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Sample records for detached divertor plasmas

  1. 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.

  2. 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.

  3. 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 » $$\

  4. 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   ×  \

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. Achieving temporary divertor plasma detachment with MARFE events by pellet injection in the EAST superconducting tokamak

    NASA Astrophysics Data System (ADS)

    Guozhong, Deng; Liang, Wang; Xiaoju, Liu; Yanmin, Duan; Jiansheng, Hu; Changzheng, Li; Ling, Zhang; Shaocheng, Liu; Huiqian, Wang; Liang, Chen; Jichan, Xu; Wei, Feng; Jianbin, Liu; Huan, Liu; Guosheng, Xu; Houyang, Guo; Xiang, Gao; the EAST Team

    2017-01-01

    A new pellet injection system has been equipped on the experimental advanced superconducting tokamak (EAST) in the 2012 campaign, with a pellet size of ϕ 2 mm × 2 mm, a frequency of 1 Hz-10 Hz and velocity of 150 m s-1-300 m s-1. The deuterium pellet is well-known for plasma fuelling as well as for triggering the edge localized mode (ELM). In the 2012 campaign, pellet injection experiments were successfully carried out on EAST. Temporary plasma detachment achieved by deuterium pellets has been observed in a double null (DN) divertor configuration, with multi-pellet injections at a repetition frequency of 2 Hz. The partial detachment of the outer divertors and complete detachment of the inner divertors was achieved after 35 ms of each pellet injection, which have a duration of 30-60 ms with the maximum degree of detachment (DOD) reaching 3.5 and 37, respectively. Meanwhile, the multifaceted asymmetric radiation from the edge (MARFE) phenomena was also observed at the high field side (HFS) near both the lower and upper X-points with radiation loss suddenly increased to about 15%-70%, which may be the main cause of divertor plasma detachment. The temporary detachment induced by pellet injection may act as a new way to study divertor detachment behaviors.

  13. 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

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

    NASA Astrophysics Data System (ADS)

    Goldston, Rj; Schwartz, Ja

    2016-10-01

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

  15. Field reversal effects on divertor plasmas under radiative and detached conditions in JT-60U

    NASA Astrophysics Data System (ADS)

    Asakura, N.; Hosogane, H.; Tsuji-Iio, S.; Itami, K.; Shimizu, K.; Shimada, M.

    1996-06-01

    Reversal effects of the toroidal field Bt on the principal divertor plasma parameters were investigated under radiative and detached divertor conditions in L mode discharges. The ion flux to the inboard separatrix strike point decreased before a MARFE occurred, irrespective of the ion Del B drift direction. The local electron temperature, Te, div, decreased to around 10 eV. The maximum fraction of power radiated in the divertor was comparable between the two directions of Bt. With the power flowing into the two divertor fans being slightly larger on the outboard than on the inboard, a nearly symmetric in-out heat load was observed for the ion Del B drift away from the target. This was due to the outboard enhanced asymmetries in the particle flux and radiation loss distributions. From the viewpoint of in-out symmetry in the target heat load and Te, div, operation with the ion Del B drift away from the target plate is desirable as long as the attached divertor condition is maintained. On the contrary, during the MARFE, although deterioration of the energy confinement as well as the increase in the fuelling efficiency were comparable, for the ion Del B drift towards the target the plasma did not detach completely, and the strong in-out asymmetry in the particle recycling was relaxed to a relatively symmetric distribution. From the viewpoint of particle exhaust to the divertor, operation with the ion Del B drift towards the target is favourable

  16. Suppression of erosion in the DIII-D divertor with detached plasmas

    SciTech Connect

    WAMPLER,WILLIAM R.; BASTASZ,ROBERT J.; WHYTE,D.G.; WONG,C.P.C.; WEST,W.P.

    2000-05-25

    The ability to withstand disruptions makes carbon-based materials attractive for use as plasma-facing components in divertors. However, such materials suffer high erosion rates during attached plasma operation which, in high power long pulse machines, would give short component lifetimes and high tritium inventories. The authors present results from recent experiments in DIII-D, in which the Divertor Materials Evaluation System (DiMES) was used to examine erosion and deposition during short exposures to well defined plasma conditions. These studies show that during operation with detached plasmas, produced by gas injection, net erosion is suppressed everywhere in the divertor. Net deposition of carbon with deuterium was observed at the inner and outer strikepoints and in the private-flux region between strikepoints. For these low temperature plasmas (T{sub e} < 2eV), physical sputtering is eliminated. These results show that with detached plasmas, the location of carbon net erosion and the carbon impurity source, probably lies outside the divertor. Physical or chemical sputtering by charge-exchange neutrals or ions in the main plasma chamber is a probable source of carbon under these plasma conditions.

  17. 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.

  18. 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

  19. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2015-11-01

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

  1. 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).

  2. 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

  3. Edge and divertor plasma: detachment, stability, and plasma-wall interactions

    NASA Astrophysics Data System (ADS)

    Krasheninnikov, S. I.; Kukushkin, A. S.; Lee, Wonjae; Phsenov, A. A.; Smirnov, R. D.; Smolyakov, A. I.; Stepanenko, A. A.; Zhang, Yanzeng

    2017-10-01

    The paper presents an overview of the results of studies on a wide range of the edge plasma related issues. The rollover of the plasma flux to the target during progressing detachment process is shown to be caused by the increase of the impurity radiation loss and volumetric plasma recombination, whereas the ion-neutral friction, although important for establishing the necessary edge plasma conditions, does not contribute per se to the rollover of the plasma flux to the target. The processes limiting the power loss by impurity radiation are discussed and a simple estimate of this limit is obtained. Different mechanisms of meso-scale thermal instabilities driven by impurity radiation and resulting in self-sustained oscillations in the edge plasma are identified. An impact of sheared magnetic field on the dynamics of the blobs and ELM filaments playing an important role in the edge and SOL plasma transport is discussed. Trapping of He, which is an intrinsic impurity for the fusion plasmas, in the plasma-facing tungsten material is considered. A newly developed model, accounting for the generation of additional He traps caused by He bubble growth, fits all the available experimental data on the layer of nano-bubbles observed in W under irradiation by low energy He plasma.

  4. 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

  5. Plasma detachment in linear devices

    NASA Astrophysics Data System (ADS)

    Ohno, N.

    2017-03-01

    Plasma detachment research in linear devices, sometimes called divertor plasma simulators, is reviewed. Pioneering works exploring the concept of plasma detachment were conducted in linear devices. Linear devices have contributed greatly to the basic understanding of plasma detachment such as volume plasma recombination processes, detached plasma structure associated with particle and energy transport, and other related issues including enhancement of convective plasma transport, dynamic response of plasma detachment, plasma flow reversal, and magnetic field effect. The importance of plasma detachment research using linear devices will be highlighted aimed at the design of future DEMO.

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

    NASA Astrophysics Data System (ADS)

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

    2017-03-01

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

  7. Progress of divertor simulation research toward the realization of detached plasma using a large tandem mirror device

    NASA Astrophysics Data System (ADS)

    Nakashima, Y.; Takeda, H.; Ichimura, K.; Hosoi, K.; Oki, K.; Sakamoto, M.; Hirata, M.; Ichimura, M.; Ikezoe, R.; Imai, T.; Iwamoto, M.; Hosoda, Y.; Katanuma, I.; Kariya, T.; Kigure, S.; Kohagura, J.; Minami, R.; Numakura, T.; Takahashi, S.; Yoshikawa, M.; Asakura, N.; Fukumoto, M.; Hatayama, A.; Hirooka, Y.; Kado, S.; Kubo, H.; Masuzaki, S.; Matsuura, H.; Nagata, S.; Nishino, N.; Ohno, N.; Sagara, A.; Shikama, T.; Shoji, M.; Tonegawa, A.; Ueda, Y.

    2015-08-01

    This paper describes the results of the experiments performed on Tandem Mirror device GAMMA 10/PDX mainly using a new "divertor simulation experimental module (D-module)" installed on one of the end mirror exits which is specially designed to investigate the physics of plasma detachment. The additional ICRF heating in the anchor-cells, connected to both ends of the central-cell, significantly increases the density in the both cells, which attained the generation of the highest particle flux up to 1023 particles/s m2 at the end-mirror exit. H2 and noble gas injection to enhance the radiation cooling in D-module was performed and a remarkable reduction of the electron temperature (from few tens eV to <3 eV) on the target plate were successfully achieved associated with the strong reduction of particle and heat flux. A significant effect of simultaneous injection with hydrogen and noble gases for detached plasma formation was recognized for the first time.

  8. 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.

  9. 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

  10. Induced carbon deposition by local hydrocarbon injection into detached divertor plasmas in JET

    NASA Astrophysics Data System (ADS)

    Jet Efda Contributors Airila, M. I.; Coad, J. P.; Brezinsek, S.; Belo, P.; Groth, M.; Kirschner, A.; Makkonen, T.; Stamp, M. F.; Strachan, J. D.; Widdowson, A. M.; Wiesen, S.

    2011-08-01

    During detachment experiments of JET in 2007 (Brezinsek et al., 2009 [1]), 1 × 1022 molecules of 12CD4 were injected at a rate of about 5 × 1020 s-1 into L-mode plasma. The injection was done at the outer strike point in the centre of the horizontal target, just before opening the machine. The deposited layers were analyzed for deuterium post mortem after the removal of tiles. The heaviest local D deposition density is found immediately upstream of the gas inlet, but downstream the deposition is found over a larger area. In total, 3.7 × 1020 deuterium atoms were found locally deposited; if the mean D/C in the deposits is 0.4, then about 10% of the injected carbon was locally deposited. Transport and local deposition of the injected carbon was modeled with the 3D Monte Carlo impurity transport code ERO (Kirschner et al., 2000 [2]). The plasma background was generated with the onion-skin solver of the DIVIMP code (Stangeby and Elder, 1992 [3]). Locally deposited fraction is reproduced with ERO but the distribution not satisfactorily. In particular the poloidal transport of carbon is several times weaker than measured.

  11. 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.

  12. 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.

  13. Model of detached plasmas

    SciTech Connect

    Yoshikawa, S.; Chance, M.

    1986-07-01

    Recently a tokamak plasma was observed in TFTR that was not limited by a limiter or a divertor. A model is proposed to explain this equilibrium, which is called a detached plasma. The model consists of (1) the core plasma where ohmic heating power is lost by anomalous heat conduction and (2) the shell plasma where the heat from the core plasma is radiated away by the atomic processes of impurity ions. A simple scaling law is proposed to test the validity of this model.

  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. 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.

  16. 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.

  17. 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 \

  18. 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

  19. 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

  20. 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.

  1. 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.

  2. 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.

  3. 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

  4. 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.

  5. 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.

  6. Kinetic dynamics simulation of the detached plasma

    NASA Astrophysics Data System (ADS)

    Pianpanit, Theerasarn; Ishiguro, Seiji; Hasegawa, Hiroki

    2015-11-01

    The detached plasma has been proposed to reduce the heat flux to the divertor. Fluid code has been widely used to investigate the detached plasma but the cooling of plasma, trapped particle effects, and other kinetic dynamics in the detached plasma has not been well understood. Particle-in-Cell (PIC) simulation with the Monte Carlo collisions and the cumulative scattering angle coulomb collision are carried out to study dynamical kinetic behavior of the plasma. The constant pressure and temperature of neutral gas box in front of the divertor target model has been used in the simulation. The results show the decrease in electron temperature in front of the divertor plate strongly relate to the Coulomb collision frequency.

