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Sample records for resolved iter divertor

  1. ELM-resolved divertor erosion in the JET ITER-Like Wall

    NASA Astrophysics Data System (ADS)

    Den Harder, N.; Brezinsek, S.; Pütterich, T.; Fedorczak, N.; Matthews, G. F.; Meigs, A.; Stamp, M. F.; van de Sanden, M. C. M.; Van Rooij, G. J.; Contributors, JET

    2016-02-01

    Tungsten erosion in H-mode plasmas is quantified in the outer divertor of the JET ITER-Like Wall environment with optical emission spectroscopy on the 400.9 nm atomic neutral tungsten line. A novel cross-calibration procedure is developed to link slow, high spectral resolution spectroscopy and fast photomultiplier tube measurements in order to obtain ELM-resolved photon fluxes. Inter-ELM W erosion is exclusively impurity sputtering by beryllium because of the high sputter threshold for deuterons. Low beryllium concentrations resulted in low inter-ELM sputter yields of around 10-4 with respect to the total flux. Intra-ELM W sources, which dominate the total W tungsten source, vary independently from the inter-ELM source. The amount of W erosion could only be partly explained by beryllium sputtering, indicating that during ELMs sputtering by fuel species is important. The total W outer divertor source is found to linearly increase with the power crossing the separatrix, whilst excessive divertor fueling can break this trend. The influence of the W source rate on the tungsten content of the core plasma is investigated using soft x-ray emission to determine the tungsten content. At low source rates the content is determined by the source, but at higher source rates, other phenomena determine the total tungsten content. Indications of impurity flushing by ELMs is seen at ELM frequencies above approximately 40 Hz. The inner/outer divertor asymmetry of the W source during ELMs is investigated, and the outer divertor W source is larger by a factor of 1.8+/- 0.7 .

  2. Modeling impurities and tilted plates in the ITER divertor

    SciTech Connect

    Rensink, M.E.; Rognlien, T.D.

    1996-07-29

    The UEDGE 2-D edge transport code is used to model the effect of impurities and tilted divertor plates for the ITER SOL/divertor region. The impurities are modeled as individual charge states using either the FMOMBAL 21-moment description or parallel force balance. Both helium and neon impurities are used together with a majority hydrogenic species. A fluid description of the neutrals is used that includes parallel inertia and neutral-neutral collisions. Effects of geometry are analyzed by using the nonorthogonal mesh capability of UEDGE to obtain solutions with the divertor plate tilted at various angles.

  3. Characteristics of divertor detachment for ITER conditions

    NASA Astrophysics Data System (ADS)

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

    2015-08-01

    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.

  4. Mechanical design issues associated with mounting, maintenance, and handling of an ITER divertor

    SciTech Connect

    Goranson, P.L.; Fogarty, P.J.; Jones, G.H.

    1991-01-01

    Several designs that address plasma-facing plate configurations and thermal-hydraulic design issues have been developed for the ITER divertor. Design criteria growing out of physics requirements, physical constraints, and remote handling requirements impose severe mechanical requirements on the support structure and its attachments. These pose a challenge to the mechanical design of a divertor, which must be addressed before a functional divertor is practical -- that is, one that can be remotely handled, aligned, and maintained; that functions reliably under thermal loading and disruptions; and that gives the required life in the nuclear environment predicted for ITER. This paper discusses the design criteria for the divertor mounting structure and identifies the mechanical design issues that need to be addressed. Achieving the criteria may require the development of new components and innovative configurations, specifically a new class of remote fasteners and electrically resistant material for mounts. The possible design of such components and an R D program to develop them are described, and issues specific to the high-aspect-ratio design (HARD) configuration are summarized. Analysis and experiments that will resolve these issues and concerns and lead to a final ITER design are identified. 2 refs., 2 figs.

  5. Is Carbon a Realistic Choice for ITER's Divertor?

    SciTech Connect

    C.H. Skinner; G. Federici

    2005-05-13

    Tritium retention by co-deposition with carbon on the divertor target plate is predicted to limit ITER's DT burning plasma operations (e.g. to about 100 pulses for the worst conditions) before the in-vessel tritium inventory limit, currently set at 350 g, is reached. At this point, ITER will only be able to continue its burning plasma program if technology is available that is capable of rapidly removing large quantities of tritium from the vessel with over 90% efficiency. The removal rate required is four orders of magnitude faster than that demonstrated in current tokamaks. Eighteen years after the observation of co-deposition on JET and TFTR, such technology is nowhere in sight. The inexorable conclusion is that either a major initiative in tritium removal should be funded or that research priorities for ITER should focus on metal alternatives.

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

    NASA Astrophysics Data System (ADS)

    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

  7. Optimization of tungsten castellated structures for the ITER divertor

    NASA Astrophysics Data System (ADS)

    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.

  8. Plasma flow interaction with ITER divertor related surfaces

    NASA Astrophysics Data System (ADS)

    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.

  9. Plasma Flow Interaction With Iter Divertor Related Surfaces

    NASA Astrophysics Data System (ADS)

    Dojcinovic, I. P.

    2010-07-01

    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 (Arkhipov et al. 2000, Federici et al. 2005). 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 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/m^2 in 0.1 ms) can be used for simulation of transient peak thermal loads during Type I ELMs and disruptions (Dojcinovic et al. 2007). Analysis of the targets erosion, brittle destruction, melting processes, and dust formation has been performed (Dojcinovic et al. 2006). 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.

  10. Facilities for technology testing of ITER divertor concepts, models, and prototypes in a plasma environment

    SciTech Connect

    Cohen, S.A.

    1991-12-01

    The exhaust of power and fusion-reaction products from ITER plasma are critical physics and technology issues from performance, safety, and reliability perspectives. Because of inadequate pulse length, fluence, flux, scrape-off layer plasma temperature and density, and other parameters, the present generation of tokamaks, linear plasma devices, or energetic beam facilities are unable to perform adequate technology testing of divertor components, though they are essential contributors to many physics issues such as edge-plasma transport and disruption effects and control. This Technical Requirements Documents presents a description of the capabilities and parameters divertor test facilities should have to perform accelerated life testing on predominantly technological divertor issues such as basic divertor concepts, heat load limits, thermal fatigue, tritium inventory and erosion/redeposition. The cost effectiveness of such divertor technology testing is also discussed.

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

    SciTech Connect

    Klepper, C Christopher; Biewer, T. M.; Graves, Van B; Andrew, P.; Marcus, Chris; Shimada, M.; Hughes, S.; Boussier, B.; Johnson, D. W.; Gardner, W. L.; Hillis, D. L.; Vayakis, G.; Vayakis, G.; Walsh, M.

    2015-01-01

    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 (H2, D2, T2). Measurement of the concentration of radiative gases, such as neon (Ne) and nitrogen (N2), is also intended. Numerical modeling of the gas flow from the sampled region to the cluster of analysis sensors, through a long (~8m long, ~110mm diameter) sampling pipe terminating in a pressure reducing orifice, confirm that the desired response time (~1s for He or D2) is achieved with the present design.

  12. Spectral resolvability of iterated rippled noise

    NASA Astrophysics Data System (ADS)

    Yost, William A.

    2005-04-01

    A forward-masking experiment was used to estimate the spectral ripple of iterated rippled noise (IRN) that is possibly resolved by the auditory system. Tonal signals were placed at spectral peaks and valleys of IRN maskers for a wide variety of IRN conditions that included different delays, number of iterations, and stimulus durations. The differences in the forward-masked thresholds of tones at spectral peaks and valleys were used to estimate spectral resolvability, and these results were compared to estimates obtained from a gamma-tone filter bank. The IRN spectrum has spectral peaks that are harmonics of the reciprocal of the delay used to generate IRN stimuli. As the number of iterations in the generation of IRN stimuli increases so does the difference in the spectral peak-to-valley ratio. For high number of iterations, long delays, and long durations evidence for spectral resolvability existed up to the 6th harmonic. For all other conditions spectral resolvability appeared to disappear at harmonics lower than the 6th, or was not measurable at all. These data will be discussed in terms of the role spectral resolvability might play in processing the pitch, pitch strength, and timbre of IRN stimuli. [Work supported by a grant from NIDCD.

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

    NASA Astrophysics Data System (ADS)

    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.

  14. Spectroscopic Measurement System for ITER Divertor Plasma: Impurity Influx Monitor (divertor)

    SciTech Connect

    Sugie, Tatsuo; Ogawa, Hiroaki; Kusama, Yoshinori; Kasai, Satoshi

    2008-03-12

    The detailed design of the Impurity Influx Monitor (divertor) has been carried out to provide the measurement capability in the harsh environment such as higher irradiation levels of neutron, gamma-ray and particles than in present devices. The in-situ calibration system using a micro retro-reflector array has been developed to monitor the sensitivity change of the optical system due to the environmental effects. The optical alignment system for the Monitor has been developed by using a dedicated optics for alignment in the collection optics for measurement.

  15. Estimation of carbon fibre composites as ITER divertor armour

    NASA Astrophysics Data System (ADS)

    Pestchanyi, S.; Safronov, V.; Landman, I.

    2004-08-01

    Exposure of the carbon fibre composites (CFC) NB31 and NS31 by multiple plasma pulses has been performed at the plasma guns MK-200UG and QSPA. Numerical simulation for the same CFCs under ITER type I ELM typical heat load has been carried out using the code PEGASUS-3D. Comparative analysis of the numerical and experimental results allowed understanding the erosion mechanism of CFC based on the simulation results. A modification of CFC structure has been proposed in order to decrease the armour erosion rate.

  16. Modelling of passive spectroscopy in the ITER divertor: the first hydrogen Balmer lines

    NASA Astrophysics Data System (ADS)

    Rosato, J.; Kotov, V.; Reiter, D.

    2010-07-01

    The first lines of the hydrogen Balmer series are investigated in ITER divertor conditions using a line shape code and a plasma edge transport code. It is shown that most of the emissivity originates from a localized, cold and dense region close to the divertor target plates, where the plasma is in the recombining regime. We simulate the signal obtained by pointing a spectrometer at this zone. The physical processes which contribute to the spectral line formation are examined, with a special emphasis on the Stark effect, photon absorption and stimulated emission. It is shown that, even though the Stark effect is significant, local information on the Doppler atomic temperature can be obtained from a fitting analysis of the Dα spectral line shape.

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

    SciTech Connect

    Takeuchi, M.; Sugie, T.; Ogawa, H.; Takeyama, S.; Itami, K.

    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.

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

    SciTech Connect

    Balboa, I.; Arnoux, G.; Kinna, D.; Thomas, P. D.; Morlock, C.; Kruezi, U.; Sergienko, G.; Rack, M.; Collaboration: JET EFDA Contributors

    2012-10-15

    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 {mu}m and up to sampling frequencies of {approx}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.

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

    NASA Astrophysics Data System (ADS)

    Takeuchi, M.; Sugie, T.; Ogawa, H.; Takeyama, S.; Itami, K.

    2014-08-01

    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.

  20. Assessment of erosion and surface tritium inventory issues for the ITER divertor

    SciTech Connect

    Brooks, J.N.; Causey, R.; Federici, G.; Ruzic, D.N.

    1996-08-01

    The authors analyzed sputtering erosion and tritium codeposition for the ITER vertical target divertor design using erosion and plasma codes (WBC/REDEP/DEGAS+) coupled to available materials data. Computations were made for a beryllium, carbon, and tungsten coated divertor plate, and for three edged plasma regimes. New data on tritium codeposition in beryllium was obtained with the TPE facility. This shows codeposited H/Be ratios of the order of 10% for surface temperatures {le} 300 C, beryllium thereby being similar to carbon in this respect. Hydrocarbon transport calculations show significant loss (10--20%) of chemically sputtered carbon for detached conditions (T{sub e} {approx} 1 eV at the divertor), compared to essentially no loss (100% redeposition) for higher temperature plasmas. Calculations also show a high, non-thermal, D-T molecular flux for detached conditions. Tritium codeposition rates for carbon are very high for detached conditions ({approximately} 20g-T/1000 s discharge), due to buildup of chemically sputtered carbon on relatively cold surfaces of the divertor cassette. Codeposition is lower ({approximately} 10X) for higher edge temperatures ({approximately} 8--30 eV) and is primarily due to divertor plate buildup of physically sputtered carbon. Peak net erosion rates for carbon are of order 30 cm/burn-yr. Erosion and codeposition rates for beryllium are much lower than for carbon at detached conditions, but are similar to carbon for the higher temperatures. Both erosion and tritium codeposition are essentially nil for tungsten for the regimes studied.

  1. The ITER divertor Thomson scattering system: engineering and advanced hardware solutions

    NASA Astrophysics Data System (ADS)

    Mukhin, E. E.; Semenov, V. V.; Razdobarin, A. G.; Tolstyakov, S. Yu; Kochergin, M. M.; Kurskiev, G. S.; Berezutsky, A. A.; Podushnikova, K. A.; Masyukevich, S. V.; Chernakov, P. V.; Borovkov, A. I.; Modestov, V. S.; Nemov, A. S.; Voinov, A. S.; Kornev, A. F.; Stupnikov, V. K.; Borisov, A. A.; Baranov, G. N.; Koval, A. N.; Makushina, A. F.; Yelizarov, B. A.; Kukushkin, A. S.; Encheva, A.; Andrew, P.

    2012-02-01

    A divertor Thomson scattering (TS) system being developed for ITER has incorporated proven solutions from currently available TS systems. On the other hand any ITER diagnostic has to operate in a hostile environment and very restricted access geometry. Therefore the operation in an environment of intensive stray light, plasma background radiation, the necessity meet the requirement using only a 20 mm gap between divertor cassettes for plasma diagnosis as well as to measure plasma temperatures as low as 1 eV severely constrain the divertor TS diagnostic design. The challenging solutions of this novel diagnostic system which has to ensure its steady performance and also the operability and maintenance are the focus of this report. One of the most demanding parts of the in-vessel diagnostic equipment development is the design assessment using different engineering analyses. The task definition and first results of thermal, e/m and seismic analyses are provided. The process of further improving of the design involves identification of susceptible areas and multiple iterations of the design, as needed. One of the key points for all Thomson scattering diagnostics are the laser capabilities. A high-performance and high-power laser system using a steady-state and high-repetitive mode Nd:YAG laser (2J, 50-100Hz, 3ns) has been developed. The reduced laser pulse duration matched with high-speed low-noise APD detector can be very important under high background light level. For diagnostics such as Thomson scattering and Raman spectroscopy, a high-degree of discrimination against stray light at the laser wavelength is required for successful detection of wavelength-shifted light from the laser-plasma interaction region. For this case of high stray light level, a triple grating polychromator characterized by high rejection and high transmission has been designed and developed. The novel polychromator design minimizes stray light while still maintaining a relatively high

  2. Hydrogen embrittlement considerations in niobium-base alloys for application in the ITER divertor

    SciTech Connect

    Peterson, D.T. ); Hull, A.B.; Loomis, B.A. )

    1991-01-01

    The ITER divertor will be subjected to hydrogen from aqueous corrosion by the coolant and by transfer from the plasma. Global hydrogen concentrations are one factor in assessing hydrogen embrittlement but local concentrations affected by source fluxes and thermotransport in thermal gradients are more important considerations. Global hydrogen concentrations is some corrosion- tested alloys will be presented and interpreted. The degradation of mechanical properties of Nb-base alloys due to hydrogen is a complex function of temperature, hydrogen concentration, stresses and alloy composition. The known tendencies for embrittlement and hydride formation in Nb alloys are reviewed.

  3. Hydrogen embrittlement considerations in niobium-base alloys for application in the ITER divertor

    SciTech Connect

    Peterson, D.T.; Hull, A.B.; Loomis, B.A.

    1991-12-31

    The ITER divertor will be subjected to hydrogen from aqueous corrosion by the coolant and by transfer from the plasma. Global hydrogen concentrations are one factor in assessing hydrogen embrittlement but local concentrations affected by source fluxes and thermotransport in thermal gradients are more important considerations. Global hydrogen concentrations is some corrosion- tested alloys will be presented and interpreted. The degradation of mechanical properties of Nb-base alloys due to hydrogen is a complex function of temperature, hydrogen concentration, stresses and alloy composition. The known tendencies for embrittlement and hydride formation in Nb alloys are reviewed.

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

    PubMed

    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. PMID:27587120

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

    NASA Astrophysics Data System (ADS)

    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.

  6. Tungsten erosion in the baffle and outboard regions of the ITER-like ASDEX Upgrade divertor

    NASA Astrophysics Data System (ADS)

    Maier, H.; ASDEX Upgrade Team

    2004-12-01

    Similar to the design of the next-step device ITER, ASDEX Upgrade is equipped with vertical divertor targets with adjacent baffles extending towards the main chamber. In ITER, it is intended to employ tungsten as a plasma-facing material in this baffle area. Tungsten-coated graphite tiles were installed in the divertor baffle and the outboard side regions of ASDEX Upgrade for a full experimental campaign. The erosion behavior of tungsten was investigated by scanning electron microscopy and by measuring the thickness of the tungsten coatings before and after exposure. The coatings had an initial thickness of approximately 450 nm. Two distinct erosion mechanisms were observed: in the outer baffle region a reduction of the coatings' thickness up to 100 nm was determined after about 6300 s of plasma discharge. On the roof baffle and on the inner baffle modules, no clear reduction of the film thickness was found. In the tracks of arcs, however, the tungsten coatings were completely removed. This represents an erosion of 5-10% of the tungsten-coated surface area in this region.

  7. Deuterium trapping and release in JET ITER-like wall divertor tiles

    NASA Astrophysics Data System (ADS)

    Likonen, J.; Heinola, K.; De Backer, A.; Koivuranta, S.; Hakola, A.; Ayres, C. F.; Baron-Wiechec, A.; Coad, P.; Matthews, G. F.; Mayer, M.; Widdowson, A.; Contributors, JET

    2016-02-01

    A selected set of samples from JET-ILW divertor tiles exposed in 2011-2012 has been analysed using thermal desorption spectrometry (TDS). The highest amount of deuterium was found on the regions with the thickest deposited layers, i.e. on the horizontal (apron) part and on the top part of Tile 1, which resides deep in the scrape-off layer. Outer divertor Tiles 6, 7 and 8 had nearly an order of magnitude less deuterium. The co-deposited layers on the JET tiles and the W coatings contain C, O and Ni impurities which may change the desorption properties. The D2 signals in the TDS spectra were convoluted and the positions of the peaks were compared with the Be and C amounts but no correlations between them were found. The remaining fractions of D in the analysed samples at ITER baking temperature 350 °C are rather high implying that co-deposited films may be difficult to be de-tritiated.

  8. Quantification of Chemical Erosion in the DIII-D Divertor and Implications for ITER

    SciTech Connect

    McLean, A. G.; Stangeby, P. C.; Bray, B. D.; Brezinsek, S.; Brooks, N. H.; Davis, J. W.; Isler, R. C.; Kirschner, A.; Laengner, R.; Lasnier, C. J.; Mu, Y.; Munoz-Burgos, J. M.; Rudakov, D. L.; Schmitz, O.; Unterberg, Ezekial A; Watkins, J. G.; Whyte, D. G.; Wong, C. P. C.

    2011-01-01

    The Porous Plug Injector (PPI) has proven to be an invaluable diagnostic for in situ characterization and quantification of erosion phenomena in DIII-D. Previous work has led to derivation of three primary figures of merit for chemical erosion (CE) in attached and cold divertor conditions: relative intensity of C+ chemical and physical sources, the CE yield (Y-chem) and effective photon efficiencies for chemically eroded products. Application of these figures for accounting of observed absolutely calibrated CI and CII emission intensities is demonstrated to produce a self-consistent solution at the DIII-D targets. Reinterpretation of the CI (C degrees) spectral lineshape profile supports the relative roles of local chemical versus physical sputtering as previously determined for CII (C+). Comparison of calculated in situ Y-chem to that measured ex situ suggests a tokamak-specific lower energy threshold for CE and has potentially major implications for prediction of tritium co-deposition near the divertor targets in ITER.

  9. A Fast Exhaust-Gas Analyzer for the ITER Fusion Experiment Divertor

    SciTech Connect

    Klepper, C Christopher; Carlson, E. P.; Moschella, J. J.; Hazelton, R C; Keitz, M D; Gardner, Walter L

    2010-01-01

    This paper presents a first demonstration of a radio-frequency (RF)-excited optical gas analyzer (RF-OGA) designed to quantitatively measure minority species inside the neutralization region of the ITER fusion experiment divertor. The sensor head, which creates its own plasma excitation and plasma light emission, is designed to operate in a strong magnetic field, and the RF coupling leads to bright light emission. It also allows for operation at low voltages, avoiding the radiation-enhanced breakdowns expected when high voltages are present in the ITER environment. Furthermore, the preferred sensor head features full isolation of the metal RF electrodes from the induced plasma. This "electrodeless" operation will permit long operation without frequent maintenance. The testing of a first experimental RF-OGA with an electrodeless design in a strong (similar to 2-T) magnetic field showed a mostly linear response of the He I-6678 angstrom line emission to the He concentration in a hydrogen background, which would produce a He concentration measurement accurate to within 2% of the helium-to-hydrogen ratio.

  10. Solid tungsten Divertor-III for ASDEX Upgrade and contributions to ITER

    NASA Astrophysics Data System (ADS)

    Herrmann, A.; Greuner, H.; Jaksic, N.; Balden, M.; Kallenbach, A.; Krieger, K.; de Marné, P.; Rohde, V.; Scarabosio, A.; Schall, G.; the ASDEX Upgrade Team

    2015-06-01

    ASDEX Upgrade became a full tungsten experiment in 2007 by coating its graphite plasma facing components with tungsten. In 2013 a redesigned solid tungsten divertor, Div-III, was installed and came into operation in 2014. The redesign of the outer divertor geometry provided the opportunity to increase the pumping efficiency in the lower divertor by increasing the gap between divertor and vessel. In parallel, a by-pass was installed into the cryo-pump in the divertor region allowing adapting of the pumping speed to the required edge density. Safe divertor operation and heat removal becomes more and more significant for future fusion devices. This requires developing ‘tools’ for divertor heat load control and to optimize the divertor design. The new divertor manipulator, DIM-II, allows retracting a relevant part of the outer divertor into a target exchange box without venting ASDEX Upgrade. Different front-ends can be installed and exposed to the plasma. At present, front-ends for probe exposition, gas puffing, electrical probes and actively cooled prototype targets are under construction. The installation of solid tungsten, the control of the pumping speed and the flexibility for testing divertor modifications on a weekly base is a unique feature of ASDEX Upgrade and offers together with the extended set of diagnostics the possibility to investigate dedicated questions for a future divertor design.

  11. Shifting the CFC/W transition point on the first ITER divertor target plates: equilibrium considerations

    NASA Astrophysics Data System (ADS)

    Kolesnikov, R. A.; Bulmer, R. H.; Lodestro, L. L.; Casper, T. A.; Pitts, R. A.

    2012-03-01

    In the 2007 ITER Design Review, the CFC/W transition point on the first divertor target plates was lowered by 10 cm to allow some experience to be gained in the non-active phases of vertical target operation with strike points on W surfaces, in preparation for a full W divertor in the nuclear phase. For operation on W just above the transition point, we use the CORSICA code to investigate the range of possible H- and L-mode equilibria, with emphasis on the maximum plasma current, achievable shapes, etc. We then investigate the operational space as the transition is lowered still further (both L- and H-mode), while still ensuring sufficient carbon vertical target extent to fulfill the requirements of the non-active phase program. The primary aim of this study is to determine if the current transition point, which can still be modified within some range if required, is optimized with respect to gaining early operational experience on an all-metal target before the nuclear phases begin. In our previous work [1], we investigated the size of feasible βp--li space for both reference and elevated strikes for operation at 14 MA (both L- and H-mode) as well as 12 MA (H-mode) currents. In this paper we present new results on the maximum achievable plasma current as a function of strikes locations. Also, we study plasma self-inductance, volt-second consumption and the vertical instability over the range of the new equilibria. [1] R.A. Kolesnikov, et al., 53^rd APS/DPP, Salt Lake City (2011)

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

    NASA Astrophysics Data System (ADS)

    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

  13. Equilibrium and vertical-instability considerations for vertical strike-point shifts on the ITER divertor targets

    NASA Astrophysics Data System (ADS)

    Kolesnikov, R. A.; Bulmer, R. H.; LoDestro, L. L.; Casper, T. A.; Pitts, R. A.

    2013-08-01

    The study of operation with raised strike points on the first ITER divertor target plates is motivated by the need to gain experience with operation with strike points on tungsten (W) surfaces during the non-active phases (in the case of an initial carbon fibre composite (CFC)/W divertor); or (if ITER begins with a full-W divertor), to gain experience with plasma control and transients while operating with raised strike points to avoid damaging the baseline strike regions in preparation for the nuclear phase, and to provide a means for operation should damage occur in the baseline strike zone. For operation with raised strike points, we use the Corsica code to investigate the range of possible H- and L-mode equilibria, with emphasis on the maximum plasma current, achievable shapes, etc. With raised strike points the maximum achievable plasma current is close to 14 MA. The operating space (βp - li) for raised strike points has been studied. The size of the βp - li operating space shrinks (compared to using standard strike-point positions) at 14 MA. For 12 MA, however, the operating space is not affected when using raised strike points. For equilibria with elevated strike points (at roughly the CFC/W transitions, following the 2007 ITER Design Review) the vertical-instability growth-rates at high plasma current (14 MA) are somewhat high but are within the 20 s-1 which studies indicate are controllable in ITER. At lower currents (12 MA) in H-mode, the vertical-instability growth rates stay below 10.0 s-1 for most of βp - li space. At 12 MA in H-mode, multiple equilibria which meet our constraints have been found in overlapping regions of the βp - li operating space.

  14. Divertor detachment

    NASA Astrophysics Data System (ADS)

    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.

  15. An experimental investigation of the post-CHF enhancement factor for a prototypical ITER divertor plate with water coolant

    SciTech Connect

    Marshall, T.D.; Watson, R.D.; McDonald, J.M.

    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.

  16. Development of a mirror-based endoscope for divertor spectroscopy on JET with the new ITER-like wall (invited)

    SciTech Connect

    Huber, A.; Brezinsek, S.; Mertens, Ph.; Schweer, B.; Sergienko, G.; Terra, A.; Clever, M.; Lambertz, H. T.; Samm, U.; Arnoux, G.; Balshaw, N.; Edlingdon, T.; Farthing, J.; Matthews, G. F.; Riccardo, V.; Sanders, S.; Stamp, M.; Williams, J.; Zastrow, K. D.; and others

    2012-10-15

    A new endoscope with optimised divertor view has been developed in order to survey and monitor the emission of specific impurities such as tungsten and the remaining carbon as well as beryllium in the tungsten divertor of JET after the implementation of the ITER-like wall in 2011. The endoscope is a prototype for testing an ITER relevant design concept based on reflective optics only. It may be subject to high neutron fluxes as expected in ITER. The operating wavelength range, from 390 nm to 2500 nm, allows the measurements of the emission of all expected impurities (W I, Be II, C I, C II, C III) with high optical transmittance ({>=}30% in the designed wavelength range) as well as high spatial resolution that is {<=}2 mm at the object plane and {<=}3 mm for the full depth of field ({+-}0.7 m). The new optical design includes options for in situ calibration of the endoscope transmittance during the experimental campaign, which allows the continuous tracing of possible transmittance degradation with time due to impurity deposition and erosion by fast neutral particles. In parallel to the new optical design, a new type of possibly ITER relevant shutter system based on pneumatic techniques has been developed and integrated into the endoscope head. The endoscope is equipped with four digital CCD cameras, each combined with two filter wheels for narrow band interference and neutral density filters. Additionally, two protection cameras in the {lambda} > 0.95 {mu}m range have been integrated in the optical design for the real time wall protection during the plasma operation of JET.

