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Sample records for resolved iter divertor

  1. Divertor interferometer diagnostic for ITER

    SciTech Connect

    Brower, D. L.; Deng, B. H.; Ding, W. X.

    2006-10-15

    In the harsh environment of the divertor region in ITER, plasmas spanning a huge density range from 10{sup 19} to 10{sup 22} m{sup -3} are anticipated making measurement of the electron density particularly challenging. For any reasonable wavelength choice, the total phase measured by a conventional two-color interferometer system is always >>2{pi} and therefore subject to fringe counting errors. This problem can be remedied by adding a polarimeter capability whereby the Cotton-Mouton effect is measured or by employing differential interferometry. Using either approach, the total phase is always <<2{pi}. The conceptual design of an interferometer system along with possible wavelength choices will be explored.

  2. ELM heat flux in the ITER divertor

    SciTech Connect

    Leonard, A.W.; Osborne, T.H.; Hermann, A.; Suttrop, W.; Itami, K.; Lingertat, J.; Loarte, A.

    1998-07-01

    Edge-Localized-Modes (ELMs) have the potential to produce unacceptable levels of erosion of the ITER divertor. Ablation of the carbon divertor target will occur if the surface temperature rises above about 2,500 C. Because a large number of ELMs, {ge}1000, are expected in each discharge it is important that the surface temperature rise due to an individual ELM remain below this threshold. Calculations that have been carried out for the ITER carbon divertor target indicate ablation will occur for ELM energy {ge}0.5MJ/m{sup 2} if it is deposited in 0.1 ms, or 1.2 MJ/m{sup 2} if the deposition time is 1.0 ms. Since {Delta}T{proportional_to}Q{Delta}t{sup {minus}1/2}, an ablation threshold can be estimated at Q{Delta}t{sup {minus}1/2}{approx}45 MJm{sup {minus}2} s{sup {minus}1/2} where Q is the divertor ELM energy density in J-m{sup {minus}2} and {Delta}t is the time in seconds for that deposition. If a significant fraction of ELMs exceed this threshold then an unacceptable level of erosion may take place. The ablation parameter in ITER can be determined by scaling four factors from present experiments: the ELM energy loss from the core plasma, the fraction of ELM energy deposited on the divertor target, the area of the ELM profile onto the target, and the time for the ELM deposition. ELM data from JET, ASDEX-Upgrade, JT-60U, DIII-D and Compass-D have been assembled by the ITER Divertor Modeling and Database expert group into a database for the purpose of predicting these factors for ELMs in the ITER divertor.

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

    NASA Astrophysics Data System (ADS)

    Schmid, K.

    2008-10-01

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

  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. Prediction of Pressure Drop in the ITER Divertor Cooling Channels

    SciTech Connect

    Yin, S.T.; Chen, J.L.

    2005-04-15

    This study investigated the pressure drop in the divertor cooling channels of the International Thermonuclear Experimental Reactor (ITER). The water in the cooling channels will encounter the following flow and boiling regimes: 1) single-phase convection, 2) highly-subcooled boiling, 3) onset of nucleate boiling (ONB), and 4) fully-developed subcooled boiling. The upper operating boundary is limited by the departure from nucleate boiling (DNB) or burnout conditions. Twisted-tape insert will be used to enhance local heat transfer. Analytical models, validated with relevant databases, were proposed for the above-identified flow regimes. A user-friendly computer code was developed to calculate the overall pressure drop and the exit pressure of a specific local segment throughout the entire flow circuit. Although the operating parameters were based on the CDA phase input the results are found in general agreement when compared with the ITER EDA results.

  7. Surface heat loads on the ITER divertor vertical targets

    NASA Astrophysics Data System (ADS)

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

    2017-04-01

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

  8. 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

  9. Evaluation of a monoblock divertor design for the ITER tokamak

    SciTech Connect

    Lee, Y.T.; Hoffman, M.A.; Hafez, M.

    1996-12-31

    A subcooled nucleate boiling computer code (with 3D heat conduction in solid and 1D forced convection in fluid) that incorporates a good estimation of the single-phase and two-phase pressure drop was developed to evaluate a monoblock design of the divertor with smooth tubes as well as a wide variety of cooling designs. Using one of the monoblock divertor designs proposed by the European International Thermonuclear Experimental Reactor (ITER) team as of March 1995, it was found that under a normal steady state operating condition with a peak heat flux of about 5 MW/m{sup 2}, the water flow remained in the single phase liquid regime. Under an abnormal operating condition with a peak heat flux of about 20 MW/m{sup 2}, the partially developed boiling (PDB) regime occurred where the local critical heat flux safety factor (SF{sub CHF}=V@CHF(z)/q{sub ({theta}}=0{degree})), was estimated to be about 1.4 using the Tong-75 CHF correlation. This indicates that further increases in the magnitude of the heat flux beyond 20 MW/m{sup 2} may raise safety concerns for the design. By increasing the mass flux, decreasing the inlet water temperature, or increasing the inlet water pressure, the CHF safety margin of the design can be increased without inserting twisted tapes inside cooling tubes. 8 refs., 6 figs.

  10. The impact of ELMs on the ITER divertor

    SciTech Connect

    Leonard, A.W.; Osborne, T.H.; Suttrop, W.; Hermann, A.; Itami, K.; Lingertat, J.; Loarte, A.

    1998-07-01

    Edge-Localized-Modes (ELMs) are expected to present a significant transient flux of energy and particles to the ITER divertor. The threshold for ablation of the graphite target will be reached if the ELM transient exceeds Q/t{sup 1/2} {approximately} 45 MJ-m{sup {minus}2}-s{sup {minus}1/2} where Q is the ELM deposition energy density and t is the ELM deposition time. The ablation parameter in ITER can be determined by scaling four factors from present experiments: the ELM energy loss from the core plasma, the fraction of ELM energy deposited on the divertor target, the area of the ELM profile onto the target, and finally the time for the ELM deposition. Review of the ELM energy loss of Type 1 ELM data suggests an ITER ELM energy loss of 2--6% of the stored energy or 25--80 MJ. The fraction of heating power crossing the separatrix due to ELMs is nearly constant (20--40%) resulting in an inverse relationship between ELM amplitude and frequency. Measurements on DIII-D and ASDEX-Upgrade indicate that 50--80% of the ELM energy is deposited on the target. There is currently no evidence for a large fraction of the ELM energy being dissipated through radiation. Profiles of the ELM heat flux are typically 1--2 times the width of steady heat flux between ELMs, with the ELM amplitude usually larger on the inboard target. The ELM deposition time varies from about 0.1 ms in JET to as high as 1.0 ms in ASDEX-Upgrade and DIII-D. The ELM deposition time for ITER will depend upon the level of conductive versus convective transport determined by the ratio of energy to particles released by the ELM. Preliminary analysis suggests that large Type 1 ELMs for low recycling H-mode may exceed the ablation parameter by a factor of 5. Promising regimes with smaller ELMS have been found at other edge operational regimes, including high density with gas puffing, use of rf heating and operation with Type 3 ELMs.

  11. Effects of ELMs on ITER divertor armour materials

    NASA Astrophysics Data System (ADS)

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

    2007-06-01

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

  12. ITER divertor performance in the low-activation phase

    NASA Astrophysics Data System (ADS)

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

    2013-12-01

    The paper presents results of SOLPS modelling of the edge plasma performance during the low-activation phase of ITER operation. The calculations show that the peak power loading of the divertor targets can reach the reactor-relevant level of 3 to 5 MW m-2, even without the fusion reactions, rendering commissioning of the high heat flux components possible in this phase. Parametrization of the output of the SOLPS runs for the predominantly helium plasma concerned by the studies reported here is performed, thus providing the boundary conditions for modelling of the core and allowing efficient integration of the core and edge models. This approach, using the ASTRA code for core simulations, is applied to the analysis of hydrogen accumulation in helium plasmas due to H pellet injection. The latter is the only available option for early testing of ELM pace-making as an ELM control tool assuming H-mode in hydrogen will not be possible. Critical dilution with H down to 70% He in the core plasma can be reached in only 0.5 to 1 s or even shorter, depending on the assumptions made.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. Thermal-hydraulic design issues and analysis for the ITER (International Thermonuclear Experimental Reactor) divertor

    SciTech Connect

    Koski, J.A.; Watson, R.D. ); Hassanien, A.M. ); Goranson, P.L. . Fusion Engineering Design Center); Salmonson, J.C. . Special Projects)

    1990-01-01

    Critical Heat Flux (CHF), also called burnout, is one of the major design limits for water-cooled divertors in tokamaks. Another important design issue is the correct thermal modeling of the divertor plate geometry where heat is applied to only one side of the plate and highly subcooled flow boiling in internal passages is used for heat removal. This paper discusses analytical techniques developed to address these design issues, and the experimental evidence gathered in support of the approach. Typical water-cooled divertor designs for the International Thermonuclear Experimental Reactor (ITER) are analyzed, and design margins estimated. Peaking of the heat flux at the tube-water boundary is shown to be an important issue, and design concerns which could lead to imposing large design safety margins are identified. The use of flow enhancement techniques such as internal twisted tapes and fins are discussed, and some estimates of the gains in the design margin are presented. Finally, unresolved issues and concerns regarding hydraulic design of divertors are summarized, and some experiments which could help the ITER final design process identified. 23 refs., 10 figs.

  19. Hydrogen lines in the infrared region and spectral background for the thomson scattering diagnostics of the iter divertor plasma

    NASA Astrophysics Data System (ADS)

    Lisitsa, V. S.; Mukhin, E. E.; Kadomtsev, M. B.; Kukushkin, A. B.; Kukushkin, A. S.; Kurskiev, G. S.; Levashova, M. G.; Tolstyakov, S. Yu.

    2012-02-01

    Calculations are made of the plasma spectral background, which is important for the Thomson scattering diagnostics in the ITER divertor. Theoretical grounds have been elaborated for computing the hydrogen spectral line shapes in the infrared spectral region for a divertor plasma in ITER. The shape of the P-7 Paschen line (transition n = 7 → n = 3) located near the laser scattering signal has been calculated for various lines of sight in the ITER divertor. Contributions from different mechanisms of broadening the P-7 line have been examined. The spectral intensities of bremsstrahlung and photorecombination continuum have been calculated. All calculations use data on the spatial distribution of temperatures and densities of all species of plasma particles computed with the SOLPS4.3 code for basic operation regimes of the ITER divertor.

  20. Divertor stray light analysis in JET-ILW and implications for the H-α diagnostic in ITER

    NASA Astrophysics Data System (ADS)

    Kukushkin, A. B.; Neverov, V. S.; Stamp, M. F.; Alekseev, A. G.; Brezinsek, S.; Gorshkov, A. V.; von Hellermann, M.; Kadomtsev, M. B.; Kotov, V.; Kukushkin, A. S.; Levashova, M. G.; Lisgo, S. W.; Lisitsa, V. S.; Shurygin, V. A.; Veshchev, E.; Vukolov, D. K.; Vukolov, K. Yu.; JET EFDA Contributors

    2014-08-01

    We report on the first results for the spectrum of divertor stray light (DSL) and the signal-to-background ratio for D-α light emitted from the far SOL and divertor in JET in the recent ITER-like wall (ILW) campaign. The results support the expectation of a strong impact of DSL upon the H-alpha (and Visible Light) Spectroscopy Diagnostic in ITER.

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

    NASA Astrophysics Data System (ADS)

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

    2007-06-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2017-04-01

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

  3. 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.

  4. 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.

  5. 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.

  6. Effects of ELMs and disruptions on ITER divertor armour materials

    NASA Astrophysics Data System (ADS)

    Federici, G.; Zhitlukhin, A.; Arkhipov, N.; Giniyatulin, R.; Klimov, N.; Landman, I.; Podkovyrov, V.; Safronov, V.; Loarte, A.; Merola, M.

    2005-03-01

    This paper describes the response of plasma facing components manufactured with tungsten (macro-brush) and CFC to energy loads characteristic of Type I ELMs and disruptions in ITER, in experiments conducted (under an EU/RF collaboration) in two plasma guns (QSPA and MK-200UG) at the TRINITI institute. Targets were exposed to a series of repetitive pulses in QSPA with heat loads in a range of 1-2 MJ/m 2 lasting 0.5 ms. Moderate tungsten erosion, of less than 0.2 μm per pulse, was found for loads of ˜1.5 MJ/m 2, consistent with ELM erosion being determined by tungsten evaporation and not by melt layer displacement. At energy densities of ˜1.8 MJ/m 2 a sharp growth of tungsten erosion was measured together with intense droplet ejection. MK-200UG experiments were focused on studying mainly vapor plasma production and impurity transport during ELMs. The conditions for removal of thin metal deposits from a carbon substrate were characterized.

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

    SciTech Connect

    Youchison, D.L.; Marshall, T.D.; McDonald, J.M.; Lutz, T.J.; Watson, R.D.; Driemeyer, D.E. Kubik, D.L.; Slattery, K.T.; Hellwig, T.H.

    1997-09-01

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

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

    NASA Astrophysics Data System (ADS)

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

    1997-12-01

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

  9. 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.

  10. 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.

  11. Thermal transients due to plasma sweeping on the monoblock divertor plate for iter

    SciTech Connect

    Renda, V.; Papa, L.; Soria, A. . Joint Research Centre)

    1992-12-01

    In this paper in the framework of the feasibility studies of the International Tokamak Experimental Reactor (ITER), the thermal behavior of the monoblock divertor plate has been investigated at the Joint Research Centre of the Commission of the European Communities. The design consists of cooling tubes embedded in a protective armor of graphite, a material that has given good results in plasma physics experiments. Previous parametric studies, based on a thermal flux peak of 15 MW/m[sup 2] and different materials, led to the choice of a Mo-Re alloy for the tubes and a high-conductivity carbon-fiber composite called SEP for the graphite armor. To comply with a design temperature of 1273 K, an allowable protective layer only 5 mm thick was indicated; however, because of the high erosion rate due to sputtering, the lifetime of such a plate would be unacceptable from an engineering standpoint. To overcome this difficulty, it has been proposed that the separatrix be swept to lower the flux peak during the transient. The nominal working condition then becomes a sweeping of the separatrix moving around the null point with a radius of 40 mm and frequency of 0.3 Hz: this generates a thermal load varying in time on the divertor plates. The results lead to the conclusion that plasma sweeping can reduce the surface temperature peak of the divertor, allowing a 16-mm-thick protective layer of the armor.

  12. Preliminary Monte Carlo simulation of beryllium migration during JET ITER-like wall divertor operation

    NASA Astrophysics Data System (ADS)

    Airila, M. I.; Järvinen, A.; Groth, M.; Belo, P.; Wiesen, S.; Brezinsek, S.; Lawson, K.; Borodin, D.; Kirschner, A.; Coad, J. P.; Heinola, K.; Likonen, J.; Rubel, M.; Widdowson, A.

    2015-08-01

    Migration of beryllium into the divertor and deposition on tungsten in the final phase of the first ITER-like-wall campaign of JET are modelled with the 3D Monte Carlo impurity transport code ERO. The simulation covers the inner wall and the inner divertor. To generate the plasma background for Monte Carlo tracing of impurity particles, we use the EDGE2D/EIRENE code set. At the relevant regions of the wall, the estimated plasma conditions vary around Te ≈ 5eV and ne ≈ 2 ×1017m-3 (far-scrape-off layer; more than 10 cm away from the LCFS). We calculate impurity distributions in the plasma using the main chamber source as a free parameter in modelling and attempt to reproduce inter-ELM spectroscopic Be II line (527 nm) profiles at the divertor. The present model reproduces the level of emission close to the inner wall, but further work is needed to match also the measured emission peak values and ultimately link the modelled poloidal net deposition profiles of beryllium to post mortem data.

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

    SciTech Connect

    Xu, J. C.; Jia, M. N.; Feng, W.; Deng, G. Z.; 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.; Hu, L. Q.; Wan, B. N.; Li, J.; Sun, Y. W.; Guo, H. Y.

    2016-08-15

    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.

  14. 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.

  15. 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.

  16. ITER divertor sputtering erosion -- recent analysis for carbon, beryllium, tungsten, and niobium surfaces

    SciTech Connect

    Brooks, J.N.

    1991-07-01

    ITER divertor plate sputtering erosion has been analyzed using current design information and updated impurity transport models. The REDEP erosion/redeposition code was used to compute erosion for a very low plasma divertor temperature (T{sub e{sub 0}} {approx} 12 eV) physics phase reference'' case, and for other plasma conditions. A high surface temperature case (T{sub s{sub 0}} = 1800{degree}C) is analyzed for a carbon surface. Niobium is analyzed using WBC near-surface transport code results for the redeposited charge state. The REDEP results show high net erosion rates ({approx gt} 20 cm/burn{sm bullet}yr) for beryllium and carbon, even at low plasma temperatures. Net erosion rates are low to moderate for niobium ({approximately}0-3 cm/burn{sm bullet}yr), depending on plasma conditions, and low for tungsten ({approximately}0-0.2 cm/burn{sm bullet}yr). 9 refs., 2 figs.

  17. International Thermonuclear Experimental Reactor (ITER) divertor plate performance and lifetime considerations

    SciTech Connect

    Mattas, R.F.

    1990-03-01

    The ITER divertor plate performance during the technology phase of operation has been analyzed. High-Z materials, such as tungsten and tantalum, have been considered as plasma side materials, and refractory metal alloys, Ta-10W, TZM, Nb-1Zr, and V-15Cr-5Ti, plus copper alloys have been considered as the structural materials. The fatigue lifetime have been predicted for structural plates and for duplex plates with the plasma side material bonded to the structure. The results indicate that refractory alloys have a comparable or improved performance to copper alloys. Peak allowable heat fluxes for these analyses are in the range of 15--20 MW/m{sup 2} for 2 mm thick structural plates and 7--11 MW/m{sup 2} for 4 mm thick duplex plates. 4 refs., 55 figs., 6 tabs.

  18. Surface composition and structure of divertor tiles following the JET tokamak operation with the ITER-like wall

    NASA Astrophysics Data System (ADS)

    Lagoyannis, A.; Tsavalas, P.; Mergia, K.; Provatas, G.; Triantou, K.; Tsompopoulou, E.; Rubel, M.; Petersson, P.; Widdowson, A.; Harissopulos, S.; Mertzimekis, T. J.; JET contributors, the

    2017-07-01

    Samples extracted from several divertor tiles following the 2011-2012 operation of JET with the ITER-Like wall were analyzed using ion beam analysis methods, x-ray fluorescence spectroscopy, scanning electron microscopy with energy dispersive spectroscopy analysis and x-ray diffraction. The emphasis was on the determination of light species and on material mixing including compound formation on the bottom and the outer divertor tiles. Deposition of deuterium, beryllium, carbon, nitrogen, oxygen, iron, chromium, nickel and molybdenum has been detected on all studied tiles. The thickest deposition, of around 4 µm, was measured on the bottom of the outer divertor, whereas the other surfaces (inner bottom and vertical outer) the co-deposits were around 1 µm. x-ray diffraction measurements have revealed the formation of the compound W2C on all specimens.

  19. 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.

  20. 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.

  1. The heat removal capability of actively cooled plasma-facing components for the ITER divertor

    NASA Astrophysics Data System (ADS)

    Missirlian, M.; Richou, M.; Riccardi, B.; Gavila, P.; Loarer, T.; Constans, S.

    2011-12-01

    Non-destructive examination followed by high-heat-flux testing was performed for different small- and medium-scale mock-ups; this included the most recent developments related to actively cooled tungsten (W) or carbon fibre composite (CFC) armoured plasma-facing components. In particular, the heat-removal capability of these mock-ups manufactured by European companies with all the main features of the ITER divertor design was investigated both after manufacturing and after thermal cycling up to 20 MW m-2. Compliance with ITER requirements was explored in terms of bonding quality, heat flux performances and operational compatibility. The main results show an overall good heat-removal capability after the manufacturing process independent of the armour-to-heat sink bonding technology and promising behaviour with respect to thermal fatigue lifetime under heat flux up to 20 MW m-2 for the CFC-armoured tiles and 15 MW m-2 for the W-armoured tiles, respectively.

  2. 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)

  3. Results of high heat flux tests of tungsten divertor targets under plasma heat loads expected in ITER and tokamaks (review)

    NASA Astrophysics Data System (ADS)

    Budaev, V. P.

    2016-12-01

    Heat loads on the tungsten divertor targets in the ITER and the tokamak power reactors reach 10MW m-2 in the steady state of DT discharges, increasing to 0.6-3.5 GW m-2 under disruptions and ELMs. The results of high heat flux tests (HHFTs) of tungsten under such transient plasma heat loads are reviewed in the paper. The main attention is paid to description of the surface microstructure, recrystallization, and the morphology of the cracks on the target. Effects of melting, cracking of tungsten, drop erosion of the surface, and formation of corrugated and porous layers are observed. Production of submicron-sized tungsten dust and the effects of the inhomogeneous surface of tungsten on the plasma-wall interaction are discussed. In conclusion, the necessity of further HHFTs and investigations of the durability of tungsten under high pulsed plasma loads on the ITER divertor plates, including disruptions and ELMs, is stressed.

  4. 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 RMP

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

    DOE PAGES

    Chang, Choong Seock; Ku, Seung -Hoe; Loarte, Alberto; ...

    2017-07-11

    Here, the XGC1 edge gyrokinetic code is used to study the width of the heat-flux to divertor plates in attached plasma condition. The flux-driven simulation is performed until an approximate power balance is achieved between the heat-flux across the steep pedestal pressure gradient and the heat-flux on the divertor plates.

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

    NASA Astrophysics Data System (ADS)

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

    2017-04-01

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

  7. The Joint European Torus (JET) pumped divertor results and their significance for the International Thermonuclear Experimental Reactor (ITER)

    NASA Astrophysics Data System (ADS)

    Watkins, M. L.; JET Team

    1996-05-01

    The effectiveness of the pumped divertor during the 1994/95 experimental campaign of the Joint European Torus (JET) [P.-H. Rebut, R. J. Bickerton, and B. E. Keen, Nucl. Fusion 25, 1011 (1985)] has allowed the pursuit of a broad-based research program that is highly relevant to the International Thermonuclear Experimental Reactor (ITER) [K. Tomabechi and the ITER Team, Nucl. Fusion 31, 1135 (1991)]. High-performance hot-ion discharges with high confinement (H-modes) free of edge localized modes (ELMs) have set a JET record neutron rate in deuterium, but are limited by various magnetohydrodynamic (MHD) phenomena to βN<1.8, where βN=β/(I/aB), β is the ratio of the plasma pressure to the toroidal field pressure, I is the plasma current, B is the toroidal field, and a is the horizontal minor radius of the plasma. Quasi-steady-state ELMy H-modes have also been studied at high power, high current, and high β. The underlying energy transport exhibits a gyro-Bohm dependence that is lost close to the H-mode threshold and at high β. ELMy H-modes with detached divertor plasmas and radiative power exhaust (the operating regime foreseen for ITER) reduce the power loading to the targets, but at the expense of main plasma confinement and purity. Beryllium has been compared with carbon fiber composite as a divertor target material and melting has been induced at ITER reference off-normal heat loads, but only a moderate degree of self-protection of the beryllium target was found.

  8. ELM resolved measurement of fuel recycling on divertor targets in DIII-D

    NASA Astrophysics Data System (ADS)

    Bykov, I.; Hollmann, E. M.; Moyer, R. A.; Watkins, J. G.; Makowski, M.; Lasnier, C. S.; McLean, A.; Wang, H.

    2016-10-01

    Simultaneous measurements of different atomic and molecular contributions are important for determining D recycling from plasma-facing components (PFCs). A splitted filtered imaging of visible-range molecular and atomic emission was applied for the first time for synchronous measurements of Dα (656 nm), D2 Fulcher- α band (600 nm), and CD (430 nm) emissions in the strike point region of the lower divertor in DIII-D. Framing rate up to 1 kHz was sufficient to resolve intra- and inter-ELM phases of H-mode discharges. Radial profiles of atomic (molecular) fluxes of recycled D were deduced using respective S(D)/XB rate coefficients. We present the results of particle flux measurements for a series of shots with varying densities (n/nGW = 0.5-0.8), which affected the degree of the divertor detachment and the balance between individual channels of D recycling from PFCs. Supported by the US DOE under DE-FG02-07ER54917, DE-FG02-04ER54758, DE-FC02-04ER54698, DE-FG03-95ER54309, and DE-FG02-04ER54762.

  9. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2017-07-01

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

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

    DOE PAGES

    Garofalo, Andrea M.; Gong, X. Z.; Qian, J.; ...

    2017-06-07

    Recent experiments on EAST have achieved the first long pulse H-mode (61 s) with zero loop voltage and an ITER-like tungsten divertor, and have demonstrated access to broad plasma current profiles by increasing the density in fully-noninductive lower hybrid current-driven discharges. These long pulse discharges reach wall thermal and particle balance, exhibit stationary good confinement (H98y2~1.1) with low core electron transport, and are only possible with optimal active cooling of the tungsten armors. In separate experiments, the electron density was systematically varied in order to study its effect on the deposition profile of the external lower hybrid current drive (LHCD),more » while keeping the plasma in fully-noninductive conditions and with divertor strike points on the tungsten divertor. A broadening of the current profile is found, as indicated by lower values of the internal inductance at higher density. A broad current profile is attractive because, among other reasons, it enables internal transport barriers at large minor radius, leading to improved confinement as shown in companion DIII-D experiments. These experiments strengthen the physics basis for achieving high performance, steady state discharges in future burning plasmas.« less

  12. 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.; Youchison, D.L.

    1995-12-31

    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 7.6 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 C, 1 m/s, and 1 MPa, q{sub i}{sup CHF} and q{sub i}{sup BO} were 600 and 1,100 W/cm{sup 2}, respectively, which gave an {eta} of 1.8.

  13. 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.

  14. Modelling of steady state erosion of CFC actively water-cooled mock-up for the ITER divertor

    NASA Astrophysics Data System (ADS)

    Ogorodnikova, O. V.

    2008-04-01

    Calculations of the physical and chemical erosion of CFC (carbon fibre composite) monoblocks as outer vertical target of the ITER divertor during normal operation regimes have been done. Off-normal events and ELM's are not considered here. For a set of components under thermal and particles loads at glancing incident angle, variations in the material properties and/or assembly of defects could result in different erosion of actively-cooled components and, thus, in temperature instabilities. Operation regimes where the temperature instability takes place are investigated. It is shown that the temperature and erosion instabilities, probably, are not a critical point for the present design of ITER vertical target if a realistic variation of material properties is assumed, namely, the difference in the thermal conductivities of the neighbouring monoblocks is 20% and the maximum allowable size of a defect between CFC armour and cooling tube is +/-90° in circumferential direction from the apex.

  15. Results of high heat flux tests of tungsten divertor targets under plasma heat loads expected in ITER and tokamaks (review)

    SciTech Connect

    Budaev, V. P.

    2016-12-15

    Heat loads on the tungsten divertor targets in the ITER and the tokamak power reactors reach ~10MW m{sup −2} in the steady state of DT discharges, increasing to ~0.6–3.5 GW m{sup −2} under disruptions and ELMs. The results of high heat flux tests (HHFTs) of tungsten under such transient plasma heat loads are reviewed in the paper. The main attention is paid to description of the surface microstructure, recrystallization, and the morphology of the cracks on the target. Effects of melting, cracking of tungsten, drop erosion of the surface, and formation of corrugated and porous layers are observed. Production of submicron-sized tungsten dust and the effects of the inhomogeneous surface of tungsten on the plasma–wall interaction are discussed. In conclusion, the necessity of further HHFTs and investigations of the durability of tungsten under high pulsed plasma loads on the ITER divertor plates, including disruptions and ELMs, is stressed.

  16. 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.

  17. 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.

  18. Development of ITER Divertor Vertical Target with Annular Flow Concept - I: Thermal-Hydraulic Characteristics of Annular Swirl Tube

    SciTech Connect

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

    2004-12-15

    Thermal-hydraulic tests for pressurized water in an annular tube with a twist fin have been performed to examine its applicability to high-heat-flux components of the International Thermonuclear Experimental Reactor (ITER) divertor. The annular swirl tube consists of two concentric tubes: an outer smooth tube and an inner tube with an external twist fin to enhance heat transfer of the cooling water in the annulus section between the outer and the inner tubes. Critical heat flux (CHF) tests under one-sided-heating conditions show that the annular swirl tube has as high removal limitation as the conventional swirl tube, the dimensions of which are similar to those of the outer tube of the annular swirl tube. A minimum axial velocity of 7.1 m/s is required for 28 MW/m{sup 2}, the ITER design value. Pressure drops in the annulus section and the end return have been measured. The applicability of the existing correlations for heat transfer and CHF to the annular swirl tube has also been examined.

  19. Thermal fatigue testing of a diffusion-bonded beryllium divertor mock-up under ITER-relevant conditions

    SciTech Connect

    Youchison, D.L.; Watson, R.D.; McDonald, J.M.

    1996-07-01

    Thermal response and thermal fatigue tests of four 5-mm-thick beryllium tiles on a Russian Federation International Thermonuclear Experimental Reactor (ITER)-relevant divertor mock-up were completed on the electron beam test system at Sandia National Laboratories. Thermal response tests were performed on the tiles to an absorbed heat flux of 5 MW/m{sup 2} and surface temperatures near 300{degree}C using 1.4 MPa water at 5 m/s flow velocity and an inlet temperature of 8 to 15{degree}C. One tile was exposed to incrementally increasing heat fluxes up to 9.5 MW/m{sup 2} and surface temperatures up to 690{degree}C before debonding at 10MW/m{sup 2}. A second tile debonded in 25 to 30 cycles at <0.5 MW/m{sup 2}. However, a third tile debonded after 9200 thermal fatigue cycles at 5 MW/m{sup 2}, while another debonded after 6800 cycles. Posttest surface analysis indicated that fatigue failure occurred in the intermetallic layers between the beryllium and copper. No fatigue cracking of the bulk beryllium was observed. It appears that microcracks growing at the diffusion bond produced the observed gradual temperature increases during thermal cycling. These experiments indicate that diffusion-bonded beryllium tiles can survive several thousand thermal cycles under ITER-relevant conditions. However, the reliability of the diffusion-bonded joint remains a serious issue. 17 refs., 25 figs., 6 tabs.

  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. Thermal fatigue testing of a diffusion-bonded beryllium divertor mock-up under ITER relevant conditions

    SciTech Connect

    Youchison, D.L.; Guiniiatouline, R.; Watson, R.D.

    1994-12-31

    Thermal response and thermal fatigue tests of four 5 mm thick beryllium tiles on a Russian divertor mock-up were completed on the Electron Beam Test System at Sandia National Laboratories. The beryllium tiles were diffusion bonded onto an OFHC copper saddleblock and a DSCu (MAGT) tube containing a porous coating. Thermal response tests were performed on the tiles to an absorbed heat flux of 5 MW/m{sup 2} and surface temperatures near 300{degrees}C using 1.4 MPa water at 5.0 m/s flow velocity and an inlet temperature of 8-15{degrees}C. One tile was exposed to incrementally increasing heat fluxes up to 9.5 MW/m{sup 2} and surface temperatures up to 690{degrees}C before debonding at 10 MW/m{sup 2}. A third tile debonded after 9200 thermal fatigue cycles at 5 MW/m{sup 2}, while another debonded after 6800 cycles. In all cases, fatigue failure occurred in the intermetallic layers between the beryllium and copper. No fatigue cracking of the bulk beryllium was observed. During thermal cycling, a gradual loss of porous coating produced increasing sample temperatures. These experiments indicate that diffusion-bonded beryllium tiles can survive several thousand thermal cycles under ITER relevant conditions without failure. However, the reliability of the diffusion bonded Joint remains a serious issue.

  2. 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.

  3. Integrated core-SOL-divertor modelling for ITER including impurity: effect of tungsten on fusion performance in H-mode and hybrid scenario

    NASA Astrophysics Data System (ADS)

    Zagórski, R.; Voitsekhovitch, I.; Ivanova-Stanik, I.; Köchl, F.; Belo, P.; Fable, E.; Garcia, J.; Garzotti, L.; Hobirk, J.; Hogeweij, G. M. D.; Joffrin, E.; Litaudon, X.; Polevoi, A. R.; Telesca, G.; contributors, JET

    2015-05-01

    The compatibility of two operational constraints—operation above the L-H power threshold and at low power to divertor—is examined for ITER long pulse H-mode and hybrid scenarios in integrated core-scrape off layer (SOL)-divertor modelling including impurities (intrinsic Be, He, W and seeded Ne). The core thermal, particle and momentum transport is simulated with the GLF23 transport model tested in the self-consistent simulations of temperatures, density and toroidal rotation velocity in JET hybrid discharges and extrapolated to ITER. The beneficial effect of toroidal rotation velocity on fusion gain is shown. The sensitivity studies with respect to operational (separatrix and pedestal density, Ne gas puff) and unknown physics (W convective velocity and perpendicular diffusion in SOL as well as W prompt re-deposition) parameters are performed to determine their influence on the operational window and fusion gain.

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

    NASA Astrophysics Data System (ADS)

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

    2011-03-01

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

  5. 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. 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.

  7. Multi-parameter scaling of divertor power load profiles in D, H and He plasmas on JET and implications for ITER

    NASA Astrophysics Data System (ADS)

    Fundamenski, W.; Eich, T.; Devaux, S.; Jachmich, S.; Jakubowski, M.; Thomsen, H.; Arnoux, G.; Militello, F.; Havlickova, E.; Moulton, D.; Brezinsek, S.; Maddison, G.; McCormick, K.; Huber, A.; EFDA Contributors, JET

    2011-08-01

    Inter-ELM and ELM divertor power loads were measured on JET in dedicated deuterium, hydrogen and helium discharges. Matched triplets (D, H, He) were obtained for different values of magnetic field, B, plasma current, Ip, line average plasma density, n, and heating power, P. In this paper, the above experiments are described and the results are presented in terms of empirical scalings of inter-ELM and ELM wetted areas (power widths) versus engineering parameters. The inter-ELM wetted area on the outer target is found to scale roughly as B^{-0.57+/- 0.32}q_{cyl}^{1.0+/- 0.31} P_{sol}^{0.23+/- 0.09} Z^{0.3+/- 0.1}n^0A^0 , where A and Z and the fuel ion mass and charge numbers, and qcyl is the cylindrical safety factor, and the ELM wetted area as B^{-0.82+/- 0.25}q_{cyl}^{0.82+/- 0.25} n^{0.24+/- 0.19}P_{sol}^{0.20+/- 0.11} E_{ELM,5}^{0.22+/- 0.1} A^0Z^0 . The obtained inter-ELM scalings are then compared with those previously reported in the literature and with a wide range of 0D theoretical predictions. For this purpose a family of scrape-off layer power width models was constructed based on a permutation of different assumptions for parallel and perpendicular transport. It is found that a combination of parallel electron conduction and drift-ordered radial convection offers the best overall match to the empirical data, closely followed by models based on marginal stability to interchange/ballooning modes and ion convection with transport ordered radial velocity. Finally, implications for ITER are tentatively drawn, and a revised estimate for the power width in ITER is proposed. Extrapolating to ITER based on the empirical JET scaling and the optimum size scaling of R0.7±0.6, obtained based on comparison with simple models, yields a median outer target inter-ELM power width of ~5.5 ± 2 mm (mapped to the outer mid-plane), in close agreement with previous estimates and the ITER design value. The most pessimistic forecasts (little or no size scaling and inverse linear

  8. Effect of boundary conditions on the neutral gas temperatures and densities in the ITER divertor and pump duct

    NASA Astrophysics Data System (ADS)

    Ruzic, D. N.; Juliano, D. R.

    1992-12-01

    The DEGAS neutral atom transport code was used to simulate helium pumping and D/T throughput in ITER. The sensitivity of the simulation to two different reflection models, four transmission probabilities from the exit of the simulation to the pump (0.0625, 0.125, 0.1875 and 0.250), and a 2-D model versus a 3-D model were analyzed. The variation in reflection model changes the densities in the duct and the recycling of D/T by a factor of 1.6. The variation in the transmission probabilities affects these same quantities by a factor of 2.5. The dimensionality of the simulation affects the density profile in the duct. A transmission probability from the exit of the DEGAS simulation to the pump of 0.110 to 0.125 was calculated from the ITER reference drawings. Using this quantity and the DEGAS results, an exhaust rate of 112 to 127 moles/h is predicted, implying that the reference pumping systems may be larger than necessary by a factor of 2.

  9. 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.

  10. FINAL REPORT FOR THE DIII-D RADIATIVE DIVERTOR PROJECT

    SciTech Connect

    O'NEIL, RC; STAMBAUGH, RD

    2002-06-01

    OAK A271 FINAL REPORT FOR THE DIII-D RADIATIVE DIVERTOR PROJECT. The Radiative Divertor Project originated in 1993 when the DIII-D Five Year Plan for the period 1994--1998 was prepared. The Project Information Sheet described the objective of the project as ''to demonstrate dispersal of divertor power by a factor of then with sufficient diagnostics and modeling to extend the results to ITER and TPX''. Key divertor components identified were: (1) Carbon-carbon and graphite armor tiles; (2) The divertor structure providing a gas baffle and cooling; and (3) The divertor cryopumps to pump fuel and impurities.

  11. The breakup of methane under ITER divertor hydrogen plasma conditions for carbon chemical erosion analysis with CH spectroscopy

    NASA Astrophysics Data System (ADS)

    Westerhout, J.; Borodin, D.; Brezinsek, S.; Lopes Cardozo, N. J.; Rapp, J.; Schram, D. C.; van Rooij, G. J.

    2010-09-01

    Methane (CH4) was injected into the high density (ne ~ 1020 m-3), low temperature (Te ~ 1 eV) hydrogen plasma in Pilot-PSI to determine the CH A-X photon efficiency in this unexplored plasma regime. The effects of particle transport and particle reflection on the emission of directly excited CH under these plasma conditions were assessed with the 3D Monte Carlo code ERO. The simulations of the inverse photon efficiency showed a difference of ~20% between full hydrocarbon sticking or no sticking (reflection). In addition it predicts that particle transport may lead to more than a factor of 10 increase. The measured inverse photon efficiency is however constant at 100 ± 30 for 0.1 < Te < 1.0 eV. The constancy is consistent with dissociative recombination of CH_4^+ , CH_3^+ and CH_2^+ to produce excited CH instead of direct excitation. These results form a framework for in situ carbon erosion measurements in future fusion reactors such as ITER.

  12. 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.

  13. Snowflake divertor configuration studies in National Spherical Torus Experiment

    SciTech Connect

    Soukhanovskii, V. A.; McLean, A. G.; Rognlien, T. D.; Ryutov, D. D.; Umansky, M. V.; Bell, R. E.; Diallo, A.; Gerhardt, S.; Kaye, S.; Kolemen, E.; LeBlanc, B. P.; Menard, J. E.; Paul, S. F.; Podesta, M.; Roquemore, A. L.; Scotti, F.; Battaglia, D.; Bell, M. G.; Gates, D. A.; Kaita, R.; and others

    2012-08-15

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

  14. 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

  15. Bolometry for divertor characterization and control

    SciTech Connect

    Leonard, A.W.; Goetz, J.; Fuchs, C.; Marashek, M.; Mast, F.; Reichle, R.

    1995-10-01

    Operation of the divertor will provide one of the greatest challenges for ITER. Up to 400 MW of power is expected to be produced in the core plasma which must then be handled by plasma facing components. Power flowing across the separatrix and into the scrape-off-layer (SOL) can lead to a heat flux in the divertor of 30 MW/m{sup 2} if nothing is done to dissipate the power. This peak heat flux must be reduced to 5 MW/m{sup 2} for an acceptable engineering design. The current plan is to use impurity radiation and other atomic processes from intrinsic or injected impurities to spread out the power onto the first wall and divertor chamber walls. It is estimated that 300 MW of radiation in the divertor and SOL will be necessary to achieve this solution. Measurement of the magnitude and distribution of this radiated power with bolometry will be important for understanding and controlling the nER divertor. Present experiments have shown intense regions of radiation both in the divertor near the separatrix and in the X-point region. The task of a divertor bolometer system will be to measure the distribution and magnitude of this radiation. First, radiation measurements can be used for machine protection. Intense divertor radiation will heat plasma facing surfaces that are not in direct view of temperature monitors. Measurement of the radiation distribution will provide information about the power flux to these components. Secondly, a bolometer diagnostic is a basic tool for divertor characterization and understanding. Radiation measurements are important for power accounting, as a cross check for other power diagnostics, and gross characterisation of the plasma behavior. A divertor bolometer system can provide a 2-D measurement of the radiation profile for comparison with theory and modeling. Finally a bolometer system can provide realtime signals for control of the divertor operation.

  16. Divertor research on the DIII-D tokamak

    SciTech Connect

    Hill, D.N.; Allen, S.L.; Brooks, N.H.

    1994-10-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2016-12-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2017-01-01

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

  19. Gradient-based iterative image reconstruction scheme for time-resolved optical tomography

    SciTech Connect

    Hielscher, A.H.; Klose, A.D.; Hanson, K.M.

    1999-03-01

    Currently available tomographic image reconstruction schemes for optical tomography (OT) are mostly based on the limiting assumptions of small perturbations and a priori knowledge of the optical properties of a reference medium. Furthermore, these algorithms usually require the inversion of large, full, ill-conditioned Jacobian matrices. In this work a gradient-based iterative image reconstruction (GIIR) method is presented that promises to overcome current limitations. The code consists of three major parts: (1) A finite-difference, time-resolved, diffusion forward model is used to predict detector readings based on the spatial distribution of optical properties; (2) An objective function that describes the difference between predicted and measured data; (3) An updating method that uses the gradient of the objective function in a line minimization scheme to provide subsequent guesses of the spatial distribution of the optical properties for the forward model. The reconstruction of these properties is completed, once a minimum of this objective function is found. After a presentation of the mathematical background, two- and three-dimensional reconstruction of simple heterogeneous media as well as the clinically relevant example of ventricular bleeding in the brain are discussed. Numerical studies suggest that intraventricular hemorrhages can be detected using the GIIR technique, even in the presence of a heterogeneous background.