  7. 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.

  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. Enhancement of cross-field transport into the private region of detached-divertor in Large Helical Device

    NASA Astrophysics Data System (ADS)

    Tanaka, H.; Ohno, N.; Tsuji, Y.; Kajita, S.; Masuzaki, S.; Kobayashi, M.; Morisaki, T.; Tsuchiya, H.; Komori, A.; LHD Experimental Group

    2010-10-01

    The fluctuation of ion saturation currents in the attached- and detached-divertor plasmas of the Large Helical Device [Fujiwara et al., Nucl. Fusion 41, 1355 (2001)] has been measured using a Langmuir probe array embedded in a divertor plate. Analytical results indicate that these fluctuation properties differ considerably from each other; for instance, the mean value distribution expands to and positive spikes propagate toward a private region from the divertor leg in the detached-divertor. We investigated the magnetic field lines traced from probe electrodes by using the KMAG code [Nakamura et al., J. Plasma Fusion Res. 69, 41 (1993)], and it is then confirmed that the propagation direction of positive spikes corresponds to that predicted by the theory of blobby plasma transport. This phenomenon is expected to lead to the broadening of plasma particle and heat fluxes to the divertor plate.

  10. Enhancement of cross-field transport into the private region of detached-divertor in Large Helical Device

    SciTech Connect

    Tanaka, H.; Ohno, N.; Tsuji, Y.; Kajita, S.; Masuzaki, S.; Kobayashi, M.; Morisaki, T.; Tsuchiya, H.; Komori, A.

    2010-10-15

    The fluctuation of ion saturation currents in the attached- and detached-divertor plasmas of the Large Helical Device [Fujiwara et al., Nucl. Fusion 41, 1355 (2001)] has been measured using a Langmuir probe array embedded in a divertor plate. Analytical results indicate that these fluctuation properties differ considerably from each other; for instance, the mean value distribution expands to and positive spikes propagate toward a private region from the divertor leg in the detached-divertor. We investigated the magnetic field lines traced from probe electrodes by using the KMAG code [Nakamura et al., J. Plasma Fusion Res. 69, 41 (1993)], and it is then confirmed that the propagation direction of positive spikes corresponds to that predicted by the theory of blobby plasma transport. This phenomenon is expected to lead to the broadening of plasma particle and heat fluxes to the divertor plate.

  11. Current understanding of divertor detachment: experiments and modelling

    SciTech Connect

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

    2008-05-23

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

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

    NASA Astrophysics Data System (ADS)

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

    2017-08-01

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

  13. 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.

  14. 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.

  15. 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.

  16. X-Divertor Geometries for Deeper Detachment Without Degrading the DIII-D H-Mode

    NASA Astrophysics Data System (ADS)

    Covele, Brent; Kotschenreuther, M. T.; Valanju, P. M.; Mahajan, S. M.; Leonard, A. W.; Hyatt, A. W.; McLean, A. G.; Thomas, D. M.; Guo, H. Y.; Watkins, J. G.; Makowski, M. A.; Hill, D. N.

    2015-11-01

    Recent DIII-D experiments comparing the standard divertor (SD) and X-Divertor (XD) geometries show heat and particle flux reduction at the divertor target plate. The XD features large poloidal flux expansion, increased connection length, and poloidal field line flaring, quantified by the Divertor Index. Both SD and XD were pushed deep into detachment with increased gas puffing, until core energy confinement and pedestal pressure were substantially reduced. As expected, outboard target heat fluxes are significantly reduced in the XD compared to the SD under similar upstream plasma conditions, even at low Greenwald fraction. The high-triangularity (floor) XD cases show larger reduction in temperature, heat, and particle flux relative to the SD in all cases, while low-triangularity (shelf) XD cases show more modest reductions over the SD. Consequently, heat flux reduction and divertor detachment may be achieved in the XD with less gas puffing and higher pedestal pressures. Further causative analysis, as well as detailed modeling with SOLPS, is underway. These initial experiments suggest the XD as a promising candidate to achieve divertor heat flux control compatible with robust H-mode operation. Work supported by US DOE under DE-FC02-04ER54698, DE-AC52-07NA27344, DE-FG02-04ER54754, and DE-FG02-04ER54742.

  17. 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.

  18. 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.

  19. SOLPS modeling of an innovative small-angle slot divertor concept for low-density detachment

    NASA Astrophysics Data System (ADS)

    Covele, B.; Sang, C.; Guo, H.; Lao, L.; Stangeby, P.; Thomas, D.

    2016-10-01

    SOLPS modeling offers insight into how a new Small-Angle Slot (SAS) divertor concept exploits the role of neutral trapping to exhaust power and particles at lower core densities than even highly slanted divertors. The special SAS baffling structure enhances volumetric power and momentum losses across the entire target profile, flattening temperatures even in the far SOL. SOLPS characterizes SAS heat and temperature handling for a spectrum of plasma and neutral source conditions, varying ne,sep, PSOL, heat flux width, gas puffing rates and locations, and pumping rates. Certain aspects of the baffling structure were also systematically varied to observe the effect on the neutral dynamics, particularly pressure gradients in D2 near the target. Radial transport coefficients were controlled to match midplane profiles to experimental H-mode profiles. The SAS divertor is an excellent testbed for probing the interplay between plasma and neutrals at the onset of detachment. The SAS concept is developed under General Atomics corporate funding.

  20. Modeling Detached Plasmas in DIII-D

    NASA Astrophysics Data System (ADS)

    Porter, Gary D.; Rognlien, T. D.; Rensink, M. E.; DIII-D Team

    1996-11-01

    The ITER divertor design relies on operation of the machine with a detached divertor plasma as a means of reducing the divertor heat load to manageable levels. This operating mode has been seen on all of the world's diverted tokamaks, and is characterized by very low plate temperatures and ion currents. Experimental results on DIII-D have shown the plate electron temperature is between 1 and 2 eV. We describe the results of modeling these detached plasmas with the UEDGE code in this paper. Plasma detachment can be achieved in a variety of ways in the code as well as in experiment. Simulations indicate the detachment process occurs in two steps: a thermal collapse in which the plate temperature drops to 1 to 2 eV, followed by a decrease in the plate ion current. When the low temperature region extends off the plate, parallel momentum of the plasma is reduced by ion/neutral interactions. The plate ion current decreases when the parallel momentum is reduced sufficiently to permit volume recombination processes to compete with ion flow to the plate.

  1. PIC Simulation of plasma detachment

    NASA Astrophysics Data System (ADS)

    Ishiguro, Seiji; Pianpanit, Theerasarn; Hasegawa, Hiroki; Kanno, Ryutaro

    2014-10-01

    The detached plasma, which is caused by gas puffing, has been proposed and it is the most promising way to reduce the heat load to the divertor plate of fusion oriented devices. Dynamical and kinetic behavior of the detached plasma is unresolved. So we are developing particle-in-cell simulation model with atomic processes such as line radiation, ionization, charge-exchange collision and recombination. As a first step, we have performed PIC simulation with Monte Carlo collisions, where spatial and velocity space distributions of charged particles, self-consistent electric field, and atomic processes such as ionization and charge exchange are included. Temperature decrease and density increase in front of the target is observed and electric potential structure along the axis is created. This work is performed with the support and under the auspices of NIFS Collaboration Research programs (NIFS14KNXN279 and 8 NIFS13KNSS038) and the Research Cooperation Program on Hierarchy and Holism in Natural Sciences at the NINS.

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

    NASA Astrophysics Data System (ADS)

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

    2017-04-01

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

  3. Scaling of midplane separatrix density with power at divertor detachment onset

    NASA Astrophysics Data System (ADS)

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

    2015-11-01

    The midplane separatrix density at divertor detachment onset is found to increase with higher parallel heat flux, q| |, flowing into the divertor, but at a slower rate than expected from simple scaling models. The separatrix density will be an important parameter in determining the compatibility of divertor heat flux control with robust pedestal operation and high core confinement in future devices. The parallel heat flux is examined by separately varying several parameters, including injected power, plasma current, toroidal field and injected impurities. Several methods are employed to locate the separatrix in this critical region of steep density gradients, including magnetic equilibrium reconstruction, power balance assumptions and spatial fiducials from other diagnostics. All methods exhibit a slower than the q|| 5 / 7 scaling predicted by a simple two point model. The nonlinear dependence of divertor radiation with power and density is one of several factors leading to this difference. Supported in part by the US DOE under DE-FC02-04ER54698 & DE-AC52-07NA27344.

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

    DOE PAGES

    Briesemeister, A. R.; Isler, R. C.; Allen, S. L.; ...

    2014-11-15

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

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

    SciTech Connect

    Briesemeister, A. R.; Isler, R. C.; Allen, S. L.; Ahn, J. -W.; McLean, A. G.; Unterberg, E. A.; Hillis, D. L.; Fenstermacher, M. E.; Meyer, W. H.

    2014-11-15

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

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

    NASA Astrophysics Data System (ADS)

    Briesemeister, A. R.; Isler, R. C.; Allen, S. L.; Ahn, J.-W.; McLean, A. G.; Unterberg, E. A.; Hillis, D. L.; Fenstermacher, M. E.; Meyer, W. H.

    2015-08-01

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

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

    DOE PAGES

    Covele, Brent; Kotschenreuther, M.; Mahajan, S.; ...

    2017-06-23

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

  8. 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.

  9. 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.

  10. 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.

  11. A Sensitivity Analysis of SOLPS Plasma Detachment

    NASA Astrophysics Data System (ADS)

    Green, D. L.; Canik, J. M.; Eldon, D.; Meneghini, O.; AToM SciDAC Collaboration

    2016-10-01

    Predicting the scrape off layer plasma conditions required for the ITER plasma to achieve detachment is an important issue when considering divertor heat load management options that are compatible with desired core plasma operational scenarios. Given the complexity of the scrape off layer, such predictions often rely on an integrated model of plasma transport with many free parameters. However, the sensitivity of any given prediction to the choices made by the modeler is often overlooked due to the logistical difficulties in completing such a study. Here we utilize an OMFIT workflow to enable a sensitivity analysis of the midplane density at which detachment occurs within the SOLPS model. The workflow leverages the TaskFarmer technology developed at NERSC to launch many instances of the SOLPS integrated model in parallel to probe the high dimensional parameter space of SOLPS inputs. We examine both predictive and interpretive models where the plasma diffusion coefficients are chosen to match an empirical scaling for divertor heat flux width or experimental profiles respectively. This research used resources of the National Energy Research Scientific Computing Center, a DOE Office of Science User Facility, and is supported under Contracts DE-AC02-05CH11231, DE-AC05-00OR22725 and DE-SC0012656.

  12. 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.

  13. 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.