  17. Development of a mirror-based endoscope for divertor spectroscopy on JET with the new ITER-like wall (invited).

    PubMed

    Huber, A; Brezinsek, S; Mertens, Ph; Schweer, B; Sergienko, G; Terra, A; Arnoux, G; Balshaw, N; Clever, M; Edlingdon, T; Egner, S; Farthing, J; Hartl, M; Horton, L; Kampf, D; Klammer, J; Lambertz, H T; Matthews, G F; Morlock, C; Murari, A; Reindl, M; Riccardo, V; Samm, U; Sanders, S; Stamp, M; Williams, J; Zastrow, K D; Zauner, C

    2012-10-01

    A new endoscope with optimised divertor view has been developed in order to survey and monitor the emission of specific impurities such as tungsten and the remaining carbon as well as beryllium in the tungsten divertor of JET after the implementation of the ITER-like wall in 2011. The endoscope is a prototype for testing an ITER relevant design concept based on reflective optics only. It may be subject to high neutron fluxes as expected in ITER. The operating wavelength range, from 390 nm to 2500 nm, allows the measurements of the emission of all expected impurities (W I, Be II, C I, C II, C III) with high optical transmittance (≥ 30% in the designed wavelength range) as well as high spatial resolution that is ≤ 2 mm at the object plane and ≤ 3 mm for the full depth of field (± 0.7 m). The new optical design includes options for in situ calibration of the endoscope transmittance during the experimental campaign, which allows the continuous tracing of possible transmittance degradation with time due to impurity deposition and erosion by fast neutral particles. In parallel to the new optical design, a new type of possibly ITER relevant shutter system based on pneumatic techniques has been developed and integrated into the endoscope head. The endoscope is equipped with four digital CCD cameras, each combined with two filter wheels for narrow band interference and neutral density filters. Additionally, two protection cameras in the λ > 0.95 μm range have been integrated in the optical design for the real time wall protection during the plasma operation of JET. PMID:23130790

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

    NASA Astrophysics Data System (ADS)

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

    2015-08-01

    The impact of the divertor geometry on global plasma confinement in type I ELMy H-mode has been investigated in the JET tokamak equipped with ITER-Like Wall. Discharges have been performed in which the position of the strike-points was changed while keeping the bulk plasma equilibrium essentially unchanged. Large variations of the global plasma confinement have been observed, the H98 factor changing from typically 0.7 when the outer strike-point is on the vertical or horizontal targets to 0.9 when it is located in the pump duct entrance. Profiles are mainly impacted in the pedestal but core gradient lengths, especially for the density, are also modified. Although substantial differences are observed in the divertor conditions, none seem to correlate directly with the confinement. Modelling with the EDGE2D-EIRENE and SOLEDGE2D-EIRENE transport codes exhibits differences in the energy losses due to neutrals inside the separatrix, but orders of magnitude are too low to explain simply the impact on the confinement.

  19. Simulation experiment of interaction of plasma facing materials and transient heat loads in ITER divertor by use of magnetized coaxial plasma gun

    NASA Astrophysics Data System (ADS)

    Nakatsuka, M.; Ando, K.; Higashi, T.; Kikuchi, Y.; Fukumoto, N.; Nagata, M.

    2009-11-01

    Interaction of plasma facing materials and transient head loads such as type I ELMs is one of the critical issues in ITER divertor. The heat load to the ITER divertor during type I ELMs is estimated to be 0.5-3 MJ/m^2 with a pulse length of 0.1-0.5 ms. We have developed a magnetized coaxial plasma gun (MCPG) for the simulation experiment of transient heat load during type I ELMs in ITER divertor. The MCPG has inner and outer electrodes made of stainless steel 304. In addition, the inner electrode is covered with molybdenum so as to suppress the release of impurities from the electrode during the discharge. The diameters of inner and outer electrodes are 0.06 m and 0.14 m, respectively. The power supply for the MCPG is a capacitor bank (7 kV, 1 mF, 25 kJ). The plasma velocity estimated by the time of flight measurement of the magnetic fields was about 50 km/s, corresponding to the ion energy of 15 eV (H) or 30 eV (D). The absorbed energy density of the plasma stream was measured a calorimeter made of graphite. It was found that the absorbed energy density was 0.9 MJ/m^2 with a pulse width of 0.5 ms at the distance of 100 mm from the inner electrode. In the conference, experimental results of plasma exposure on the plasma facing materials in ITER divertor will be shown.

  20. Modeling of divertor particle and heat loads during application of resonant magnetic perturbation fields for ELM control in ITER

    NASA Astrophysics Data System (ADS)

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

    2013-07-01

    First results from three-dimensional modeling of the divertor heat and particle flux pattern during application of resonant magnetic perturbation fields as ELM control scheme in ITER with the EMC3-Eirene fluid plasma and kinetic neutral transport code are discussed. The formation of a helical magnetic footprint breaks the toroidal symmetry of the heat and particle fluxes. Expansion of the flux pattern as far as 60 cm away from the unperturbed strike line is seen with vacuum RMP fields, resulting in a preferable heat flux spreading. Inclusion of plasma response reduces the radial extension of the heat and particle fluxes and results in a heat flux peaking closer to the unperturbed level. A strong reduction of the particle confinement is found. 3D flow channels are identified as a consistent reason due to direct parallel outflow from inside of the separatrix. Their radial inward expansion and hence the level of particle pump out is shown to be dependent on the perturbation level.

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

    SciTech Connect

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

    2011-04-05

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

  2. Resolving the iterated prisoner's dilemma: theory and reality.

    PubMed

    Raihani, N J; Bshary, R

    2011-08-01

    Pairs of unrelated individuals face a prisoner's dilemma if cooperation is the best mutual outcome, but each player does best to defect regardless of his partner's behaviour. Although mutual defection is the only evolutionarily stable strategy in one-shot games, cooperative solutions based on reciprocity can emerge in iterated games. Among the most prominent theoretical solutions are the so-called bookkeeping strategies, such as tit-for-tat, where individuals copy their partner's behaviour in the previous round. However, the lack of empirical data conforming to predicted strategies has prompted the suggestion that the iterated prisoner's dilemma (IPD) is neither a useful nor realistic basis for investigating cooperation. Here, we discuss several recent studies where authors have used the IPD framework to interpret their data. We evaluate the validity of their approach and highlight the diversity of proposed solutions. Strategies based on precise accounting are relatively uncommon, perhaps because the full set of assumptions of the IPD model are rarely satisfied. Instead, animals use a diverse array of strategies that apparently promote cooperation, despite the temptation to cheat. These include both positive and negative reciprocity, as well as long-term mutual investments based on 'friendships'. Although there are various gaps in these studies that remain to be filled, we argue that in most cases, individuals could theoretically benefit from cheating and that cooperation cannot therefore be explained with the concept of positive pseudo-reciprocity. We suggest that by incorporating empirical data into the theoretical framework, we may gain fundamental new insights into the evolution of mutual reciprocal investment in nature. PMID:21599777

  3. An experimental examination of the loss-of-flow accident phenomenon for prototypical ITER divertor channels of Y=0 and Y=2

    SciTech Connect

    Marshall, T.D.; McDonald, J.M.; Cadwallader, L.C.; Steiner, D.

    2000-01-01

    This paper discusses the thermal response of two prototypical International Thermonuclear Experimental Reactor (ITER) divertor channels during simulated loss-of-flow-accident (LOFA) experiments. The thermal response was characterized by the time-to-burnout (TBO), which is a figure of merit on the mockups' survivability. Data from the LOFA experiments illustrate that (a) the pre-LOFA inlet velocity does not significantly influence the TBO, (b) the incident heat flux (IHF) does influence the TBO, and (c) a swirl tape insert significantly improves the TBO and promotes the initiation of natural circulation. This natural circulation enabled the mockup to absorb steady-state IHFs after the coolant circulation pump was disabled. Several methodologies for thermal-hydraulic modeling of the LOFA were attempted.

  4. An Experimental Examination of the Loss-of-Flow Accident Phenomenon for Prototypical ITER Divertor Channels of Y = 0 and Y = 2

    SciTech Connect

    Marshall, Theron D.; McDonald, Jimmie M.; Cadwallader, Lee C.; Steiner, Don

    2000-01-15

    This paper discusses the thermal response of two prototypical International Thermonuclear Experimental Reactor (ITER) divertor channels during simulated loss-of-flow-accident (LOFA) experiments. The thermal response was characterized by the time-to-burnout (TBO), which is a figure of merit on the mockups' survivability. Data from the LOFA experiments illustrate that (a) the pre-LOFA inlet velocity does not significantly influence the TBO, (b) the incident heat flux (IHF) does influence the TBO, and (c) a swirl tape insert significantly improves the TBO and promotes the initiation of natural circulation. This natural circulation enabled the mockup to absorb steady-state IHFs after the coolant circulation pump was disabled. Several methodologies for thermal-hydraulic modeling of the LOFA were attempted.

  5. Material deposition on inner divertor quartz-micro balances during ITER-like wall operation in JET

    NASA Astrophysics Data System (ADS)

    Esser, H. G.; Philipps, V.; Freisinger, M.; Widdowson, A.; Heinola, K.; Kirschner, A.; Möller, S.; Petersson, P.; Brezinsek, S.; Huber, A.; Matthews, G. F.; Rubel, M.; Sergienko, G.

    2015-08-01

    The migration of beryllium, tungsten and carbon to remote areas of the inner JET-ILW divertor and the accompanying co-deposition of deuterium has been investigated using post-mortem analysis of the housings of quartz-micro balances (QMBs) and their quartz crystals. The analysis of the deposition provides that the rate of beryllium atoms is significantly reduced compared to the analogue deposition rate of carbon during the carbon wall conditions (JET-C) at the same locations of the QMBs. A reduction factor of 50 was found at the entrance gap to the cryo-pumps while it was 14 under tile 5, the semi-horizontal target plate. The deposits consist of C/Be atomic ratios of typically 0.1-0.5 showing an enrichment of carbon in remote areas compared to directly exposed areas with less carbon. The deuterium retention fraction D/Be is between 0.3 and 1 at these unheated locations in the divertor.

  6. Deposition in the inner and outer corners of the JET divertor with carbon wall and metallic ITER-like wall

    NASA Astrophysics Data System (ADS)

    Beal, J.; Widdowson, A.; Heinola, K.; Baron-Wiechec, A.; Gibson, K. J.; Coad, J. P.; Alves, E.; Lipschultz, B.; Kirschner, A.; Esser, H. G.; Matthews, G. F.; Brezinsek, S.; Contributors, JET

    2016-02-01

    Rotating collectors and quartz microbalances (QMBs) are used in JET to provide time-dependent measurements of erosion and deposition. Rotation of collector discs behind apertures allows recording of the long term evolution of deposition. QMBs measure mass change via the frequency deviations of vibrating quartz crystals. These diagnostics are used to investigate erosion/deposition during JET-C carbon operation and JET-ILW (ITER-like wall) beryllium/tungsten operation. A simple geometrical model utilising experimental data is used to model the time-dependent collector deposition profiles, demonstrating good qualitative agreement with experimental results. Overall, the JET-ILW collector deposition is reduced by an order of magnitude relative to JET-C, with beryllium replacing carbon as the dominant deposit. However, contrary to JET-C, in JET-ILW there is more deposition on the outer collector than the inner. This reversal of deposition asymmetry is investigated using an analysis of QMB data and is attributed to the different chemical properties of carbon and beryllium.

  7. Development of ITER Divertor Vertical Target with Annular Flow Concept - II: Development of Brazing Technique for CFC/CuCrZr Joint and Heating Test of Large-Scale Mock-Up

    SciTech Connect

    Ezato, K.; Dairaku, M.; Taniguchi, M.; Sato, K.; Suzuki, S.; Akiba, M.; Ibbott, C.; Tivey, R.

    2004-12-15

    The first fabrication and heating test of a large-scale carbon-fiber-composite (CFC) monoblock divertor mock-up using an annular flow concept has been performed to demonstrate its manufacturability and thermomechanical performance. This mock-up is based on the design of the lower part of the vertical target of the International Thermonuclear Experimental Reactor (ITER) divertor adapted for the annular flow concept. The annular cooling tube consists of two concentric tubes: an outer tube made of CuCrZr and an inner stainless steel tube with a twisted external fin. Prior to the fabrication of the mock-up, brazed joint tests between the CFC monoblock and the CuCrZr tube have been carried out to find the suitable heat treatment mitigating loss of the high mechanical strength of the CuCrZr material. A basic mechanical examination of CuCrZr undergoing the brazing heat treatment and finite element method analyses are also performed to support the design of the mock-up. High heat flux tests on the large-scale divertor mock-up have been performed in an ion beam facility. The mock-up has successfully withstood more than 1000 thermal cycles of 20 MW/m{sup 2} for 15 s and 3000 cycles of >10 MW/m{sup 2} for 15 s, which simulates the heat load condition of the ITER divertor. No degradation of the thermal performance of the mock-up has been observed throughout the thermal cycle test although in the tile with exposure to the heat flux of 20 MW/m{sup 2}, the erosion depth has been measured as 5.8 and 8.8 mm at the 300th and 500th cycles.

  8. Modelling of the material transport and layer formation in the divertor of JET: Comparison of ITER-like wall with full carbon wall conditions

    NASA Astrophysics Data System (ADS)

    Kirschner, A.; Matveev, D.; Borodin, D.; Airila, M.; Brezinsek, S.; Groth, M.; Wiesen, S.; Widdowson, A.; Beal, J.; Esser, H. G.; Likonen, J.; Bekris, N.; Ding, R.

    2015-08-01

    Impurity transport within the inner JET divertor has been modelled with ERO to estimate the transport to and the resulting deposition at remote areas. Various parametric studies involving divertor plasma conditions and strike point position have been performed. In JET-ILW (beryllium main chamber and tungsten divertor) beryllium, flowing from the main chamber into the divertor and then effectively reflected at the tungsten divertor tiles, is transported to remote areas. The tungsten flux to remote areas in L-Mode is in comparison to the beryllium flux negligible due to small sputtering. However, tungsten is sputtered during ELMs in H-Mode conditions. Nevertheless, depending on the plasma conditions, strike point position and the location of the remote area, the maximum resulting tungsten flux to remote areas is at least ∼3 times lower than the corresponding beryllium flux. Modelled beryllium and tungsten deposition on a rotating collector probe located below tile 5 is in good agreement with measurements if the beryllium influx into the inner divertor is assumed to be in the range of 0.1% relative to the deuterium ion flux and erosion due to fast charge exchange neutrals is considered. Comparison between JET-ILW and JET-C is presented.

  9. Iter

    NASA Astrophysics Data System (ADS)

    Iotti, Robert

    2015-04-01

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

  10. Investigation on the erosion/deposition processes in the ITER-like wall divertor at JET using glow discharge optical emission spectrometry technique

    NASA Astrophysics Data System (ADS)

    Ruset, C.; Grigore, E.; Luculescu, C.; Tiseanu, I.; Likonen, J.; Mayer, M.; Rubel, M.; Matthews, G. F.; contributors, JET

    2016-02-01

    As a complementary method to Rutherford back scattering (RBS), glow discharge optical emission spectrometry (GDOES) was used to investigate the depth profiles of W, Mo, Be, O and C concentrations into marker coatings (CFC/Mo/W/Mo/W) and the substrate of divertor tiles up to a depth of about 100 μm. A number of 10 samples cored from particular areas of the divertor tiles were analyzed. The results presented in this paper are valid only for those areas and they cannot be extrapolated to the entire tile. Significant deposition of Be was measured on Tile 3 (near to the top), Tile 6 (at about 40 mm from the innermost edge) and especially on Tile 0 (HFGC). Preliminary experiments seem to indicate a penetration of Be through the pores and imperfections of CFC material up to a depth of 100 μm in some cases. No erosion and a thin layer of Be (<1 μm) was detected on Tiles 4, 7 and 8. On Tile 1 no erosion was found at about 1/3 from bottom.

  11. Actively convected liquid metal divertor

    NASA Astrophysics Data System (ADS)

    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.

  12. Modeling of divertor geometry effects in China fusion engineering testing reactor by SOLPS/B2-Eirene

    SciTech Connect

    Zhao, M. L.; Chen, Y. P.; Li, G. Q.; Luo, Z. P.; Guo, H. Y.; Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031; General Atomics, P.O. Box 85608, San Diego, California 92186 ; Ye, M. Y.; Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031 ; Tendler, M.

    2014-05-15

    The China Fusion Engineering Testing Reactor (CFETR) is currently under design. The SOLPS/B2-Eirene code package is utilized for the design and optimization of the divertor geometry for CFETR. Detailed modeling is carried out for an ITER-like divertor configuration and one with relatively open inner divertor structure, to assess, in particular, peak power loading on the divertor target, which is a key issue for the operation of a next-step fusion machine, such as ITER and CFETR. As expected, the divertor peak heat flux greatly exceeds the maximum steady-state heat load of 10 MW/m{sup 2}, which is a limit dictated by engineering, for both divertor configurations with a wide range of edge plasma conditions. Ar puffing is effective at reducing divertor peak heat fluxes below 10 MW/m{sup 2} even at relatively low densities for both cases, favoring the divertor configuration with more open inner divertor structure.

  13. Divertor parameters and divertor operation in ASDEX

    NASA Astrophysics Data System (ADS)

    Fussmann, G.; Ditte, U.; Eckstein, W.; Grave, T.; Keilhacker, M.; McCormick, K.; Murmann, H.; Röhr, H.; Elshaer, M.; Steuer, K.-H.; Szymanski, Z.; Wagner, F.; Becker, G.; Bernhardi, K.; Eberhagen, A.; Gehre, O.; Gernhardt, J.; Gierke, G. V.; Glock, E.; Gruber, O.; Haas, G.; Hesse, M.; Janeschitz, G.; Karger, F.; Kissel, S.; Klüber, O.; Kornherr, M.; Lisitano, G.; Mayer, H. M.; Meisel, D.; Müller, E. R.; Poschenrieder, W.; Ryter, F.; Rapp, H.; Schneider, F.; Siller, G.; Smeulders, P.; Söldner, F.; Speth, E.; Stäbler, A.; Vollmer, O.

    1984-12-01

    Recent measurements of plasma boundary and divertor scrape-off parameters for ohmically and neutral injection heated plasmas are presented. For these data the power flow onto the divertor plates and the sputtering rates at the plates are calculated and compared with separate measurements. The impurity behaviour in front of the plates is also discussed.

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

    SciTech Connect

    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.

  15. Divertor heat flux mitigation in the National Spherical Torus Experiment

    SciTech Connect

    Soukhanovskii, V. A.; Maingi, R.; Gates, D.A.; Menard, J.E.; Bush, C.E.

    2009-01-01

    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 , 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(-2) to 0.5-2 MW m(-2) in small-ELM 0.8-1.0 MA, 4-6 MW neutral beam injection-heated H-mode discharges. A self-consistent picture of the 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.

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

    NASA Astrophysics Data System (ADS)

    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

  17. Impurity-induced divertor plasma oscillations

    DOE PAGESBeta

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

    2016-01-07

    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 ionmore » 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. As a result, the implications of the impurity-induced plasma oscillations for divertor operation in the next generation tokamaks are also discussed.« less

  18. Impurity-induced divertor plasma oscillations

    NASA Astrophysics Data System (ADS)

    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.

  19. Models for poloidal divertors

    SciTech Connect

    Post, D.E.; Heifetz, D.; Petravic, M.

    1982-07-01

    Recent progress in models for poloidal divertors has both helped to explain current divertor experiments and contributed significantly to design efforts for future large tokamak (INTOR, etc.) divertor systems. These models range in sophistication from zero-dimensional treatments and dimensional analysis to two-dimensional models for plasma and neutral particle transport which include a wide variety of atomic and molecular processes as well as detailed treatments of the plasma-wall interaction. This paper presents a brief review of some of these models, describing the physics and approximations involved in each model. We discuss the wide variety of physics necessary for a comprehensive description of poloidal divertors. To illustrate the progress in models for poloidal divertors, we discuss some of our recent work as typical examples of the kinds of calculations being done.

  20. Divertor efficiency in ASDEX

    NASA Astrophysics Data System (ADS)

    Engelhardt, W.; Becker, G.; Behringer, K.; Campbell, D.; Eberhagen, A.; Fussmann, G.; Gehre, O.; Gierke, G. V.; Glock, E.; Haas, G.; Huang, M.; Karger, F.; Keilhacker, M.; KlÜber, O.; Kornherr, M.; Lisitano, G.; Mayer, H.-M.; Meisel, D.; Müller, E. R.; Murmann, H.; Niedermeyer, H.; Poschenrieder, W.; Rapp, H.; Schneider, F.; Siller, G.; Steuer, K.-H.; Venus, G.; Vernickel, H.; Wagner, F.

    1982-12-01

    The divertor efficiency in ASDEX is discussed for ohmically heated plasmas. The parameters of the boundary layer both in the torus midplane and the divertor chamber have been measured. The results are reasonably well understood in terms of parallel and perpendicular transport. A high pressure of neutral hydrogen builds up in the divertor chamber and Franck-Condon particles recycle back through the divertor throat. Due to dissociation processes the boundary plasma is effectively cooled before it reaches the neutralizer plates. The shielding property of the boundary layer against impurity influx is comparable to that of a limiter plasma. The transport of iron is numerically simulated for an iron influx produced by sputtering of charge exchange neutrals at the wall. The results are consistent with the measured iron concentration. First results from a comparison of the poloidal divertor with toroidally closed limiters (stainless steel, carbon) are given. Diverted discharges are considerably cleaner and easier to create.

  1. Spectroscopy of divertor plasmas

    SciTech Connect

    Isler, R.C.

    1995-12-31

    The requirements for divertor spectroscopy are treated with respect to instrumentation and observations on present machines. Emphasis is placed on quantitative measurements.of impurity concentrations from the interpretation of spectral line intensities. The possible influence of non-Maxwellian electron distributions on spectral line excitation in the divertor is discussed. Finally the use of spectroscopy for determining plasma temperature, density, and flows is examined.

  2. Divertor design for the Tokamak Physics Experiment

    SciTech Connect

    Hill, D.N.; Braams, B.; Brooks, J.N.

    1994-05-01

    In this paper we discuss the present divertor design for the planned TPX tokamak, which will explore the physics and technology of steady-state (1000s pulses) heat and particle removal in high confinement (2--4{times} L-mode), high beta ({beta}{sub N} {ge} 3) divertor plasmas sustained by non-inductive current drive. The TPX device will operate in the double-null divertor configuration, with actively cooled graphite targets forming a deep (0.5 m) slot at the outer strike point. The peak heat flux on, the highly tilted (74{degrees} from normal) re-entrant (to recycle ions back toward the separatrix) will be in the range of 4--6 MW/m{sup 2} with 18 MW of neutral beams and RF heating power. The combination of active pumping and gas puffing (deuterium plus impurities), along with higher heating power (45 MW maximum) will allow testing of radiative divertor concepts at ITER-like power densities.

  3. Divertor IR thermography on Alcator C-Moda)

    NASA Astrophysics Data System (ADS)

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

    2010-10-01

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

  4. Divertor IR thermography on Alcator C-Mod.

    PubMed

    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. PMID:21034041

  5. Divertor IR thermography on Alcator C-Mod

    SciTech Connect

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

    2010-10-15

    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 deg. toroidal sector has been given a 2 deg. 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.

  6. Development of a Spatially Resolving X-Ray Crystal Spectrometer (XCS) for Measurement of Ion-Temperature (Ti) and Rotation-Velocity (v) Profiles in ITER

    SciTech Connect

    Hill, K W; Delgado-Aprico, L; Johnson, D; Feder, R; Beiersdorfer,; Dunn, J; Morris, K; Wang, E; Reinke, M; Podpaly, Y; Rice, J E; Barnsley, R; O'Mullane, M; Lee, S G

    2010-05-21

    Imaging XCS arrays are being developed as a US-ITER activity for Doppler measurement of Ti and v profiles of impurities (W, Kr, Fe) with ~7 cm (a/30) and 10-100 ms resolution in ITER. The imaging XCS, modeled after a PPPL-MIT instrument on Alcator C-Mod, uses a spherically bent crystal and 2d x-ray detectors to achieve high spectral resolving power (E/dE>6000) horizontally and spatial imaging vertically. Two arrays will measure Ti and both poloidal and toroidal rotation velocity profiles. Measurement of many spatial chords permits tomographic inversion for inference of local parameters. The instrument design, predictions of performance, and results from C-Mod will be presented.

  7. Development of a spatially resolving x-ray crystal spectrometer for measurement of ion-temperature (T{sub i}) and rotation-velocity (v) profiles in ITER

    SciTech Connect

    Hill, K. W.; Bitter, M.; Delgado-Aparicio, L.; Johnson, D.; Feder, R.; Beiersdorfer, P.; Dunn, J.; Morris, K.; Wang, E.; Reinke, M.; Podpaly, Y.; Rice, J. E.; Barnsley, R.; O'Mullane, M.; Lee, S. G.

    2010-10-15

    Imaging x-ray crystal spectrometer (XCS) arrays are being developed as a US-ITER activity for Doppler measurement of T{sub i} and v profiles of impurities (W, Kr, and Fe) with {approx}7 cm (a/30) and 10-100 ms resolution in ITER. The imaging XCS, modeled after a prototype instrument on Alcator C-Mod, uses a spherically bent crystal and 2D x-ray detectors to achieve high spectral resolving power (E/dE>6000) horizontally and spatial imaging vertically. Two arrays will measure T{sub i} and both poloidal and toroidal rotation velocity profiles. The measurement of many spatial chords permits tomographic inversion for the inference of local parameters. The instrument design, predictions of performance, and results from C-Mod are presented.

  8. Development Of a Spatially Resolving X-ray Crystal Spectrometer For Measurement Of Ion-temperature (Ti) And Rotation-velocity (v) Profiles in ITER

    SciTech Connect

    Hill, K W; Delgado-Aparico, L; Johnson, David; Feder, R; Beiersdorfer, P; Dunn, James; Morris, K; Wang, E; Reinke, M; Podpaly, Y; Rice, J E; Barnsley, R; O'Mullane, M

    2010-12-15

    Imaging x-ray crystal spectrometer XCS arrays are being developed as a US-ITER activity for Doppler measurement of Ti and v profiles of impurities (W, Kr, and Fe) with ~ 7 cm (a/30) and 10-100 ms resolution in ITER. The imaging XCS, modeled after a prototype instrument on Alcator C-Mod, uses a spherically bent crystal and 2D x-ray detectors to achieve high spectral resolving power (E / dE > 6000) horizontally and spatial imaging vertically. Two arrays will measure Ti and both poloidal and toroidal rotation velocity profiles. The measurement of many spatial chords permits tomographic inversion for the inference of local parameters. The instrument design, predictions of performance, and results from C-Mod are presented.

  9. Diagnostic options for radiative divertor feedback control on NSTX-Ua)

    NASA Astrophysics Data System (ADS)

    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).

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

    SciTech Connect

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

    2012-10-15

    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{sub peak} Less-Than-Or-Slanted-Equal-To 15 MW/m{sup 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{sub 2} or CD{sub 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{sup 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).

  11. The snowflake divertor

    DOE PAGESBeta

    Ryutov, D. D.; Soukhanovskii, V. A.

