This layer effectively absorbs the energy of the incident plasma flow, turning it into its own kinetic and internal energy and radiating it. This so called vapor shielding is a phenomenon that may help mitigating the erosion during transient events. In particular, the vapor shielding results in saturation of energy (per unit surface area) accumulated by the target during single pulse of heat load at some level E{sub max}. Matching this value is one of the possible tests to verify complicated numerical codes, developed to calculate the erosion rate during abnormal events in tokamaks. The paper presents three very different models of vapor shielding, demonstrating that E{sub max} depends strongly on the heat pulse duration, thermodynamic properties, and evaporation energy of the irradiated target material. While its dependence on the other shielding details such as radiation capabilities of material and dynamics of the vapor cloud is logarithmically weak. The reason for this is a strong (exponential) dependence of the target material evaporation rate, and therefore the “strength” of vapor shield on the target surface temperature. As a result, the influence of the vapor shielding phenomena details, such as radiation transport in the vapor cloud and evaporated material dynamics, on the E{sub max} is virtually completely masked by the strong dependence of the evaporation rate on the target surface temperature. However, the very same details define the amount of evaporated particles, needed to provide an effective shielding

  8. Fluid-particle hybrid simulation on the transports of plasma, recycling neutrals, and carbon impurities in the Korea Superconducting Tokamak Advanced Research divertor region

    Science.gov (United States)

    Kim, Deok-Kyu; Hong, Sang Hee

    2005-06-01

    A two-dimensional simulation modeling that has been performed in a self-consistent way for analysis on the fully coupled transports of plasma, recycling neutrals, and intrinsic carbon impurities in the divertor domain of tokamaks is presented. The numerical model coupling the three major species transports in the tokamak edge is based on a fluid-particle hybrid approach where the plasma is described as a single magnetohydrodynamic fluid while the neutrals and impurities are treated as kinetic particles using the Monte Carlo technique. This simulation code is applied to the KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak [G. S. Lee, J. Kim, S. M. Hwang et al., Nucl. Fusion 40, 575 (2000)] to calculate the peak heat flux on the divertor plate and to explore the divertor plasma behavior depending on the upstream conditions in its base line operation mode for various values of input heating power and separatrix plasma density. The numerical modeling for the KSTAR tokamak shows that its full-powered operation is subject to the peak heat loads on the divertor plate exceeding an engineering limit, and reveals that the recycling zone is formed in front of the divertor by increasing plasma density and by reducing power flow into the scrape-off layer. Compared with other researchers' work, the present hybrid simulation more rigorously reproduces severe electron pressure losses along field lines by the presence of recycling zone accounting for the transitions between the sheath limited and the detached divertor regimes. The substantial profile changes in carbon impurity population and ionic composition also represent the key features of this divertor regime transition.

  9. On Split Lie Triple Systems II

    Indian Academy of Sciences (India)

    Antonio J Calderón Martín; M Forero Piulestán

    2010-04-01

    In [4] it is studied that the structure of split Lie triple systems with a coherent 0-root space, that is, satisfying $[T_0,T_0,T]=0$ and $[T_0,T_,T_0]≠ 0$ for any nonzero root and where $T_0$ denotes the 0-root space and $T_$ the -root space, by showing that any of such triple systems with a symmetric root system is of the form $T=\\mathcal{U}+\\sum_j I_j$ with $\\mathcal{U}$ a subspace of the 0-root space $T_0$ and any $I_j$ a well described ideal of , satisfying $[I_j,T,I_k]=0$ if $j≠ k$. It is also shown in [4] that under certain conditions, a split Lie triple system with a coherent 0-root space is the direct sum of the family of its minimal ideals, each one being a simple split Lie triple system, and the simplicity of is characterized. In the present paper we extend these results to arbitrary split Lie triple systems with no restrictions on their 0-root spaces.

  10. A sensitive and versatile "signal-on" electrochemical aptasensor based on a triple-helix molecular switch.

    Science.gov (United States)

    Wang, Xiuzhong; Jiang, Aiwen; Hou, Ting; Li, Feng

    2014-12-07

    In the present study, a versatile "signal-on" electrochemical aptasensor based on a triple-helix molecular switch has been developed. An aptamer probe is designed to hybridize with the methylene blue (MB)-modified DNA capture probe immobilized on the gold electrode to form rigid triple-helix DNA, impeding the efficient electron transfer of MB to the electrode and resulting in the decreased oxidation peak current of MB. However, upon introduction of the perfectly matched target, for example, human α-thrombin (Tmb), the interaction between Tmb and the aptamer probe leads to the dissociation of the triple-helix DNA structure and thereby liberates the MB-modified end of the capture probe, allowing the MB to collide with the electrode surface and resulting in an increase of the oxidation peak currents of MB. Therefore, the sensitive signal-on detection of Tmb is realized, and the detection limit of Tmb is 0.12 nM. The proposed approach also demonstrates excellent regenerability, reproducibility and stability. Additionally, it also has the advantages of simplicity in design and easy operation. The success in the present biosensor provides a promising alternative to the electrochemical detection of a variety of analytes and may have potential applications in point-of-care testing and clinical diagnosis.

  11. Proximal Probes Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Proximal Probes Facility consists of laboratories for microscopy, spectroscopy, and probing of nanostructured materials and their functional properties. At the...

  12. Triple helix interactions for eco-innovation

    DEFF Research Database (Denmark)

    Hermann, Roberto Rivas; Riisgaard, Henrik; Remmen, Arne

    Authority with insights from consultants, universities and donnor agencies. The proximity of the science park to the canal, has hitherto not yielded with the creation of a “green cluster”, which could be a precedent to promote eco-innovations. These findings suggest that, Triple Helix interactions...... the role of science parks in promoting eco-innovation. This study uses qualitative data gathered in two units of analysis: Panama Canal Authority and City of Knowledge Science Park. The study examines how Triple Helix interactions have built the regional system of eco-innovation at the Panama Canal...... are not institutionalized but take place through adhoc projects. Further, science parks could become mediators in Triple Helix interactions between industry, universities and governments....

  13. Double Fell bundles and Spectral triples

    CERN Document Server

    Martins, Rachel A D

    2007-01-01

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

  14. The triple point of sulfur hexafluoride

    Science.gov (United States)

    Rourke, P. M. C.

    2016-04-01

    A cryogenic fixed point cell has been filled with high purity (99.999%) sulfur hexafluoride (SF6) and measured in an adiabatic closed-cycle cryostat system. Temperature measurements of the SF6 melting curve were performed using a capsule-type standard platinum resistance thermometer (CSPRT) calibrated over the International Temperature Scale of 1990 (ITS-90) subrange from the triple point of equilibrium hydrogen to the triple point of water. The measured temperatures were corrected by 0.37 mK for the effects of thermometer self-heating, and the liquidus-point temperature estimated by extrapolation to melted fraction F  =  1 of a simple linear regression versus melted fraction F in the range F  =  0.53 to 0.84. Based on this measurement, the temperature of the triple point of sulfur hexafluoride is shown to be 223.555 23(49) K (k  =  1) on the ITS-90. This value is in excellent agreement with the best prior measurements reported in the literature, but with considerably smaller uncertainty. An analysis of the detailed uncertainty budget of this measurement suggests that if the triple point of sulfur hexafluoride were to be included as a defining fixed point of the next revision of the International Temperature Scale, it could do so with a total realization uncertainty of approximately 0.43 mK, slightly larger than the realization uncertainties of the defining fixed points of the ITS-90. Since the combined standard uncertainty of this SF6 triple point temperature determination is dominated by chemical impurity effects, further research exploring gas purification techniques and the influence of specific impurity species on the SF6 triple point temperature may bring the realization uncertainty of SF6 as a fixed point material into the range of the defining fixed points of the ITS-90.