  14. 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

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. Observation of Plasma Recombination with the Negative Ions in Detached Plasma

    NASA Astrophysics Data System (ADS)

    Tonegawa, Akira; Shirota, Isao; Yoshida, Kenichi; Sugimoto, Tatunori; Kawamura, Kazutaka; Watanabe, Tsuguhiro; Ohyabu, Nobuyoshi; Takayama, Kazuo

    2000-10-01

    Detached divertors regimes are characterized by a low temperature (few eV) and high density plasma near the divertor plates. Recently, a new recombination process associated with excited hydrogen molecule, that is, molecular activated recombination (MAR), is expected to lead to an enhancement of the reduction of ion particle flux, and to modify the structure of detached recombining plasmas. In particular, negative ions play a key role in detached divertors regimes in charge exchange recombination of MAR. We present the experimental investigation of effects of the negative ions on detached plasma with MAR in the linear divertor plasma simulator, TPDSHEET-IV (Test Plasma produced by Directed current for SHEET plasma) device. The hydrogen plasma were generated with the hydrogen gas flow of 100 sccm at the discharge current of 50 A and the magnetic field of 0.7 kG. The negative ion density of hydrogen atom was measure by a probe-assisted laser photodetachment method. The reduction of the heat load to the target plate was clearly observed in hydrogen plasma with the hydrogen gas puff. At the same time, negative ions of hydrogen atom are localized in the region of cold electrons (2 eV) of the circumference of the sheet plasma. The charge exchange recombination rate of MAR is about 4 times of magnitude larger than three body recombination rate coefficients at this temperature. These experimental results suggest that the plasma recombination process comes from the negative ion of hydrogen atom.

  1. 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].

  2. Evolution of 2D Visible and VUV Divertor Emission Profiles During DIII-D H-MODE Detachment Transitions

    NASA Astrophysics Data System (ADS)

    Fenstermacher, M. E.; Jalufka, N.; Meyer, W. H.; Nilson, D. G.; Gafert, J.

    1999-07-01

    The peak heat flux to divertor target surfaces (Pdiv) must be reduced, compared with present experimental levels, before a tokamak operating in the high confinement regime (H-mode) can be extrapolated to a reactor. Partially Detached Divertor (PDD) operation (1), in which deuterium gas is injected into an H-mode plasma, reduces Pdiv by factors of 3-5 in DIII-D. A key element in the physics model of PDD operation is that carbon radiation near the X-point dissipates the energy flowing in the scrape-off-layer (SOL) before it enters the divertor (2). This allows the divertor temperature to be low, density to be high and thereby reduces the heat flux and ion particle flux to the targets both by reduced recycling and increased recombination. Previous line integrated SPRED measurements (3) and computer simulations indicated that the 155 nm (Delta)n = 0 transition of C3+ was the main power radiator from carbon during PDD operation. This paper presents the first 2D profiles of 155 nm CIV emission in any tokamak divertor. The images were obtained on DIII-D with a new tangentially viewing VUV camera (4) and established image reconstruction techniques (5). The discharges were lower single null configurations with, I p = 1.75 MA, BT = 2.1 (Tau), q95 = 3.2, Pinj = 9 MW, (kappa) = 1.9 and the (nabla)B drift toward the lower divertor. After establishing an ELMing H-mode with neutral beam injection, deuterium gas was injected at 17 Pa m3/s (130 T (ell)/s) to increase the divertor density and produce a transition to PDD operation. A practical discharge scenario is given in Ref. (2).

  3. 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.

  4. Study of near SOL decay lengths in ASDEX Upgrade under attached and detached divertor conditions

    NASA Astrophysics Data System (ADS)

    Sun, H. J.; Wolfrum, E.; Kurzan, B.; Eich, T.; Lackner, K.; Scarabosio, A.; Paradela Pérez, I.; Kardaun, O.; Faitsch, M.; Potzel, S.; Stroth, U.; the ASDEX Upgrade Team

    2017-10-01

    A database with attached, partially detached and completely detached divertors has been constructed in ASDEX Upgrade discharges in both H-mode and L-mode plasmas with Thomson Scattering data suitable for the analysis of the upstream SOL electron profiles. By comparing upstream temperature decay width, {λ }{Te,u}, with the scaling of the SOL power decay width, {λ }{q\\parallel e}, based on the downstream IR measurements, it is found that a simple relation based on classical electron conduction can relate {λ }{Te,u} and {λ }{q\\parallel e} well. The combined dataset can be described by both a single scaling and a separate scaling for H-modes and L-modes. For the single scaling, a strong inverse dependence of, {λ }{Te,u} on the separatrix temperature, {T}e,u, is found, suggesting the classical parallel Spitzer-Harm conductivity as dominant mechanism controlling the SOL width in both L-mode and H-mode over a large set of plasma parameters. This dependence on {T}e,u explains why, for the same global plasma parameters, {λ }{q\\parallel e} in L-mode is approximately twice that in H-mode and under detached conditions, the SOL upstream electron profile broadens when the density reaches a critical value. Comparing the derived scaling from experimental data with power balance, gives the cross-field thermal diffusivity as {χ }\\perp \\propto {T}e{1/2}/{n}e, consistent with earlier studies on Compass-D, JET and Alcator C-Mod. However, the possibility of the separate scalings for different regimes cannot be excluded, which gives results similar to those previously reported for the H-mode, but here the wider SOL width for L-mode plasmas is explained simply by the larger premultiplying coefficient. The relative merits of the two scalings in representing the data and their theoretical implications are discussed.

  5. Plasma Detachment Study in VASIMR

    NASA Astrophysics Data System (ADS)

    Ilin, A. V.; Díaz, F. R. Chang; Squire, J. P.; Breizman, B. N.; Novakovski, S. V.; Sagdeev, R. Z.

    2000-10-01

    We present kinetic and MHD simulations of plasma detachment in the exhaust of the Variable Specific Impulse Magnetoplasma Rocket (VASIMR). The detachment is associated with a transition from subalfvenic to superalfvenic plasma flow in the magnetic nozzle. As a result, the kinetic energy of the outgoing plasma flow is greater than the magnetic field energy in the exhaust area, so that the plasma is no longer confined by the magnetic field. We model the outgoing plasma flow under the assumptions that the plasma is collisionless and has a constant electron temperature. Particle simulations show that the ion motion may become nonadiabatic in the exhaust area as the magnetic field decreases downstream. This effect should facilitate the detachment.

  6. 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.

  7. 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

  8. 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.

  9. 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.

  10. Influence of expanding and contracting magnetic field configurations on detached plasma formation in a linear plasma device

    NASA Astrophysics Data System (ADS)

    Hayashi, Y.; Ohno, N.; Kajita, S.; Tanaka, H.

    2017-06-01

    We investigated the effects of magnetic field structure on detached plasma formation by simulating magnetically expanding and contracting plasma in a linear plasma device. The present study helps to characterize the geometries of a conventional poloidal divertor and advanced divertors, e.g., super-X divertor. The total ion particle flux measured with a large-diameter target plate dramatically changed under the detached plasma condition compared to that in attached plasma. Under the detached plasma condition, the magnetically expanding plasma clearly exhibited a significant influence on the degradation of detached plasma formation. Further, the magnetically contracting plasma slightly enhanced the electron-ion recombination (EIR) processes. By changing the magnetic field structure from contraction to expansion, the electron density (ne) decreased and the electron temperature (Te) increased upstream from the recombination front, leading to the degradation of the EIR processes. The effect of the decrease in parallel flow velocity under the magnetically contracting plasma on the plasma detachment was not observed because the driven flow due to pressure gradient compensated the effect.

  11. Experimental results from detached plasmas in TFTR

    SciTech Connect

    Strachan, J.D.; Boody, F.P.; Bush, C.E.; Cohen, S.A.; Grek, B.; Grisham, L.; Jobes, F.C.; Johnson, D.W.; Mansfield, D.K.; Medley, S.S.

    1986-10-01

    Detached plasmas are formed in TFTR which have the principal property of the boundary to the high temperature plasma core being defined by a radiating layer. This paper documents the properties of TFTR ohmic-detached plasmas with a range of plasma densities at two different plasma currents.

  12. Comprehensive 2D measurements of radiative divertor plasmas in DIII-D

    SciTech Connect

    Fenstermacher, M.E.; Wood, R.D.; Allen, S.L.; Hill, D.N.

    1997-07-01

    This paper presents a comparison of the total radiated power profile and impurity line emission distributions in the SOL and divertor of DIII-D. This is done for ELMing H-mode plasmas with heavy deuterium injection (Partially Detached Divertor operation, PDD) and those without deuterium puffing. Results are described from a series of dedicated experiments performed on DIII-D to systematically measure the 2-D (R,Z) structure of the divertor plasma. The discharges were designed to optimize measurements with new divertor diagnostics including a divertor Thomson scattering system. Discharge sequences were designed to produce optimized data sets against which SOL and divertor theories and simulation codes could be benchmarked. During PDD operation the regions of significant radiated power shift from the inner divertor leg and SOL to the outer leg and X-point regions. D{alpha} emission shifts from the inner strikepoint to the outer strikepoint. Carbon emissions (visible CII and CIII) shift from the inner SOL near the X-point to a distributed region from the X-point to partially down the outer leg during moderate D2 puffing. In heavy puffing discharges the carbon emission coalesces on the outer separatrix near the X-point and for very heavy puffing it appears inside the last closed flux surface above the X-point. Calibrated spectroscopic measurements indicate that hydrogenic and carbon radiation can account for all of the radiated power. L{alpha} and CIV radiation are comparable and when combined account for as much as 90% of the total radiated power along chords viewing the significant radiating regions of the outer leg.

  13. 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.

  14. Detachment of helium plasmas in DIII-D

    NASA Astrophysics Data System (ADS)

    Briesemeister, A. R.; Canik, J. M.; Isler, R. C.; Hillis, D. L.; Shafer, M. W.; McLean, A. G.; Makowski, M. A.; Fenstermacher, M. E.; Meyer, W. H.; Moser, A. L.; Wang, H.; Watkins, J. G.

    2016-10-01

    In low power, Ohmic plasmas, the high recycling regime typically seen as core density is increased in deuterium plasmas, was not observed in helium plasmas. When neutral helium beam heating was added, the high recycling regime returned as predicted using the SOLPS code. Modeling shows that increasing input power reduces neutral penetration across the separatrix, reducing radiation from the confined plasma and allowing a larger fraction of the input power to be carried into the scrape off layer. Detailed measurements including divertor ion and electron temperatures and densities and radiation emission patterns are presented at a variety of heating levels including both H and L-mode plasmas. An understanding of the differences between detachment in the helium and deuterium plasmas will be important for understanding how divertor conditions in the helium startup phase of ITER will differ from those in deuterium operation. Work supported by the US DOE under DE-AC05-00OR22725, DE-AC52-07NA27344, DE-FC02-04ER54698, DE-AC04-94AL85000.

  15. Particle Simulations of Plasma Detachment in VASIMR

    NASA Technical Reports Server (NTRS)

    Ilin, A. V.; ChangDiaz, F. R.; Squire, J. P.; Breizman, B. N.; Carter, M. D.; Novakovski, S. V.

    1999-01-01

    The particle simulations in a Variable Specific Impulse Magnetoplasma Rocket (VASIMR) currently include self-consistent calculation of. 1) stationary magnetic field in plasma, 2) ion density and velocity, 3) ion-cyclotron radio-frequency heating, 4) ambipolar electric field. The assumptions of quasineutral and collissionless plasma are based on the range of operating VASIMR parameters. The main motivation for the particle simulation in VASIMR is plasma detachment from the magnetic field in the exhaust area. The plasma detachment is caused mainly by the Larmor radius increase. The plasma beta effect on detachment is observed and investigated as well. The results of particle simulations are compared with those from MHD simulations.