    2015-11-17

    The snowflake magnetic configuration is characterized by the presence of two closely spaced poloidal field nulls that create a characteristic hexagonal (reminiscent of a snowflake) separatrix structure. The magnetic field properties and the plasma behaviour in the snowflake are determined by the simultaneous action of both nulls, this generating a lot of interesting physics, as well as providing a chance for improving divertor performance. One of the most interesting effects of the snowflake geometry is the heat flux sharing between multiple divertor channels. The authors summarise experimental results obtained with the snowflake configuration on several tokamaks. Wherever possible, relation tomore » the existing theoretical models is described. Divertor concepts utilizing the properties of a snowflake configuration are briefly discussed.« less

  12. The snowflake divertor

    SciTech Connect

    Ryutov, D. D.; Soukhanovskii, V. A.

    2015-11-17

    The snowflake magnetic configuration is characterized by the presence of two closely spaced poloidal field nulls that create a characteristic hexagonal (reminiscent of a snowflake) separatrix structure. The magnetic field properties and the plasma behaviour in the snowflake are determined by the simultaneous action of both nulls, this generating a lot of interesting physics, as well as providing a chance for improving divertor performance. One of the most interesting effects of the snowflake geometry is the heat flux sharing between multiple divertor channels. The authors summarise experimental results obtained with the snowflake configuration on several tokamaks. Wherever possible, relation to the existing theoretical models is described. Divertor concepts utilizing the properties of a snowflake configuration are briefly discussed.

  13. Development of a radiative divertor for DIII-D

    SciTech Connect

    Allen, S.L.; Brooks, N.H.; Campbell, R.B.

    1994-07-01

    We have used experiments and modeling to develop a new radiative divertor configuration for DIII-D. Gas puffing experiments with the existing open divertor have shown the creation of a localized ({approximately}10 cm diameter) radiation zone which results in substantial reduction (3--10) in the divertor heat flux while {delta}{sub E} remains {approximately}2 times ITER-89P scaling. However, ne increases with D{sub 2} puffing, and Z{sub eff} increases with neon puffing. Divertor structures are required to minimize the effects on the core plasma. The UEDGE fluid code, benchmarked with DIII-D data, and the DEGAS neutrals transport code are used to estimate the effectiveness of divertor configurations; slots reduce the core ionization more than baffles. The overall divertor shape is set by confinement studies which indicate that high triangularity ({delta} {approximately}0.8) is important for high {tau}{sub E} VH-modes. Results from engineering feasibility studies, including diagnostic access, will be presented.

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

    NASA Astrophysics Data System (ADS)

    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.

  15. Divertor plasma detachment

    NASA Astrophysics Data System (ADS)

    Krasheninnikov, S. I.; Kukushkin, A. S.; Pshenov, A. A.

    2016-05-01

    Regime with the plasma detached from the divertor targets (detached divertor regime) is a natural continuation of the high recycling conditions to higher density and stronger impurity radiation loss. Both the theoretical considerations and experimental data show clearly that the increase of the impurity radiation loss and volumetric plasma recombination causes the rollover of the plasma flux to the target when the density increases, which is the manifestation of detachment. Plasma-neutral friction (neutral viscosity effects), although important for the sustainment of high density/pressure plasma upstream and providing the conditions for efficient recombination and power loss, is not directly involved in the reduction of the plasma flux to the targets. The stability of detachment is also discussed.

  16. The snowflake divertor

    SciTech Connect

    Ryutov, D. D.; Soukhanovskii, V. A.

    2015-11-15

    The snowflake magnetic configuration is characterized by the presence of two closely spaced poloidal field nulls that create a characteristic hexagonal (reminiscent of a snowflake) separatrix structure. The magnetic field properties and the plasma behaviour in the snowflake are determined by the simultaneous action of both nulls, this generating a lot of interesting physics, as well as providing a chance for improving divertor performance. Among potential beneficial effects of this geometry are: increased volume of a low poloidal field around the null, increased connection length, and the heat flux sharing between multiple divertor channels. The authors summarise experimental results obtained with the snowflake configuration on several tokamaks. Wherever possible, relation to the existing theoretical models is described.

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

    NASA Astrophysics Data System (ADS)

    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.

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

    SciTech Connect

    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.

  19. Design, R&D and commissioning of EAST tungsten divertor

    NASA Astrophysics Data System (ADS)

    Yao, D. M.; Luo, G. N.; Zhou, Z. B.; Cao, L.; Li, Q.; Wang, W. J.; Li, L.; Qin, S. G.; Shi, Y. L.; Liu, G. H.; Li, J. G.

    2016-02-01

    After commissioning in 2005, the EAST superconducting tokamak had been operated with its water cooled divertors for eight campaigns up to 2012, employing graphite as plasma facing material. With increase in heating power over 20 MW in recent years, the heat flux going to the divertors rises rapidly over 10 MW m-2 for steady state operation. To accommodate the rapid increasing heat load in EAST, the bolting graphite tile divertor must be upgraded. An ITER-like tungsten (W) divertor has been designed and developed; and firstly used for the upper divertor of EAST. The EAST upper W divertor is modular structure with 80 modules in total. Eighty sets of W/Cu plasma-facing components (PFC) with each set consisting of an outer vertical target (OVT), an inner vertical target (IVT) and a DOME, are attached to 80 stainless steel cassette bodies (CB) by pins. The monoblock W/Cu-PFCs have been developed for the strike points of both OVT and IVT, and the flat type W/Cu-PFCs for the DOME and the baffle parts of both OVT and IVT, employing so-called hot isostatic pressing (HIP) technology for tungsten to CuCrZr heat sink bonding, and electron beam welding for CuCrZr to CuCrZr and CuCrZr to other material bonding. Both monoblock and flat type PFC mockups passed high heat flux (HHF) testing by means of electron beam facilities. The 80 divertor modules were installed in EAST in 2014 and results of the first commissioning are presented in this paper.

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

    SciTech Connect

    Lomanowski, B. A. Sharples, R. M.; Meigs, A. G.; Conway, N. J.; Zastrow, K.-D.; Heesterman, P.; Kinna, D. [EURATOM Collaboration: JET-EFDA Team

    2014-11-15

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

  1. Modeling results for a linear simulator of a divertor

    SciTech Connect

    Hooper, E.B.; Brown, M.D.; Byers, J.A.; Casper, T.A.; Cohen, B.I.; Cohen, R.H.; Jackson, M.C.; Kaiser, T.B.; Molvik, A.W.; Nevins, W.M.; Nilson, D.G.; Pearlstein, L.D.; Rognlien, T.D.

    1993-06-23

    A divertor simulator, IDEAL, has been proposed by S. Cohen to study the difficult power-handling requirements of the tokamak program in general and the ITER program in particular. Projections of the power density in the ITER divertor reach {approximately} 1 Gw/m{sup 2} along the magnetic fieldlines and > 10 MW/m{sup 2} on a surface inclined at a shallow angle to the fieldlines. These power densities are substantially greater than can be handled reliably on the surface, so new techniques are required to reduce the power density to a reasonable level. Although the divertor physics must be demonstrated in tokamaks, a linear device could contribute to the development because of its flexibility, the easy access to the plasma and to tested components, and long pulse operation (essentially cw). However, a decision to build a simulator requires not just the recognition of its programmatic value, but also confidence that it can meet the required parameters at an affordable cost. Accordingly, as reported here, it was decided to examine the physics of the proposed device, including kinetic effects resulting from the intense heating required to reach the plasma parameters, and to conduct an independent cost estimate. The detailed role of the simulator in a divertor program is not explored in this report.

  2. Heat loads to divertor nearby components from secondary radiation evolved during plasma instabilities

    NASA Astrophysics Data System (ADS)

    Sizyuk, V.; Hassanein, A.

    2015-01-01

    A fundamental issue in tokamak operation related to power exhaust during plasma instabilities is the understanding of heat and particle transport from the core plasma into the scrape-off layer and to plasma-facing materials. During abnormal and disruptive operation in tokamaks, radiation transport processes play a critical role in divertor/edge-generated plasma dynamics and are very important in determining overall lifetimes of the divertor and nearby components. This is equivalent to or greater than the effect of the direct impact of escaped core plasma on the divertor plate. We have developed and implemented comprehensive enhanced physical and numerical models in the upgraded HEIGHTS package for simulating detailed photon and particle transport in the evolved edge plasma during various instabilities. The paper describes details of a newly developed 3D Monte Carlo radiation transport model, including optimization methods of generated plasma opacities in the full range of expected photon spectra. Response of the ITER divertor's nearby surfaces due to radiation from the divertor-developed plasma was simulated by using actual full 3D reactor design and magnetic configurations. We analyzed in detail the radiation emission spectra and compared the emission of both carbon and tungsten as divertor plate materials. The integrated 3D simulation predicted unexpectedly high damage risk to the open stainless steel legs of the dome structure in the current ITER design from the intense radiation during a disruption on the tungsten divertor plate.

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

    NASA Astrophysics Data System (ADS)

    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.

  4. Erosion and deposition in the JET divertor during the first ILW campaign

    NASA Astrophysics Data System (ADS)

    Mayer, M.; Krat, S.; Van Renterghem, W.; Baron-Wiechec, A.; Brezinsek, S.; Bykov, I.; Coad, P.; Gasparyan, Yu; Heinola, K.; Likonen, J.; Pisarev, A.; Ruset, C.; de Saint-Aubin, G.; Widdowson, A.; Contributors, JET

    2016-02-01

    Erosion and deposition were studied in the JET divertor during the first JET ITER-like wall campaign 2011 to 2012 using marker tiles. An almost complete poloidal section consisting of tiles 0, 1, 3, 4, 6, 7, 8 was studied. The data from divertor tile surfaces were completed by the analysis of samples from remote divertor areas and from the inner wall cladding. The total mass of material deposited in the divertor decreased by a factor of 4-9 compared to the deposition of carbon during all-carbon JET operation before 2010. Deposits in 2011 to 2012 consist mainly of beryllium with 5-20 at.% of carbon and oxygen, respectively, and small amounts of Ni, Cr, Fe and W. This decrease of material deposition in the divertor is accompanied by a decrease of total deuterium retention inside the JET vessel by a factor of 10 to 20. The detailed erosion/deposition pattern in the divertor with the ITER-like wall configuration shows rigorous changes compared to the pattern with the all-carbon JET configuration.

  5. Asymmetric divertor biasing in MAST

    NASA Astrophysics Data System (ADS)

    Helander, P.; Cohen, R.; Counsell, G. C.; Ryutov, D. D.

    2002-11-01

    Experiments are being carried out on the Mega-Ampere Spherical Tokamak (MAST) where the divertor tiles are electrically biased in a toroidally alternating way. The aim is to induce convective cells in the divertor plasma, broaden the SOL and reduce the divertor heat load. This paper describes the underlying theory and experimental results. Criteria are presented for achieving strong broadening and exciting shear-flow turbulence in the SOL, and properties of the expected turbulence are derived. It is also shown that magnetic shear near the X-point is likely to confine the potential perturbations to the divertor region, leaving the part of the SOL that is in direct contact with the core plasma intact. Preliminary comparison of the theory with MAST data is encouraging: the distortion of the heat deposition pattern, its broadening, and the incremental heat load are qualitatively in agreement; quantitative comparisons are underway.

  6. Defining the infrared systems for ITER.

    PubMed

    Reichle, R; Andrew, P; Counsell, G; Drevon, J-M; Encheva, A; Janeschitz, G; Johnson, D; Kusama, Y; Levesy, B; Martin, A; Pitcher, C S; Pitts, R; Thomas, D; Vayakis, G; Walsh, M

    2010-10-01

    The International Thermonuclear Experimental Reactor will have wide angle viewing systems and a divertor thermography diagnostic, which shall provide infrared coverage of the divertor and large parts of the first wall surfaces with spatial and temporal resolution adequate for operational purposes and higher resolved details of the divertor and other areas for physics investigations. We propose specifications for each system such that they jointly respond to the requirements. Risk analysis driven priorities for future work concern mirror degradation, interfaces with other diagnostics, radiation damage to refractive optics, reflections, and the development of calibration and measurement methods for varying optical and thermal target properties. PMID:21033997

  7. Defining the infrared systems for ITER

    SciTech Connect

    Reichle, R.; Andrew, P.; Drevon, J.-M.; Encheva, A.; Janeschitz, G.; Levesy, B.; Martin, A.; Pitcher, C. S.; Pitts, R.; Thomas, D.; Vayakis, G.; Walsh, M.; Counsell, G.; Johnson, D.; Kusama, Y.

    2010-10-15

    The International Thermonuclear Experimental Reactor will have wide angle viewing systems and a divertor thermography diagnostic, which shall provide infrared coverage of the divertor and large parts of the first wall surfaces with spatial and temporal resolution adequate for operational purposes and higher resolved details of the divertor and other areas for physics investigations. We propose specifications for each system such that they jointly respond to the requirements. Risk analysis driven priorities for future work concern mirror degradation, interfaces with other diagnostics, radiation damage to refractive optics, reflections, and the development of calibration and measurement methods for varying optical and thermal target properties.

  8. Survivability of dust in tokamaks: Dust transport in the divertor sheath

    SciTech Connect

    Delzanno, Gian Luca; Tang, Xianzhu

    2014-02-15

    The survivability of dust being transported in the magnetized sheath near the divertor plate of a tokamak and its impact on the desired balance of erosion and redeposition for a steady-state reactor are investigated. Two different divertor scenarios are considered. The first is characterized by an energy flux perpendicular to the plate q{sub 0}≃1 MW/m{sup 2} typical of current short-pulse tokamaks. The second has q{sub 0}≃10 MW/m{sup 2} and is relevant to long-pulse machines like ITER or Demonstration Power Plant. It is shown that micrometer dust particles can survive rather easily near the plates of a divertor plasma with q{sub 0}≃1 MW/m{sup 2} because thermal radiation provides adequate cooling for the dust particle. On the other hand, the survivability of micrometer dust particles near the divertor plates is drastically reduced when q{sub 0}≃10 MW/m{sup 2}. Micrometer dust particles redeposit their material non-locally, leading to a net poloidal mass migration across the divertor. Smaller particles (with radius ∼0.1 μm) cannot survive near the divertor and redeposit their material locally. Bigger particle (with radius ∼10 μm) can instead survive partially and move outside the divertor strike points, thus causing a net loss of divertor material to dust accumulation inside the chamber and some non-local redeposition. The implications of these results for ITER are discussed.

  9. Innovative tokamak DEMO first wall and divertor material concepts

    NASA Astrophysics Data System (ADS)

    Wong, C. P. C.

    2009-06-01

    ITER has selected Be as the first wall and C and W as the divertor surface materials. When extrapolated to the DEMO design, C and Be layers will not be suitable due to radiation damage. The remaining material, W, could also suffer radiation damage from helium ion implantation and experience blistering at the first wall and form submicron fine structure at the divertor. In this paper we introduce a new concept called the boron W-mesh (BW-mesh) in which B is infiltrated into a W-mesh. The goal is to use a thin coating of B to protect the W-mesh from helium ion damage and to maintain a sufficient amount of B to protect the W from transient events like edge localized modes (ELMs) and disruptions. Critical issues and corresponding development of this BW-mesh concept have been identified, including the need for real time boronization.

  10. ITER tokamak device

    NASA Astrophysics Data System (ADS)

    Doggett, J.; Salpietro, E.; Shatalov, G.

    1991-07-01

    The results of the Conceptual Design Activities for the International Thermonuclear Experimental Reactor (ITER) are summarized. These activities, carried out between April 1988 and December 1990, produced a consistent set of technical characteristics and preliminary plans for co-ordinated research and development support of ITER, a conceptual design, a description of design requirements and a preliminary construction schedule and cost estimate. After a description of the design basis, an overview is given of the tokamak device, its auxiliary systems, facility and maintenance. The interrelation and integration of the various subsystems that form the ITER tokamak concept are discussed. The 16 ITER equatorial port allocations, used for nuclear testing, diagnostics, fueling, maintenance, and heating and current drive, are given, as well as a layout of the reactor building. Finally, brief descriptions are given of the major ITER sub-systems, i.e., (1) magnet systems (toroidal and poloidal field coils and cryogenic systems), (2) containment structures (vacuum and cryostat vessels, machine gravity supports, attaching locks, passive loops and active coils), (3) first wall, (4) divertor plate (design and materials, performance and lifetime, a.o.), (5) blanket/shield system, (6) maintenance equipment, (7) current drive and heating, (8) fuel cycle system, and (9) diagnostics.

  11. Electric field divertor plasma pump

    DOEpatents

    Schaffer, M.J.

    1994-10-04

    An electric field plasma pump includes a toroidal ring bias electrode positioned near the divertor strike point of a poloidal divertor of a tokamak, or similar plasma-confining apparatus. For optimum plasma pumping, the separatrix of the poloidal divertor contacts the ring electrode, which then also acts as a divertor plate. A plenum or other duct near the electrode includes an entrance aperture open to receive electrically-driven plasma. The electrode is insulated laterally with insulators, one of which is positioned opposite the electrode at the entrance aperture. An electric field E is established between the ring electrode and a vacuum vessel wall, with the polarity of the bias applied to the electrode being relative to the vessel wall selected such that the resultant electric field E interacts with the magnetic field B already existing in the tokamak to create an E [times] B/B[sup 2] drift velocity that drives plasma into the entrance aperture. The pumped plasma flow into the entrance aperture is insensitive to variations, intentional or otherwise, of the pump and divertor geometry. Pressure buildups in the plenum or duct connected to the entrance aperture in excess of 10 mtorr are achievable. 11 figs.

  12. Electric field divertor plasma pump

    DOEpatents

    Schaffer, Michael J.

    1994-01-01

    An electric field plasma pump includes a toroidal ring bias electrode (56) positioned near the divertor strike point of a poloidal divertor of a tokamak (20), or similar plasma-confining apparatus. For optimum plasma pumping, the separatrix (40) of the poloidal divertor contacts the ring electrode (56), which then also acts as a divertor plate. A plenum (54) or other duct near the electrode (56) includes an entrance aperture open to receive electrically-driven plasma. The electrode (56) is insulated laterally with insulators (63,64), one of which (64) is positioned opposite the electrode at the entrance aperture. An electric field E is established between the ring electrode (56) and a vacuum vessel wall (22), with the polarity of the bias applied to the electrode being relative to the vessel wall selected such that the resultant electric field E interacts with the magnetic field B already existing in the tokamak to create an E.times.B/B.sup.2 drift velocity that drives plasma into the entrance aperture. The pumped plasma flow into the entrance aperture is insensitive to variations, intentional or otherwise, of the pump and divertor geometry. Pressure buildups in the plenum or duct connected to the entrance aperture in excess of 10 mtorr are achievable.

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

    NASA Astrophysics Data System (ADS)

    Groth, M.; Brezinsek, S.; Belo, P.; Brix, M.; Calabro, G.; Chankin, A.; Clever, M.; Coenen, J. W.; Corrigan, G.; Drewelow, P.; Guillemaut, C.; Harting, D.; Huber, A.; Jachmich, S.; Järvinen, A.; Kruezi, U.; Lawson, K. D.; Lehnen, M.; Maggi, C. F.; Marchetto, C.; Marsen, S.; Maviglia, F.; Meigs, A. G.; Moulton, D.; Silva, C.; Stamp, M. F.; Wiesen, S.

    2015-08-01

    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.

  14. A review of ELMs in divertor tokamaks

    SciTech Connect

    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.

  15. The ITER project construction status

    NASA Astrophysics Data System (ADS)

    Motojima, O.

    2015-10-01

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

  16. Exploration of magnetic perturbation effects on advanced divertor configurations in NSTX-U

    NASA Astrophysics Data System (ADS)

    Frerichs, H.; Schmitz, O.; Waters, I.; Canal, G. P.; Evans, T. E.; Feng, Y.; Soukhanovskii, V. A.

    2016-06-01

    The control of divertor heat loads - both steady state and transient - remains a key challenge for the successful operation of ITER and FNSF. Magnetic perturbations provide a promising technique to control ELMs (Edge Localized Modes) (transients), but understanding their detailed impact is difficult due to their symmetry breaking nature. One approach for reducing steady state heat loads is so called "advanced divertors" which aim at optimizing the magnetic field configuration: the snowflake and the (super-)X-divertor. It is likely that both concepts - magnetic perturbations and advanced divertors - will have to work together, and we explore their interaction based on the NSTX-U setup. An overview of different divertor configurations under the impact of magnetic perturbations is presented, and the resulting impact on plasma edge transport is investigated with the EMC3-EIRENE code. Variations in size of the magnetic footprint of the perturbed separatrix are found, which are related to the level of flux expansion on the divertor target. Non-axisymmetric peaking of the heat flux related to the perturbed separatrix is found at the outer strike point, but only in locations where flux expansion is not too large.

  17. UEDGE Simulation of Triple-X Divertors

    NASA Astrophysics Data System (ADS)

    Wiley, J.; Kotschenreuther, M.; Valanju, P.; Pekker, M.; Rognlien, T.

    2006-04-01

    Novel magnetic divertors with additional X-points downstream from the main plasma X-point have been proposed to overcome reactor heat flux limitations. These divertor designs may allow a fully detached state at the divertor plate - without the poor confinement and disruptive tendencies by avoiding x-point MARFEs found in conventional divertor magnetic geometries. These new configurations are examined using UEDGE for existing machines that are considering experimental implementation of these divertors: PEGASUS, MAST, and EAST(China's new long-pulse, superconducting tokamak) as well as proposed reactor designs.

  18. Experimental developments towards an ITER thermography diagnostic

    NASA Astrophysics Data System (ADS)

    Reichle, R.; Brichard, B.; Escourbiac, F.; Gardarein, J. L.; Hernandez, D.; Le Niliot, C.; Rigollet, F.; Serra, J. J.; Badie, J. M.; van Ierschot, S.; Jouve, M.; Martinez, S.; Ooms, H.; Pocheau, C.; Rauber, X.; Sans, J. L.; Scheer, E.; Berghmans, F.; Decréton, M.

    2007-06-01

    In the course of the development of a concept for a spectrally resolving thermography diagnostic for the ITER divertor using optical fibres experimental development work has been carried out in three different areas. Firstly ZrF4 fibres and hollow fibres (silica capillaries with internal AG/AgJ coating) were tested in a Co60 irradiation facility under γ irradiation up to doses of 5 kGy and 27 kGy, respectively. The ZrF4 fibres suffered more radiation induced degradation (>1 db/m) then the hollow fibres (0-0.4 db/m). Secondly multi-colour pyroreflectometry is being developed towards tokamak applicability. The emissivity and temperature of tungsten samples were measured in the range of 700-1500 °C. The angular working range for off normal observation of the method was 20-30°. The working distance of the method has been be increased from cm to the m range. Finally, encouraging preliminary results have been obtained concerning the application of pulsed and modulated active thermography.

  19. Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake

    SciTech Connect

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

    2013-10-15

    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. A primary result of our analysis is the emergence of a physical “metric,” the Divertor Index DI, which 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 (DI = 1), and two advanced geometries—the X-Divertor (XD, DI > 1) and the Snowflake (DI < 1). The XD, therefore, cannot be classified as one variant of the Snowflake. By this measure, recent National Spherical Torus Experiment and DIIID experiments are X-Divertors, not Snowflakes.

  20. Kinetic Modeling of Divertor Plasma

    NASA Astrophysics Data System (ADS)

    Ishiguro, Seiji; Hasegawa, Hiroki; Pianpanit, Theerasarn

    2015-11-01

    Particle-in-Cell (PIC) simulation with the Monte Carlo collisions and the cumulative scattering angle coulomb collision can present kinetic dynamics of divertor plasmas. We are developing two types of PIC codes. The first one is the three dimensional bounded PIC code where three dimensional kinetic dynamics of blob is studied and current flow structures related to sheath formation are unveiled. The second one is the one spatial three velocity space dimensional (1D3V) PIC code with the Monte Carlo collisions where formation of detach plasma is studied. First target of our research is to construct self-consistent full kinetic simulation modeling of the linear divertor simulation experiments. This work is performed with the support and under the auspices of NIFS Collaboration Research program (NIFS15KNSS059, NIFS14KNXN279, and NIFS13KNSS038) and the Research Cooperation Program on Hierarchy and Holism in Natural Science at NINS.

  1. The lithium vapor box divertor

    NASA Astrophysics Data System (ADS)

    Goldston, R. J.; Myers, R.; Schwartz, J.

    2016-02-01

    It has long been recognized that volumetric dissipation of the plasma heat flux from a fusion power system is preferable to its localized impingement on a material surface. Volumetric dissipation mitigates both the anticipated very high heat flux and intense particle-induced damage due to sputtering. Recent projections to a tokamak demonstration power plant suggest an immense upstream parallel heat flux, of order 20 GW m-2, implying that fully detached operation may be a requirement for the success of fusion power. Building on pioneering work on the use of lithium by Nagayama et al and by Ono et al as well as earlier work on the gas box divertor by Watkins and Rebut, we present here a concept for a lithium vapor box divertor, in which lithium vapor extracts momentum and energy from a fusion-power-plant divertor plasma, using fully volumetric processes. At the high powers and pressures that are projected this requires a high density of lithium vapor, which must be isolated from the main plasma in order to avoid lithium build-up on the chamber walls or in the plasma. Isolation is achieved through a powerful multi-box differential pumping scheme available only for condensable vapors. The preliminary box-wise calculations are encouraging, but much more work is required to demonstrate the practical viability of this scheme, taking into account at least 2D plasma and vapor flows within and between the vapor boxes and out of the vapor boxes to the main plasma.

  2. Conceptual design of divertor and first wall for DEMO-FNS

    NASA Astrophysics Data System (ADS)

    Sergeev, V. Yu.; Kuteev, B. V.; Bykov, A. S.; Gervash, A. A.; Glazunov, D. A.; Goncharov, P. R.; Dnestrovskij, A. Yu.; Khayrutdinov, R. R.; Klishchenko, A. V.; Lukash, V. E.; Mazul, I. V.; Molchanov, P. A.; Petrov, V. S.; Rozhansky, V. A.; Shpanskiy, Yu. S.; Sivak, A. B.; Skokov, V. G.; Spitsyn, A. V.

    2015-11-01

    Key issues of design of the divertor and the first wall of DEMO-FNS are presented. A double null closed magnetic configuration was chosen with long external legs and V-shaped corners. The divertor employs a cassette design similar to that of ITER. Water-cooled first wall of the tokamak is made of Be tiles and CuCrZr-stainless steel shells. Lithium injection and circulation technologies are foreseen for protection of plasma facing components. Simulations of thermal loads onto the first wall and divertor plates suggest a possibility to distribute heat loads making them less than 10 MW m-2. Evaluations of sputtering and evaporation of plasma-facing materials suggest that lithium may protect the first wall. To prevent Be erosion at the outer divertor plates either the full detached divertor operation or arrangement of the renewal lithium flow on targets should be implemented. Test bed experiments on the Tsefey-M facility with the first wall mockup coated by Ве tiles and cooled by water are presented. The temperature of the surface of tiles reached 280-300 °С at 5 MW m-2 and 600-650 °С at 10.5 MW m-2. The mockup successfully withstood 1000 cycles with the lower thermal loading and 100 cycles with higher thermal loading.

  3. ITER diagnostic systems in development in Ioffe Institute

    SciTech Connect

    Petrov, M.; Afanasyev, V.; Petrov, S.; Mironov, M.; Mukhin, E.; Tolstyakov, S.; Chugunov, I.; Shevelev, A.