  15. Normalization for triple-target microarray experiments

    Directory of Open Access Journals (Sweden)

    Magniette Frederic

    2008-04-01

    Full Text Available Abstract Background Most microarray studies are made using labelling with one or two dyes which allows the hybridization of one or two samples on the same slide. In such experiments, the most frequently used dyes are Cy3 and Cy5. Recent improvements in the technology (dye-labelling, scanner and, image analysis allow hybridization up to four samples simultaneously. The two additional dyes are Alexa488 and Alexa494. The triple-target or four-target technology is very promising, since it allows more flexibility in the design of experiments, an increase in the statistical power when comparing gene expressions induced by different conditions and a scaled down number of slides. However, there have been few methods proposed for statistical analysis of such data. Moreover the lowess correction of the global dye effect is available for only two-color experiments, and even if its application can be derived, it does not allow simultaneous correction of the raw data. Results We propose a two-step normalization procedure for triple-target experiments. First the dye bleeding is evaluated and corrected if necessary. Then the signal in each channel is normalized using a generalized lowess procedure to correct a global dye bias. The normalization procedure is validated using triple-self experiments and by comparing the results of triple-target and two-color experiments. Although the focus is on triple-target microarrays, the proposed method can be used to normalize p differently labelled targets co-hybridized on a same array, for any value of p greater than 2. Conclusion The proposed normalization procedure is effective: the technical biases are reduced, the number of false positives is under control in the analysis of differentially expressed genes, and the triple-target experiments are more powerful than the corresponding two-color experiments. There is room for improving the microarray experiments by simultaneously hybridizing more than two samples.

  16. Viewgraphs presented at the ASDEX/DOE workshop on disruptions in divertor tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Granetz, R.; Gruber, O.; Zohm, H. [and others

    1994-09-01

    The emphasis of this year`s ASDEX/DOE workshop was on disruptions in diverted tokamaks. The meeting was held here at MIT on 14--15 March. It is particularly appropriate that MIT hosted the workshop this year, since Alcator C-Mod had just recently completed its very first run campaign, and disruptions are one of the key areas of research in our program. There were a total of 14 speakers, with participants from IPP (Garching), CRPP (Lausanne), Culham, General Atomics, PPPL, Sandia, ORNL, the ITER JCT, and MIT. The subjects addressed included statistical analysis of disruption probabilities in ASDEX, modelling of the vertical axisymmetric plasma motion in DIII-D, impact of disruptions on the design of the ITER divertors, modelling of runaway electrons, and TSC calculations of disruption-induced currents and forces in TPX, etc. One item of particular interest to us was the experimental correlation of halo current magnitude with plasma current on ASDEX-Upgrade. The data indicates at least a linear, and possibly even a quadractic dependence. This has important implications for Alcator C-Mod, since it would predict halo currents of order 1 MA or more at full performance. At the conclusion of the talks, an informal discussion of disruption databases was held, primarily for the purpose of helping us develop a useful one for C-Mod.

  17. Design of a water cooled monoblock divertor for DEMO using Eurofer as structural material

    Energy Technology Data Exchange (ETDEWEB)

    Richou, Marianne, E-mail: marianne.richou@cea.fr [CEA, IRFM, F-13108 Saint Paul Lez Durance (France); Li-Puma, Antonella [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Visca, Eliseo [Associazione EURATOM-ENEA sulla Fusione, C.R. Frascati, IT-00044 Frascati (Italy)

    2014-10-15

    The performed investigation focus on a monoblock type design for a water cooled DEMO divertor using Eurofer as structural material. In 2013, a study case of such a concept was presented. It was shown that basic concepts using Eurofer as structural material are limited to an incident heat flux of 8 MW m{sup −2}. Since, the EFDA agency issued new specifications. In this study, the conceptual design is reassessed with regard to specifications. Then, steady state thermal analyses and thermo-mechanical elastic analyses have been performed to define an upgrade of the geometry taking into account new specifications, design criteria and the maximum heat flux requirement of 10 MW m{sup −2}. An analysis of the influence of each adjustable geometrical parameter on thermo-mechanical design criteria was performed. As a consequence, geometrical parameters were set in order to fit to materials requirements. For defined hydraulic conditions taken in the most favourable configuration, the limit of this design is estimated to an incident heat flux of 10 MW m{sup −2}. Margin to critical heat flux and rules against progressive deformation/ratcheting in structural material limit the design.

  18. Tearing mode physics studies applying the dynamic ergodic divertor on TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Koslowski, H R [Forschungszentrum Juelich GmbH, Institut fuer Plasmaphysik, Association EURATOM-FZ Juelich, D-52425 Juelich (Germany); Westerhof, E [FOM-Institute for Plasmaphysics ' Rijnhuizen' , Association EURATOM-FOM, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Bock, M de [FOM-Institute for Plasmaphysics ' Rijnhuizen' , Association EURATOM-FOM, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Classen, I [FOM-Institute for Plasmaphysics ' Rijnhuizen' , Association EURATOM-FOM, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Jaspers, R [FOM-Institute for Plasmaphysics ' Rijnhuizen' , Association EURATOM-FOM, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Kikuchi, Y [Forschungszentrum Juelich GmbH, Institut fuer Plasmaphysik, Association EURATOM-FZ Juelich, D-52425 Juelich (Germany); Kraemer-Flecken, A [Forschungszentrum Juelich GmbH, Institut fuer Plasmaphysik, Association EURATOM-FZ Juelich, D-52425 Juelich (Germany); Lazaros, A [FOM-Institute for Plasmaphysics ' Rijnhuizen' , Association EURATOM-FOM, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Liang, Y [Forschungszentrum Juelich GmbH, Institut fuer Plasmaphysik, Association EURATOM-FZ Juelich, D-52425 Juelich (Germany); Loewenbrueck, K [Forschungszentrum Juelich GmbH, Institut fuer Plasmaphysik, Association EURATOM-FZ Juelich, D-52425 Juelich (Germany); Varshney, S [FOM-Institute for Plasmaphysics ' Rijnhuizen' , Association EURATOM-FOM, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Hellermann, M von [FOM-Institute for Plasmaphysics ' Rijnhuizen' , Association EURATOM-FOM, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Wolf, R [Forschungszentrum Juelich GmbH, Institut fuer Plasmaphysik, Association EURATOM-FZ Juelich, D-52425 Juelich (Germany); Zimmermann, O [Forschungszentrum Juelich GmbH, Institut fuer Plasmaphysik, Association EURATOM-FZ Juelich, D-52425 Juelich (Germany)

    2006-12-15

    The dynamic ergodic divertor (DED) on the TEXTOR tokamak allows for the reproducible destabilization of the m/n = 2/1 tearing mode which is phase locked to the external static or rotating perturbation field. In combination with its flexible heating systems (co- and counter-neutral beam injection, ion cyclotron resonance heating, electron cyclotron resonance heating (ECRH) with steerable launcher) dedicated experiments to study the mode onset, properties of large islands and mode stabilization can be performed. The dependence of the mode excitation threshold (field penetration) on the plasma rotation shows a resonance character, with minimum threshold when the external perturbation frequency matches the MHD frequency of the 2/1 mode. Mode stabilization by ECRH heating shows that for the TEXTOR plasma heating is more effective than the current drive in O-point. Extrapolation to ITER yields a significant contribution to the mode suppression originating from the temperature increase within the island. Alfven-like modes, which have been previously identified in the vicinity of large islands on FTU (Buratti et al 2005 Nuclear Fusion 45 1446), are found to be created already before island formation above a certain threshold of the externally applied perturbation field.