  16. Particle Simulations of Plasma Detachment in VASIMR

    NASA Technical Reports Server (NTRS)

    Ilin, A. V.; ChangDiaz, F. R.; Squire, J. P.; Breizman, B. N.; Carter, M. D.; Novakovski, S. V.

    1999-01-01

    The particle simulations in a Variable Specific Impulse Magnetoplasma Rocket (VASIMR) currently include self-consistent calculation of. 1) stationary magnetic field in plasma, 2) ion density and velocity, 3) ion-cyclotron radio-frequency heating, 4) ambipolar electric field. The assumptions of quasineutral and collissionless plasma are based on the range of operating VASIMR parameters. The main motivation for the particle simulation in VASIMR is plasma detachment from the magnetic field in the exhaust area. The plasma detachment is caused mainly by the Larmor radius increase. The plasma beta effect on detachment is observed and investigated as well. The results of particle simulations are compared with those from MHD simulations.

  17. Detached plasma in Saturn's front side magnetosphere

    NASA Technical Reports Server (NTRS)

    Goertz, C. K.

    1983-01-01

    Plasma observations in the outer front side Saturnian magnetosphere are discussed which indicate the existence of dense flux tubes outside the plasma sheets. It is suggested that flux tubes are detached from the plasma sheet by a centifugally driven flute instability. The same instability leads to a dispersal of Titan-injected plasma. It is shown that the detached flux tubes will probably break open as they convect into the nightside magnetotail and lose their content in the form of a planetary wind.

  18. On plasma detachment in propulsive magnetic nozzles

    SciTech Connect

    Ahedo, Eduardo; Merino, Mario

    2011-05-15

    Three detachment mechanisms proposed in the literature (via resistivity, via electron inertia, and via induced magnetic field) are analyzed with an axisymmetric model of the expansion of a small-beta, weakly collisional, near-sonic plasma in a diverging magnetic nozzle. The model assumes cold, partially magnetized ions and hot, isothermal, fully magnetized electrons. Different conditions of the plasma beam at the nozzle throat are considered. A central feature is that a positive thrust gain in the nozzle of a plasma thruster is intimately related to the azimuthal current in the plasma being diamagnetic. Then, and contrary to existing expectations, the three aforementioned detachment mechanisms are divergent, that is, the plasma beam diverges outwards of the guide nozzle, further hindering its axial expansion and the thrust efficiency. The rate of divergent detachment is quantified for the small-parameter range of the three mechanisms. Alternative mechanisms for a convergent detachment of the plasma beam are suggested.

  19. Plasma detachment study of high density helium plasmas in the Pilot-PSI device

    NASA Astrophysics Data System (ADS)

    Hayashi, Y.; Ješko, K.; van der Meiden, H. J.; Vernimmen, J. W. M.; Morgan, T. W.; Ohno, N.; Kajita, S.; Yoshikawa, M.; Masuzaki, S.

    2016-12-01

    We have investigated plasma detachment phenomena of high-density helium plasmas in the linear plasma device Pilot-PSI, which can realize a relevant ITER SOL/Divertor plasma condition. The experiment clearly indicated plasma detachment features such as drops in the plasma pressure and particle flux along the magnetic field lines that were observed under the condition of high neutral pressure; a feature of flux drop was parameterized using the degree of detachment (DOD) index. Fundamental plasma parameters such as electron temperature (T e) and electron density in the detached recombining plasmas were measured by different methods: reciprocating electrostatic probes, Thomson scattering (TS), and optical emission spectroscopy (OES). The T e measured using single and double probes corresponded to the TS measurement. No anomalies in the single probe I-V characteristics, observed in other linear plasma devices [16, 17, 36], appeared under the present condition in the Pilot-PSI device. A possible reason for this difference is discussed by comparing the different linear devices. The OES results are also compared with the simulation results of a collisional radiative (CR) model. Further, we demonstrated more than 90% of parallel particle and heat fluxes were dissipated in a short length of 0.5 m under the high neutral pressure condition in Pilot-PSI.

  20. 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.

  1. On detachment asymmetry and stability

    NASA Astrophysics Data System (ADS)

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

    2017-07-01

    Three issues related to the physics of divertor detachment are analyzed in detail: the criteria for the onset of complete divertor detachment, the role of neutrals in "symmetryzation" of detachment in the inner and outer divertors, and the transition to divertor detachment. The results of comprehensive 2D numerical simulations with the SOLPS4.3 package are compared with some experimental data and predictions based on simplified analytical models. It is shown that it is the ratio of the upstream plasma pressure to the specific power flux entering the recycling region that controls the local onset of detachment on a specific flux tube. Pup/qreclycl≥(Pup/qreclycl) crit remains the valid criterion also in the presence of seeded impurity, if the impurity radiation and hydrogen recycling regions are spatially separated. Detailed analysis indicates that the reverse plasma flow forming on the most heat loaded flux tubes in the outer divertor under the influence of the neutrals coming from the deeply detached inner divertor plays the key role in the detachment "symmetryzation" and allows the outer divertor to reach the detached regime. Finally, it is demonstrated that a gradual increase of the perpendicular heat transport in the edge plasma during transition to the detached regime can make this transition bifurcation-like.

  2. 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.

  3. 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.

  4. A 0D stationary model for the evaluation of the degree of detachment on the divertor plates

    NASA Astrophysics Data System (ADS)

    Siccinio, M.; Fable, E.; Lackner, K.; Scarabosio, A.; Wenninger, R. P.; Zohm, H.

    2016-12-01

    A 0D analytical stationary model to estimate the degree of detachment on the divertor plates is derived, starting from previous works of Igitkhanov (Igitkhanov et al 1994 21st EPS (Montpellier) ECA v.18B, 1995 22th EPS (Bournemouth) ECA v.19C, Igitkhanov 2014 KIT Sci. Rep. 7661). It accounts for heat convection, conduction, impurity radiation and contains a simplified balance for the neutrals. The upstream particle flux {{ Γ }up} (or, alternatively, the density upstream n up ), the heat flux in the scrape-off layer (SOL) q up and the impurity concentrations are required as input, whereas the temperatures at the plate and at the separatrix are not fixed a priori. The routine has been mainly developed for system codes, or more in general for scopy studies in the framework of the preliminary design phase for the reactor DEMO, its simplicity being therefore justified by the purpose of keeping the calculation times as low as possible. A calibration against a more detailed 1D routine (Kallenbach et al 2016 Plasma Phys. Control. Fusion 58 045013) has been performed, finding a reasonable agreement. Furthermore, a coupling of the model with the 1.5D transport code ASTRA (Pereverzev 1991 IPP Report 5/42, Fable et al 2013 Plasma Phys. Control. Fusion 55 124028) to illustrate its possible usage is also presented.

  5. 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.

  6. 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.

  7. 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.

  8. 3D plasma turbulence and neutral simulations using the Hermes model in BOUT + + : a study of linear devices and the tokamak edge and divertor region

    NASA Astrophysics Data System (ADS)

    Leddy, Jarrod; Dudson, Ben

    2016-10-01

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

  9. Inferring divertor plasma properties from hydrogen Balmer and Paschen series spectroscopy in JET-ILW

    NASA Astrophysics Data System (ADS)

    Lomanowski, B. A.; Meigs, A. G.; Sharples, R. M.; Stamp, M.; Guillemaut, C.; Contributors, JET

    2015-11-01

    A parametrised spectral line profile model is formulated to investigate the diagnostic scope for recovering plasma parameters from hydrogenic Balmer and Paschen series spectroscopy in the context of JET-ILW divertor plasmas. The separate treatment of Zeeman and Stark contributions in the line model is tested against the PPP-B code which accounts for their combined influence on the spectral line shape. The proposed simplified model does not fully reproduce the Stark-Zeeman features for the α and β transitions, but good agreement is observed in the line width and wing profiles, especially for n  >  5. The line model has been applied to infer radial density profiles in the JET-ILW divertor with generally good agreement between the D 5\\to 2 , 5\\to 3 , 6\\to 2 , 7\\to 2 and 9\\to 2 lines for high recycling and detached conditions. In an L-mode detached plasma pulse the Langmuir probe measurements typically underestimated the density by a factor 2-3 and overestimated the electron temperature by a factor of 5-10 compared to spectroscopically derived values. The line model is further used to generate synthetic high-resolution spectra for low-n transitions to assess the potential for parameter recovery using a multi-parametric fitting technique. In cases with 4 parameter fits with a single Maxwellian neutral temperature component the D 4\\to 3 line yields the best results with parameter estimates within 10% of the input values. For cases with 9 parameter fits inclusive of a multi-component neutral velocity distribution function the quality of the fits is degraded. Simultaneous fitting of the D 3\\to 2 and 4\\to 3 profiles improves the fit quality significantly, highlighting the importance of complementary spectroscopic measurements for divertor plasma emission studies.

  10. DiMES studies of temperature dependence of carbon erosion and re-deposition in the lower divertor of DIII-D under detachment

    NASA Astrophysics Data System (ADS)

    Rudakov, D. L.; Jacob, W.; Krieger, K.; Litnovsky, A.; Philipps, V.; West, W. P.; Wong, C. P. C.; Allen, S. L.; Bastasz, R. J.; Boedo, J. A.; Brooks, N. H.; Boivin, R. L.; DeTemmerman, G.; Fenstermacher, M. E.; Groth, M.; Hollmann, E. M.; Lasnier, C. J.; McLean, A. G.; Moyer, R. A.; Stangeby, P. C.; Wampler, W. R.; Watkins, J. G.; Wienhold, P.; Whaley, J.

    2007-03-01

    A strong effect of a moderately elevated surface temperature on net carbon deposition and deuterium co-deposition in the DIII-D divertor was observed under detached conditions. A graphite DiMES sample with a 2 mm wide, 18 mm deep gap lined with silicon catcher plates was exposed to lower-single-null (LSN) L-mode plasmas first at room temperature, and then pre-heated to 200 °C by a built-in electrical heater. At the elevated temperature, deuterium co-deposition in the gap was reduced by an order of magnitude. At the plasma-facing surface of the pre-heated sample net carbon erosion was measured at a rate of 3 nm s-1, whereas without pre-heating net deposition is normally observed under detachment. In a related experiment three sets of molybdenum mirrors recessed 2 cm below the divertor floor were exposed to identical LSN ELMy H-mode discharges. The first set of mirrors exposed at ambient temperature exhibited net carbon deposition at a rate of up to 3.7 nm s-1 and suffered a significant drop in reflectivity. In contrast, two other mirror sets exposed at elevated temperatures between 90 and 175 °C exhibited practically no carbon deposition and their optical reflectivity in the wavelength range above 500 nm was essentially preserved.

  11. 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.

  12. 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

  13. 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.

  14. Erosion/redeposition analysis : status of modeling and code validation for semi-detached tokamak edge plasmas.

    SciTech Connect

    Brooks, J. N.