    2014-08-21

    Three diagnostic systems are being developed in Ioffe Institute for ITER. Those are Neutral Particle Analysis (NPA), Thomson Scattering in Divertor (TSD) and Gamma Spectroscopy (GS). The main objective of NPA in ITER is to measure D/T fuel ration in plasma on the basis of measurement of neutralized fluxes of D and T ions [1]. Fuel ratio is one of the key parameters needed by ITER control system to provide the optimal conditions in plasma and the most effective plasma burning. Another objective is to measure the distribution function of fast ions (including alpha particles) generated as a result of the additional heating and nuclear fusion reactions. Thomson Scattering in Divertor (TSD) [2] will be used to measure electron temperature and density in the scrape-off layer in outer leg of ITER divertor. The main task of TSD is to protect the machine from divertor overloading. Gamma Spectroscopy (GS) [3] is based on the measurement of spectral lines of MeV range gammas generated in nuclear reactions in plasma. 2-D gamma-ray emission measurements give valuable information on the confined alpha particles in DT plasma. They also provide important information on the location of MeV range runaway electron beams in ITER plasma. For all three cases the physical basis and instrumentation are presented. The simple NPA version for measurements of D/T ratio in DEMO is also briefly described.

  4. Gamma-irradiation tests of IR optical fibres for ITER thermography--a case study

    SciTech Connect

    Reichle, R.; Pocheau, C.; Jouve, M.

    2008-03-12

    In the course of the development of a concept for a spectrally resolving infrared thermography diagnostic for the ITER divertor we have tested 3 types of infrared (IR) fibres in Co{sup 60} irradiation facilities under {gamma} irradiation. The fibres were ZrF{sub 4} (and HfF{sub 4}) fibres from different manufacturers, hollow fibres (silica capillaries with internal Ag/AgJ coating) and a sapphire fibre. For the IR range, only the latter fibre type encourages to go further for neutron tests in a reactor. If one restricted the interest onto the near infrared range, high purity core silica fibres could be used. This study might be seen as a typical example of the relation between diagnostic development for a nuclear environment and irradiation experiments.

  5. Overview of neutron and confined escaping alpha diagnostics planned for ITER

    NASA Astrophysics Data System (ADS)

    Sasao, M.; Krasilnikov, A. V.; Nishitani, T.; Batistoni, P.; Zaveryaev, V.; Kaschuck, Yu A.; Popovichev, S.; Iguchi, T.; Jarvis, O. N.; Kallne, J.; Fiore, C. L.; Roquemore, L.; Heidbrink, W. W.; Donne, A. J. H.; Costley, A. E.; Walker, C.

    2004-07-01

    Fusion product measurements planned for ITER are reviewed from the viewpoint of alpha particle-related physics studies. Recent advances in fusion plasma physics have extended the desirable measurement requirements to the megahertz region for neutron emission rate, better resolution of neutron profiles for the study of internal transport barriers (ITBs), etc. Employing threshold counters and/or scintillation detectors confers megahertz capability on neutron emission rate measurement. The changes in the neutron/alpha particle birth profile due to the formation of ITB and its deviation from uniformity on the magnetic flux surface can be measured by addition of eight viewing chords in an equatorial port plug and seven viewing chords from the divertor to the original radial neutron camera. On the other hand, it is still difficult to measure the distributions of confined and escaping alpha particles. Several proposals to resolve these difficulties are currently under investigation.

  6. Effect of Divertor Shaping on Divertor Plasma Behavior on DIII-D

    NASA Astrophysics Data System (ADS)

    Petrie, T. W.; Leonard, A. W.; Luce, T. C.; Mahdavi, M. A.; Holcomb, C. T.; Fenstermacher, M. E.; Hill, D. N.; Lasnier, C. J.; Watkins, J. G.; Moyer, R. A.; Stangeby, P. C.

    2012-10-01

    Recent experiments examined the dependence of divertor density (nTAR), temperature (TTAR), and heat flux at the outer divertor separatrix target on changes in the divertor separatrix geometry. The responses of nTAR and TTAR to changes in the parallel connection length in the scrape-off layer (SOL) (L||) are consistent with the predictions of the Two Point Model (TPM). However, nTAR and TTAR display a more complex response to changes in the radial location of the outer divertor strike point (RTAR) than expected based on the TPM. SOLPS transport analysis indicates that small differences in divertor geometry can change neutral trapping sufficient to explain differences between experiment and TPM predictions. The response of the core and divertor plasmas to changes in L|| and RTAR, under both radiating and non-radiating divertor conditions, will be shown.

  7. Moving Divertor Plates in a Tokamak

    SciTech Connect

    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.

  8. Nonlinear Impact of Edge Localized Modes on Carbon Erosion in the Divertor of the JET Tokamak

    SciTech Connect

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

    2009-01-30

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

  9. Rapidly Moving Divertor Plates In A Tokamak

    SciTech Connect

    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.

  10. Final Report on ITER Task Agreement 81-08

    SciTech Connect

    Richard L. Moore

    2008-03-01

    As part of an ITER Implementing Task Agreement (ITA) between the ITER US Participant Team (PT) and the ITER International Team (IT), the INL Fusion Safety Program was tasked to provide the ITER IT with upgrades to the fusion version of the MELCOR 1.8.5 code including a beryllium dust oxidation model. The purpose of this model is to allow the ITER IT to investigate hydrogen production from beryllium dust layers on hot surfaces inside the ITER vacuum vessel (VV) during in-vessel loss-of-cooling accidents (LOCAs). Also included in the ITER ITA was a task to construct a RELAP5/ATHENA model of the ITER divertor cooling loop to model the draining of the loop during a large ex-vessel pipe break followed by an in-vessel divertor break and compare the results to a simular MELCOR model developed by the ITER IT. This report, which is the final report for this agreement, documents the completion of the work scope under this ITER TA, designated as TA 81-08.

  11. Development of the ITER magnetic diagnostic set and specification

    SciTech Connect

    Vayakis, G.; Delhom, D.; Encheva, A.; Giacomin, T.; Jones, L.; Patel, K. M.; Portales, M.; Prieto, D.; Simrock, S.; Snipes, J. A.; Udintsev, V. S.; Watts, C.; Winter, A.; Zabeo, L.; Arshad, S.; Perez-Lasala, M.; Sartori, F.

    2012-10-15

    ITER magnetic diagnostics are now in their detailed design and R and D phase. They have passed their conceptual design reviews and a working diagnostic specification has been prepared aimed at the ITER project requirements. This paper highlights specific design progress, in particular, for the in-vessel coils, steady state sensors, saddle loops and divertor sensors. Key changes in the measurement specifications, and a working concept of software and electronics are also outlined.

  12. Impact of Resonant Magnetic Perturbation Fields on NSTX-U Advanced Divertor Topologies

    NASA Astrophysics Data System (ADS)

    Waters, Ian; Frerichs, Heinke; Schmitz, Oliver; Ahn, Joon-Wook; Canal, Gustavo; Evans, Todd; Soukhanovskii, Vlad

    2015-11-01

    Explorations are under way to optimize the magnetic topology in the plasma edge of NSTX-U with the goal of improving neutral and impurity fueling and exhaust. The use of magnetic perturbation fields is being considered to spread heat and particle fluxes in the divertor, adjust plasma refueling, control impurity transport, and improve coupling to the exhaust systems. Also, advanced divertor configurations are being considered to improve peak heat loads on divertors. An assessment is made of the topologies of a number of representative NSTX-U advanced divertor configurations: lower single null, exact snowflake, and snowflake minus. Wall to wall magnetic connection lengths for each configuration are assessed in both their perturbed and axisymmetric configurations with perturbation coil currents of 1kA and 3kA. The magnetic perturbations yield complex strike patterns on divertor elements that are expected to be resolvable experimentally. The EMC3-EIRENE fluid plasma and kinetic neutral transport code will be used to study neutral and impurity transport and exhaust in these topologies. This work was funded in part by the Department of Energy under grant DE-SC0012315 and by startup funds of the Department of Engineering Physics at the University of Wisconsin-Madison.

  13. ITER helium ash accumulation

    SciTech Connect

    Hogan, J.T.; Hillis, D.L.; Galambos, J.; Uckan, N.A. ); Dippel, K.H.; Finken, K.H. . Inst. fuer Plasmaphysik); Hulse, R.A.; Budny, R.V. . Plasma Physics Lab.)

    1990-01-01

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

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

    NASA Astrophysics Data System (ADS)

    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)

  15. Dust divertor for a tokamak fusion reactor

    SciTech Connect

    Tang, X Z; Delzanno, G L

    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.

  16. Rescheduling with iterative repair

    NASA Technical Reports Server (NTRS)

    Zweben, Monte; Davis, Eugene; Daun, Brian; Deale, Michael

    1992-01-01

    This paper presents a new approach to rescheduling called constraint-based iterative repair. This approach gives our system the ability to satisfy domain constraints, address optimization concerns, minimize perturbation to the original schedule, produce modified schedules, quickly, and exhibits 'anytime' behavior. The system begins with an initial, flawed schedule and then iteratively repairs constraint violations until a conflict-free schedule is produced. In an empirical demonstration, we vary the importance of minimizing perturbation and report how fast the system is able to resolve conflicts in a given time bound. We also show the anytime characteristics of the system. These experiments were performed within the domain of Space Shuttle ground processing.

  17. Rescheduling with iterative repair

    NASA Technical Reports Server (NTRS)

    Zweben, Monte; Davis, Eugene; Daun, Brian; Deale, Michael

    1992-01-01

    This paper presents a new approach to rescheduling called constraint-based iterative repair. This approach gives our system the ability to satisfy domain constraints, address optimization concerns, minimize perturbation to the original schedule, and produce modified schedules quickly. The system begins with an initial, flawed schedule and then iteratively repairs constraint violations until a conflict-free schedule is produced. In an empirical demonstration, we vary the importance of minimizing perturbation and report how fast the system is able to resolve conflicts in a given time bound. These experiments were performed within the domain of Space Shuttle ground processing.

  18. Stochasticity about a poloidal divertor separatrix

    SciTech Connect

    Skinner, D.A.; Osborne, T.H.; Prager, S.C.; Park, W.

    1986-10-01

    The stochasticization of the magnetic separatrix due to the presence of a helical perturbation in a poloidal divertor tokamak is illustrated by a numerical computation which traces magnetic field lines.

  19. Stochasticity about a poloidal divertor separatrix

    SciTech Connect

    Skinner, D.A.; Osborne, T.H.; Prager, S.C.; Park, W.

    1987-04-01

    The stochasticization of the magnetic separatrix caused by the presence of a helical perturbation in a poloidal divertor tokamak is illustrated by a numerical computation that traces magnetic field lines.

  20. High flux expansion divertor studies in NSTX

    SciTech Connect

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

    2009-06-29

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

  1. Parametric analysis of the thermal effects on the divertor in tokamaks during plasma disruptions

    SciTech Connect

    Bruhn, M.L.

    1988-04-01

    Plasma disruptions are an ever present danger to the plasma-facing components in today's tokamak fusion reactors. This threat results from our lack of understanding and limited ability to control this complex phenomenon. In particular, severe energy deposition occurs on the divertor component of the double-null configured tokamak reactor during such disruptions. A hybrid computational model developed to estimate and graphically illustrate global thermal effects of disruptions on the divertor plates is described in detail. The quasi-two-dimensional computer code, TADDPAK (Thermal Analysis Divertor during Disruptions PAcKage), is used to conduct parametric analysis for the TIBER II Tokamak Engineering Test Reactor Design. The dependence of these thermal effects on divertor material choice, disruption pulse length, disruption pulse shape, and the characteristic thickness of the plasma scrape-off layer is investigated for this reactor design. Results and conclusions from this analysis are presented. Improvements to this model and issues that require further investigation are discussed. Cursory analysis for ITER (International Thermonuclear Experimental Reactor) is also presented in the appendix. 75 refs., 49 figs., 10 tabs.

  2. Evaluation of performance for the EAST upgraded divertor targets during type I ELMy H-mode

    NASA Astrophysics Data System (ADS)

    Qian, X. Y.; Peng, X. B.; Wang, L.; Song, Y. T.; Ye, M. Y.; Zhang, J. W.; Li, W. X.; Zhu, C. C.

    2016-02-01

    The long-pulse high-confinement (H-mode) plasma regime is considered to be a preferable scenario in future fusion devices, and in the period of normal operation during H-mode, edge-localised modes (ELMs) are one of the most serious threats to the performance and capability of divertor targets. The EAST recently achieved a variety of H-mode regimes with ELMs. For the purpose of studying the performance of the EAST upgraded divertor during type I ELMs, a series of simulations were performed by using three-dimensional (3D) finite element code. To make a visible outcome of the direct ELM impact on the divertor targets, a preliminary evaluation system with three indices to exhibit the influence has been developed. The indices that comprise temperature evolution, thermal penetration depth and crack initiation life, which could reveal the process of micro-crack formation, are calculated in both low and high-power scenarios for type I ELMs. The initial results indicate that the transient heat load has a significant influence in a very short thickness layer along the direction perpendicular to the plasma-facing surface throughout its duration. The conclusion could offer a pertinent guide to the next-step high-power long-pulse operation in EAST and would also be helpful for scientifically studying the damage and fatigue mechanism of the divertor in ITER and future fusion power reactors.

  3. Dust in the divertor sheath: a problem or a possible solution to a problem?

    NASA Astrophysics Data System (ADS)

    Delzanno, Gian Luca; Tang, Xianzhu

    2012-03-01

    In this work, we will present results on dust transport in the magnetized sheath near the divertor plate for micron-sized dust. We consider conditions relevant to present short-pulse tokamak machines as well as conditions for long-pulse ITER/DEMO reactors. We solve the dust charging equation, the dust equation of motion and the equations for dust heating and mass loss in the magnetized sheath. We present parametric studies changing the divertor plasma conditions and the angle of the equilibrium magnetic field relative to the wall. Our main result is that, for conditions relavant to DEMO, the stronger heat flux to the wall severely limits the dust survivability and mobility. We discuss the implications of this result for the divertor plates of long-pulse fusion reactors. We will also discuss two fusion technology solutions to DEMO PFC, the dust patch and the dust shield, based on externally introduced solid particulates to patch areas of net erosion and to provide the primary heat exhaust for the divertor.

  4. ITER's woes

    NASA Astrophysics Data System (ADS)

    jjeherrera; Duffield, John; ZoloftNotWorking; esromac; protogonus; mleconte; cmfluteguy; adivita

    2014-07-01

    In reply to the physicsworld.com news story “US sanctions on Russia hit ITER council” (20 May, http://ow.ly/xF7oc and also June p8), about how a meeting of the fusion experiment's council had to be moved from St Petersburg and the US Congress's call for ITER boss Osamu Motojima to step down.

  5. Simulation of 3D effects on partially detached divertor conditions in NSTX and Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Lore, Jeremy

    2014-10-01

    Establishing a validated, predictive capability for the divertor plasma is critical for future fusion reactors, which must operate with detached divertors to reduce peak heat fluxes to the plasma facing components (PFC) and to mitigate net material erosion. This is challenging even for existing 2D codes, and is complicated further by non-axisymmetric effects due to divertor-localized gas injection, 3D magnetic fields, and 3D PFCs, modeling of which requires 3D simulations. New experiments performed on C-Mod at the request of the ITER organization to examine the consequence of localized nitrogen gas injection, show clear toroidal asymmetries in radiated power, impurity radiation, and divertor pressure. The 3D plasma/neutral transport code EMC3-EIRENE has been applied to model these experiments in the first attempt to benchmark the code against tokamak experimental data under detached conditions. The measured pressure modulation and the impurity radiation trends in the edge are qualitatively reproduced by the simulations, which also predict a ~2x modulation in heat flux at the outer strike point. Discrepancies are found in comparison with the measured private region radiation, and the simulations also indicate colder, denser divertor conditions than measured, suggesting that drifts and kinetic corrections may be required for more quantitative agreement. In separate experiments on NSTX, detached divertor plasmas are observed to reattach when 3D fields are applied. Modeling of the NSTX experiments reproduces these trends, with an increased peak heat flux with 3D fields and qualitative agreement of the striated flux patterns. The experimental identification of toroidal asymmetries in detached plasmas highlights the need for reliable 3D models for projecting the impact for ITER and beyond. Support from USDOE DE-AC05-00OR22725, DE-AC02-09CH11466, DE-FC02-99ER54512.

  6. Snowflake divertor configuration studies for NSTX-Upgrade

    SciTech Connect

    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.

  7. Fusion Specific Features in ITER Accident Analysis

    NASA Astrophysics Data System (ADS)

    Bartels, H.-W.; Gordon, C. W.; Piet, S. J.; Poucet, A. E.; Saji, G.; Topilski, L. N.

    1997-06-01

    Fusion specific features like inherent plasma shutdown, low decay heat densities, cryogenic temperatures, and limited source terms were considered during the safety design process of ITER. Uncertainties in plasma disruptions motivates a robust design to cope with multiple failures of in-vessel cooling piping. A vacuum vessel pressure suppression system mitigates pressure transients and effectively captures mobilized radioactivity. In case of pump trips or ex-vessel coolant losses in the divertor the plasma needs to be actively terminated in a few seconds. Failure to do so might damage the divertor but radiological consequences will be minor due to the intact first confinement barrier. Tritium plant inventories are protected by several layers of confinement. Uncontrolled release of magnet energy will be prevented by design. Postulated damage from magnets to confinement barriers causes fluid ingress (air, water, helium) into the cryostat. The cold environment limits pressurization. Most tritium and dust is captured by condensation.

  8. A "Snowflake" Divertor and its Properties

    SciTech Connect

    Ryutov, D

    2007-06-21

    Handling the power and particle exhaust in fusion reactors based on tokamaks is a challenging problem [1,2]. To bring the energy flux to the divertor plates to an acceptable level (< 10 MW/m2), it is desirable to significantly increase poloidal flux expansion in the divertor area. Some recent ideas include that of a so-called X divertor [3] and a 'snowflake' divertor [4]. We use an acronym SF to designate the latter. In this paper we concentrate on the SF divertor. The general idea behind this configuration is that, by a proper selection of divertor (poloidal field) coils, one can make the null point of the second, not of the first order as in the standard divertor. The separatrix in the vicinity of the X point then acquires a characteristic hexapole structure (Fig. 1), reminiscent of a snowflake, whence the name. The fact that the field has a second-order null, leads to a significant increase of the flux expansion. It was noted in Ref. [4] that the SF configuration is topologically unstable: if the current in the divertor coils is somewhat higher than the one that provides the SF configuration, it becomes a single-null X-point configuration. Conversely, if the coil current becomes somewhat lower, there appear two separate X-points. To solve this problem, one can operate the divertor at the current by roughly 5% higher than the value needed to create the second-order null. Then, configuration becomes robust enough and the shape of the separatrix does not change significantly if the coil current varies by 2-3%. At the same time, the flux expansion still remained by a factor of {approx}3 larger compared to a 'canonical' divertor. Following Ref. [4], we call this configuration a 'SF-plus' configuration. Specific examples in Ref. [4] were given for simple magnetic geometries The aim of this paper is to demonstrate that the SF concept will also work for a strongly shaped plasma. The other set of issues considered in the present paper relates to the possible presence of

  9. The ITER in-vessel system

    SciTech Connect

    Lousteau, D.C.

    1994-09-01

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

  10. ITER Experts' meeting on density limits

    SciTech Connect

    Borrass, K.; Igitkhanov, Y.L.; Uckan, N.A.

    1989-12-01

    The necessity of achieving a prescribed wall load or fusion power essentially determines the plasma pressure in a device like ITER. The range of operation densities and temperatures compatible with this condition is constrained by the problems of power exhaust and the disruptive density limit. The maximum allowable heat loads on the divertor plates and the maximum allowable sheath edge temperature practically impose a lower limit on the operating densities, whereas the disruptive density limit imposes an upper limit. For most of the density limit scalings proposed in the past an overlap of the two constraints or at best a very narrow accessible density range is predicted for ITER. Improved understanding of the underlying mechanisms is therefore a crucial issue in order to provide a more reliable basis for extrapolation to ITER and to identify possible ways of alleviating the problem.

  11. Optical dumps for H-alpha and visible spectroscopy in ITER

    SciTech Connect

    Andreenko, E. N.; Alekseev, A. G.; Gorshkov, A. V.; Orlovskiy, I. I.

    2014-08-21

    High-reflective Beryllium cover of ITER first wall (R≈30–60%) causes remarkable increase of divertor stray light component (DSL). Optical dumps are well-known solution for DSL attenuation. In this work few types of optical dumps have been examined both by modeling and experimental studies. Taking into account the limitations, induced by ITER first wall design, OD optimized design has been proposed which could decrease divertor stray light component by 10..100 times depending on incidence angle of light.

  12. Liquid metal cooled divertor for ARIES

    SciTech Connect

    Muraviev, E.

    1995-01-01

    A liquid metal, Ga-cooled divertor design was completed for the double null ARIES-II divertor design. The design analysis indicated a surface heat flux removal capability of up to 15 MW/m{sup 2}, and its relative easy maintenance. Design issues of configuration, thermal hydraulics, thermal stresses, liquid metal loop and safety effects were evaluated. For coolant flow control, it was found that it is necessary to use some part of the blanket cooling ducts for the draining of liquid metal from the top divertor. In order to minimize the inventory of Ga, it was recommended that the liquid metal loop equipment should be located as close to the torus as possible. More detailed analysis of transient conditions especially under accident conditions was identified as an issue that will need to be addressed.

  13. Designing divertor targets for uniform power load

    NASA Astrophysics Data System (ADS)

    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.

  14. Synchronized multiartifact reduction with tomographic reconstruction (SMART-RECON): A statistical model based iterative image reconstruction method to eliminate limited-view artifacts and to mitigate the temporal-average artifacts in time-resolved CT

    PubMed Central

    Chen, Guang-Hong; Li, Yinsheng

    2015-01-01

    Purpose: In x-ray computed tomography (CT), a violation of the Tuy data sufficiency condition leads to limited-view artifacts. In some applications, it is desirable to use data corresponding to a narrow temporal window to reconstruct images with reduced temporal-average artifacts. However, the need to reduce temporal-average artifacts in practice may result in a violation of the Tuy condition and thus undesirable limited-view artifacts. In this paper, the authors present a new iterative reconstruction method, synchronized multiartifact reduction with tomographic reconstruction (SMART-RECON), to eliminate limited-view artifacts using data acquired within an ultranarrow temporal window that severely violates the Tuy condition. Methods: In time-resolved contrast enhanced CT acquisitions, image contrast dynamically changes during data acquisition. Each image reconstructed from data acquired in a given temporal window represents one time frame and can be denoted as an image vector. Conventionally, each individual time frame is reconstructed independently. In this paper, all image frames are grouped into a spatial–temporal image matrix and are reconstructed together. Rather than the spatial and/or temporal smoothing regularizers commonly used in iterative image reconstruction, the nuclear norm of the spatial–temporal image matrix is used in SMART-RECON to regularize the reconstruction of all image time frames. This regularizer exploits the low-dimensional structure of the spatial–temporal image matrix to mitigate limited-view artifacts when an ultranarrow temporal window is desired in some applications to reduce temporal-average artifacts. Both numerical simulations in two dimensional image slices with known ground truth and in vivo human subject data acquired in a contrast enhanced cone beam CT exam have been used to validate the proposed SMART-RECON algorithm and to demonstrate the initial performance of the algorithm. Reconstruction errors and temporal fidelity

  15. Utilization of vanadium alloys in the DIII-D Radiative Divertor Program

    SciTech Connect

    Smith, J.P.; Johnson, W.R.; Stambaugh, R.D.; Trester, P.W.; Smith, D.; Bloom, E.

    1995-10-01

    Vanadium alloys are attractive candidate structural materials for fusion power plants because of their potential for minimum environmental impact due to low neutron activation and rapid activation decay. They also possess favorable material properties for operation in a fusion environment. General Atomics (GA), in conjunction with Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL), has developed a plan for the utilization of vanadium alloys as part of the Radiative Divertor (RD) upgrade for the DIII-D tokamak. The plan will be carried out in conjunction with General Atomics and the Materials Program of the US Department of Energy (DOE). This application of a vanadium alloy will provide a meaningful step in the development of advanced materials for fusion power devices by: (1) developing necessary materials processing technology for the fabrication of large vanadium alloy components, and (2) demonstrating the in-service behavior of a vanadium alloy (V-4Cr-4Ti) in a tokamak environment. The program consists of three phases: first, small vanadium alloy coupon samples will be exposed in DIII-D at positions in the vessel floor and within the pumping plenum region of the existing divertor structure; second, a small vanadium alloy component will be installed in the existing divertor, and third, during the forthcoming Radiative Divertor modification, scheduled for completion in mid-1997, the upper section of the new double-null, slotted divertor will be fabricated from vanadium alloy product forms. This program also includes research and development (R and D) efforts to support fabrication development and to resolve key issues related to environmental effects.

  16. Heat Load on Divertors in NCSX

    NASA Astrophysics Data System (ADS)

    Kaiser, T. B.; Hill, D. N.; Maingi, R.; Monticello, D.; Zarnstorff, M.; Grossman, A.

    2006-10-01

    We have continued our study[1-3] of the effect of divertors in NCSX, using magnetic field data generated by both the PIES and VMEC/MFBE equilibrium codes. Results for comparable equilibria from the two codes agree to within statistical uncertainty. We follow field lines from a surface just outside and conformal with the LCMS until they strike a divertor plate or the first wall, or exceed 1000m in length, with effects of particle scattering mimicked by field-line diffusion. Current candidate divertor designs efficiently collect field lines, allowing fewer than 0.1% to reach the wall. The sensitivity of localized power deposition, assumed to be proportional to the density of field- line strike-points, to adjustments in the divertor configuration is under investigation.1. T.B. Kaiser, et al, Bull. Am. Phys. Soc., 48, paper RP1-20, 2003.2. T.B. Kaiser, et al, Bull. Am. Phys. Soc., 49, paper PP1-73, 2004.3. R. Maingi, et al, EPS Conf. Rome, Italy, paper P5.116, 2006.

  17. Divertor target for magnetic containment device

    DOEpatents

    Luzzi, Jr., Theodore E.

    1982-01-01

    In a plasma containment device of a type having superconducting field coils for magnetically shaping the plasma into approximately the form of a torus, an improved divertor target for removing impurities from a "scrape off" region of the plasma comprises an array of water cooled swirl tubes onto which the scrape off flux is impinged. Impurities reflected from the divertor target are removed from the target region by a conventional vacuum getter system. The swirl tubes are oriented and spaced apart within the divertor region relative to the incident angle of the scrape off flux to cause only one side of each tube to be exposed to the flux to increase the burnout rating of the target. The divertor target plane is oriented relative to the plane of the path of the scrape off flux such that the maximum heat flux onto a swirl tube is less than the tube design flux. The containment device is used to contain the plasma of a tokamak fusion reactor and is applicable to other long pulse plasma containment systems.

  18. An X-point ergodic divertor

    SciTech Connect

    Chu, M.S.; Jensen, T.H.; La Haye, R.J.; Taylor, T.S.; Evans, T.E.

    1991-10-01

    A new ergodic divertor is proposed. It utilizes a system of external (n = 3) coils arranged to generate overlapping magnetic islands in the edge region of a diverted tokamak and connect the randomized field lines to the external (cold) divertor plate. The novel feature in the configuration is the placement of the external coils close to the X-point. A realistic design of the external coil set is studied by using the field line tracing method for a low aspect ratio (A {approx equal} 3) tokamak. Two types of effects are observed. First, by placing the coils close to the X-point, where the poloidal magnetic field is weak and the rational surfaces are closely packed only a moderate amount of current in the external coils is needed to ergodize the edge region. This ergodized edge enhances the edge transport in the X-point region and leads to the potential of edge profile control and the avoidance of edge localized modes (ELMs). Furthermore, the trajectories of the field lines close to the X-point are modified by the external coil set, causing the hit points on the external divertor plates to be randomized and spread out in the major radius direction. A time-dependent modulation of the currents in the external (n = 3) coils can potentially spread the heat flux more uniformly on the divertor plate avoiding high concentration of the heat flux. 10 refs., 9 figs.