  20. 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.

  1. Engineering design of a radiative divertor for DIII-D

    NASA Astrophysics Data System (ADS)

    Smith, J. P.; Anderson, P. M.; Baxi, C. B.; Chin, E.; Hollerbach, M. A.; Hyatt, A. W.; Junge, R.; Mahdavi, M. A.; Redler, K.; Reis, E. E.

    1994-10-01

    A new divertor called the Radiative Divertor is presently being designed for the DIII-D tokamak. Input from tokamak experiments and modeling form the basis for the new design. The Radiative Divertor is intended to reduce the heat flux on the divertor plates by dispersing the power with radiation. Gas puffing experiments in the current open divertor have shown a reduction of the divertor heat flux with either deuterium or impurity puffing. However, either the plasma density (D(sub 2)) or the core Z(sub eff) (impurities) increases in these experiments. The radiative divertor uses a slot structure to isolate the divertor plasma region from the area surrounding the core plasma. Modeling has shown that the Radiative Divertor hardware will provide better baffling and particle control and thereby minimize the effect of the gas puffing in the divertor region on the plasma core. In addition, the Radiative Divertor structure will allow density control in plasma shapes with high triangularity (greater than 0.8) required for advanced tokamak operation. The divertor structure allows for operation in either double or single-null plasma configurations. Four independently controlled divertor cryopumps will enable pumping at either the inboard (upper and lower) or the outboard (upper and lower) divertor plates. Biasing is an integral part of the design and is based on experience at the Tokamak de Varennes (TdeV) and DIII-D. Boron nitride tiles electrically insulate the inner and outer strike points and a low current electrode is used to apply a radial electric field to the scrape-off layer. TdeV has shown that biasing can provide particle and impurity control. The design is extremely flexible, and will allow physics studies of the effect of slot width and height. This is extremely important as the amount of chamber volume needed for the divertor in future machines such as International Thermonuclear Experiment Reactor (ITER) and Tokamak Physics Experiment (TPX) must be determined.

  2. DIII-D divertor reflectometer system

    SciTech Connect

    Rhodes, T.L.; Doyle, E.J.; Nguyen, X.V.; Kim, K.W.; Peebles, W.A.; Doane, J.L.

    1997-01-01

    Divertor density profiles, asymmetries, turbulence, and MARFE diagnosis are extremely important and affect the divertor design process for ITER and other future devices. In addition, a functioning divertor density profile system will be essential for the operation of these machines. It is thus critical to prototype and demonstrate diagnostics capable of operating in a divertor environment. To meet these needs a divertor reflectometer system has been designed and installed on DIII-D. The design stresses flexibility, modularity, and simplicity. It consists of a circular, smoothwall, overmoded waveguide followed by a TE{sub 11}{R_arrow}HE{sub 11} mode converter (the HE{sub 11} mode is a low loss Gaussian mode with a very symmetric radiation pattern, optimal for this use) thus allowing use of an arbitrary polarization (f{sub pe},f{sub LH},f{sub RH}). The design provides for testing of a variety of antennas/probing directions including: upward to probe the X-point region, including MARFEs, sideways to probe outboard/inboard divertor legs, and oppositely directed to probe both divertor legs simultaneously. System design, operational considerations, and experimental data are presented. {copyright} {ital 1997 American Institute of Physics.}

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

    NASA Astrophysics Data System (ADS)

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

    2014-12-01

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

  4. Preliminary assessment of the tritium inventory and permeation in the plasma facing components of ITER

    SciTech Connect

    Federici, G.; Holland, D.; Brooks, J.; Causey, R.; Dolan, T.J.; Longhurst, G.

    1995-12-31

    This paper discusses preliminary quantitative predictions for the tritium inventory in- and permeation through the first-wall and divertor PFC`s of ITER. The primary plasma facing material under consideration is beryllium, with possible use of tungsten or carbon fiber composites (CFC`s) on high-heat-flux surfaces. They use state-of-the-art tritium transport models, in conjunction with design parameters, and loading conditions anticipated for the first-wall, baffle, limiter and divertor. The analysis includes the synergistic effects of erosion on tritium implantation and trapping, which are expected to play a key role, particularly in the divertor regions where the interaction of the plasma with the surfaces will be most severe. The influence of several key parameters that strongly affect tritium build-up and release is assessed. Finally, they discuss the uncertainties in materials properties under ITER operating conditions and the R and D needed to resolve these uncertainties.

  5. 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.; Ye, M. Y.; 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.

  6. TPX divertor modeling studies

    SciTech Connect

    Rensink, M.E.; Braams, B.J.; Brooks, J.N.

    1995-06-20

    The Tokamak Physics Experiment (TPX) is designed to demonstrate features of an economically attractive steady state tokamak reactor. In this paper we present recent results from numerical studies of the proposed TPX divertor design (1), focusing on particle control and on radiative divertor scenarios for reducing the peak divertor heat flux. The configuration is an up/down symmetric double-null with a deep re-entrant slot geometry for the outer divertor legs.

  7. DIII-D research towards resolving key issues for ITER and steady-state tokamaks

    NASA Astrophysics Data System (ADS)

    Hill, D. N.; the DIII-D Team

    2013-10-01

    The DIII-D research program is addressing key ITER research needs and developing the physics basis for future steady-state tokamaks. Pellet pacing edge-localized mode (ELM) control in the ITER configuration reduces ELM energy loss in proportion to 1/fpellet by inducing ELMs at up to 12× the natural ELM rate. Complete suppression of ELMs with resonant magnetic perturbations has been extended to the q95 expected for ITER baseline scenario discharges, and long-duration ELM-free QH-mode discharges have been produced with ITER-relevant co-current neutral-beam injection (NBI) using external n = 3 coils to generate sufficient counter-Ip torque. ITER baseline discharges at βN ˜ 2 and scaled NBI torque have been maintained in stationary conditions for more than four resistive times using electron cyclotron current drive (ECCD) for tearing mode suppression and disruption avoidance; active tracking with steerable launchers and feedback control catch these modes at small amplitude, reducing the ECCD power required to suppress them. Massive high-Z gas injection into disruption-induced 300-600 kA 20 MeV runaway electron (RE) beams yield dissipation rates ˜10× faster than expected from e-e collisions and demonstrate the possibility of benign dissipation of such REs should they occur in ITER. Other ITER-related experiments show measured intrinsic plasma torque in good agreement with a physics-based model over a wide range of conditions, while first-time main-ion rotation measurements show it to be lower than expected from neoclassical theory. Core turbulence measurements show increased temperature fluctuations correlated with sharply enhanced electron transport when \

  8. 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.

  9. Divertor experiment in large helical device

    NASA Astrophysics Data System (ADS)

    Motojima, O.; Ohyabu, N.; Komori, A.; Noda, N.; Yamazaki, K.; Yamada, H.; Sagara, A.; Kubota, Y.; Suzuki, H.; Inoue, N.; Morisaki, T.; Masuzaki, S.; Sakamoto, R.; Matsuoka, K.; Fujiwara, M.; Iiyoshi, A.

    1996-12-01

    This paper describes the major objectives of the LHD divertor experiment which is proposed to produce currentless-steady-state plasmas with high performance and without any current disruption. Since further improvement in confinement is a common and general requirement for fusion research including the LHD project, it is also necessary to develop the edge plasma control techniques and to understand the physical behaviour in the LHD divertor, i.e. the newly developed continuous helical divertor and a local island divertor (LID) concepts. In order to achieve these objectives, there were several key issues in physics and technology, which had to be resolved through careful investigation before the LHD experiment could start. In this paper, we summarize the recent progress of the physics understanding of divertor functions, divertor plasma operation scenarios, and properties of the LHD magnetic field structure in addition to the experimental planning. We also describe the recent result of an LID experiment in the CHS device, which demonstrated the possibility of edge particle and heat control by the LID.

  10. Alternative divertor target concepts for next step fusion devices

    NASA Astrophysics Data System (ADS)

    Mazul, I. V.

    2016-12-01

    The operational conditions of a divertor target in the next steps of fusion devices are more severe in comparison with ITER. The current divertor designs and technologies have a limited application concerning these conditions, and so new design concepts/technologies are required. The main reasons which practically prevent the use of the traditional motionless solid divertor target are analyzed. We describe several alternative divertor target concepts in this paper. The comparative analysis of these concepts (including the advantages and the drawbacks) is made and the prospects for their practical implementation are prioritized. The concept of the swept divertor target with a liquid metal interlayer between the moving armour and motionless heat-sink is presented in more detail. The critical issues of this design are listed and outlined, and the possible experiments are presented.

  11. Impurity-induced divertor plasma oscillations

    SciTech Connect

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

  12. Impurity-induced divertor plasma oscillations

    SciTech Connect

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

    2016-01-15

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

  13. Impurity-induced divertor plasma oscillations

    DOE PAGES

    Smirnov, R. D.; Kukushkin, A. S.; Krasheninnikov, S. I.; ...

    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

  14. 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.

  15. Divertor design for the tokamak physics experiment

    NASA Astrophysics Data System (ADS)

    Hill, D. N.; Braams, B.; Brooks, J. N.; Ruzic, D. N.; Ulrickson, M.; Werley, K. A.; Campbell, R.; Goldston, R.; Kaiser, T.; Neilson, G. H.; Mioduszewski, P.; Rensink, M. E.; Rognlien, T. D.

    1995-04-01

    In this paper we discuss the divertor design for the planned TPX tokamak, which will explore the physics and technology of steady state (1000 s pulses) heat and particle removal in high confinement (up to 4 × L-mode), high beta (up to βN = 5) divertor plasmas sustained by non-inductive current drive. TPX will operate in the double-null divertor configuration, with actively cooled graphite targets forming a deep (0.57 m) slot at the outer strike point. The peak heat flux on the highly tilted (74° from normal) re-entrant divertor plate (tilted to recycle ions back toward the separatrix) will be in the range of 4-6 MW/m 2 with 17.5 MW of auxiliary heating power. The combination of pumping and gas puffing (D 2 plus impurities), along with higher heating power (45 MW maximum) will allow testing of radiative divertor concepts at ITER-like power densities.

  16. Carbon Deposition in the Inner JET Divertor Measured by Means of Quartz Microbalance

    NASA Astrophysics Data System (ADS)

    Esser, H. G.; Philipps, V.; Freisinger, M.; Coad, P.; Matthews, G. F.; Neill, G.; JET EFDA Contributors

    A Quartz Microbalance (QMB) system was implemented in the inner divertor region of JET in order to measure in situ and time resolved (minimum exposure time ≥0.1 s) material fluxes (mainly carbon) and layer deposition. The system has been developed to operate at temperatures up to 200°C. The aim is to investigate carbon transport to the remote areas, and hence the tritium retention in dependence on plasma conditions. This question is still a major concern for the ITER operation. The mass sensitivity of the system is Sm = 1.5 A~— 10-8 [g/Hz cm2]. First reliable measurements were made during the C5 campaign (March–May 2002; â‰e1000 plasma discharges). The results presented are based on 74 selected exposures (694 s) under various conditions (strike point position, input power, neutral pressure, ELM frequency). Most influencing on the carbon deposition in the remote area seems to be the geometry i.e. the strike point position on the divertor tiles. In average 1.9 A~— 10-4 C-atom are deposited per deuterium ion flowing into the inner divertor.

  17. Divertor Configurations which Optimize Helium Pumping

    NASA Astrophysics Data System (ADS)

    Strachan, James

    2008-11-01

    Helium accumulation in DT plasmas is often presumed to be one limitation to the fusion power production. The core helium density has an unavoidable central source and a confinement time which tends to be long as is consistent with the required energy confinement times. Any pumping of the helium can only act to reduce the helium recycling. Within that constraint, however, it is still valuable to efficiently pump helium. Helium pumping can be aided by optimal placement of the helium pump in the divertor. The pump should be on the SOL side of the separatrix displaced into the region where the current of impurity particles enters into the divertor and initially strike the target. A numerical example will be given of helium pumping by the ITER divertor. A factor-of-two reduction in core helium densities is possible by optimal pump placement. One difficulty is the need for low temperatures along the targets to prevent their erosion. On ITER, recycled DT near the strike points is hoped to cool this region. The angle between the separatrix and the target is such that recycled neutrals cause ionization, excitation, and dissociation power losses along the target. The ITER solution constrains the choice of pump locations. Alternatively, the strike point cooling can be achieved by local DT (or low Z impurity) injection at the strike point.

  18. Distribution of Hydrogen Isotopes, Carbon and Beryllium on In-Vessel Surfaces in the Various JET Divertors

    SciTech Connect

    Coad, J.P.; Rubel, M.; Bekris, N.; Brennan, D.; Hole, D.; Likonen, J.; Vainonen-Ahlgren, E

    2005-07-15

    JET has operated with divertors of differing geometries since 1994. Impurities accumulated in the inner leg of all the divertors, and operation of the first (Mk I) divertor with beryllium tiles demonstrated that most are eroded from the main chamber walls and swept along the scrape-off layer to the inner divertor. Carbon deposited at the inner divertor is then locally transported to shadowed regions such as the inner louvres, where, for example, most of the tritium was trapped during the deuterium-tritium experiment (DTE1). Factors affecting these transport processes (e.g. temperature) are important for ITER, but are not well understood.

  19. Divertor research on the DIII-D tokamak

    NASA Astrophysics Data System (ADS)

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

    1994-10-01

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

  20. 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.

  1. 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.

  2. Divertor design for the Tokamak Physics Experiment

    NASA Astrophysics Data System (ADS)

    Hill, D. N.; Braams, B.; Brooks, J. N.; Ruzic, D. N.; Ulrickson, M.; Werley, K. A.; Campbell, R.; Goldston, R.; Kaiser, T.; Nellson, G. H.

    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 x L-mode), high beta (beta(sub N) greater than or equal to 3) divertor plasmas sustained by non-induct ive 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 deg) from normal) re-entrant (to recycle ions back toward the separatrix) will be in the range of 4-6 MW/sq m 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-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.

  4. 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.

  5. 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.

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

    SciTech Connect

    1995-10-01

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

  7. JET helps prepare for ITER operation

    NASA Astrophysics Data System (ADS)

    Watkins, Michael

    2005-10-01

    The main focus of the JET programme (2006-10) in preparation of ITER operation is a new ITER-like ICRH antenna (total RF power increased to ˜15MW), a new ITER-like first wall (beryllium in the main chamber, tungsten in the divertor, and possibly CFC at the strike points), upgraded NB power (to 35MW/20s or 17.5MW/10s), and an improved diagnostic and control capability. Mass flows for ITER Scenarios with the ITER-like first wall will be optimised, particularly to minimise in-vessel tritium inventory, since this must be controlled strictly in ITER and has been shown on JET with a carbon first wall to depend sensitively on plasma conditions. Higher power will allow confinement scalings to be resolved for normalised parameters closer to ITER (beta dependence of ELMy H-modes, confinement of improved H-modes at low ρ*) and offers the prospect of high beta operation at high current and density, and new fully non-inductive, high performance, ITB discharges sustained to long pulse by real time current and pressure profile control, particularly in bootstrap current dominated regimes. Together, the first wall and increased heating power place strict constraints on the optimisation of ITER scenarios for long pulse operation with low melt damage. Large ELMs (in excess of 1MJ; marginally accessible on JET at present) and disruptions could cause melt severe damage which must be studied and controlled. The testing and optimisation of techniques for ELM mitigation (impurity seeding, demonstrated on JET; use of a new high frequency pellet injector (10-60Hz) to prevent large ELMs, demonstrated on ASDEX Upgrade) and disruption mitigation (fast gas injection from a new disruption mitigation valve, demonstrated on DIII-D) will be even more relevant under the ITER-like edge plasma conditions accessible with the increased power. Acknowledgement : Contributors to EFDA-JET Workprogramme

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

    SciTech Connect

    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).

  9. 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).

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

    PubMed

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

    2012-10-01

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

  11. Divertor Materials Evaluation System (DiMES)

    SciTech Connect

    Wong, C.P.; West, W.P.; Whyte, D.G.; Bastasz, R.J.; Brooks, J.; Wampler, W.R.

    1997-11-01

    The mission of the Divertor Materials Evaluation System (DiMES) in DIII-D is to establish an integrated data base from measurements in the divertor of a tokamak in order to address some of the ITER and fusion power reactor plasma material interaction issues. Carbon and metal coatings of Be, W, V, and Mo were exposed to the steady-state outer strike point on DIII-D for 4-18 s. These short exposure times ensure controlled exposure conditions, and the extensive arrays of DIII-D divertor diagnostics provide a well-characterized plasma for modeling efforts. Postexposure analysis provides a direct measure of surface material erosion rates and the amount of retained deuterium. For carbon, these results match closely with the results of accumulated carbon deposition and erosion, and the corresponding deuterium retention of long term exposure tiles in DIII-D. Deuterium retention of different materials was measured using the {sup 3}He(d,p) {sup 4}He nuclear reaction. For carbon, these measurements showed peak deuterium areal density of about 8 {times} 10 {sup 18} D/cm{sup 2} in a co-deposited layer about 6 {micro}m deep, mainly at the usually detached inboard divertor leg. That layer of carbon near the inner divertor strike point has an atomic saturation concentration of D/C {approx} 0.25, which is not significantly lower than the laboratory-measured saturation retention of 0.4. Under the carbon contaminated background plasma of DIII-D, metal coatings of Be, V, Mo, and W were exposed to the steady state outer strike point under ELMing and ELM-free H-mode discharges. The rate of material erosion and tritium retention were measured. As expected, W shows the lowest erosion rate at 0.1 nm/s and the lowest deuterium uptake.

  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. The snowflake divertor

    DOE PAGES

    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

  14. R&D on tungsten plasma facing components for the JET ITER-like wall project

    NASA Astrophysics Data System (ADS)

    Piazza, G.; Matthews, G. F.; Pamela, J.; Altmann, H.; Coad, J. P.; Hirai, T.; Lioure, A.; Maier, H.; Mertens, Ph.; Philipps, V.; Riccardo, V.; Rubel, M.; Villedieu, E.; JET ITER-like Project

    2007-08-01

    Currently, the primary ITER materials choice is a full beryllium main wall with carbon fibre composite at the divertor strike points and tungsten on the upper vertical targets and dome. The full tungsten divertor option is a possibility for the subsequent D-T phase. Neither of the ITER material combinations of first wall and divertor materials has ever been tested in a tokamak. To collect operational experience at JET with ITER relevant material combination (Be, C and W) would reduce uncertainties and focus the preparation for ITER operations. Therefore, the ITER-like wall project has been launched to install in JET a tungsten divertor and a beryllium main wall. This paper describes the R&D activities carried out for the project to develop an inertially cooled bulk tungsten divertor tile, to fully characterise tungsten coating technologies for CFC divertor tiles and to develop erosion markers for use as diagnostics on beryllium tiles.

  15. 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.

  16. 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.

  17. Taming the heat flux problem: Advanced divertors towards fusion power

    SciTech Connect

    Kotschenreuther, M.; Mahajan, S.; Valanju, P. M.; Covele, B.; Waelbroeck, F. L.; Canik, John M.; LaBombard, Brian

    2015-09-11

    The next generation fusion machines are likely to face enormous heat exhaust problems. In addition to summarizing major issues and physical processes connected with these problems, we discuss how advanced divertors, obtained by modifying the local geometry, may yield workable solutions. We also point out that: (1) the initial interpretation of recent experiments show that the advantages, predicted, for instance, for the X-divertor (in particular, being able to run a detached operation at high pedestal pressure) correlate very well with observations, and (2) the X-D geometry could be implemented on ITER (and DEMOS) respecting all the relevant constraints. As a result, a roadmap for future research efforts is proposed.

  18. A new scaling for divertor detachment

    DOE PAGES

    Goldston, R. J.; Reinke, M. L.; Schwartz, J. A.

    2017-03-29

    The ITER design, and future reactor designs, depend on divertor `detachment,'whether partial, pronounced or complete, to limit heat flux to plasma-facing components and to limit surface erosion due to sputtering. It would be valuable to have a measure of the difficulty of achieving detachment as a function of machine parameters, such as input power, magnetic field, major radius, etc. Frequently the parallel heat flux, estimated typically as proportional to P-sep/R or PsepB/R, is used as a proxy for this difficulty. Here we argue that impurity cooling is dependent on the upstream density, which itself must be limited by a Greenwald-likemore » scaling. Taking this into account self-consistently, we find the impurity fraction required for detachment scales dominantly as power divided by poloidal magnetic field. The absence of any explicit scaling with machine size is concerning, as P-sep surely must increase greatly for an economic fusion system, while increases in the poloidal field strength are limited by coil technology and plasma physics. This result should be challenged by comparison with 2D divertor codes and with measurements on existing experiments. Nonetheless, it suggests that higher magnetic field, stronger shaping, double-null operation, `advanced' divertor configurations, as well as alternate means to handle heat flux such as metallic liquid and/or vapor targets merit greater attention.« less

  19. Divertor materials evaluation system (DiMES)

    SciTech Connect

    Wong, C.P.C.; West, W.P.; Whyte, D.G.; Bastasz, R.J.; Brooks, J.; Wampler, W.R.

    1997-12-31

    The mission of the Divertor Materials Evaluation System (DiMES) in DIII-D is to establish an integrated data base from measurements in the divertor of a tokamak in order to address some of the ITER and fusion power reactor plasma material interaction issues. Carbon and metal coatings of Be, W, V, and Mo were exposed to the steady-state outer strike point on DIII-D for 4--18 s. These short exposure times ensure controlled exposure conditions, and the extensive arrays of DIII-D divertor diagnostics provide a well-characterized plasma for modeling efforts. Post-exposure analysis provides a direct measure of surface material erosion rates and the amount of retained deuterium. For carbon, these results match closely with the results of accumulated carbon deposition and erosion, and the corresponding deuterium retention of long term exposure tiles in DIII-D. Under the carbon-contaminated background plasma of DIII-D, metal coatings of Be, V, Mo, and W were exposed to the steady-state outer strike point under ELMing and ELM-free H-mode discharges. The rate of material erosion and deuterium retention were measured. As expected, W shows the lowest erosion rate at 0.1 mm/s and the lowest deuterium uptake of 2 {times} 10{sup 20}/m{sup 2}.

  20. 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.

    2010-01-01

    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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. Divertor plasma detachment

    SciTech Connect

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

    2016-05-15

    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.

  6. 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.

  7. 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.

  8. Variation of Particle Control with Changes in Divertor Geometry

    SciTech Connect

    Petrie, T W; Allen, S L; Brooks, N H; Fenstermacher, M E; Ferron, J R; Greenfield, C M; Groth, M; Hyatt, A W; Leonard, A W; Luce, T C; Mahdavi, M A; Murakami, M; Porter, G D; Rensink, M E; Schaffer, M J; Wade, M R; Watkins, J G; West, W P; Wolf, N S

    2004-10-18

    Recent experiments on DIII-D point to the importance of two factors in determining how effectively the deuterium particle inventory in a tokamak plasma can be controlled through pumping at the divertor target(s): (1) the divertor magnetic balance, i.e., the degree to which the divertor topology is single-null (SN) or double-null (DN), and (2) the direction of the of Bx{divergent}B ion drift with respect to the X-point(s). Changes in divertor magnetic balance near the DN shape have a much stronger effect on the particle exhaust rate at the inner divertor target(s) than on the particle exhaust rate at the outer divertor target(s). The particle exhaust rate for the DN shape is strongest at the outer strike point opposite the Bx{divergent}B ion particle drift direction. Our data suggests that the presence of Bx{divergent}B and ExB ion particle drifts in the scrapeoff layer (SOL) and divertors play an important role in the particle exhaust rates of DN and near-DN plasmas. Particle exhaust rates are shown to depend strongly on the edge (pedestal) density n{sub e,PED}. In the lower range of densities considered in this study, i.e., n{sub e,PED}/ n{sub GREENWALD}<0.4, particle exhaust rates are also found to be approximately proportional to the deuterium recycling intensity in front of the respective plenum entrance. Our results are shown to have implications for particle control in ITER and other future tokamaks.

  9. Overview of the DIII-D Divertor Tungsten Rings Campaign

    NASA Astrophysics Data System (ADS)

    Unterberg, E. A.; Thomas, D. M.; Petrie, T. W.; Abrams, T.; Garofalo, A. M.; Stangeby, P. C.; Rudakov, D. L.; Schmitz, O.; Grierson, B. A.; Victor, B.

    2016-10-01

    Experiments have recently been carried out with toroidal arrays of W-coated metal inserts at two distinct locations in the lower divertor region. The purpose of the experiments is to determine the high-Z divertor erosion and migration, and its effect on core contamination in high performance, ELM-y H-mode, tokamak discharges in a mixed-material, i.e. C and W, environment. The experiments focused on characterizing the sputtering source from each location, the SOL transport of W, and the subsequent impact on core performance. A wide range of ELM-y conditions was studied, including ELM controlled and ELM-free regimes, to determine the importance of the divertor strike point position relative to W sources in these various regimes. The W penetration efficiency was characterized by using a far-SOL collector probe related to core W density. Correlations between source strength (as measured by W-I spectroscopy) relative to the distance of the strikepoint to each W array, the divertor target magnetic flux expansion, and ELM frequency was seen. These experiments aid in understanding the impact of high-Z divertor source location on core performance in future mixed-material fusion devices, e.g. ITER. Supported by US DOE under DE- AC05-00OR22725, DE-FC02-04ER54698, DE-FG02-07ER54917, DE-SC0013911, DE-AC02-09CH11466, DE-AC52-07NA27344.

  10. Analysis of sweeping heat loads on divertor plate materials

    SciTech Connect

    Hassanein, A.

    1991-12-31

    The heat flux on the divertor plate of a fusion reactor is probably one of the most limiting constraints on its lifetime. The current heat flux profile on the outer divertor plate of a device like ITER is highly peaked with narrow profile. The peak heat flux can be as high as 30--40 MW/m{sup 2} with full width at half maximum (FWHM) is in the order of a few centimeters. Sweeping the separatrix along the divertor plate is one of the options proposed to reduce the thermomechanical effects of this highly peaked narrow profile distribution. The effectiveness of the sweeping process is investigated parametrically for various design values. The optimum sweeping parameters of a particular heat load will depend on the design of the divertor plate as well as on the profile of such a heat load. In general, moving a highly peaked heat load results in substantial reduction of the thermomechanical effects on the divertor plate. 3 refs., 8 figs.

  11. Analysis of sweeping heat loads on divertor plate materials

    SciTech Connect

    Hassanein, A.

    1991-01-01

    The heat flux on the divertor plate of a fusion reactor is probably one of the most limiting constraints on its lifetime. The current heat flux profile on the outer divertor plate of a device like ITER is highly peaked with narrow profile. The peak heat flux can be as high as 30--40 MW/m{sup 2} with full width at half maximum (FWHM) is in the order of a few centimeters. Sweeping the separatrix along the divertor plate is one of the options proposed to reduce the thermomechanical effects of this highly peaked narrow profile distribution. The effectiveness of the sweeping process is investigated parametrically for various design values. The optimum sweeping parameters of a particular heat load will depend on the design of the divertor plate as well as on the profile of such a heat load. In general, moving a highly peaked heat load results in substantial reduction of the thermomechanical effects on the divertor plate. 3 refs., 8 figs.

  12. 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.

  13. Enhanced visible and near-infrared capabilities of the JET mirror-linked divertor spectroscopy systema)

    NASA Astrophysics Data System (ADS)

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

    2014-11-01

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

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

    SciTech Connect

    Sizyuk, V. Hassanein, A.

    2015-01-15

    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.

  15. 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.

  16. 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.

  17. Design, Engineering, and Testing for the Alcator C-Mod Outer Divertor Upgrade

    NASA Astrophysics Data System (ADS)

    Harrison, S.; Vieira, R.; Lipschultz, B.; Ellis, R.; Karnes, D.; Doody, J.; Zhou, L.; Titus, P.; Zhang, H.; Beck, W.; Granetz, R.

    2012-10-01

    Alcator C-mod's major outer divertor upgrade will enable significant advances in our understanding of reactor relevant physics and operations. Two primary features of the new outer divertor are its toroidally continuous design (electrical and mechanical), and ability to be operated up to or independently heated to 600 C. Full control of the divertor PFC temperature from ambient vessel temperature to 600 C, will enable new and important tokamak research into the temperature dependence of fuel retention, PFC deposition and erosion, and divertor recycling. Significant design, analysis, and testing is underway to complete this important and challenging upgrade, which will provide valuable information for ITER and future reactors. Among other aspects of the innovative approach, the divertor plate supports, halo current shunts, and thermal shield assemblies will be discussed. The divertor supports enable pure radial motion of the divertor ring as it expands thermally and robustness to massive disruption induced electro-mechanical loads. Halo current shunts conduct 400kA in an 8T magnetic field and allow for divertor displacement relative to the vessel. Thermal shielding significantly reduces radiation and conduction to surrounding vessel structures.

  18. Multiplexing thermography for International Thermonuclear Experimental Reactor divertor targets

    SciTech Connect

    Itami, K.; Sugie, T.; Vayakis, G.; Walker, C.

    2004-10-01

    The concept of multiplexing thermography is applied to the design of the divertor thermography system for International Thermonuclear Experimental Reactor (ITER). The combination of the front mirror with multiellipticity and a Czerney-Turner spectrometer with a 0.2 mm pitched multichannel detector enables a spatial resolution of 3 mm and a time resolution of 20 {mu}s above a target temperature of 300 deg. C to be achieved. This should be sufficient to measure ELM heat fluxes to the targets in ITER. To satisfy the measurement requirement, it is very important to keep an accurate alignment around the optical axis against movement of the vessel during the plasma discharges. Several key engineering problems, such as the survivability of components against mirror coating by redeposited divertor material, remain to be solved. Potential solutions have been identified.

  19. Compatibility of the Radiating Divertor with High Performance Plasmas in DIII-D

    SciTech Connect

    Petrie, T; Wade, M; Allen, S; Brooks, N; Fenstermacher, M; Ferron, J; Greenfield, C; Groth, M; Hyatt, A; Lasnier, C; Leonard, A; Luce, T; Mahdavi, M; Schaffer, M; Watkins, J; West, W

    2005-06-24

    Excessive thermal power loading on the divertor structures presents a design difficulty for future-generation, high powered tokamaks. This difficulty may be mitigated by ''seeding'' the divertor with impurities which radiate a significant fraction of the power upstream of the divertor targets. For this ''radiating divertor'' concept to be practical, however, the confinement and stability of the plasma cannot be compromised by excessive leakage of the seeded impurities into the core plasma. One proposed way of reducing impurity influx is to enhance the directed scrape-off layer (SOL) flow of deuterium ions toward the divertor [1-5]. We report here on the successful application of the radiating divertor scenario to high performance plasma operation in a DIII-D ''hybrid'' H-mode regime. The ''hybrid'' regime [6,7] has many features in common with conventional ELMing H-mode regimes, such as high confinement, e.g., H{sub ITER89P} > 2, where H{sub ITER89P} is the energy confinement normalized to the 1989 ITER L-mode scaling [8]. The main difference is the absence of sawtooth activity in the hybrid. Argon was selected as the seeded impurity for this experiment because argon radiates effectively at both the divertor and pedestal temperatures found in DIII-D hybrid H-mode operation and has a relatively short ionization mean free path. Carbon is also present as the dominant intrinsic impurity in DIII-D discharges. The geometry of this experiment is shown in Fig. 1. A double-null cross-sectional shape was biased upward (dRsep = +1.0 cm). To increase the deuterium ion flow toward the divertor at the top of the vessel, deuterium gas was introduced near the bottom. Argon was injected directly into the private flux region (PFR) of the upper divertor. In-vessel pumping of deuterium and argon was done by cryopumps located in the two upper divertor plenums, shown in cross-hatching [9]. The upper divertor, which we hereafter will simply refer to as the ''divertor'', is the region

  20. Selection of plasma facing materials for ITER

    SciTech Connect

    Ulrickson, M.; Barabash, V.; Chiocchio, S.

    1996-10-01

    ITER will be the first tokamak having long pulse operation using deuterium-tritium fuel. The problem of designing heat removal structures for steady state in a neutron environment is a major technical goal for the ITER Engineering Design Activity (EDA). The steady state heat flux specified for divertor components is 5 MW/m{sup 2} for normal operation with transients to 15 MW/m{sup 2} for up to 10 s. The selection of materials for plasma facing components is one of the major research activities. Three materials are being considered for the divertor; carbon fiber composites, beryllium, and tungsten. This paper discusses the relative advantages and disadvantages of these materials. The final section of plasma facing materials for the ITER divertor will not be made until the end of the EDA.

  1. Analysis of rotating collectors from the private region of JET 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.; Matthews, G. F.; Brezinsek, S.

    2015-08-01

    Rotating collectors are used in JET to provide time-resolved measurements of erosion and redeposition of vessel materials. The silicon collecting discs rotate behind an aperture, driven by pulsing of the toroidal magnetic field, with the deposits analysed ex-situ by nuclear reaction analysis. The angular dependence of deposition is mapped to discharge number using the B-field history, allowing the influence of different plasma configurations and parameters to be investigated. A simple geometrical model using sputtering and reflection from the strike point has qualitatively reproduced the deposition found on collectors located under the central divertor tile and facing towards the inner strike point. The beryllium deposition on the ITER-like wall (ILW) collector showed an order of magnitude reduction in deposition compared to carbon deposition on the JET-C collector. This decreased deposition is attributed to low long range divertor transport due to reduced chemical sputtering/erosion and codeposition of beryllium relative to carbon.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. The role of atomic and molecular physics for dissipative divertor operation in helium and deuterium plasmas

    NASA Astrophysics Data System (ADS)

    Canik, J. M.

    2016-10-01

    Recent experiments in DIII-D helium plasmas are examined to resolve the role of atomic and molecular physics in major discrepancies between experiment and modeling of dissipative divertor operation. Helium operation removes the complicated molecular processes of deuterium plasmas that are a prime candidate for the inability of standard fluid models (SOLPS, UEDGE) to reproduce dissipative divertor operation, primarily the consistent under-prediction of radiated power. With helium fueling, a high-recycling divertor was established with divertor densities increasing to ne,div >= 3 ×1020m-3 and temperatures decreasing to Te,div <= 2 eV as measured by divertor Thomson scattering (DTS). The electron pressure, pe , div decreased gradually with increasing density to less than 30% of the low density value. However, the ion flux to the divertor target did not decrease until the highest densities and lowest temperatures, Te,div <= 2 eV. In contrast, with deuterium operation, increasing density leads to a rapid transition from Te,div >= 10 eV to Te,div <= 3 eV, though both pe , div and ion flux do not decrease until Te,div <= 2 eV. These differences indicate an important role for molecular and atomic physics in the dynamics of divertor dissipation. Initial SOLPS modeling has reproduced ne and Te profiles at the midplane and divertor target, as well as the spatial structure of radiation patterns measured in moderate density helium plasmas. However, the modeled divertor radiation is less than measured, similar to deuterium simulations, suggesting processes more universal than species-specific atomic or molecular physics may be the source of radiation deficit. Detailed assessments of ne, Te profiles in the divertor volume, uniquely determined at DIII-D using DTS, are made along with analysis of measured and modeled line radiation to shed more light on these intriguing findings. Supported by the US DOE under DE-AC05-00OR22725.

  7. Attainment of a stable, fully detached plasma state in innovative divertor configurations

    NASA Astrophysics Data System (ADS)

    Umansky, Maxim

    2016-10-01

    The heat load on plasma facing components is a critical engineering constraint for future tokamaks, which has stimulated the community to consider innovative magnetic divertor geometries for future high power devices. Present-day advanced divertor scenarios generally rely on partially detached regimes, also planned for ITER; a fully detached state would usually lead to MARFE and degradation of core confinement. Modeling reveals that novel magnetic geometries can have a major impact on plasma detachment and power handling. Using the UEDGE tokamak edge transport model for configurations with tightly baffled long divertor legs, extended radially, or vertically, we find stable, fully detached divertor operation. Including a secondary X-point in the outer leg volume extends the attainment of a stable detached state to the highest power. As the input power is reduced to a threshold value, the outer leg transitions to a fully detached state with the detachment front localized at the secondary X-point or in the leg volume; reducing the power further results in the detachment front steady-state location shifting upstream. As the power is reduced, the detachment front eventually moves to the primary X-point, which sets the lower power limit for the range of stable operation. Still, for a long-legged divertor, a fully detached, stable divertor regime is maintained over an order-of-magnitude variation in exhaust power. In contrast, a standard divertor has a much smaller detachment operational window. These results suggest that stable fully detached divertor operation can be realized in tokamaks with extended divertor legs.

  8. Critical Assessment of Pressure Gauges for ITER

    SciTech Connect

    Tabares, Francisco L.; Tafalla, David; Garcia-Cortes, Isabel

    2008-03-12

    The density and flux of molecular species in ITER, largely dominated by the molecular form of the main plasma components and the He ash, is a valuable parameter of relevance not only for operation purposes but also for validating existing neutral particle models of direct implications in divertor performance. An accurate and spatially resolved monitoring of this parameter implies the proper selection of pressure gauges able to cope with the very unique and aggressive environment to be expected in a fusion reactor. To date, there is no standard gauge fulfilling all the requirements, which encompass high neutron and gamma fluxes, together with strong magnetic field and temperature excursions and dusty environment. In the present work, a review of the challenges to face in the measurement of neutral pressure in ITER, together with existing technologies and developments to be made in some of them for their application to the task is presented. Particular attention is paid to R and D needs of existing concepts with potential use in future designs.

  9. Spectroscopic problems in ITER diagnostics

    NASA Astrophysics Data System (ADS)

    Lisitsa, V. S.; Bureyeva, L. A.; Kukushkin, A. B.; Kadomtsev, M. B.; Krupin, V. A.; Levashova, M. G.; Medvedev, A. A.; Mukhin, E. E.; Shurygin, V. A.; Tugarinov, S. N.; Vukolov, K. Yu

    2012-12-01

    Problems of spectroscopic diagnostics of ITER plasma are under consideration. Three types of diagnostics are presented: 1) Balmer lines spectroscopy in the edge and divertor plasmas; 2) Thomson scattering, 3) charge exchange recombination spectroscopy. The Zeeman-Stark structure of line shapes is discussed. The overlapping of isotopes H-D-T spectral line shapes are presented for the SOL and divertor conditions. The polarization measurements of H-alpha spectral lines for H-D mixture on T-10 tokamak are shown in order to separate Zeeman splitting in more details. The problem of plasma background radiation emission for Thomson scattering in ITER is discussed in details. The line shape of P-7 hydrogen spectral line having a wave length close to laser one is presented together with continuum radiation. The charge exchange recombination spectroscopy (CXRS) is discussed in details. The data on Dα, HeII and CVI measurements in CXRS experiments on T-10 tokamak are presented.

  10. 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.

  11. 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.

  12. 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.

  13. Initial Results from the C-Mod Divertor Thomson Scattering System

    NASA Astrophysics Data System (ADS)

    Grek, B.; Johnson, D.; Paladino, R.; Bartolick, J.; Dimock, D.; Lowrance, J.; Lipshultz, B.; Labombard, B.

    1996-11-01

    Thomson scattering system has been installed recently to diagnose the x-point and divertor plasma regions with a resolution of 2-3 mm over a 12 cm field. The light scattered from a 30 HZ Nd:YAG laser is viewed from below through a slot in the outer divertor plate with a reentrant, high throughput collection system. A compact laser dump is located inside the inner divertor plate. Laser alignment is maintained under feedback control to track vessel motion. A filter polychromator spectrally resolves the scattered light from 25 spatial positions onto four 25 element avalanche photodiode arrays. System performance is described in terms of both calibration results and initial measurements of divertor plasma parameters. Supported by U.S. DOE Contract No. DE-AC02-78ET51013, DE-AC02-76-CHO-3073 and SBIR Grant No. 20431-92-II.

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

    DOE Data Explorer

    Frerichs, H. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Waters, I. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Schmitz, O. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Canal, G. P. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Evans, T. E. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Feng, Y. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Soukhanovskii, V. A. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)

    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 (transients), but understanding their detailed impact is difficult due to their symmetry breaking nature. One approach for reducing steady state heat loads are 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 is 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.

  15. VUV Spectroscopy in DIII-D Divertor

    SciTech Connect

    Alkesh Punjabi; Nelson Jalufka

    2004-11-04

    The research carried out on this grant was motivated by the high power emission from the CIV doublet at 155 nm in the DIII-D divertor and to study the characteristics of the radiative divertor. The radiative divertor is designed to reduce the heat load to the target plates of the divertor by reducing the energy in the divertor plasma using upstream scrape-off-layer (SOL) radiation. In some cases, particularly in Partially Detached Divertor (PDD) operations, this emission accounts for more than 50% of the total radiation from the divertor. In PDD operation, produced by neutral gas injection, the particle flow to the target plate and the divertor temperature are significantly reduced. A father motivation was to study the CIV emission distribution in the lower, open divertor and the upper baffled divertor. Two Vacuum Ultra Violet Tangential viewing Television cameras (VUV TTV) were constructed and installed in the upper, baffled and the lower, open divertor. The images recorded by these cameras were then inverted to produce two-dimensional distributions of CIV in the poloidal plane. Results obtained in the project are summarized in this report.