    1999-01-19

    We are analyzing erosion and tritium codeposition for ITER, DIII-D, and other devices with a focus on carbon divertor and metallic wall sputtering, for detached and semi-detached edge plasmas. Carbon chemical-sputtering hydrocarbon-transport is computed in detail using upgraded models for sputtering yields, species, and atomic and molecular processes. For the DIII-D analysis this includes proton impact and dissociative recombination for the full methane and higher hydrocarbon chains. Several mixed material (Si-C doping and Be/C) effects on erosion are examined. A semi-detached reactor plasma regime yields peak net wall erosion rates of {approximately}1.0 (Be), {approximately}0.3 (Fe), and {approximately}0.01 (W) cm/burn-yr, and {approximately}50 cm/burn-yr for a carbon divertor. Net carbon erosion is dominated by chemical sputtering in the {approximately}1-3 eV detached plasma zone. Tritium codeposition in divertor-sputtered redeposited carbon is high ({approximately}10-20 g-T/1000 s ). Silicon and beryllium mixing tends to reduce carbon erosion. Initial hydrocarbon transport calculations for the DIII-D DiMES-73 detached plasma experiment show a broad spectrum of redeposited molecules with {approximately}90% redeposition fraction.

  15. Plasma Detachment Mechanisms in Propulsive Magnetic Nozzles

    DTIC Science & Technology

    2013-03-07

    Merino and E. Ahedo, Physics of Plasmas 20, 023502 (2013). 26E. Ahedo, J. Gallardo , and M. Mart́ınez-Sánchez, Physics of Plasmas 10, 3397 (2003...Wiley-Blackwell, 2005. 11Ahedo, E., Gallardo , J., and Mart́ınez-Sánchez, M., “Effects of the radial-plasma wall interaction on the axial Hall...detachment References [1] Ahedo, E., Gallardo , J. M., & Mart́ınez-Sánchez, M. (2003). Effects of the radial-plasma wall interaction on the axial Hall

  16. 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.

  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. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. Observation of mutual neutralization in detached plasma

    NASA Astrophysics Data System (ADS)

    Akira, Tonegawa; Isao, Shirota; Ken'ichi, Yoshida; Masataka, Ono; Kazutaka, Kawamura; Tuguhiro, Watanabe; Nobuyoshi, Ohyabu; Hajime, Suzuki; Kazuo, Takayama

    2001-10-01

    Mutual neutralization in collisions between negative ions and positive ions in molecular activated recombination (MAR) has been observed in a high density magnetized sheet plasma source TPDSHEET-IV(Test Plasma produced by Directed current for SHEET plasma) device. Measurements of the negative ion density of hydrogen atom, the electron density, electron temperature, and the heat load to the target plate were carried out in hydrogen high density plasma with hydrogen gas puff. A cylindrical probe made of tungsten ( 0.4 x 2 cm) was used to measure the spatial profiles of H- by a probe-assisted laser photodetachment The Balmer spectra of visible light emission from hydrogen or helium atoms were detected in front of the target plate. A small amount of secondary hydrogen gas puffing into a hydrogen plasma reduced strongly the heat flux to the target and increased rapidly the density of negative ions of hydrogen atom in the circumference of the plasma, while the conventional radiative and three-body recombination processes were disappeared. These results can be well explained by taking the charge exchange recombination of MAR in the detached plasma into account.

  5. 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.

  6. 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.

  7. Magnetic Nozzle and Plasma Detachment Experiment

    NASA Technical Reports Server (NTRS)

    Chavers, Gregory; Dobson, Chris; Jones, Jonathan; Martin, Adam; Bengtson, Roger D.; Briezman, Boris; Arefiev, Alexey; Cassibry, Jason; Shuttpelz, Branwen; Deline, Christopher

    2006-01-01

    High power plasma propulsion can move large payloads for orbit transfer (such as the ISS), lunar missions, and beyond with large savings in fuel consumption owing to the high specific impulse. At high power, lifetime of the thruster becomes an issue. Electrodeless devices with magnetically guided plasma offer the advantage of long life since magnetic fields confine the plasma radially and keep it from impacting the material surfaces. For decades, concerns have been raised about the plasma remaining attached to the magnetic field and returning to the vehicle along the closed magnetic field lines. Recent analysis suggests that this may not be an issue of the magnetic field is properly shaped in the nozzle region and the plasma has sufficient energy density to stretch the magnetic field downstream. An experiment was performed to test the theory regarding the Magneto-hydrodynamic (MHD) detachment scenario. Data from this experiment will be presented. The Variable Specific Impulse Magnetoplasma Rocket (VASIMR) being developed by the Ad Astra Rocket Company uses a magnetic nozzle as described above. The VASIMR is also a leading candidate for exploiting an electric propulsion test platform being considered for the ISS.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. Kinetic effects on parallel heat flow and ionization rate in divertor plasmas

    NASA Astrophysics Data System (ADS)

    Allais, Fabrice; Kim, Chang-Geun; Alouani Bibi, Fathallah; Matte, Jean-Pierre; Stotler, Daren; Rognlien, Thomas

    2004-11-01

    1-D simulations of parallel heat flow in divertor plasmas, with and without recycling are made with the UEDGE fluid code, comparing runs using classical flux limited heat flow to nonlocal heat transport [1], now implemented in UEDGE. Comparative simulations are made with the electron kinetic code FPI. For the latter, we prescribe the power input source, which emulates cross field transport, to be identical to that of our UEDGE runs. But, the temperature profile computed by FPI is found to depend very strongly on the assumed velocity dependence of this source, even if the integrated power is the same. The atomic hydrogen ionization module in FPI uses cross sections such that, for Maxwellian plasmas, the rates are the same as those used by UEDGE and DEGAS; this is necessary because step-wise ionization is dominant. There is strong enhancement of the total ionization rate (including stepwise ionization) in cold, detached plasmas, due to nonlocal transport effects. [1] F. Alouani Bibi and J.P. Matte, Phys. Rev. E 66, 066414 (2002)

  13. Magnetic Field Structure near the Plasma Boundary in Helical Systems and Divertor Tokamaks

    NASA Astrophysics Data System (ADS)

    Nagasaki, Kazunobu; Itoh, Kimitaka

    1990-07-01

    The magnetic field structure of the scrape-off layer (SOL) region in both helical systems and divertor tokamaks is studied numerically by using model fields. The connection length of the field line to the wall, L, is calculated. In helical systems, L has logarithmic properties in the SOL region. The effect of axisymmetric fields on the field structure is determined. In divertor tokamaks, L also has logarithmic properties near the separatrix. Even when the perturbations which resonate to rational surfaces near the plasma boundary are added, the logarithmic properties are not changed. The connection length of torsatron/helical-heliotron systems is compared with that of divertor tokamaks.

  14. Time evolution of the particle and heat flux of the detached plasma

    NASA Astrophysics Data System (ADS)

    Pianpanit, Theerasarn; Ishiguro, Seiji; Hasegawa, Hiroki

    2016-10-01

    The detached plasma is a regime when the particle and heat flux of the plasma are largely reduced before reaching the divertor target. Linear devices experiment data show that when the neutral gas pressure in front of the target increases the heat flux to the target largely decreases. The 1D-3V particle simulation with Monte Carlo collision and cumulative scattering angle Coulomb collision has been developed to study the kinetic effect of the detached plasma. The simulation was performed with the constant temperature and pressure of neutral gas in front of the target. A large decrease in the electron temperature from 5eV to below 1 eV follows a large decrease in the ion temperature inside the neutral gas area in the case with high neutral gas pressure in front of the target. The energy flux at the target decreases in the process of attaining the detached state. This work was performed with the support and under the auspices of the NIFS Collaboration Research programs NIFS14KNXN279 and NIFS14KNSS059.

  15. Magnetic Detachment and Plume Control in Escaping Magnetized Plasma

    SciTech Connect

    P. F. Schmit and N. J. Fisch

    2008-11-05

    The model of two-fluid, axisymmetric, ambipolar magnetized plasma detachment from thruster guide fields is extended to include plasmas with non-zero injection angular velocity profiles. Certain plasma injection angular velocity profiles are shown to narrow the plasma plume, thereby increasing exhaust efficiency. As an example, we consider a magnetic guide field arising from a simple current ring and demonstrate plasma injection schemes that more than double the fraction of useful exhaust aperture area, more than halve the exhaust plume angle, and enhance magnetized plasma detachment.

  16. Magnetic Detachment and Plume Control in Escaping Magnetized Plasma

    SciTech Connect

    P. F. Schmit and N. J. Fisch

    2008-11-05

    The model of two-fluid, axisymmetric, ambipolar magnetized plasma detachment from thruster guide fields is extended to include plasmas with non-zero injection angular velocity profiles. Certain plasma injection angular velocity profiles are shown to narrow the plasma plume, thereby increasing exhaust efficiency. As an example, we consider a magnetic guide field arising from a simple current ring and demonstrate plasma injection schemes that more than double the fraction of useful exhaust aperture area, more than halve the exhaust plume angle, and enhance magnetized plasma detachment.

  17. 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

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

    SciTech Connect

    Petrie, T W; Wade, M R; Brooks, N H; Fenstermacher, M E; Groth, M; Hyatt, A W; Isler, R C; Lasnier, C J; Leonard, A W; Mahdavi, M A; Porter, G D; Schaffer, M J; Watkins, J G; West, W P

    2006-05-18

    A radiating divertor approach was successfully applied to high performance 'hybrid' plasmas [M.R. Wade, et al., Proc. 20th IAEA Fusion Energy Conf., Vilamoura, (2004)]. Our technique included: (1) injecting argon near the outer divertor target, (2) enhancing the plasma flow into the inner and outer divertors by a combination of particle pumping and deuterium gas puffing upstream of the divertor targets, and (3) isolating the inner divertor from the outer by a structure in the private flux region. Good hybrid conditions were maintained, as the peak heat flux at the outer divertor target was reduced by a factor of 2.5; the peak heat flux at the inner target decreased by 20%. This difference was caused by a higher concentration of argon at the outer target than at the inner target. Argon accumulation in the main plasma was modest (n{sub AR}/n{sub e} {le}0.004 on axis), although the argon profile was more peaked than the electron profile.

  19. 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.

  20. 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.

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

    SciTech Connect

    Ryutov, D D; Cohen, R H

    2007-05-16

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

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

    NASA Astrophysics Data System (ADS)

    Lee, J. S.; Orlov, D. M.; Moyer, R. A.; Bykov, I.; Evans, T. E.; Wu, W.; Lyons, B. C.; Sugiyama, L. E.

    2016-10-01

    Magnetic field perturbations (RMPs) split the strike points in divertor tokamaks. This splitting is measured using fast imaging of filtered visible light from the divertor. We compare the observed splitting during n=3 RMP experiments to vacuum and plasma response modeling to determine if the measured splitting provides a sensitive diagnostic for the plasma response to the RMP. We also investigate the sensitivity of the computed plasma response to uncertainties in the initial 2D equilibrium. Strike point splitting was also observed in ELMing H-mode without the RMP, possibly due to n=1 error- and error-field correction fields. We compare the measured splitting during ELMs to linear plasma response modeling of the divertor footprints, and to nonlinear M3D ELM simulations. Work supported by U.S. DOE under Grant Numbers DE-FG02-07ER54917, DE-FG02-05ER54809.