  19. The tungsten divertor experiment at ASDEX Upgrade

    NASA Astrophysics Data System (ADS)

    Neu, R.; Asmussen, K.; Krieger, K.; Thoma, A.; Bosch, H.-S.; Deschka, S.; Dux, R.; Engelhardt, W.; García-Rosales, C.; Gruber, O.; Herrmann, A.; Kallenbach, A.; Kaufmann, M.; Mertens, V.; Ryter, F.; Rohde, V.; Roth, J.; Sokoll, M.; Stäbler, A.; Suttrop, W.; Weinlich, M.; Zohm, H.; Alexander, M.; Becker, G.; Behler, K.; Behringer, K.; Behrisch, R.; Bergmann, A.; Bessenrodt-Weberpals, M.; Brambilla, M.; Brinkschulte, H.; Büchl, K.; Carlson, A.; Chodura, R.; Coster, D.; Cupido, L.; de Blank, H. J.; de Peña Hempel, S.; Drube, R.; Fahrbach, H.-U.; Feist, J.-H.; Feneberg, W.; Fiedler, S.; Franzen, P.; Fuchs, J. C.; Fußmann, G.; Gafert, J.; Gehre, O.; Gernhardt, J.; Haas, G.; Herppich, G.; Herrmann, W.; Hirsch, S.; Hoek, M.; Hoenen, F.; Hofmeister, F.; Hohenöcker, H.; Jacobi, D.; Junker, W.; Kardaun, O.; Kass, T.; Kollotzek, H.; Köppendörfer, W.; Kurzan, B.; Lackner, K.; Lang, P. T.; Lang, R. S.; Laux, M.; Lengyel, L. L.; Leuterer, F.; Manso, M. E.; Maraschek, M.; Mast, K.-F.; McCarthy, P.; Meisel, D.; Merkel, R.; Müller, H. W.; Münich, M.; Murmann, H.; Napiontek, B.; Neu, G.; Neuhauser, J.; Niethammer, M.; Noterdaeme, J.-M.; Pasch, E.; Pautasso, G.; Peeters, A. G.; Pereverzev, G.; Pitcher, C. S.; Poschenrieder, W.; Raupp, G.; Reinmüller, K.; Riedl, R.; Röhr, H.; Salzmann, H.; Sandmann, W.; Schilling, H.-B.; Schlögl, D.; Schneider, H.; Schneider, R.; Schneider, W.; Schramm, G.; Schweinzer, J.; Scott, B. D.; Seidel, U.; Serra, F.; Speth, E.; Silva, A.; Steuer, K.-H.; Stober, J.; Streibl, B.; Treutterer, W.; Troppmann, M.; Tsois, N.; Ulrich, M.; Varela, P.; Verbeek, H.; Verplancke, Ph; Vollmer, O.; Wedler, H.; Wenzel, U.; Wesner, F.; Wolf, R.; Wunderlich, R.; Zasche, D.; Zehetbauer, T.; Zehrfeld, H.-P.

    1996-12-01

    Tungsten-coated tiles, manufactured by plasma spray on graphite, were mounted in the divertor of the ASDEX Upgrade tokamak and cover almost 90% of the surface facing the plasma in the strike zone. Over 600 plasma discharges have been performed to date, around 300 of which were auxiliary heated with heating powers up to 10 MW. The production of tungsten in the divertor was monitored by a W I line at 400.8 nm. In the plasma centre an array of spectral lines at 5 nm emitted by ionization states around W XXX was measured. From the intensity of these lines the W content was derived. Under normal discharge conditions W-concentrations around 0741-3335/38/12A/013/img12 or even lower were found. The influence on the main plasma parameters was found to be negligible. The maximum concentrations observed decrease with increasing heating power. In several low power discharges accumulation of tungsten occurred and the temperature profile was flattened. The concentrations of the intrinsic impurities carbon and oxygen were comparable to the discharges with the graphite divertor. Furthermore, the density and the 0741-3335/38/12A/013/img13 limits remained unchanged and no negative influence on the energy confinement or on the H-mode threshold was found. Discharges with neon radiative cooling showed the same behaviour as in the graphite divertor case.

  20. Divertor Coil Design and Implementation on Pegasus

    NASA Astrophysics Data System (ADS)

    Shriwise, P. C.; Bongard, M. W.; Cole, J. A.; Fonck, R. J.; Kujak-Ford, B. A.; Lewicki, B. T.; Winz, G. R.

    2012-10-01

    An upgraded divertor coil system is being commissioned on the Pegasus Toroidal Experiment in conjunction with power system upgrades in order to achieve higher β plasmas, reduce impurities, and possibly achieve H-mode operation. Design points for the divertor coil locations and estimates of their necessary current ratings were found using predictive equilibrium modeling based upon a 300 kA target plasma. This modeling represented existing Pegasus coil locations and current drive limits. The resultant design calls for 125 kA-turns from the divertor system to support the creation of a double null magnetic topology in plasmas with Ip<=300 kA. Initial experiments using this system will employ 900 V IGBT power supply modules to provide IDIV<=4 kA. The resulting 20 kA-turn capability of the existing divertor coil will be augmented by a new coil providing additional A-turns in series. Induced vessel wall current modeling indicates the time response of a 28 turn augmentation coil remains fast compared to the poloidal field penetration rate through the vessel. First results operating the augmented system are shown.

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

    SciTech Connect

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

    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.

  2. Recent results from tokamak divertor plasma measurements

    SciTech Connect

    Allen, S.L.

    1996-05-01

    New diagnostics have been developed to address key divertor physics questions, including: target plate heat flux reduction by radiation, basic edge transport issues, and plasma wall interactions (PWI) such as erosion. A system of diagnostics measures the target plate heat flux (imaging IR thermography) and particle flux (probes, pressure and Penning gauges, and visible emission arrays). Recently, T{sub e},n{sub e}, and P{sub e} (electron pressure) have been measured in 2-D with divertor Thomson Scattering. During radiative divertor operation T{sub e} is less than 2 eV, indicating that new atomic processes are important. Langmuir probes measure higher T{sub e} in some cases. In addition, the measured P{sub e} near the separatrix at the target plate is lower than the midplane pressure, implying radial momentum transport. Bolometer arrays, inverted with reconstruction algorithms, provide the 2-D core and divertor radiation profiles. Spectroscopic measurements identify the radiating species and provide information on impurity transport; both absolute chordal measurements and tomographic reconstructions of images are used. Either intrinsic carbon or an inert species (e.g., injected Ne) are usually observed, and absolute particle inventories are obtained. Computer codes are both benchmarked with the experimental data and provide important consistency checks. Several techniques are used to measure fundamental plasma transport and fluctuations, including probes and reflectometry. PWI issues are studied with in-situ coupons and insertable samples (DiMES). Representative divertor results from DIII-D with references to results on other tokamaks will be presented.

  3. Findings of pre-ELM structures through the observation of divertor heat load patterns at JET with applied n = 2 perturbation fields

    NASA Astrophysics Data System (ADS)

    Rack, M.; Sieglin, B.; Eich, T.; Pearson, J.; Liang, Y.; Balboa, I.; Jachmich, S.; Wingen, A.; Pamela, S. J. P.; EFDA Contributors, JET

    2014-07-01

    Resonant magnetic perturbation experiments at JET with the ITER-like wall have shown the formation of radially propagating pre-ELM structures in the heat flux profile on the outer divertor. These appear a few milliseconds before the major divertor heat load, caused by type-I edge-localized modes (ELMs). The formation of the pre-ELM structures is accompanied by an increase in the Dα emission. For some pronounced examples, the propagation appears to end at the positions where an increased heat load is seen during the ELM crash a few milliseconds later. These observations are presented and discussed along with a comparison of a thermoelectric edge currents model.

  4. Analysis of a multi-machine database on divertor heat fluxesa)

    NASA Astrophysics Data System (ADS)

    Makowski, M. A.; Elder, D.; Gray, T. K.; LaBombard, B.; Lasnier, C. J.; Leonard, A. W.; Maingi, R.; Osborne, T. H.; Stangeby, P. C.; Terry, J. L.; Watkins, J.

    2012-05-01

    A coordinated effort to measure divertor heat flux characteristics in fully attached, similarly shaped H-mode plasmas on C-Mod, DIII-D, and NSTX was carried out in 2010 in order to construct a predictive scaling relation applicable to next step devices including ITER, FNSF, and DEMO. Few published scaling laws are available and those that have been published were obtained under widely varying conditions and divertor geometries, leading to conflicting predictions for this critically important quantity. This study was designed to overcome these deficiencies. Analysis of the combined data set reveals that the primary dependence of the parallel heat flux width is robustly inverse with Ip, which all three tokamaks independently demonstrate. An improved Thomson scattering system on DIII-D has yielded very accurate scrape off layer (SOL) profile measurements from which tests of parallel transport models have been made. It is found that a flux-limited model agrees best with the data at all collisionalities, while a Spitzer resistivity model agrees at higher collisionality where it is more valid. The SOL profile measurements and divertor heat flux scaling are consistent with a heuristic drift based model as well as a critical gradient model.

  5. The role of plasma response in divertor footprint modification by 3D fields in NSTX

    NASA Astrophysics Data System (ADS)

    Ahn, Joonwook; Kim, Kimin; Canal, Gustavo; Gan, Kaifu; Gray, Travis; McLean, Adam; Park, Jong-Kyu; Scotti, Filippo

    2015-11-01

    In NSTX, the divertor footprints of both heat and particle fluxes are found to be significantly modified by externally applied 3D magnetic perturbations. Striations on the divertor surface, indicating separatrix splitting and formation of magnetic lobes, are observed for both n = 1 and n = 3 perturbation fields. These striations can lead to localized heating of the divertor plates and to the re-attachment of detached plasmas, both of which have to be avoided in ITER for successful heat flux management. In this work, the role of plasma response on the formation of separatrix splitting has been investigated in the ideal framework by comparing measured heat and particle flux footprints with field line tracing calculations with and without contributions from the plasma response calculated by the ideal code IPEC. Simulations show that, n = 3 fields are slightly shielded by the plasma, with the measured helical pattern of striations in good agreement with the results from the vacuum approximation. The n = 1 fields are, however, significantly amplified by the plasma response, which provides a better agreement with the measurements. Resistive plasma response calculations by M3D-C1 are also in progress and the results will be compared with those from the ideal code IPEC. This work was supported by DoE Contracts: DE-AC05-00OR22725, DE-AC52-07NA27344 and DE-AC02-09CH11466.

  6. Gyrokinetic study of edge blobs and divertor heat-load footprint

    NASA Astrophysics Data System (ADS)

    Chang, C.-S.; Ku, S.-H.; Churchill, M.; Zweben, S.

    2014-10-01

    In an attempt to better understand the complicated physics of the inter-related ``intermittent plasma objects (blobs)'' and divertor heat-load footprint, the full-function gyrokinetic PIC code XGC1 has been used in realistic diverted geometry. Neoclassical and turbulence physics are simulated together self-consistently in the presence of Monte Carlo neutral particles. Blobs are modeled here as electrostatic nonlinear turbulence phenomenon. It is found that the ``blobs'' are generated, together with the ``holes,'' around the steep density gradient region. XGC1 reasserts the previous findings that blobs move out convectively into the scrape-off layer, while the holes move inward toward plasma core. The measured radial width of the divertor heat load, mapped to the outer midplane, is found to be much less than the median radial size of the intermittent plasma objects, but is rather closer to the width of neoclassical orbit excursion from pedestal to divertor, yielding approximately the 1/Ip-type scaling found from our previous pure neoclassical simulation or a heuristic neoclassical argument by Goldston. However, it also shows some spreading by the intermittent turbulence. In ITER plasma edge, where the ion banana width at separatrix becomes negligibly small compared to the meso-scale blob size, blobs may saturate the 1/Ip scaling.

  7. Suppression of Tritium Retention in Remote Areas of ITER by Nonperturbative Reactive Gas Injection

    SciTech Connect

    Tabares, F. L.; Ferreira, J. A.; Ramos, A.; Rooij, G. van; Westerhout, J.; Al, R.; Rapp, J.; Drenik, A.; Mozetic, M.

    2010-10-22

    A technique based on reactive gas injection in the afterglow region of the divertor plasma is proposed for the suppression of tritium-carbon codeposits in remote areas of ITER when operated with carbon-based divertor targets. Experiments in a divertor simulator plasma device indicate that a 4 nm/min deposition can be suppressed by addition of 1 Pa{center_dot}m{sup 3} s{sup -1} ammonia flow at 10 cm from the plasma. These results bolster the concept of nonperturbative scavenger injection for tritium inventory control in carbon-based fusion plasma devices, thus paving the way for ITER operation in the active phase under a carbon-dominated, plasma facing component background.

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

    NASA Astrophysics Data System (ADS)

    McLean, Adam Gordon

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

  9. Safety characteristics of options for plasma-facing components for ITER and beyond

    SciTech Connect

    Piet, S.J.; McCarthy, K.A.; Holland, D.F.; Longhurst, G.R.; Merrill, B.J.

    1991-01-01

    Plasma-facing components (PFC) likely dominate the safety hazards of the International Thermonuclear Experimental Reactor (ITER) and post-ITER machines. To gain regulatory approval and for fusion energy to fulfill its ultimate attractive safety and environmental potential, safety must be considered when selecting among PFC options. This paper summarizes current PFC safety information. PFC safety issues fall into seven areas: disruption tolerance, disruption severity, tritium inventory and permeation, accidental energy release, activation/toxin hazards, cooling disturbances, and system issues. RFC options include current ITER mainline options (Be or W coating, C tiles), variants on current ITER options, and liquid metal (LM) divertors. No PFC option that we have examined is free of critical safety concerns. There are also innovative ideas that may improve any PFC's performance -- super-permeable vacuum ducts, helium self-pumping, and gaseous divertors. We conclude with recommendations and a future strategy. 17 refs., 1 fig., 3 tabs.

  10. Ion target impact energy during Type I edge localized modes in JET ITER-like Wall

    NASA Astrophysics Data System (ADS)

    Guillemaut, C.; Jardin, A.; Horacek, J.; Autricque, A.; Arnoux, G.; Boom, J.; Brezinsek, S.; Coenen, J. W.; De La Luna, E.; Devaux, S.; Eich, T.; Giroud, C.; Harting, D.; Kirschner, A.; Lipschultz, B.; Matthews, G. F.; Moulton, D.; O'Mullane, M.; Stamp, M.

    2015-08-01

    The ITER baseline scenario, with 500 MW of DT fusion power and Q = 10, will rely on a Type I ELMy H-mode, with ΔW = 0.7 MJ mitigated edge localized modes (ELMs). Tungsten (W) is the material now decided for the divertor plasma-facing components from the start of plasma operations. W atoms sputtered from divertor targets during ELMs are expected to be the dominant source under the partially detached divertor conditions required for safe ITER operation. W impurity concentration in the plasma core can dramatically degrade its performance and lead to potentially damaging disruptions. Understanding the physics of plasma-wall interaction during ELMs is important and a primary input for this is the energy of incoming ions during an ELM event. In this paper, coupled Infrared thermography and Langmuir Probe (LP) measurements in JET-ITER-Like-Wall unseeded H-mode experiments with ITER relevant ELM energy drop have been used to estimate the impact energy of deuterium ions (D+) on the divertor target. This analysis gives an ion energy of several keV during ELMs, which makes D+ responsible for most of the W sputtering in unseeded H-mode discharges. These LP measurements were possible because of the low electron temperature (Te) during ELMs which allowed saturation of the ion current. Although at first sight surprising, the observation of low Te at the divertor target during ELMs is consistent with the ‘Free-Streaming’ kinetic model which predicts a near-complete transfer of parallel energy from electrons to ions in order to maintain quasi-neutrality of the ELM filaments while they are transported to the divertor targets.

  11. The effect of L mode filaments on divertor heat flux profiles as measured by infrared thermography on MAST

    NASA Astrophysics Data System (ADS)

    Thornton, A. J.; Fishpool, G.; Kirk, A.; the MAST Team; the EUROfusion MST1 Team

    2015-11-01

    Filamentary transport across the scrape off layer is a key issue for the design and operation of future devices, such as ITER, DEMO and MAST-U, as it sets the power loadings to the divertor and first wall of the machine. Analysis has been performed on L mode filaments in MAST in order to gain an understanding of the spatial structure and attempt to reconcile the different scales of the filament width and the power fall off length ({λq} ). The L mode filament heat flux arriving at the divertor has been measured using high spatial resolution (1.5 mm) infrared (IR) thermography. The filaments form discrete spiral patterns at the divertor which can be seen as bands of increased heat flux in the IR measurements. Analysis of the width and spacing of these bands at the divertor has allowed the toroidal mode number of the filaments to be determined (7≤slant n≤slant 22 ). The size of the filaments at the midplane has been determined using the target filament radial width and the magnetic field geometry. The filament width perpendicular to the magnetic field at the midplane has been found to be between 3 and 5 cm. Direct calculation of the filament width from midplane visible imaging gives a range of 4-6 cm which agrees well with the IR data.

  12. Power Radiated from ITER and CIT by Impurities

    DOE R&D Accomplishments Database

    Cummings, J.; Cohen, S. A.; Hulse, R.; Post, D. E.; Redi, M. H.; Perkins, J.

    1990-07-01

    The MIST code has been used to model impurity radiation from the edge and core plasmas in ITER and CIT. A broad range of parameters have been varied, including Z{sub eff}, impurity species, impurity transport coefficients, and plasma temperature and density profiles, especially at the edge. For a set of these parameters representative of the baseline ITER ignition scenario, it is seen that impurity radiation, which is produced in roughly equal amounts by the edge and core regions, can make a major improvement in divertor operation without compromising core energy confinement. Scalings of impurity radiation with atomic number and machine size are also discussed.

  13. Novel aspects of plasma control in ITER

    SciTech Connect

    Humphreys, D.; Jackson, G.; Walker, M.; Welander, A.; Ambrosino, G.; Pironti, A.; Felici, F.; Kallenbach, A.; Raupp, G.; Treutterer, W.; Kolemen, E.; Lister, J.; Sauter, O.; Moreau, D.; Schuster, E.

    2015-02-15

    ITER plasma control design solutions and performance requirements are strongly driven by its nuclear mission, aggressive commissioning constraints, and limited number of operational discharges. In addition, high plasma energy content, heat fluxes, neutron fluxes, and very long pulse operation place novel demands on control performance in many areas ranging from plasma boundary and divertor regulation to plasma kinetics and stability control. Both commissioning and experimental operations schedules provide limited time for tuning of control algorithms relative to operating devices. Although many aspects of the control solutions required by ITER have been well-demonstrated in present devices and even designed satisfactorily for ITER application, many elements unique to ITER including various crucial integration issues are presently under development. We describe selected novel aspects of plasma control in ITER, identifying unique parts of the control problem and highlighting some key areas of research remaining. Novel control areas described include control physics understanding (e.g., current profile regulation, tearing mode (TM) suppression), control mathematics (e.g., algorithmic and simulation approaches to high confidence robust performance), and integration solutions (e.g., methods for management of highly subscribed control resources). We identify unique aspects of the ITER TM suppression scheme, which will pulse gyrotrons to drive current within a magnetic island, and turn the drive off following suppression in order to minimize use of auxiliary power and maximize fusion gain. The potential role of active current profile control and approaches to design in ITER are discussed. Issues and approaches to fault handling algorithms are described, along with novel aspects of actuator sharing in ITER.

  14. Novel aspects of plasma control in ITER

    NASA Astrophysics Data System (ADS)

    Humphreys, D.; Ambrosino, G.; de Vries, P.; Felici, F.; Kim, S. H.; Jackson, G.; Kallenbach, A.; Kolemen, E.; Lister, J.; Moreau, D.; Pironti, A.; Raupp, G.; Sauter, O.; Schuster, E.; Snipes, J.; Treutterer, W.; Walker, M.; Welander, A.; Winter, A.; Zabeo, L.

    2015-02-01

    ITER plasma control design solutions and performance requirements are strongly driven by its nuclear mission, aggressive commissioning constraints, and limited number of operational discharges. In addition, high plasma energy content, heat fluxes, neutron fluxes, and very long pulse operation place novel demands on control performance in many areas ranging from plasma boundary and divertor regulation to plasma kinetics and stability control. Both commissioning and experimental operations schedules provide limited time for tuning of control algorithms relative to operating devices. Although many aspects of the control solutions required by ITER have been well-demonstrated in present devices and even designed satisfactorily for ITER application, many elements unique to ITER including various crucial integration issues are presently under development. We describe selected novel aspects of plasma control in ITER, identifying unique parts of the control problem and highlighting some key areas of research remaining. Novel control areas described include control physics understanding (e.g., current profile regulation, tearing mode (TM) suppression), control mathematics (e.g., algorithmic and simulation approaches to high confidence robust performance), and integration solutions (e.g., methods for management of highly subscribed control resources). We identify unique aspects of the ITER TM suppression scheme, which will pulse gyrotrons to drive current within a magnetic island, and turn the drive off following suppression in order to minimize use of auxiliary power and maximize fusion gain. The potential role of active current profile control and approaches to design in ITER are discussed. Issues and approaches to fault handling algorithms are described, along with novel aspects of actuator sharing in ITER.

  15. Recent ASDEX Upgrade research in support of ITER and DEMO

    NASA Astrophysics Data System (ADS)

    H. Zohmthe ASDEX Upgrade Team; the EUROfusion MST1 Team

    2015-10-01

    Recent experiments on the ASDEX Upgrade tokamak aim at improving the physics base for ITER and DEMO to aid the machine design and prepare efficient operation. Type I edge localized mode (ELM) mitigation using resonant magnetic perturbations (RMPs) has been shown at low pedestal collisionality (νped\\ast <0.4) . In contrast to the previous high ν* regime, suppression only occurs in a narrow RMP spectral window, indicating a resonant process, and a concomitant confinement drop is observed due to a reduction of pedestal top density and electron temperature. Strong evidence is found for the ion heat flux to be the decisive element for the L-H power threshold. A physics based scaling of the density at which the minimum PLH occurs indicates that ITER could take advantage of it to initiate H-mode at lower density than that of the final Q = 10 operational point. Core density fluctuation measurements resolved in radius and wave number show that an increase of R/LTe introduced by off-axis electron cyclotron resonance heating (ECRH) mainly increases the large scale fluctuations. The radial variation of the fluctuation level is in agreement with simulations using the GENE code. Fast particles are shown to undergo classical slowing down in the absence of large scale magnetohydrodynamic (MHD) events and for low heating power, but show signs of anomalous radial redistribution at large heating power, consistent with a broadened off-axis neutral beam current drive current profile under these conditions. Neoclassical tearing mode (NTM) suppression experiments using electron cyclotron current drive (ECCD) with feedback controlled deposition have allowed to test several control strategies for ITER, including automated control of (3,2) and (2,1) NTMs during a single discharge. Disruption mitigation studies using massive gas injection (MGI) can show an increased fuelling efficiency with high field side injection, but a saturation of the fuelling efficiency is observed at high injected

  16. Main challenges for ITER optical diagnostics

    NASA Astrophysics Data System (ADS)

    Vukolov, K. Yu.; Orlovskiy, I. I.; Alekseev, A. G.; Borisov, A. A.; Andreenko, E. N.; Kukushkin, A. B.; Lisitsa, V. S.; Neverov, V. S.

    2014-08-01

    The review is made of the problems of ITER optical diagnostics. Most of these problems will be related to the intensive neutron radiation from hot plasma. At a high level of radiation loads the most types of materials gradually change their properties. This effect is most critical for optical diagnostics because of degradation of optical glasses and mirrors. The degradation of mirrors, that collect the light from plasma, basically will be induced by impurity deposition and (or) sputtering by charge exchange atoms. Main attention is paid to the search of glasses for vacuum windows and achromatic lens which are stable under ITER irradiation conditions. The last results of irradiation tests in nuclear reactor of candidate silica glasses KU-1, KS-4V and TF 200 are presented. An additional problem is discussed that deals with the stray light produced by multiple reflections from the first wall of the intense light emitted in the divertor plasma.

  17. Performance of the INTOR poloidal divertor

    SciTech Connect

    Post, D.E.; Petravic, M.; Schmidt, J.A.; Heifetz, D.

    1981-10-01

    The next generation of large tokamak experiments is expected to have large particle and heat outfluxes (approx. 10/sup 23/ particles/sec and 80 MW). These outfluxes must be controlled to provide adequate pumping of the helium ash and to minimize the sputtering erosion of the vacuum vessel walls, limiters, and neutralizer plates. A poloidal divertor design to solve these problems for INTOR has been done using a two-dimensional code which models the plasma as a fluid and solves equations for the flow of particles, momentum and energy, and calculates the neutral gas transport with Monte-Carlo techniques. These calculations show that there is a regime of operation where the density in the divertor is high and the temperature is low, thus easing the heat load and erosion problems. The neutral pressure at the plate is high, resulting in high gas throughputs, with modest pumping speeds.

  18. Electron beam facility for divertor target experiments

    SciTech Connect

    Anisimov, A.; Gagen-Torn, V.; Giniyatulin, R.N.

    1994-12-31

    To test different concepts of divertor targets and bumpers an electron beam facility was assembled in Efremov Institute. It consists of a vacuum chamber (3m{sup 3}), vacuum pump, electron beam gun, manipulator to place and remove the samples, water loop and liquid metal loop. The following diagnostics of mock-ups is stipulated: (1) temperature distribution on the mock-up working surface (scanning pyrometer and infra-red imager); (2) temperature distribution over mocked-up thickness in 3 typical cross-sections (thermo-couples); (3) cracking dynamics during thermal cycling (acoustic-emission method), (4) defects in the mock-up before and after tests (ultra-sonic diagnostics, electron and optical microscopes). Carbon-based and beryllium mock-ups are made for experimental feasibility study of water and liquid-metal-cooled divertor/bumper concepts.

  19. NSTX Plasma Response to Lithium Coated Divertor

    SciTech Connect

    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.

  20. Modeling detachment physics in the NSTX snowflake divertor

    NASA Astrophysics Data System (ADS)

    Meier, E. T.; Soukhanovskii, V. A.; Bell, R. E.; Diallo, A.; Kaita, R.; LeBlanc, B. P.; McLean, A. G.; Podestà, M.; Rognlien, T. D.; Scotti, F.

    2015-08-01

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

  1. Flute mode fluctuations in the divertor mirror cell

    SciTech Connect

    Katanuma, I.; Yagi, K.; Nakashima, Y.; Ichimura, M.; Imai, T.

    2010-03-15

    The computer code by reduced magnetohydrodynamic equations were made which can simulate the flute interchange modes (similar to the Rayleigh-Taylor instability) and the instability associated with the presence of nonuniform plasma flows (similar to the Kelvin-Helmholtz instability). This code is applied to a model divertor and the GAMMA10 [M. Inutake et al., Phys. Rev. Lett. 55, 939 (1985)] with divertor in order to investigate the flute modes in these divertor cells. The linear growth rate of the flute instability determined by the nonlocal linear analysis agrees with that in the linear phase of the simulations. There is a stable nonlinear steady state in both divertor cells, but the nonlinear steady state is different between the model divertor and the GAMMA10 with divertor.