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

    NASA Astrophysics Data System (ADS)

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

    2017-08-01

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

  17. Simulating Divertor Detachment of Ohmic Discharges in ASDEX Upgrade Using SOLPS: the Role of Carbon

    SciTech Connect

    Wischmeier, M; Coster, D; Chankin, A; Fuchs, C; Groth, M; Harhausen, J; Kallenbach, A; Muller, H; Tsalas, M; Wolfrum, E

    2007-06-27

    With divertor detachment being a prerequisite for burning plasma operation in ITER, numerical codes such as SOLPS [1] have been developed for predicting and interpreting the divertor performance at all operational regimes in current tokamaks and ITER. In ITER complete detachment from the outer divertor target is not permitted as this might result in an X-point MARFE, imposing an upper limit for the upstream separatrix density, n{sub e}{sup sep}. Despite the knowledge of the basic mechanisms required for achieving detachment, such as radiative power exhaust, volumetric momentum and charge removal [1], a quantitative evaluation of experimentally observed detached regimes proves to be particularly difficult for several tokamaks. In particular the strong asymmetry of the ion flux density between the inner, {Lambda}{sub it}, and the outer target {Lambda}{sub ot} with increasing line averaged density, {bar n}{sub e}, and in particular ''vanishing'' of the ion flux, defined as full/complete detachment, at the inner target cannot be reproduced. It is unclear how this is related to divertor target plates or other plasma facing components containing carbon. As part of a combined effort at various experimental devices this paper contributes to the validation of the SOLPS code against experimental data from ASDEX Upgrade, AUG, at the onset of divertor detachment. In the framework established under the International Tokamak Physics Activity (ITPA) Divertor and SOL working group a series of ohmic discharges have been performed in AUG, which had as similar as possible plasma parameters as companion discharges undertaken in DIII-D [2]. The effect of activating drift terms, the influence of the chemical sputtering yield at the inner target and in addition to [3] the role of impurity influx from the inner heat shield are analyzed.

  18. Tokamak Physics Experiment divertor design

    SciTech Connect

    Anderson, P.M.

    1995-12-31

    The Tokamak Physics Experiment (TPX) tokamak requires a symmetric up/down double-null divertor capable of operation with steady-state heat flux as high as 7.5 MW/m{sup 2}. The divertor is designed to operate in the radiative mode and employs a deep slot configuration with gas puffing lines to enhance radiative divertor operation. Pumping is provided by cryopumps that pump through eight vertical ports in the floor and ceiling of the vessel. The plasma facing surface is made of carbon-carbon composite blocks (macroblocks) bonded to multiple parallel copper tubes oriented vertically. Water flowing at 6 m/s is used, with the critical heat flux (CHF) margin improved by the use of enhanced heat transfer surfaces. In order to extend the operating period where hands on maintenance is allowed and to also reduce dismantling and disposal costs, the TPX design emphasizes the use of low activation materials. The primary materials used in the divertor are titanium, copper, and carbon-carbon composite. The low activation material selection and the planned physics operation will allow personnel access into the vacuum vessel for the first 2 years of operation. The remote handling system requires that all plasma facing components (PFCs) are configured as modular components of restricted dimensions with special provisions for lifting, alignment, mounting, attachment, and connection of cooling lines, and instrumentation and diagnostics services.

  19. Controlling marginally detached divertor plasmas

    DOE PAGES

    Eldon, David; Kolemen, Egemen; Barton, Joseph L.; ...

    2017-05-04

    A new control system at DIII-D has stabilized the inter-ELM detached divertor plasma state for H-mode in close proximity to the threshold for reattachment, thus demonstrating the ability to maintain detachment with minimal gas puffing. When the same control system was instead ordered to hold the plasma at the threshold (here defined as Te = 5 eV near the divertor target plate), the resulting Te profiles separated into two groups with one group consistent with marginal detachment, and the other with marginal attachment. The plasma dithers between the attached and detached states when the control system attempts to hold at the threshold. The control system is upgraded from the one described in and it handles ELMing plasmas by using real time Dα measurements to remove during-ELM slices from real time Te measurements derived from divertor Thomson scattering. The difference between measured and requested inter-ELM Te is passed to a PID (proportionalintegral-derivative) controller to determine gas puff commands. While some degree of detachment is essential for the health of ITER’s divertor, more deeply detached plasmas have greater radiative losses and, at the extreme, confinement degradation, making it desirable to limit detachment to the minimum level needed to protect the target plate. However, the observed bifurcation in plasma conditions at the outer strike point with the ion B ×more » $$\

  20. Surface erosion issues and analysis for dissipative divertors

    SciTech Connect

    Brooks, J.N.; Ruzic, D.N.; Hayden, D.B.; Turkot, R.B. Jr.

    1994-08-01

    Erosion/redeposition is examined for the sidewall of a dissipative divertor using coupled impurity transport, charge exchange, and sputtering codes, applied to a plasma solution for the ITER design. A key issue for this regime is possible runaway self-sputtering, due to the effect of a low boundary density and nearly parallel field geometry on redeposition parameters. Net erosion rates, assuming finite self-sputtering, vary with wall location, boundary conditions, and plasma solution, and are roughly of the following order: 200--2000 {angstrom}/s for beryllium, 10--100 {angstrom}/s for vanadium, and 0.3--3 {angstrom}/s for tungsten.

  1. 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.

  2. Tile profiling analysis of samples from the JET ITER-like wall and carbon wall

    NASA Astrophysics Data System (ADS)

    Heinola, K.; Ayres, C. F.; Baron-Wiechec, A.; Coad, J. P.; Likonen, J.; Matthews, G. F.; Widdowson, A.; Contributors, JET-EFDA

    2014-04-01

    A complete global balance for material transport in JET requires knowledge of the net erosion in the main chamber, net deposition in the divertor and the amount of dust in the divertor region. Following the end of the first JET ITER-like wall campaign a set of tiles has been removed from the main chamber and the divertor. This paper describes the initial tile surface profiling results for evaluating the erosion in the main chamber and deposition in the divertor. Tile profiling was performed on upper dump plate tiles and inner wall guard limiters made of beryllium and on inner divertor tiles made of tungsten coated carbon (C)-fibre composites. Additionally, the mass of dust collected from the JET divertor is also reported. Present results are compared with JET-C campaign results with the all-carbon C wall.

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

    NASA Astrophysics Data System (ADS)

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

    1996-11-01

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

  4. 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.

  5. Controlling marginally detached divertor plasmas

    NASA Astrophysics Data System (ADS)

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

    2017-06-01

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

  6. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2016-12-01

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

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

    SciTech Connect

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

    2016-09-14

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

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

    DOE PAGES

    Guo, H. Y.; Hill, D. N.; Leonard, A. W.; ...

    2016-09-14

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

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

    SciTech Connect

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

    2016-09-14

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

  11. 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.

  12. Reduction in resonant magnetic field induced heat flux splitting caused by detachment of the divertor

    NASA Astrophysics Data System (ADS)

    Briesemeister, A. R.; Ahn, J.-W.; Hillis, D. L.; Lore, J. D.; Shafer, M. W.; Unterberg, E. A.; Wingen, A.; Schmitz, O.; Frerichs, H.; Makowski, M. A.; McLean, A. G.; Ferraro, N. M.

    2015-11-01

    Measurements in DIII-D show that in high-density detached divertor conditions, the inter-ELM non-axisymmetric heat flux striations generated by resonant magnetic perturbations (RMPs) are eliminated. Non-axisymmetric heat loads caused by the RMP fields used to mitigate ELMs could reduce the lifetime of divertor components in ITER and future devices. It is shown that for RMPs with an n=3 toroidal mode number low levels of gas puffing can cause an increase in the heat flux splitting, but at high densities where the divertor becomes detached this splitting is eliminated. VUV imaging and 2D divertor Thomson scattering are used to measure RMP induced perturbations to the plasma conditions above the target plates. Modeling performed with the 3D fluid transport code EMC3-EIRENE both with and without the plasma response calculated by M3D-C1 is compared to the measured divertor conditions. Work supported by the US DOE under DE-AC05-00OR22725, DE-FC02-04ER54698, DE-AC52-07NA27344 & DE-FG02-92ER54139.

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

    NASA Astrophysics Data System (ADS)

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

    2016-10-01

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

  14. 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.

  15. 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.

  16. 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.

  17. The lithium vapor box divertor

    SciTech Connect

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

    2016-01-13

    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. Our 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. Furthermore, 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 in order 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.

  18. The lithium vapor box divertor

    DOE PAGES

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

    2016-01-13

    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. Our 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 asmore » 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. Furthermore, 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 in order 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.« less

  19. Testing the role of molecular physics in dissipative divertor operations through helium plasmas at DIII-D

    DOE PAGES

    Canik, John M.; Briesemeister, Alexis R.; McLean, Adam G.; ...

    2017-05-10

    Recent experiments in DIII-D helium plasmas are examined to resolve the role of atomic and molecular physics in major discrepancies between experiment and modeling of dissipative divertor operation. Helium operation removes the complicated molecular processes of deuterium plasmas that are a prime candidate for the inability of standard fluid models to reproduce dissipative divertor operation, primarily the consistent under-prediction of radiated power. Modeling of these experiments shows that the full divertor radiation can be accounted for, but only if measures are taken to ensure that the model reproduces the measured divertor density. Relying on upstream measurements instead results in amore » lower divertor density and radiation than is measured, indicating a need for improved modeling of the connection between the diverter and the upstream scrape-off layer. Furthermore, these results show that fluid models are able to quantitatively describe the divertor-region plasma, including radiative losses, and indicate that efforts to improve the fidelity of the molecular deuterium models are likely to help resolve the discrepancy in radiation for deuterium plasmas.« less

  20. ELM-induced transient tungsten melting in the JET divertor

    NASA Astrophysics Data System (ADS)

    Coenen, J. W.; Arnoux, G.; Bazylev, B.; Matthews, G. F.; Autricque, A.; Balboa, I.; Clever, M.; Dejarnac, R.; Coffey, I.; Corre, Y.; Devaux, S.; Frassinetti, L.; Gauthier, E.; Horacek, J.; Jachmich, S.; Komm, M.; Knaup, M.; Krieger, K.; Marsen, S.; Meigs, A.; Mertens, Ph.; Pitts, R. A.; Puetterich, T.; Rack, M.; Stamp, M.; Sergienko, G.; Tamain, P.; Thompson, V.; Contributors, JET-EFDA

    2015-02-01

    The original goals of the JET ITER-like wall included the study of the impact of an all W divertor on plasma operation (Coenen et al 2013 Nucl. Fusion 53 073043) and fuel retention (Brezinsek et al 2013 Nucl. Fusion 53 083023). ITER has recently decided to install a full-tungsten (W) divertor from the start of operations. One of the key inputs required in support of this decision was the study of the possibility of W melting and melt splashing during transients. Damage of this type can lead to modifications of surface topology which could lead to higher disruption frequency or compromise subsequent plasma operation. Although every effort will be made to avoid leading edges, ITER plasma stored energies are sufficient that transients can drive shallow melting on the top surfaces of components. JET is able to produce ELMs large enough to allow access to transient melting in a regime of relevance to ITER. Transient W melt experiments were performed in JET using a dedicated divertor module and a sequence of IP = 3.0 MA/BT = 2.9 T H-mode pulses with an input power of PIN = 23 MW, a stored energy of ˜6 MJ and regular type I ELMs at ΔWELM = 0.3 MJ and fELM ˜ 30 Hz. By moving the outer strike point onto a dedicated leading edge in the W divertor the base temperature was raised within ˜1 s to a level allowing transient, ELM-driven melting during the subsequent 0.5 s. Such ELMs (δW ˜ 300 kJ per ELM) are comparable to mitigated ELMs expected in ITER (Pitts et al 2011 J. Nucl. Mater. 415 (Suppl.) S957-64). Although significant material losses in terms of ejections into the plasma were not observed, there is indirect evidence that some small droplets (˜80 µm) were released. Almost 1 mm (˜6 mm3) of W was moved by ˜150 ELMs within 7 subsequent discharges. The impact on the main plasma parameters was minor and no disruptions occurred. The W-melt gradually moved along the leading edge towards the high-field side, driven by j × B forces. The evaporation rate determined

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

    NASA Astrophysics Data System (ADS)

    Chaffin, Michael; Maingi, Rajesh

    2007-11-01

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

  2. Co-deposited layers in the divertor region of JET-ILW

    NASA Astrophysics Data System (ADS)

    Petersson, P.; Rubel, M.; Esser, H. G.; Likonen, J.; Koivuranta, S.; Widdowson, A.

    2015-08-01

    Tungsten-coated carbon tiles from a poloidal cross-section of the divertor and several types of erosion-deposition probes from the shadowed areas in the divertor were studied using heavy ion elastic recoil detection to obtain quantitative and depth-resolved deposition patterns. Deuterium, beryllium, carbon, nitrogen and oxygen along with tungsten and Inconel components are the main species detected in the studied surface region. The top of Tile 1 in the inner divertor is the main deposition area where the greatest amounts of deposited species are measured. Beryllium and tungsten-containing deposits on the probes (test mirrors and quartz microbalance) indicate that both low-Z and high-Z metals are transported to remote areas. Deposition of nitrogen-15 tracer used for edge cooling only at the end of experimental campaigns in 2012 was also detected giving evidence that nitrogen is effectively retained in wall components.

  3. 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.

  4. Extending Helium Partial Pressure Measurement Technology to JET DTE2 and ITER

    SciTech Connect

    Klepper, C Christopher; Biewer, Theodore M; Douai, D.; Hillis, Donald Lee; Marcus, Chris; Kruezi, Uron

    2016-01-01

    The detection limit for helium (He) partial pressure monitoring via the Penning discharge optical emission diagnostic, mainly used for tokamak divertor effluent gas analysis, is shown here to be possible for He concentrations down to 0.1% in predominantly deuterium effluents. This result from a dedicated laboratory study means that the technique can now be extended to intrinsically (non-injected) He produced as fusion reaction ash in deuterium-tritium experiments. The paper also examines threshold ionization mass spectroscopy as a potential backup to the optical technique, but finds that further development is needed to attain with plasma pulse-relevant response times. Both these studies are presented in the context of continuing development of plasma pulse-resolving, residual gas analysis for the upcoming JET deuterium-tritium campaign (DTE-2) and for ITER.

  5. Extending helium partial pressure measurement technology to JET DTE2 and ITER

    NASA Astrophysics Data System (ADS)

    Klepper, C. C.; Biewer, T. M.; Kruezi, U.; Vartanian, S.; Douai, D.; Hillis, D. L.; Marcus, C.

    2016-11-01

    The detection limit for helium (He) partial pressure monitoring via the Penning discharge optical emission diagnostic, mainly used for tokamak divertor effluent gas analysis, is shown here to be possible for He concentrations down to 0.1% in predominantly deuterium effluents. This result from a dedicated laboratory study means that the technique can now be extended to intrinsically (non-injected) He produced as fusion reaction ash in deuterium-tritium experiments. The paper also examines threshold ionization mass spectroscopy as a potential backup to the optical technique, but finds that further development is needed to attain with plasma pulse-relevant response times. Both these studies are presented in the context of continuing development of plasma pulse-resolving, residual gas analysis for the upcoming JET deuterium-tritium campaign (DTE2) and for ITER.

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

    NASA Astrophysics Data System (ADS)

    Reichle, R.; Brichard, B.; Pocheau, C.; Jouve, M.; van Ierschot, S.; Martinez, S.; Ooms, H.; Berghmans, F.; Decréton, M.

    2008-03-01

    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 Co60 irradiation facilities under γ irradiation. The fibres were ZrF4 (and HfF4) 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.

  7. In-vessel dust and tritium control strategy in ITER

    NASA Astrophysics Data System (ADS)

    Shimada, M.; Pitts, R. A.; Ciattaglia, S.; Carpentier, S.; Choi, C. H.; Dell Orco, G.; Hirai, T.; Kukushkin, A.; Lisgo, S.; Palmer, J.; Shu, W.; Veshchev, E.

    2013-07-01

    A baseline strategy for dust and tritium-inventory control and recovery in ITER has been established and preparations are underway for its implementation. Limits on dust and tritium-inventory are an integral part of the ITER safety case and are fixed at 1 kg for tritium, 1000 kg for mobilisable dust and 11 kg (beryllium)/76 kg (tungsten) for dust on hot surfaces. Maximum average T-retention rates of ˜1 g/shot are estimated for baseline inductive operation at QDT = 10, suggesting that the in-vessel T-retention could reach the administrative limit of 640 g in as little as ˜2 months of operation. Baking is expected to remove a significant fraction of the T co-deposited on the divertor targets. Despite large uncertainties, dust quantities are expected to remain well below safety limits over the divertor cassette lifetime. In situ aspiration during divertor cassette exchange is foreseen as the main dust removal technique.

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

    NASA Astrophysics Data System (ADS)

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

    2017-01-01

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

  9. Advanced Divertor Developments at DIII-D

    NASA Astrophysics Data System (ADS)

    Kolemen, E.; Allen, S. L.; Makowski, M. A.; Soukhanovskii, V. A.; Bray, B. D.; Eldon, D.; Humphreys, D. A.; Johnson, R.; Leonard, A. W.; Liu, C.; Penaflor, B. G.; Petrie, T. W.; McLean, A. G.; Unterberg, E. A.

    2013-10-01

    Novel divertor configurations and control schemes have been implemented at DIII-D to test and optimize heat and particle handling capabilities for advanced tokamaks. The snowflake configuration is stabilized by first calculating the position of the two null-points using real-time equilibrium reconstruction and then regulating the shaping coil currents. Experiments in which the snowflake divertor is stabilized for many confinement times show that it is compatible with high-performance operation and results in greatly reduced divertor heat flux. An advanced divertor control system regulates the gas injection to achieve partial or full detachment by using the divertor temperature measurements from real-time Thomson diagnostics and a line ratio measurement, and adjusts the core and divertor radiation via measurement of the real-time bolometer diagnostics. Prospects of achieving acceptable divertor target heat fluxes for future fusion reactors are analyzed and challenges are presented. Work supported by the US DOE under DE-AC02-09CH11466, DE-AC52-07NA27344, DE-FC02-04ER54698 and DE-AC05-00OR22725.

  10. Engineering design of a radiative divertor for DIII-D

    SciTech Connect

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

    1995-10-01

    A new divertor configuration is being developed for the DIII-D tokamak. This divertor will operate in the radiative mode. Experiments and modeling form the basis for the new design. The Radiative Divertor reduces the heat flux on the divertor plates by dispersing the power with radiation in the divertor region. In addition, the Radiative Divertor structure will allow density control in plasma shapes required for advanced tokamak operation. The divertor structure allows for operation in either double-null or single-null plasma configurations. Four independently controlled divertor cryopumps will enable pumping at either the inboard (upper and lower) or the outboard (upper and lower) divertor plates. An upgrade to the DIII-D cryogenic system is part of this project. The increased capabilities of the cryogenic system will allow delivery of liquid helium and nitrogen to the three new cryopumps. The Radiative Divertor design is very flexible, and will allow physics studies of the effects of slot width and length. Radiative Divertor diagnostics are being designed in parallel to provide comprehensive measurements for diagnosing the divertor. The Radiative divertor installation is scheduled for late 1996. Engineering experience gained in the DIII-D Advanced Divertor program form a foundation for the design work on the Radiative Divertor.

  11. Material migration patterns and overview of first surface analysis of the JET ITER-like wall

    NASA Astrophysics Data System (ADS)

    Widdowson, A.; Alves, E.; Ayres, C. F.; Baron-Wiechec, A.; Brezinsek, S.; Catarino, N.; Coad, J. P.; Heinola, K.; Likonen, J.; Matthews, G. F.; Mayer, M.; Rubel, M.; Contributors, JET-EFDA

    2014-04-01

    Following the first JET ITER-like wall operations a detailed in situ photographic survey of the main chamber and divertor was completed. In addition, a selection of tiles and passive diagnostics were removed from the vessel and made available for post mortem analysis. From the photographic survey and results from initial analysis, the first conclusions regarding erosion, deposition, fuel retention and material transport during divertor and limiter phases have been drawn. The rate of deposition on inner and outer base divertor tiles and remote divertor corners was more than an order of magnitude less than during the preceding carbon wall operations, as was the concomitant deuterium retention. There was however beryllium deposition at the top of the inner divertor. The net beryllium erosion rate from the mid-plane inner limiters was found to be higher than for the previous carbon wall campaign although further analysis is required to determine the overall material balance due to erosion and re-deposition.

  12. ITER research plan of plasma-wall interaction

    NASA Astrophysics Data System (ADS)

    Shimada, M.; Pitts, R.; Loarte, A.; Campbell, D. J.; Sugihara, M.; Mukhovatov, V.; Kukushkin, A.; Chuyanov, V.

    2009-06-01

    This paper describes an important part of ITER Research Plan, on plasma-wall interaction (PWI). In order to maximize the flexibility of the machine during the initial operation (H and D phases), CFC will be used for the targets. Tungsten will be used for the other plasma-facing components of the divertor. In order to minimize the tritium retention, tungsten will fully cover the divertor targets before the DT phase. Extrapolation of heat loads on plasma-facing components (PFCs) during disruption and ELMs to ITER parameters indicates serious consequences of these phenomena. Therefore schemes for prediction and mitigation or avoidance of these phenomena need to be developed during construction and demonstrated in the early phase of ITER operation. T-retention and dust have important impacts on safety. Therefore the methods of measurement and removal of tritium and dust must be developed during construction and demonstrated in the early phase of ITER operation.

  13. 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.

  14. Design of divertor plate and measurements of double-null open divertor plasma in the JFT-2M tokamak

    NASA Astrophysics Data System (ADS)

    Yanagisawa, Ichiro; Shoji, Teruaki; Mori, Masahiro; Odajima, Kazuo; Ohtsuka, Hideo; Suzuki, Norio; Hasegawa, Mitsuru; Ohta, Kanji; Sugihara, Masayoshi; Uesugi, Yoshihiko

    1987-10-01

    The Design of the divertor plate, the results of the computational simulation and the experimental results on the compact diverter of the JFT-2 tokamak are described. Graphite divertor plates have showed a good performance as divertor target materials through divertor discharges. The H-mode plasma and low temperature, high density divertor plasma are obtained. From computational results, this is in the intermediate region between low and high recycling region.

  15. 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)

  16. Near infrared spectroscopy of the DIII-D divertor

    NASA Astrophysics Data System (ADS)

    McLean, A. G.; Soukhanovskii, V. A.; Brooks, N. H.; Bray, B. D.; Carlstrom, T. N.

    2012-10-01

    A high speed, high resolution near infrared (NIR) spectrometer has been installed at DIII-D to make first-of-its-kind observations of the 0.8-2.2 μm region in a tokamak divertor. The goals of this diagnostic are (1) to study Paschen spectra for line-averaged measurement of low temperature plasma parameters, (2) to benchmark the chemical and physically sputtered sources of neutral carbon using the lineshape of the CI, 910 nm multiplet, and (3) to quantify contamination of the 0.75-1.1 μm region where Thomson-shifted laser light is measured by the Thomson scattering diagnostic. Diagnostic capabilities include a 300 mm, f/3.9 design, 300-2400 Gr/mm gratings providing optical resolution of ˜0.65-0.04 nm, and readout at up to 900 frames/second. Data are presented in L-mode plasmas, and in H-mode between ELMs and during the ELM peak. Results acquired by this diagnostic will be applied to design of a proposed divertor Thomson diagnostic for NSTX-U and aid validation of the Thomson system on ITER.

  17. 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.

  18. Effect of Divertors in NCSX

    NASA Astrophysics Data System (ADS)

    Kaiser, Thomas B.; Hill, David N.

    2004-11-01

    We have used magnetic field data generated by the PIES 3D MHD equilibrium code (M50 coil set) and a new vacuum field code [1] together with the latest numerical model of the first wall [2] to compute wall heat-loading in the National Compact Stellarator Experiment (NCSX). Heat flow is traced by following field lines, with field-line diffusion used to mimic the effect of particle scattering, and the local heat flux estimated from the strike-point density of escaping field lines. This extends our earlier work [3] by including the effect of divertors, whose size, location and configuration are varied to minimize estimated wall damage. Error scaling of the field-line integrator is also presented. 1. Michael Drevlak, MPIPP, Greifswald, Germany, private communication 2. Art Brooks, PPPL, private communication. 3. T. B. Kaiser, et al, BAPPS 48, 304 (2003).

  19. 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

  20. 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.

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

    SciTech Connect

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

    2016-06-15

    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.

  2. Design of Divertor Scraper Elements for the W7-X Stellarator

    NASA Astrophysics Data System (ADS)

    Harris, Jeffrey; Lumsdaine, Arnold; Canik, John; Lore, Jeremy; McGinnis, Dean; Peacock, Alan; Hurd, Fred; Boscary, Jean; Geiger, Joachim; Tipton, Joseph

    2011-10-01

    A PPPL/ORNL/LANL team is partnering with the Max-Planck Institut für Plasmaphysik in the Wendelstein 7-X (W7-X) stellarator project. W7-X is a large superconducting, steady-state stellarator (R = 5.5, a = 0.5, B = 3T) with P =15-30 MW that will begin operation in 2015. The US team is focusing on control of the magnetic configuration and divertor heat flux. The W7-X divertor consists of cooled CFC plates arranged as a magnetic island divertor outside the last closed flux surface. While the W7-X configuration is optimized to minimize both Pfirsch-Schlüter and bootstrap currents, the ~30 sec evolution of the plasma to its final equilibrium drives bootstrap currents which transiently alter the distribution of divertor heat flux. This necessitates the addition of 10 actively cooled scraper elements (dimensions ~0.2 m x 1 m) capable of absorbing localized heat fluxes < 12 MW/m2. ORNL/IPP are developing an engineering design for the scraper elements using ITER CFC monoblock technology. Work supported by US Department of Energy.

  3. ARIES-III divertor engineering design

    SciTech Connect

    Wong, C.P.C.; Schultz, K.R.; Cheng, E.T.; Grotz, S.; Hasan, M.A.; Najmabadi, F.; Sharafat, S.; Brooks, J.N.; Ehst, D.A.; Sze, D.K.; Herring, J.S.; Valenti, M.; Steiner, D.

    1992-01-01

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

  4. Recent DIII-D divertor research

    SciTech Connect

    Allen, S.L.; Bozek, A.S.; Brooks, N.H.

    1995-07-01

    DIII-D currently operates with a single- or double-null open divertor and graphite walls. Active particle control with a divertor cryopump has demonstrated density control, efficient helium exhaust, and reduction of the inventory of particles in the wall. Gas puffing of D{sub 2} and impurities has demonstrated reduction of the peak divertor beat flux by factors of 3--5 by radiation. A combination of active cryopumping and feedback-controlled D{sub 2} gas puffing has produced similar divertor heat flux reduction with density control. Experiments with neon puffing have shown that the radiation is equally-divided between a localized zone near the X-point and a mantle around the plasma core. The density in these experiments has also been controlled with cryopumping. These experimental results combined with modeling were used to develop the new Radiative Divertor for DIII-D. This is a double-null slot divertor with four cryopumps to provide particle control and neutral shielding for high-triangularity advanced tokamak discharges. UEDGE and DEGAS simulations, benchmarked to experimental data, have been used to optimize the design.

  5. Predictive modelling for EAST divertor operation

    NASA Astrophysics Data System (ADS)

    Chen, YiPing

    2011-06-01

    The predictive modelling study of the divertor operation in EAST tokamak [B. Wan et al., Nucl. Fusion 49, 104011 (2009)] with double null (DN) configuration is carried out by using the two-dimensional edge plasma code B2.5-SOLPS5.0 [D. P. Coster, X. Bonnin et al., J. Nucl. Mater. 337-339, 366 (2005)]. The modelling study includes the particle and power balance in the scrape-off-layer (SOL), the operation parameters of plasma density, temperature and plasma heat fluxes at the separatrix, the target plates and the wall, and the effect of the gas puffing, drifts, and vertical target plate on the divertor operation. The fluid model for the edge plasma is applied using the real magnetohydrodynamic (MHD) equilibrium from the MHD equilibrium code EFIT [L. L. Lao et al., Nucl. Fusion 25, 1611 (1985)] and the real divertor geometry in the device. Before EAST tokamak starts its experimental programme of divertor operation, the modelling plays an important role in the design of its experimental programme and the optimization of the divertor operation parameters. Based on the modelling results, EAST divertor can operate over a large wide of plasma parameters with different regimes. For a heating power of 8 MW and an edge density at core-SOL interface Nedge = 0.8 × 10191/m3 and Nedge = 1.3 × 10191/m3, the EAST divertor begins access to the high recycling operation regime at the outer and inner target plates, respectively, where the plasma temperature and the heat fluxes at the target plates decrease. The gas puffing can increase the plasma density at the separatrix and trigger the transition from the high recycling operation into detachment at the target plates. When E × B and B × ▿B drifts are taken into account, the asymmetry of plasma parameters and heat fluxes between up-down SOLs can be found. The vertical target plate in EAST divertor can reduce the peak values of heat fluxes at the target plate and enables detachment at lower plasma density. The divertor with the

  6. Predictive modelling for EAST divertor operation

    SciTech Connect

    Chen Yiping

    2011-06-15

    The predictive modelling study of the divertor operation in EAST tokamak [B. Wan et al., Nucl. Fusion 49, 104011 (2009)] with double null (DN) configuration is carried out by using the two-dimensional edge plasma code B2.5-SOLPS5.0 [D. P. Coster, X. Bonnin et al., J. Nucl. Mater. 337-339, 366 (2005)]. The modelling study includes the particle and power balance in the scrape-off-layer (SOL), the operation parameters of plasma density, temperature and plasma heat fluxes at the separatrix, the target plates and the wall, and the effect of the gas puffing, drifts, and vertical target plate on the divertor operation. The fluid model for the edge plasma is applied using the real magnetohydrodynamic (MHD) equilibrium from the MHD equilibrium code EFIT [L. L. Lao et al., Nucl. Fusion 25, 1611 (1985)] and the real divertor geometry in the device. Before EAST tokamak starts its experimental programme of divertor operation, the modelling plays an important role in the design of its experimental programme and the optimization of the divertor operation parameters. Based on the modelling results, EAST divertor can operate over a large wide of plasma parameters with different regimes. For a heating power of 8 MW and an edge density at core-SOL interface N{sub edge} = 0.8 x 10{sup 19}1/m{sup 3} and N{sub edge} = 1.3 x 10{sup 19}1/m{sup 3}, the EAST divertor begins access to the high recycling operation regime at the outer and inner target plates, respectively, where the plasma temperature and the heat fluxes at the target plates decrease. The gas puffing can increase the plasma density at the separatrix and trigger the transition from the high recycling operation into detachment at the target plates. When E x B and B x {nabla}B drifts are taken into account, the asymmetry of plasma parameters and heat fluxes between up-down SOLs can be found. The vertical target plate in EAST divertor can reduce the peak values of heat fluxes at the target plate and enables detachment at lower

  7. 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.

  8. 3D modeling of toroidal asymmetry due to localized divertor nitrogen puffing on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Lore, J. D.; Reinke, M. L.; Labombard, B.; Lipschultz, B.; Pitts, R.

    2013-10-01

    For inductive operation at Q =10, ITER will need to run with partially detached divertor plasmas in order to maximize target lifetime and remain below engineering heat-flux limits. The radiated power fraction will be controlled via a divertor gas injection system that consists of six valves. To investigate the effect of potential toroidal asymmetry introduced by a finite number of gas valves, or the failure of one or more valves, experiments were performed on Alcator C-Mod. Nitrogen was injected through each of five toroidally spaced divertor gas valves into Ohmic L-mode plasmas with a high recycling divertor. Clear, reproducible toroidal variation in divertor radiated power and impurity line radiation was measured. The 3D scrape-off-layer transport code EMC3-EIRENE is being used to model and interpret these experiments. Initial results indicate that trends in the radiated power and nitrogen emission asymmetry are reproduced. Both experimental and modeling results will be presented. Work supported by D.O.E. contracts DE-AC05-00OR22725 and DE-FC02-99ER54512.

  9. A method of interpreting the Balmer-alpha high-resolution spectroscopy for tokamak edge plasmas with account of divertor stray light

    NASA Astrophysics Data System (ADS)

    Neverov, V. S.; Kukushkin, A. B.; Alekseev, A. G.

    2016-01-01

    A method is suggested for interpreting the data from the Balmer-alpha high- resolution spectroscopy diagnostics of the edge plasma in the tokamak main chamber, which additionally uses the data from direct observation of the divertor. Such an extension of the diagnostics is motivated by the fact that in a tokamak-reactor with the metal first wall, like ITER tokamak, a significant role of the divertor stray light (DSL), which is emitted by the plasma in the divertor in the same spectral line and reflected from the first wall of the vacuum chamber to a spectrometer in the main chamber, is expected. The results of the first applications of the developed model to interpret the data from the JET-ILW tokamak experiments, which simulate the conditions of occurrence of the DSL in ITER, are discussed.

  10. ITER safety challenges and opportunities

    SciTech Connect

    Piet, S.J.

    1991-01-01

    Results of the Conceptual Design Activity (CDA) for the International Thermonuclear Experimental Reactor (ITER) suggest challenges and opportunities. ITER is capable of meeting anticipated regulatory dose limits,'' but proof is difficult because of large radioactive inventories needing stringent radioactivity confinement. We need much research and development (R D) and design analysis to establish that ITER meets regulatory requirements. We have a further opportunity to do more to prove more of fusion's potential safety and environmental advantages and maximize the amount of ITER technology on the path toward fusion power plants. To fulfill these tasks, we need to overcome three programmatic challenges and three technical challenges. The first programmatic challenge is to fund a comprehensive safety and environmental ITER R D plan. Second is to strengthen safety and environment work and personnel in the international team. Third is to establish an external consultant group to advise the ITER Joint Team on designing ITER to meet safety requirements for siting by any of the Parties. The first of the three key technical challenges is plasma engineering -- burn control, plasma shutdown, disruptions, tritium burn fraction, and steady state operation. The second is the divertor, including tritium inventory, activation hazards, chemical reactions, and coolant disturbances. The third technical challenge is optimization of design requirements considering safety risk, technical risk, and cost. Some design requirements are now too strict; some are too lax. Fuel cycle design requirements are presently too strict, mandating inappropriate T separation from H and D. Heat sink requirements are presently too lax; they should be strengthened to ensure that maximum loss of coolant accident temperatures drop.

  11. Volume Recombination in Alcator C-Mod Divertor Plasmas

    NASA Astrophysics Data System (ADS)

    Terry, J. L.

    1997-11-01

    Volume recombination has been predicted(See, for example, A. Loarte, Proc. 12th PSI Conf, J. Nucl. Mater (1996) I9, in press.) to be a significant sink for plasma ions under the detached divertor conditions achieved on many tokamaks. This volume recombination sink was observed initially in Alcator C-Mod and shown to be a major fraction of the ion loss. Signatures of recombination have now been observed on DIII-D(R.C. Isler, et al., paper submitted for publication), Asdex-UG (B. Napiontek, et al. 24th EPS Conf., Berchtesgaden, Germany, 1997, P4.007, in press.), and JET(R.D. Monk, et al. 24th EPS Conf., Berchtesgaden, Germany, 1997, P1.030, in press.). It is important primarily because the recombined atoms are not accelerated through the sheath - thus reducing divertor plate sputtering, and because most of the potential energy of recombination (13.6 eV) is released as radiation before the ion strikes the plate. The Alcator C-Mod measurements show that the recombination occurs in low Te ( ~1 eV), high ne ( ~1× 10^21 m-3) regions, and is significantly larger in detached regions. At the inboard, detached divertor plate the measured volume recombination rate is typically greater than the rate of ion collection at that plate and is about an order of magnitude higher than on the attached, outer plate. These spatially resolved measurements also show that the recombination rate is peaked near the strike point and imply that the recombination is occurring close to the plate surface. The C-Mod observations about the magnitude and spatial distribution of the recombination are consistent with the modelling of similar discharges(F. Wising et al., Contrib. Plasma Phys. 36, p 136 (1996).). The experimental evidence for recombination is found in the deuterium emission spectra from the divertor, in particular in the Balmer- and/or Lyman-series. The spectra show that the dominant recombination mechanism is 3-body recombination into excited states of deuterium and that the populations

  12. Impact of W on scenario simulations for ITER

    NASA Astrophysics Data System (ADS)

    Hogeweij, G. M. D.; Leonov, V.; Schweinzer, J.; Sips, A. C. C.; Angioni, C.; Calabrò, G.; Dux, R.; Kallenbach, A.; Lerche, E.; Maggi, C.; Pütterich, Th.; ITPA Integrated Operating Scenarios topical Group; ASDEX Upgrade Team; Contributors, JET

    2015-06-01

    In preparation of ITER operation, large machines have replaced their wall and divertor material to W (ASDEX Upgrade) or a combination of Be for the wall and W for the divertor (JET). Operation in these machines has shown that the influx of W can have a significant impact on the discharge evolution, which has made modelling of this impact for ITER an urgent task. This paper reports on such modelling efforts. Maximum tolerable W concentrations have been determined for various scenarios, both for the current ramp-up and flat-top phase. Results of two independent methods are presented, based on the codes ZIMPUR plus ASTRA and CRONOS, respectively. Both methods have been tested and benchmarked against ITER-like Ip RU experiments at JET. It is found that W significantly disturbs the discharge evolution when the W concentration approaches ˜10-4 this critical level varies somewhat between scenarios.

  13. Code development for ITER edge modelling - SOLPS5.1

    NASA Astrophysics Data System (ADS)

    Bonnin, X.; Kukushkin, A. S.; Coster, D. P.

    2009-06-01

    Most ITER divertor modelling work to date used the B2-EIRENE (SOLPS4) code package, coupling a 2D fluid description of the charged plasma species (B2) to a Monte-Carlo kinetic description of the neutrals (EIRENE). In recent years, the emphasis at ITER has been on completing the neutral model, including neutral-neutral collisions, opacity effects, radiation transport, etc. Elsewhere, new physics, numerics, and algorithmic improvements, such as E × B and diamagnetic drifts, electric currents, ion and neutral heat and particle flux limits, wall material mixing and surface temperature evolution, and bundling of heavy ions species, as well as switching to cell-centred velocities and using an internal energy instead of a total energy equation, gave birth to the B2.5 code, combined with EIRENE as SOLPS5. We report on work in progress to merge these advances with the ITER-specific model of the edge and divertor.

  14. Effects of low-Z and high-Z impurities on divertor detachment and plasma confinement

    DOE PAGES

    Wang, H. Q.; Guo, Houyang Y.; Petrie, Thomas W.; ...

    2017-03-18

    The impurity-seeded detached divertor is essential for heat exhaust in ITER and other reactor-relevant devices. Dedicated experiments with injection of N2, Ne and Ar have been performed in DIII-D to assess the impact of the different impurities on divertor detachment and confinement. Seeding with N2, Ne and Ar all promote divertor detachment, greatly reducing heat flux near the strike point. The upstream plasma density at the onset of detachment decreases with increasing impurity-puffing flow rates. For all injected impurity species, the confinement and pedestal pressure are correlated with the impurity content and the ratio of separatrix loss power to themore » L-H transition threshold power. As the divertor plasma approaches detachment, the high-Z impurity seeding tends to degrade the core confinement owing to the increased core radiation. In particular, Ar injection leads to an increase in core radiation, up to 50% of the injected power, and a reduction in pedestal temperature over 60%, thus significantly degrading the confinement, i.e., with H98 reducing from 1.1 to below 0.7. As for Ne seeding, H98 near 0.8 can be maintained during the detachment phase with the pedestal temperature being reduced by about 50%. In contrast, in the N2 seeded plasmas, radiation is predominately confined in the boundary plasma, with up to 50% of heating power being radiated in the divertor region and less than 25% in the core at the onset of detachment. In the case of strong N2 gas puffing, the confinement recovers during the detachment, from ~20% reduction at the onset of the detachment to greater than that before the seeding. The core and pedestal temperatures feature a reduction of 30% from the initial attached phase and remain nearly constant during the detachment phase. The improvement in confinement appears to arise from the increase in pedestal and core density despite the temperature reduction.« less

  15. 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

  16. Overall Features of EAST Operation Space by Using Simple Core-SOL-Divertor Model

    NASA Astrophysics Data System (ADS)

    Hiwatari, R.; Hatayama, A.; Zhu, Sizheng; Takizuka, T.; Tomita, Y.

    2006-01-01

    A simple Core-SOL-Divertor (C-S-D) model has been developed to investigate qualitatively the overall features of the operational space for the integrated core and edge plasma. To construct the simple C-S-D model, a simple core plasma model of ITER physics guidelines and a two-point SOL-divertor model are used. The simple C-S-D model is applied to the study of the EAST operational space with lower hybrid current drive experiments under various kinds of trade-off for the basic plasma parameters. Effective methods for extending the operational space are also presented. From this study for the EAST operational space, it is evident that the C-S-D model is a useful tool for understanding qualitatively the overall features of the plasma operational space.

  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, 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.

  18. 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.

  19. Plasma power recycling at the divertor surface

    SciTech Connect

    Tang, Xian -Zhu; Guo, Zehua

    2016-12-03

    With a divertor made of solid materials like carbon and tungsten, plasma ions are expected to be recycled at the divertor surface with a time-averaged particle recycling coefficient very close to unity in steady-state operation. This means that almost every plasma ion (hydrogen and helium) will be returned to the plasma, mostly as neutrals. The power flux deposited by the plasma on the divertor surface, on the other hand, can have varying recycling characteristics depending on the material choice of the divertor; the run-time atomic composition of the surface, which can be modified by material mix due to impurity migration in the chamber; and the surface morphology change over time. In general, a high-Z–material (such as tungsten) surface tends to reflect light ions and produce stronger power recycling, while a low-Z–material (such as carbon) surface tends to have a larger sticking coefficient for light ions and hence lower power recycling. Here, an explicit constraint on target plasma density and temperature is derived from the truncated bi-Maxwellian sheath model, in relation to the absorbed power load and power recycling coefficient at the divertor surface. Lastly, it is shown that because of the surface recombination energy flux, the attached plasma has a sharper response to power recycling in comparison to a detached plasma.