  3. Exposures of tungsten nanostructures to divertor plasmas in DIII-D

    SciTech Connect

    Rudakov, D. L.; Wong, C. P. C.; Doerner, R. P.; Wright, G. M.; Abrams, T.; Baldwin, M. J.; Boedo, J. A.; Briesemeister, A. R.; Chrobak, C. P.; Guo, H. Y.; Hollmann, E. M.; McLean, A. G.; Fenstermacher, M. E.; Lasnier, C. J.; Leonard, A. W.; Moyer, R. A.; Pace, D. C.; Thomas, D. M.; Watkins, J. G.

    2016-01-22

    Tungsten nanostructures (W-fuzz) prepared in the PISCES-A linear device have been found to survive direct exposure to divertor plasmas in DIII-D. W-fuzz was exposed in the lower divertor of DIII-D using the divertor material evaluation system. Two samples were exposed in lower single null (LSN) deuterium H-mode plasmas. The first sample was exposed in three discharges terminated by vertical displacement event disruptions, and the second in two discharges near the lowered X-point. More recently, three samples were exposed near the lower outer strike point in predominantly helium H-mode LSN plasmas. In all cases, the W-fuzz survived plasma exposure with little obvious damage except in the areas where unipolar arcing occurred. In conclusion, arcing is effective in W-fuzz removal, and it appears that surfaces covered with W-fuzz can be more prone to arcing than smooth W surfaces.

  4. Development of ion source for simulation of edge localized mode in divertor plasma.

    PubMed

    Daibo, A; Okamoto, A; Takahashi, H; Kumagai, T; Takahashi, T; Tsubota, S; Kitajima, S

    2014-02-01

    A helium ion beam is injected into a linear plasma device for the development of an ion beam source simulating high energy particle flux in divertor plasma. Beam current density more than 10 mA/cm(2) is extracted. Measurement of beam currents indicates that the beam is transported along the linear device and reaches to the downstream end plate.

  5. Comparison of radiating divertor behaviour in single-null and double-null plasmas in DIII-D

    SciTech Connect

    Petrie, T W; Brooks, N. H.; Fenstermacher, M. E.; Groth, M.; Hyatt, A. W.; Isler, R.C.; Lasnier, C. J.; Leonard, A. W.; Porter, G. D.; Schaffer, M. J.; Watkins, J. G.; Wade, M R; West, W.P.

    2008-01-01

    Puff-and-pump' radiating divertor scenarios, applied to both upper single-null (SN) and double-null (DN) H-mode plasmas, result in a 30-60% increase in radiated power with little or no decrease in tau(E). Argon was injected into the private flux region of the upper divertor, and plasma flow into the upper divertor was enhanced by a combination of deuterium gas puffing upstream of the divertor targets and particle pumping at the targets. For the same constant deuterium injection rate, argon penetrated the main plasma of SNs more rapidly and reached a higher steady-state concentration when the B x del B-ion drift direction was towards the divertor (V-del B up arrow) rather than away from the divertor (V-del B down arrow). We also found that the initial rate at which argon accumulated inside DN plasmas was more than twice that of comparable SN plasmas having the same B x del B-ion drift direction. In DNs, the radiated power was not shared equally between divertors during argon injection. Only when the B x del B ion drift direction was away from the divertor were both significant increases in divertor radiated power and an accumulation of argon in the divertor observed, based on spectroscopic measurements of Ar II. Our data suggest that an unbalanced DN shape where the B x del B-ion drift is directed away from the dominant divertor may provide the best chance of successfully coupling a radiating divertor approach with a higher performance H-mode plasma.

  6. 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).

  7. 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).

  8. 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).

  9. Deuterium and tritium fuelding in an ETF/INTOR plasma with divertor

    SciTech Connect

    Houlberg, W.A.; Howe, H.C.; Attenberger, S.E.

    1980-01-01

    Fueling by pellets and neutral gas in the presence of a divertor is examined with a one-dimensional multispecies transport code. Deuterium, tritium, and alpha particles are treated as independent thermal species. With an efficiently operating divertor, it becomes impossible to maintain high plasma density (anti n approx. 10/sup 14/ cm/sup -3/) with neutral gas fueling alone because of the high probability of the gas being ionized in the scrapeoff layer. Pellet fueling significantly reduces the particle load on the divertor and, with feedback control, can maintain the plasma density at the desired level. A low level of deuterium gas fueling can then be used to maintain the density of the scrapeoff layer and increase shielding against sputtered impurities. Even with an effective shielding divertor, the energy and particle fluxes to the first wall from both charged and neutral particles may be significant. The fluctuations at the plasma edge and in the scrapeoff layer introduced by the pellets cause the particle and energy fluxes to the divertor and first wall to fluctuate. This makes simulation more difficult but may be used to experimentally determine radial and parallel transport properties in the scrapeoff layer. Recommendations for further study are made.

  10. Numerical Investigation of Plasma Detachment in Magnetic Nozzle Experiments

    NASA Technical Reports Server (NTRS)

    Sankaran, Kamesh; Polzin, Kurt A.

    2008-01-01

    At present there exists no generally accepted theoretical model that provides a consistent physical explanation of plasma detachment from an externally-imposed magnetic nozzle. To make progress towards that end, simulation of plasma flow in the magnetic nozzle of an arcjet experiment is performed using a multidimensional numerical simulation tool that includes theoretical models of the various dispersive and dissipative processes present in the plasma. This is an extension of the simulation tool employed in previous work by Sankaran et al. The aim is to compare the computational results with various proposed magnetic nozzle detachment theories to develop an understanding of the physical mechanisms that cause detachment. An applied magnetic field topology is obtained using a magnetostatic field solver (see Fig. I), and this field is superimposed on the time-dependent magnetic field induced in the plasma to provide a self-consistent field description. The applied magnetic field and model geometry match those found in experiments by Kuriki and Okada. This geometry is modeled because there is a substantial amount of experimental data that can be compared to the computational results, allowing for validation of the model. In addition, comparison of the simulation results with the experimentally obtained plasma parameters will provide insight into the mechanisms that lead to plasma detachment, revealing how they scale with different input parameters. Further studies will focus on modeling literature experiments both for the purpose of additional code validation and to extract physical insight regarding the mechanisms driving detachment.

  11. 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.

  12. Simulation of turbulence in the divertor region of tokamak edge plasma

    NASA Astrophysics Data System (ADS)

    Umansky, M. V.; Rognlien, T. D.; Xu, X. Q.

    2005-03-01

    Results are presented for turbulence simulations with the fluid edge turbulence code BOUT [X.Q. Xu, R.H. Cohen, Contr. Plas. Phys. 36 (1998) 158]. The present study is focussed on turbulence in the divertor leg region and on the role of the X-point in the structure of turbulence. Results of the present calculations indicate that the ballooning effects are important for the divertor fluctuations. The X-point shear leads to weak correlation of turbulence across the X-point regions, in particular for large toroidal wavenumber. For the saturated amplitudes of the divertor region turbulence it is found that amplitudes of density fluctuations are roughly proportional to the local density of the background plasma. The amplitudes of electron temperature and electric potential fluctuations are roughly proportional to the local electron temperature of the background plasma.

  13. Evolution of 2D deuterium and impurity radiation profiles during transitions from attached to detached divertor operation in DIII-D

    SciTech Connect

    Fenstermacher, M.E.; Allen, S.L.; Hill, D.N.

    1998-08-01

    This paper presents the detailed evolution of conditions along both the inner and outer divertor legs during the transition from attached ELMing H-mode to partially detached divertor (PDD) operation in DIII-D. Visible emission profiles in a poloidal plane show that in ELMing H-mode prior to deuterium gas injection, CIII emission peaks in the inner SOL near the X-point and deuterium emission (from ionization and recombination) peaks at the inner target plate near the inner strike point (ISP). The spatial profiles of the recombination and ionization zones, determined by forming images of the ratio of intensities from simultaneous images of D{sub {alpha}} and D{sub {gamma}} emission, show that recombination dominates the inner leg emission near the target; ionization dominates in a poloidally narrow zone upstream in the inner leg. After deuterium injection, when the PDD transition begins, the profiles of carbon visible emission show first an increase in the inner SOL near the X-point, followed by increases in emission in the lower regions of the outer leg. Deuterium emission at the transition onset decreases at th4e ISP and increases across the private flux region below the X-point. As the transition to PDD conditions proceeds the deuterium emission increases in the private flux region; recombination dominates near the floor and ionization higher near the X-point. Carbon emission appears along both divertor legs and at the X-point. In the final quasi-steady PDD state, the recombination emission in the outer leg is near the separatrix and along the target plate; emission from collisional excitation dominates in the upper part of the outer leg just below the X-point, and carbon emission is localized at the X-point. These results suggest that transport of neutral deuterium between the inner and outer divertor legs through the private flux region plays an important role in the initiation of outer leg detachment in DIII-D.

  14. Evolution of 2D deuterium and impurity radiation profiles during transitions from attached to detached divertor operation in DIII-D

    NASA Astrophysics Data System (ADS)

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

    This paper presents the detailed evolution of conditions along both the inner and outer divertor legs during the transition from attached ELMing H-mode to partially detached divertor (PDD) operation in DIII-D. Visible emission profiles in a poloidal plane show that in ELMing H-mode prior to deuterium gas injection, CIII emission peaks in the inner SOL near the X-point and deuterium emission (from ionization and recombination) peaks at the inner target plate near the inner strike point (ISP). The spatial profiles of the recombination and ionization zones, determined by forming images of the ratio of intensities from simultaneous images of D α and D γ emission, show that recombination dominates the inner leg emission near the target; ionization dominates in a poloidally narrow zone upstream in the inner leg. After deuterium injection, when the PDD transition begins, the profiles of carbon visible emission show first an increase in the inner SOL near the X-point, followed by increases in emission in the lower regions of the outer leg. Deuterium emission at the transition onset decreases at the ISP and increases across the private flux region below the X-point. As the transition to PDD conditions proceeds the deuterium emission increases in the private flux region; recombination dominates near the floor and ionization higher near the X-point. Carbon emission appears along both divertor legs and at the X-point. In the final quasi-steady PDD state, the recombination emission in the outer leg is near the separatrix and along the target plate; emission from collisional excitation dominates in the upper part of the outer leg just below the X-point, and carbon emission is localized at the X-point. These results suggest that transport of neutral deuterium between the inner and outer divertor legs through the private flux region plays an important role in the initiation of outer leg detachment in DIII-D.

  15. Sensitivity of detachment extent to magnetic configuration and external parameters

    NASA Astrophysics Data System (ADS)

    Lipschultz, Bruce; Parra, Felix I.; Hutchinson, Ian H.