  2. Thomson scattering diagnostic systems in ITER

    NASA Astrophysics Data System (ADS)

    Bassan, M.; Andrew, P.; Kurskiev, G.; Mukhin, E.; Hatae, T.; Vayakis, G.; Yatsuka, E.; Walsh, M.

    2016-01-01

    Thomson scattering (TS) is a proven diagnostic technique that will be implemented in ITER in three independent systems. The Edge TS will measure electron temperature Te and electron density ne profiles at high resolution in the region with r/a>0.8 (with a the minor radius). The Core TS will cover the region r/a<0.85 and shall be able to measure electron temperatures up to 40 keV . The Divertor TS will observe a segment of the divertor plasma more than 700 mm long and is designed to detect Te as low as 0.3 eV . The Edge and Core systems are primary contributors to Te and ne profiles. Both are installed in equatorial port 10 and very close together with the toroidal distance between the two laser beams of less than 600 mm at the first wall (~ 6° toroidal separation), a characteristic that should allow to reliably match the two profiles in the region 0.8ITER environment is imposing specific loads (e.g. gamma and neutron radiation, temperatures, disruption-induced stresses) and also access and reliability constraints that require new designs for many of the sub-systems. The challenges and the proposed solutions for all three TS systems are presented.

  3. Divertor and scoop limiter experiments on PDX

    SciTech Connect

    McGuire, K.; Beiersdorfer, P.; Bell, M.; Bol, K.; Boyd, D.; Buchenauer, D.; Budny, R.; Cavallo, A.; Couture, P.; Crowley, T.

    1985-01-01

    Routine operation in the enhanced energy confinement (or H-mode) regime during neutral beam injection was achieved by modifying the PDX divertor hardware to inhibit the influx of neutral gas from the divertor region to the main plasma chamber. A particle scoop limiter has been studied as a mechanical means of controlling particles at the plasma edge, and neutral beam heated discharges with this limiter show similar confinement times (normalized to tau/sub E//I/sub p/) to average H-mode plasmas. Two new instabilities are observed near the plasma edge in PDX during H-mode operation. The first, a quasicoherent fluctuation, occurred in bursts at well-defined frequencies (..delta omega../..omega.. less than or equal to 0.1) in the range 50 to 180 kHz, and had no obvious effects on confinement. The second instability, the edge relaxation phenomena (ERP), did cause deterioration in the global confinement time. The ERP's are characterized by sharp spikes in the divertor plasma density, H/sub ..cap alpha../ emission, and on the x-ray signals they appear as sawtoothlike relaxations at the plasma edge with an inversion radius near the separatrix. Attempts to obtain high ..beta../sub T/ in the H-mode discharges were hampered by a deterioration in the H-mode confinement and major disruptions which limited the achievable ..beta../sub T/. A study of the stability of both the limiter L-mode and divertor H-mode discharges close to the theoretical ..beta.. boundary, showed that the major disruptions observed there are sometimes caused by a fast growing m/n = 1/1 mode with no observable external precursor oscillations.

  4. JET divertor coils, manufacture, assembly and testing

    NASA Astrophysics Data System (ADS)

    Dolgetta, N.; Bertolini, E.; D'Urzo, C.; Last, J. R.; Laurenti, A.; Presle, P.; Sannazzaro, G.; Tait, J.; Tesini, A.

    1994-07-01

    Four coils have been built and installed in the JET vacuum vessel to produce divertor plasmas. The coils are copper with glass epoxy insulation and are enclosed in vacuum tight Inconel cases. At the coil contractor's factory, the coil parts were manufactured and process techniques qualified. In the JET vacuum vessel the conductor bars were brazed to form the coils, which were inserted in the casings and impregnated and cured with epoxy resin.

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

    SciTech Connect

    Linke, J.; Duwe, R.; Kuehnlein, W.

    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.

  6. Constrained ripple optimization of Tokamak bundle divertors

    SciTech Connect

    Hively, L.M.; Rome, J.A.; Lynch, V.E.; Lyon, J.F.; Fowler, R.H.; Peng, Y-K.M.; Dory, R.A.

    1983-02-01

    Magnetic field ripple from a tokamak bundle divertor is localized to a small toroidal sector and must be treated differently from the usual (distributed) toroidal field (TF) coil ripple. Generally, in a tokamak with an unoptimized divertor design, all of the banana-trapped fast ions are quickly lost due to banana drift diffusion or to trapping between the 1/R variation in absolute value vector B ..xi.. B and local field maxima due to the divertor. A computer code has been written to optimize automatically on-axis ripple subject to these constraints, while varying up to nine design parameters. Optimum configurations have low on-axis ripple (<0.2%) so that, now, most banana-trapped fast ions are confined. Only those ions with banana tips near the outside region (absolute value theta < or equal to 45/sup 0/) are lost. However, because finite-sized TF coils have not been used in this study, the flux bundle is not expanded.

  7. THERMAL HYDRAULIC ANALYSIS OF FIRE DIVERTOR

    SciTech Connect

    C.B. bAXI; M.A. ULRICKSON; D.E. DRIMEYER; P. HEITZENROEDER

    2000-10-01

    The Fusion Ignition Research Experiment (FIRE) is being designed as a next step in the US magnetic fusion program. The FIRE tokamak has a major radius of 2 m, a minor radius of 0.525 m, and liquid nitrogen cooled copper coils. The aim is to produce a pulse length of 20 s with a plasma current of 6.6 MA and with alpha dominated heating. The outer divertor and baffle of FIRE are water cooled. The worst thermal condition for the outer divertor and baffle is the baseline D-T operating mode (10 T, 6.6 MA, 20 s) with a plasma exhaust power of 67 MW and a peak heat flux of 20 MW/m{sup 2}. A swirl tape (ST) heat transfer enhancement method is used in the outer divertor cooling channels to increase the heat transfer coefficient and the critical heat flux (CHF). The plasma-facing surface consists of tungsten brush. The finite element (FE) analysis shows that for an inlet water temperature of 30 C, inlet pressure of 1.5 MPa and a flow velocity of 10 m/s, the incident critical heat flux is greater than 30 MW/m{sup 2}. The peak copper temperature is 490 C, peak tungsten temperature is 1560 C, and the pressure drop is less than 0.5 MPa. All these results fulfill the design requirements.

  8. ADX - Advanced Divertor and RF Tokamak Experiment

    NASA Astrophysics Data System (ADS)

    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.

  9. An Overview Of The ITER In-Vessel Coil Systems

    SciTech Connect

    Heitzenroeder, P J; Chrzanowski, J H; Dahlgren, F; Hawryluk, R J; Loesser, G D; Neumeyer, C; Mansfield, C; Smith, J P; Schaffer, M; Humphreys, D; Cordier, J J; Campbell, D; Johnson, G A; Martin, A; Rebut, P H; Tao, J O; Fogarty, P J; Nelson, B E; Reed, R P

    2009-09-24

    ELM mitigation is of particular importance in ITER in order to prevent rapid erosion or melting of the divertor surface, with the consequent risk of water leaks, increased plasma impurity content and disruptivity. Exploitable "natural" small or no ELM regimes might yet be found which extrapolate to ITER but this cannot be depended upon. Resonant Magnetic Perturbation has been added to pellet pacing as a tool for ITER to mitigate ELMs. Both are required, since neither method is fully developed and much work remains to be done. In addition, in-vessel coils enable vertical stabilization and RWM control. For these reasons, in-vessel coils (IVCs) are being designed for ITER to provide control of Edge Localized Modes (ELMs) in addition to providing control of moderately unstable resistive wall modes (RWMs) and the vertical stability (VS) of the plasma.

  10. Divertor bypass in the Alcator C-Mod tokamak

    NASA Astrophysics Data System (ADS)

    Pitcher, C. S.; LaBombard, B.; Danforth, R.; Pina, W.; Silveira, M.; Parkin, B.

    2001-01-01

    The Alcator C-Mod divertor bypass has for the first time allowed in situ variations to the mechanical baffle design in a tokamak. The design utilizes small coils which interact with the ambient magnetic field inside the vessel to provide the torque required to control small flaps of a Venetian blind geometry. Plasma physics experiments with the bypass have revealed the importance of the divertor baffling to maintain high divertor gas pressures. These experiments have also indicated that the divertor baffling has only a limited effect on the main chamber pressure in C-Mod.

  11. The effects of an open and closed divertor on particle exhaust during edge-localized mode suppression by resonant magnetic perturbations in DIII-D

    SciTech Connect

    Unterberg, E. A.; Schmitz, O.; Evans, T.E.; Maingi, Rajesh; Brooks, N. H.; Fenstermacher, M. E.; Mordijck, S.; Moyer, R.A.; Orlov, D. M.

    2010-01-01

    This paper compares the effects of divertor geometry on particle exhaust characteristics during the suppression of ELM using resonant magnetic perturbations (RMPs) on DIII-D. The subject is timely, particularly for ITER, because the combination of techniques to control or mitigate ELMs and control particle exhaust can provide confidence in the ability of an external pumping system to fully remove the particle exhaust. The differences between an open and closed divertor magnetic topology show a strong coupling of the perturbed strikepoint to the pumping manifold in closed divertor configurations, which can increase the particle exhaust by a factor of four. There is also an observed dependence on q(95) in this configuration, which is a common feature of RMP ELM suppression. Neutral density in both the active and non-active divertors is seen to increase during the RMP in the ISS configuration, and edge plasma conditions (i.e. n(e,sep) and midplane profile of D(alpha)) are seen to increase in the closed divertor configuration. Finally, the pumping exhaust is also shown to have a strong dependence on local measurements of the recycling flux. These observations, when taken as a whole, point to a substantial change in the plasma edge conditions, i.e. near the LCFS, throughout the poloidal cross-section of the vacuum vessel. This is coincident with the application of the RMP affecting the pumping capability of the system.

  12. Super-saturated hydrogen effects on radiation damages in tungsten under the high-flux divertor plasma irradiation

    NASA Astrophysics Data System (ADS)

    Kato, D.; Iwakiri, H.; Watanabe, Y.; Morishita, K.; Muroga, T.

    2015-08-01

    Tungsten is a prime candidate as the divertor material of the ITER and DEMO reactors, which would be exposed to unprecedentedly high-flux plasmas as well as neutrons. For a better characterization of radiation damages in the tungsten under the divertor condition, we examine influences of super-saturated hydrogen on vacancies in the tungsten. The present calculations based on density functional theory (DFT) reveal unusual phenomena predicted at a super-saturated hydrogen concentration: (1) strongly enhanced vacancy concentration with the super-saturated hydrogen concentration is predicted by a thermodynamics model assuming multiple-hydrogen trapping, i.e. hydrogen clusters formation, in the vacancies; and (2) DFT molecular dynamics revealed that hydrogen clusters can prevent a vacancy from recombining with the neighboring crowdion-type self-interstitial-atom. This suggests that neutron damage effects will be increased in the presence of the hydrogen clusters.

  13. First results on modeling of ITER infrared images

    NASA Astrophysics Data System (ADS)

    Kočan, M.; Reichle, R.; Aumeunier, M.-H.; Gunn, J. P.; Kajita, S.; Le Guern, F.; Lisgo, S. W.; Loarer, T.; Kukushkin, A. S.; Sashala Naik, A.; Rigollet, F.; Stratton, B.

    2016-02-01

    Infrared (IR) images of the ITER wide angle viewing system are modeled for the baseline plasma equilibrium and partially detached tungsten divertor, taking into account the three-dimensional structure of the first wall and the divertor. The modeling includes a comprehensive chain of calculations from the heat load specifications up to the synthetic, reflection-free IR images of the surface temperature, T surf. The effect of the optical blur due to finite IR detector size and diffraction/aberrations—approximated by a Gaussian filter—on the measured T surf is investigated. The optical blur characterized by σ = 0.7 pixel (approximately twice the diffraction limit) leads to underestimation of T surf,max on the inner vertical divertor target and near the upper X-point by <6% and <4%, respectively. This is within the required measurement accuracy of 10%. Larger underestimation of T surf,max (<12%) is observed on the outer vertical divertor target. The study demonstrates the importance of keeping the performance of the optical system as close as possible to the diffraction limit.

  14. The DIII-D Radiative Divertor Project: Status and plans

    SciTech Connect

    Smith, J.P.; Baxi, C.B.; Bozek, A.S.

    1996-10-01

    New divertor hardware is being designed and fabricated for the Radiative Divertor modification of the DIII-D tokamak. The installation of the hardware has been separated into two phases, the first phase starting in October of 1996 and the second and final phase, in 1998. The phased approach enables the continuation of the divertor characterization research in the lower divertor while providing pumping for density control in high triangularity, single- or double-null advanced tokamak discharges. When completed, the Radiative Divertor Project hardware will provide pumping at all four strike points of a double-null, high triangularity discharge and provide baffling of the neutral particles from transport back to the core plasma. By puffing neutral gas into the divertor region, a reduction in the heat flux on the target plates will be be demonstrated without a large rise in core density. This reduction in heat flux is accomplished by dispersing the power with radiation in the divertor region. Experiments and modeling have formed the basis for the new design. The capability of the DIII-D cryogenic system is being upgraded as part of this project. The increased capability of the cryogenic system will allow delivery of liquid helium and nitrogen to three new cryopumps. Physics studies on the effects of slot width and length can be accomplished easily with the design of the Radiative Divertor. The slot width can be varied by installing graphite tiles of different geometry. The change in slot length, the distance from the X-point to the target plate, requires relocating the structure vertically and can be completed in about 6-8 weeks. Radiative Divertor diagnostics are being designed to provide comprehensive measurements for diagnosing the divertor. Required diagnostic modifications will be minimal for Phase 1, but extensive for Phase 2 installation. These Phase 2 diagnostics will be required to fully diagnose the high triangularity discharges in the divertor slots.

  15. Wall conditioning for ITER: Current experimental and modeling activities

    NASA Astrophysics Data System (ADS)

    Douai, D.; Kogut, D.; Wauters, T.; Brezinsek, S.; Hagelaar, G. J. M.; Hong, S. H.; Lomas, P. J.; Lyssoivan, A.; Nunes, I.; Pitts, R. A.; Rohde, V.; de Vries, P. C.

    2015-08-01

    Wall conditioning will be required in ITER to control fuel and impurity recycling, as well as tritium (T) inventory. Analysis of conditioning cycle on the JET, with its ITER-Like Wall is presented, evidencing reduced need for wall cleaning in ITER compared to JET-CFC. Using a novel 2D multi-fluid model, current density during Glow Discharge Conditioning (GDC) on the in-vessel plasma-facing components (PFC) of ITER is predicted to approach the simple expectation of total anode current divided by wall surface area. Baking of the divertor to 350 °C should desorb the majority of the co-deposited T. ITER foresees the use of low temperature plasma based techniques compatible with the permanent toroidal magnetic field, such as Ion (ICWC) or Electron Cyclotron Wall Conditioning (ECWC), for tritium removal between ITER plasma pulses. Extrapolation of JET ICWC results to ITER indicates removal comparable to estimated T-retention in nominal ITER D:T shots, whereas GDC may be unattractive for that purpose.

  16. Effect of plasma response on the fast ion losses due to ELM control coils in ITER

    NASA Astrophysics Data System (ADS)

    Varje, Jari; Asunta, Otto; Cavinato, Mario; Gagliardi, Mario; Hirvijoki, Eero; Koskela, Tuomas; Kurki-Suonio, Taina; Liu, Yueqiang; Parail, Vassili; Saibene, Gabriella; Sipilä, Seppo; Snicker, Antti; Särkimäki, Konsta; Äkäslompolo, Simppa

    2016-04-01

    Mitigating edge localized modes (ELMs) with resonant magnetic perturbations (RMPs) can increase energetic particle losses and resulting wall loads, which have previously been studied in the vacuum approximation. This paper presents recent results of fusion alpha and NBI ion losses in the ITER baseline scenario modelled with the Monte Carlo orbit following code ASCOT in a realistic magnetic field including the effect of the plasma response. The response was found to reduce alpha particle losses but increase NBI losses, with up to 4.2% of the injected power being lost. Additionally, some of the load in the divertor was found to be shifted away from the target plates toward the divertor dome.

  17. Concept development for the ITER equatorial port visible/infrared wide angle viewing system

    SciTech Connect

    Reichle, R.; Beaumont, B.; Boilson, D.; Bouhamou, R.; Direz, M.-F.; Encheva, A.; Henderson, M.; Kazarian, F.; Lamalle, Ph.; Lisgo, S.; Mitteau, R.; Patel, K. M.; Pitcher, C. S.; Pitts, R. A.; Prakash, A.; Raffray, R.; Schunke, B.; Snipes, J.; Diaz, A. Suarez; Udintsev, V. S.; and others

    2012-10-15

    The ITER equatorial port visible/infrared wide angle viewing system concept is developed from the measurement requirements. The proposed solution situates 4 viewing systems in the equatorial ports 3, 9, 12, and 17 with 4 views each (looking at the upper target, the inner divertor, and tangentially left and right). This gives sufficient coverage. The spatial resolution of the divertor system is 2 times higher than the other views. For compensation of vacuum-vessel movements, an optical hinge concept is proposed. Compactness and low neutron streaming is achieved by orienting port plug doglegs horizontally. Calibration methods, risks, and R and D topics are outlined.

  18. Concept development for the ITER equatorial port visible/infrared wide angle viewing systema)

    NASA Astrophysics Data System (ADS)

    Reichle, R.; Beaumont, B.; Boilson, D.; Bouhamou, R.; Direz, M.-F.; Encheva, A.; Henderson, M.; Huxford, R.; Kazarian, F.; Lamalle, Ph.; Lisgo, S.; Mitteau, R.; Patel, K. M.; Pitcher, C. S.; Pitts, R. A.; Prakash, A.; Raffray, R.; Schunke, B.; Snipes, J.; Diaz, A. Suarez; Udintsev, V. S.; Walker, C.; Walsh, M.

    2012-10-01

    The ITER equatorial port visible/infrared wide angle viewing system concept is developed from the measurement requirements. The proposed solution situates 4 viewing systems in the equatorial ports 3, 9, 12, and 17 with 4 views each (looking at the upper target, the inner divertor, and tangentially left and right). This gives sufficient coverage. The spatial resolution of the divertor system is 2 times higher than the other views. For compensation of vacuum-vessel movements, an optical hinge concept is proposed. Compactness and low neutron streaming is achieved by orienting port plug doglegs horizontally. Calibration methods, risks, and R&D topics are outlined.

  19. Divertor for use in fusion reactors

    DOEpatents

    Christensen, Uffe R.

    1979-01-01

    A poloidal divertor for a toroidal plasma column ring having a set of poloidal coils co-axial with the plasma ring for providing a space for a thick shielding blanket close to the plasma along the entire length of the plasma ring cross section and all the way around the axis of rotation of the plasma ring. The poloidal coils of this invention also provide a stagnation point on the inside of the toroidal plasma column ring, gently curving field lines for vertical stability, an initial plasma current, and the shaping of the field lines of a separatrix up and around the shielding blanket.

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

    SciTech Connect

    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.

  1. Super-X divertors and high power density fusion devices

    SciTech Connect

    Valanju, P. M.; Kotschenreuther, M.; Mahajan, S. M.; Canik, J.

    2009-05-15

    The Super-X Divertor (SXD), a robust axisymmetric redesign of the divertor magnetic geometry that can allow a fivefold increase in the core power density of toroidal fusion devices, is presented. With small changes in poloidal coils and currents for standard divertors, the SXD allows the largest divertor plate radius inside toroidal field coils. This increases the plasma-wetted area by 2-3 times over all flux-expansion-only methods (e.g., plate near main X point, plate tilting, X divertor, and snowflake), decreases parallel heat flux and hence plasma temperature at plate, and increases connection length by 2-5 times. Examples of high-power-density fusion devices enabled by SXD are discussed; the most promising near-term device is a 100 MW modular compact fusion neutron source 'battery' small enough to fit inside a conventional fission blanket.

  2. RELAP5 MODEL OF THE DIVERTOR PRIMARY HEAT TRANSFER SYSTEM

    SciTech Connect

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

    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.

  3. OEDGE Modeling of Divertor Fueling at DIII-D

    NASA Astrophysics Data System (ADS)

    Bray, B. D.; Leonard, A. W.; Elder, J. D.; Stangeby, P. C.

    2015-11-01

    Onion-skin-modeling (OSM) is used to assess the affect of divertor closure on pedestal fueling sources. The OSM includes information from a wide range of diagnostic measurements at DIII-D to constrain the model background plasma for better simulation of neutrals and impurity ions and spectroscopy to compare to the results of the simulation. DIII-D has open lower divertor and closed upper divertor configurations which can be run with similar discharges. Progress toward modeling the pedestal fueling in low density plasmas for these cases will be presented as well as initial comparisons of recent lower single null discharges with the outer leg on the divertor shelf (fully open) and divertor floor (partially open). Work supported by the US DOE under DE-FC02-04ER54698 and DE-AC52-07NA27344.

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

    SciTech Connect

    Ulrickson, M.A.; Manly, W.D.; Dombrowski, D.E.

    1995-08-01

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

  5. The physics role of ITER

    SciTech Connect

    Rutherford, P.H.

    1997-04-01

    Experimental research on the International Thermonuclear Experimental Reactor (ITER) will go far beyond what is possible on present-day tokamaks to address new and challenging issues in the physics of reactor-like plasmas. First and foremost, experiments in ITER will explore the physics issues of burning plasmas--plasmas that are dominantly self-heated by alpha-particles created by the fusion reactions themselves. Such issues will include (i) new plasma-physical effects introduced by the presence within the plasma of an intense population of energetic alpha particles; (ii) the physics of magnetic confinement for a burning plasma, which will involve a complex interplay of transport, stability and an internal self-generated heat source; and (iii) the physics of very-long-pulse/steady-state burning plasmas, in which much of the plasma current is also self-generated and which will require effective control of plasma purity and plasma-wall interactions. Achieving and sustaining burning plasma regimes in a tokamak necessarily requires plasmas that are larger than those in present experiments and have higher energy content and power flow, as well as much longer pulse length. Accordingly, the experimental program on ITER will embrace the study of issues of plasma physics and plasma-materials interactions that are specific to a reactor-scale fusion experiment. Such issues will include (i) confinement physics for a tokamak in which, for the first time, the core-plasma and the edge-plasma are simultaneously in a reactor-like regime; (ii) phenomena arising during plasma transients, including so-called disruptions, in regimes of high plasma current and thermal energy; and (iii) physics of a radiative divertor designed for handling high power flow for long pulses, including novel plasma and atomic-physics effects as well as materials science of surfaces subject to intense plasma interaction. Experiments on ITER will be conducted by researchers in control rooms situated at major

  6. Analysis of a Multi-Machine Database on Divertor Heat Fluxes

    NASA Astrophysics Data System (ADS)

    Makowski, M. A.

    2011-10-01

    A coordinated effort to measure divertor heat flux characteristics in fully attached, similarly shaped H-mode plasmas on C-Mod, DIII-D and NSTX was carried out in 2010 in order to construct a predictive scaling relation applicable to next step devices including ITER, FNSF, and DEMO. Few published scaling laws are available and those that have been published were obtained under widely varying conditions and divertor geometries, leading to conflicting predictions for this critically important quantity. This study was designed to overcome these deficiencies. Corresponding plasma parameters were systematically varied in each tokamak, resulting in a combined data set in which Ip varies by a factor 3, Bt varies by a factor of 14.5, and major radius varies by a factor of 2.6. The derived scaling relation consistently predicts narrower heat flux widths than relations currently in use. Analysis of the combined data set reveals that the primary dependence of the parallel heat flux width is robustly inverse with Ip. All three tokamaks independently demonstrate this dependence. The midplane SOL profiles in DIII-D are also found to steepen with higher Ip, similar to the divertor heat flux profiles. Weaker dependencies on the toroidal field and normalized Greenwald density, fGW, are also found, but vary across devices and with the measure of the heat flux width used, either FWHM or integral width. In the combined data set, the strongest size scaling is with minor radius resulting in an approximately linear dependence on a /Ip . This suggests a scaling correlated with the inverse of the poloidal field, as would be expected for critical gradient or drift-based transport. Supported by the US DOE under DE-AC52-07NA27344 and DE-FC02-04ER54698.

  7. Electron Density Measurements in the National Spherical Torus Experiment Detached Divertor Region Using Stark Broadening of Deuterium Infrared Paschen Emission Lines

    SciTech Connect

    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.

  8. A super-cusp divertor configuration for tokamaks

    DOE PAGESBeta

    Ryutov, D. D.

    2015-08-26

    Our study demonstrates a remarkable flexibility of advanced divertor configurations created with the remote poloidal field coils. The emphasis here is on the configurations with three poloidal field nulls in the divertor area. We are seeking the structures where all three nulls lie on the same separatrix, thereby creating two zones of a very strong flux expansion, as envisaged in the concept of Takase’s cusp divertor. It turns out that the set of remote coils can produce a cusp divertor, with additional advantages of: (i) a large stand-off distance between the divertor and the coils and (ii) a thorough controlmore » that these coils exert over the fine features of the configuration. In reference to these additional favourable properties acquired by the cusp divertor, the resulting configuration could be called ‘a super-cusp’. General geometrical features of the three-null configurations produced by remote coils are described. Furthermore, issues on the way to practical applications include the need for a more sophisticated control system and possible constraints related to excessively high currents in the divertor coils.« less

  9. A super-cusp divertor configuration for tokamaks

    SciTech Connect

    Ryutov, D. D.

    2015-08-26

    Our study demonstrates a remarkable flexibility of advanced divertor configurations created with the remote poloidal field coils. The emphasis here is on the configurations with three poloidal field nulls in the divertor area. We are seeking the structures where all three nulls lie on the same separatrix, thereby creating two zones of a very strong flux expansion, as envisaged in the concept of Takase’s cusp divertor. It turns out that the set of remote coils can produce a cusp divertor, with additional advantages of: (i) a large stand-off distance between the divertor and the coils and (ii) a thorough control that these coils exert over the fine features of the configuration. In reference to these additional favourable properties acquired by the cusp divertor, the resulting configuration could be called ‘a super-cusp’. General geometrical features of the three-null configurations produced by remote coils are described. Furthermore, issues on the way to practical applications include the need for a more sophisticated control system and possible constraints related to excessively high currents in the divertor coils.

  10. SOLPS Modeling of Slot Divertor Configuration on DIII-D

    NASA Astrophysics Data System (ADS)

    Sang, C. F.; Stangeby, P. C.; Guo, H. Y.; Lao, L. L.

    2015-11-01

    A major thrust of the DIII-D boundary/PMI initiative is to develop an advanced divertor configuration for next-step devices, such as FNSF and DEMO. We are adopting an integrated approach by optimizing both divertor structure and magnetic shape. Initial SOLPS modeling was carried out to optimize divertor structure shape to enhance divertor power dissipation, focusing on slot configurations. In particular, four different slot divertor structures, i.e., orthogonal-target slot, slanted-target slot, very narrow slot and v-shaped slot have been analyzed and comparisons made with an open divertor structure. It is found that the slot helps to trap recycling neutrals and impurities thus increasing radiative power dissipation in the divertor, reducing the electron temperature Te and the perpendicular heat flux q⊥ at the target plate. As expected, a narrower slot leads to lower Te and q⊥ than a less narrow one. The v-shaped slot appears to be especially effective at redirecting and concentrating recycling neutrals and impurities near the separatrix, thus promoting detachment at a lower upstream density than the other configurations. Work supported by US DOE under DE-FC02-04ER54698.