  19. Manufacturing and testing in reactor relevant conditions of brazed plasma facing components of the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Bisio, M. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy); Branca, V. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy); Marco, M. Di [FN s.p.a., ss 35 bis dei Giovi km 15, I-15062 Bosco Marengo (Albania) (Italy); Federici, A. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy); Grattarola, M. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy)]. E-mail: grattarola@ansaldo.it; Gualco, G. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy); Guarnone, P. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy); Luconi, U. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy); Merola, M. [EFDA, Boltzmanstr. 2, D-85748 Garching (Germany); Ozzano, C. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy); Pasquale, G. [FN s.p.a., ss 35 bis dei Giovi km 15, I-15062 Bosco Marengo (AL) (Italy); Poggi, P. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy); Rizzo, S. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy); Varone, F. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy)

    2005-11-15

    A fabrication route based on brazing technology has been developed for the realization of the high heat flux components for the ITER vertical target and Dome-Liner. The divertor vertical target is armoured with carbon fiber reinforced carbon and tungsten in the lower straight part and in the upper curved part, respectively. The armour material is joined to heat sinks made of precipitation hardened copper-chromium-zirconium alloy. The plasma facing units of the dome component are based on a tungsten flat tile design with hypervapotron cooling. An innovative brazing technique based on the addition of carbon fibers to the active brazing alloy, developed by Ansaldo Ricerche for applications in the field of the energy production, has been used for the carbon fiber composite to copper joint to reduce residual stresses. The tungsten-copper joint has been realized by direct casting. A proper brazing thermal cycle has been studied to guarantee the required mechanical properties of the precipitation hardened alloy after brazing. The fabrication route of plasma facing components for the ITER vertical target and dome based on the brazing technology has been proved by means of thermal fatigue tests performed on mock-ups in reactor relevant conditions.

  20. Study and simulation of carbon impurity dynamics near the ergodic divertor in Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Giannella, R.; Cordier, J.J.; Corre, Y.; Ghendrih, P.; Guirlet, R.; Gunn, J. [Association Euratom-CEA, CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee (DRFC), 13 - Saint-Paul-lez-Durance (France); Hogan, J. [Oak Ridge National Lab., TN (United States)

    1999-10-15

    In the past few years, effects induced by the ergodic dive such as impurity screening and transport modifications in the plasma edge have been used to achieve high radiation, low contamination regimes. A crucial issue in understanding these effects is that of impurity generation and propagation across the plasma edge, especially in the vicinity of the Ergodic Divertor (ED) neutralizer plates. A variety of diagnostic tools and techniques are used for this purpose. In the case of Tore Supra, interpretation of spectroscopic data is strongly complicated by the complex geometry of the ED, leading among other effects to the total lack of uniformity of the sources. Indeed, due to the specific pattern of impurity sources on the neutralizers and to their particular orientation with respect to the local magnetic field, densities of lowly ionised impurities are deeply modulated on the sub-centimeter scale in both directions perpendicular to the magnetic field. Because of this, accurate 3-D simulations are essential for the evaluation of experimental signals. (authors)

  1. Kinetic studies of divertor heat fluxes in Alcator C-Mod

    Science.gov (United States)

    Pankin, A. Y.; Bateman, G.; Kritz, A. H.; Rafiq, T.; Park, G. Y.; Chang, C. S.; Brunner, D.; Hughes, J. W.; Labombard, B.; Terry, J.

    2010-11-01

    The kinetic XGC0 code [C.S. Chang et al, Phys. Plasmas 11 (2004) 2649] is used to model the H- mode pedestal and SOL regions in Alcator C-Mod discharges. The self-consistent simulations in this study include kinetic neoclassical physics and anomalous transport models along with the ExB flow shear effects. The heat fluxes on the divertor plates are computed and the fluxes to the outer plate are compared with experimental observations. The dynamics of the radial electric field near the separatrix and in the SOL region are computed with the XGC0 code, and the effect of the anomalous transport on the heat fluxes in the SOL region is investigated. In particular, the particle and thermal diffusivities obtained in the analysis mode are compared with predictions from the theory-based anomalous transport models such as MMM95 [G. Bateman et al, Phys. Plasmas 5 (1998) 1793] and DRIBM [T. Rafiq et al, to appear in Phys. Plasmas (2010)]. It is found that there is a notable pinch effect in the inner separatrix region. Possible physical mechanisms for the particle and thermal pinches are discussed.

  2. 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.

  3. 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.

  4. Characterization of Impurities in Tokamak Divertor Plasmas from Analysis of Spectral Profiles

    Science.gov (United States)

    Isler, R. C.; Brooks, N. H.; Zaniol, B.

    2002-12-01

    Studies of the production, transport, and radiative losses of impurities in present-day tokamak divertors provide input necessary for the design of future burning- plasma machines. Several types of information rely on detailed analysis of emission profiles. These include ion temperatures, ion flows along field lines, and impurity production mechanisms. Temperatures and flows are determined from Doppler broadening and shifts by comparing measured line shapes to theoretical profiles that include the nonlinear Zeeman/Paschen-Back effect. The two major production mechanisms for atomic carbon are physical and chemical sputtering. These processes can be distinguished by comparing atomic and molecular fluxes, which requires modeling the band emissions of CD and C2. They can also be differentiated from measurements of effective temperatures of C I (best profile fits to thermal distributions). Careful inspection of profiles that give high effective temperatures reveals that they are not actually Gaussian but have asymmetries and shifts that can be correlated to energy distributions expected for physical sputtering. Examples of all these applications are discussed in this review.

  5. Applicability of the dielectric barrier discharge for helium ash measurements in the divertor region

    Directory of Open Access Journals (Sweden)

    Książek Ireneusz

    2016-06-01

    Full Text Available Controlled fusion based on the magnetic confinement of the plasma is one of the main aims of the Euro-fusion programme. In the fusion device, the hydrogen isotopes, in nuclear reactions, will produce helium nuclei. The products, as the ash, will be removed from the plasma in the region of the so-called divertor. Controlling the helium to hydrogen ratio in this ‘exhaust gas’ will provide information about the efficiency of the fusion process as well as of the efficiency of the helium removal system. One of the methods to perform this task is to study the properties of the discharge conducted in such exhaust gas. In this paper, the applicability of the dielectric barrier discharge (DBD is studied. This preliminary experiment shows a great potential in applicability of this kind of discharge. The optical as well as pulse-height spectra were studied, both revealing very promising properties. In the optical spectrum, one can observe well separated hydrogen and helium spectral lines, with intensities of the same order of magnitude. Moreover, in the registered spectral region, the molecular spectra are negligible. The pulse-height spectra reveal very distinct shape in helium and hydrogen. Checking of this spectrum could provide parallel (redundant information about the partial pressure of helium in the magnetic confinement fusion (MCF device exhaust gas.

  6. 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.)

  7. Interfacial metallurgy study of brazed joints between tungsten and fusion related materials for divertor design

    Science.gov (United States)

    Zhang, Yuxuan; Galloway, Alexander; Wood, James; Robbie, Mikael Brian Olsson; Easton, David; Zhu, Wenzhong

    2014-11-01

    In the developing DEMO divertor, the design of joints between tungsten to other fusion related materials is a significant challenge as a result of the dissimilar physical metallurgy of the materials to be joined. This paper focuses on the design and fabrication of dissimilar brazed joints between tungsten and fusion relevant materials such as EUROFER 97, oxygen-free high thermal conductivity (OFHC) Cu and SS316L using a gold based brazing foil. The main objectives are to develop acceptable brazing procedures for dissimilar joining of tungsten to other fusion compliant materials and to advance the metallurgical understanding within the interfacial region of the brazed joint. Four different butt-type brazed joints were created and characterised, each of which were joined with the aid of a thin brazing foil (Au80Cu19Fe1, in wt.%). Microstructural characterisation and elemental mapping in the transition region of the joint was undertaken and, thereafter, the results were analysed as was the interfacial diffusion characteristics of each material combination produced. Nano-indentation tests are performed at the joint regions and correlated with element composition information in order to understand the effects of diffused elements on mechanical properties. The experimental procedures of specimen fabrication and material characterisation methods are presented. The results of elemental transitions after brazing are reported. Elastic modulus and nano-hardness of each brazed joints are reported.

  8. Triple arthrodesis for adult acquired flatfoot.

    Science.gov (United States)

    Catanzariti, Alan R; Dix, Brian T; Richardson, Phillip E; Mendicino, Robert W

    2014-07-01

    The primary goal of triple arthrodesis for stage III and IV adult acquired flatfoot is to obtain a well-aligned plantigrade foot that will support the ankle in optimal alignment. Ancillary procedures including posterior muscle group lengthening, medial displacement calcaneal osteotomy, medial column stabilization, peroneus brevis tenotomy, or transfer and harvest of regional bone graft are often necessary to achieve adequate realignment. Image intensification is helpful in confirming optimal realignment before fixation. Results of triple arthrodesis are enhanced with adequate preparation of joint surfaces, bone graft/orthobiologics, 2-point fixation of all 3 tritarsal joints, and a vertical heel position.