    2016-05-01

    Divertor detachment may be essential to reduce heat loads to magnetic fusion tokamak reactor divertor surfaces. Yet in experiments it is difficult to control the extent of the detached, low pressure, plasma region. At maximum extent the front edge of the detached region reaches the X-point and can lead to degradation of core plasma properties. We define the ‘detachment window’ in a given position control variable C (for example, the upstream plasma density) as the range in C within which the front location can be stably held at any position from the target to the X-point; increased detachment window corresponds to better control. We extend a 1D analytic model [1] to determine the detachment window for the following control variables: the upstream plasma density, the impurity concentration and the power entering the scrape-off layer (SOL). We find that variations in magnetic configuration can have strong effects; increasing the ratio of the total magnetic field at the X-point to that at the target, {{B}×}/{{B}t} , (total flux expansion, as in the super-x divertor configuration) strongly increases the detachment window for all control variables studied, thus strongly improving detachment front control and the capability of the divertor plasma to passively accommodate transients while still staying detached. Increasing flux tube length and thus volume in the divertor, through poloidal flux expansion (as in the snowflake or x-divertor configurations) or length of the divertor, also increases the detachment window, but less than the total flux expansion does. The sensitivity of the detachment front location, z h , to each control variable, C, defined as \\partial {{z}h}/\\partial C , depends on the magnetic configuration. The size of the radiating volume and the total divertor radiation increase \\propto {{≤ft({{B}×}/{{B}t}\\right)}2} and \\propto {{B}×}/{{B}t} , respectively, but not by increasing divertor poloidal flux expansion or field line length. We

  16. Initial Study Comparing the Radiating Divertor Behavior in Single-Null and Double-Null Plasmas in DIII-D

    SciTech Connect

    Petrie, T; Brooks, N; Fenstermacher, M; Groth, M; Hyatt, A; Isler, R; Lasnier, C; Leonard, A; Porter, G; Schaffer, M; Watkins, J; Wade, M; West, W

    2007-06-27

    'Puff and pump' radiating divertor scenarios [1,2] were applied to upper SN and DN H-mode plasmas. Under similar operating conditions, argon (Ar) accumulated in the main plasma of single-null (SN) plasmas more rapidly and reached a higher steady-state concentration when the B x {del}B ion drift direction was toward the divertor than when the B x {del}B ion drift direction was out of the divertor. The initial rate that Ar accumulated inside double-null (DN) plasmas was more than twice that of comparably-prepared SNs with the same B x {del}B direction. One way to reduce power loading at the divertor targets is to 'seed' the divertor plasma with impurities that radiatively reduce the conducted power. Studies have shown that the concentration of impurities in the divertor are increased by raising the flow of deuterium ions (D{sup +}) into the divertor by a combination of upstream deuterium gas puffing and active particle exhaust at the divertor targets, i.e., puff-and-pump. An enhanced D{sup +} particle flow toward the divertor targets exerts a frictional drag on impurities, and inhibits their escape from the divertor. A puff-and-pump approach using Ar as the impurity was successfully applied in recent DIII-D experiments to SN plasmas [3] while maintaining good H-mode performance. Studies on DIII-D and other tokamaks have shown that both the direction of the toroidal magnetic field B{sub T} and the degree of magnetic balance between divertors [i.e., the degree to which the plasma shape is considered SN or DN] are important factors in determining recycling and particle pumping [4,5]. It is unclear whether the favorable results of Ref. [3] can be extended to cases with different magnetic balance and/or B{sub T} direction. We show in this paper that reversing the direction of B{sub T} or altering the divertor magnetic balance does have an impact on how plasmas behave under puff-and-pump conditions. Our study takes advantage of DIII-D's capabilities to actively pump SN and

  17. A Comparison of Radiating Divertor Behavior in Single- and Double-Null Plasmas in DIII-D

    SciTech Connect

    Petrie, T W; Brooks, N H; Fenstermacher, M E; Groth, M; Hyatt, A W; Isler, R C; Lasnier, C J; Leonard, A W; Porter, G D; Schaffer, M J; Watkins, J G; Wade, M R; West, W P

    2008-03-25

    'Puff and pump' radiating divertor scenarios, applied to both upper single-null (SN) and double-null (DN) H-mode plasmas, result in a 30-60% increase in radiated power with little or no decrease in {tau}{sub E}. Argon was injected into the private flux region of the upper divertor, and plasma flow into the upper divertor was enhanced by a combination of deuterium gas puffing upstream of the divertor targets and particle pumping at the targets. For the same constant deuterium injection rate, argon penetrated the main plasma of SNs more rapidly and reached a higher steady-state concentration when the Bx{del}B-ion drift direction was toward the divertor (V{sub {del}B{up_arrow}}) rather than away from the divertor (V{sub {del}B{down_arrow}}). We also found that the initial rate at which argon accumulated inside DN plasmas was more than twice that of comparable SN plasmas having the same Bx{del}B-ion drift direction. In DNs, the radiated power was not shared equally between divertors during argon injection. Only in the divertor opposite Bx{del}B ion drift direction were both significant increases in divertor radiated power and an accumulation of argon, based on spectroscopic measurements of ArII, observed. Our data suggests that a DN shape that is biased in the direction away from the Bx{del}B-ion drift direction may provide the best prospect of successfully coupling a radiating divertor approach with a higher performance H-mode plasma.

  18. 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

  19. 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

  20. Effect of RMP spectrum on ELM suppression and the divertor plasma in KSTAR

    NASA Astrophysics Data System (ADS)

    Ahn, Joon-Wook; Park, J.-K.; in, Y.; Loarte, A.; Kim, J.; Jeon, Y. M.; Park, G. Y.; Choe, W.; Hong, J. H.; Hong, S. H.; Lee, H. H.; Kang, C. S.; Ko, W. H.; Yoon, S. W.

    2016-10-01

    ELM suppression by n =1 and n =2 magnetic perturbations have been robustly obtained in KSTAR, and effects of various coil configurations for applied magnetic perturbations (MPs) on ELM suppression as well as divertor plasma conditions have been investigated. The 4 toroidal and 3 poloidal sectors of internal coils allow to fully scan the phase difference (Δφ) of n =1 between different rows of coils, where it is shown that ideal plasma response can either shield or amplify applied MPs, depending on Δφ , which leads respectively to the weakening and strengthening of divertor footprint striations compared to the vacuum case. On the other hand, shielding is found to be the dominant plasma response for all possible cases of n =2 configuration (Δφ =0o and 90o, and mid-plane coil only), which weakens footprint striations. Spectra of applied MPs have been varied by changing Δφ as well as modifying the ratio of coil currents between different row of coils, e.g. IU/IL, in order to find optimal conditions for ELM suppression and divertor heat and particle flux dispersal. Effects of divertor conditions in various density and impurity levels on the ELM behavior and footprint striations are also being investigated. Work supported by the U.S. DOE, contract # DE-AC05-00OR22725.

  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. Detached scrape-off layer tokamak plasmas

    SciTech Connect

    Kesner, J.

    1995-06-01

    The equilibrium and stability of scrape-off layer plasmas are considered using a one-dimensional treatment of coupled heat conduction and pressure balance equations. It is found that, for sufficiently low-temperature and high neutral density, a region of greatly reduced power flux to the endplate can be achieved. The plasma in the vicinity of the end wall is characterized by a sharp plasma pressure gradient and a relatively low temperature, 1{lt}{ital T}{sub 0}{lt}10 eV.

  3. Fusion plasma theory. Task 3: ECRH and transport modeling in tandem mirrors and divertor physics

    NASA Astrophysics Data System (ADS)

    Emmert, G. A.

    1984-06-01

    The research performed under Tank II of this contact has focused on: (1) the coupling of an ECRH ray tracing and absorption code to a tandem mirror transport code in order to self-consistently model the temporal and spatial evolution of the plasma, and (2) the further development of semi-analytical models for plasma flow in divertors and pumped limiters. Work on these topics is briefly summarized.

  4. 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

  5. Divertor heat loads in RMP ELM controlled H-mode plasmas on DIII-D*

    SciTech Connect

    Jakubowski, M; Lasnier, C; Schmitz, O; Evans, T; Fenstermacher, M; Groth, M; Watkins, J; Eich, T; Moyer, R; Wolf, R; Baylor, L; Boedo, J; Burrell, K; Frerichs, H; deGrassie, J; Gohil, P; Joseph, I; Lehnen, M; Leonard, A; Petty, C; Pinsker, R; Reiter, D; Rhodes, T; Samm, U; Snyder, P; Stoschus, H; Osborne, T; Unterberg, B; West, W

    2008-10-13

    In this paper the manipulation of power deposition on divertor targets at DIII-D by application of resonant magnetic perturbations (RMPs) is analyzed. It has been found that heat transport shows a different reaction to the applied RMP depending on the plasma pedestal collisionality. At pedestal electron collisionality above 0.5 the heat flux during the ELM suppressed phase is of the same order as the inter-ELM in the non-RMP phase. Below this collisionality value we observe a slight increase of the total power flux to the divertor. This can be caused by much more negative potential at the divertor surface due to hot electrons reaching the divertor surface from the pedestal area and/or so called pump out effect. In the second part we discuss modification of ELM behavior due to the RMP. It is shown, that the width of the deposition pattern in ELMy H-mode depends linearly on the ELM deposited energy, whereas in the RMP phase of the discharge those patterns seem to be controlled by the externally induced magnetic perturbation. D{sub 2} pellets injected into the plasma bulk during ELM-free RMP H-mode lead in some cases to a short term small transients, which have very similar properties to ELMs in the initial RMP-on phase.

  6. Experimental Characterization of Plasma Detachment from Magnetic Nozzles

    NASA Astrophysics Data System (ADS)

    Olsen, Christopher Scott

    Magnetic nozzles, like Laval nozzles, are observed in several natural systems and have application in areas such as electric propulsion and plasma processing. Plasma flowing through these nozzles is inherently tied to the field lines and must separate for momentum redirection or particle transport to occur. Plasma detachment and associated mechanisms from a magnetic nozzle are investigated. Experimental results are presented from the plume of the VASIMRRTM VX-200 device flowing along an axisymmetric magnetic nozzle and operated at two ion energies to explore momentum dependent detachment. The argon plume expanded into a 150m3 vacuum chamber where the background pressure was low enough that charge-exchange mean-free-paths were longer than experiment scale lengths. This magnetic nozzle system is demonstrated to hydrodynamically scale up to astrophysical plasmas, particularly the solar chromosphere, implying general relevance to many systems. Plasma parameters were mapped over a large spatial range using measurements from multiple plasma diagnostics. The data show that the plume does not follow the magnetic field lines. A mapped integration of the ion flux shows the plume may be divided into three regions where 1) the plume briefly follows the magnetic flux, 2) diverges quadratically before 3) expanding with linear trajectories. Transitioning from region 1→2, the ion flux departs from the magnetic flux suggesting ion detachment. An instability forms in region 2 driving an oscillating electric field that causes ions to expand before enhancing electron cross-field transport through anomalous resistivity. Transitioning from region 2→3 the electric field dissipates, the trajectories linearize, and the plume effectively detaches. A delineation of sub-to-super Alfvenic flow aligns well with the inflection points of the linearization without a change in magnetic topology. The detachment process is best described as a two part process: First, ions detach by a breakdown of

  7. Transport calculations of chemically sputtered carbon near a plasma divertor surface

    SciTech Connect

    Brooks, J.N.