  20. Plasma power recycling at the divertor surface

    DOE PAGES

    Tang, Xian -Zhu; Guo, Zehua

    2016-12-03

    With a divertor made of solid materials like carbon and tungsten, plasma ions are expected to be recycled at the divertor surface with a time-averaged particle recycling coefficient very close to unity in steady-state operation. This means that almost every plasma ion (hydrogen and helium) will be returned to the plasma, mostly as neutrals. The power flux deposited by the plasma on the divertor surface, on the other hand, can have varying recycling characteristics depending on the material choice of the divertor; the run-time atomic composition of the surface, which can be modified by material mix due to impurity migrationmore » in the chamber; and the surface morphology change over time. In general, a high-Z–material (such as tungsten) surface tends to reflect light ions and produce stronger power recycling, while a low-Z–material (such as carbon) surface tends to have a larger sticking coefficient for light ions and hence lower power recycling. Here, an explicit constraint on target plasma density and temperature is derived from the truncated bi-Maxwellian sheath model, in relation to the absorbed power load and power recycling coefficient at the divertor surface. Lastly, it is shown that because of the surface recombination energy flux, the attached plasma has a sharper response to power recycling in comparison to a detached plasma.« less

  1. Innovative Divertor Development to Solve the Plasma Heat-Flux Problem

    SciTech Connect

    Rognlien, T; Ryutov, D; Makowski, M; Soukhanovskii, V; Umansky, M; Cohen, R; HIll, D; Joseph, I

    2009-02-26

    Large, localized plasma heat exhaust continues to be one of the critical problems for the development of tokamak fusion reactors. Excessive heat flux erodes and possibly melts plasma-facing materials, thereby dramatically shortening their lifetime and increasing the impurity contamination of the core plasma. A detailed assessment by the ITER team for their divertor has revealed substantial limitations on the operational space imposed by the divertor performance. For a fusion reactor, the problem becomes worse in that the divertor must accommodate 20% of the total fusion power (less any broadly radiated loss), while not allowing excess buildup of tritium in the walls nor excessive impurity production. This is an extremely challenging set of problems that must be solved for fusion to succeed as a power source; it deserves a substantial research investment. Material heat-flux constraints: Results from present-day tokamaks show that there are two major limitations of peak plasma heat exhaust. The first is the continuous flow of power to the divertor plates and nearby surfaces that, for present technology, is limited to 10-20 MW/m{sup 2}. The second is the transient peak heat-flux that can be tolerated in a short time, {tau}{sub m}, before substantial ablation and melting of the surface occurs; such common large transient events are Edge Localized Mode (ELMs) and disruptions. The material limits imposed by these events give a peak energy/{tau}{sub m}{sup 1/2} parameter of {approx} 40 MJ/m{sup 2}s{sup 1/2} [1]. Both the continuous and transient limits can be approached by input powers in the largest present-day devices, and future devices are expected to substantially exceed the limits unless a solution can be found. Since the early 90's LLNL has developed the analytic and computational foundation for analyzing divertor plasmas, and also suggested and studied a number of solid and liquid material concepts for improving divertor/wall performance, with the most recent

  2. A design method of divertor in tokamak reactors

    NASA Astrophysics Data System (ADS)

    Ueda, N.; Itoh, S.-I.; Tanaka, M.; Itoh, K.

    1990-08-01

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

  3. 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.

  4. 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.

  5. Liquid Surface Divertor Designs for Fusion Reactors

    SciTech Connect

    Nygren, R; Rognlien, T; Rensink, M

    2003-11-11

    As part of work in the US on free flowing liquid surfaces facing the plasma, we are studying issues of integrating a liquid surface divertor into a configuration based upon an advanced tokamak (ARIES-RS). The simplest form of such a divertor is to extend the flow of the liquid first wall and avoid introducing any separate fluid streams. A design and some of the issues in design integration are presented for a divertor (and first wall) with the molten salt Flinabe, a mixture of lithium and sodium fluorides. Thermal performance and the interactions with the plasma edge are treated. Sn and Sn-Li have also been considered, although the complicated 3-D MHD flows cannot yet be fully modeled.

  6. Design Integration of Liquid Surface Divertors

    SciTech Connect

    Nygren, R E; Cowgill, D F; Ulrickson, M A; Nelson, B E; Fogarty, P J; Rognlien, T D; Rensink, M E; Hassanein, A; Smolentsev, S S; Kotschenreuther, M

    2003-11-13

    The US Enabling Technology Program in fusion is investigating the use of free flowing liquid surfaces facing the plasma. We have been studying the issues in integrating a liquid surface divertor into a configuration based upon an advanced tokamak, specifically the ARIES-RS configuration. The simplest form of such a divertor is to extend the flow of the liquid first wall into the divertor and thereby avoid introducing additional fluid streams. In this case, one can modify the flow above the divertor to enhance thermal mixing. For divertors with flowing liquid metals (or other electrically conductive fluids) MHD (magneto-hydrodynamics) effects are a major concern and can produce forces that redirect flow and suppress turbulence. An evaluation of Flibe (a molten salt) as a working fluid was done to assess a case in which the MHD forces could be largely neglected. Initial studies indicate that, for a tokamak with high power density, an integrated Flibe first wall and divertor does not seem workable. We have continued work with molten salts and replaced Flibe with Flinabe, a mixture of lithium and sodium fluorides, that has some potential because of its lower melting temperature. Sn and Sn-Li have also been considered, and the initial evaluations on heat removal with minimal plasma contamination show promise, although the complicated 3-D MHD flows cannot yet be fully modeled. Particle pumping in these design concepts is accomplished by conventional means (ports and pumps). However, trapping of hydrogen in these flowing liquids seems plausible and novel concepts for entrapping helium are also being studied.

  7. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2014-08-01

    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.

  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. Small angle slot divertor concept for long pulse advanced tokamaks

    NASA Astrophysics Data System (ADS)

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

    2017-04-01

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

  11. 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.

  12. Divertor-leg instability for finite beta and radially-tilted divertor plate

    NASA Astrophysics Data System (ADS)

    Cohen, R. H.; Ryutov, D. D.

    2004-11-01

    Plasma in the divertor leg may experience a fast instability caused by sheath boundary conditions (BC). Perturbations cannot penetrate beyond the X point because of very strong shearing in its vicinity. Accordingly, this instability could increase cross-field transport in the divertor leg, and thereby reduce the heat load on the divertor plate, without having any appreciable negative effect on core plasma confinement. A way of describing the role of shearing in terms of the surface resistivity attributed to a ``control plane'' below the X point has recently been suggested (Contr. Plasma Phys., v. 44, p. 168, 2004). We use this BC, plus sheath BC at the divertor plate. We include effects of finite beta and of the radial tilt of the divertor plate. We optimize the radial tilt in order to maximize radial transport in divertor legs. We discuss experimental signatures of the instability: i) phase velocity and wave-numbers of the most unstable modes; ii) correlations between fluctuations of various parameters; and iii) the differences between fluctuations in the common and private flux regions.

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

    NASA Astrophysics Data System (ADS)

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

    2015-09-01

    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 Ip. 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/Ip 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/Ip scaling, the inclusion of the anomalous transport that is driven by ballooning-type instabilities results in recovering the neoclassical 1/Ip 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.

  14. 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.

  15. 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.

  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. Divertor erosion in DIII-D

    SciTech Connect

    Whyte, D.G.; Bastasz, R.; Wampler, W.R.; Brooks, J.N.; West, W.P.; Wong, C.P.C.; Buzhinskij, O.I.; Opimach, I.V.

    1998-08-01

    Net erosion rates of carbon target plates have been measured in situ for the DIII-D lower divertor. The principal method of obtaining this data is the DiMES sample probe. Recent experiments have focused on erosion at the outer strike-point (OSP) of two divertor plasma conditions: attached (T{sub e} > 40 eV) ELMing plasmas, and detached (T{sub e} < 2 eV) ELMing plasmas. For the attached cases, the erosion rates exceed 10 cm/exposure-year, even with incident heat flux < 1 MW/m{sup 2}. In this case, measurements and modeling agree for both gross and net carbon erosion, showing the near-surface transport and redeposition of the carbon is well understood. In the attached cases, physical sputtering (with enhancement from self-sputtering and oblique incidence) is dominant, and the effective sputtering yield, Y, is greater than 10%. In ELM-free discharges, the total OSP net erosion rate is equal to the rate of carbon accumulation in the core plasma. For the detached divertor cases, the cold incident plasma eliminates physical sputtering. Attempts to measure chemically eroded hydrocarbon molecules spectroscopically indicate an upper limit of Y {le} 0.1% for the chemical sputtering yield. Net erosion is suppressed at the outer strike-point, which becomes a region of net redeposition ({approximately} 4 cm/exposure-year). The private flux wall is measured to be a region of net redeposition with dense, high neutral pressure, attached divertor plasmas. Leading edges intercepting parallel heat flux ({approximately} 50 MW/m{sup 2}) have very high net erosion rates at the OSP of an attached plasma ({approximately} 10 {micro}m/s > 1,000x erosion rate of aligned surfaces). Leading edge erosion, and subsequent carbon redeposition, caused by tile gaps can account for half of the deuterium codeposition in the DIII-D divertor.

  19. Design considerations for ITER (International Thermonuclear Experimental Reactor) plasma facing components

    SciTech Connect

    McGrath, R.T.; Koski, J.A.; Watson, R.D.; Causey, R.A.; Croessmann, C.D.; Dempsey, J.F.; Hosking, M.; Neimer, K.A.; Russo, A.J.; Salmonson, J.C.; Stephens, J.; Smith, M.F.; Watkins, J.G.; Whitley, J.B.

    1989-07-01

    The International Thermonuclear Experimental Reactor (ITER) is a joint design and R D project involving the USA, the Soviet Union, Japan and the European Community. These international partners are working together on the design of a fusion tokamak reactor that will operate in the D-T ignition regime. This report compiles the contributions to ITER made by Sandia National Laboratories in the area of design and R D for plasma facing components, such as the first wall and divertor. The following topics are discussed: divertor fabrication issues, divertor thermal-hydraulic analysis, separatrix sweeping effects, divertor tile 2-D stress analysis, electromechanical disruption effects, runaway electron and intense energy deposition analyses, lifetime analysis and tritium retention in plasma facing materials. Material properties for pyrolytic graphite and beryllium are presented. Use of pyrolytic graphite as the plasma facing material allows for operation with thicker graphite armor at the design heat flux level of 10 MW/m/sup 2/. The design of a divertor coated with plasma sprayed beryllium is presented as an attractive alternative to pyrolytic graphite armor tiles. Finally, the Sandia research and development plan for ITER is discussed. 82 figs.

  20. Analysis of a multi-machine database on divertor heat fluxes

    SciTech Connect

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

    2012-05-15

    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 I{sub p}, 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.

  1. Divertors for Helical Devices: Concepts, Plans, Results, and Problems

    SciTech Connect

    Koenig, R.; Grigull, P.; McCormick, K.

    2004-07-15

    With Large Helical Device (LHD) and Wendelstein 7-X (W7-X), the development of helical devices is now taking a large step forward on the path to a steady-state fusion reactor. Important issues that need to be settled in these machines are particle flux and heat control and the impact of divertors on plasma performance in future continuously burning fusion plasmas. The divertor concepts that will initially be explored in these large machines were prepared in smaller-scale devices like Heliotron E, Compact Helical System (CHS), and Wendelstein 7-AS (W7-AS). While advanced divertor scenarios relevant for W7-X were already studied in W7-AS, other smaller-scale experiments like Heliotron-J, CHS, and National Compact Stellarator Experiment will be used for the further development of divertor concepts. The two divertor configurations that are being investigated are the helical and the island divertor, as well as the local island divertor, which was successfully demonstrated on CHS and just went into operation on LHD. At present, on its route to a fully closed helical divertor, LHD operates in an open helical divertor configuration. W7-X will be equipped right from the start with an actively cooled discrete island divertor that will allow quasi-continuous operation. The divertor design is very similar to the one explored on W7-AS. For sufficiently large island sizes and not too long field line connection lengths, this divertor gives access to a partially detached quasi-steady-state operating scenario in a newly found high-density H-mode operating regime, which benefits from high energy and low impurity confinement times, with edge radiation levels of up to 90% and sufficient neutral compression in the subdivertor region (>10) for active pumping. The basic physics of the different divertor concepts and associated implementation problems, like asymmetries due to drifts, accessibility of essential operating scenarios, toroidal asymmetries due to symmetry breaking error fields

  2. ITER vacuum vessel design (D201 subtask 1.3 and subtask 3). Final report

    SciTech Connect

    1996-08-01

    ITER Task No. D201, Vacuum Vessel Design (Subtask 1.3 and Subtask 3), was initiated to propose and evaluate local vacuum vessel reinforcement alternatives in proximity to the Neutral Beam, Radial Mid-Plane, Top, and Divertor Ports. These areas were reported to be highly stressed regions based on the results of preliminary stress analyses performed by the USHT (US Home Team) and the ITER Joint Central Team (JCT) at the Garching JWS (Joint Work Site). Initial design activities focused on the divertor port region which was reported to experience the highest stress intensities. Existing stress analysis models and results were reviewed with the USHT stress analysts to obtain an overall understanding of the vessel response to the various applied loads. These reviews indicated that the reported stress intensities in the divertor port region were significantly affected by the loads applied to the vessel in adjacent regions.

  3. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2010-10-01

    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 4nm/min deposition can be suppressed by addition of 1Pa·m3s-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.

  5. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2017-04-01

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

  7. Irradiation damage analysis on the flat plate type target plate of the divertor for fusion experimental reactors

    NASA Astrophysics Data System (ADS)

    Ishiyama, S.; Akiba, M.; Eto, M.

    1996-04-01

    To design the relevant plasma facing components of fusion experimental reactors such as ITER, irradiation damage analysis, especially on divertor structures exposed to high heat flux and heavy neutron irradiation, is one of the most important problems. This paper presents finite element analytical results of the thermal and irradiation induced stresses which occurred in the divertor structures which are exposed to neutron irradiation at 0-1 dpa with a high heat flux up to 15 MW/m 2. A type of target plate model of the divertor structure studied in present study e.g. flat plate model has bonded structure of one-dimensional high thermal conductivity carbon-carbon composite (C/C) and oxygen-free high conductivity copper (OFHC), as armor and substrate/heat sink materials, respectively. These results show that irradiation induced stresses at edges of bonded interface between an armor and a substrate/heat sink, become higher with increase of dpa and reach up to the critical values of the materials at 0 and 1 dpa. This indicates that drop-off of armor tiles from substrate structure is one of very serious problems for the safety design of target plate; thus the reduction of service conditions and change of divertor materials are important to extend lifetime of the model.

  8. 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.

  9. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2016-10-01

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

  11. 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.

  12. Factors affecting the measurement accuracy of ITER neutron activation system

    NASA Astrophysics Data System (ADS)

    Cheon, M. S.; Ahn, Y. H.; Pak, S.; Seon, C.; Krasilnikov, V.; Bertalot, L.

    2017-08-01

    One of the main purposes of the ITER2 neutron activation system (NAS) is to evaluate the total neutron production rate from all over the plasma. The measurement accuracy depends on the position and profile of the plasma and the material in front of the irradiation end. It is required to minimize the amount of material and its density variation across the field of view between the plasma and the irradiation end. Due to the radiation and thermal environment of the ITER in-vessel, however, the measurement from ITER NAS cannot avoid the strong influence from in-vessel materials such as the diagnostic first wall, blanket modules, and divertor cassettes, those are located near the irradiation ends. In order to improve the reliability of the measurement in such environment, special cutouts in the diagnostic first wall are introduced near the irradiation end structures located in the port plugs. The effect of the materials and the position and profile of the neutron source in the plasma are evaluated for these irradiation locations, as well as the ones under the divertor cassettes and between blanket modules, by the neutron transport calculation. Calculation results show that simultaneous measurements at upper port and divertor location can provide highly accurate results even without a position or profile correction from other diagnostics.

  13. 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.

  14. 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.

  15. Two-point model for divertor transport

    SciTech Connect

    Galambos, J.D.; Peng, Y.K.M.

    1984-04-01

    Plasma transport along divertor field lines was investigated using a two-point model. This treatment requires considerably less effort to find solutions to the transport equations than previously used one-dimensional (1-D) models and is useful for studying general trends. It also can be a valuable tool for benchmarking more sophisticated models. The model was used to investigate the possibility of operating in the so-called high density, low temperature regime.

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

    SciTech Connect

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

    1999-07-01

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

  17. Divertor bypass in the Alcator C-Mod tokamak

    SciTech Connect

    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.

  18. 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.

  19. 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.

  20. Island divertor studies on W7-AS

    NASA Astrophysics Data System (ADS)

    Sardei, F.; Feng, Y.; Grigull, P.; Herre, G.; Hildebrandt, D.; Hofmann, J. V.; Kisslinger, J.; Brakel, R.; Das, J.; Geiger, J.; Heinrich, O.; Kühner, G.; Niedermeyer, H.; Reiter, D.; Richter-Glötzl, M.; Runov, A.; Schneider, R.; Stroth, U.; Verbeek, H.; Wagner, F.; Wolf, R.; W7-AS Team; NBI Group

    1997-02-01

    Basic topological features of the island divertor concept for low shear stellarators are discussed with emphasis on the differences to tokamak divertors. Extensive measurements of the edge structures by two-dimensional plasma spectroscopy and by target calorimetry are in excellent agreement with predicted vacuum and equilibrium configurations, which are available up to central β values of ˜ 1%. For this β value the calculated field-line pitch inside the islands is twice that of the corresponding vacuum case. Video observations of the strike points indicate stability of the island structures for central β values up to ˜ 3.7%. The interpretation of the complex island divertor physics of W7-AS has become possible by the development of the three-dimensional plasma transport code EMC3 (Edge Monte Carlo 3D), which has been coupled self-consistently to the EIRENE neutral gas code. Analysis of high density NBI discharges gives strong indications of stable high recycling conditions for overlinene ≥ 10 20 m -3. The observations are reproduced by the EMC3/EIRENE code and supported by calculations with the B2/EIRENE code adapted to W7-AS. Improvement of recycling, pumping and target load distribution is expected from the new optimized target plates and baffles to be installed in W7-AS.

  1. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2013-10-01

    New diagnostic, modelling and plant capability on the Mega Ampère Spherical Tokamak (MAST) have delivered important results in key areas for ITER/DEMO and the upcoming MAST Upgrade, a step towards future ST devices on the path to fusion currently under procurement. Micro-stability analysis of the pedestal highlights the potential roles of micro-tearing modes and kinetic ballooning modes for the pedestal formation. Mitigation of edge localized modes (ELM) using resonant magnetic perturbation has been demonstrated for toroidal mode numbers n = 3, 4, 6 with an ELM frequency increase by up to a factor of 9, compatible with pellet fuelling. The peak heat flux of mitigated and natural ELMs follows the same linear trend with ELM energy loss and the first ELM-resolved Ti measurements in the divertor region are shown. Measurements of flow shear and turbulence dynamics during L-H transitions show filaments erupting from the plasma edge whilst the full flow shear is still present. Off-axis neutral beam injection helps to strongly reduce the redistribution of fast-ions due to fishbone modes when compared to on-axis injection. Low-k ion-scale turbulence has been measured in L-mode and compared to global gyro-kinetic simulations. A statistical analysis of principal turbulence time scales shows them to be of comparable magnitude and reasonably correlated with turbulence decorrelation time. Te inside the island of a neoclassical tearing mode allow the analysis of the island evolution without assuming specific models for the heat flux. Other results include the discrepancy of the current profile evolution during the current ramp-up with solutions of the poloidal field diffusion equation, studies of the anomalous Doppler resonance compressional Alfvén eigenmodes, disruption mitigation studies and modelling of the new divertor design for MAST Upgrade. The novel 3D electron Bernstein synthetic imaging shows promising first data sensitive to the edge current profile and flows.

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

    NASA Astrophysics Data System (ADS)

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

    2017-09-01

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

  4. 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.

  5. Concept for a beryllium divertor with in-situ plasma spray surface regeneration

    NASA Astrophysics Data System (ADS)

    Smith, M. F.; Watson, R. D.; McGrath, R. T.; Croessmann, C. D.; Whitley, J. B.; Causey, R. A.

    1990-04-01

    Two serious problems with the use of graphite tiles on the ITER divertor are the limited lifetime due to erosion and the difficulty of replacing broken tiles inside the machine. Beryllium is proposed as an alternative low-Z armor material because the plasma spray process can be used to make in-situ repairs of eroded or damaged surfaces. Recent advances in plasma spray technology have produced beryllium coatings of 98% density with a 95% deposition efficiency and strong adhesion to the substrate. With existing technology, the entire active region of the ITER divertor surface could be coated with 2 mm of beryllium in less than 15 h using four small plasma spray guns. Beryllium also has other potential advantages over graphite, e.g., efficient gettering of oxygen, ten times less tritium inventory, reduced problems of transient fueling from D/T exchange and release, no runaway erosion cascades from self-sputtering, better adhesion of redeposited material, as well as higher strength, ductility, and fracture toughness than graphite. A 2-D finite element stress analysis was performed on a 3 mm thick Be tile brazed to an OFHC soft-copper saddle block, which was brazed to a high-strength copper tube. Peak stresses remained 50% below the ultimate strength for both brazing and in-service thermal stresses.

  6. 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.

  7. Modeling of Hydrogen Line Shapes for the Diagnostic of ITER

    NASA Astrophysics Data System (ADS)

    Rosato, J.; Kotov, V.; Reiter, D.; Capes, H.; Godbert-Mouret, L.; Koubiti, M.; Marandet, Y.; Stamm, R.

    2010-10-01

    The state of art of the line shape modeling techniques involved in tokamak edge plasma spectroscopy is reported, in the context of the preparation for ITER. Hydrogen spectra are calculated assuming a line-of-sight crossing a 2D-plasma background obtained from numerical simulations. The Doppler, Zeeman and Stark effects are retained. Ion dynamics effects are accounted for by using the numerical simulation method. The possibility for a line shape-based diagnostic of the ITER divertor plasma is examined through fittings of simulated spectra and comparison with the input plasma fields.

  8. Heat flux management via advanced magnetic divertor configurations and divertor detachment

    NASA Astrophysics Data System (ADS)

    Kolemen, E.; Allen, S. L.; Bray, B. D.; Fenstermacher, M. E.; Humphreys, D. A.; Hyatt, A. W.; Lasnier, C. J.; Leonard, A. W.; Makowski, M. A.; McLean, A. G.; Maingi, R.; Nazikian, R.; Petrie, T. W.; Soukhanovskii, V. A.; Unterberg, E. A.

    2015-08-01

    The snowflake divertor (SFD) control and detachment control to manage the heat flux at the divertor are successfully demonstrated at DIII-D. Results of the development and implementation of these two heat flux reduction control methods are presented. The SFD control algorithm calculates the position of the two null-points in real-time and controls shaping coil currents to achieve and stabilize various snowflake configurations. Detachment control stabilizes the detachment front fixed at specified distance between the strike point and the X-point throughout the shot.

  9. Transport analysis of tungsten impurity in ITER

    NASA Astrophysics Data System (ADS)

    Murakami, Y.; Amano, T.; Shimizu, K.; Shimada, M.

    2003-03-01

    The radial distribution of tungsten impurity in ITER is calculated by using the 1.5D transport code TOTAL coupled with NCLASS, which can solve the neo-classical impurity flux considering arbitrary aspect ratio and collisionality. An impurity screening effect is observed when the density profile is flat and the line radiation power is smaller than in the case without impurity transport by a factor of 2. It is shown that 90 MW of line radiation power is possible without significant degradation of plasma performance ( HH98( y,2) ˜1) when the fusion power is 700 MW (fusion gain Q=10). The allowable tungsten density is about 7×10 15/m 3, which is 0.01% of the electron density and the increase of the effective ionic charge Zeff is about 0.39. In this case, the total radiation power is more than half of the total heating power 210 MW, and power to the divertor region is less than 100 MW. This operation regime gives an opportunity for high fusion power operation in ITER with acceptable divertor conditions. Simulations for the case with an internal transport barrier (ITB) are also performed and it is found that impurity shielding by an ITB is possible with density profile control.

  10. 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.

  11. Divertor plasma studies on DIII-D: Experiment and modeling

    SciTech Connect

    West, W.P.; Brooks, N.H.; Allen, S.L.

    1996-09-01

    In a magnetically diverted tokamak, the scrape-off layer (SOL) and divertor plasma provides separation between the first wall and the core plasma, intercepting impurities generated at the wall before they reach the core plasma. The divertor plasma can also serve to spread the heat and particle flux over a large area of divertor structure wall using impurity radiation and neutral charge exchange, thus reducing peak heat and particle fluxes at the divertor strike plate. Such a reduction will be required in the next generation of tokamaks, for without it, the divertor engineering requirements are very demanding. To successfully demonstrate a radiative divertor, a highly radiative condition with significant volume recombination must be achieved in the divertor, while maintaining a low impurity content in the core plasma. Divertor plasma properties are determined by a complex interaction of classical parallel transport, anomalous perpendicular transport, impurity transport and radiation, and plasma wall interaction. In this paper the authors describe a set of experiments on DIII-D designed to provide detailed two dimensional documentation of the divertor and SOL plasma. Measurements have been made in operating modes where the plasma is attached to the divertor strike plate and in highly radiating cases where the plasma is detached from the divertor strike plate. They also discuss the results of experiments designed to influence the distribution of impurities in the plasma using enhanced SOL plasma flow. Extensive modeling efforts will be described which are successfully reproducing attached plasma conditions and are helping to elucidate the important plasma and atomic physics involved in the detachment process.

  12. Plasma flow in the DIII-D divertor

    SciTech Connect

    Boedo, J.A.; Porter, G.D.; Schaffer, M.J.

    1998-07-01

    Indications that flows in the divertor can exhibit complex behavior have been obtained from 2-D modeling but so far remain mostly unconfirmed by experiment. An important feature of flow physics is that of flow reversal. Flow reversal has been predicted analytically and it is expected when the ionization source arising from neutral or impurity ionization in the divertor region is large, creating a high pressure zone. Plasma flows arise to equilibrate the pressure. A radiative divertor regime has been proposed in order to reduce the heat and particle fluxes to the divertor target plates. In this regime, the energy and momentum of the plasma are dissipated into neutral gas introduced in the divertor region, cooling the plasma by collisional, radiative and other atomic processes so that the plasma becomes detached from the target plates. These regimes have been the subject of extensive studies in DIII-D to evaluate their energy and particle transport properties, but only recently it has been proposed that the energy transport over large regions of the divertor must be dominated by convection instead of conduction. It is therefore important to understand the role of the plasma conditions and geometry on determining the region of convection-dominated plasma in order to properly control the heat and particle fluxes to the target plates and hence, divertor performance. The authors have observed complex structures in the deuterium ion flows in the DIII-D divertor. Features observed include reverse flow, convective flow over a large volume of the divertor and stagnant flow. They have measured large gradients in the plasma potential across the separatrix in the divertor and determined that these gradients induce poloidal flows that can potentially affect the particle balance in the divertor.

  13. 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

  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

    Chen, Guang-Hong; Li, Yinsheng

    2015-08-01

    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. 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 of the reconstructed

  15. 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

    SciTech Connect

    Chen, Guang-Hong; Li, Yinsheng

    2015-08-15

    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

  16. 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.

  17. Novel aspects of plasma control in ITER

    DOE PAGES

    Humphreys, David; Ambrosino, G.; de Vries, Peter; ...

    2015-02-12

    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 formore » 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 suppression (TM)), 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. Finally, issues and approaches to fault handling algorithms are described, along with novel aspects of actuator sharing in ITER.« less

  18. Novel aspects of plasma control in ITER

    SciTech Connect

    Humphreys, David; Ambrosino, G.; Felici, Federico; Kim, Sun H.; Jackson, Gary; Kallenbach, A.; Kolemen, Egemen; Lister, J.; Moreau, D.; Pironti, A.; Sauter, O.; Schuster, E.; Snipes, J.; Treutterer, W.; Walker, M.; Welander, A.; Winter, A.; Zabeo, L.

    2015-02-12

    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 suppression (TM)), 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. Finally, issues and approaches to fault handling algorithms are described, along with novel aspects of actuator sharing in ITER.

  19. 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.

  20. 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

  1. Divertor impurity monitor for the International Thermonuclear Experimental Reactor

    NASA Astrophysics Data System (ADS)

    Sugie, T.; Ogawa, H.; Nishitani, T.; Kasai, S.; Katsunuma, J.; Maruo, M.; Ebisawa, K.; Ando, T.; Kita, Y.

    1999-01-01

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

  2. Integrated core-edge tokamak simulations using a novel coordinate system for divertor detachment and heat-load studies

    NASA Astrophysics Data System (ADS)

    Leddy, Jarrod; Dudson, Ben; Romanelli, Michele

    2015-11-01

    Simulating tokamak edge plasmas can often be difficult due to the X-point and divertor region having a different geometry than the rest of the plasma. For edge simulations, a field-aligned coordinate system is normally utilized so that the elongated structures along the field line can be resolved using less grid points while maintaining high resolution perpendicular to the field line. This introduces a singularity at the X-point and constrains the radial coordinate and the poloidal projection of the field-aligned coordinate to be orthogonal. We propose a new coordinate system that relaxes this constraint to allow arbitrary geometries to be matched in the poloidal plane while maintaining a field-aligned coordinate. This is useful at the divertor plate where field lines are not perpendicular to the surface and at the X-point where a close approach is desired. We implement a collisional two-fluid turbulence model using BOUT + + to simulate an isolated divertor leg and investigate the effect of divertor plate angle on detachment and heat loads. We then couple edge simulations in BOUT + + with CENTORI, a core plasma fluid code, to study the evolution of the full plasma with these improved boundary conditions. This work has received funding from the RCUK Energy Programme [grant number EP/I501045].

  3. Electron density measurements in the ITER fusion plasma

    NASA Astrophysics Data System (ADS)

    Watts, Christopher; Udintsev, Victor; Andrew, Philip; Vayakis, George; Van Zeeland, Michael; Brower, David; Feder, Russell; Mukhin, Eugene; Tolstyakov, Sergey

    2013-08-01

    The operation of ITER requires high-quality estimates of the plasma electron density over multiple regions in the plasma for plasma evaluation, plasma control and machine protection purposes. Although the density regimes of ITER are not very different from those of existing tokamaks (1018-1021 m-3), the severe conditions of the fusion plasma environment present particular challenges to implementing these density diagnostics. In this paper we present an overview of the array of ITER electron density diagnostics designed to measure over the entire ITER domain: plasma core, pedestal, edge, scrape-off layer and divertor. It will focus on the challenges faced in making these measurements, and the technical solutions of the current designs.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. High heat flux experiments of saddle type divertor module

    NASA Astrophysics Data System (ADS)

    Suzuki, Satoshi; Akiba, Masato; Araki, Masanori; Satoh, Kazuyoshi; Yokoyama, Kenji; Dairaku, Masayuki

    1994-09-01

    JAERI has been extensively developing plasma facing components for next tokomak devices. The authors have developed a saddle type divertor module which consists of saddle-shaped armor tiles brazed on metal heat sink. This paper presents the experimental and analytical results of thermal cycling experiments of the saddle type divertor module. The divertor module has unidirectional CFC armor tiles brazed on OFHC copper heat sink. A twisted tape was inserted in the cooling tube to enhance the heat transfer. In the experiments, thermal response of the divertor module was monitored by an infrared camera and thermocouples. The maximum incident heat flux was 24.5 MW/m 2 for a duration of 30 s. No degradation of thermal response was observed during the experiment. As a result, the saddle type divertor module successfully endured at an incident heat flux of over 20 MW/m 2 under steady state conditions for 1000 cycles.

  10. Plasma vertical stabilisation in ITER

    NASA Astrophysics Data System (ADS)

    Gribov, Y.; Kavin, A.; Lukash, V.; Khayrutdinov, R.; Huijsmans, G. T. A.; Loarte, A.; Snipes, J. A.; Zabeo, L.

    2015-07-01

    This paper describes the progress in analysis of the ITER plasma vertical stabilisation (VS) system since its design review in 2007-2008. Two indices characterising plasma VS were studied. These are (1) the maximum value of plasma vertical displacement due to free drift that can be stopped by the VS system and (2) the maximum root mean square value of low frequency noise in the dZ/dt measurement signal used in the VS feedback loop. The first VS index was calculated using the PET code for 15 MA plasmas with the nominal position and shape. The second VS index was studied with the DINA code in the most demanding simulations for plasma magnetic control of 15 MA scenarios with the fastest plasma current ramp-up and early X-point formation, the fastest plasma current ramp-down in a divertor configuration, and an H to L mode transition at the current flattop. The studies performed demonstrate that the VS in-vessel coils, adopted recently in the baseline design, significantly increase the range of plasma controllability in comparison with the stabilising systems VS1 and VS2, providing operating margins sufficient to achieve ITER's goals specified in the project requirements. Additionally two sets of the DINA code simulations were performed with the goal of assessment of the capability of the PF system with the VS in-vessel coils: (i) to control the position of runaway electrons generated during disruptions in 15 MA scenarios and (ii) to trigger ELMs in H-mode plasmas of 7.5 MA/2.65 T scenarios planned for the early phase of ITER operation. It was also shown that ferromagnetic structures of the vacuum vessel (ferromagnetic inserts) and test blanket modules insignificantly affect the plasma VS.

  11. Magnetic configuration flexibility of snowflake divertor for HL-2M [Analysis of snowflake divertor configurations for HL-2M

    SciTech Connect

    Zheng, G. Y.; Xu, X. Q.; Ryutov, D. D.; Pan, Y. D.; Xia, T. Y.

    2014-07-09

    HL-2M (Li, 2013 [1]) is a tokamak device that is under construction. Based on the magnetic coils design of HL-2M, four kinds of divertor configurations are calculated by CORSICA code (Pearlstein et al., 2001 [2]) with the same main plasma parameters, which are standard divertor, exact snowflake divertor, snowflake-plus divertor and snowflake-minus divertor configurations. The potential properties of these divertors are analyzed and presented in this paper: low poloidal field area around X-point, connection length from outside mid-plane to the primary X-point, target plate design and magnetic field shear. The results show that the snowflake configurations not only can reduce the heat load at divertor target plates, but also may improve the magneto-hydrodynamic stability by stronger magnetic shear at the edge. Furthermore, a new divertor configuration, named “tripod divertor”, is designed by adjusting the positions of the two X-points according to plasma parameters and magnetic coils current of HL-2M.

  12. Magnetic configuration flexibility of snowflake divertor for HL-2M [Analysis of snowflake divertor configurations for HL-2M

    DOE PAGES

    Zheng, G. Y.; Xu, X. Q.; Ryutov, D. D.; ...

    2014-07-09

    HL-2M (Li, 2013 [1]) is a tokamak device that is under construction. Based on the magnetic coils design of HL-2M, four kinds of divertor configurations are calculated by CORSICA code (Pearlstein et al., 2001 [2]) with the same main plasma parameters, which are standard divertor, exact snowflake divertor, snowflake-plus divertor and snowflake-minus divertor configurations. The potential properties of these divertors are analyzed and presented in this paper: low poloidal field area around X-point, connection length from outside mid-plane to the primary X-point, target plate design and magnetic field shear. The results show that the snowflake configurations not only can reducemore » the heat load at divertor target plates, but also may improve the magneto-hydrodynamic stability by stronger magnetic shear at the edge. Furthermore, a new divertor configuration, named “tripod divertor”, is designed by adjusting the positions of the two X-points according to plasma parameters and magnetic coils current of HL-2M.« less

  13. 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.

  14. Using the Tritium Plasma Experiment to evaluate ITER PFC safety

    SciTech Connect

    Longhurst, G.R.; Anderl, R.A.; Bartlit, J.R.; Causey, R.A.; Haines, J.R.

    1993-07-01

    The Tritium Plasma Experiment was assembled at Sandia National Laboratories, Livermore to investigate interactions between dense plasmas at low energies and plasma-facing component materials. This apparatus has the unique capability of replicating plasma conditions in a tokamak divertor with particle flux densities of 2 {times} 10{sup 19} ions/cm{sup 2} {center_dot} s and a plasma temperature of about 15 eV using a plasma that includes tritium. With the closure of the Tritium Research Laboratory at Livermore, the experiment was moved to the Tritium Systems Test Assembly facility at Los Alamos National Laboratory. An experimental program has been initiated there using the Tritium Plasma Experiment to examine safety issues related to tritium in plasma-facing components, particularly the ITER divertor. Those issues include tritium retention and release characteristics, tritium permeation rates and transient times to coolant streams, surface modification and erosion by the plasma, the effects of thermal loads and cycling, and particulate production. A considerable lack of data exists in these areas for many of the materials, especially beryllium, being considered for use in ITER. Not only will basic material behavior with respect to safety issues in the divertor environment be examined, but innovative techniques for optimizing performance with respect to tritium safety by material modification and process control will be investigated. Supplementary experiments will be carried out at the Idaho National Engineering Laboratory and Sandia National Laboratory to expand and clarify results obtained on the Tritium Plasma Experiment.

  15. Imaging divertor strike point splitting in RMP ELM suppression experiments

    NASA Astrophysics Data System (ADS)

    Moyer, R. A.; Bykov, I.; Orlov, D. M.; Lee, J. S.; Evans, T. E.; Nazikian, R.; Makowski, M.; Lasnier, C. S.; Wang, H.; Abrams, T.; Watkins, J. G.

    2016-10-01

    Fast visible imaging of the lower divertor has been implemented at DIII-D to study the structure and dynamics of lobes induced by 3D fields in RMP ELM suppression experiments. The sharpest imaging was obtained with spatially localized molecular D2 emission indicative of the D flux to the surface. Multiple D2 emission peaks are readily resolved during RMPs, in contrast to the heat flux profile (from IR), which often shows little structure. The brightest D2 lobe is often farthest from the primary inner strike point (ISP). Mitigated ELMs perturb the position and intensity of the ISP lobes and spread the outer strike point emission into the far SOL, where it may be caused by ELM filament propagation. RMP current ramps affect the lobe locations and separations. Implications of the lobe dynamics for plasma response is being studied. Work supported by U.S. DOE under Grants DE-FG02-07ER54917 and DE-FG02-05ER54809, and Contracts DE-FC02-04ER54698, DE-AC52-07NA27344, DE-AC04-94AL85000, DE-AC05-06OR23100 and DE-AC02-09CH11466.

  16. Overview of erosion-deposition diagnostic tools for the ITER-Like Wall in the JET tokamak

    NASA Astrophysics Data System (ADS)

    Rubel, M.; Coad, J. P.; Widdowson, A.; Matthews, G. F.; Esser, H. G.; Hirai, T.; Likonen, J.; Linke, J.; Lungu, C. P.; Mayer, M.; Pedrick, L.; Ruset, C.; JET-EFDA Contributors

    2013-07-01

    This paper presents scientific and technical issues related to the development of erosion-deposition diagnostic tools for JET operated with the ITER-Like Wall: beryllium and tungsten marker tiles and several types of wall probes installed in the main chamber and in the divertor. Markers tiles are the standard limiter and divertor components additionally coated first with a thin sandwich of Ni-Be and Mo-W for, beryllium and tungsten markers, respectively. Both types of markers are embedded in regular arrays of limiter and divertor tiles. Coated W-Be probes are also inserted in the Be-covered Inconel cladding tiles on the central column. Other types of erosion-deposition diagnostic tools are: rotating collectors, deposition traps, louver clips, quartz microbalance and mirrors for the First Mirror Test at JET for ITER. The specific role of these tools is discussed in detail.

  17. Extinguishing ELMs in detached radiative divertor plasmas

    NASA Astrophysics Data System (ADS)

    Pigarov, Alexander; Krasheninnikov, Sergei; Rognlien, Thomas

    2016-10-01

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

  18. 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.

  19. Guidance of the divertor channel outside the main coil system for heliotron/torsatron devices

    NASA Astrophysics Data System (ADS)

    Takase, H.; Ohyabu, N.

    1995-02-01

    A divertor magnetic configuration is proposed that significantly reduces heat load on the divertor plates in heliotron/torsatron devices. The proposed configuration utilizes an octupole field for guiding the divertor channels to a remote area outside the main coil system, where the magnetic field is weak. This allows a significant reduction of the heat load due to expansion of the divertor channels as well as substantially easier access to the divertor plates for maintenance, the key requirements for toroidal fusion reactor designs

  20. Snowflake divertor experiments in the DIII-D, NSTX, and NSTX-U tokamaks aimed at the development of the divertor power exhaust solution

    DOE PAGES

    Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; ...

    2016-11-16

    Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in the future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment was performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large Edge Localized Modes (ELMs). However, a stable partial detachment ofmore » the outer strike point was also achieved where inter-ELM peak heat flux was reduced by factors 3-5, and peak ELM heat flux was reduced by up to 80% (see standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower ne, and ELM energy and divertor peak heat flux reduction, especially prominent in radiative D2-seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D2-seeded SF divertor at PSOL = 3 - 4 MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multifluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected PSOL ≃ 9 MW case. Furthermore, the radiative SF divertor with carbon impurity provides a wider ne operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar qpeak reduction factors (see standard divertor).« less

  1. Snowflake divertor experiments in the DIII-D, NSTX, and NSTX-U tokamaks aimed at the development of the divertor power exhaust solution

    SciTech Connect

    Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; Lasnier, C. J.; Makowski, M. A.; McLean, A. G.; Meier, E. T.; Meyer, W. H.; Rognlien, T. D.; Ryutov, D. D.; Scotti, F.; Kolemen, E.; Bell, R. E.; Diallo, A.; Gerhardt, S.; Kaita, R.; Kaye, S.; LeBlanc, B. P.; Maingi, R.; Menard, J. E.; Podesta, M.; Roquemore, A. L.; Groebner, R. J.; Hyatt, A. W.; Leonard, A. W.; Osborne, T. H.; Petrie, T. W.; Ahn, J. -W.; Raman, R.; Watkins, J. G.

    2016-11-16

    Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in the future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment was performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large Edge Localized Modes (ELMs). However, a stable partial detachment of the outer strike point was also achieved where inter-ELM peak heat flux was reduced by factors 3-5, and peak ELM heat flux was reduced by up to 80% (see standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower ne, and ELM energy and divertor peak heat flux reduction, especially prominent in radiative D2-seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D2-seeded SF divertor at PSOL = 3 - 4 MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multifluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected PSOL ≃ 9 MW case. Furthermore, the radiative SF divertor with carbon impurity provides a wider ne operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar qpeak reduction factors (see standard divertor).