  9. Triple Bragg diffraction in paratellurite crystal

    Science.gov (United States)

    Kotov, V. M.; Averin, S. V.; Voronko, A. I.; Kotov, E. V.; Tikhomirov, S. A.

    2017-07-01

    Triple Bragg diffraction in a paratellurite crystal has been considered for the case when the plane of diffraction is oblique to the optical axis of the crystal. It has been shown that effective photoelastic constants for isotropic and anisotropic diffraction depend on the inclination of the plane of diffraction insignificantly. Triple Bragg diffraction of 0.63-μm coherent radiation in paratellurite at a 47.3-MHz slow acoustic wave has been experimentally demonstrated. For an optical power of 0.69 W delivered to a piezoconverter, the relative intensities of diffraction orders equal 0.4, 0.4, 0.1, and 0.1, respectively.

  10. $\\kappa$-Deformation and Spectral Triples

    CERN Document Server

    Iochum, B; Schücker, Th; Sitarz, A; 10.5506/APhysPolBSupp.4.305

    2011-01-01

    The aim of the paper is to answer the following question: does $\\kappa$-deformation fit into the framework of noncommutative geometry in the sense of spectral triples? Using a compactification of time, we get a discrete version of $\\kappa$-Minkowski deformation via $C^*$-algebras of groups. The dynamical system of the underlying groups (including some Baumslag--Solitar groups) is used in order to construct \\emph{finitely summable} spectral triples. This allows to bypass an obstruction to finite-summability appearing when using the common regular representation.

  11. A PZT Actuated Triple-Finger Gripper for Multi-Target Micromanipulation

    Directory of Open Access Journals (Sweden)

    Tao Chen

    2017-01-01

    Full Text Available This paper presents a triple-finger gripper driven by a piezoceramic (PZT transducer for multi-target micromanipulation. The gripper consists of three fingers assembled on adjustable pedestals with flexible hinges for a large adjustable range. Each finger has a PZT actuator, an amplifying structure, and a changeable end effector. The moving trajectories of single and double fingers were calculated and finite element analyses were performed to verify the reliability of the structures. In the gripping experiment, various end effectors of the fingers such as tungsten probes and fibers were tested, and different micro-objects such as glass hollow spheres and iron spheres with diameters ranging from 10 to 800 μm were picked and released. The output resolution is 145 nm/V, and the driven displacement range of the gripper is 43.4 μm. The PZT actuated triple-finger gripper has superior adaptability, high efficiency, and a low cost.

  12. T-Odd Triple-Product Correlations in Hadronic b Decays

    OpenAIRE

    Bensalem, W.; London, D.

    2000-01-01

    We study T-violating triple-product asymmetries in the quark-level decay b -> s u u(bar) within the standard model (SM). We find that only two types of triple products are non-negligible. The asymmetry in p_u . [s_u x s_u(bar)] or p_u(bar) . [s_u x s_u(bar)] can be as large as about 5%. It can be probed in B -> V_1 V_2 decays, where V_1 and V_2 are vector mesons. And the asymmetry in s_b . [p_u x p_s] can be in the range 1%--3%. One can search for this signal in the decay \\Lambda_b -> \\Lambda...

  13. 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.

  14. Electron Density Measurements in the National Spherical Torus Experiment Detached Divertor Region Using Stark Broadening of Deuterium Infrared Paschen Emission Lines

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V A; Johnson, D W; Kaita, R; Roquemore, A L

    2007-04-27

    Spatially resolved measurements of deuterium Balmer and Paschen line emission have been performed in the divertor region of the National Spherical Torus Experiment using a commercial 0.5 m Czerny-Turner spectrometer. While the Balmer emission lines, Balmer and Paschen continua in the ultraviolet and visible regions have been extensively used for tokamak divertor plasma temperature and density measurements, the diagnostic potential of infrared Paschen lines has been largely overlooked. We analyze Stark broadening of the lines corresponding to 2-n and 3-m transitions with principle quantum numbers n = 7-12 and m = 10-12 using recent Model Microfield Method calculations (C. Stehle and R. Hutcheon, Astron. Astrophys. Supl. Ser. 140, 93 (1999)). Densities in the range (5-50) x 10{sup 19} m{sup -3} are obtained in the recombining inner divertor plasma in 2-6 MW NBI H-mode discharges. The measured Paschen line profiles show good sensitivity to Stark effects, and low sensitivity to instrumental and Doppler broadening. The lines are situated in the near-infrared wavelength domain, where optical signal extraction schemes for harsh nuclear environments are practically realizable, and where a recombining divertor plasma is optically thin. These properties make them an attractive recombining divertor density diagnostic for a burning plasma experiment.

  15. MEIR-KEELER TYPE CONTRACTIONS FOR TRIPLED FIXED POINTS

    Institute of Scientific and Technical Information of China (English)

    Hassen Aydi; Erdal Karapinar; Calogero Vetro

    2012-01-01

    In 2011,Berinde and Borcut [6] introduced the notion of tripled fixed point in partially ordered metric spaces.In our paper,we give some new tripled fixed point theorems by using a generalization of Meir-Keeler contraction.

  16. Aptamer/target binding-induced triple helix forming for signal-on electrochemical biosensing.

    Science.gov (United States)

    Mao, Yinfei; Liu, Jinquan; He, Dinggen; He, Xiaoxiao; Wang, Kemin; Shi, Hui; Wen, Li

    2015-10-01

    Owing to its diversified structures, high affinity, and specificity for binding a wide range of non-nucleic acid targets, aptamer is a useful molecular recognition tool for the design of various biosensors. Herein, we report a new signal-on electrochemical biosensing platform which is based on an aptamer/target binding-induced strand displacement and triple-helix forming. The biosensing platform is composed of a signal transduction probe (STP) modified with a methylene blue (MB) and a sulfhydryl group, a triplex-forming oligonucleotides probe (TFO) and a target specific aptamer probe (Apt). Through hybridization with the TFO probe and the Apt probe, the self-assembled STP on Au electrode via Au-S bonding keeps its rigid structure. The MB on the STP is distal to the Au electrode surface. It is eT off state. Target binding releases the Apt probe and liberates the end of the MB tagged STP to fold back and form a triplex-helix structure with TFO (STP/TFO/STP), allowing MB to approach the Au electrode surface and generating measurable electrochemical signals (eT ON). As test for the feasibility and universality of this signal-on electrochemical biosensing platform, two aptamers which bind to adenosine triphosphate (ATP) and human α-thrombin (Tmb), respectively, are selected as models. The detection limit of ATP was 7.2 nM, whereas the detection limit of Tmb was 0.86 nM.

  17. Triple Antibiotic Polymer Nanofibers for Intracanal Drug Delivery: Effects on Dual Species Biofilm and Cell Function.