  11. Long-term fuel retention in JET ITER-like wall

    NASA Astrophysics Data System (ADS)

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

    2016-02-01

    Post-mortem studies with ion beam analysis, thermal desorption, and secondary ion mass spectrometry have been applied for investigating the long-term fuel retention in the JET ITER-like wall components. The retention takes place via implantation and co-deposition, and the highest retention values were found to correlate with the thickness of the deposited impurity layers. From the total amount of retained D fuel over half was detected in the divertor region. The majority of the retained D is on the top surface of the inner divertor, whereas the least retention was measured in the main chamber on the mid-plane of the inner wall limiter. The recessed areas of the inner wall showed significant contribution to the main chamber total retention. Thermal desorption spectroscopy analysis revealed the energetic T from DD reactions being implanted in the divertor. The total T inventory was assessed to be \\gt 0.3 {{mg}}.

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

    NASA Astrophysics Data System (ADS)

    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

  13. Current and Potential Distribution in a Divertor with Torioidally-Asymmetric Biasing of the Divertor Plate

    SciTech Connect

    Cohen, R H; Ryutov, D D; Counsell, G F; Helander, P

    2006-06-06

    Toroidally-asymmetric biasing of the divertor plate may increase convective cross-field transport in SOL and thereby reduce the divertor heat load. Experiments performed with the MAST spherical tokamak generally agree with a simple theory of non-axisymmetric biasing. However, some of the experimental results have not yet received a theoretical explanation. In particular, existing theory seems to overestimate the asymmetry between the positive and the negative biasing. Also lacking a theoretical explanation is experimentally observed increase of the average floating potential in the main SOL in the presence of biasing. In this paper we attempt to solve these problems by accounting for the closing of the currents (driven by the biasing) in a strong-shear region near the X-point. We come up with the picture which, at least qualitatively, agrees with these experimental results.

  14. Optics design of the divertor infrared television of KSTAR.

    PubMed

    Oh, S; Lee, K; Lee, H H; Wi, H M; Kim, Y S; Kang, C S

    2014-11-01

    The divertor Infrared television (IR TV) system for monitoring the temperature of a divertor and localized hot spots will be installed on the upper port of the N-port in the Korea Superconducting Tokamak Advanced Research (KSTAR). The cassette of KSTAR makes a periscope inevitable for the divertor IR TV. In this article, 4 design concepts for the periscope were examined, and the design based on Keplerian was shown to have better stabilities in alignment and the vibration. The final optics design based on an f-theta lens, Keplerian, and telecentric lens was derived. PMID:25430316

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

    NASA Astrophysics Data System (ADS)

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

    2004-11-01

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

  16. Application of best-fit survey techniques throughout design, manufacturing and installation of the MKII divertor at JET

    SciTech Connect

    Macklin, B.; Celentano, G.; Tait, J.; Lente, E. van; Brade, R.; Shaw, R.

    1995-12-31

    The precise installation and alignment of large components in an activated and beryllium contaminated fusion device is a problem which must be faced in JET as well as future devices such as ITER. To guarantee the successful alignment of the MKII Divertor in JET it was essential that, early in the design phase, realistic manufacturing and installation tolerances and restrictions were identified and considered. The main components of the MKII divertor structure are an inner and outer ring mounted on a base plate. The overall diameter of the assembly is 6m and is dismantled into 24 sub-assemblies for installation. The structure must be installed very accurately whilst wearing full pressurized suits. As the other major in-vessel components remain unchanged it is important that the new divertor be installed to the same center as these components. Major considerations in the design process were the installation accuracy required, the installation method and restrictions imposed by the existing in-vessel structure. Joints between modules could only be made from one side due to access restrictions. Design of the support system had to be such that minimal modification to the existing in-vessel structure would be required. The tight tolerances necessary to ensure the mechanical integrity of the module joints were compromised by the necessity to have realistic assembly tolerances.

  17. Potential collector surface materials for divertors

    NASA Astrophysics Data System (ADS)

    Prebble, H. E.; Forty, C. B. A.; Butterworth, G. J.

    1992-09-01

    Twelve refractory materials have been investigated to assess their suitability for use as collector target materials for divertors. The steady state limiting heat flux to avoid melting of the collector material has been calculated as a function of thickness using a simple one-dimensional thermal-hydraulics model. Similarly, the limiting heat flux to avoid melting following a plasma disruption has been calculated as a function of collector surface temperature just prior to the disruption event. Finally, the resistance of each collector material to thermal shock was estimated. The calculations indicate diamond, graphite and tungsten as favourable materials, BN, AlN, TiN, V 2C and beryllium as unsuitable and BeO, SiC, TiC and TIB 2 as exhibiting combinations of favourable and unfavourable properties.

  18. Divertor conditions relevant for fusion reactors achieved with linear plasma generator

    SciTech Connect

    Eck, H. J. N. van; Lof, A.; Meiden, H. J. van der; Rooij, G. J. van; Scholten, J.; Zeijlmans van Emmichoven, P. A.; Kleyn, A. W.

    2012-11-26

    Intense magnetized hydrogen and deuterium plasmas have been produced with electron densities up to 3.6 Multiplication-Sign 10{sup 20} m{sup -3} and electron temperatures up to 3.7 eV with a linear plasma generator. Exposure of a W target has led to average heat and particle flux densities well in excess of 4 MW m{sup -2} and 10{sup 24} m{sup -2} s{sup -1}, respectively. We have shown that the plasma surface interactions are dominated by the incoming ions. The achieved conditions correspond very well to the projected conditions at the divertor strike zones of fusion reactors such as ITER. In addition, the machine has an unprecedented high gas efficiency.

  19. EXB-Drift, Current, and Kinetic Effects on Divertor Plasma Profiles During ELMs

    SciTech Connect

    Rognlien, T.D.; Shimada, M.

    2002-05-23

    The transient heat load on divertor surfaces from Edge-Localized Modes (ELMs) in tokamaks can be very large and thus of concern for a large device such as ITER. Models for kinetic modifications to fluid models are discussed that should allow them to reasonably describe the long mean-free path regime encountered owing to the high electron and ion temperatures in the SOL during large ELMs. A set of two-dimensional (2D) simulations of the dynamic response of the scrape-off layer (SOL) plasma to an ELM is presented. The role of plasma currents and E x B motion is emphasized, which cause large changes in the response compared to models neglecting them.

  20. Evidence for enhanced cross-field transport mechanisms in the TCV Snowflake divertor

    NASA Astrophysics Data System (ADS)

    Vijvers, Wouter

    2015-11-01

    TCV experiments demonstrate that cross-field plasma transport is enhanced in the Snowflake divertor (SFD) compared to a standard single-null divertor (SND). This enhanced cross-field transport spreads the exhaust power over a larger surface area than can be achieved by magnetic geometry alone and, thereby, reduces the peak heat flux. Comparison of the experiments with modelling identifies steepened radial gradients, ExB drift effects, and βp-driven instabilities as the responsible transport mechanisms. The uncovered physics is also relevant to the SND and may help improve predictive models for the target profiles in ITER and DEMO. In SFD variants with an X-point in the scrape-off layer (SOL), part of the heat flux profile is split off and redirected to an additional target. The resulting steepened radial gradients enhance cross-field diffusion. This is confirmed by EMC3-Eirene simulations, which show a factor two reduction of the parallel heat flux, even if diffusivities remain constant. Theoretical analysis predicts enhanced ExB drifts in the SFD by increased poloidal gradients of the temperature and density. The predictions are confirmed by target heat and particle flux measurements in dedicated experiments with both toroidal field directions. Cross-field convection by curvature-driven modes at high βp (``churning modes'') explains the large fluxes into the private flux region of the SFD. This activates the extra targets and reduces the peak power to the primary targets up to a factor four. This mechanism is expected to be most effective when the divertor conditions are most severe: near the separatrix of a narrow, high-pressure SOL of a large tokamak. These and other alternative divertor configurations thus provide potential solutions to the power exhaust challenge, as well as laboratories to study SOL transport, one of the most important topics in tokamak research. This project was carried out with financial support from NWO. The work was carried out within

  1. Status of National Spherical Torus Experiment Liquid Lithium Divertor

    NASA Astrophysics Data System (ADS)

    Kugel, H. W.; Viola, M.; Ellis, R.; Bell, M.; Gerhardt, S.; Kaita, R.; Kallman, J.; Majeski, R.; Mansfield, D.; Roquemore, A. L.; Schneider, H.; Timberlake, J.; Zakharov, L.; Nygren, R. E.; Allain, J. P.; Maingi, R.; Soukhanovskii, V.

    2009-11-01

    Recent NSTX high power divertor experiments have shown significant and recurring benefits of solid lithium coatings on plasma facing components to the performance of divertor plasmas in both L- and H- mode confinement regimes heated by high-power neutral beams. The next step in this work is the 2009 installation of a Liquid Lithium Divertor (LLD). The 20 cm wide LLD located on the lower outer divertor, consists of four, 80 degree sections; each section is separated by a row of graphite diagnostic tiles. The temperature controlled LLD structure consists of a 0.01cm layer of vacuum flame-sprayed, 50 percent porous molybdenum, on top of 0.02 cm, 316-SS brazed to a 1.9 cm Cu base. The physics design of the LLD encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.

  2. Local Effects of Biased Electrodes in the Divertor of NSTX

    SciTech Connect

    Zweben, S.; Campanell, M. D.; Lyons, B. C.; Maqueda, R. J.; Raitses, Y.; Roquemore, A. L.; Scotti, F.

    2012-05-07

    The goal of this paper is to characterize the effects of small non-axisymmetric divertor plate electrodes on the local scrape-off layer plasma. Four small rectangular electrodes were installed into the outer divertor plates of NSTX. When the electrodes were located near the outer divertor strike point and biased positively, there was an increase in the nearby probe currents and probe potentials and an increase in the LiI light emission at the large major radius end of these electrodes. When an electrode located farther outward from the outer divertor strike point was biased positively, there was sometimes a significant decrease in the LiI light emission at the small major radius end of this electrode, but there were no clear effects on the nearby probes. No non-local effects were observed with the biasing of these electrodes.

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

    NASA Astrophysics Data System (ADS)

    Leonard, A. W.; Makowski, M. A.; McLean, A. G.; Osborne, T. H.; Snyder, P. B.

    2015-08-01

    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, Te ⩽ 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.

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

    NASA Astrophysics Data System (ADS)

    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

  5. DIII-D Research in Support of ITER

    SciTech Connect

    Strait, E

    2008-10-14

    several approaches to mitigation of disruptions, including injection of low-Z gas and low-Z pellets, and have shown the conditions that minimize core impurity accumulation during radiative divertor operation. Investigation of carbon erosion, transport, and co-deposition with hydrogenic species, and methods for the removal of co-deposits, will contribute to the physics basis for initial operation of ITER with a carbon divertor.

  6. Structural design of the DIII-D radiative divertor

    SciTech Connect

    Reis, E.E.; Smith, J.P.; Baxi, C.B.; Bozek, A.S.; Chin, E.; Hollerbach, M.A.; Laughon, G.J.; Sevier, D.L.

    1996-10-01

    The divertor of the DIII-D tokamak is being modified to operate as a slot type, dissipative divertor. This modification, called the Radiative Divertor Program (RDP) is being carried out in two phases. The design and analysis is complete and hardware is being fabricated for the first phase. This first phase consists of an upper divertor baffle and cryopump to provide some density control for high triangularity, single or double null discharges. Installation of the first phase is scheduled to start in October, 1996. The second phase provides pumping at all four divertor strike points of double null high triangularity discharges and baffling of the neutral particles from transport back to the core plasma. Studies of the effects of varying the slot length and width of the divertor can be easily accomplished with the design of RDP hardware. Static and dynamic analyses of the baffle structures, new cryopumps, and feedlines were performed during the preliminary and final design phases. Disruption loads and differential thermal displacements must be accommodated in the design of these components. With the full RDP hardware installed, the plasma current in DIII-D will be a maximum of 3.0 MA. Plasma disruptions induce toroidal currents in the cryopump, producing complex dynamic loads. Simultaneously, the vacuum vessel vibrations impose a sinusoidal base excitation to the supports for the cryopump. Static and dynamic analyses of the cryopump demonstrate that the stresses due to disruption and thermal loadings satisfy the stress and deflection criteria.

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

    NASA Astrophysics Data System (ADS)

    Kamada, Y.; Barabaschi, P.; Ishida, S.; Ide, S.; Lackner, K.; Fujita, T.; Bolzonella, T.; Suzuki, T.; Matsunaga, G.; Yoshida, M.; Shinohara, K.; Urano, H.; Nakano, T.; Sakurai, S.; Kawashima, H.; JT-60SA Team

    2011-07-01

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

  8. Evaluation of pumping and fueling requirements for the ITER EDA

    NASA Astrophysics Data System (ADS)

    Houlberg, W. A.; Attenberger, S. E.

    1994-06-01

    The relationships between fueling (gas injection and pellets of various sizes and velocities), pumping in the divertor chamber (constrained by fuel processing and divertor design), core density (constrained by the desired fusion power and helium ash accumulation), separatrix density (constrained by divertor operation and density limits) and plasma confinement models are examined for the International Engineering Tokamak Reactor (ITER) Engineering Design Activity (EDA) for guidance in the definition of design requirements for the pumping and fueling systems. Various combinations of gas and pellet injection are found to meet the constraints for operation at 1,500 MW of fusion power and 1 bar(center dot)l/s (5.3 x 10(exp 22) atoms/s) of DT pumping. Very low pumping reduces fuel processing requirements, but can lead to excessive helium accumulation depending on the particle transport properties. Isotopic tailoring of the fuel sources, e.g., 20-30% of the input fuel stream as tritium pellets and the rest as deuterium gas, can maintain the core fuel species mixture in the optimum range for fusion production (at least a 40-60 mixture) while reducing the tritium concentration in the edge region to 20-30%. This should reduce the tritium inventory in the plasma facing components, since that is typically governed by the fuel density mix near the plasma edge. A high density, low temperature ignited regime supported by deep pellet injection is shown to exist under some low confinement conditions.

  9. Comment on “Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake” [Phys. Plasmas 20, 102507 (2013)

    SciTech Connect

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

    2014-05-15

    In the recently published paper “Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake” [Phys. Plasmas 20, 102507 (2013)], the authors raise interesting and important issues concerning divertor physics and design. However, the paper contains significant errors: (a) The conceptual framework used in it for the evaluation of divertor “quality” is reduced to the assessment of the magnetic field structure in the outer Scrape-Off Layer. This framework is incorrect because processes affecting the pedestal, the private flux region and all of the divertor legs (four, in the case of a snowflake) are an inseparable part of divertor operation. (b) The concept of the divertor index focuses on only one feature of the magnetic field structure and can be quite misleading when applied to divertor design. (c) The suggestion to rename the divertor configurations experimentally realized on NSTX (National Spherical Torus Experiment) and DIII-D (Doublet III-D) from snowflakes to X-divertors is not justified: it is not based on comparison of these configurations with the prototypical X-divertor, and it ignores the fact that the NSTX and DIII-D poloidal magnetic field geometries fit very well into the snowflake “two-null” prescription.

  10. Comment on "Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake" [Phys. Plasmas 20, 102507 (2013)

    NASA Astrophysics Data System (ADS)

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

    2014-05-01

    In the recently published paper "Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake" [Phys. Plasmas 20, 102507 (2013)], the authors raise interesting and important issues concerning divertor physics and design. However, the paper contains significant errors: (a) The conceptual framework used in it for the evaluation of divertor "quality" is reduced to the assessment of the magnetic field structure in the outer Scrape-Off Layer. This framework is incorrect because processes affecting the pedestal, the private flux region and all of the divertor legs (four, in the case of a snowflake) are an inseparable part of divertor operation. (b) The concept of the divertor index focuses on only one feature of the magnetic field structure and can be quite misleading when applied to divertor design. (c) The suggestion to rename the divertor configurations experimentally realized on NSTX (National Spherical Torus Experiment) and DIII-D (Doublet III-D) from snowflakes to X-divertors is not justified: it is not based on comparison of these configurations with the prototypical X-divertor, and it ignores the fact that the NSTX and DIII-D poloidal magnetic field geometries fit very well into the snowflake "two-null" prescription.

  11. US ITER Moving Forward

    ScienceCinema

    US ITER / ORNL

    2012-03-16

    US ITER Project Manager Ned Sauthoff, joined by Wayne Reiersen, Team Leader Magnet Systems, and Jan Berry, Team Leader Tokamak Cooling System, discuss the U.S.'s role in the ITER international collaboration.

  12. Spectroscopic measurements and modeling of tungsten erosion in the DIII-D divertor

    NASA Astrophysics Data System (ADS)

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

    2015-11-01

    In situ time-resolved measurements of the gross W erosion rate have been performed in DIII-D by monitoring W/I (400.9 nm) emission in the divertor via a filtered camera and high-resolution spectrometer. The erosion rate of a thin W coating on DiMES, inferred via the S/XB method, was found to be ~ 0.7 nm/s during deuterim L-mode exposure, in fair agreement with post-mortem IBA analysis but lower than REDEP/WBC modeling. During H-mode He bombardment of W disks, average erosion rates of ~ 2.9 nm/s and ~ 9.0 nm/s were estimated during the inter-ELM and intra-ELM phases, using ne and Te from divertor Thomson scattering and Langmuir probes. Results will also be presented from additional W erosion experiments in preparation for the DIII-D mini-campaign to measure high-Z transport in the edge plasma. Comparisons will be made with ERO modeling Supported by US DOE DE-AC05-06OR23100, DE-FC02-04ER54698, DE-AC52-07NA27344, DE-AC05-00OR22725, DE-SC0001961, DE-AC04-94AL85000.

  13. Vanadium alloys for the radiative divertor program of DIII-D

    SciTech Connect

    Smith, J.P.; Johnson, W.R.; Stambaugh, R.D.; Trester, P.W.; Smith, D.; Bloom, E.

    1995-10-01

    Vanadium alloys provide an attractive solution for fusion power plants as they exhibit a potential for low environmental impact due to low level of activation from neutron fluence and a relatively short half-life. They also have attractive material properties for use in a reactor. General Atomics along with Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL), has developed a plan to utilize vanadium alloys as part of the Radiative Divertor Project (RDP) modification for the DIII-D tokamak. The goal for using vanadium alloys is to provide a meaningful step towards developing advanced materials for fusion power applications by demonstrating the in-service behavior of a vanadium alloy (V-4Cr-4Ti) in a tokamak in conjunction with developing essential fabrication technology for the manufacture of full-scale vanadium alloy components. A phased approach towards utilizing vanadium in DIII-D is being used starting with small coupons and samples, advancing to a small component, and finally a portion of the new double-null, slotted divertor will be fabricated from vanadium alloy product forms. A major portion of the program is research and development to support fabrication and resolve key issues related to environmental effects.

  14. Carbon flows in attached divertor plasmas

    SciTech Connect

    Isler, R.C.; Brooks, N.H.; West, W.P.; Porter, G.D. |; The DIII-D Divertor Team

    1999-05-01

    Parallel flow velocities of carbon ions in the DIII-D divertor [J. Luxon {ital et al.}, {ital Plasma Physics Controlled Nuclear Fusion Research}, 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159; S. L. Allen {ital et al.}, {ital Controlled Fusion and Plasma Physics}, 1987 (Proc. 24th European Conf. Berchtesgaden, 1997), Vol. 21 A, Part III, p. 1129] have been studied under various operating conditions: L-mode (low-confinement mode), H-mode (high-confinement mode) with low-frequency ELMs (edge-localized modes), and H-mode with high-frequency ELMs. Both normal and reversed flows (toward the target plate and away from the target plate, respectively) are observed under all conditions, with the reversed speeds being as much as a factor of four greater than normal speeds. Magnitudes are approximately the same for L-mode and H-mode operation with high-frequency ELMs. In H-mode conditions with low-frequency ELMs, normal velocities are frequently observed to decline while reversed velocities increase in comparison to the other two conditions. {copyright} {ital 1999 American Institute of Physics.}

  15. Plasma-surface interaction issues of an all-metal ITER

    NASA Astrophysics Data System (ADS)

    Brooks, J.N.; Allain, J.P.; Doerner, R.P.; Hassanein, A.; Nygren, R.; Rognlien, T.D.; Whyte, D.G.

    2009-03-01

    We assess key plasma-surface interaction issues of an all-metal plasma facing component (PFC) system for ITER, in particular a tungsten divertor, and a beryllium or tungsten first wall. Such a system eliminates problems with carbon divertor erosion and T/C codeposition, and for an all-tungsten system would better extrapolate to post-ITER devices. The issues studied are sputtering, transport and formation of mixed surface layers, tritium codeposition, plasma contamination, edge-localized mode (ELM) response and He-on-W irradiation effects. Code package OMEGA computes PFC sputtering erosion/redeposition in an ITER full power D-T plasma with convective edge transport. The HEIGHTS package analyses plasma transient response. PISCES and other data are used with code results to assess PFC performance. Predicted outer-wall sputter erosion rates are acceptable for Be (0.3 nm s-1) or bare (stainless steel/Fe) wall (0.05 nm s-1) for the low duty factor ITER, and are very low (0.002 nm s-1) for W. T/Be codeposition in redeposited wall material could be significant (~2 gT/400 s-ITER pulse). Core plasma contamination from wall sputtering appears acceptable for Be (~2%) and negligible for W (or Fe). A W divertor has negligible sputter erosion, plasma contamination and T/W codeposition. Be can grow at/near the strike point region of a W divertor, but for the predicted maximum surface temperature of ~800 °C, deleterious Be/W alloy formation as well as major He/W surface degradation will probably be avoided. ELMs are a serious challenge to the divertor, but this is true for all materials. We identify acceptable ELM parameters for W. We conclude that an all-metal PFC system is likely a much better choice for ITER D-T operation than a system using C. We discuss critical R&D needs, testing requirements, and suggest employing a 350-400 °C baking capability for T/Be reduction and using a deposited tungsten first wall test section.

  16. Beryllium migration in JET ITER-like wall plasmas

    NASA Astrophysics Data System (ADS)

    Brezinsek, S.; Widdowson, A.; Mayer, M.; Philipps, V.; Baron-Wiechec, P.; Coenen, J. W.; Heinola, K.; Huber, A.; Likonen, J.; Petersson, P.; Rubel, M.; Stamp, M. F.; Borodin, D.; Coad, J. P.; Carrasco, A. G.; Kirschner, A.; Krat, S.; Krieger, K.; Lipschultz, B.; Linsmeier, Ch.; Matthews, G. F.; Schmid, K.; contributors, JET

    2015-06-01

    JET is used as a test bed for ITER, to investigate beryllium migration which connects the lifetime of first-wall components under erosion with tokamak safety, in relation to long-term fuel retention. The (i) limiter and the (ii) divertor configurations have been studied in JET-ILW (JET with a Be first wall and W divertor), and compared with those for the former JET-C (JET with carbon-based plasma-facing components (PFCs)). (i) For the limiter configuration, the Be gross erosion at the contact point was determined in situ by spectroscopy as between 4% (Ein = 35 eV) and more than 100%, caused by Be self-sputtering (Ein = 200 eV). Chemically assisted physical sputtering via BeD release has been identified to contribute to the effective Be sputtering yield, i.e. at Ein = 75 eV, erosion was enhanced by about 1/3 with respect to the bare physical sputtering case. An effective gross yield of 10% is on average representative for limiter plasma conditions, whereas a factor of 2 difference between the gross erosion and net erosion, determined by post-mortem analysis, was found. The primary impurity source in the limiter configuration in JET-ILW is only 25% higher (in weight) than that for the JET-C case. The main fraction of eroded Be stays within the main chamber. (ii) For the divertor configuration, neutral Be and BeD from physically and chemically assisted physical sputtering by charge exchange neutrals and residual ion flux at the recessed wall enter the plasma, ionize and are transported by scrape-off layer flows towards the inner divertor where significant net deposition takes place. The amount of Be eroded at the first wall (21 g) and the Be amount deposited in the inner divertor (28 g) are in fair agreement, though the balancing is as yet incomplete due to the limited analysis of PFCs. The primary impurity source in the JET-ILW is a factor of 5.3 less in comparison with that for JET-C, resulting in lower divertor material deposition, by more than one order of

  17. Review of the ITER diagnostics suite for erosion, deposition, dust and tritium measurements

    NASA Astrophysics Data System (ADS)

    Reichle, R.; Andrew, P.; Bates, P.; Bede, O.; Casal, N.; Choi, C. H.; Barnsley, R.; Damiani, C.; Bertalot, L.; Dubus, G.; Ferreol, J.; Jagannathan, G.; Kocan, M.; Leipold, F.; Lisgo, S. W.; Martin, V.; Palmer, J.; Pearce, R.; Philipps, V.; Pitts, R. A.; Pampin, R.; Passedat, G.; Puiu, A.; Suarez, A.; Shigin, P.; Shu, W.; Vayakis, G.; Veshchev, E.; Walsh, M.

    2015-08-01

    Dust and tritium inventories in the vacuum vessel have upper limits in ITER that are set by nuclear safety requirements. Erosion, migration and re-deposition of wall material together with fuel co-deposition will be largely responsible for these inventories. The diagnostic suite required to monitor these processes, along with the set of the corresponding measurement requirements is currently under review given the recent decision by the ITER Organization to eliminate the first carbon/tungsten (C/W) divertor and begin operations with a full-W variant Pitts et al. [1]. This paper presents the result of this review as well as the status of the chosen diagnostics.

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

    NASA Astrophysics Data System (ADS)

    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.

  19. Optimal thermal-hydraulic performance for helium-cooled divertors

    SciTech Connect

    Izenson, M.G.; Martin, J.L.

    1996-07-01

    Normal flow heat exchanger (NFHX) technology offers the potential for cooling divertor panels with reduced pressure drops (<0.5% {Delta}p/p), reduced pumping power (<0.75% pumping/thermal power), and smaller duct sizes than conventional helium heat exchangers. Furthermore, the NFHX can easily be fabricated in the large sizes required for divertors in large tokamaks. Recent experimental and computational results from a program to develop NFHX technology for divertor coolings using porous metal heat transfer media are described. We have tested the thermal and flow characteristics of porous metals and identified the optimal heat transfer material for the divertor heat exchanger. Methods have been developed to create highly conductive thermal bonds between the porous material and a solid substrate. Computational fluid dynamics calculations of flow and heat transfer in the porous metal layer have shown the capability of high thermal effectiveness. An 18-kW NFHX, designed to meet specifications for the international Thermonuclear Experimental Reactor divertor, has been fabricated and tested for thermal and flow performance. Preliminary results confirm design and fabrication methods. 11 refs., 12 figs., 1 tab.

  20. Radiative snowflake divertor studies in DIII-D

    NASA Astrophysics Data System (ADS)

    Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; Hill, D. N.; Lasnier, C. J.; Makowski, M. A.; McLean, A. G.; Meyer, W. H.; Kolemen, E.; Groebner, R. J.; Hyatt, A. W.; Leonard, A. W.; Osborne, T. H.; Petrie, T. W.

    2015-08-01

    Recent DIII-D experiments assessed the snowflake divertor (SF) configuration in a radiative regime in H-mode discharges with D2 seeding. The SF configuration was maintained for many energy confinement times (2-3 s) in H-mode discharges (Ip = 1.2 MA, PNBI = 4- 5 MW, and B × ∇B down (favorable direction toward the divertor)), and found to be compatible with high performance operation (H98y2 ⩾ 1). The two studied SF configurations, the SF-plus and the SF-minus, have a small finite distance between the primary X-point and the secondary Bp null located in the private flux region or the common flux region, respectively. In H-mode discharges with the SF configurations (cf. H-mode discharges with the standard divertor with similar conditions) the stored energy lost per the edge localized mode (ELM) was reduced, and significant divertor heat flux reduction between and during ELMs was observed over a range of collisionalities, from lower density conditions toward a higher density H-modes with the radiative SF divertor.