  2. 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.

  3. A super-cusp divertor configuration for tokamaks

    DOE PAGES

    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

  4. 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.

  5. Modeling of extinguishing ELMs in detached divertor plasmas

    NASA Astrophysics Data System (ADS)

    Pigarov, A.; Krasheninnikov, S.; Hollmann, E.; Rognlien, T.

    2015-11-01

    Detached plasmas, the primary operational regime for divertors in next-step fusion devices, should be compatible with both good H-mode confinement and relatively small ELMs providing tolerable heat power loads on divertor targets. Here, dynamics of boundary plasma, impurities and material walls over a sequence of many type-I ELM events under detached divertor plasma conditions is studied with UEGDE-MB-W, the newest version of 2D edge plasma transport code, which incorporates Macro-Blob (MB) approach to simulate non-diffusive filamentary transport and various ``Wall'' (W) models for time-dependent hydrogen wall inventory and recycling. We present the results of multi-parametric analysis on the impact of the size and frequency of ELMs on the divertor plasma parameters where we vary the MB characteristics under different pedestals and divertor configurations. We discuss the conditions, under which small but frequent type-I ELMs (typical for high-power H-mode discharges on current tokamaks with hard deuterium gas puff) are not ``burning through'' the formed detached divertor plasma. In this case, the inner and outer divertors are filled by sub-eV, recombining, highly-impure plasma. Variations of impurity plasma content, radiation pattern, and deuterium wall inventory over the ELM cycle are analyzed. UEDGE-MB-W modeling results are compared to available experimental data.

  6. Divertor Optimization via Control at DIII-D

    NASA Astrophysics Data System (ADS)

    Kolemen, E.; Allen, S. L.; Makowski, M. A.; Soukhanovskii, V. A.; Bray, B. D.; Humphreys, D. A.; Johnson, R.; Leonard, A. W.; Liu, C.; Penaflor, B. G.; Petrie, T. W.; Eldon, D.; McLean, A. G.; Unterberg, E. A.

    2014-10-01

    DIII-D divertor performance and heat-handling capabilities are optimized using advanced control techniques. The world's first real-time snowflake divertor detection and control system was implemented on DIII-D in order to stabilize and optimize this configuration. A new control system was implemented to regulate and study detachment and radiation, since future fusion reactors will require detached or partially detached plasmas to achieve acceptable divertor target heat fluxes. The algorithm regulates the D2 and impurity gas injection level by using the divertor temperature measurements from real-time Thomson diagnostics to compute the detachment level, and the real-time bolometer diagnostics to determine core and divertor radiation. This control allows the optimization of the detachment and radiation from the core and the divertor to achieve high core performance compatible with reduced heat-flux to the divertor. Work supported by the US DOE under DE-AC02-09CH11466, DE-AC52-07NA27344, DE-FC02-04ER54698 and DE-AC05-00OR22725.

  7. A super-cusp divertor configuration for tokamaks

    NASA Astrophysics Data System (ADS)

    Ryutov, D. D.

    2015-10-01

    > This 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 indeed 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. 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.

  8. 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.

  9. Simulations of NSTX with a Liquid Lithium Divertor Module

    SciTech Connect

    D. P. Stotler, R. Maingi, H.W. Kugel, A. Yu. Pigarov, T.D. Rognlien, V.A. Soukhanovskii

    2008-07-08

    The UEDGE edge plasma transport code is used to model the effect of the reduced recycling provided by the Liquid Lithium Divertor (LLD) module that will be installed in NSTX. UEDGE's transport coefficients are calibrated against an existing NSTX shot using midplane and divertor diagnostic data. The LLD is then incorporated into the simulations as a reduction in the recycling coefficient over the outer divertor. Heat transfer calculations performed using the resulting heat flux profiles indicate that lithium evaporation will be negligible for pulse lengths < 2 s at low (~ 2 MW) input power. At high input power (~ 7 MW), the pulse length may have to be restricted.

  10. Simulations of NSTX with a Liquid Lithium Divertor Module

    SciTech Connect

    Stotler, D. P.; Maingi, R.; Zakharov, L. E.; Kugel, H. W.; Pigarov, A. Yu.; Rognlien, T. D.; Soukhanovskii, V. A.

    2010-02-18

    A strategy to develop self-consistent simulations of the behavior of lithium in the Liquid Lithium Divertor (LLD) module to be installed in NSTX is described. In this initial stage of the plan, the UEDGE edge plasma transport code is used to simulate an existing NSTX shot, with UEDGE's transport coefficients set using midplane and divertor diagnostic data. The LLD is incorporated into the simulations as a reduction in the recycling coefficient over the outer divertor. Heat transfer calculations performed using the resulting heat flux profiles provide preliminary estimates on operating limits for the LLD as well as input data for subsequent steps in the LLD modeling effort.

  11. Two-chamber model for divertors with plasma recycling

    SciTech Connect

    Langer, W.D.; Singer, C.E.

    1984-11-01

    To model particle and heat loss terms at the edge of a tokamak with a divertor or pumped limiter, a simple two-chamber formulation of the scrapeoff has been constructed by integrating the fluid equations, including sources, along open field lines. The model is then solved for a wide range of density and temperature conditions in the scrapeoff, using geometrical parameters typical of the PDX poloidal divertor. The solutions characterize four divertor operating conditions for beam-heated plasmas: plugged, unplugged, blowthrough, and blowback.

  12. Disruption characteristics in PDX with limiter and divertor discharges

    SciTech Connect

    Couture, P.; McGuire, K.

    1986-09-01

    A comparison has been made between the characteristics of disruptions with limiter and divertor configurations in PDX. A large data base on disruptions has been collected over four years of machine operation, and a total of 15,000 discharges are contained in the data file. It was found that divertor discharges have less disruptions during ramp up and flattop of the plasma current. However, for divertor discharges a large number of fast, low current disruptions take place during the current ramp down. These disruptions are probably caused by the deformation of the plasma shape.

  13. Evaluation of cooling concepts and specimen geometries for high heat flux tests on neutron irradiated divertor elements

    SciTech Connect

    Linke, J.; Bolt. H.; Breitbach, G.

    1994-12-31

    To assess the lifetime and the long term heat removal capabilities of plasma facing components in future thermonuclear fusion reactors such as ITER, neutron irradiation and subsequent high heat flux tests will be most essential. The effect of neutron damage will be simulated in material test reactors (such as the HFR-Petten) in a fission neutron environment. To investigate the heat loads during normal and off-normal operation scenarios a 60 kW electron beam test stand (Juelich Divertor Test Facility in Hot Cells, JUDITH) has been installed in a hot cell which can be operated by remote handling techniques. In this facility inertially cooled test coupons can be handled as well as small actively cooled divertor mock-ups. A special clamping mechanism for small test coupons (25 mm x 25 mm x 35 mm) with an integrated coolant channel within a copper or TZM heat sink has been developed and tested in an electron beam test bed. This method is an attractive alternative to costly large scale tests on complete divertor modules. The temperature and stress fields in individual CFC or beryllium tiles brazed to metallic heat sink (e.g. copper or TZM) can be investigated before and after neutron irradiation with moderate efforts.

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

    NASA Astrophysics Data System (ADS)

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

    2016-02-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2012-07-01

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

  16. The effect of resonant magnetic perturbations on the divertor heat and particle fluxes in MAST

    NASA Astrophysics Data System (ADS)

    Thornton, A. J.; Kirk, A.; Cahyna, P.; Chapman, I. T.; Harrison, J. R.; Liu, Yueqiang; the MAST Team

    2014-06-01

    Edge localized modes (ELMs) are a concern for future devices, such as ITER, due to the large transient heat loads they generate on the divertor surfaces which could limit the operational lifetime of the device. This paper discusses the application of resonant magnetic perturbations (RMPs) as a mechanism for ELM control on Mega Amp Spherical Tokamak (MAST). Experiments have been performed using an n = 3 toroidal mode number perturbation and measurements of the strike point splitting performed. The measurements have been made using both infrared and visible imaging to measure the heat and particle flux to the divertor. The measured profiles have shown clear splitting in L-mode which compares well with the predication of the splitting location from modelling including the effect of screening. The splitting of the strike point has also been studied as a function of time during the ELM. The splitting varies during the ELM, being the strongest at the time of the peak heat flux and becoming more filamentary at the end of the ELM (200 µs after the peak midplane Dα emission). Variation in the splitting profiles has also been seen, with some ELMs showing clear splitting and others no splitting. A possible explanation of this effect is proposed, and supported by modelling, which concerns the relative phase between the RMP field and the ELM filament location.

  17. Design and thermal-hydraulic analysis of tokamak divertor armor tiles

    SciTech Connect

    Sharpe, J.P.; Carter, T.A.; Bourham, M.A.; Gilligan, J.G.

    1995-12-31

    A prototype divertor armor tile design has been investigated using water-cooled ATJ graphite tiles fitted to a copper heat sink. Two-dimensional steady-state and 1-D time dependent heat transfer codes were developed to determine thermal design characteristics. A steady-state heat flux of 5 MW/m{sup 2} and a transient disruption load of 140 MJ/m{sup 2} over 100 {micro}s were assumed for an ITER-type device operating in a radiative divertor configuration. For a tile fitted to the heat sink by a bonded-pin mechanism, the optimal armor thickness was determined to be 1.0 cm, with a 2.2 cm diameter coolant channel. The maximum steady state and disruption temperatures of the tile were determined to be 1,760 K and 4,800 K, respectively. LOCA analysis yielded that a 7 second response time would be needed after loss-of-coolant in the armor tile. The design is predicted to survive approximately 6 disruptions before tile replacement would be necessary.

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

    NASA Astrophysics Data System (ADS)

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

    2017-01-01

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

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

    SciTech Connect

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

    2017-01-01

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

  20. 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

  1. Nuclear modules of ITER tokamak systems code

    SciTech Connect

    Gohar, Y.; Baker, C.; Brooks, J.; Finn, P.; Hassanein, A.; Willms, S.; Barr, W.; Bushigin, A.; Kalyanam, K.M.; Haines, J.

    1987-10-01

    Nuclear modules were developed to model various reactor components in the ITER systems code. Several design options and cost algorithms are included for each component. The first wall, blanket and shield modules calculate the beryllium zone thickness, the disruptions results, the nuclear responses in different components including the toroidal field coils. Tungsten shield/water coolant/steel structure and steel shield/water coolant are the shield options for the inboard and outboard sections of the reactor. Lithium nitrate dissolved in the water coolant with a variable beryllium zone thickness in the outboard section of the reactor provides the tritium breeding capability. The reactor vault module defines the thickness of the reactor wall and the roof based on the dose equivalent during operation including skyshine contribution. The impurity control module provides the design parameters for the divertor including plate design, heat load, erosion rate, tritium permeation through the plate material to the coolant, plasma contamination by sputtered impurities, and plate lifetime. Several materials: Be, C, V, Mo, and W can be used for the divertor plate to cover a range of plasma edge temperatures. The tritium module calculates tritium and deuterium flow rates for the reactor plant. The tritium inventory in the fuelers, neutral beams, vacuum pumps, impurity control, first wall, and blanket is calculated. Tritium requirements are provided for different operating conditions. The nuclear models are summarized in this paper including the different design options and key analyses of each module. 39 refs., 3 tabs.

  2. 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

  3. Beryllium accumulation at the inner divertor of JET

    NASA Astrophysics Data System (ADS)

    Likonen, J.; Vainonen-Ahlgren, E.; Coad, J. P.; Zilliacus, R.; Renvall, T.; Hole, D. E.; Rubel, M.; Arstila, K.; Matthews, G. F.; Stamp, M.; JET-EFDA Contributors

    2005-03-01

    MkIIGB divertor tiles exposed in JET for the 1998-2001 and 1999-2001 campaigns have been used to assess the amount of beryllium and carbon deposited at the inner divertor wall. Total amount of Be at the inner divertor tiles was determined and integrated toroidally. Results were compared with data obtained with optical spectroscopy and good agreement was obtained. The amount of deposited C was computed from the amount of deposited Be assuming that the Be/C ratio arriving in the divertor is the same as the Be/C ratio in the main chamber. On the basis of this analysis we would expect there to be ˜0.4 kg of C deposited. This gives an average C deposition rate lower than during the MkIIA phase.

  4. 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.

  5. Carbon fiber composites application in ITER plasma facing components

    NASA Astrophysics Data System (ADS)

    Barabash, V.; Akiba, M.; Bonal, J. P.; Federici, G.; Matera, R.; Nakamura, K.; Pacher, H. D.; Rödig, M.; Vieider, G.; Wu, C. H.

    1998-10-01

    Carbon Fiber Composites (CFCs) are one of the candidate armour materials for the plasma facing components of the International Thermonuclear Experimental Reactor (ITER). For the present reference design, CFC has been selected as armour for the divertor target near the plasma strike point mainly because of unique resistance to high normal and off-normal heat loads. It does not melt under disruptions and might have higher erosion lifetime in comparison with other possible armour materials. Issues related to CFC application in ITER are described in this paper. They include erosion lifetime, tritium codeposition with eroded material and possible methods for the removal of the codeposited layers, neutron irradiation effect, development of joining technologies with heat sink materials, and thermomechanical performance. The status of the development of new advanced CFCs for ITER application is also described. Finally, the remaining R&D needs are critically discussed.

  6. Evaluation of helium cooling for fusion divertors

    SciTech Connect

    Baxi, C.B.

    1993-09-01

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

  7. Diagnostics for the DIII-D radiative divertor

    SciTech Connect

    Nilson, D.G.; Brooks, N.H.; Smith, J.P.; Snider, R.T.

    1995-10-01

    This paper reviews the design of new diagnostics and the modifications to existing diagnostics needed to carry out radiative divertor experiments in DIII-D following installation in late 1996 of a set of baffle structures that will restrict the backflow to the core plasma of neutral deuterium atoms and impurity gases. The divertor slots formed by the new baffle structures will inhibit the easy view of the divertor legs and target plates that the open divertor geometry in DIII-D currently affords. We review a basic set of diagnostics that are needed to demonstrate the reduction of divertor heat loading and radiative dissipation of energy within the divertor. This will include IR cameras, bolometry, foil bolometers, and Langmuir probes. Within the limits of available funding, we will implement a supplemental set of instruments which provide a more detailed understanding of the underlying physical processes. Many existing diagnostics require only re-aiming to provide proper coverage of the initial 23 cm long divertor plasma configuration (X- point to floor distance). Other diagnostics need extensive reconfiguration using in-vessel fiber-optic bundles or high power laser mirrors. The new divertor baffle panels provide a protective shelf for diagnostic hardware mounted underneath them, but the water cooling channels in the panels limit the permissible size of through holes and, thereby, restrict the available views of under-the- baffle diagnostics. The successful resolution of the design and implementation of these diagnostic modifications is dependent on a strong coordination between GA and its many diagnostic collaborators.

  8. Impact of divertor geometry on H-mode confinement in the JET metallic wall

    NASA Astrophysics Data System (ADS)

    Joffrin, E.; Tamain, P.; Belonohy, E.; Bufferand, H.; Buratti, P.; Challis, C. D.; Delabie, E.; Drewelow, P.; Dodt, D.; Frassinetti, L.; Garcia, J.; Giroud, C.; Groth, M.; Hobirk, J.; Jarvinen, A. E.; Kim, H.-T.; Koechl, F.; Kruezi, U.; Lipschutz, B.; Lomas, P. J.; de la Luna, E.; Loarer, T.; Maget, P.; Maggi, C.; Matthews, G.; Maviglia, F.; Meigs, A.; Nunes, I.; Pucella, G.; Rimini, F.; Saarelma, S.; Solano, E.; Sips, A. C. C.; Tsalas, M.; Voitsekhovitch, I.; Weisen, H.; JET Contributors, the

    2017-08-01

    Recent experiments with the ITER-like wall have demonstrated that changes in divertor strike point position are correlated with strong modification of the global energy confinement. The impact on energy confinement is observable both on the pedestal confinement and core normalised gradients. The corner configuration shows an increased core density gradient length and ion pressure indicating a better ion confinement. The study of neutral re-circulation indicates the neutral pressure in the main chamber varies inversely with the energy confinement and a correlation between the pedestal total pressure and the neutral pressure in the main chamber can be established. It does not appear that charge exchange losses nor momentum losses could explain this effect, but it may be that changes in edge electric potential are playing a role at the plasma edge. This study emphasizes the importance of the scrape-off layer (SOL) conditions on the pedestal and core confinement.

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

    NASA Astrophysics Data System (ADS)

    Arnoux, G.; Coenen, J.; Bazylev, B.; Corre, Y.; Matthews, G. F.; Balboa, I.; Clever, M.; Dejarnac, R.; Devaux, S.; Eich, T.; Gauthier, E.; Frassinetti, L.; Horacek, J.; Jachmich, S.; Kinna, D.; Marsen, S.; Mertens, Ph.; Pitts, R. A.; Rack, M.; Sergienko, G.; Sieglin, B.; Stamp, M.; Thompson, V.

    2015-08-01

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

  10. 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.

  11. Control of particle and power exhaust in pellet fuelled ITER DT scenarios employing integrated models

    NASA Astrophysics Data System (ADS)

    Wiesen, S.; Köchl, F.; Belo, P.; Kotov, V.; Loarte, A.; Parail, V.; Corrigan, G.; Garzotti, L.; Harting, D.

    2017-07-01

    The integrated model JINTRAC is employed to assess the dynamic density evolution of the ITER baseline scenario when fuelled by discrete pellets. The consequences on the core confinement properties, α-particle heating due to fusion and the effect on the ITER divertor operation, taking into account the material limitations on the target heat loads, are discussed within the integrated model. Using the model one can observe that stable but cyclical operational regimes can be achieved for a pellet-fuelled ITER ELMy H-mode scenario with Q  =  10 maintaining partially detached conditions in the divertor. It is shown that the level of divertor detachment is inversely correlated with the core plasma density due to α-particle heating, and thus depends on the density evolution cycle imposed by pellet ablations. The power crossing the separatrix to be dissipated depends on the enhancement of the transport in the pedestal region being linked with the pressure gradient evolution after pellet injection. The fuelling efficacy of the deposited pellet material is strongly dependent on the E  ×  B plasmoid drift. It is concluded that integrated models like JINTRAC, if validated and supported by realistic physics constraints, may help to establish suitable control schemes of particle and power exhaust in burning ITER DT-plasma scenarios.

  12. Overview of the JET ITER-like Wall, First Results and Scientific Programme

    NASA Astrophysics Data System (ADS)

    Matthews, Guy; JET-EFDA Collaboration

    2011-10-01

    The ITER-like Wall (ILW) is the first integrated tokamak experiment with a beryllium main chamber wall and tungsten divertor as foreseen for the activated operational phase of ITER: The ILW will study plasma-wall interaction (PWI) processes (material erosion, material mixing etc.), and the compatibility of the ITER materials with low fuel retention and high power operation. Replacement of the JET CFC first wall by solid Be limiters, and a combination of bulk W and W-coated CFC divertor tiles was performed by remote handling and completed in May 2011 in parallel with a neutral beam heating upgrade to 35 MW and enhancement of diagnostic capabilities. Mitigation of the power and energy loads in the divertor to acceptable levels at high power plasma performance will require high-density plasmas and radiative cooling via impurity seeding. Experiments were carried out with the carbon wall in preparation for the ILW to operate plasmas within ILW limits and provide reference plasmas for key physics studies. Although first plasma is scheduled for mid-August, the scientific programme in support of ITER will start earlier with machine conditioning. In this paper, an overview of the ILW, first results and the outlook for the scientific programme will be presented.

  13. Overview of the JET results in support to ITER

    NASA Astrophysics Data System (ADS)

    Litaudon, X.; Abduallev, S.; Abhangi, M.; Abreu, P.; Afzal, M.; Aggarwal, K. M.; Ahlgren, T.; Ahn, J. H.; Aho-Mantila, L.; Aiba, N.; Airila, M.; Albanese, R.; Aldred, V.; Alegre, D.; Alessi, E.; Aleynikov, P.; Alfier, A.; Alkseev, A.; Allinson, M.; Alper, B.; Alves, E.; Ambrosino, G.; Ambrosino, R.; Amicucci, L.; Amosov, V.; Andersson Sundén, E.; Angelone, M.; Anghel, M.; Angioni, C.; Appel, L.; Appelbee, C.; Arena, P.; Ariola, M.; Arnichand, H.; Arshad, S.; Ash, A.; Ashikawa, N.; Aslanyan, V.; Asunta, O.; Auriemma, F.; Austin, Y.; Avotina, L.; Axton, M. D.; Ayres, C.; Bacharis, M.; Baciero, A.; Baião, D.; Bailey, S.; Baker, A.; Balboa, I.; Balden, M.; Balshaw, N.; Bament, R.; Banks, J. W.; Baranov, Y. F.; Barnard, M. A.; Barnes, D.; Barnes, M.; Barnsley, R.; Baron Wiechec, A.; Barrera Orte, L.; Baruzzo, M.; Basiuk, V.; Bassan, M.; Bastow, R.; Batista, A.; Batistoni, P.; Baughan, R.; Bauvir, B.; Baylor, L.; Bazylev, B.; Beal, J.; Beaumont, P. S.; Beckers, M.; Beckett, B.; Becoulet, A.; Bekris, N.; Beldishevski, M.; Bell, K.; Belli, F.; Bellinger, M.; Belonohy, É.; Ben Ayed, N.; Benterman, N. A.; Bergsåker, H.; Bernardo, J.; Bernert, M.; Berry, M.; Bertalot, L.; Besliu, C.; Beurskens, M.; Bieg, B.; Bielecki, J.; Biewer, T.; Bigi, M.; Bílková, P.; Binda, F.; Bisoffi, A.; Bizarro, J. P. S.; Björkas, C.; Blackburn, J.; Blackman, K.; Blackman, T. R.; Blanchard, P.; Blatchford, P.; Bobkov, V.; Boboc, A.; Bodnár, G.; Bogar, O.; Bolshakova, I.; Bolzonella, T.; Bonanomi, N.; Bonelli, F.; Boom, J.; Booth, J.; Borba, D.; Borodin, D.; Borodkina, I.; Botrugno, A.; Bottereau, C.; Boulting, P.; Bourdelle, C.; Bowden, M.; Bower, C.; Bowman, C.; Boyce, T.; Boyd, C.; Boyer, H. J.; Bradshaw, J. M. A.; Braic, V.; Bravanec, R.; Breizman, B.; Bremond, S.; Brennan, P. D.; Breton, S.; Brett, A.; Brezinsek, S.; Bright, M. D. J.; Brix, M.; Broeckx, W.; Brombin, M.; Brosławski, A.; Brown, D. P. D.; Brown, M.; Bruno, E.; Bucalossi, J.; Buch, J.; Buchanan, J.; Buckley, M. A.; Budny, R.; Bufferand, H.; Bulman, M.; Bulmer, N.; Bunting, P.; Buratti, P.; Burckhart, A.; Buscarino, A.; Busse, A.; Butler, N. K.; Bykov, I.; Byrne, J.; Cahyna, P.; Calabrò, G.; Calvo, I.; Camenen, Y.; Camp, P.; Campling, D. C.; Cane, J.; Cannas, B.; Capel, A. J.; Card, P. J.; Cardinali, A.; Carman, P.; Carr, M.; Carralero, D.; Carraro, L.; Carvalho, B. B.; Carvalho, I.; Carvalho, P.; Casson, F. J.; Castaldo, C.; Catarino, N.; Caumont, J.; Causa, F.; Cavazzana, R.; Cave-Ayland, K.; Cavinato, M.; Cecconello, M.; Ceccuzzi, S.; Cecil, E.; Cenedese, A.; Cesario, R.; Challis, C. D.; Chandler, M.; Chandra, D.; Chang, C. S.; Chankin, A.; Chapman, I. T.; Chapman, S. C.; Chernyshova, M.; Chitarin, G.; Ciraolo, G.; Ciric, D.; Citrin, J.; Clairet, F.; Clark, E.; Clark, M.; Clarkson, R.; Clatworthy, D.; Clements, C.; Cleverly, M.; Coad, J. P.; Coates, P. A.; Cobalt, A.; Coccorese, V.; Cocilovo, V.; Coda, S.; Coelho, R.; Coenen, J. W.; Coffey, I.; Colas, L.; Collins, S.; Conka, D.; Conroy, S.; Conway, N.; Coombs, D.; Cooper, D.; Cooper, S. R.; Corradino, C.; Corre, Y.; Corrigan, G.; Cortes, S.; Coster, D.; Couchman, A. S.; Cox, M. P.; Craciunescu, T.; Cramp, S.; Craven, R.; Crisanti, F.; Croci, G.; Croft, D.; Crombé, K.; Crowe, R.; Cruz, N.; Cseh, G.; Cufar, A.; Cullen, A.; Curuia, M.; Czarnecka, A.; Dabirikhah, H.; Dalgliesh, P.; Dalley, S.; Dankowski, J.; Darrow, D.; Davies, O.; Davis, W.; Day, C.; Day, I. E.; De Bock, M.; de Castro, A.; de la Cal, E.; de la Luna, E.; De Masi, G.; de Pablos, J. L.; De Temmerman, G.; De Tommasi, G.; de Vries, P.; Deakin, K.; Deane, J.; Degli Agostini, F.; Dejarnac, R.; Delabie, E.; den Harder, N.; Dendy, R. O.; Denis, J.; Denner, P.; Devaux, S.; Devynck, P.; Di Maio, F.; Di Siena, A.; Di Troia, C.; Dinca, P.; D'Inca, R.; Ding, B.; Dittmar, T.; Doerk, H.; Doerner, R. P.; Donné, T.; Dorling, S. E.; Dormido-Canto, S.; Doswon, S.; Douai, D.; Doyle, P. T.; Drenik, A.; Drewelow, P.; Drews, P.; Duckworth, Ph.; Dumont, R.; Dumortier, P.; Dunai, D.; Dunne, M.; Ďuran, I.; Durodié, F.; Dutta, P.; Duval, B. P.; Dux, R.; Dylst, K.; Dzysiuk, N.; Edappala, P. V.; Edmond, J.; Edwards, A. M.; Edwards, J.; Eich, Th.; Ekedahl, A.; El-Jorf, R.; Elsmore, C. G.; Enachescu, M.; Ericsson, G.; Eriksson, F.; Eriksson, J.; Eriksson, L. G.; Esposito, B.; Esquembri, S.; Esser, H. G.; Esteve, D.; Evans, B.; Evans, G. E.; Evison, G.; Ewart, G. D.; Fagan, D.; Faitsch, M.; Falie, D.; Fanni, A.; Fasoli, A.; Faustin, J. M.; Fawlk, N.; Fazendeiro, L.; Fedorczak, N.; Felton, R. C.; Fenton, K.; Fernades, A.; Fernandes, H.; Ferreira, J.; Fessey, J. A.; Février, O.; Ficker, O.; Field, A.; Fietz, S.; Figueiredo, A.; Figueiredo, J.; Fil, A.; Finburg, P.; Firdaouss, M.; Fischer, U.; Fittill, L.; Fitzgerald, M.; Flammini, D.; Flanagan, J.; Fleming, C.; Flinders, K.; Fonnesu, N.; Fontdecaba, J. M.; Formisano, A.; Forsythe, L.; Fortuna, L.; Fortuna-Zalesna, E.; Fortune, M.; Foster, S.; Franke, T.; Franklin, T.; Frasca, M.; Frassinetti, L.; Freisinger, M.; Fresa, R.; Frigione, D.; Fuchs, V.; Fuller, D.; Futatani, S.; Fyvie, J.; Gál, K.; Galassi, D.; Gałązka, K.; Galdon-Quiroga, J.; Gallagher, J.; Gallart, D.; Galvão, R.; Gao, X.; Gao, Y.; Garcia, J.; Garcia-Carrasco, A.; García-Muñoz, M.; Gardarein, J.-L.; Garzotti, L.; Gaudio, P.; Gauthier, E.; Gear, D. F.; Gee, S. J.; Geiger, B.; Gelfusa, M.; Gerasimov, S.; Gervasini, G.; Gethins, M.; Ghani, Z.; Ghate, M.; Gherendi, M.; Giacalone, J. C.; Giacomelli, L.; Gibson, C. S.; Giegerich, T.; Gil, C.; Gil, L.; Gilligan, S.; Gin, D.; Giovannozzi, E.; Girardo, J. B.; Giroud, C.; Giruzzi, G.; Glöggler, S.; Godwin, J.; Goff, J.; Gohil, P.; Goloborod'ko, V.; Gomes, R.; Gonçalves, B.; Goniche, M.; Goodliffe, M.; Goodyear, A.; Gorini, G.; Gosk, M.; Goulding, R.; Goussarov, A.; Gowland, R.; Graham, B.; Graham, M. E.; Graves, J. P.; Grazier, N.; Grazier, P.; Green, N. R.; Greuner, H.; Grierson, B.; Griph, F. S.; Grisolia, C.; Grist, D.; Groth, M.; Grove, R.; Grundy, C. N.; Grzonka, J.; Guard, D.; Guérard, C.; Guillemaut, C.; Guirlet, R.; Gurl, C.; Utoh, H. H.; Hackett, L. J.; Hacquin, S.; Hagar, A.; Hager, R.; Hakola, A.; Halitovs, M.; Hall, S. J.; Hallworth Cook, S. P.; Hamlyn-Harris, C.; Hammond, K.; Harrington, C.; Harrison, J.; Harting, D.; Hasenbeck, F.; Hatano, Y.; Hatch, D. R.; Haupt, T. D. V.; Hawes, J.; Hawkes, N. C.; Hawkins, J.; Hawkins, P.; Haydon, P. W.; Hayter, N.; Hazel, S.; Heesterman, P. J. L.; Heinola, K.; Hellesen, C.; Hellsten, T.; Helou, W.; Hemming, O. N.; Hender, T. C.; Henderson, M.; Henderson, S. S.; Henriques, R.; Hepple, D.; Hermon, G.; Hertout, P.; Hidalgo, C.; Highcock, E. G.; Hill, M.; Hillairet, J.; Hillesheim, J.; Hillis, D.; Hizanidis, K.; Hjalmarsson, A.; Hobirk, J.; Hodille, E.; Hogben, C. H. A.; Hogeweij, G. M. D.; Hollingsworth, A.; Hollis, S.; Homfray, D. A.; Horáček, J.; Hornung, G.; Horton, A. R.; Horton, L. D.; Horvath, L.; Hotchin, S. P.; Hough, M. R.; Howarth, P. J.; Hubbard, A.; Huber, A.; Huber, V.; Huddleston, T. M.; Hughes, M.; Huijsmans, G. T. A.; Hunter, C. L.; Huynh, P.; Hynes, A. M.; Iglesias, D.; Imazawa, N.; Imbeaux, F.; Imríšek, M.; Incelli, M.; Innocente, P.; Irishkin, M.; Ivanova-Stanik, I.; Jachmich, S.; Jacobsen, A. S.; Jacquet, P.; Jansons, J.; Jardin, A.; Järvinen, A.; Jaulmes, F.; Jednoróg, S.; Jenkins, I.; Jeong, C.; Jepu, I.; Joffrin, E.; Johnson, R.; Johnson, T.; Johnston, Jane; Joita, L.; Jones, G.; Jones, T. T. C.; Hoshino, K. K.; Kallenbach, A.; Kamiya, K.; Kaniewski, J.; Kantor, A.; Kappatou, A.; Karhunen, J.; Karkinsky, D.; Karnowska, I.; Kaufman, M.; Kaveney, G.; Kazakov, Y.; Kazantzidis, V.; Keeling, D. L.; Keenan, T.; Keep, J.; Kempenaars, M.; Kennedy, C.; Kenny, D.; Kent, J.; Kent, O. N.; Khilkevich, E.; Kim, H. T.; Kim, H. S.; Kinch, A.; king, C.; King, D.; King, R. F.; Kinna, D. J.; Kiptily, V.; Kirk, A.; Kirov, K.; Kirschner, A.; Kizane, G.; Klepper, C.; Klix, A.; Knight, P.; Knipe, S. J.; Knott, S.; Kobuchi, T.; Köchl, F.; Kocsis, G.; Kodeli, I.; Kogan, L.; Kogut, D.; Koivuranta, S.; Kominis, Y.; Köppen, M.; Kos, B.; Koskela, T.; Koslowski, H. R.; Koubiti, M.; Kovari, M.; Kowalska-Strzęciwilk, E.; Krasilnikov, A.; Krasilnikov, V.; Krawczyk, N.; Kresina, M.; Krieger, K.; Krivska, A.; Kruezi, U.; Książek, I.; Kukushkin, A.; Kundu, A.; Kurki-Suonio, T.; Kwak, S.; Kwiatkowski, R.; Kwon, O. J.; Laguardia, L.; Lahtinen, A.; Laing, A.; Lam, N.; Lambertz, H. T.; Lane, C.; Lang, P. T.; Lanthaler, S.; Lapins, J.; Lasa, A.; Last, J. R.; Łaszyńska, E.; Lawless, R.; Lawson, A.; Lawson, K. D.; Lazaros, A.; Lazzaro, E.; Leddy, J.; Lee, S.; Lefebvre, X.; Leggate, H. J.; Lehmann, J.; Lehnen, M.; Leichtle, D.; Leichuer, P.; Leipold, F.; Lengar, I.; Lennholm, M.; Lerche, E.; Lescinskis, A.; Lesnoj, S.; Letellier, E.; Leyland, M.; Leysen, W.; Li, L.; Liang, Y.; Likonen, J.; Linke, J.; Linsmeier, Ch.; Lipschultz, B.; Liu, G.; Liu, Y.; Lo Schiavo, V. P.; Loarer, T.; Loarte, A.; Lobel, R. C.; Lomanowski, B.; Lomas, P. J.; Lönnroth, J.; López, J. M.; López-Razola, J.; Lorenzini, R.; Losada, U.; Lovell, J. J.; Loving, A. B.; Lowry, C.; Luce, T.; Lucock, R. M. A.; Lukin, A.; Luna, C.; Lungaroni, M.; Lungu, C. P.; Lungu, M.; Lunniss, A.; Lupelli, I.; Lyssoivan, A.; Macdonald, N.; Macheta, P.; Maczewa, K.; Magesh, B.; Maget, P.; Maggi, C.; Maier, H.; Mailloux, J.; Makkonen, T.; Makwana, R.; Malaquias, A.; Malizia, A.; Manas, P.; Manning, A.; Manso, M. E.; Mantica, P.; Mantsinen, M.; Manzanares, A.; Maquet, Ph.; Marandet, Y.; Marcenko, N.; Marchetto, C.; Marchuk, O.; Marinelli, M.; Marinucci, M.; Markovič, T.; Marocco, D.; Marot, L.; Marren, C. A.; Marshal, R.; Martin, A.; Martin, Y.; Martín de Aguilera, A.; Martínez, F. J.; Martín-Solís, J. R.; Martynova, Y.; Maruyama, S.; Masiello, A.; Maslov, M.; Matejcik, S.; Mattei, M.; Matthews, G. F.; Maviglia, F.; Mayer, M.; Mayoral, M. L.; May-Smith, T.; Mazon, D.; Mazzotta, C.; McAdams, R.; McCarthy, P. J.; McClements, K. G.; McCormack, O.; McCullen, P. A.; McDonald, D.; McIntosh, S.; McKean, R.; McKehon, J.; Meadows, R. C.; Meakins, A.; Medina, F.; Medland, M.; Medley, S.; Meigh, S.; Meigs, A. G.; Meisl, G.; Meitner, S.; Meneses, L.; Menmuir, S.; Mergia, K.; Merrigan, I. R.; Mertens, Ph.; Meshchaninov, S.; Messiaen, A.; Meyer, H.; Mianowski, S.; Michling, R.; Middleton-Gear, D.; Miettunen, J.; Militello, F.; Militello-Asp, E.; Miloshevsky, G.; Mink, F.; Minucci, S.; Miyoshi, Y.; Mlynář, J.; Molina, D.; Monakhov, I.; Moneti, M.; Mooney, R.; Moradi, S.; Mordijck, S.; Moreira, L.; Moreno, R.; Moro, F.; Morris, A. W.; Morris, J.; Moser, L.; Mosher, S.; Moulton, D.; Murari, A.; Muraro, A.; Murphy, S.; Asakura, N. N.; Na, Y. S.; Nabais, F.; Naish, R.; Nakano, T.; Nardon, E.; Naulin, V.; Nave, M. F. F.; Nedzelski, I.; Nemtsev, G.; Nespoli, F.; Neto, A.; Neu, R.; Neverov, V. S.; Newman, M.; Nicholls, K. J.; Nicolas, T.; Nielsen, A. H.; Nielsen, P.; Nilsson, E.; Nishijima, D.; Noble, C.; Nocente, M.; Nodwell, D.; Nordlund, K.; Nordman, H.; Nouailletas, R.; Nunes, I.; Oberkofler, M.; Odupitan, T.; Ogawa, M. T.; O'Gorman, T.; Okabayashi, M.; Olney, R.; Omolayo, O.; O'Mullane, M.; Ongena, J.; Orsitto, F.; Orszagh, J.; Oswuigwe, B. I.; Otin, R.; Owen, A.; Paccagnella, R.; Pace, N.; Pacella, D.; Packer, L. W.; Page, A.; Pajuste, E.; Palazzo, S.; Pamela, S.; Panja, S.; Papp, P.; Paprok, R.; Parail, V.; Park, M.; Parra Diaz, F.; Parsons, M.; Pasqualotto, R.; Patel, A.; Pathak, S.; Paton, D.; Patten, H.; Pau, A.; Pawelec, E.; Soldan, C. Paz; Peackoc, A.; Pearson, I. J.; Pehkonen, S.-P.; Peluso, E.; Penot, C.; Pereira, A.; Pereira, R.; Pereira Puglia, P. P.; Perez von Thun, C.; Peruzzo, S.; Peschanyi, S.; Peterka, M.; Petersson, P.; Petravich, G.; Petre, A.; Petrella, N.; Petržilka, V.; Peysson, Y.; Pfefferlé, D.; Philipps, V.; Pillon, M.; Pintsuk, G.; Piovesan, P.; Pires dos Reis, A.; Piron, L.; Pironti, A.; Pisano, F.; Pitts, R.; Pizzo, F.; Plyusnin, V.; Pomaro, N.; Pompilian, O. G.; Pool, P. J.; Popovichev, S.; Porfiri, M. T.; Porosnicu, C.; Porton, M.; Possnert, G.; Potzel, S.; Powell, T.; Pozzi, J.; Prajapati, V.; Prakash, R.; Prestopino, G.; Price, D.; Price, M.; Price, R.; Prior, P.; Proudfoot, R.; Pucella, G.; Puglia, P.; Puiatti, M. E.; Pulley, D.; Purahoo, K.; Pütterich, Th.; Rachlew, E.; Rack, M.; Ragona, R.; Rainford, M. S. J.; Rakha, A.; Ramogida, G.; Ranjan, S.; Rapson, C. J.; Rasmussen, J. J.; Rathod, K.; Rattá, G.; Ratynskaia, S.; Ravera, G.; Rayner, C.; Rebai, M.; Reece, D.; Reed, A.; Réfy, D.; Regan, B.; Regaña, J.; Reich, M.; Reid, N.; Reimold, F.; Reinhart, M.; Reinke, M.; Reiser, D.; Rendell, D.; Reux, C.; Reyes Cortes, S. D. A.; Reynolds, S.; Riccardo, V.; Richardson, N.; Riddle, K.; Rigamonti, D.; Rimini, F. G.; Risner, J.; Riva, M.; Roach, C.; Robins, R. J.; Robinson, S. A.; Robinson, T.; Robson, D. W.; Roccella, R.; Rodionov, R.; Rodrigues, P.; Rodriguez, J.; Rohde, V.; Romanelli, F.; Romanelli, M.; Romanelli, S.; Romazanov, J.; Rowe, S.; Rubel, M.; Rubinacci, G.; Rubino, G.; Ruchko, L.; Ruiz, M.; Ruset, C.; Rzadkiewicz, J.; Saarelma, S.; Sabot, R.; Safi, E.; Sagar, P.; Saibene, G.; Saint-Laurent, F.; Salewski, M.; Salmi, A.; Salmon, R.; Salzedas, F.; Samaddar, D.; Samm, U.; Sandiford, D.; Santa, P.; Santala, M. I. K.; Santos, B.; Santucci, A.; Sartori, F.; Sartori, R.; Sauter, O.; Scannell, R.; Schlummer, T.; Schmid, K.; Schmidt, V.; Schmuck, S.; Schneider, M.; Schöpf, K.; Schwörer, D.; Scott, S. D.; Sergienko, G.; Sertoli, M.; Shabbir, A.; Sharapov, S. E.; Shaw, A.; Shaw, R.; Sheikh, H.; Shepherd, A.; Shevelev, A.; Shumack, A.; Sias, G.; Sibbald, M.; Sieglin, B.; Silburn, S.; Silva, A.; Silva, C.; Simmons, P. A.; Simpson, J.; Simpson-Hutchinson, J.; Sinha, A.; Sipilä, S. K.; Sips, A. C. C.; Sirén, P.; Sirinelli, A.; Sjöstrand, H.; Skiba, M.; Skilton, R.; Slabkowska, K.; Slade, B.; Smith, N.; Smith, P. G.; Smith, R.; Smith, T. J.; Smithies, M.; Snoj, L.; Soare, S.; Solano, E. R.; Somers, A.; Sommariva, C.; Sonato, P.; Sopplesa, A.; Sousa, J.; Sozzi, C.; Spagnolo, S.; Spelzini, T.; Spineanu, F.; Stables, G.; Stamatelatos, I.; Stamp, M. F.; Staniec, P.; Stankūnas, G.; Stan-Sion, C.; Stead, M. J.; Stefanikova, E.; Stepanov, I.; Stephen, A. V.; Stephen, M.; Stevens, A.; Stevens, B. D.; Strachan, J.; Strand, P.; Strauss, H. R.; Ström, P.; Stubbs, G.; Studholme, W.; Subba, F.; Summers, H. P.; Svensson, J.; Świderski, Ł.; Szabolics, T.; Szawlowski, M.; Szepesi, G.; Suzuki, T. T.; Tál, B.; Tala, T.; Talbot, A. R.; Talebzadeh, S.; Taliercio, C.; Tamain, P.; Tame, C.; Tang, W.; Tardocchi, M.; Taroni, L.; Taylor, D.; Taylor, K. A.; Tegnered, D.; Telesca, G.; Teplova, N.; Terranova, D.; Testa, D.; Tholerus, E.; Thomas, J.; Thomas, J. D.; Thomas, P.; Thompson, A.; Thompson, C.-A.; Thompson, V. K.; Thorne, L.; Thornton, A.; Thrysøe, A. S.; Tigwell, P. A.; Tipton, N.; Tiseanu, I.; Tojo, H.; Tokitani, M.; Tolias, P.; Tomeš, M.; Tonner, P.; Towndrow, M.; Trimble, P.; Tripsky, M.; Tsalas, M.; Tsavalas, P.; Tskhakaya jun, D.; Turner, I.; Turner, M. M.; Turnyanskiy, M.; Tvalashvili, G.; Tyrrell, S. G. J.; Uccello, A.; Ul-Abidin, Z.; Uljanovs, J.; Ulyatt, D.; Urano, H.; Uytdenhouwen, I.; Vadgama, A. P.; Valcarcel, D.; Valentinuzzi, M.; Valisa, M.; Vallejos Olivares, P.; Valovic, M.; Van De Mortel, M.; Van Eester, D.; Van Renterghem, W.; van Rooij, G. J.; Varje, J.; Varoutis, S.; Vartanian, S.; Vasava, K.; Vasilopoulou, T.; Vega, J.; Verdoolaege, G.; Verhoeven, R.; Verona, C.; Verona Rinati, G.; Veshchev, E.; Vianello, N.; Vicente, J.; Viezzer, E.; Villari, S.; Villone, F.; Vincenzi, P.; Vinyar, I.; Viola, B.; Vitins, A.; Vizvary, Z.; Vlad, M.; Voitsekhovitch, I.; Vondráček, P.; Vora, N.; Vu, T.; Pires de Sa, W. W.; Wakeling, B.; Waldon, C. W. F.; Walkden, N.; Walker, M.; Walker, R.; Walsh, M.; Wang, E.; Wang, N.; Warder, S.; Warren, R. J.; Waterhouse, J.; Watkins, N. W.; Watts, C.; Wauters, T.; Weckmann, A.; Weiland, J.; Weisen, H.; Weiszflog, M.; Wellstood, C.; West, A. T.; Wheatley, M. R.; Whetham, S.; Whitehead, A. M.; Whitehead, B. D.; Widdowson, A. M.; Wiesen, S.; Wilkinson, J.; Williams, J.; Williams, M.; Wilson, A. R.; Wilson, D. J.; Wilson, H. R.; Wilson, J.; Wischmeier, M.; Withenshaw, G.; Withycombe, A.; Witts, D. M.; Wood, D.; Wood, R.; Woodley, C.; Wray, S.; Wright, J.; Wright, J. C.; Wu, J.; Wukitch, S.; Wynn, A.; Xu, T.; Yadikin, D.; Yanling, W.; Yao, L.; Yavorskij, V.; Yoo, M. G.; Young, C.; Young, D.; Young, I. D.; Young, R.; Zacks, J.; Zagorski, R.; Zaitsev, F. S.; Zanino, R.; Zarins, A.; Zastrow, K. D.; Zerbini, M.; Zhang, W.; Zhou, Y.; Zilli, E.; Zoita, V.; Zoletnik, S.; Zychor, I.; JET Contributors

    2017-10-01

    The 2014-2016 JET results are reviewed in the light of their significance for optimising the ITER research plan for the active and non-active operation. More than 60 h of plasma operation with ITER first wall materials successfully took place since its installation in 2011. New multi-machine scaling of the type I-ELM divertor energy flux density to ITER is supported by first principle modelling. ITER relevant disruption experiments and first principle modelling are reported with a set of three disruption mitigation valves mimicking the ITER setup. Insights of the L-H power threshold in Deuterium and Hydrogen are given, stressing the importance of the magnetic configurations and the recent measurements of fine-scale structures in the edge radial electric. Dimensionless scans of the core and pedestal confinement provide new information to elucidate the importance of the first wall material on the fusion performance. H-mode plasmas at ITER triangularity (H  =  1 at β N ~ 1.8 and n/n GW ~ 0.6) have been sustained at 2 MA during 5 s. The ITER neutronics codes have been validated on high performance experiments. Prospects for the coming D-T campaign and 14 MeV neutron calibration strategy are reviewed.