    Science.gov (United States)

    Pankajakshan, Divya; Albuquerque, Maria T P; Evans, Joshua D; Kamocka, Malgorzata M; Gregory, Richard L; Bottino, Marco C

    2016-10-01

    Root canal disinfection and the establishment of an intracanal microenvironment conducive to the proliferation/differentiation of stem cells play a significant role in regenerative endodontics. This study was designed to (1) investigate the antimicrobial efficacy of triple antibiotic-containing nanofibers against a dual-species biofilm and (2) evaluate the ability of dental pulp stem cells (DPSCs) to adhere to and proliferate on dentin upon nanofiber exposure. Seven-day-old dual-species biofilm established on dentin specimens was exposed for 3 days to the following: saline (control), antibiotic-free nanofibers (control), and triple antibiotic-containing nanofibers or a saturated triple antibiotic paste (TAP) solution (50 mg/mL in phosphate buffer solution). Bacterial viability was assessed using the LIVE/DEAD assay (Molecular Probes, Inc, Eugene, OR) and confocal laser scanning microscopy. For cytocompatibility studies, dentin specimens after nanofiber or TAP (1 g/mL in phosphate buffer solution) exposure were evaluated for cell adhesion and spreading by actin-phalloidin staining. DPSC proliferation was assessed on days 1, 3, and 7. Statistics were performed, and significance was set at the 5% level. Confocal laser scanning microscopy showed significant bacterial death upon antibiotic-containing nanofiber exposure, differing significantly (P antibiotic-free fibers and the control (saline). DPSCs showed enhanced adhesion/spreading on dentin specimens treated with antibiotic-containing nanofibers when compared with its TAP counterparts. The DPSC proliferation rate was similar on days 1 and 3 in antibiotic-free nanofibers, triple antibiotic-containing nanofibers, and TAP-treated dentin. Proliferation was higher (9-fold) on dentin treated with antibiotic-containing nanofibers on day 7 compared with TAP. Triple antibiotic-containing polymer nanofibers led to significant bacterial death, whereas they did not affect DPSC attachment and proliferation on dentin

  18. Fast response of the optical nonlinearity in a GaAs/AlGaAs asymmetric triple quantum well structure

    CERN Document Server

    Ahn, S H; Sawaki, N

    1999-01-01

    The time response of the optical nonlinear behavior in a GaAs/AlGaAs asymmetric triple quantum well structure is estimated by using a picosecond pump-probe method at 77 K. From the results of the transmission of the probe pulse as a function of the delay time at the excitation wavelengths, a rise time of 5 approx 10 ps and a fall time of 8 approx 16 ps are obtained. The nonlinear behavior is attributed to the triple resonance of the electronic states due to the build-up of the internal field induced by the separation of photo-excited electrons and holes. It is found that the rise time is determined by the tunneling transfer time of the electrons in the narrowest well to an adjacent well separated by a thin potential barrier.

  19. Probe tip heating assembly

    Science.gov (United States)

    Schmitz, Roger William; Oh, Yunje

    2016-10-25

    A heating assembly configured for use in mechanical testing at a scale of microns or less. The heating assembly includes a probe tip assembly configured for coupling with a transducer of the mechanical testing system. The probe tip assembly includes a probe tip heater system having a heating element, a probe tip coupled with the probe tip heater system, and a heater socket assembly. The heater socket assembly, in one example, includes a yoke and a heater interface that form a socket within the heater socket assembly. The probe tip heater system, coupled with the probe tip, is slidably received and clamped within the socket.

  20. Revised Reynolds Stress and Triple Product Models

    Science.gov (United States)

    Olsen, Michael E.; Lillard, Randolph P.

    2017-01-01

    Revised versions of Lag methodology Reynolds-stress and triple product models are applied to accepted test cases to assess the improvement, or lack thereof, in the prediction capability of the models. The Bachalo-Johnson bump flow is shown as an example for this abstract submission.

  1. Targeting EGFR in Triple Negative Breast Cancer

    Directory of Open Access Journals (Sweden)

    Naoto T. Ueno, Dongwei Zhang

    2011-01-01

    Full Text Available Our preliminary data show that erlotinib inhibits Triple-negative breast cancer (TNBC in a xenograft model. However, inhibition of metastasis by erlotinib is accompanied by nonspecific effects because erlotinib can inhibit other kinases; thus, more direct targets that regulate TNBC metastasis need to be identified to improve its therapeutic efficacy.

  2. Triple-axis spectrometer DruechaL

    Energy Technology Data Exchange (ETDEWEB)

    Buehrer, W.; Keller, P. [Lab. for Neutron Scattering ETH Zurich, Zurich (Switzerland) and Paul Scherrer Institute, Villigen (Switzerland)

    1996-11-01

    DruechaL is a triple-axis spectrometer located at a cold guide. The characteristics of guide and instrument allow the use of a broad spectral range of neutrons. The resolution in momentum and energy transfer can be tuned to match the experimental requirements by using either collimators or focusing systems (monochromator, antitrumpet, analyser). (author) figs., tabs., refs.

  3. Resolution of a triple axis spectrometer

    DEFF Research Database (Denmark)

    Nielsen, Mourits; Bjerrum Møller, Hans

    1969-01-01

    A new method for obtaining the resolution function for a triple-axis neutron spectrometer is described, involving a combination of direct measurement and analytical calculation. All factors which contribute to the finite resolution of the instrument may be taken into account, and Gaussian...

  4. Triple mode filters with coaxial excitation

    NARCIS (Netherlands)

    Gerini, G.; Bustamante, F.D.; Guglielmi, M.

    2004-01-01

    In this paper we describe triple mode filters in a square waveguide with integrated coaxial input/output excitation. An important feature of the structure proposed is that it is easily amenable to an accurate full wave analysis. In addition to theory, a practical six pole filter with two transmissio

  5. A New Parity Formula: Triple T.

    Science.gov (United States)

    Newton, Richard F.

    Triple T, a graduate project to educate (train) teacher trainers, allows participants to view the schools as a total system and gives educators a means of training people to change that system. It offers an opportunity to develop an alternative means for graduate education. An important element in this program is parity. While parity is more a…

  6. Detecting Triple Systems with Gravitational Wave Observations

    Science.gov (United States)

    Meiron, Yohai; Kocsis, Bence; Loeb, Abraham

    2017-01-01

    The Laser Interferometer Gravitational Wave Observatory (LIGO) has recently discovered gravitational waves (GWs) emitted by merging black hole binaries. We examine whether future GW detections may identify triple companions of merging binaries. Such a triple companion causes variations in the GW signal due to: (1) the varying path length along the line of sight during the orbit around the center of mass; (2) relativistic beaming, Doppler, and gravitational redshift; (3) the variation of the “light”-travel time in the gravitational field of the triple companion; and (4) secular variations of the orbital elements. We find that the prospects for detecting a triple companion are the highest for low-mass compact object binaries which spend the longest time in the LIGO frequency band. In particular, for merging neutron star binaries, LIGO may detect a white dwarf or M-dwarf perturber at a signal-to-noise ratio of 8, if it is within 0.4 {R}ȯ distance from the binary and the system is within a distance of 100 Mpc. Stellar mass (supermassive) black hole perturbers may be detected at a factor 5 × (103×) larger separations. Such pertubers in orbit around a merging binary emit GWs at frequencies above 1 mHz detectable by the Laser Interferometer Space Antenna in coincidence.

  7. Detecting triple systems with gravitational wave observations

    CERN Document Server

    Meiron, Yohai; Loeb, Abraham

    2016-01-01

    The Laser Interferometer Gravitational Wave Observatory (LIGO) has recently discovered gravitational waves (GWs) emitted by merging black hole binaries. We examine whether future GW detections may identify triple companions of merging binaries. Such a triple companion causes variations in the GW signal due to (1) the varying path length along the line of sight during the orbit around the center of mass, (2) relativistic beaming, Doppler, and gravitational redshift, and (3) the variation of the "light"-travel time in the gravitational field of the triple companion, known respectively as Roemer-, Einstein-, and Shapiro-delays in pulsar binaries. We find that the prospects for detecting the triple companion are the highest for low-mass compact object binaries which spend the longest time in the LIGO frequency band with circular orbits. In particular, for merging neutron star binaries, LIGO may detect a white dwarf or M-dwarf perturber at signal to noise ratio of 8, if it is within 0.4 solar radius distance from ...

  8. Discovering Steiner Triple Systems through Problem Solving

    Science.gov (United States)

    Sriraman, Bharath

    2004-01-01

    An attempt to implement problem solving as a teacher of ninth grade algebra is described. The problems selected were not general ones, they involved combinations and represented various situations and were more complex which lead to the discovery of Steiner triple systems.

  9. 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.