  1. Atomic Physics in the Quest for Fusion Energy and ITER

    SciTech Connect

    Charles H. Skinner

    2008-02-27

    The urgent quest for new energy sources has led developed countries, representing over half of the world population, to collaborate on demonstrating the scientific and technological feasibility of magnetic fusion through the construction and operation of ITER. Data on high-Z ions will be important in this quest. Tungsten plasma facing components have the necessary low erosion rates and low tritium retention but the high radiative efficiency of tungsten ions leads to stringent restrictions on the concentration of tungsten ions in the burning plasma. The influx of tungsten to the burning plasma will need to be diagnosed, understood and stringently controlled. Expanded knowledge of the atomic physics of neutral and ionized tungsten will be important to monitor impurity influxes and derive tungsten concentrations. Also, inert gases such as argon and xenon will be used to dissipate the heat flux flowing to the divertor. This article will summarize the spectroscopic diagnostics planned for ITER and outline areas where additional data is needed.

  2. ITER physics design guidelines at high aspect ratio

    NASA Astrophysics Data System (ADS)

    Uckan, N. A.

    1991-09-01

    The physics requirements for the International Thermonuclear Experimental Reactor (ITER) design are formulated in a set of physics design guidelines. These guidelines, established by the ITER Physics Group during the Conceptual Design Activity (CDA, 1988--90), were based on credible extrapolations of the tokamak physics database as assessed during the CDA, and defined a class of tokamak designs (with plasma current I is approximately 20 MA and aspect ratio A is approximately 2.5--3.5) that meet the ITER objectives. Recent U.S. studies have indicated that there may be significant benefits if the ITER-CDA design point is moved from the low aspect ratio, high current baseline (A = 2.79, I = 22 MA) to a high aspect ratio machine at Ais approximately 4, I is approximately 15 MA, especially regarding steady-state, technology-testing performance. To adequately assess the physics and technology testing capability of higher aspect ratio design options, several changes are proposed to the original ITER guidelines to reflect the latest developments in physics understanding at higher aspect ratios. The critical issues for higher aspect ratio design options are the uncertainty in scaling of confinement with aspect ratio, the variation of vertical stability with elongation and aspect ratio, plasma shaping requirements, ability to control and maintain plasma current and q-profiles for MHD stability (and volt-second consumption), access for current drive, restrictions on field ripple and divertor plate incident angles, etc.

  3. Scaling of divertor power footprint width in RF-heated type-III ELMy H-mode on the EAST superconducting tokamak

    NASA Astrophysics Data System (ADS)

    Wang, L.; Guo, H. Y.; Xu, G. S.; Liu, S. C.; Gan, K. F.; Wang, H. Q.; Gong, X. Z.; Liang, Y.; Zou, X. L.; Hu, J. S.; Chen, L.; Xu, J. C.; Liu, J. B.; Yan, N.; Zhang, W.; Chen, R.; Shao, L. M.; Ding, S.; Hu, G. H.; Feng, W.; Zhao, N.; Xiang, L. Y.; Liu, Y. L.; Li, Y. L.; Sang, C. F.; Sun, J. Z.; Wang, D. Z.; Ding, H. B.; Luo, G. N.; Chen, J. L.; Gao, X.; Hu, L. Q.; Wan, B. N.; Li, J.; the EAST Team

    2014-11-01

    Dedicated experiments for the scaling of divertor power footprint width have been performed in the ITER-relevant radio-frequency (RF)-heated H-mode scheme under the lower single null, double null and upper single null divertor configurations in the Experimental Advanced Superconducting Tokamak (EAST) under lithium wall coating conditioning. A strong inverse scaling of the edge localized mode (ELM)-averaged power fall-off width with the plasma current (equivalently the poloidal field) has been demonstrated for the attached type-III ELMy H-mode as λq \\propto Ip-1.05 by various heat flux diagnostics including the divertor Langmuir probes (LPs), infra-red (IR) thermograph and reciprocating LPs on the low-field side. The IR camera and divertor LP measurements show that λq,IR ≈ {λq,div{-LPs}}/{1.3}=1.15Bp,omp-1.25 , in good agreement with the multi-machine scaling trend during the inter-ELM phase between type-I ELMs or ELM-free enhanced Dα (EDA). H-mode. However, the magnitude is nearly doubled, which may be attributed to the different operation scenarios or heating schemes in EAST, i.e., dominated by electron heating. It is also shown that the type-III ELMs only broaden the power fall-off width slightly, and the ELM-averaged width is representative for the inter-ELM period. Furthermore, the inverse Ip (Bp) scaling appears to be independent of the divertor configurations in EAST. The divertor power footprint integral width, fall-off width and dissipation width derived from EAST IR camera measurements follow the relation, λint ≅ λq + 1.64S, yielding λ_intEAST =(1.39+/- 0.03)λqEAST +(0.97+/- 0.35) mm . Detailed analysis of these three characteristic widths was carried out to shed more light on their extrapolation to ITER.

  4. Gyrokinetic simulation of edge blobs and divertor heat-load footprint

    NASA Astrophysics Data System (ADS)

    Chang, C. S.; Ku, S.; Hager, R.; Churchill, M.; D'Azevedo, E.; Worley, P.

    2015-11-01

    Gyrokinetic study of divertor heat-load width Lq has been performed using the edge gyrokinetic code XGC1. Both neoclassical and electrostatic turbulence physics are self-consistently included in the simulation with fully nonlinear Fokker-Planck collision operation and neutral recycling. Gyrokinetic ions and drift kinetic electrons constitute the plasma in realistic magnetic separatrix geometry. The electron density fluctuations from nonlinear turbulence form blobs, as similarly seen in the experiments. DIII-D and NSTX geometries have been used to represent today's conventional and tight aspect ratio tokamaks. XGC1 shows that the ion neoclassical orbit dynamics dominates over the blob physics in setting Lq in the sample DIII-D and NSTX plasmas, re-discovering the experimentally observed 1/Ip type scaling. Magnitude of Lq is in the right ballpark, too, in comparison with experimental data. However, in an ITER standard plasma, XGC1 shows that the negligible neoclassical orbit excursion effect makes the blob dynamics to dominate Lq. Differently from Lq 1mm (when mapped back to outboard midplane) as was predicted by simple-minded extrapolation from the present-day data, XGC1 shows that Lq in ITER is about 1 cm that is somewhat smaller than the average blob size. Supported by US DOE and the INCITE program.

  5. Broadening of divertor heat flux profile with increasing number of ELM filaments in NSTX

    DOE PAGESBeta

    Ahn, J. -W.; Maingi, R.; Canik, J. M.; Gan, K. F.; Gray, T. K.; McLean, A. G.

    2014-11-13

    Edge localized modes (ELMs) represent a challenge to future fusion devices, owing to cyclical high peak heat fluxes on divertor plasma facing surfaces. One ameliorating factor has been that the heat flux characteristic profile width has been observed to broaden with the size of the ELM, as compared with the inter-ELM heat flux profile. In contrast, the heat flux profile has been observed to narrow during ELMs under certain conditions in NSTX. Here we show that the ELM heat flux profile width increases with the number of filamentary striations observed, i.e., profile narrowing is observed with zero or very fewmore » striations. Because NSTX often lies on the long wavelength current-driven mode side of ideal MHD instabilities, few filamentary structures can be expected under many conditions. Lastly, ITER is also projected to lie on the current driven low-n stability boundary, and therefore detailed projections of the unstable modes expected in ITER and the heat flux driven in ensuing filamentary structures is needed.« less

  6. Viewgraphs presented at the ASDEX/DOE workshop on disruptions in divertor tokamaks

    SciTech Connect

    Granetz, R.; Gruber, O.; Zohm, H.

    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.

  7. Development and qualification of a bulk tungsten divertor row for JET

    NASA Astrophysics Data System (ADS)

    Mertens, Ph.; Altmann, H.; Hirai, T.; Philipps, V.; Pintsuk, G.; Rapp, J.; Riccardo, V.; Schweer, B.; Uytdenhouwen, I.; Samm, U.

    2009-06-01

    A bulk tungsten divertor row has been developed in the frame of the ITER-like Wall project at JET. It consists of 96 tiles grouped in 48 modules around the torus. The outer strike point is located on those tiles for most of the ITER-relevant, high triangularity plasmas. High power loads (locally up to 10-20 MW/m 2) and erosion rates are expected, even a risk of melting, especially with the transients or ELM loads. These are demanding conditions for an inertially cooled design as prescribed. A lamella design has been selected for the tungsten, arranged to control the eddy and halo current flows. The lamellae must also withstand high temperature gradients (2200 to 220 °C over 40 mm height), without overheating the supporting carrier (600-700 °C maximum). As a consequence of the tungsten emissivity, the radiative cooling drops appreciably in comparison with the current CFC tiles, calling for interleaved plasma scenarios in terms of performance. The compromise between shadowing and power handling is discussed, as well as the consequences for operation. Prototypes have been exposed in TEXTOR and in an electron beam facility (JUDITH-2) to the nominal power density of 7 MW/m 2 for 10 s and, in addition, to higher loads leading to surface temperatures above 2000 °C.

  8. RAMI Analysis for Designing and Optimizing Tokamak Cooling Water System (TCWS) for the ITER's Fusion Reactor

    SciTech Connect

    Ferrada, Juan J; Reiersen, Wayne T

    2011-01-01

    U.S.-ITER is responsible for the design, engineering, and procurement of the Tokamak Cooling Water System (TCWS). TCWS is designed to provide cooling and baking for client systems that include the first wall/blanket, vacuum vessel, divertor, and neutral beam injector. Additional operations that support these primary functions include chemical control of water provided to client systems, draining and drying for maintenance, and leak detection/localization. TCWS interfaces with 27 systems including the secondary cooling system, which rejects this heat to the environment. TCWS transfers heat generated in the Tokamak during nominal pulsed operation - 850 MW at up to 150 C and 4.2 MPa water pressure. Impurities are diffused from in-vessel components and the vacuum vessel by water baking at 200-240 C at up to 4.4 MPa. TCWS is complex because it serves vital functions for four primary clients whose performance is critical to ITER's success and interfaces with more than 20 additional ITER systems. Conceptual design of this one-of-a-kind cooling system has been completed; however, several issues remain that must be resolved before moving to the next stage of the design process. The 2004 baseline design indicated cooling loops that have no fault tolerance for component failures. During plasma operation, each cooling loop relies on a single pump, a single pressurizer, and one heat exchanger. Consequently, failure of any of these would render TCWS inoperable, resulting in plasma shutdown. The application of reliability, availability, maintainability, and inspectability (RAMI) tools during the different stages of TCWS design is crucial for optimization purposes and for maintaining compliance with project requirements. RAMI analysis will indicate appropriate equipment redundancy that provides graceful degradation in the event of an equipment failure. This analysis helps demonstrate that using proven, commercially available equipment is better than using custom-designed equipment

  9. Tungsten dust impact on ITER-like plasma edge

    SciTech Connect

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

    2015-01-12

    The impact of tungsten dust originating from divertor plates on the performance of edge plasma in ITER-like discharge is evaluated using computer modeling with the coupled dust-plasma transport code DUSTT-UEDGE. Different dust injection parameters, including dust size and mass injection rates, are surveyed. It is found that tungsten dust injection with rates as low as a few mg/s can lead to dangerously high tungsten impurity concentrations in the plasma core. Dust injections with rates of a few tens of mg/s are shown to have a significant effect on edge plasma parameters and dynamics in ITER scale tokamaks. The large impact of certain phenomena, such as dust shielding by an ablation cloud and the thermal force on tungsten ions, on dust/impurity transport in edge plasma and consequently on core tungsten contamination level is demonstrated. Lastly, it is also found that high-Z impurities provided by dust can induce macroscopic self-sustained plasma oscillations in plasma edge leading to large temporal variations of edge plasma parameters and heat load to divertor target plates.

  10. Tungsten dust impact on ITER-like plasma edge

    DOE PAGESBeta

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

    2015-01-12

    The impact of tungsten dust originating from divertor plates on the performance of edge plasma in ITER-like discharge is evaluated using computer modeling with the coupled dust-plasma transport code DUSTT-UEDGE. Different dust injection parameters, including dust size and mass injection rates, are surveyed. It is found that tungsten dust injection with rates as low as a few mg/s can lead to dangerously high tungsten impurity concentrations in the plasma core. Dust injections with rates of a few tens of mg/s are shown to have a significant effect on edge plasma parameters and dynamics in ITER scale tokamaks. The large impactmore » of certain phenomena, such as dust shielding by an ablation cloud and the thermal force on tungsten ions, on dust/impurity transport in edge plasma and consequently on core tungsten contamination level is demonstrated. Lastly, it is also found that high-Z impurities provided by dust can induce macroscopic self-sustained plasma oscillations in plasma edge leading to large temporal variations of edge plasma parameters and heat load to divertor target plates.« less

  11. Tungsten dust impact on ITER-like plasma edge

    NASA Astrophysics Data System (ADS)

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

    2015-01-01

    The impact of tungsten dust originating from divertor plates on the performance of edge plasma in ITER-like discharge is evaluated using computer modeling with the coupled dust-plasma transport code DUSTT-UEDGE. Different dust injection parameters, including dust size and mass injection rates, are surveyed. It is found that tungsten dust injection with rates as low as a few mg/s can lead to dangerously high tungsten impurity concentrations in the plasma core. Dust injections with rates of a few tens of mg/s are shown to have a significant effect on edge plasma parameters and dynamics in ITER scale tokamaks. The large impact of certain phenomena, such as dust shielding by an ablation cloud and the thermal force on tungsten ions, on dust/impurity transport in edge plasma and consequently on core tungsten contamination level is demonstrated. It is also found that high-Z impurities provided by dust can induce macroscopic self-sustained plasma oscillations in plasma edge leading to large temporal variations of edge plasma parameters and heat load to divertor target plates.

  12. Tungsten dust impact on ITER-like plasma edge

    SciTech Connect

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

    2015-01-15

    The impact of tungsten dust originating from divertor plates on the performance of edge plasma in ITER-like discharge is evaluated using computer modeling with the coupled dust-plasma transport code DUSTT-UEDGE. Different dust injection parameters, including dust size and mass injection rates, are surveyed. It is found that tungsten dust injection with rates as low as a few mg/s can lead to dangerously high tungsten impurity concentrations in the plasma core. Dust injections with rates of a few tens of mg/s are shown to have a significant effect on edge plasma parameters and dynamics in ITER scale tokamaks. The large impact of certain phenomena, such as dust shielding by an ablation cloud and the thermal force on tungsten ions, on dust/impurity transport in edge plasma and consequently on core tungsten contamination level is demonstrated. It is also found that high-Z impurities provided by dust can induce macroscopic self-sustained plasma oscillations in plasma edge leading to large temporal variations of edge plasma parameters and heat load to divertor target plates.

  13. Dual transmission grating based imaging radiometer for tokamak edge and divertor plasmas

    SciTech Connect

    Kumar, Deepak; Clayton, Daniel J.; Parman, Matthew; Stutman, Dan; Tritz, Kevin; Finkenthal, Michael

    2012-10-15

    The designs of single transmission grating based extreme ultraviolet (XUV) and vacuum ultraviolet (VUV) imaging spectrometers can be adapted to build an imaging radiometer for simultaneous measurement of both spectral ranges. This paper describes the design of such an imaging radiometer with dual transmission gratings. The radiometer will have an XUV coverage of 20-200 A with a {approx}10 A resolution and a VUV coverage of 200-2000 A with a {approx}50 A resolution. The radiometer is designed to have a spatial view of 16 Degree-Sign , with a 0.33 Degree-Sign resolution and a time resolution of {approx}10 ms. The applications for such a radiometer include spatially resolved impurity monitoring and electron temperature measurements in the tokamak edge and the divertor. As a proof of principle, the single grating instruments were used to diagnose a low temperature reflex discharge and the relevant data is also included in this paper.

  14. Developing a Roadmap for US Divertor and PMI Research in the ITER Era

    NASA Astrophysics Data System (ADS)

    Hill, D. N.; Lipschultz, B.; Whyte, D. G.; Garofalo, A. M.; Leonard, A. W.; Maingi, R.

    2013-10-01

    The role of existing and candidate future facilities for developing driven core, boundary plasma and plasma-facing components (PFCs) solutions for burning plasma experiments will be discussed in light of scientific and technical challenges, testing capabilities, scheduling implications, and cost. Present experiments point to likely integrated core-edge solutions which may enable steady-state high-gain, high power density operation; focused research on existing tokamak facilities could strengthen confidence significantly. In parallel, both existing and new candidate materials suitable for testing under high neutron fluence can be developed and qualified. We will also discuss the potential role of new facilities in closing the knowledge gaps to a Fusion Nuclear Science Facility (FNSF), and what form the final step of integrating core and edge solutions will be (separate, or as part of an FNSF) in terms of size, goals and cost. Supported by the US DOE under DE-AC52-07NA27344, DE-FC02-99ER54512, DE-SC00-02060, DE-FG02-04ER54762, DE-FC-02-04ER54698, and DE-AC02-09CH11466.

  15. Radiative divertor plasmas with convection in DIII-D

    SciTech Connect

    Leornard, A.W.; Porter, G.D.; Wood, R.D.

    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.

  16. A survey of problems in divertor and edge plasma theory

    SciTech Connect

    Boozer, A. ); Braams, B.; Weitzner, H. . Courant Inst. of Mathematical Sciences); Cohen, R. ); Hazeltine, R. . Inst. for Fusion Studies); Hinton, F. ); Houlberg, W. (Oak

    1992-12-22

    Theoretical physics problems related to divertor design are presented, organized by the region in which they occur. Some of the open questions in edge physics are presented from a theoretician's point of view. After a cursory sketch of the fluid models of the edge plasma and their numerical realization, the following topics are taken up: time-dependent problems, non-axisymmetric effects, anomalous transport in the scrape-off layer, edge kinetic theory, sheath effects and boundary conditions in divertors, electric field effects, atomic and molecular data issues, impurity transport in the divertor region, poloidally localized power dissipation (MARFEs and dense gas targets), helium ash removal, and neutral transport. The report ends with a summary of selected problems of particular significance and a brief bibliography of survey articles and related conference proceedings.

  17. A survey of problems in divertor and edge plasma theory

    SciTech Connect

    Boozer, A.; Braams, B.; Weitzner, H.; Cohen, R.; Hazeltine, R.; Hinton, F.; Houlberg, W.; Oktay, E.; Sadowski, W.; Post, D.; Sigmar, D.; Wootton, A.

    1992-12-22

    Theoretical physics problems related to divertor design are presented, organized by the region in which they occur. Some of the open questions in edge physics are presented from a theoretician`s point of view. After a cursory sketch of the fluid models of the edge plasma and their numerical realization, the following topics are taken up: time-dependent problems, non-axisymmetric effects, anomalous transport in the scrape-off layer, edge kinetic theory, sheath effects and boundary conditions in divertors, electric field effects, atomic and molecular data issues, impurity transport in the divertor region, poloidally localized power dissipation (MARFEs and dense gas targets), helium ash removal, and neutral transport. The report ends with a summary of selected problems of particular significance and a brief bibliography of survey articles and related conference proceedings.

  18. A novel approach to magnetic divertor configuration design

    NASA Astrophysics Data System (ADS)

    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.

  19. Plasma transport in a simulated magnetic-divertor configuration

    SciTech Connect

    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.

  20. Toroidally symmetric plasma vortex at tokamak divertor null point

    NASA Astrophysics Data System (ADS)

    Umansky, M. V.; Ryutov, D. D.

    2016-03-01

    Reduced MHD equations are used for studying toroidally symmetric plasma dynamics near the divertor null point. Numerical solution of these equations exhibits a plasma vortex localized at the null point with the time-evolution defined by interplay of the curvature drive, magnetic restoring force, and dissipation. Convective motion is easier to achieve for a second-order null (snowflake) divertor than for a regular x-point configuration, and the size of the convection zone in a snowflake configuration grows with plasma pressure at the null point. The trends in simulations are consistent with tokamak experiments which indicate the presence of enhanced transport at the null point.

  1. Turbulence studies in Tokamak boundary plasmas with realistic divertor geometry

    SciTech Connect

    Xu, X.Q.

    1998-10-14

    Results are presented from the 3D nonlocal electromagnetic turbulence code BOUT [1] and the linearized shooting code BAL[2] to study turbulence in tokamak boundary plasmas and its relationship to the L-H transition, in a realistic divertor plasma geometry. The key results include: (1) the identification of the dominant, resistive X-point mode in divertor geometry and (2) turbulence suppression in the L-H transition by shear in the ExB drift speed, ion diamagnetism and finite polarization. Based on the simulation results, a parameterization of the transport is given that includes the dependence on the relevant physical parameters.

  2. Non-ambipolar transport in a magnetic divertor

    SciTech Connect

    Strawitch, C M; Emmert, G A

    1980-02-01

    Plasma transport is studied in a simulated magnetic divertor in the Wisconsin single ring DC machine. The transport perpendicular and parallel to the magnetic field is shown to be non-ambipolar by a variety of measurements, but can be forced to be ambipolar by an appropriately designed divertor target plate. The density profile in the scrape-off zone agrees with the predictions of a one-dimensional diffusion equation that assumes classical cross-field transport and plasma flow parallel to the field at the local ion acoustic velocity.

  3. Vapor shielding models and the energy absorbed by divertor targets during transient events

    NASA Astrophysics Data System (ADS)

    Skovorodin, D. I.; Pshenov, A. A.; Arakcheev, A. S.; Eksaeva, E. A.; Marenkov, E. D.; Krasheninnikov, S. I.

    2016-02-01

    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 Emax. 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 Emax 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 Emax 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 to the target, and

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

    NASA Astrophysics Data System (ADS)

    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.

  5. ITER EDA project status

    NASA Astrophysics Data System (ADS)

    Chuyanov, V. A.

    1996-10-01

    The status of the ITER design is as presented in the Interim Design Report accepted by the ITER council for considerations by ITER parties. Physical and technical parameters of the machine, conditions of operation of main nuclear systems, corresponding design and material choices are described, with conventional materials selected. To fully utilize the safety and economical potential of fusion advanced materials are necessary. ITER shall and can be built with materials already available. The ITER project and advanced fusion material developments can proceed in parallel. The role of ITER is to establish (experimentally) requirements to these materials and to provide a test bed for their final qualification in fusion reactor environment. To achieve this goal, the first wall/blanket modules test program is foreseen.

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

    SciTech Connect

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

    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.

  7. The Corrected Simulation Method of Critical Heat Flux Prediction for Water-Cooled Divertor Based on Euler Homogeneous Model

    NASA Astrophysics Data System (ADS)

    Zhang, Jingyang; Han, Le; Chang, Haiping; Liu, Nan; Xu, Tiejun

    2016-02-01

    An accurate critical heat flux (CHF) prediction method is the key factor for realizing the steady-state operation of a water-cooled divertor that works under one-sided high heating flux conditions. An improved CHF prediction method based on Euler's homogeneous model for flow boiling combined with realizable k-ɛ model for single-phase flow is adopted in this paper in which time relaxation coefficients are corrected by the Hertz-Knudsen formula in order to improve the calculation accuracy of vapor-liquid conversion efficiency under high heating flux conditions. Moreover, local large differences of liquid physical properties due to the extreme nonuniform heating flux on cooling wall along the circumference direction are revised by formula IAPWS-IF97. Therefore, this method can improve the calculation accuracy of heat and mass transfer between liquid phase and vapor phase in a CHF prediction simulation of water-cooled divertors under the one-sided high heating condition. An experimental example is simulated based on the improved and the uncorrected methods. The simulation results, such as temperature, void fraction and heat transfer coefficient, are analyzed to achieve the CHF prediction. The results show that the maximum error of CHF based on the improved method is 23.7%, while that of CHF based on uncorrected method is up to 188%, as compared with the experiment results of Ref. [12]. Finally, this method is verified by comparison with the experimental data obtained by International Thermonuclear Experimental Reactor (ITER), with a maximum error of 6% only. This method provides an efficient tool for the CHF prediction of water-cooled divertors. supported by the National Magnetic Confinement Fusion Science Program of China (No. 2010GB104005) and National Natural Science Foundation of China (No. 51406085)

  8. Neoclassical and Initial Divertor-Geometry Tests of COGENT

    NASA Astrophysics Data System (ADS)

    Cohen, R. H.; Dorf, M.; Compton, J. C.; Dorr, M.; Rognlien, T. D.; Colella, P.; McCorquodale, P.; Angus, J.; Krasheninnikov, S.

    2012-03-01

    COGENT is a full-f continuum kinetic code being developed for study of edge physics phenomena in tokamaks. The code is distinguished by 4th order conservative discretization and mapped multiblock grid technology to handle the geometric complexity of the tokamak edge. We discuss a number of recent neoclassical results in closed-flux-surface geometry, in particular self-consistent neoclassical simulations with increasingly complete collision operators (Lorentz, full test-particle, and adding model momentum- and energy-conserving terms). We also examine the effects of strong radial electric fields on neoclassical transport and decay of geodesic acoustic modes (GAM's). The code is being upgraded to full single-null divertor geometry, with numerical geometric coefficients imported from an external MHD equilibrium calculation. We discuss several initial tests of the divertor code: advection of phase-space blobs through the x-point region, and neoclassical transport and flows in the presence of divertor losses. We also summarize progress on code-development activities needed to complete the divertor code.

  9. Theoretical design of a compact energy recovering divertor

    NASA Astrophysics Data System (ADS)

    Baver, D. A.

    2015-11-01

    An energy recovering divertor (ERD) is a type of plasma direct converter (PDC) designed to fit in the divertor channel of a tokamak. Such a device reduces the heat load to the divertor plate by converting a portion of it into electrical energy. This recovered energy can then be used for auxiliary heating and current drive, fundamentally altering the relationship between scientific and engineering breakeven and reducing dependence on bootstrap current. Previous work on the ERD concept focused on amplification of Alfven waves in a manner similar to a free-electron laser. While conceptually straightforward, this concept was also bulky, thus limiting its applicability to existing tokamak experiments. A design is presented for an ERD based on sheath-localized waves. This makes possible a device sufficiently compact to fit in the divertor channel of many existing tokamak experiments, and moreover requires no new shaping coils to achieve the desired magnetic geometry or topology. In addition, incidental advantages of this concept will be discussed.

  10. Overview of Recent Developments in Pellet Injection for ITER

    SciTech Connect

    Combs, Stephen Kirk; Baylor, Larry R; Meitner, Steven J; Caughman, John B; Rasmussen, David A; Maruyama, So

    2012-01-01

    Pellet injection is the primary fueling technique planned for core fueling of ITER burning plasmas. Also, the injection of relatively small pellets to purposely trigger rapid small edge localized modes (ELMs) has been proposed as a possible solution to the heat flux damage from larger natural ELMs likely to be an issue on the ITER divertor surfaces. The ITER pellet injection system is designed to inject pellets into the plasma through both inner and outer wall guide tubes. The inner wall guide tubes will provide high throughput pellet fueling while the outerwall guide tubes will be used primarily to trigger ELMs at a high frequency (>15 Hz). The pellet fueling rate ofeach injector is to be up to 120 Pa-m3/s, which will require the formation of solid D-T at a volumetric rate of ~1500 mm3/s. Two injectors are to be provided for ITER at the startup with a provision for up to six injectorsduring the D-T phase. The required throughput of each injector is greater than that of any injector built to date, and a novel twin-screw continuous extrusion system is being developed to meet the challenging design parameters. Status of the development activities will be presented, highlighting recent progress.