  14. Multi-Fluid Modeling of Low-Recycling Divertor Regimes

    SciTech Connect

    Smirnov, R. D.; Pigarov, A. Y.; Krasheninnikov, S. I.; Rognlien, T. D.; Soukhanovskii, V. A.; Rensink, M. E.; Maingi, Rajesh; Skinner, C. H.; Stotler, D. P.; Bell, R. E.; Kugel, H. W.

    2010-01-01

    The low-recycling regimes of divertor operation in a single-null NSTX magnetic configuration are studied using computer simulations with the edge plasma transport code UEDGE. The edge plasma transport properties pertinent to the low-recycling regimes are demonstrated. These include the flux-limited character of the parallel heat transport and the high plasma temperatures with the flattened profiles in the scrape-off-layer. It is shown that to maintain the balance of particle fluxes at the core interface the deuterium gas puffing rate should increase as the divertor recycling coefficient decreases. The radial profiles of the heat load to the outer divertor plate, the upstream radial plasma profiles, and the effects of the cross-field plasma transport in the low-recycling regimes are discussed. It is also shown that recycling of lithium impurities evaporating from the divertor plate at high surface temperatures can reverse the low-recycling divertor operational regime to the high-recycling one and may cause thermal instability of the divertor plate. (C) 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim

  15. Divertor conditions near double null in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Brunner, Dan; Labombard, Brian; Kuang, Adam; Terry, Jim; Mumgaard, Bob; Wolfe, Steve

    2016-10-01

    Many tokamak reactor designs utilize a double-null equilibrium for the boundary plasma because of the expected benefits of heat flux sharing between the two outer divertor leg as well as the attractiveness of the high-field side scrape-off layer plasma in double-null for RF actuators. However, there has been very little reported on boundary plasma conditions near double null, especially at the divertor plate. And, due to the narrow boundary plasma width, there is concern of the precision to which a double-null equilibrium must be controlled to maintain divertor heat flux sharing. To this end, a series of experiments were performed varying the magnetic balance around double null. The magnetic balance between the two nulls was scanned shot-to-shot in L-, I-, and H-mode plasmas. In addition, current and density scans were performed in L-mode plasmas. Results will be presented for relative balances of divertor particle and energy fluxes to the four divertors (inboard/outboard, upper/lower) as well as the sensitivity of changes in divertor conditions to the magnetic balance. Supported by USDoE Award DE-FC02-99ER54512.

  16. The Magnetic Field Structure of a Snowflake Divertor

    SciTech Connect

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

    2008-05-30

    The snowflake divertor exploits a tokamak geometry in which the poloidal magnetic field null approaches second order; the name stems from the characteristic hexagonal, snowflake-like, shape of the separatrix for an exact second-order null. The proximity of the poloidal field structure to that of a second-order null substantially modifies edge magnetic properties compared to the standard X-point geometry; this, in turn, affects the edge plasma behavior. Modifications include: (1) The flux expansion near the null-point becomes 2-3 times larger; (2) The connection length between the equatorial plane and divertor plate significantly increases; (3) Magnetic shear just inside the separatrix becomes much larger; and (4) In the open-field-line region, the squeezing of the flux-tubes near the null-point increases, thereby causing stronger decoupling of the plasma turbulence in the divertor legs and in the main SOL. These effects can be used to reduce the power load on the divertor plates and/or to suppress the 'bursty' component of the heat flux. It is emphasized that the snowflake divertor can be created by a relatively simple set of poloidal field coils situated beyond the toroidal field coils. Analysis of the robustness of the proposed divertor configuration with respect to changes of the plasma current distribution is presented and it is concluded that, even if the null is close to the second order, the configuration is quite robust.

  17. Snowflake Divertor Configuration Studies in DIII-D Tokamak

    NASA Astrophysics Data System (ADS)

    Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; 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.

    2014-10-01

    Recent DIII-D studies show that the snowflake (SF) divertor enables significant manipulation of divertor heat transport for power exhaust in attached and radiative divertor conditions, between and during edge localized modes (ELMs), while maintaining good H-mode confinement. Results include: 1) Increased scrape-off layer (SOL) width suggesting enhanced divertor heat transport; 2) Direct measurements of divertor null-region poloidal beta βp >> 1 in support of the theoretically proposed instability mechanism leading to fast convective plasma redistribution, especially efficient during ELMs, and contribution to 1); 3) Weak effect on pedestal profile and stability resulting in essentially unchanged ELM regime; 4) Reduction of Type-I ELM energy loss; 5) In radiative SF divertor regimes with D2 seeding, a significant reduction of peak heat fluxes between and during ELMs, as in standard H-modes. Work supported by the US Department of Energy under DE-AC52-07NA27344, DE-AC02-09CH11466, DE-FC02-04ER54698, and DE-AC04-94AL85000.

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

    NASA Astrophysics Data System (ADS)

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

    2011-10-01

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

  19. Plasma-sprayed beryllium for ITER

    SciTech Connect

    Castro, R.G.; Stanek, P.W.; Elliott, K.E.; Youchison, D.L.; Watson, R.D.; Walsh, D.S.

    1995-12-31

    Plasma-spray technology is under investigation as a method for producing high thermal conductivity beryllium coatings for use in magnetic fusion applications. Recent investigations have focused on optimizing the plasma-spray process for depositing beryllium coatings on damaged beryllium surfaces. Of particular interest has been optimizing the processing parameters to maximize the through-thickness thermal conductivity of the beryllium coatings. Experimental results will be reported on the use of secondary H{sub 2} gas additions to improve the melting of the beryllium powder and negative transferred-arc cleaning to prepare beryllium surfaces prior to depositing beryllium. Information will also b presented on thermal fatigue tests which were performed on beryllium coated ISX-B beryllium limiter tiles using 10 sec cycle times with 60 sec cooldowns and an International Thermonuclear Experimental Reactor (ITER) relevant divertor heat flux slightly in excess of 5 MW/m{sup 2}.

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

    DOE PAGES

    Rognlien, Thomas D.; McLean, Adam G.; Fenstermacher, Max E.; ...

    2017-01-27

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

  1. Developing snowflake divertor physics basis in the DIII-D, NSTX and NSTX-U tokamaks aimed at the divertor power exhaust solution [Snowflake divertor experiments in the DIII-D, NSTX and NSTX-U tokamaks aimed at the development of the divertor power exhaust solution

    DOE PAGES

    Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; ...

    2016-06-02

    Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment were performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large ELMs. However, a stable partial detachment of the outer strike pointmore » was also achieved where inter-ELM peak heat flux was reduced by factors 3-5, and peak ELM heat flux was reduced by up to 80% (cf. standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower ne, and ELM energy and divertor peak heat flux reduction, especially prominent in radiative D2-seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D2-seeded SF divertor at PSOL = 3 - 4 MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multi-fluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected PSOL ≃9 MW case. In conclusion, the radiative SF divertor with carbon impurity provides a wider ne operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar qpeak reduction factors (cf. standard divertor).« less

  2. Recent analysis of key plasma wall interactions issues for ITER

    NASA Astrophysics Data System (ADS)

    Roth, Joachim; Tsitrone, E.; Loarte, A.; Loarer, Th.; Counsell, G.; Neu, R.; Philipps, V.; Brezinsek, S.; Lehnen, M.; Coad, P.; Grisolia, Ch.; Schmid, K.; Krieger, K.; Kallenbach, A.; Lipschultz, B.; Doerner, R.; Causey, R.; Alimov, V.; Shu, W.; Ogorodnikova, O.; Kirschner, A.; Federici, G.; Kukushkin, A.; EFDA PWI Task Force, ITER PWI Team, FusionEnergy, ITPA SOL/DIV

    2009-06-01

    Plasma wall interaction (PWI) is important for the material choice in ITER and for the plasma scenarios compatible with material constraints. In this paper, different aspects of the PWI are assessed in their importance for the initial wall materials choice: CFC for the strike point tiles, W in the divertor and baffle and Be on the first wall. Further material options are addressed for comparison, such as W divertor/Be first wall and all-W or all-C. One main parameter in this evaluation is the particle flux to the main vessel wall. One detailed plasma scenario exists for a Q = 10 ITER discharge [G. Federici et al., J. Nucl. Mater. 290-293 (2001) 260] which was taken as the basis of further erosion and tritium retention evaluations. As the assessment of steady state wall fluxes from a scaling of present fusion devices indicates that global wall fluxes may be a factor of 4 ± 3 higher, this margin has been adopted as uncertainty of the scaling. With these wall and divertor fluxes, important PWI processes such as erosion and tritium accumulation have been evaluated: It was found that the steady state erosion is no problem for the lifetime of plasma-facing divertor components. Be wall erosion may pose a problem in case of a concentration of the wall fluxes to small wall areas. ELM erosion may drastically limit the PFC lifetime if ELMs are not mitigated to energies below 0.5 MJ. Dust generation is still a process which requires more attention. Conversion from gross or net erosion to dust and the assessment of dust on hot surfaces need to be investigated. For low- Z materials the build-up of the tritium inventory is dominated by co-deposition with eroded wall atoms. For W, where erosion and tritium co-deposition are small, the implantation, diffusion and bulk trapping constitute the dominant retention processes. First extrapolations with models based on laboratory data show small contributions to the inventory. For later ITER phases and the extrapolation to DEMO

  3. 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.

  4. Model for particle balance in pumped divertors (pre-VORTEX)

    SciTech Connect

    Hogan, J.T.

    1990-08-01

    An internally consistent model for particle transport in an open divertor geometry has been developed. Embodied in a new code, pre-VORTEX, the model couples the particle balance in the plasma core, the scrape-off layer, the open divertor channels, and the vacuum'' regions. This mutual coupling is particularly important in determining the conditions required for high recycling in the divertor. The plasma core is considered to have a relatively quiescent core region and a less well confined edge-localized mode''(ELM) region. The scrape-off layer is modeled with one-dimensional parallel and perpendicular transport. A two-point divertor channel model is used; it is similar to previous models, but with the addition of new physical processes: hydrogen charge exchange, impurity thermal charge exchange, and flux-limited parallel transport. Wall recycling data are required to describe the differing recycling properties of the wall regions and the divertor plates. Given local plasma diffusivities and wall recycling properties, the model predicts the volume-averaged density and global particle confinement time. The input data are uncertain, and a major use for the model is to permit comparison with data. The final model, VORTEX, is intended for application to the analysis of divertor confinement experiments; it is coupled to a one-and-one-half--dimensional transport code and uses detailed geometric input from equilibrium fitting codes, experimentally measured core profiles, and such parameters as can be measured in the scrape-off layer. The pre-VORTEX model is compared as a stand-alone code with typical data from the DIII-D experiment and applied to the proposed DIII-D Advanced Divertor Project.

  5. Comparison study of toroidal-field divertors for a compact reversed-field pinch reactor

    SciTech Connect

    Bathke, C.G.; Krakowski, R.A.; Miller, R.L.

    1985-01-01

    Two divertor configurations for the Compact Reversed-Field Pinch Reactor (CRFPR) based on diverting the minority (toroidal) field have been reported. A critical factor in evaluating the performance of both poloidally symmetric and bundle divertor configurations is the accurate determination of the divertor connection length and the monitoring of magnetic islands introduced by the divertors, the latter being a three-dimensional effect. To this end the poloidal-field, toroidal-field, and divertor coils and the plasma currents are simulated in three dimensions for field-line tracings in both the divertor channel and the plasma-edge regions. The results of this analysis indicate a clear preference for the poloidally symmetric toroidal-field divertor. Design modifications to the limiter-based CRFPR design that accommodate this divertor are presented.

  6. 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.

  7. Long-term fuel retention and release in JET ITER-Like Wall at ITER-relevant baking temperatures

    NASA Astrophysics Data System (ADS)

    Heinola, K.; Likonen, J.; Ahlgren, T.; Brezinsek, S.; De Temmerman, G.; Jepu, I.; Matthews, G. F.; Pitts, R. A.; Widdowson, A.; Contributors, JET

    2017-08-01

    The fuel outgassing efficiency from plasma-facing components exposed in JET-ILW has been studied at ITER-relevant baking temperatures. Samples retrieved from the W divertor and Be main chamber were annealed at 350 and 240 °C, respectively. Annealing was performed with thermal desoprtion spectrometry (TDS) for 0, 5 and 15 h to study the deuterium removal effectiveness at the nominal baking temperatures. The remained fraction was determined by emptying the samples fully of deuterium by heating W and Be samples up to 1000 and 775 °C,respectively. Results showed the deposits in the divertor having an increasing effect to the remaining retention at temperatures above baking. Highest remaining fractions 54 and 87 % were observed with deposit thicknesses of 10 and 40 μm, respectively. Substantially high fractions were obtained in the main chamber samples from the deposit-free erosion zone of the limiter midplane, in which the dominant fuel retention mechanism is via implantation: 15 h annealing resulted in retained deuterium higher than 90 % . TDS results from the divertor were simulated with TMAP7 calculations. The spectra were modelled with three deuterium activation energies resulting in good agreement with the experiments.

  8. Analysis of fuelling requirements in ITER H-modes with SOLPS-EPED1 derived scalings

    NASA Astrophysics Data System (ADS)

    Polevoi, A. R.; Loarte, A.; Kukushkin, A. S.; Pacher, H. D.; Pacher, G. W.; Köchl, F.

    2017-02-01

    Fuelling requirements for ITER are analysed in relation to pellet fuelling and ELM pacing, and a divertor power load control consistent with the ITER pumping and fuel throughput capabilities. The plasma parameters at the separatrix and the particle sources are derived from scalings based on SOLPS simulations. Effective transport coefficients in the H-mode pedestal are derived from EPED1 + SOLPS scalings for the pedestal height and width. 1.5D transport is simulated in the ASTRA framework. The operating window for ITER DT plasmas with the required fusion performance and level of ELM, and divertor power load control compatible with ITER fuelling and pumping capabilities, is determined. It is shown that the flexibility of the ITER fuelling systems, comprising pellet and gas injection systems, enables operation with Q  =  10, which was found to be marginal in previous studies following a similar approach but with different assumptions. The present assessment shows that a reduction of < {{n}e}> by a factor ~2 (from 9 to 5  ×  1019 m-3) in 15 MA H-mode plasmas leads to a reduction in the required pellet fuelling rate by a factor of four. Results of the analysis of the fuelling requirements for a range of ITER scenarios are found to be similar to those obtained with the JINTRAC code that included 2D modelling of the edge plasma.

  9. 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.

  10. Compatibility of Detached Divertor Operation with Robust Edge Pedestal Performance

    NASA Astrophysics Data System (ADS)

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

    2014-10-01

    The compatibility of radiative detached divertor operation with the maintenance of a robust H-mode pedestal is examined in DIII-D. A density scan with deuterium injection into H-mode spanned a range of divertor conditions from fully attached, ~30 eV at the target, with little divertor radiation to a fully detached with Te < 5 eV throughout the divertor up to the X-point. Over this scan of pedestal density from n /nGW = 30% to 60% the pedestal Te was reduced from 800 eV to 350 eV, representing a ~20% reduction in pedestal pressure with a similar reduction in normalized energy confinement. The reduction in pedestal pressure at high density was found to be consistent with a reduced pedestal ELM MHD stability limit at high collisionality. The scaling of the pedestal top pressure with density was also consistent with the EPED model, which assumes an additional constraint on the local pressure gradient. The MHD stability limit at the highest collisionality depends on details of the ELM instability growth rate normalization. This result is encouraging for future burning plasmas where a low collisionality pedestal is expected to be maintained even for high density detached divertor operation. Work supported by the US DOE under DE-FC02-04ER54698 and DE-AC52-07NA27344.

  11. Upgraded divertor Thomson scattering system on DIII-D

    SciTech Connect

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

    2016-11-15

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

  12. Upgraded divertor Thomson scattering system on DIII-D

    NASA Astrophysics Data System (ADS)

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

    2016-11-01

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

  13. 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

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

    NASA Astrophysics Data System (ADS)

    Soukhanovskii, V. A.

    2017-06-01

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

  15. Progress on the application of ELM control schemes to ITER scenarios from the non-active phase to DT operation

    NASA Astrophysics Data System (ADS)

    Loarte, A.; Huijsmans, G.; Futatani, S.; Baylor, L. R.; Evans, T. E.; Orlov, D. M.; Schmitz, O.; Becoulet, M.; Cahyna, P.; Gribov, Y.; Kavin, A.; Sashala Naik, A.; Campbell, D. J.; Casper, T.; Daly, E.; Frerichs, H.; Kischner, A.; Laengner, R.; Lisgo, S.; Pitts, R. A.; Saibene, G.; Wingen, A.

    2014-03-01

    Progress in the definition of the requirements for edge localized mode (ELM) control and the application of ELM control methods both for high fusion performance DT operation and non-active low-current operation in ITER is described. Evaluation of the power fluxes for low plasma current H-modes in ITER shows that uncontrolled ELMs will not lead to damage to the tungsten (W) divertor target, unlike for high-current H-modes in which divertor damage by uncontrolled ELMs is expected. Despite the lack of divertor damage at lower currents, ELM control is found to be required in ITER under these conditions to prevent an excessive contamination of the plasma by W, which could eventually lead to an increased disruptivity. Modelling with the non-linear MHD code JOREK of the physics processes determining the flow of energy from the confined plasma onto the plasma-facing components during ELMs at the ITER scale shows that the relative contribution of conductive and convective losses is intrinsically linked to the magnitude of the ELM energy loss. Modelling of the triggering of ELMs by pellet injection for DIII-D and ITER has identified the minimum pellet size required to trigger ELMs and, from this, the required fuel throughput for the application of this technique to ITER is evaluated and shown to be compatible with the installed fuelling and tritium re-processing capabilities in ITER. The evaluation of the capabilities of the ELM control coil system in ITER for ELM suppression is carried out (in the vacuum approximation) and found to have a factor of ˜2 margin in terms of coil current to achieve its design criterion, although such a margin could be substantially reduced when plasma shielding effects are taken into account. The consequences for the spatial distribution of the power fluxes at the divertor of ELM control by three-dimensional (3D) fields are evaluated and found to lead to substantial toroidal asymmetries in zones of the divertor target away from the separatrix

  16. Influence of atomic physics on EDGE2D-EIRENE simulations of JET divertor detachment with carbon and beryllium/tungsten plasma-facing components

    NASA Astrophysics Data System (ADS)

    Guillemaut, C.; Pitts, R. A.; Kukushkin, A. S.; Gunn, J. P.; Bucalossi, J.; Arnoux, G.; Belo, P.; Brezinsek, S.; Brix, M.; Corrigan, G.; Devaux, S.; Flanagan, J.; Groth, M.; Harting, D.; Huber, A.; Jachmich, S.; Kruezi, U.; Lehnen, M.; Marchetto, C.; Marsen, S.; Meigs, A. G.; Meyer, O.; Stamp, M.; Strachan, J. D.; Wiesen, S.; Wischmeier, M.; EFDA Contributors, JET

    2014-09-01

    The EDGE2D-EIRENE code is applied for simulation of divertor detachment during matched density ramp experiments in high triangularity, L-mode plasmas in both JET-Carbon (JET-C) and JET-ITER-Like Wall (JET-ILW). The code runs without drifts and includes either C or Be as impurity, but not W, assuming that the W targets have been coated with Be via main chamber migration. The simulations reproduce reasonably well the observed particle flux detachment as density is raised in both JET-C and JET-ILW experiments and can better match the experimental in-out divertor target power asymmetry if the heat flux entering the outer divertor is artificially set at around 2-3 times that entering the inner divertor. A careful comparison between different sets of atomic physics processes used in EIRENE shows that the detachment modelled by EDGE2D-EIRENE relies only on an increase of the particle sinks and not on a decrease of the ionization source. For the rollover and the beginning of the partially detached phase, the particle losses by perpendicular transport and the molecular activated recombination processes are mainly involved. For a deeper detachment with significant target ion flux reduction, volume recombination appears to be the main contributor. The elastic molecule-ion collisions are also important to provide good neutral confinement in the divertor and thus stabilize the simulations at low electron temperature (Te), when the sink terms are strong. Comparison between EDGE2D-EIRENE and SOLPS4.3 simulations of the density ramp in C shows similar detachment trends, but the importance of the elastic ion-molecule collisions is reduced in SOLPS4.3. Both codes suggest that any process capable of increasing the neutral confinement in the divertor should help to improve the modelling of the detachment. A further outcome of this work has been to demonstrate that key JET divertor diagnostic signals—Langmuir probe Te and bolometric tomographic reconstructions—are running beyond

  17. Surface analysis of tiles and samples exposed to the first JET campaigns with the ITER-like wall

    NASA Astrophysics Data System (ADS)

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

    2014-04-01

    This paper reports on the first post-mortem analyses of tiles removed from JET after the first campaigns with the ITER-like wall (ILW) during 2011-12 [1]. Tiles from the divertor have been analysed by ion beam analysis techniques and by secondary ion mass spectrometry to determine the amount of beryllium deposition and deuterium retention in the tiles exposed to the scrape-off layer. Films 10-20 μm thick were present at the top of tile 1, but only very thin films (< 1 μm) were found in the shadowed areas and on other divertor tiles. The total amount of Be found in the divertor following the ILW campaign was a factor of ˜ 9 less than the material deposited in the 2007-09 carbon campaign, after allowing for the longer operations in 2007-09.

  18. 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.

  19. 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.

  20. Innovative divertor concept development on DIII-D and EAST

    SciTech Connect

    Guo, H. Y.; Allen, S.; Canik, J.; Hill, D. N.; Leonard, T.; Sang, C. F.; Stangeby, P. C.; Thomas, D. M.; Unterberg, Z.; Luo, G. N.; Wang, L.; Wan, B. N.; Xu, G. S.

    2016-06-02

    A critical issue facing the design and operation of next-step high-power steady-state fusion devices is the control of heat fluxes and erosion at the plasma-facing components, in particular, the divertor target plates. A new initiative has been launched on DIII-D to develop and demonstrate innovative boundary plasma-materials interface solutions. The central purposes of this new initiative are to advance scientific understanding in this critical area and develop an advanced divertor concept for application to next-step fusion devices. Finally, DIII-D will leverage strong collaborative efforts on the EAST superconducting tokamak for extending integrated high performance advanced divertor solutions to true steady-state.

  1. 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.

  2. 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.

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

    DOE PAGES

    Ahn, J. -W.; Maingi, R.; Canik, J. M.; ...

    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

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

    NASA Astrophysics Data System (ADS)

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

    2014-12-01

    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 few 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. 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.

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

    SciTech Connect

    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 few 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.

  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. 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.

  8. Type-I ELM substructure on the divertor target plates in ASDEX Upgrade

    NASA Astrophysics Data System (ADS)

    Eich, T.; Herrmann, A.; Neuhauser, J.; Dux, R.; Fuchs, J. C.; Günter, S.; Horton, L. D.; Kallenbach, A.; Lang, P. T.; Maggi, C. F.; Maraschek, M.; Rohde, V.; Schneider, W.; ASDEX Upgrade Team

    2005-06-01

    In the ASDEX Upgrade tokamak, the power deposition structures on the divertor target plates during type-I edge localized modes (ELMs) have been investigated by infrared thermography. In addition to the axisymmetric strike line, several poloidally displaced stripes are resolved, identifying an ELM as a composite of several subevents. This pattern is interpreted as being a signature of the helical perturbations in the low field side edge during the non-linear ELM evolution. Based on this observation, the ELM related magnetic perturbation in the midplane can be derived from the target load pattern. In the start phase of an ELM collapse, average toroidal mode numbers around n ap 3-5 are found evolving to values of n ap 12-14 during the ELM power deposition maximum. Further information about the non-linear evolution of the ELM mode structure is obtained from statistical analyses of the spatial distribution, heat flux amplitudes and number of single stripes.

  9. 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.

  10. 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.

  11. Deuterium and tritium separation in a tokamak reactor divertor layer

    NASA Astrophysics Data System (ADS)

    Tokar', M. Z.

    1989-04-01

    It's shown that the plasma isotope composition in a tokamak reactor divertor layer changes along the magnetic field and can notable differ from the gas composition in a pumping chamber. Heavier tritium must concentrate in the hot plasma far from the divertor plate due to thermal force stipulated by mutial collisions of deuterium and tritium ions. This circumstance is favourable from the point of view of tritium cycle optimization and must facilitate solution of the problem of tritium accumulation in the reactor construction elements.

  12. 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.

  13. 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.

  14. US ITER Moving Forward

    ScienceCinema

    US ITER / ORNL

    2016-07-12

    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.

  15. 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.

  16. Tungsten dust impact on ITER-like plasma edge

    DOE PAGES

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

    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

  17. 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.

  18. ITER edge-plasma conditions versus pump configuration

    SciTech Connect

    Werley, K.A.; Cohen, S.A. |

    1991-12-11

    Two-dimensional fluid simulations of ITER double-null divertor scrape-off-layer plasma conditions have been restricted to examining a single outboard divertor plate with up/down symmetry assumed. The present work evaluates the effect of pumping at only one plate on particle flow patterns and other parameters of interest. Pumping only at one plate results in reduced sheath temperatures at both plates but an increased heat flux at the pumped plate. The physics assumptions for separatrix density (n{sub SEP}={l_angle}n{r_angle}/3.5) and for radial particle diffusivity (D=0.66m{sup 2}/s) used in the simulation of ITER edge plasma result in particle throughputs two orders of magnitude greater than that required for acceptable fusion-product ash removal rates. The particle confinement time, however, is an order of magnitude shorter than the transport energy confinement time, {tau}{sub E}. Plasmas (D=0.04m{sup 2}/s) which would have {tau}{sub p} {approximately} {tau}{sub E} are evaluated and found to have unacceptably high plasma temperatures and heat flux at the plate. Ash removal rates are still acceptable. A plasma recycle coefficient of R=0.965 reduces the particle throughput by a factor of 2.8 below the no-recycle case.

  19. 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

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

    NASA Astrophysics Data System (ADS)

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

    2007-08-01

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

  1. 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.

  2. 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.

  3. High Flux FRC Facility for the Stability, Confinement and ITER Divertor Studies

    SciTech Connect

    Hoffman, Alan L.; Milroy, Richard D.

    2014-01-31

    The TCS (Translation, Confinement, & Sustainment) program was begun on 7 August, 1996 to renew basic studies of the Field Reversed Configuration (FRC). The program made use of the old LSX (Large s Experiment) device, which was constructed at STI during the period from 1986 to 1990, but only operated for one year due to a DOE decision at the time to focus exclusively on the tokamak configuration. LSX was transferred to the University of Washington in 1992 and modified (LSX/mod) to perform Tokamak Refueling by Accelerated Plasmoids (TRAP) experiments. The TRAP program was funded from 7 August, 1992 until 6 August, 1996, but was utilized for an additional year while TCS was being constructed. During the first TCS funding period TCS was completed and initial experiments were begun. A large multi-megawatt RF power supply was built by Los Alamos National Laboratory (LANL) for use with a Rotating Magnetic Field (RMF) system, and LANL has been a continuing participant in our experimental program. A smaller prototype facility, called the Star Thrust Experiment (STX) was also built and operated in this period, partly with NASA funding, before TCS came on-line. A final report for this construction period was submitted in September 2000. A first renewal period (2.5 years) provided operating funds for the period between July 7, 2000 and January 6, 2003. A great deal of progress was made in understanding the use of RMF to both form and sustain FRCs during this period. The principal result of the experimental program was the formation of quasi steady-state (as long as RMF power was available) FRCs with densities in the 1-3x1019 m-3 range. However, the plasma temperature (Te or Ti) was limited to sub-25 eV, except transiently during start-up, by the rapid accumulation of impurities. This is not surprising since TCS was only designed to demonstrate RMF flux build-up and was not provided with either fueling capabilities or modern vacuum conditioning technology. (The unplanned for long time steady-state operation was due entirely to recycling.) TCS employed a multi-section quartz vacuum vessel with greased “O”-ring seals. A final report for this second funding period was submitted in May of 2003.

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

    DOE PAGES

    Soukhanovskii, V. A.

    2017-04-28

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

  5. Attainment of a stable, fully detached plasma state in innovative divertor configurations

    NASA Astrophysics Data System (ADS)

    Umansky, M. V.; LaBombard, B.; Brunner, D.; Rensink, M. E.; Rognlien, T. D.; Terry, J. L.; Whyte, D. G.

    2017-05-01

    A computational study of long-legged tokamak divertor configurations is performed with the edge transport code UEDGE. Several divertor configurations are considered, with radially or vertically extended, tightly baffled, outer divertor legs and with or without a secondary X-point in the divertor leg volume. For otherwise identical conditions, a scan of the input power from the core plasma is performed. As the power is reduced to a threshold value, the plasma in the outer leg transitions to a fully detached state, which defines the upper limit on the power for detached divertor operation. Reducing the power further results in the detachment front shifting upstream but remains stable. At low power, the detachment front eventually moves all the way to the primary X-point, which is usually associated with degradation of the core plasma, and this defines the lower limit on the power for the detached divertor operation. For the studied parameters, for long-legged divertors, the detached operation window is quite large, in particular, for the X-point target configuration using a secondary X-point in the divertor leg volume, allowing a factor of 5-10 variations in the input power. For the same parameters, for the standard divertor configuration, the detached operation window is very small or even non-existent. The present modeling results suggest the possibility of stable fully detached divertor operation for a tokamak with tightly baffled extended divertor legs.

  6. Iter and Ornl

    NASA Astrophysics Data System (ADS)

    Uckan, N. A.; Milora, S. L.

    2004-11-01

    ITER (means ``the way''), a tokamak burning plasma experiment, is the next step device toward making fusion energy a reality. The programmatic objective of ITER is to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes. ITER began in 1985 as collaboration between the Russian Federation (former Soviet Union), the USA, European Union, and Japan. ITER conceptual and engineering design activities led to a detailed design in 2001. The USA opted out of the project between 1999-2003, but rejoined in 2004 for site selection and construction negotiations. China and Korea joined the project in 2003. Negotiations are continuing and a decision on the site for ITER construction [France versus Japan] is pending. The ITER international undertaking is an unprecedented scale and the six ITER parties represent 40% of the world population. By 2018, ITER will produce a fusion power of 500 million Watts for time periods up to an hour with one-tenth of the power needed to sustain it. Steady state operation is also possible at lower power levels with higher fraction of circulated power. The ITER parties invested about $1 billion into the research and development (R) and related fusion experiments to establish the ITER's feasibility. ORNL has been a key player in the ITER project and contributed to its physics and engineering design and related R since its inception. Recently, the U.S. DOE selected the PPPL/ORNL partnership to lead the U.S. project office for ITER.

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

    NASA Astrophysics Data System (ADS)

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

    2016-10-01

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

  8. Reconstruction in 3D of the fast wave fields in ITER, DIII-D, C-Mod and NSTX, including the coupling of full-wave and particle codes to resolve finite orbit effects

    SciTech Connect

    Green, David L; Jaeger, Erwin Frederick; Berry, Lee A; Choi, M.

    2009-01-01

    The rf-SciDAC collaboration is developing computer simulations to predict the damping of radio frequency (rf) waves in fusion plasmas. Here we extend self-consistent quasi-linear calculations of ion cyclotron resonant heating to include the finite drift of ions from magnetic flux surfaces and rf induced spatial transport. The all-orders spectral wave solver AORSA is iteratively coupled with a particle based update of the plasma distribution function using a quasi-linear diffusion tersor representative of the (k) over right arrow spectrum. Initial results are presented for a high power minority heating scenario on the Alcator C-Mod tokamak and a high harmonic beam heating scenario on DIII-D. Finite orbit effects are shown to give a less peaked perpendicular energy profile and rf induced transport.

  9. 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)

  10. The effects of the Snowflake Divertor on upstream SOL profiles

    NASA Astrophysics Data System (ADS)

    Tsui, C. K.; Boedo, J. A.; Coda, S.; Labit, B.; Maurizio, R.; Nespoli, F.; Reimerdes, H.; Theiler, C.; Spolaore, M.; Vianello, N.; Lunt, T.; Vijvers, W. A. J.; Walkden, N.; the EUROfusion MST1 Team Team; the TCV Team Team

    2016-10-01

    The Snowflake Divertor creates separated volumes within the SOL and divertor that feature strikingly different ne, Te profiles, and decay lengths, as measured with a scanning probe. Profiles were taken at the outer midplane of TCV plasmas with snowflake divertors as well as just above the X-points within the region of enhanced βpol. Density shoulders in the far SOL in single null plasmas are relaxed by secondary X-points, while effects are more complex in the near SOL. These changes were observed whether the secondary X-point was placed in the low field side SOL, or in the high field side SOL. Additionally, target profiles measured with IR camera and Langmiur probes that were taken in the divertor leg opposite the secondary X-point also show features on the flux surface corresponding to the secondary X-point. Fluctuation statistics from the reciprocating probe as well as comparisons made between upstream and downstream measurements are considered for their implications on SOL transport. Support from EUROfusion Grant 633053 and US DOE Grant DE-SC0010529 are gratefully acknowledged.

  11. Line Shapes and Opacity Studies in Divertor Plasmas

    SciTech Connect

    Rosato, J.

    2008-10-22

    Large or dense divertor plasmas of magnetic fusion devices can be optically thick to the resonance lines of the hydrogen isotopes. In this work we examine the sensitivity of the line radiation transport to the detailed structure of the spectral profiles.

  12. Visible spectroscopy in the DIII-D divertor

    SciTech Connect

    Brooks, N.H.; Fehling, D.; Hillis, D.L.; Klepper, C.C.; Naumenko, N.; Tugarinov, S.; Whyte, D.G.

    1996-06-01

    Spectroscopy measurements in the DIII-D divertor have been carried out with a survey spectrometer which provides simultaneous registration of the visible spectrum over the region 400--900 nm with a resolution of 0.2 nm. Broad spectral coverage is achieved through use of a fiberoptic transformer assembly to map the curved focal plane of a fast (f/3) Rowland-circle spectrograph into a rastered format on the rectangular sensor area of a two-dimensional CCD camera. Vertical grouping of pixels during CCD readout integrates the signal intensity over the height of each spectral segment in the rastered image, minimizing readout time. For the full visible spectrum, readout time is 50 ms. Faster response time (< 10 ms) may be obtained by selecting for readout just a small number of the twenty spectral segments in the image on the CCD. Simultaneous recording of low charge states of carbon, oxygen and injected impurities has yielded information about gas recycling and impurity behavior at the divertor strike points. Transport of lithium to the divertor region during lithium pellet injection has been studied, as well as cumulative deposition of lithium on the divertor targets from pellet injection over many successive discharges.

  13. Mechanical Design of the NSTX Liquid Lithium Divertor

    SciTech Connect

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

    2009-02-19

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

  14. Snowflake Divertor Configuration Studies in DIII-D Tokamak

    NASA Astrophysics Data System (ADS)

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

    2013-10-01

    Experiments in DIII-D show the snowflake divertor (SFD) configuration is compatible with high performance operation (H98 y 2 >= 1) and results in greatly reduced divertor heat flux between and during edge localized modes (ELMs). The SFD was sustained for many energy confinement times using the standard poloidal field shaping coils in 3-5 MW neutral beam injection-heated discharges. Pedestal and divertor effects resulting from a large region of reduced poloidal magnetic field in the SFD are measured and studied using the 2D multi-fluid code UEDGE. The pedestal pressure appeared to be unchanged, while the energy loss per ELM was reduced by 50%. Partial detachment of the SFD was observed at higher ne, with an expanded divertor radiation zone and peak ELM heat flux reduced by up to 80%. Work supported by the US Department of Energy under DE-AC52-07NA27344, DE-AC02-09CH11466, DE-FC02-04ER54698, DE-FG02-07ER54917, and DE-AC04-94AL85000.

  15. 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.

  16. 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

  17. Iterated fractional Tikhonov regularization

    NASA Astrophysics Data System (ADS)

    Bianchi, Davide; Buccini, Alessandro; Donatelli, Marco; Serra-Capizzano, Stefano

    2015-05-01

    Fractional Tikhonov regularization methods have been recently proposed to reduce the oversmoothing property of the Tikhonov regularization in standard form, in order to preserve the details of the approximated solution. Their regularization and convergence properties have been previously investigated showing that they are of optimal order. This paper provides saturation and converse results on their convergence rates. Using the same iterative refinement strategy of iterated Tikhonov regularization, new iterated fractional Tikhonov regularization methods are introduced. We show that these iterated methods are of optimal order and overcome the previous saturation results. Furthermore, nonstationary iterated fractional Tikhonov regularization methods are investigated, establishing their convergence rate under general conditions on the iteration parameters. Numerical results confirm the effectiveness of the proposed regularization iterations.

  18. A method for the inhibition of carbon-film formation in hidden areas of the divertor for tritium inventory control in tokamak

    NASA Astrophysics Data System (ADS)

    Vassallo, E.; Cremona, A.; Laguardia, L.; Grosso, G.

    2011-03-01

    Carbon fibre composites (CFCs) planned to be used in the International Thermonuclear Experimental Reactor (ITER) divertor can be eroded due to hydrogen species from the plasma and the resulting hydrocarbons can be redeposited in other locations. Tritium retention in redeposited materials is the major concern due to the limits imposed for safety reasons by the nuclear licensing authorities. The scavenging effect has already been proposed to decrease the carbon redeposited materials using nitrogen as a scavenger. To date, only fairly limited data are available on the use of NH3 as a scavenger. In this brief communication, the possibility of performing the inhibition of carbon-film formation using NH3 was examined.