  10. An experimental investigation of the post-CHF enhancement factor for a prototypical ITER divertor plate with water coolant

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, T.D. [Rensselaer Polytechnic Institute, Troy, NY (United States); Watson, R.D.; McDonald, J.M. [Sandia National Lab., Albuquerque, NM (United States)] [and others

    1995-09-01

    In an off-normal event, water-cooled copper divertor plates in the International Thermonuclear Experimental Reactor (ITER) may either experience heat loads beyond their design basis, or the normal heat loads may be accompanied by low coolant pressure and velocity. The purpose of this experiment was to illustrate that during one-sided heating, as in ITER, a copper divertor plate with the proper side wall thickness, at low system pressure and velocity can absorb without failing an incident heat flux, q{sub i}, that significantly exceed the value, q{sub i}{sup CHF}, which is associated with local CHF at the wall of the coolant channel. The experiment was performed using a 30 kW electron beam test system for heating of a square cross-section divertor heat sink with a smooth circular channel of 7.63 mm diameter. The heated width, length, and wall thickness were 16, 40, and 3 mm, respectively. Stable surface temperatures were observed at incident heat fluxes greater than the local CHF point, presumably due to circumferential conduction around the thick tube walls when q{sub i}{sup CHF} was exceeded. The Post-CHF enhancement factor, {eta}, is defined as the ratio of the incident burnout heat flux, q{sub i}{sup BO}, to q{sub i}{sup CHF}. For this experiment with water at inlet conditions of 70{degrees}C, 1 m/s, and 1 MPa, q{sub i}{sup CHF} and q{sub i}{sup BO} were 600 and 1100 W/cm{sup 2}, respectively, which gave an {eta} of 1.8.

  11. Development of remote pipe cutting tool for divertor cassettes in JT-60SA

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Takao, E-mail: hayashi.takao@jaea.go.jp; Sakurai, Shinji; Shibanuma, Kiyoshi; Sakasai, Akira

    2014-10-15

    Remote pipe cutting tool accessing from inside pipe has been newly developed for JT-60SA. The tool head equips a disk-shaped cutter blade and four rollers which are subjected to the reaction force. The tool pushes out the cutter blade by decreasing the distance between two cams. The tool cuts a cooling pipe by both pushing out the cutter blade and rotating the tool head itself. The roller holder is not pushed out anymore after touching the inner wall of the pipe. In other words, only cutter blade is pushed out after bringing the tool axis into the pipe axis. Outer diameter of the cutting tool head is 44 mm. The cutting tool is able to push out the cutter blade up to 32.5 mm in radius, i.e. 65 mm in diameter, which is enough to cut the pipe having an outer diameter of 59.8 mm. The thickness and material of the cooling pipe are 2.8 mm and SUS316L, respectively. The length of the cutting tool head is about 1 m. The tool is able to cut a pipe locates about 480 mm in depth from the mounting surface on the divertor cassette. The pipe cutting system equips two cutting heads and they are able to cut two pipes at the same time in order to remove the inner target plate. Reproducibility of the cross-sectional shape of the cut pipe is required for re-welding. The degree of reproducibility is inside 0.1 mm except for burr at outside of the pipe, which is enough to re-weld the cut pipe. Some swarf is generated during cutting the double-layered pipe assuming a plug located on the top of the pipe. The swarf is deposited on the bottom of the plug and collected by pulling out the plug in the actual equipment.

  12. Interfacial metallurgy study of brazed joints between tungsten and fusion related materials for divertor design

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Yuxuan, E-mail: yuxuan.zhang@strath.ac.uk [Department of Mechanical and Aerospace Engineering, University of Strathclyde, Glasgow G1 1XJ (United Kingdom); Galloway, Alexander; Wood, James; Robbie, Mikael Brian Olsson; Easton, David [Department of Mechanical and Aerospace Engineering, University of Strathclyde, Glasgow G1 1XJ (United Kingdom); Zhu, Wenzhong [School of Engineering, University of the West of Scotland, Paisley PA1 2BE (United Kingdom)

    2014-11-15

    Highlights: • We created brazed joints between tungsten and EUROFER 97, Cu and SS316L with Au80Cu19Fe1 filler. • No elemental transitions were detected between the W and the AuCuFe filler in either direction. • Transition regions between filler to EUROFER97/316L showed similar elastic modulus and hardness to the filler. • Smooth elemental and mechanical properties transition were detected between the filler and Cu. - Abstract: In the developing DEMO divertor, the design of joints between tungsten to other fusion related materials is a significant challenge as a result of the dissimilar physical metallurgy of the materials to be joined. This paper focuses on the design and fabrication of dissimilar brazed joints between tungsten and fusion relevant materials such as EUROFER 97, oxygen-free high thermal conductivity (OFHC) Cu and SS316L using a gold based brazing foil. The main objectives are to develop acceptable brazing procedures for dissimilar joining of tungsten to other fusion compliant materials and to advance the metallurgical understanding within the interfacial region of the brazed joint. Four different butt-type brazed joints were created and characterised, each of which were joined with the aid of a thin brazing foil (Au80Cu19Fe1, in wt.%). Microstructural characterisation and elemental mapping in the transition region of the joint was undertaken and, thereafter, the results were analysed as was the interfacial diffusion characteristics of each material combination produced. Nano-indentation tests are performed at the joint regions and correlated with element composition information in order to understand the effects of diffused elements on mechanical properties. The experimental procedures of specimen fabrication and material characterisation methods are presented. The results of elemental transitions after brazing are reported. Elastic modulus and nano-hardness of each brazed joints are reported.

  13. Mechanical examination and analysis of W7-X divertor module sub-structures

    Energy Technology Data Exchange (ETDEWEB)

    Smirnow, M., E-mail: michael.smirnow@gmail.com; Boscary, J.; Tittes, H.; Schubert, W.; Peacock, A.

    2015-10-15

    Highlights: • A thermo-structural simulation model of the W7-X target element. • Strain gauge measurements. • Mechanical testing. - Abstract: For the long pulse operation phase, the W7-X stellarator is equipped with an actively water cooled high heat flux (HHF) divertor, consisting of parallel cooled target elements mounted in individual target modules. Due to the thermal deformation of these target elements during heat loading, the pipework that connects the target elements to the water supply manifold is subject to significant forces. Finite element calculations, for target modules TMh1–TMh2, show the superimposed forces of the whole pipework structure on to the manifold resulting in a torsional torque on the manifold support structure and weld. During manufacture, welding of the manifold to its support structure produces thermal induced distortion, resulting in difficulty in maintaining the accuracy of the manifolds. The welding between manifold and support structure was thus minimised in order to reduce this distortion. Finite element calculations showed that the nominal welds were acceptable; however, mechanical stress test on the manifolds mount point was carried out to prove the weld performance under the calculated loading conditions to ensure the safety of the component. For the remaining modules under design TMh1–TMh4 a parametric finite element calculation design study on the effect of the pipe length and routing on the stiffness helped to define minimum requirements for the design. The status of the manifolds for these modules will be shown. The manifolds are also mechanically connected to the port plug-in, therefore the impact of the thermal displacements on this pipework coming from plasma radiation affecting the target elements and from power loads coming from Electron-Cyclotron-Resonance Heating (ECRH) stray field radiation have been calculated. The paper discusses the results of the calculations and presents the outcomes of the stress

  14. Performance of electro-plated and joined components for divertor application

    Energy Technology Data Exchange (ETDEWEB)

    Krauss, Wolfgang, E-mail: wolfgang.krauss@kit.edu; Lorenz, Julia; Konys, Jürgen

    2013-10-15

    Highlights: • Active interlayers of Ni and Pd were electroplated on W to assist joining. • Demonstrator types of W-steel and W–W joints were successfully fabricated. • Diffusion processes increase operation temperature above brazing temperature. • Ni electro-plating is less sensitive to variation of deposition parameters than Pd. • Shear tests showed values in resistance comparable to those of commercial fillers. -- Abstract: A general challenge in divertor development, independently of design type and cooling medium water or helium, is the reliable and adapted joining of components. Depending on the design variants, the characteristics of the joints will be focused on functional or structural behavior to guarantee e.g. good thermal conductivity and sufficient mechanical strength. All variants will have in common that tungsten is the plasma facing material. Thus, material combinations to be joined will range from Cu base over steel to tungsten. Especially tungsten shows lacks in adapted joining due to its metallurgical behavior ranging from immiscibility over bad wetting up to brittle intermetallic phase formation. Joining assisted by electro-chemical deposition of functional and filler layers showed that encouraging progress was achieved in wetting applying nickel interlayers. Nickel proved to be a good reference material but alternative elements (e.g. Pd, Fe) may be more attractive in fusion to manufacture suitable joints. Replacing of Ni as activator element by Pd for W/W or W/steel joints was achieved and joining with Cu-filler was successfully performed. Manufactured joints were characterized applying metallurgical testing and SEM/EDX analyses demonstrating the applicability of Pd activator. First shear tests showed that the joints exhibit mechanical stability sufficient for technical application.