  19. Three-dimensional simulation of H-mode plasmas with localized divertor impurity injection on Alcator C-Mod using the edge transport code EMC3-EIRENE

    DOE PAGES

    Lore, Jeremy D.; Reinke, M. L.; Brunner, D.; ...

    2015-04-28

    We study experiments in Alcator C-Mod to assess the level of toroidal asymmetry in divertor conditions resulting from poloidally and toroidally localized extrinsic impurity gas seeding show a weak toroidal peaking (~1.1) in divertor electron temperatures for high-power enhanced D-alpha H-modeplasmas. This is in contrast to similar experiments in Ohmically heated L-modeplasmas, which showed a clear toroidal modulation in the divertor electron temperature. Modeling of these experiments using the 3D edge transport code EMC3-EIRENE [Y. Feng et al., J. Nucl. Mater. 241, 930 (1997)] qualitatively reproduces these trends, and indicates that the different response in the simulations is due tomore » the ionization location of the injected nitrogen. Low electron temperatures in the private flux region (PFR) in L-mode result in a PFR plasma that is nearly transparent to neutral nitrogen, while in H-mode the impurities are ionized in close proximity to the injection location, with this latter case yielding a largely axisymmetric radiation pattern in the scrape-off-layer. In conclusion, the consequences for the ITER gas injection system are discussed. Quantitative agreement with the experiment is lacking in some areas, suggesting potential areas for improving the physics model in EMC3-EIRENE.« less

  20. Three-dimensional simulation of H-mode plasmas with localized divertor impurity injection on Alcator C-Mod using the edge transport code EMC3-EIRENE

    SciTech Connect

    Lore, Jeremy D.; Reinke, M. L.; Brunner, D.; LaBombard, B. A.; Lipschultz, B.; Terry, J. L.; Pitts, R. A.; Feng, Y.

    2015-04-28

    We study experiments in Alcator C-Mod to assess the level of toroidal asymmetry in divertor conditions resulting from poloidally and toroidally localized extrinsic impurity gas seeding show a weak toroidal peaking (~1.1) in divertor electron temperatures for high-power enhanced D-alpha H-modeplasmas. This is in contrast to similar experiments in Ohmically heated L-modeplasmas, which showed a clear toroidal modulation in the divertor electron temperature. Modeling of these experiments using the 3D edge transport code EMC3-EIRENE [Y. Feng et al., J. Nucl. Mater. 241, 930 (1997)] qualitatively reproduces these trends, and indicates that the different response in the simulations is due to the ionization location of the injected nitrogen. Low electron temperatures in the private flux region (PFR) in L-mode result in a PFR plasma that is nearly transparent to neutral nitrogen, while in H-mode the impurities are ionized in close proximity to the injection location, with this latter case yielding a largely axisymmetric radiation pattern in the scrape-off-layer. In conclusion, the consequences for the ITER gas injection system are discussed. Quantitative agreement with the experiment is lacking in some areas, suggesting potential areas for improving the physics model in EMC3-EIRENE.

  1. A new visible spectroscopy diagnostic for the JET ITER-like wall main chambera)

    NASA Astrophysics Data System (ADS)

    Maggi, C. F.; Brezinsek, S.; Stamp, M. F.; Griph, S.; Heesterman, P.; Hogben, C.; Horton, A.; Meigs, A.; Morlock, C.; Studholme, W.; Zastrow, K.-D.; JET-EFDA Contributors

    2012-10-01

    In preparation for ITER, JET has been upgraded with a new ITER-like wall (ILW), whereby the main plasma facing components, previously of carbon, have been replaced by mainly Be in the main chamber and W in the divertor. As part of the many diagnostic enhancements, a new, survey, visible spectroscopy diagnostic has been installed for the characterization of the ILW. An array of eight lines-of-sight (LOS) view radially one of the two JET neutral beam shine through areas (W coated carbon fibre composite tiles) at the inner wall. In addition, one vertical LOS views the solid W tile at the outer divertor. The light emitted from the plasma is coupled to a series of compact overview spectrometers, with overall wavelength range of 380-960 nm and to one high resolution Echelle overview spectrometer covering the wavelength range 365-720 nm. The new survey diagnostic has been absolutely calibrated in situ by means of a radiometric light source placed inside the JET vessel in front of the whole optical path and operated by remote handling. The diagnostic is operated in every JET discharge, routinely monitoring photon fluxes from intrinsic and extrinsic impurities (e.g., Be, C, W, N, and Ne), molecules (e.g., BeD, D2, ND) and main chamber and divertor recycling (typically Dα, Dβ, and Dγ). The paper presents a technical description of the diagnostic and first measurements during JET discharges.

  2. DiMES PMI research at DIII-D in support of ITER and beyond

    DOE PAGES

    Rudakov, Dimitry L.; Abrams, Tyler; Ding, Rui; ...

    2017-03-27

    An overview of recent Plasma-Material Interactions (PMI) research at the DIII-D tokamak using the Divertor Material Evaluation System (DiMES) is presented. The DiMES manipulator allows for exposure of material samples in the lower divertor of DIII-D under well-diagnosed ITER-relevant plasma conditions. Plasma parameters during the exposures are characterized by an extensive diagnostic suite including a number of spectroscopic diagnostics, Langmuir probes, IR imaging, and Divertor Thomson Scattering. Post-mortem measurements of net erosion/deposition on the samples are done by Ion Beam Analysis, and results are modelled by the ERO and REDEP/WBC codes with plasma background reproduced by OEDGE/DIVIMP modelling based onmore » experimental inputs. This article highlights experiments studying sputtering erosion, re-deposition and migration of high-Z elements, mostly tungsten and molybdenum, as well as some alternative materials. Results are generally encouraging for use of high-Z PFCs in ITER and beyond, showing high redeposition and reduced net sputter erosion. Two methods of high-Z PFC surface erosion control, with (i) external electrical biasing and (ii) local gas injection, are also discussed. Furthermore, these techniques may find applications in the future devices.« less

  3. A new visible spectroscopy diagnostic for the JET ITER-like wall main chamber.

    PubMed

    Maggi, C F; Brezinsek, S; Stamp, M F; Griph, S; Heesterman, P; Hogben, C; Horton, A; Meigs, A; Morlock, C; Studholme, W; Zastrow, K-D

    2012-10-01

    In preparation for ITER, JET has been upgraded with a new ITER-like wall (ILW), whereby the main plasma facing components, previously of carbon, have been replaced by mainly Be in the main chamber and W in the divertor. As part of the many diagnostic enhancements, a new, survey, visible spectroscopy diagnostic has been installed for the characterization of the ILW. An array of eight lines-of-sight (LOS) view radially one of the two JET neutral beam shine through areas (W coated carbon fibre composite tiles) at the inner wall. In addition, one vertical LOS views the solid W tile at the outer divertor. The light emitted from the plasma is coupled to a series of compact overview spectrometers, with overall wavelength range of 380-960 nm and to one high resolution Echelle overview spectrometer covering the wavelength range 365-720 nm. The new survey diagnostic has been absolutely calibrated in situ by means of a radiometric light source placed inside the JET vessel in front of the whole optical path and operated by remote handling. The diagnostic is operated in every JET discharge, routinely monitoring photon fluxes from intrinsic and extrinsic impurities (e.g., Be, C, W, N, and Ne), molecules (e.g., BeD, D(2), ND) and main chamber and divertor recycling (typically Dα, Dβ, and Dγ). The paper presents a technical description of the diagnostic and first measurements during JET discharges.

  4. Direct ion orbit loss near the plasma edge of a divertor tokamak in the presence of a radial electric field

    NASA Astrophysics Data System (ADS)

    Miyamoto, K.

    1996-07-01

    The loss region in the initial velocity space of the direct orbit loss ions near the plasma edge of tokamaks with the divertor configuration is studied analytically. The results of this analysis are compared with the numerical results of the loss region in the JET case obtained by Chankin and McCracken (1993). The results agree with each other semiquantitatively in several cases involving the presence of a radial electric field. A measure of the direct ion orbit loss Gamma is calculated from the given loss region in the initial velocity space for JET, JT-60U and ITER. When the initial position of an ion is located in the outside torus (r>Rx, where Rx is the radius at the null X point), the dependence of Gamma on the radial electric field shows the existence of a local maximum and a local minimum in the negative region of the radial electric field

  5. Developing snowflake divertor physics basis in the DIII-D, NSTX and NSTX-U tokamaks aimed at the divertor power exhaust solution [Snowflake divertor experiments in the DIII-D, NSTX and NSTX-U tokamaks aimed at the development of the divertor power exhaust solution

    SciTech Connect

    Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; Lasnier, C. J.; Makowski, M. A.; McLean, A. G.; Meier, E. T.; Meyer, W. H.; Rognlien, T. D.; Ryutov, D. D.; Scotti, F.; Kolemen, E.; Bell, R. E.; Diallo, A.; Gerhardt, S.; Kaita, R.; Kaye, S.; LeBlanc, B. P.; Maingi, R.; Menard, J. E.; Podesta, M.; Roquemore, A. L.; Groebner, R. J.; Hyatt, A. W.; Leonard, A. W.; Osborne, T. H.; Petrie, T. W.; Ahn, J. -W.; Raman, R.; Watkins, J. G.

    2016-06-02

    Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment were performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large ELMs. However, a stable partial detachment of the outer strike point was also achieved where inter-ELM peak heat flux was reduced by factors 3-5, and peak ELM heat flux was reduced by up to 80% (cf. standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower ne, and ELM energy and divertor peak heat flux reduction, especially prominent in radiative D2-seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D2-seeded SF divertor at PSOL = 3 - 4 MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multi-fluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected PSOL ≃9 MW case. In conclusion, the radiative SF divertor with carbon impurity provides a wider ne operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar qpeak reduction factors (cf. standard divertor).

  6. European Technological Effort in Preparation of ITER Construction

    SciTech Connect

    Andreani, Roberto

    2005-04-15

    Europe has started since the '80s with the preparatory work done on NET, the Next European Torus, the successor of JET, to prepare for the construction of the next generation experiment on the road to the fusion reactor. In 2000 the European Fusion Development Agreement (EFDA) has been signed by sixteen countries, including Switzerland, not a member of the Union. Now the signatory countries have increased to twenty-five. A vigorous programme of design and R and D in support of ITER construction has been conducted by EFDA through the coordinated effort of the national institutes and laboratories supported financially, in the framework of the VI European Framework Research Programme (2002-2006), by contracts of association with EURATOM. In the last three years, with the expenditure of 160 M[Euro], the accent has been particularly put on the preparation of the industrial manufacturing activities of components and systems for ITER. Prototypes and manufacturing methods have been developed in all the main critical areas of machine construction with the objective of providing sound and effective solutions: vacuum vessel, toroidal field coils, poloidal field coils, remote handling equipment, plasma facing components and divertor components, electrical power supplies, generators and power supplies for the Heating and Current Drive Systems and other minor subsystems.Europe feels to be ready to host the ITER site and to provide adequate support and guidance for the success of construction to our partners in the ITER collaboration, wherever needed.

  7. 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.

  8. Engineering, installation, testing, and initial operation of the DIII-D Advanced Divertor

    SciTech Connect

    Andersen, P.M.; Baxi, C.B.; Reis, E.E.; Schaffer, M.J.; Smith, J.P.

    1990-09-01

    The Advanced Divertor (AD) for General Atomics tokamak, DIII-D, was installed in the summer of 1990. The AD has enabled two classes of physics experiments to be run: divertor biasing and divertor baffling. Both are new experiments for DIII-D. The AD has two principal components: (1) a continuous ring electrode; and (2) a toroidally symmetric baffle. The tokamak can be run in bias baffle or standard DIII-D divertor modes by accurate positioning of the outer divertor strike point through the use of the DIII-D control system. The paper covers design, analysis, fabrication, installation, instrumentation, testing, initial operation, and future plans for the Advanced Divertor from an engineering viewpoint. 2 refs., 5 figs.

  9. Preliminary study of divertor particle exhaust in the EAST superconducting tokamak

    NASA Astrophysics Data System (ADS)

    Liu, Huan; Wang, Liang; Xu, Guosheng; Ding, Fang; Liu, Jianbin; Xu, Jichan; Feng, Wei; Deng, Guozhong; Zheng, Xingwei; Yu, Yaowei; Si, Hang; Liu, Haiqing; Yang, Qingquan; Sun, Zhen; Guo, Houyang

    2017-09-01

    The particle exhaust of the upper tungsten and lower carbon divertors in EAST has been preliminarily studied during the 2016 experimental campaign. The density decay time during terminating gas puffing has been employed as a key parameter to evaluate the divertor particle exhaust performance. Comparative plasma discharges have been carried out on the particle exhaust performance between two toroidal field directions in the upper single null and lower single null divertor configurations. This work has enhanced the understanding of the effects of the in-out asymmetry and divertor geometry on the efficiency of the divertor particle exhaust. In addition, the sensitivity of the particle exhaust capability on different strike point locations has been analyzed. The experimental results are expected to provide important information on the future upgrade of EAST bottom divertor and facilitate the realization of longer pulse operation.

  10. Sensitivity analysis of upstream plasma condition for SST-1 X-Divertor configuration with SOLPS

    NASA Astrophysics Data System (ADS)

    Himabindu, M.; Tyagi, Anil K.; Sharma, Deepti; Sharma, Devendra; Srinivasan, R.

    2017-04-01

    Extensive power exhausts and target heat loads are anticipated in reactor grade fusion devices. Prototyping of an X-Divertor based power exhaust scheme is being attempted by means of simulations of Scrape-off Layer plasma transport in the diverted plasma equilibria of SST-1 tokamak using SOLPS5.1. Evaluation of the relative advantages of an X-Divertor configuration involves simulating the SST-1 standard divertor scheme plasma transport for the reference and then achieving equivalent upstream plasma conditions in the X-divertor equilibrium to ensure equivalent core plasma in both the cases. The first optimization is to be achieved by simulating effects of an external gas puff in the SOL region for controlling separatrix density in the X-divertor configuration with visible modifications in the downstream plasma conditions. The present work analyzes sensitivity of the upstream SOL plasma conditions to the gas puff intensity and its effect on the plasma neutral transport in the divertor region

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

    NASA Astrophysics Data System (ADS)

    Takizuka, T.

    2017-03-01

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

  12. Impact of potential narrow SOL heat flux on H-mode access in ITER

    NASA Astrophysics Data System (ADS)

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

    2013-12-01

    The paper presents results of a first analysis of the divertor performance during the L-H transition in ITER. The integrated model consists of the SOLPS4.3 code suite for the SOL and divertor, and the ASTRA code for the core and pedestal regions. The results of SOLPS4.3 are parametrized and used as the boundary conditions for ASTRA, ensuring a consistent description of the plasma core and the edge. Boundary conditions switch from those for wide (L-mode) to narrow (H-mode) SOL once the transition criterion is met. The results show that, for conditions for which a full-power operational space with acceptable power loading of the targets exists, a transition from the initial L-mode operation to H-mode can be found for the same assumptions, i.e. the full-power H-mode regime is accessible.

  13. Upgrade of the infrared camera diagnostics for the JET ITER-like wall divertora)

    NASA Astrophysics Data System (ADS)

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

    2012-10-01

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

  14. Progress in the development of deposition prevention and cleaning techniques of in-vessel optics in ITER

    NASA Astrophysics Data System (ADS)

    Mukhin, E.; Vukolov, K.; Semenov, V.; Tolstyakov, S.; Kochergin, M.; Kurskiev, G.; Podushnikova, K.; Razdobarin, A.; Gorodetsky, A.; Zalavutdinov, R.; Bukhovets, V.; Zakharov, A.; Bulovich, S.; Veiko, V.; Shakshno, E.

    2009-08-01

    The lifetime of front optical components unprotected from reactor grade plasmas may be very short due to intensive contamination with carbon and beryllium-based materials eroded by the plasma from beryllium walls and carbon tiles. Deposits result in a significant reduction and spectral alterations of optical transmission. In addition, even rather thin and transparent deposits can dramatically change the shape of reflectance spectra, especially for mirrors with rather low reflectivity, such as W or Mo. The distortion of data obtained with various optical diagnostics may affect the safe operation of ITER. Therefore, the development of optics-cleaning and deposition-mitigating techniques is a key factor in the construction and operation of optical diagnostics in ITER. The problem is of particular concern for optical elements positioned in the divertor region. The latest achievements in protection of in-vessel optics are presented using the example of deposition prevention/cleaning techniques for in-machine components of the Thomson scattering system in the divertor. Careful consideration of well-known and novel protection approaches shows that neither of them alone provides guaranteed survivability of the first in-vessel optics in the divertor. Only a set of complementary prevention/cleaning techniques, which include special materials for mirrors and inhibition additives for plasma, is able to manage the challenging task. The essential issue, which needs to be addressed in the immediate future, is an extensive development of techniques tested under experimental conditions (exposure time and contamination fluxes) similar to those expected in ITER.

  15. Tokamak power exhaust with the snowflake divertor: Present results and outstanding issues

    SciTech Connect

    Soukhanovskii, V. A.; Xu, X.

    2015-09-15

    Here, a snowflake divertor magnetic configuration (Ryutov in Phys Plasmas 14(6):064502, 2007) with the second-order poloidal field null offers a number of possible advantages for tokamak plasma heat and particle exhaust in comparison with the standard poloidal divertor with the first-order null. Results from snowflake divertor experiments are briefly reviewed and future directions for research in this area are outlined.

  16. An automated approach to magnetic divertor configuration design

    NASA Astrophysics Data System (ADS)

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

    2015-01-01

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

  17. Modelling of Divertor Plasma Transport in Stochastic Magnetic Boundary

    SciTech Connect

    Kobayashi, Masahiro

    2010-05-20

    Impacts of stochastic magnetic field structure on divertor functions are discussed based on analyses with the three dimensional (3D) edge transport code package EMC3-EIRENE with Braginskii type fluid equations, in the Large Helical Device (LHD), in comparison with the experimental data. It is shown that the three dimensional field line topology introduced by the stochasticity provides controllability of the edge plasma transport such as divertor regime, impurity transport. The observations in other devices with stochastic magnetic boundary regarding these issues are discussed as well. Also presented are the traditional formulation of the magnetic field and the transport in the stochastic layer based on diffusive picture, which are contrasted with the 3D treatment of the flux tube topology and of the transport.

  18. Diagnosing transient plasma status: from solar atmosphere to tokamak divertor

    NASA Astrophysics Data System (ADS)

    Giunta, A. S.; Henderson, S.; O'Mullane, M.; Harrison, J.; Doyle, J. G.; Summers, H. P.

    2016-09-01

    This work strongly exploits the interdisciplinary links between astrophysical (such as the solar upper atmosphere) and laboratory plasmas (such as tokamak devices) by sharing the development of a common modelling for time-dependent ionisation. This is applied to the interpretation of solar flare data observed by the UVSP (Ultraviolet Spectrometer and Polarimeter), on-board the Solar Maximum Mission and the IRIS (Interface Region Imaging Spectrograph), and also to data from B2-SOLPS (Scrape Off Layer Plasma Simulations) for MAST (Mega Ampère Spherical Tokamak) Super-X divertor upgrade. The derived atomic data, calculated in the framework of the ADAS (Atomic Data and Analysis Structure) project, allow equivalent prediction in non-stationary transport regimes and transients of both the solar atmosphere and tokamak divertors, except that the tokamak evolution is about one thousand times faster.

  19. Modelling of Divertor Plasma Transport in Stochastic Magnetic Boundary

    NASA Astrophysics Data System (ADS)

    Kobayashi, Masahiro

    2010-05-01

    Impacts of stochastic magnetic field structure on divertor functions are discussed based on analyses with the three dimensional (3D) edge transport code package EMC3-EIRENE with Braginskii type fluid equations, in the Large Helical Device (LHD), in comparison with the experimental data. It is shown that the three dimensional field line topology introduced by the stochasticity provides controllability of the edge plasma transport such as divertor regime, impurity transport. The observations in other devices with stochastic magnetic boundary regarding these issues are discussed as well. Also presented are the traditional formulation of the magnetic field and the transport in the stochastic layer based on diffusive picture, which are contrasted with the 3D treatment of the flux tube topology and of the transport.

  20. Edge exposure of poloidal divertor target plate tiles

    SciTech Connect

    Mohanti, R.B.; Gilligan, J.G.; Bourham, M.A.

    1996-12-01

    Exposure to near normal surfaces of poloidal divertor target plate tiles is a limiting feature of the power handling capability of the tiles. The problems associated with the design of poloidal divertor tiles, with beryllium chosen as the tile material, and possible methods of solving the problem are discussed. Thermal two- and three-dimensional analyses are carried out for the assessment of relative merits in performance due to modifications to the surface. The power handling capability (time to reach melting temperature of beryllium) of the target plate tiles is presented for unswept and swept plasma cases. Results have shown that sweeping the plasma improves the power handling capability by a factor of up to 10. 20 refs., 7 figs., 3 tabs.

  1. Characterizing the Outer Divertor Leg Transition to Full Detachment

    NASA Astrophysics Data System (ADS)

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

    2013-10-01

    Experiments at DIII-D have explored the transition from an attached to fully detached divertor condition in L- and H-mode with an unprecedented level of detail. Improved divertor Thomson scattering capturing Te <= 1 eV, coupled with high resolution spectroscopic studies of molecular and neutral emissions, and Stark broadening of the deuterium Paschen series provide essential data for modeling the transition to detachment. 2D Te and ne profiles of the outer leg reveal movement of the ionization front away from the plate not replicated in modeling. Measured Paschen and molecular emissions suggest the onset of recombination occurs prior to, and to a greater extent than modeled. These data help guide and expose any missing physics in predictions for detached operation in future devices. This work supported in part by the US Department of Energy under DE-AC52-07NA27344 and DE-FC02-04ER54698.

  2. Preparation of the liquid lithium divertor plates for NSTX

    NASA Astrophysics Data System (ADS)

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

    2011-10-01

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

  3. ICRF Specific Plasma Wall Interactions in JET with the ITER-Like Wall

    SciTech Connect

    Bobkov, V.; Arnoux, G.; Brezinsek, S.; Coenen, J. W.; Colas, L.; Clever, M.; Czarnecka, A.; Braun, F.; Dux, R.; Huber, Alexander; Lerche, E.; Maggi, C.; Marcotte, F.; Maslov, M.; Matthews, G.; Mayoral, M.-L.; Meigs, A. G.; Monakhov, I.; Putterich, Th.; Rimini, F.; Rooj, G. Van; Sergienko, G.; Van Eester, D.

    2013-01-01

    A variety of plasma wall interactions (PWIs) during operation of the so-called A2 ICRF antennas is observed in JET with the ITER-like wall. Amongst effects of the PWIs, the W content increase is the most significant, especially at low plasma densities. No increase of W source from the main divertor and entrance of the outer divertor during ICRF compared to NBI phases was found by means of spectroscopic and WI (400.9 nm) imaging diagnostics. In contrary, the W flux there is higher during NBI. Charge exchange neutrals of hydrogen isotopes could be excluded as considerable contributors to the W source. The high W content in ICRF heated limiter discharges suggests the possibility of other W sources than the divertor alone. Dependencies of PWIs to individual ICRF antennas during q95-scans, and intensification of those for the 90 phasing, indicate a link between the PWIs and the antenna near-fields. The PWIs include heat loads and Be sputtering pattern on antenna limiters. Indications of some PWIs at the outer divertor entrance are observed which do not result in higher W flux compared to the NBI phases, but are characterized by small antenna-specific (up to 25% with respect to ohmic phases) bipolar variations of WI emission. The first TOPICA calculations show a particularity of the A2 antennas compared to the ITER antenna, due to the presence of long antenna limiters in the RF image current loop and thus high near-fields across the most part of the JET outer wall.

  4. Thermal and structural analysis of the TPX divertor

    SciTech Connect

    Reis, E.E.; Baxi, C.B.; Chin, E.; Redler, K.M.

    1995-12-31

    The high heat flux on the surfaces of the TPX divertor will require a design in which a carbon-carbon (C-C) tile material is brazed to water cooled copper tubes. Thermal and structural analyses were performed to assist in the design selection of a divertor tile concept and C-C material. The relevancy of finite element analysis (FEA) for evaluating tile design was examined by conducting a literature survey to compare FEA stress results to subsequent brazing and thermal test results. The thermal responses for five tile concepts and four C-C materials were analyzed for a steady-state heat flux of 7.5 MW/m{sup 2}. Elastic-plastic stress analyses were performed to calculate the residual stresses due to brazing C-C tiles to soft copper heat sinks for the various tile designs. Monoblock and archblock divertor tile concepts were analyzed for residual stresses in which elevated temperature creep effects were included with the elastic-plastic behavior of the copper heat sink for an assumed braze cooldown cycle. As a result of these 2D studies, the archblock concept with a 3D fine weave C-C was initially found to be a preferred design for the divertor. A 3D elastic-plastic analysis for brazing of the arch block tile was performed to investigate the singularity effects at the C-C to copper interface in the direction of the tube axis. This analysis showed that the large residual stresses at the tube and tile edge intersection would produce cracks in the C-C and possible delamination along the braze interface. These results, coupled with the difficulties experienced in brazing archblocks for the Tore Supra Limiter, required that other tile designs be considered.

  5. Iteration, Not Induction

    ERIC Educational Resources Information Center

    Dobbs, David E.

    2009-01-01

    The main purpose of this note is to present and justify proof via iteration as an intuitive, creative and empowering method that is often available and preferable as an alternative to proofs via either mathematical induction or the well-ordering principle. The method of iteration depends only on the fact that any strictly decreasing sequence of…

  6. Iteration, Not Induction

    ERIC Educational Resources Information Center

    Dobbs, David E.

    2009-01-01

    The main purpose of this note is to present and justify proof via iteration as an intuitive, creative and empowering method that is often available and preferable as an alternative to proofs via either mathematical induction or the well-ordering principle. The method of iteration depends only on the fact that any strictly decreasing sequence of…

  7. Experimental investigation of the natural divertor configuration in Heliotron-E

    SciTech Connect

    Hillis, D.L.; Mioduszewski, P.K.; Fowler, R.H.; Rome, J.A.; Motojima, O.; Mizuuchi, T.; Noda, N.; Mutoh, T.; Zushi, H.; Takahashi, R.; Obiki, T.; Iiyoshi, A.; Uo, K.

    1988-01-01

    Particle control with pump limiters and divertors has been successfully demonstrated in a number of present-day tokamaks. In a heliotron/stellarator configuration, plasma flows to the wall in distinct flux bundles, often called ''divertor stripes''. This complicated three-dimensional characteristic of the plasma edge presents a new challenge for active particle control systems such as pump limiters and divertors. The experiment described here has obtained data with an instrumented pump particle collector that is located in the ''natural'' magnetic divertor stripe of Heliotron-E. The particle collector consists of a moveable graphite assembly with single-sided particle collection and active pumping. By scanning the particle collector assembly through the plasma edge of Heliotron-E, the divertor stripe is observed to be about 2-3 cm (FWHM) in width, and pressure rises of 0.01-0.01 mTorr are observed in the particle collector pumping chamber. These measurements have demonstrated that particles leaving the bulk plasma via the divertor stripes can be collected and provide a basis for developing a divertor scheme for particle control in helical systems. Modelling of the Heliotron-E magnetic configuration at the plasma edge is used to determine the collection efficiency of the particle collector in the divertor stripes. The modeling is further extended to describe a helical divertor concept. 18 refs., 6 figs.

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

    NASA Astrophysics Data System (ADS)

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

    2017-07-01

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

  9. Modelling of radiative divertor operation towards detachment in experimental advanced superconducting tokamak

    NASA Astrophysics Data System (ADS)

    Chen, YiPing; Wang, F. Q.; Zha, X. J.; Hu, L. Q.; Guo, H. Y.; Wu, Z. W.; Zhang, X. D.; Wan, B. N.; Li, J. G.

    2013-02-01

    In order to actively control power load on the divertor target plates and study the effect of radiative divertor on plasma parameters in divertor plasmas and heat fluxes to the targets, dedicated experiments with Ar impurity seeding have been performed on experimental advanced superconducting tokamak in typical L-mode discharge with single null divertor configuration, ohmic heating power of 0.5 MW, and lower hybrid wave heating power of 1.0 MW. Ar is puffed into the divertor plasma at the outer target plate near the separatrix strike point with the puffing rate 1.26×1020 s-1. The radiative divertor is formed during the Ar puffing. The SOL/divertor plasma in the L-mode discharge with radiative divertor has been modelled by using SOLPS5.2 code package [V. Rozhansky et al., Nucl. Fusion 49, 025007 (2009)]. The modelling shows the cooling of the divertor plasma due to Ar seeding and is compared with the experimental measurement. The changes of peak electron temperature and heat fluxes at the targets with the shot time from the modelling results are similar to the experimental measurement before and during the Ar impurity seeding, but there is a major difference in time scales when Ar affects the plasma in between experiment and modelling.

  10. Measuring the effect of divertor closure on detachment in DIII-D

    NASA Astrophysics Data System (ADS)

    Moser, Auna; Leonard, A. W.; Petrie, T. W.; Sang, C. F.; Allen, S. L.; McLean, A. G.; Fenstermacher, M. E.; Joseph, I.; Lasnier, C. J.; Makowski, M. A.; Watkins, J. G.; Briesemeister, A. R.

    2015-11-01

    Recent experiments compared the open lower divertor and semi-closed upper divertor in DIII-D to measure the effect of divertor closure on detachment onset and heat flux control, extending past work showing reduced core fueling with the more-closed upper DIII-D divertor. Experiments were performed to determine the extent to which closure may facilitate detachment at collisionalities more relevant to future devices. This work builds on previous experiments that quantified effects of divertor magnetic geometry, including connection length, ∇B-drift direction, incidence angle, and flux expansion; efforts were made to match these parameters while comparing single null configurations in the upper and lower divertor in order to isolate the effects of closure. Experimental measurements coupled with simulation results will help weigh the benefits of a more-closed divertor in facilitating detachment and reducing heat flux against the constraints imposed on the magnetic geometry by a more-closed divertor tile structure, aiding in the design of a future advanced divertor for DIII-D. Supported by the US Department of Energy under DE-FC02-04ER54698, DE-AC52-07NA27344, DE-AC04-94AL85000, and DE-AC05-00OR22725.

  11. Effect of divertor closure and impurities on detachment onset in DIII-D

    NASA Astrophysics Data System (ADS)

    Moser, A. L.; Leonard, A. W.; Groebner, R. J.; Petrie, T. W.; Sang, C. F.; Wang, H.; Allen, S. L.; McLean, A. G.; Fenstermacher, M. E.; Lasnier, C. J.; Makowski, M.; Watkins, J. G.; Briesemeister, A. R.

    2016-10-01

    Heat flux control in future devices requires a detached divertor with upstream parameters compatible with core performance, e.g., at a lower upstream density than presently achievable. Comparison between matched H-mode discharges in the upper and lower divertors of DIII-D demonstrates onset of detachment at a reduced pedestal density for the more-closed geometry of the upper divertor. The upper divertor also produces a lower pedestal density with a less-steep profile than the lower divertor for matched discharges with no additional fueling, presumably due to a reduction in ionization source for the upper divertor. Recent experiments further compare the upper and lower divertors with the addition of impurities injected into the private flux region. These experiments measure the interplay between increased closure and radiating impurities and the effect on divertor detachment, as well as the ability of the more-closed divertor geometry to prevent the accumulation of impurities in the core. Work supported by US DOE under DE-FC02-04ER54698, DE-AC52-07NA27344, DE-AC04-94AL85000, DE-AC05-00OR22725.

  12. Assessment of issues for the MAST divertor biasing experiment

    NASA Astrophysics Data System (ADS)

    Helander, P.; Cohen, R. H.; Fielding, S.; Ryutov, D.

    2001-10-01

    A biasing experiment is being undertaken in the MAST scrape-off layer; the goal is to induce intense convection by a toroidally alternating biasing of divertor tiles. This would lead to a thickening of the SOL and a reduction of the heat load on the divertor plates. In addition, by studying the reaction of a plasma to a varying bias, one can collect new information regarding pre-existing SOL turbulence. We consider the following issues: 1. The bias amplitude required to produce significant SOL broadening; 2. Excitation of shear-flow turbulence in convective cells; 3. The role of magnetic shear; 4. Effects of electrostatic sheaths at the divertor plates; 5. Redistribution of heat fluxes during biasing. We show that a significant effect of the biasing on the SOL structure can be reached at relatively small bias voltages 30 V. We also show that the potential perturbations will be limited to a zone between the X-point and the biased tiles, and will be essentially decoupled from the main SOL plasma. Preliminary experimental results may be shown.

  13. Modeling of Divertor Plates in the Compact Toroidal Hybrid

    NASA Astrophysics Data System (ADS)

    Hartwell, G. J.; Small, C. M.; Ennis, D. A.; Hanson, J. D.; Knowlton, S. F.; Maurer, D. A.

    2014-10-01

    In long pulse length stellarator experiments, edge island divertors can be used as a method of plasma particle and heat exhaust. Knowledge of the detailed power loading on these structures and its relationship to the long connection length scrape off layer physics is a new Compact Toroidal Hybrid research thrust. We report the results of connection length studies for divertor plates to be installed in the Compact Toroidal Hybrid (CTH), a five field period torsatron with R0 = 0 . 75 m, ap ~ 0 . 2 m, and B <= 0 . 7 T. For these studies, CTH will be operated as a pure stellarator with no ohmically generated plasma current. The CTH edge rotational transform can be varied from tvac (a) = 0.02-0.35 by adjusting the ratio of currents in the helical and toroidal field coils. A poloidal field coil is used to adjust the shear of the rotational transform profile, and hence the size of edge islands, while the phase of the island is rotated with a set of five error coils producing an n = 1 perturbation. For the studies conducted, a magnetic configuration with a large n = 1, m = 3 magnetic island at the edge is generated. Results from multiple possible divertor plate locations relative to the island structure will be presented. This work is supported by U.S. Department of Energy Grant No. DE-FG02-00ER54610.

  14. Fast reciprocating Langmuir probe for the DIII-D divertor

    NASA Astrophysics Data System (ADS)

    Watkins, J. G.; Hunter, J.; Tafoya, B.; Ulrickson, M.; Watson, R. D.; Moyer, R. A.; Cuthbertson, J. W.; Gunner, G.; Lehmer, R.; Luong, P.; Hill, D. N.; Mascaro, M.; Robinson, J. I.; Snider, R.; Stambaugh, R.

    1997-01-01

    A new reciprocating Langmuir probe was used to measure density and temperature profiles, ion flow, and potential fluctuation levels from the lower divertor floor up to the X point on the DIII-D Tokamak. This probe is designed to make fast (2 kHz swept, 20 kHz Mach, 500 kHz Vfloat) measurements with 2 mm spatial resolution in the region where the largest gradients on the plasma open flux tubes are found and therefore provide the best benchmarks for scrap-off layer and divertor numerical models. Profiles are constructed using the 300 ms time history of the probe measurements during the 25 cm reciprocating stroke. Both single and double null plasmas can be measured and compared with a 20 Hz divertor Thomson scattering system. The probe head is constructed of four different kinds of graphite to optimize the electrical and thermal characteristics. Electrically insulated pyrolytic graphite rings act as a heat shield to absorb the plasma heat flux on the probe shaft and are mounted on a carbon/carbon composite core for mechanical strength. The Langmuir probe sampling tips are made of a linear carbon fiber composite. The mechanical, electrical, data acquisition, and power supply systems will be described. Initial measurements will also be presented.

  15. High heat flux performance of brazed tungsten macro-brush test mock-up for divertors

    NASA Astrophysics Data System (ADS)

    Patil, Yashashri; Khirwadkar, S. S.; Krishnan, D.; Patel, A.; Tripathi, S.; Singh, K. P.; Belsare, S. M.

    2013-06-01

    Plasma facing components (PFCs) of divertor will be exposed to steady state and transient heat loads up to 20 MW/m2, during operation of ITER-like plasma fusion device. The critical task in fusion research is to design, fabricate and test of PFCs. To withstand high heat loads, PFCs are designed and fabricated in flat tile, mono-block type geometries using tungsten as plasma facing material and CuCrZr alloy is used as a heat sink. These fabricated mock-ups are tested under thermal cyclic heat loads using intense electron beam in pulsed mode. Tungsten macro-brush type of mock-up has been developed by vacuum furnace brazing route. Mock-up was tested to the absorbed heat flux in the range of 0.5-9 MW/m2. Simulation of high heat flux (HHF) test under steady state and cyclic heat loads has been done using ANSYS12 finite element analysis (FEA) software. HHF tests have been successfully performed on the tungsten mock-up.

  16. 2-D thermal response calculations of the liquid lithium divertor on NSTX*

    NASA Astrophysics Data System (ADS)

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

    2011-10-01

    The liquid lithium divertor (LLD) in NSTX was installed for particle and impurity control in NSTX, and its effectiveness was predicted to vary with the lithium surface temperature. It is therefore important to know the temperature evolution of the LLD during plasma discharges. A 2-D implicit finite difference code (``Li_enthalpy'') was written to simulate the lithium temperature with an accurate description of the LLD components, which include a surface lithium layer, a porous molybdenum mesh that is ~ 50% filled with lithium, a thin stainless steel layer, and a thick underlying copper substrate. The heat flux on the graphite was measured with a recently developed dual-band infrared camera; we use the same heat flux profile on the LLD at the same major radius, because of toroidal symmetry. The code ``Li_enthalpy'' computes the LLD thermal response to this heat flux profile; a Gauss-Seidel iterative procedure was implemented to solve the phase-change problem as lithium melted in response to plasma heating. The computed LLD temperature response is then compared and calibrated with the measured surface temperature on the LLD with the dual-band camera. From this the dynamics of the spatially and time varying liquid lithium layer thickness are extracted. Analysis from a number of plasma discharges is presented. *Supported in part by U.S. DoE contracts DE-AC05-00OR22725 and DE-AC02-09CH11466.

  17. ITER in-vessel system design and performance

    NASA Astrophysics Data System (ADS)

    Parker, R. R.

    2000-03-01

    The article reviews the design and performance of the in-vessel components of ITER as developed for the Engineering Design Activities (EDA) Final Design Report. The double walled vacuum vessel is the first confinement boundary and is designed to maintain its integrity under all normal and off-normal conditions, e.g. the most intense vertical displacement events (VDEs) and seismic events. The shielding blanket consists of modules connected to a toroidal backplate by flexible connectors which allow differential displacements due to temperature non-uniformities. Breeding blanket modules replace the shield modules for the Enhanced Performance Phase. The divertor concept is based on a cassette structure which is convenient for remote installation and removal. High heat flux (HHF) components are mechanically attached and can be removed and replaced in the hot cell. Operation of the divertor is based on achieving partially detached plasma conditions along and near the separatrix. Nominal heat loads of 5-10 MW/m2 are expected on the target. These are accommodated by HHF technology developed during the EDA. Disruptions and VDEs can lead to melting of the first wall armour but no damage to the underlying structure. Stresses in the main structural components remain within allowable ranges for all postulated disruption and seismic events.

  18. A Fully Noninductive, ELM-Suppressed Scenario for ITER

    NASA Astrophysics Data System (ADS)

    Petty, C. C.; Petrie, T. W.; Nazikian, R.; Turco, F.; Lasnier, C.

    2016-10-01

    An attractive regime with beta, collisionality and plasma shape relevant to the ITER steady-state mission has been attained in DIII-D using the hybrid scenario, including complete ELM suppression using resonant magnetic perturbation (RMP) coils. Fully noninductive hybrids with simultaneous high beta (βN <= 3.1) and high confinement (H98 y 2 <= 1.4) have achieved zero surface loop voltage for up to two current relaxation times using efficient central current drive from ECCD and NBCD. This steady-state regime has been successfully integrated with ELM suppression by applying an odd parity n=3 RMP, which has only a minor impact on the pedestal pressure ( 15 %) and H98 y 2 ( 10 %) In radiating divertor experiments in hybrids, the combination of Argon seeding and strong Deuterium puffing more than doubles the plasma radiative power, up to 55% of the input power, with less than 10% increase in Zeff. IR camera measurements find that the peak heat flux in the upper, outer divertor falls by a factor of 2 (from 4.6 to 2.3 MW /m2). Work supported by USDOE under DE-FC02-04ER54698, DE-AC02-09CH11466, DE-FG02-04ER54761, and DE-AC52-07NA27344.

  19. The ITER design

    NASA Astrophysics Data System (ADS)

    Aymar, R.; Barabaschi, P.; Shimomura, Y.

    2002-05-01

    In 1998, after six years of joint work originally foreseen under the ITER engineering design activities (EDA) agreement, a design for ITER had been developed fulfilling all objectives and the cost target adopted by the ITER parties in 1992 at the start of the EDA. While accepting this design, the ITER parties recognized the possibility that they might be unable, for financial reasons, to proceed to the construction of the then foreseen device. The focus of effort in the ITER EDA since 1998 has been the development of a new design to meet revised technical objectives and a cost reduction target of about 50% of the previously accepted cost estimate. The rationale for the choice of parameters of the design has been based largely on system analysis drawing on the design solutions already developed and using the latest physics results and outputs from technology R&D projects. In so doing the joint central team and home teams converge towards a new design which will allow the exploration of a range of burning plasma conditions. The new ITER design, whilst having reduced technical objectives from its predecessor, will nonetheless meet the programmatic objective of providing an integrated demonstration of the scientific and technological feasibility of fusion energy. Background, design features, performance, safety features, and R&D and future perspectives of the ITER design are discussed.

  20. Compatibility of separatrix density scaling for divertor detachment with H-mode pedestal operation in DIII-D

    NASA Astrophysics Data System (ADS)

    Leonard, A. W.; McLean, A. G.; Makowski, M. A.; Stangeby, P. C.

    2017-08-01

    The midplane separatrix density is characterized in response to variations in upstream parallel heat flux density and central density through deuterium gas injection. The midplane density is determined from a high spatial resolution Thomson scattering diagnostic at the midplane with power balance analysis to determine the separatrix location. The heat flux density is varied by scans of three parameters, auxiliary heating, toroidal field with fixed plasma current, and plasma current with fixed safety factor, q 95. The separatrix density just before divertor detachment onset is found to scale consistent with the two-point model when radiative dissipation is taken into account. The ratio of separatrix to pedestal density, n e,sep/n e,ped varies from  ⩽30% to  ⩾60% over the dataset, helping to resolve the conflicting scaling of core plasma density limit and divertor detachment onset. The scaling of the separatrix density at detachment onset is combined with H-mode power threshold scaling to obtain a scaling ratio of minimum n e,sep/n e,ped expected in future devices.

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

    NASA Astrophysics Data System (ADS)

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

    2016-12-01

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

  2. Wall conditioning on ITER

    NASA Astrophysics Data System (ADS)

    Shimada, Michiya; Pitts, Richard A.

    2011-08-01

    Like all tokamaks, ITER will require wall conditioning systems and strategies for successful operation from the point of view of plasma-facing surface preparation. Unlike today's devices however, ITER will have to manage large quantities of tritium fuel, imposing on wall conditioning a major responsibility for tritium inventory control. It will also feature the largest plasma-facing beryllium surface ever used in a tokamak and its high duty cycle and long pulse are expected to lead to the rapid formation of deposited layers in which tritium can accumulate. This paper summarises the currently planned ITER wall conditioning systems and describes the strategy for their use throughout exploitation of the device.

  3. Plasma-surface interaction in the Be/W environment: Conclusions drawn from the JET-ILW for ITER

    NASA Astrophysics Data System (ADS)

    Brezinsek, S.

    2015-08-01

    The JET ITER-Like Wall experiment (JET-ILW) provides an ideal test bed to investigate plasma-surface interaction (PSI) and plasma operation with the ITER plasma-facing material selection employing beryllium in the main chamber and tungsten in the divertor. The main PSI processes: material erosion and migration, (b) fuel recycling and retention, (c) impurity concentration and radiation have be1en studied and compared between JET-C and JET-ILW. The current physics understanding of these key processes in the JET-ILW revealed that both interpretation of previously obtained carbon results (JET-C) and predictions to ITER need to be revisited. The impact of the first-wall material on the plasma was underestimated. Main observations are: (a) low primary erosion source in H-mode plasmas and reduction of the material migration from the main chamber to the divertor (factor 7) as well as within the divertor from plasma-facing to remote areas (factor 30 - 50). The energetic threshold for beryllium sputtering minimises the primary erosion source and inhibits multi-step re-erosion in the divertor. The physical sputtering yield of tungsten is low as 10-5 and determined by beryllium ions. (b) Reduction of the long-term fuel retention (factor 10 - 20) in JET-ILW with respect to JET-C. The remaining retention is caused by implantation and co-deposition with beryllium and residual impurities. Outgassing has gained importance and impacts on the recycling properties of beryllium and tungsten. (c) The low effective plasma charge (Zeff = 1.2) and low radiation capability of beryllium reveal the bare deuterium plasma physics. Moderate nitrogen seeding, reaching Zeff = 1.6 , restores in particular the confinement and the L-H threshold behaviour. ITER-compatible divertor conditions with stable semi-detachment were obtained owing to a higher density limit with ILW. Overall JET demonstrated successful plasma operation in the Be/W material combination and confirms its advantageous PSI behaviour

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

    SciTech Connect

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

    2016-02-15

    The erosion of divertor targets caused by high heat fluxes during transients is a serious threat to ITER operation, as it is going to be the main factor determining the divertor lifetime. Under the influence of extreme heat fluxes, the surface temperature of plasma facing components can reach some certain threshold, leading to an onset of intense material evaporation. The latter results in formation of cold dense vapor and secondary plasma cloud. This layer effectively absorbs the energy of the incident plasma flow, turning it into its own kinetic and internal energy and radiating it. This so called vapor shielding is a phenomenon that may help mitigating the erosion during transient events. In particular, the vapor shielding results in saturation of energy (per unit surface area) accumulated by the target during single pulse of heat load at some level E{sub max}. Matching this value is one of the possible tests to verify complicated numerical codes, developed to calculate the erosion rate during abnormal events in tokamaks. The paper presents three very different models of vapor shielding, demonstrating that E{sub max} depends strongly on the heat pulse duration, thermodynamic properties, and evaporation energy of the irradiated target material. While its dependence on the other shielding details such as radiation capabilities of material and dynamics of the vapor cloud is logarithmically weak. The reason for this is a strong (exponential) dependence of the target material evaporation rate, and therefore the “strength” of vapor shield on the target surface temperature. As a result, the influence of the vapor shielding phenomena details, such as radiation transport in the vapor cloud and evaporated material dynamics, on the E{sub max} is virtually completely masked by the strong dependence of the evaporation rate on the target surface temperature. However, the very same details define the amount of evaporated particles, needed to provide an effective shielding

  5. Upper wide-angle viewing system for ITER.

    PubMed

    Lasnier, C J; McLean, A G; Gattuso, A; O'Neill, R; Smiley, M; Vasquez, J; Feder, R; Smith, M; Stratton, B; Johnson, D; Verlaan, A L; Heijmans, J A C

    2016-11-01

    The Upper Wide Angle Viewing System (UWAVS) will be installed on five upper ports of ITER. This paper shows major requirements, gives an overview of the preliminary design with reasons for some design choices, examines self-emitted IR light from UWAVS optics and its effect on accuracy, and shows calculations of signal-to-noise ratios for the two-color temperature output as a function of integration time and divertor temperature. Accurate temperature output requires correction for vacuum window absorption vs. wavelength and for self-emitted IR, which requires good measurement of the temperature of the optical components. The anticipated signal-to-noise ratio using presently available IR cameras is adequate for the required 500 Hz frame rate.

  6. Upper wide-angle viewing system for ITER

    DOE PAGES

    Lasnier, C. J.; McLean, A. G.; Gattuso, A.; ...

    2016-08-15

    The Upper Wide Angle Viewing System (UWAVS) will be installed on five upper ports of ITER. Here, this paper shows major requirements, gives an overview of the preliminary design with reasons for some design choices, examines self-emitted IR light from UWAVS optics and its effect on accuracy, and shows calculations of signal-to-noise ratios for the two-color temperature output as a function of integration time and divertor temperature. Accurate temperature output requires correction for vacuum window absorption vs. wavelength and for self-emitted IR, which requires good measurement of the temperature of the optical components. The anticipated signal-to-noise ratio using presently availablemore » IR cameras is adequate for the required 500 Hz frame rate.« less

  7. Modelling of radiation impact on ITER Beryllium wall

    NASA Astrophysics Data System (ADS)

    Landman, I. S.; Janeschitz, G.

    2009-04-01

    In the ITER H-Mode confinement regime, edge localized instabilities (ELMs) will perturb the discharge. Plasma lost after each ELM moves along magnetic field lines and impacts on divertor armour, causing plasma contamination by back propagating eroded carbon or tungsten. These impurities produce enhanced radiation flux distributed mainly over the beryllium main chamber wall. The simulation of the complicated processes involved are subject of the integrated tokamak code TOKES that is currently under development. This work describes the new TOKES model for radiation transport through confined plasma. Equations for level populations of the multi-fluid plasma species and the propagation of different kinds of radiation (resonance, recombination and bremsstrahlung photons) are implemented. First simulation results without account of resonance lines are presented.

  8. Upper wide-angle viewing system for ITER

    SciTech Connect

    Lasnier, C. J.; McLean, A. G.; Gattuso, A.; O’Neill, R.; Smiley, M.; Vasquez, J.; Feder, R.; Smith, M.; Stratton, B.; Johnson, D.; Verlaan, A. L.; Heijmans, J. A. C.

    2016-08-15

    The Upper Wide Angle Viewing System (UWAVS) will be installed on five upper ports of ITER. Here, this paper shows major requirements, gives an overview of the preliminary design with reasons for some design choices, examines self-emitted IR light from UWAVS optics and its effect on accuracy, and shows calculations of signal-to-noise ratios for the two-color temperature output as a function of integration time and divertor temperature. Accurate temperature output requires correction for vacuum window absorption vs. wavelength and for self-emitted IR, which requires good measurement of the temperature of the optical components. The anticipated signal-to-noise ratio using presently available IR cameras is adequate for the required 500 Hz frame rate.

  9. Perl Modules for Constructing Iterators

    NASA Technical Reports Server (NTRS)

    Tilmes, Curt

    2009-01-01

    The Iterator Perl Module provides a general-purpose framework for constructing iterator objects within Perl, and a standard API for interacting with those objects. Iterators are an object-oriented design pattern where a description of a series of values is used in a constructor. Subsequent queries can request values in that series. These Perl modules build on the standard Iterator framework and provide iterators for some other types of values. Iterator::DateTime constructs iterators from DateTime objects or Date::Parse descriptions and ICal/RFC 2445 style re-currence descriptions. It supports a variety of input parameters, including a start to the sequence, an end to the sequence, an Ical/RFC 2445 recurrence describing the frequency of the values in the series, and a format description that can refine the presentation manner of the DateTime. Iterator::String constructs iterators from string representations. This module is useful in contexts where the API consists of supplying a string and getting back an iterator where the specific iteration desired is opaque to the caller. It is of particular value to the Iterator::Hash module which provides nested iterations. Iterator::Hash constructs iterators from Perl hashes that can include multiple iterators. The constructed iterators will return all the permutations of the iterations of the hash by nested iteration of embedded iterators. A hash simply includes a set of keys mapped to values. It is a very common data structure used throughout Perl programming. The Iterator:: Hash module allows a hash to include strings defining iterators (parsed and dispatched with Iterator::String) that are used to construct an overall series of hash values.

  10. ITER Cryoplant Infrastructures

    NASA Astrophysics Data System (ADS)

    Fauve, E.; Monneret, E.; Voigt, T.; Vincent, G.; Forgeas, A.; Simon, M.

    2017-02-01

    The ITER Tokamak requires an average 75 kW of refrigeration power at 4.5 K and 600 kW of refrigeration Power at 80 K to maintain the nominal operation condition of the ITER thermal shields, superconducting magnets and cryopumps. This is produced by the ITER Cryoplant, a complex cluster of refrigeration systems including in particular three identical Liquid Helium Plants and two identical Liquid Nitrogen Plants. Beyond the equipment directly part of the Cryoplant, colossal infrastructures are required. These infrastructures account for a large part of the Cryoplants lay-out, budget and engineering efforts. It is ITER Organization responsibility to ensure that all infrastructures are adequately sized and designed to interface with the Cryoplant. This proceeding presents the overall architecture of the cryoplant. It provides order of magnitude related to the cryoplant building and utilities: electricity, cooling water, heating, ventilation and air conditioning (HVAC).

  11. Near-infrared spectroscopy for divertor plasma diagnosis and control in DIII-D tokamaka)

    NASA Astrophysics Data System (ADS)

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

    2014-11-01

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

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

    SciTech Connect

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

    2014-11-15

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

  13. Diagnostics for ITER

    SciTech Connect

    Donne, A. J. H.; Hellermann, M. G. von; Barnsley, R.

    2008-10-22

    After an introduction into the specific challenges in the field of diagnostics for ITER (specifically high level of nuclear radiation, long pulses, high fluxes of particles to plasma facing components, need for reliability and robustness), an overview will be given of the spectroscopic diagnostics foreseen for ITER. The paper will describe both active neutral-beam based diagnostics as well as passive spectroscopic diagnostics operating in the visible, ultra-violet and x-ray spectral regions.

  14. Features and Initial Results of the DIII-D Advanced Tokamak Radiative Divertor

    SciTech Connect

    R.C. O'Neill; A.S. Bozek; M.E. Friend; C.B. Baxi; E.E. Reis; M.A. Mahdavi; D.G. Nilson; S.L. Allen; W.P. West

    1999-11-01

    The Radiative Divertor Program of DIII-D is in its final phase with the installation of the cryopump and baffle structure (Phase 1B Divertor) in the upper inner radius of the DIII-D vacuum vessel at the end of this calendar year. This divertor, in conjunction with the Advanced Divertor and the Phase 1A Divertor, located in the lower and upper outer radius of the DIII-D vacuum vessel respectively, provides pumping for density control of the plasma while minimizing the effects on the core confinement. Each divertor consists of a cryobelium cooling ring and a shielded protective structure. The cryo/helium-cooled pumps of all three diverters exhaust helium from the plasma. The protective shielded structure or baffle structure, in the case of the diverters located at the top of the vacuum vessel, provides baffling of neutral charged particles and minimize the flow of impurities back into the core of the plasma. The baffles, which consist of water-cooled panels that allow for the attachment of tiles of various sizes and shapes, house gas puff systems. The intent of the puffing systems is to inject gas in and around the divertor to minimize the heat flux on specific areas on the divertor and its components. The reduction of the heat flux on the divertor minimizes the impurities that are generated from excess heat on divertor components, specifically tiles. Experiments involving the gas puff systems and the divertor structures have shown the heat flux can be spread over a large area of the divertor, reducing the peak heat flux in specific areas. The three diverters also incorporate a variety of diagnostic tools such as halo current monitors, magnetic probes and thermocouples to monitor certain plasma characteristics as well as determine the effectiveness of the cryopumps and baffle configurations. The diverters were designed to optimize pumping performance and to withstand the electromagnetic loads from both halo currents and toroidal induced currents. Incorporated also

  15. HL-2M Divertor Geometry Exploration with SOLPS5.0

    NASA Astrophysics Data System (ADS)

    Cui, Xuewu; Pan, Yudong; Cui, Zhengying; Li, Jiaxian; Zhang, Jinhua; Mao, Rui

    2013-12-01

    One of the critical issues to be solved for HL-2M is the power and particle exhaust. Divertor target plate geometry strongly influences the plasma profiles by controlling the neutral recycling pattern, which has in turn a strong effect on the symmetry and stability of the divertor plasma and finally on the whole edge region. The numerical simulation software SOLPS5.0 Package is used to design and explore the divertor target plates for HL-2M. We choose two divertor geometries, and assess the heat flux on the target plates and first wall, then further discuss the divertor plasma parameters, and how private flux baffling affects both neutral recirculation pattern and pumping efficiency.

  16. The dynamics of coherent scrape-off layer structures in a snowflake divertor

    NASA Astrophysics Data System (ADS)

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

    2008-11-01

    A characteristic feature of a snowflake divertor is the quadratic dependence of the poloidal magnetic field strength vs the distance from the field null. Compared to a standard X-point divertor, where the magnetic field dependence over distance is linear, this leads to significant changes in the geometry of flux tubes passing in the vicinity of the null. In particular, squeezing of flux tubes by the magnetic shear becomes stronger; the field line mapping from the midplane to the divertor plate indicates much higher poloidal velocities of plasma filaments near the divertor plates. Thus, significant changes are expected in the dynamics of coherent structures (sometimes called ``blobs'') in the scrape-off layer. An analysis of the dynamical effects associated with curvature drive, divertor boundary conditions, and strong magnetic shearing is presented. Regimes of enhanced blob transport are identified. Prepared by LLNL under Contract DE-AC52-07NA27344.

  17. Reconstruction of Detached Divertor Plasma Conditions in DIII-D Using Spectroscopic and Probe Data

    SciTech Connect

    Stangeby, P; Fenstermacher, M

    2004-12-03

    For some divertor aspects, such as detached plasmas or the private flux zone, it is not clear that the controlling physics has been fully identified. This is a particular concern when the details of the plasma are likely to be important in modeling the problem--for example, modeling co-deposition in detached inner divertors. An empirical method of ''reconstructing'' the plasma based on direct experimental measurements may be useful in such situations. It is shown that a detached plasma in the outer divertor leg of DIII-D can be reconstructed reasonably well using spectroscopic and probe data as input to a simple onion-skin model and the Monte Carlo hydrogenic code, EIRENE. The calculated 2D distributions of n{sub e} and T{sub e} in the detached divertor were compared with direct measurements from the divertor Thomson scattering system, a diagnostic capability unique to DIII-D.

  18. Compatibility of ITER scenarios with an all-W wall

    NASA Astrophysics Data System (ADS)

    Sips, A. C. C.; Gruber, O.; ASDEX Upgrade Team

    2008-12-01

    In 2008, ASDEX Upgrade has started its second experimental campaign with full tungsten coverage of the plasma facing components. In the transition from a partially W-coated device (69% tungsten coverage in 2004/2005) to a full tungsten device (since 2007), post campaign analyses show a reduction by an order of magnitude in both the carbon deposition and deuterium retention for the experimental campaigns. Spectroscopic measurements show that the outer divertor is by far the strongest tungsten source region. However, influxes from the outboard limiters are the most important source for the tungsten content in the plasma. The application of ICRH results in large W influxes due to sputtering from light impurities accelerated by electrical fields at the ICRH antennas. In H-mode discharges, ELMs reduce the inward transport of tungsten in the H-mode edge transport barrier. Central heating provided by neutral beams and the upgraded ECRH systems is used to avoid tungsten accumulation in the core of the plasma. Stationary, ITER relevant H-modes (H98 ~ 1, βN ~ 1.6-2), with W concentrations below 3 × 10-5, were routinely achieved. High performance H-modes have been obtained before the first boronization, achieving H98 = 1.1-1.2 and βN up to 2.6 as required for advanced scenarios in ITER. Specific ITER studies performed in 2008 include the demonstration of low voltage plasma start-up with ECRH and heating during the current rise to q95 = 3, to achieve a range of plasma inductance of 0.71-0.97. The new results reported here form the basis of further enhancing the operational space of ASDEX Upgrade with the full tungsten wall, preparing for ITER and the ITER-like wall project in JET.

  19. Calorimeter probe for the DIII-D divertor

    NASA Astrophysics Data System (ADS)

    Watkins, J. G.; Lasnier, C. J.; Whyte, D. G.; Stangeby, P. C.; Ulrickson, M. A.

    2003-03-01

    Heat flux measurements of the DIII-D divertor plate have been obtained with 6 mm spatial resolution using a calorimeter probe. These measurements complement the infrared camera system normally used for heat flux measurements on DIII-D but at higher-spatial resolution. The calorimeter probe is inserted into the tokamak from below to a position which is flush with the lower divertor plate tiles using the divertor materials experimental station (DiMES). The DiMES mechanism allows for retraction of the probe behind a gate valve and removal from the tokamak for modification or calibration. A 6 mm diameter insulated graphite cylinder for collecting energy is mounted within a standard DiMES sample. A 0.8 mm diameter thermocouple, installed 4 mm below the surface, provides a measurement of the temperature during and after the plasma exposure. The 80 ms time constant for the measurement is fast enough to determine heat flux changes during the 5 s plasma discharge and heat flux profiles have been obtained using both fixed strike points and slow strike point sweeps across the calorimeter. Special electronics and isolation is necessary as the sample is in direct electrical contact with the plasma. The calorimeter observes approximately 100 °C temperature rise over one tokamak discharge. The thermocouple signals are typically less than 1 mV and must be amplified near the vacuum feedthrough, passed through a low-pass filter to eliminate magnetic pickup, isolated, and sent to the data acquisition system approximately 8 m away. Initial measurements are included.

  20. X-ray crystal spectrometer upgrade for ITER-like wall experiments at JET.

    PubMed

    Shumack, A E; Rzadkiewicz, J; Chernyshova, M; Jakubowska, K; Scholz, M; Byszuk, A; Cieszewski, R; Czarski, T; Dominik, W; Karpinski, L; Kasprowicz, G; Pozniak, K; Wojenski, A; Zabolotny, W; Conway, N J; Dalley, S; Figueiredo, J; Nakano, T; Tyrrell, S; Zastrow, K-D; Zoita, V

    2014-11-01

    The high resolution X-Ray crystal spectrometer at the JET tokamak has been upgraded with the main goal of measuring the tungsten impurity concentration. This is important for understanding impurity accumulation in the plasma after installation of the JET ITER-like wall (main chamber: Be, divertor: W). This contribution provides details of the upgraded spectrometer with a focus on the aspects important for spectral analysis and plasma parameter calculation. In particular, we describe the determination of the spectrometer sensitivity: important for impurity concentration determination.

  1. The Tungsten Project: Dielectronic recombination data for collisional-radiative modelling in ITER

    NASA Astrophysics Data System (ADS)

    Preval, S. P.; Badnell, N. R.; O'Mullane, M. G.

    2017-03-01

    Tungsten is an important metal in nuclear fusion reactors. It will be used in the divertor component of ITER (Latin for `the way'). The Tungsten Project aims to calculate partial and total DR rate coefficients for the isonuclear sequence of Tungsten. The calculated data will be made available as and when they are produced via the open access database OPEN-ADAS in the standard adf09 and adf48 file formats. We present our progress thus far, detailing calculational methods, and showing comparisons with other available data. We conclude with plans for the future.

  2. X-ray crystal spectrometer upgrade for ITER-like wall experiments at JET

    SciTech Connect

    Shumack, A. E.; Rzadkiewicz, J.; Chernyshova, M.; Czarski, T.; Karpinski, L.; Jakubowska, K.; Scholz, M.; Byszuk, A.; Cieszewski, R.; Kasprowicz, G.; Pozniak, K.; Wojenski, A.; Zabolotny, W.; Dominik, W.; Conway, N. J.; Dalley, S.; Tyrrell, S.; Zastrow, K.-D.; Figueiredo, J. [EFDA-CSU, Culham Science Centre, Abingdon OX14 3DB; Associação EURATOM and others

    2014-11-15

    The high resolution X-Ray crystal spectrometer at the JET tokamak has been upgraded with the main goal of measuring the tungsten impurity concentration. This is important for understanding impurity accumulation in the plasma after installation of the JET ITER-like wall (main chamber: Be, divertor: W). This contribution provides details of the upgraded spectrometer with a focus on the aspects important for spectral analysis and plasma parameter calculation. In particular, we describe the determination of the spectrometer sensitivity: important for impurity concentration determination.

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

    SciTech Connect

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

    2008-10-14

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

  4. Ballooning Modes in the Systems Stabilized by Divertors

    SciTech Connect

    Arsenin, V.V.; Skovoroda, A.A.; Zvonkov, A.V.

    2005-01-15

    MHD stability of a plasma in systems with closed magnetic field lines and open systems containing the nonparaxial stabilizing cells with large field lines curvature, in particular, divertors is analyzed. It is shown that population of particles trapped in such cells has a stabilizing effect not only on flute modes, but also on ballooning modes that determine the {beta} limit. At kinetic description that accounts for different effect of trapped and passing particles on perturbations, {beta} limit permitted by stability may be much greater then it follows from MHD model.

  5. Performance characteristics of the DIII-D advanced divertor cryopump

    SciTech Connect

    Menon, M.M.; Maingi, R.; Wade, M.R.; Baxi, C.B.; Campbell, G.L.; Holtrop, K.L.; Hyatt, A.W.; Laughon, G.J.; Makariou, C.C.; Mahdavi, M.A.; Reis, E.E.; Schaffer, M.J.; Schaubel, K.M.; Scoville, J.T.; Smith, J.P.; Stambaugh, R.D.

    1993-10-01

    A cryocondensation pump, cooled by forced flow of two-phase helium, has been installed for particle exhaust from the divertor region of the DIII-D tokamak. The Inconel pumping surface is of coaxial geometry, 25.4 mm in outer diameter and 11.65 m in length. Because of the tokamak environment, the pump is designed to perform under relatively high pulsed heat loads (300 Wm{sup {minus}2}). Results of measurements made on the pumping characteristics for D{sub 2}, H{sub 2}, and Ar are discussed.

  6. Crossed-field divertor for a plasma device

    DOEpatents

    Kerst, Donald W.; Strait, Edward J.

    1981-01-01

    A divertor for removal of unwanted materials from the interior of a magnetic plasma confinement device includes the division of the wall of the device into segments insulated from each other in order to apply an electric field having a component perpendicular to the confining magnetic field. The resulting crossed-field drift causes electrically charged particles to be removed from the outer part of the confinement chamber to a pumping chamber. This method moves the particles quickly past the saddle point in the poloidal magnetic field where they would otherwise tend to stall, and provides external control over the rate of removal by controlling the magnitude of the electric field.

  7. Overview of experimental preparation for the ITER-Like Wall at JET

    NASA Astrophysics Data System (ADS)

    Jet Efda Contributors Brezinsek, S.; Fundamenski, W.; Eich, T.; Coad, J. P.; Giroud, C.; Huber, A.; Jachmich, S.; Joffrin, E.; Krieger, K.; McCormick, K.; Lehnen, M.; Loarer, T.; de La Luna, E.; Maddison, G.; Matthews, G. F.; Mertens, Ph.; Nunes, I.; Philipps, V.; Riccardo, V.; Rubel, M.; Stamp, M. F.; Tsalas, M.

    2011-08-01

    Experiments in JET with carbon-based plasma-facing components have been carried out in preparation of the ITER-Like Wall with beryllium main chamber and full tungsten divertor. The preparatory work was twofold: (i) development of techniques, which ensure safe operation with the new wall and (ii) provision of reference plasmas, which allow a comparison of operation with carbon and metallic wall. (i) Compatibility with the W divertor with respect to energy loads could be achieved in N2 seeded plasmas at high densities and low temperatures, finally approaching partial detachment, with only moderate confinement reduction of 10%. Strike-point sweeping increases the operational space further by re-distributing the load over several components. (ii) Be and C migration to the divertor has been documented with spectroscopy and QMBs under different plasma conditions providing a database which will allow a comparison of the material transport to remote areas with metallic walls. Fuel retention rates of 1.0-2.0 × 1021 D s-1 were obtained as references in accompanied gas balance studies.

  8. Impact of the JET ITER-like wall on H-mode plasma fueling

    NASA Astrophysics Data System (ADS)

    Wiesen, S.; Brezinsek, S.; Wischmeier, M.; De la Luna, E.; Groth, M.; Jaervinen, A. E.; de la Cal, E.; Losada, U.; de Aguilera, A. M.; Frassinetti, L.; Gao, Y.; Guillemaut, C.; Harting, D.; Meigs, A.; Schmid, K.; Sergienko, G.; contributors, JET

    2017-06-01

    JET ITER-like wall (ILW) experiments show that the edge density evolution is strongly linked with the poloidal distribution of the ionization source. The fueling profile in the JET-ILW is more delocalized as compared to JET-C (JET with carbon-based plasma-facing components PFCs). Compared to JET-C the H-mode pedestal fueling cycle is dynamically influenced by a combination of plasma-wall interaction features, in particular: (1) edge-localized modes (ELMs) induced energetic particles are kinetically reflected on W divertor PFCs leading to distributed refueling away from the divertor depending on the divertor plasma configuration, (2) delayed molecular re-emission and outgassing of particles being trapped in W PFCs (bulk-W at the high field side and W-coated CFCs at the low field side) with different fuel content and (3) outgassing from Be co-deposits located on top of the high-field side baffle region shortly after the ELM. In view of the results of a set of well diagnosed series of JET-ILW type-I ELMy H-mode discharges with good statistics, the aforementioned effects are discussed in view of H-mode pedestal fueling capacity. The ongoing modelling activities with the focus on coupled core-edge plasma simulations and plasma-wall interaction are described and discussed also in view of possible code improvements required.

  9. An iterative decoupling solution method for large scale Lyapunov equations

    NASA Technical Reports Server (NTRS)

    Athay, T. M.; Sandell, N. R., Jr.

    1976-01-01

    A great deal of attention has been given to the numerical solution of the Lyapunov equation. A useful classification of the variety of solution techniques are the groupings of direct, transformation, and iterative methods. The paper summarizes those methods that are at least partly favorable numerically, giving special attention to two criteria: exploitation of a general sparse system matrix structure and efficiency in resolving the governing linear matrix equation for different matrices. An iterative decoupling solution method is proposed as a promising approach for solving large-scale Lyapunov equation when the system matrix exhibits a general sparse structure. A Fortran computer program that realizes the iterative decoupling algorithm is also discussed.

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

    NASA Astrophysics Data System (ADS)

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

    2017-02-01

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

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

    DOE PAGES

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

    2016-12-15

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

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

    SciTech Connect

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

    2016-12-15

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

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

    NASA Astrophysics Data System (ADS)

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

    2001-03-01

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

  14. Robust iterative methods

    SciTech Connect

    Saadd, Y.

    1994-12-31

    In spite of the tremendous progress achieved in recent years in the general area of iterative solution techniques, there are still a few obstacles to the acceptance of iterative methods in a number of applications. These applications give rise to very indefinite or highly ill-conditioned non Hermitian matrices. Trying to solve these systems with the simple-minded standard preconditioned Krylov subspace methods can be a frustrating experience. With the mathematical and physical models becoming more sophisticated, the typical linear systems which we encounter today are far more difficult to solve than those of just a few years ago. This trend is likely to accentuate. This workshop will discuss (1) these applications and the types of problems that they give rise to; and (2) recent progress in solving these problems with iterative methods. The workshop will end with a hopefully stimulating panel discussion with the speakers.

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

    NASA Astrophysics Data System (ADS)

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

    2016-12-01

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

  16. Axisymmetric curvature-driven instability in a model divertor geometry

    SciTech Connect

    Farmer, W. A.; Ryutov, D. D.

    2013-09-15

    A model problem is presented which qualitatively describes a pressure-driven instability which can occur near the null-point in the divertor region of a tokamak where the poloidal field becomes small. The model problem is described by a horizontal slot with a vertical magnetic field which plays the role of the poloidal field. Line-tying boundary conditions are applied at the planes defining the slot. A toroidal field lying parallel to the planes is assumed to be very strong, thereby constraining the possible structure of the perturbations. Axisymmetric perturbations which leave the toroidal field unperturbed are analyzed. Ideal magnetohydrodynamics is used, and the instability threshold is determined by the energy principle. Because of the boundary conditions, the Euler equation is, in general, non-separable except at marginal stability. This problem may be useful in understanding the source of heat transport into the private flux region in a snowflake divertor which possesses a large region of small poloidal field, and for code benchmarking as it yields simple analytic results in an interesting geometry.

  17. ALPS - advanced limiter-divertor plasma-facing systems.

    SciTech Connect

    Allain, J. P.; Bastasz, R.; Brooks, J. N.; Evans, T.; Hassanein, A.; Luckhardt, S.; Maingi, R.; Mattas, R. F.; McCarthy, K.; Mioduszewski, P.; Mogahed, E.; Moir, R.; Molokov, S.; Morely, N.; Nygren, R.; Reed, C.; Rognlien, T.; Ruzic, D.; Sviatoslavsky, I.; Sze, D.; Tillack, M.; Ulrickson, M.; Wade, P. M.; Wong, C.; Wooley, R.

    1999-09-15

    The Advanced Limiter-divertor Plasma-facing Systems (ALPS) program was initiated in order to evaluate the potential for improved performance and lifetime for plasma-facing systems. The main goal of the program is to demonstrate the advantages of advanced limiter/divertor systems over conventional systems in terms of power density capability, component lifetime, and power conversion efficiency, while providing for safe operation and minimizing impurity concerns for the plasma. Most of the work to date has been applied to free surface liquids. A multi-disciplinary team from several institutions has been organized to address the key issues associated with these systems. The main performance goals for advanced limiters and diverters are a peak heat flux of >50 MW/m{sup 2},elimination of a lifetime limit for erosion, and the ability to extract useful heat at high power conversion efficiency ({approximately}40%). The evaluation of various options is being conducted through a combination of laboratory experiments, modeling of key processes, and conceptual design studies. The current emphasis for the work is on the effects of free surface liquids on plasma edge performance.

  18. Iterated multidimensional wave conversion

    SciTech Connect

    Brizard, A. J.; Tracy, E. R.; Johnston, D.; Kaufman, A. N.; Richardson, A. S.; Zobin, N.

    2011-12-23

    Mode conversion can occur repeatedly in a two-dimensional cavity (e.g., the poloidal cross section of an axisymmetric tokamak). We report on two novel concepts that allow for a complete and global visualization of the ray evolution under iterated conversions. First, iterated conversion is discussed in terms of ray-induced maps from the two-dimensional conversion surface to itself (which can be visualized in terms of three-dimensional rooms). Second, the two-dimensional conversion surface is shown to possess a symplectic structure derived from Dirac constraints associated with the two dispersion surfaces of the interacting waves.

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

    NASA Astrophysics Data System (ADS)

    Chen, L.; Xu, G. S.; Gao, W.; Zhang, L.; Nielsen, A. H.; Luo, Z. P.; Si, H.; Wang, Y. M.; Qu, H.; Sun, Z.; Duan, Y. M.; Liu, H. Q.; Wang, S. X.; Li, M. H.; Zhang, X. J.; Wu, B.; Chen, R.; Wang, L.; Wang, H. Q.; Ding, S. Y.; Yan, N.; Liu, S. C.; Shao, L. M.; Zhang, W.; Hu, G. H.; Li, J.; Li, Y. L.; Wu, X. Q.; Zhao, N.; Jia, M. N.

    2016-05-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2012-10-01

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

  1. A comprehensive 2-D divertor data set from DIII-D for edge theory validation

    SciTech Connect

    Fenstermacher, M.E.; Allen, S.L.; Hill, D.N.

    1996-02-01

    A comprehensive set of experiments has been carried out on the DIII-D tokamak to measure the 2-D (R,Z) structure of the divertor plasma in a systematic way using new diagnostics. Measurements cover the divertor radially from inside the X-point to the outer target plate and vertically from the target plate to above the X-point. Identical, repeatable shots were made, each having radial sweeps of the X-point and divertor strike points, to allow complete plasma and radiation profile measurements. Data have been obtained in ohmic, L-mode, ELMing H-mode, and reversed B{sub T} operation ({gradient}B drift away from the X-point). In addition, complete measurements were made of radiative divertor plasmas with a Partially Detached Divertor (PDD) induced by D{sub 2} injection and with a Radiating Mantle induced by Impurity injection (RMI) using neon and nitrogen. The data set includes first observations of the radial and poloidal profiles of the X-point, inner and outer leg plasmas in PDD and RMI radiative divertor operation. Preliminary data analysis shows that intrinsic impurities play a critical role in determining the SOL and divertor conditions.

  2. Direct measurement of divertor exhaust neo enrichment in DIII-D

    SciTech Connect

    Schaffer, M.J.; Wade, M.R.; Maingi, R.; Monier-Garbet, P.; West, W.P.; Whyte, D.G.; Wood, R.D.; Mahdavi, M.A.

    1996-06-01

    We report first direct measurements of divertor exhaust gas impurity enrichment, {eta}{sub exh}=(exhaust impurity concentration){divided_by}(core impurity concentration), for both unpumped and D{sub 2} puff-with-divertor-pump conditions. The experiment was performed with neutral beam heated, ELMing H-mode, single-null diverted deuterium plasmas with matched core and exhaust parameters in the DIII-D tokamak. Neon gas impurity was puffed into the divertor. Neon density was measured in the exhaust by a specially modified Penning gauge and in the core by absolute charge exchange recombination spectroscopy. Neon particle accounting indicates that much of the puffed neon entered a temporary unmeasured reservoir, inferred to be the graphite divertor target, which makes direct measurements necessary to calculate divertor enrichments. D{sub 2} puff into the SOL (scrape-off layer) with pumping increased {eta}{sub exh} threefold over either unpumped conditions or D{sub 2} puff directly into the divertor with pumping. These results show that SOL flow plays an important role in divertor exhaust impurity enrichment.

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

    NASA Astrophysics Data System (ADS)

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

    2012-08-01

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

  4. Role of cross-field drifts in the onset of divertor detachment

    NASA Astrophysics Data System (ADS)

    Groth, Mathias; Allen, S. L.; Fenstermacher, M. E.; Hill, D. H.; Makowski, M. A.; McLean, A. G.; Lasnier, C. J.; Porter, G. D.; Rognlien, T. D.; Briesemeister, A. R.; Unterberg, E. A.; Leonard, A. W.; Watkins, J. G.

    2015-11-01

    The impact of cross-field drifts in divertor configurations was investigated in DIII-D L and H-mode discharges. The studies show that the electron temperature at the outer divertor plate is reduced to below 2 eV at about 20 % lower pedestal density in configurations with the ion Bx ∇B direction toward the divertor X-point. When attached, these plasmas have significantly lower electron temperatures and and higher densities in the inner than in the outer divertor as directly measured with divertor Thomson scattering and inferred from line emission imaging using tangentially viewing cameras. Upon reversal of the toroidal field direction, the divertor conditions were observed in-out symmetric. Simulations with the edge fluid code UEDGE show that poloidal flows due to the radial electric field in the private flux region dominate the divertor asymmetries. Work supported by US DOE under DE-AC52-07NA27344, DE-FC02-04ER54698, DE-AC05-00OR22725, and DE-AC04-94AL85000.

  5. Impact of real-time magnetic axis sweeping on steady state divertor operation in LHD

    NASA Astrophysics Data System (ADS)

    Nakamura, Y.; Masuzaki, S.; Morisaki, T.; Ogawa, H.; Watanabe, T.; Kubota, Y.; Sakamoto, R.; Ashikawa, N.; Sato, K.; Chikaraishi, H.; Saito, K.; Seki, T.; Kumazawa, R.; Mutoh, T.; Kubo, S.; Takeiri, Y.; Peterson, B. J.; Komori, A.; Motojima, O.; LHD experimental Group

    2006-07-01

    Steady state divertor operation with high performance plasmas (ne ~ 0.7 × 1019 cm-3, Ti ~ 2 keV) was demonstrated for half an hour in the Large Helical Device (LHD), the superconducting helical device (R = 3.6-3.9 m, a = 0.6 m, B = 3 T, l/m = 2/10). The high performance plasmas have been sustained with an averaged heating power of 680 kW and achieved an injected energy of 1.3 GJ. This required both advanced technological integration of heating systems and divertor heat flux control. In particular, optimization of divertor heat flux distribution along the divertor leg trace on divertor plates and real-time magnetic axis sweeping (R = 3.67-3.7 m) have allowed LHD to access a steady state regime with a margin of safety for the actively cooled divertor plates. The distribution of divertor heat load along the traces was investigated with calorimetric measurements and it was found that there was a localized heat load connected with the loss of high-energy ions produced by ion cyclotron radio frequency near-fields. Orbit analysis shows that the behaviour of high-energy ions is qualitatively in good agreement with the experimental result. Long-pulse discharges were terminated by radiation collapse due to penetration of metallic flakes into the plasma.

  6. Experimental estimation of tungsten impurity sputtering due to Type I ELMs in JET-ITER-like wall using pedestal electron cyclotron emission and target Langmuir probe measurements

    NASA Astrophysics Data System (ADS)

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

    2016-02-01

    The ITER baseline scenario, with 500 MW of DT fusion power and Q = 10, will rely on a Type I ELMy H-mode and will be achieved with a tungsten (W) divertor. W atoms sputtered from divertor targets during mitigated ELMs are expected to be the dominant source in ITER. W impurity concentration in the plasma core can dramatically degrade its performance and lead to potentially damaging disruptions. Understanding the physics of the target W source due to sputtering during ELMs and inter-ELMs is important and can be helped by experimental measurements with improved precision. It has been established that the ELMy target ion impact energy has a simple linear dependence with the pedestal electron temperature measured by Electron Cyclotron Emission (ECE). It has also been shown that Langmuir Probes (LP) ion flux measurements are reliable during ELMs due to the surprisingly low electron temperature. Therefore, in this paper, LP and ECE measurements in JET-ITER-Like-Wall (ILW) unseeded Type I ELMy H-mode experiments have been used to estimate the W sputtering flux from divertor targets in ELM and inter-ELM conditions. Comparison with similar estimates using W I spectroscopy measurements shows a reasonable agreement for the ELM and inter-ELM W source. The main advantage of the method involving LP measurements is the very high time resolution of the diagnostic (˜10 μs) allowing very precise description of the W sputtering source during ELMs.

  7. The influence of Filaments in the Private Flux Region on Divertor Power and Particle Deposition

    NASA Astrophysics Data System (ADS)

    Harrison, James

    2014-10-01

    Recent advances in imaging of the MAST divertor have revealed, for the first time, evidence for filaments in the private flux region (PFR). Detailed analysis of the image data shows 3 distinct types of fluctuations occurring within the divertor volume: highly sheared filaments in the SOL originating from the outer midplane, high frequency (>50 kHz) filaments near the separatrix of the outer divertor leg and filaments in the private flux region originating from inner divertor leg. With the need to extrapolate divertor performance from existing machines to future devices, these observations can contribute to our quantitative understanding of transport in the PFR. In particular, they suggest that transport in the PFR is, at least in part, driven by turbulence, which may not be well captured by the Eich/Wagner description of the divertor footprint, expressed in terms of exponential decay in space above the X-point and Gaussian spreading below the X-point. The PFR filaments are observed to move largely parallel with the flux surfaces in a way equivalent to a toroidal angular velocity of order 2 ×104 rad/s in H-mode, and slower by a factor of order 2 in L-mode. During their transit parallel to the flux surfaces across the PFR, the filaments eject plasma in bursts, away from the separatrix, deeper into the private flux region. Correlation analysis suggests that they are generated by processes local to the inner divertor leg, as there is a weak correlation between fluctuations in the SOL and PFR above what is expected from line integration effects. Scaling of filament properties with machine operating parameters, such as plasma current, density and auxiliary heating power will be presented, together with a comparison with data from divertor Langmuir probes and IR thermography to estimate the role PFR filaments play in determining the width of the divertor footprint.

  8. Non-axisymmetric ideal equilibrium and stability of ITER plasmas with rotating RMPs

    NASA Astrophysics Data System (ADS)

    Ham, C. J.; Cramp, R. G. J.; Gibson, S.; Lazerson, S. A.; Chapman, I. T.; Kirk, A.

    2016-08-01

    The magnetic perturbations produced by the resonant magnetic perturbation (RMP) coils will be rotated in ITER so that the spiral patterns due to strike point splitting which are locked to the RMP also rotate. This is to ensure even power deposition on the divertor plates. VMEC equilibria are calculated for different phases of the RMP rotation. It is demonstrated that the off harmonics rotate in the opposite direction to the main harmo