  15. 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  <  10 eV and extending over two orders of magnitude for each correlate: {{T}\\text{et}}   =  ~6.14× {{10}13}n{{D2}t}-0.68 with R 2 = 0.98. The values of T et, and n{{D2}t}{} are for each individual flux tube of the computational 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}}} .

  16. MELCOR Analyses of Divertor Ex-vessel LOCA During Normal Operation. Contract EFDA 01/599, Deliverable 3 - Final Report

    Energy Technology Data Exchange (ETDEWEB)

    ChunHong Sheng

    2002-06-01

    A MELCOR model of ITER-FEAT divertor cooling system has been developed for the analyses of thermal-hydraulic accidents as specified in the Accident Analysis Specifications (AAS-3) for the ITER-FEAT Generic Site Safety Report (GSSR). The model is based on data from the Safety Analysis Data List (SADL-3). The report presents the results of DV ex-vessel LOCA with plasma shutdown from MELCOR calculations. The intention is to verify previous analyses with ATHENA and INTRA to update parts of GSSR documenting the analysis of representative accident sequences for ITER.

  17. A reconstruction theorem for almost-commutative spectral triples

    CERN Document Server

    Ćaćić, Branimir

    2011-01-01

    We propose an expansion of the definition of almost-commutative spectral triple that accommodates non-trivial fibrations and is stable under inner fluctuation of the metric, and then prove a reconstruction theorem for almost-commutative spectral triples under this definition as a simple consequence of the reconstruction theorem for commutative spectral triples. Along the way, we weaken the orientability hypothesis in Connes's reconstruction theorem for commutative spectral triples, and, following Chakraborty and Mathai, prove a number of results concerning the stability of properties of spectral triples under suitable perturbation of the Dirac operator.

  18. Thermal stability of collagen triple helix.

    Science.gov (United States)

    Xu, Yujia

    2009-01-01

    Chief among the challenges of characterizing the thermal stability of the collagen triple helix are the lack of the reversibility of the thermal transition and the presence of multiple folding-unfolding steps during the thermal transition which rarely follows the simple two-state, all-or-none mechanism. Despite of the difficulties inherited in the quantitative depiction of the thermal transition of collagen, biophysical studies combined with proteolysis and mutagenesis approaches using full-chain collagens, short synthetic peptides, and recombinant collagen fragments have revealed molecular features of the thermal unfolding of the subdomains of collagen and led to a better understanding of the diverse biological functions of this versatile protein. The subdomain of collagen generally refers to a segment of the long, rope-like triple helical molecule that can unfold cooperatively as an independent unit whose properties (their size, location, and thermal stability) are considered essential for the molecular recognition during the self-assembly of collagen and during the interactions of collagen with other macromolecules. While the unfolding of segments of the triple helix at temperatures below the apparent melting temperature of the molecule has been used to interpret much of the features of the thermal unfolding of full-chain collagens, the thermal studies of short, synthetic peptides have firmly established the molecular basis of the subdomains by clearly demonstrating the close dependence of the thermal stability of a triple helix on the constituent amino acid residues at the X and the Y positions of the characteristic Gly-X-Y repeating sequence patterns of the triple helix. Studies using recombinant collagen fragments further revealed that in the context of the long, linear molecule, the stability of a segment of the triple helix is also modulated by long-range impact of the local interactions such as the interchain salt bridges. Together, the combined approaches

  19. Real structures on almost-commutative spectral triples

    CERN Document Server

    Ćaćić, Branimir

    2012-01-01

    We refine the reconstruction theorem for almost-commutative spectral triples to a result for real almost-commutative spectral triples, clarifying, in the process, both concrete and abstract definitions of real commutative and almost-commutative spectral triples. In particular, we find that a real almost-commutative spectral triple algebraically encodes the commutative *-algebra of the base manifold in a canonical way, and that a compact oriented Riemannian manifold admits real (almost-)commutative spectral triples of arbitrary KO-dimension. Moreover, we define a notion of smooth family of real finite spectral triples and of the twisting of a concrete real commutative spectral triple by such a family, with interesting KK-theoretic and gauge-theoretic implications.

  20. Mobile Game Probes

    DEFF Research Database (Denmark)

    Borup Lynggaard, Aviaja

    2006-01-01

    This paper will examine how probes can be useful for game designers in the preliminary phases of a design process. The work is based upon a case study concerning pervasive mobile phone games where Mobile Game Probes have emerged from the project. The new probes are aimed towards a specific target...... group and the goal is to specify the probes so they will cover the most relevant areas for our project. The Mobile Game Probes generated many interesting results and new issues occurred, since the probes came to be dynamic and favorable for the process in new ways....

  1. Is there a difference between the effects of single and triple indirect moxibustion stimulations on skin temperature changes of the posterior trunk surface?

    Science.gov (United States)

    Mori, Hidetoshi; Kuge, Hiroshi; Tanaka, Tim Hideaki; Taniwaki, Eiichi; Ohsawa, Hideo

    2011-06-01

    To determine whether any difference exists in responses to indirect moxibustion (IM) relative to thermal stimulation duration. In experiment 1, 9 subjects attended two experimental sessions consisting of single stimulation with IM or triple stimulation with IM, using a crossover design. A K-type thermocouple temperature probe was fixed on the skin surface at the GV14 acupuncture point. IM stimulation was administered to the top of the probe in order to measure the temperature curve. In addition, each subject evaluated his or her subjective feeling of heat on a visual analogue scale after each stimulation. Experiment 2 was conducted on 42 participants, divided into three groups according to the envelope allocation method: single stimulation with IM (n=20), triple stimulation with IM (n=11) and a control group (n=11). A thermograph was used to obtain the skin temperature on the posterior trunk of the participant. To analyse skin temperature, four arbitrary frames (the scapular, interscapular, lumbar and vertebral regions) were made on the posterior trunk. In experiment 1, no significant difference in maximum temperature was found in IM and subjective feeling of heat intensity between single and triple stimulation with IM. In experiment 2, increases in skin temperature occurred on the posterior trunk, but no differences in skin temperature occurred between the groups receiving single and triple stimulation with IM. No difference exists in the skin temperature response to moxibustion between the single and triple stimulation with IM.

  2. On Spectral Triples in Quantum Gravity I

    DEFF Research Database (Denmark)

    Aastrup, Johannes; M. Grimstrup, Jesper; Nest, Ryszard

    2009-01-01

    This paper establishes a link between Noncommutative Geometry and canonical quantum gravity. A semi-finite spectral triple over a space of connections is presented. The triple involves an algebra of holonomy loops and a Dirac type operator which resembles a global functional derivation operator....... The interaction between the Dirac operator and the algebra reproduces the Poisson structure of General Relativity. Moreover, the associated Hilbert space corresponds, up to a discrete symmetry group, to the Hilbert space of diffeomorphism invariant states known from Loop Quantum Gravity. Correspondingly......, the square of the Dirac operator has, in terms of canonical quantum gravity, the form of a global area-squared operator. Furthermore, the spectral action resembles a partition function of Quantum Gravity. The construction is background independent and is based on an inductive system of triangulations...

  3. Secular Evolution of Hierarchical Triple Star Systems

    CERN Document Server

    Ford, E B; Kozinsky, B

    1999-01-01

    We derive octupole-level secular perturbation equations for hierarchical triple systems, using classical Hamiltonian perturbation techniques. Our equations describe the secular evolution of the orbital eccentricities and inclinations over timescales long compared to the orbital periods. By extending previous work done to leading (quadrupole) order to octupole level (i.e., including terms of order $\\alpha^3$, where $\\alpha\\equiv a_1/a_2<1$ is the ratio of semimajor axes) we obtain expressions that are applicable to a much wider range of parameters. For triple systems containing a close inner binary, we also discuss the possible interaction between the classical Newtonian perturbations and the general relativistic precession of the inner orbit. In some cases we show that this interaction can lead to resonances and a significant increase in the maximum amplitude of eccentricity perturbations. We establish the validity of our analytic expressions by providing detailed comparisons with the results of direct num...

  4. Triple A Syndrome: A Case Report

    Directory of Open Access Journals (Sweden)

    Murat Atmaca

    2014-09-01

    Full Text Available Triple A syndrome is a rarely seen autosomal recessive disease characterized by achalasia, adrenal failure and alacrima. The syndrome is frequently seen in childhood. The appearance of its components are usually ordered as alacrima, achalasia and adrenal failure. The majority of the patients diagnosed in the later stages predominantly present with neurological symptoms. In this study, a 21-year-old male who was referred to our clinic with clinical findings of chronic adrenal failure and was diagnosed WİTH triple A syndrome is presented. This patient had been operated three years ago due to achalasia. The diagnosis and treatment of adrenal failure in this syndrome is the most important determinant and indicator in the prognosis of the disease. Turk Jem 2014; 18: 97-99

  5. On Triple-Cut of Scattering Amplitudes

    CERN Document Server

    Mastrolia, Pierpaolo

    2007-01-01

    It is analysed the triple-cut of one-loop amplitudes in dimensional regularisation within spinor-helicity representation. The triple-cut is defined as a difference of two double-cuts with the same particle content, and a same propagator carrying, respectively, causal and anti-causal prescription in each of the two cuts. That turns out into an effective tool for extracting the coefficients of the three-point functions (and higher-point ones) from one-loop-amplitudes. The phase-space integration is oversimplified by using residues theorem to perform the integration over the spinor variables, via the holomorphic anomaly, and a trivial integration on the Feynman parameter. The results are valid for arbitrary values of dimensions.

  6. Potential and limits of water cooled divertor concepts based on monoblock design as possible candidates for a DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Li-Puma, Antonella, E-mail: antonella.lipuma@cea.fr [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Richou, Marianne; Magaud, Philippe; Missirlian, Marc [CEA, IRFM, F-13108 Saint Paul Lez Durance (France); Visca, Eliseo [Associazione EURATOM-ENEA sulla Fusione, C.R. Frascati, IT-00044 Frascati (Italy); Ridolfini, Vincenzo Pericoli [EFDA-CSU Garching, PPPT department, D-85748 Garching bei München (Germany)

    2013-10-15

    In this paper water-cooled divertor concepts based on tungsten monoblock design identified in previous studies as candidate for fusion power plant have been reviewed to assess their potential and limits as possible candidates for a DEMO concept deliverable in a short to medium term (“conservative baseline design”). The rationale and technology development assumptions that have led to their selection are revisited taking into account present factual information on reactor parameters, materials properties and manufacturing technologies. For that purpose, main parameters impacting the divertor design are identified and their relevance discussed. The state of the art knowledge on materials and relevant manufacturing techniques is reviewed. Particular attention is paid to material properties change after irradiation; phenomenon thresholds (if any) and possible operating ranges are identified (in terms of temperature and damage dose). The suitability of various proposed heat sink/structural and sacrificial layer materials, as proposed in the past, are re-assessed (e.g. with regard to the possibility of reducing peak heat flux and/or neutron radiation damages). As a result, potential and limits of various proposed concepts are highlighted, ranges in which they could operate (if any) defined and possible improvements are proposed. Identified missing point in materials database and/or manufacturing techniques knowledge that should be uppermost investigated in future R and D activities are reported. This work has been carried out in the frame of EFDA PPPT Work Programme activities.

  7. 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.

  8. 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.

  9. 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 ...

  10. Metallurgical bonding development of V-4Cr-4Ti alloy for the DIII-D radiative divertor program

    Energy Technology Data Exchange (ETDEWEB)

    Smith, J.P.; Johnson, W.R.; Trester, P.W. [General Atomics, San Diego, CA (United States)

    1998-10-01

    General atomics (GA), in conjunction with the Department of Energy`s (DOE) DIII-D Program, is carrying out a plan to utilize a vanadium alloy in the DIII-D tokamak as part of the DIII-D radiative divertor (RD) upgrade. The V-4Cr-4Ti alloy has been selected in the U.S. as the leading candidate vanadium alloy for fusion applications. This alloy will be used for the divertor fabrication. Manufacturing development with the V-4Cr-4Ti alloy is a focus of the DIII-D RD Program. The RD structure, part of which will be fabricated from V-4Cr-4Ti alloy, will require many product forms and types of metal/metal bonded joints. Metallurgical bonding methods development on this vanadium alloy is therefore a key area of study by GA. Several solid-state (non-fusion weld) and fusion weld joining methods are being investigated. To date, GA has been successful in producing ductile, high-strength, vacuum leak-tight joints by all of the methods under investigation. The solid-state joining was accomplished in air, i.e., without the need for a vacuum or inert gas environment to prevent interstitial impurity contamination of the V-4Cr-4Ti alloy. (orig.) 7 refs.

  11. 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.

  12. Triple Negative Breast Cancer and Metabolic Regulation

    Science.gov (United States)

    2015-08-01

    Lactate Dehydrogenase A is an isoform of lactate dehydrogenase, which catalyzes the conversion of pyruvate to lactate . LDHA is expressed in cancer ...AWARD NUMBER: W81XWH-13-1-0167 TITLE: Triple Negative Breast Cancer and Metabolic Regulation PRINCIPAL INVESTIGATOR: Amy S. Yee, Ph.D...Negative Breast Cancer and Metabolic Regulation 5a. CONTRACT NUMBER 5b. GRANT NUMBER W81XWH-13-1-0167 5c. PROGRAM ELEMENT NUMBER 6. AUTHOR(S) Amy S

  13. Triple plasmon resonance of bimetal nanoshell

    Energy Technology Data Exchange (ETDEWEB)

    Shirzaditabar, Farzad [Department of Physics, Razi University, Kermanshah (Iran, Islamic Republic of); Saliminasab, Maryam, E-mail: m.saliminasab@yahoo.com [Young Researchers Club, Kermanshah Branch, Islamic Azad University, Kermanshah (Iran, Islamic Republic of); Arghavani Nia, Borhan [Department of Physics, Kermanshah Branch, Islamic Azad University, Kermanshah (Iran, Islamic Republic of)

    2014-07-15

    In this paper, light absorption spectra properties of a bimetal multilayer nanoshell based on quasi-static approach are investigated. Comparing with silver-dielectric-silver and silver-dielectric-gold nanoshells, gold-dielectric-silver nanoshells have three intense and separated plasmon peaks which are more suitable for multiplex biosensing. Calculations show that relatively small thickness of outer silver shell and large dielectric constant of middle dielectric layer of gold-dielectric-silver nanoshell are suitable to obtain the triple plasmon resonance.

  14. Triple plasmon resonance of bimetal nanoshell

    Science.gov (United States)

    Shirzaditabar, Farzad; Saliminasab, Maryam; Arghavani Nia, Borhan

    2014-07-01

    In this paper, light absorption spectra properties of a bimetal multilayer nanoshell based on quasi-static approach are investigated. Comparing with silver-dielectric-silver and silver-dielectric-gold nanoshells, gold-dielectric-silver nanoshells have three intense and separated plasmon peaks which are more suitable for multiplex biosensing. Calculations show that relatively small thickness of outer silver shell and large dielectric constant of middle dielectric layer of gold-dielectric-silver nanoshell are suitable to obtain the triple plasmon resonance.