    1992-01-01

    The transport of chemically sputtered carbon near a tokamak divertor surface has been analyzed with the Monte Carlo code WBC. The code follows the motion of sputtered methane atoms and the resulting carbon and hydrocarbon derivatives. Ion transport due to the magnetic field, sheath electric field, and collisions with the plasma is computed. Redeposition fractions, impinging species type, charge state, and velocity have been analyzed. For plasma temperatures {ge} 10 eV, and for typical divertor plasma densities, local redeposition of chemically sputtered carbon approaches 100%. Redeposition fractions are lower ({approximately}80%) for lower temperatures and/or lower density. Physical sputtering of carbon due to redeposition of chemically sputtered material is low but a hydrocarbon reflection cascade due to redeposition may be high.

  8. Exposures of tungsten nanostructures to divertor plasmas in DIII-D

    DOE PAGES

    Rudakov, D. L.; Wong, C. P. C.; Doerner, R. P.; ...

    2016-01-22

    Tungsten nanostructures (W-fuzz) prepared in the PISCES-A linear device have been found to survive direct exposure to divertor plasmas in DIII-D. W-fuzz was exposed in the lower divertor of DIII-D using the divertor material evaluation system. Two samples were exposed in lower single null (LSN) deuterium H-mode plasmas. The first sample was exposed in three discharges terminated by vertical displacement event disruptions, and the second in two discharges near the lowered X-point. More recently, three samples were exposed near the lower outer strike point in predominantly helium H-mode LSN plasmas. In all cases, the W-fuzz survived plasma exposure with littlemore » obvious damage except in the areas where unipolar arcing occurred. In conclusion, arcing is effective in W-fuzz removal, and it appears that surfaces covered with W-fuzz can be more prone to arcing than smooth W surfaces.« less

  9. Coexistent Regime of H-mode with a Dense & Cold Divorter plasma in JFT-2M Closed Divertor

    NASA Astrophysics Data System (ADS)

    Sengoku, Seio; JFT-2m Group

    1996-11-01

    A possibility of stable coexistence of an H-mode with a dense & cold divertor plasma had been demonstrated using a strong gas puffing in divertor chamber at the lower density limit for the H-mode transition ( 2x10^19m-3: the regime of no spontaneous dense & cold state) by modifying the divertor shape of JFT-2M to a closed configuration.footnote S. Sengoku et al., Bull. Amer. Phys. Soc. 40, 1675 (1995) A build up of neutral pressure occurs only in the divertor chamber without degrading the H-factor, and the divertor plasma results in a dense & cold state (n_e=1.3 2.5x 10^19m-3, T_e=4 15eV). In order to improve baffling effect and to extend operational regime of the coexistence, the divertor baffle plates of JFT-2M had been modified from relatively wide baffle-opening to more closed one. Studies on fueling and exhaustion, particle control, neutral buildup scaling and SOL plasma behaviors are being carried out with the modified divertor shape.

  10. Interpretation of radiative divertor studies with impurity seeding in type-I ELMy H-mode plasmas in JET-ILW using EDGE2D-EIRENE

    NASA Astrophysics Data System (ADS)

    Jaervinen, A. E.; Groth, M.; Airila, M.; Belo, P.; Beurskens, M.; Brezinsek, S.; Clever, M.; Corrigan, G.; Devaux, S.; Drewelow, P.; Eich, T.; Giroud, C.; Harting, D.; Huber, A.; Jachmich, S.; Lawson, K.; Lipschultz, B.; Maddison, G.; Maggi, C.; Makkonen, T.; Marchetto, C.; Marsen, S.; Matthews, G. F.; Meigs, A. G.; Moulton, D.; Stamp, M. F.; Wiesen, S.; Wischmeier, M.

    2015-08-01

    Nitrogen seeded JET-ILW H-mode plasmas have been investigated with EDGE2D-EIRENE. The simulations reproduce the experimentally observed factor of 10 reduction in the outer target power deposition when the normalized divertor radiation, Praddiv/PSOL, increases from the unseeded levels of 15% up to the 50% levels required for detachment. At these radiation levels, nitrogen is predicted dominate the total radiation with a contribution of 85%, consistent with previous measurements in JET-C. Due to the low radiative potential of nitrogen at the electron temperatures above 100 eV, more than 80% of the radiation is predicted to occur in the scrape-off layer, making nitrogen a suitable divertor radiator for typical JET divertor conditions with Te around 30 eV. The simulations reproduce the experimentally observed particle flux reduction at the low-field side target without the need for strong recombination. This is due to strong impurity radiation reducing the power levels entering the deuterium ionization front.

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

    NASA Astrophysics Data System (ADS)

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

    2016-11-01

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

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

    SciTech Connect

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

    2016-11-15

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

  13. 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.

  14. Plasma flow in peripheral region of detached plasma in linear plasma device

    SciTech Connect

    Hayashi, Y. Ohno, N.; Kajita, S.; Tanaka, H.

    2016-01-15

    A plasma flow structure is investigated using a Mach probe under detached plasma condition in a linear plasma device NAGDIS-II. A reverse flow along the magnetic field is observed in a steady-state at far-peripheral region of the plasma column in the upstream side from the recombination front. These experimental results indicate that plasma near the recombination front should strongly diffuse across the magnetic field, and it should be transported along the magnetic field in the reverse flow direction. Furthermore, bursty plasma density fluctuations associated with intermittent convective plasma transport are observed in the far-peripheral region of the plasma column in both upstream and downstream sides from the recombination front. Such a nondiffusive transport can contribute to the intermittent reverse plasma flow, and the experimental results indicate that intermittent transports are frequently produced near the recombination front.

  15. Influence of the divertor plate material on the plasma performance in DEMO

    NASA Astrophysics Data System (ADS)

    Ivanova-Stanik, I.; Zagórski, R.

    2015-07-01

    This paper describes integrated numerical modeling applied to DEMO discharges with different divertor plate materials (nickel, molybdenum and tungsten) in a slab geometry, using the COREDIV code. Calculations are performed for the inductive DEMO scenario with argon seeding. Simulations shown that for Mo and Ni, it is possible to achieve H-mode plasma operation (power to SOL > power threshold for L-H transition) with acceptable levels of power to the target plates.

  16. 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.

  17. Status of Magnetic Nozzle and Plasma Detachment Experiment

    SciTech Connect

    Chavers, D. Gregory; Dobson, Chris; Jones, Jonathan; Lee, Michael; Martin, Adam; Gregory, Judith; Cecil, Jim; Bengtson, Roger D.; Breizman, Boris; Arefiev, Alexey; Chang-Diaz, Franklin; Squire, Jared; Glover, Tim; McCaskill, Greg; Cassibry, Jason; Li Zhongmin

    2006-01-20

    High power plasma propulsion can move large payloads for orbit transfer, lunar missions, and beyond with large savings in fuel consumption owing to the high specific impulse. At high power, lifetime of the thruster becomes an issue. Electrodeless devices with magnetically guided plasma offer the advantage of long life since magnetic fields confine the plasma radially and keep it from impacting the material surfaces. For decades, concerns have been raised about the plasma remaining attached to the magnetic field and returning to the vehicle along the closed magnetic field lines. Recent analysis suggests that this may not be an issue if the magnetic field is properly shaped in the nozzle region and the plasma has sufficient energy density to stretch the magnetic field downstream. An experiment is being performed to test the theory regarding the MHD detachment scenario. The status of that experiment will be discussed in this paper.

  18. 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

  19. Plasma flow and electron losses in the expander divertor of FRC

    NASA Astrophysics Data System (ADS)

    Yushmanov, P.; Barnes, D.; Dettrick, S.; Gupta, S.; Ryutov, D.; Krasheninnikov, S.; Necas, A.; Putvinski, S.

    2014-10-01

    Expander divertor is planned to be used in the design of next generation FRC device. The main goal of magnetic field expansion is to decrease heat load on the target plates and slow down heat losses through electron channel. A comprehensive study of expander divertor physics is initiated in Tri Alpha. It started with revision of pre-sheath electrostatic potential formation in the expander using both analytic and numerical means. An adaptation of 3D code KSOL has been developed to analyze electron physics and electrostatic potential formation. Initial results are presented. The key issue of the study is the analysis of the interaction of plasma with neutrals. Presence of neutrals affects expander physics in several ways. First of all, charge exchange and ionization modify pattern of ion flow in the expander magnetic field. That changes plasma density profile and affects formation of pre-sheath electrostatic potential. Second, ionization (as well as secondary electron emission) creates population of cold electrons in the expander which flow into confinement vessel and enhance out-flux of hot electrons. Distribution of neutrals is calculated in realistic geometry of expander divertor and effect on electron losses is evaluated.

  20. 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.

  1. Modeling of Plasma Detachment from a Magnetic Nozzle

    NASA Astrophysics Data System (ADS)

    Glesner, Colin; Srinivasan, Bhuvana

    2015-11-01

    The detachment of plasma from a magnetic nozzle is examined using numerical simulations based on the discontinuous galerkin method. Plasma detachment is of interest for its role in the development of plasma based space propulsion systems. The simulation parameters used, modeled after the computational and experimental work of Winglee et al. result in β ~ 0 . 01 , and Rem ~ 0 . 3 . In this low- β regime perturbation of the initially imposed magnetic field is expected to be small. To more effectively study these perturbations, the ideal magnetohydrodynamic equations are modified by linearizing the magnetic field. The perturbative component is then evolved in the simulation rather than the total magnetic field, allowing for a clearer resolution of changes in the magnetic field produced by the plasma. Because of the intermediate range of magnetic Reynolds number present in this plasma configuration, the effect of introducing resistivity in the simulation is also examined. Further work will investigate the effect of varying the configuration of the magnetic field. Supported by a grant from the Virginia Space Grant Consortium.

  2. Plasma Detachment Studies in the VASIMR Magnetic Nozzle

    NASA Astrophysics Data System (ADS)

    Tarditi, Alfonso G.; Shebalin, John

    2004-11-01

    Two important issues related to the VASIMR (Variable Specific Impulse Magnetoplasma Rocket, [1]) experiment are the plasma detachment and the collimation of the plume in the magnetic nozzle. These issues are being investigated both through theory/simulation studies and now also experimentally. A 3D, nonlinear MHD/2-fluid model of the magnetic nozzle has been implemented with the NIMROD code. The model has been run both with the actual VASIMR geometry and for an ideal De Laval nozzle configuration. The simulations indicate a distortion of the external field due to the plasma exhaust flow (carrying an azimuthal diamagnetic current) that may to lead to plasma detachment through the formation of magnetic islands. This is also being investigated experimentally. A Hall-effect, one-axis, gaussmeter has shown the local low-frequency magnetic field fluctuations during a plasma pulse. A 2D array of 3-axis "B-dot" probes is being developed for a fast mapping of the field perturbations in the nozzle (on the order of the Alfven time). Finally, a Rogowski coil probe is being designed to measure the azimuthal current profile in the exhaust plasma. [1] F. R. Chang-Diaz et al, Scientific American, p. 90, Nov. 2000

  3. Theory Issues for Induced Plasma Convection Experiments in the Divertor of the MAST Spherical Tokamak

    SciTech Connect

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

    2001-09-05

    This paper surveys theory issues associated with inducing convective cells through divertor tile biasing in a tokamak to broaden the scrape-off layer (SOL). The theory is applied to the Mega-Ampere Spherical Tokamak (MAST), where such experiments are planned in the near future. Criteria are presented for achieving strong broadening and for exciting shear-flow turbulence in the SOL; these criteria are shown to be attainable in practice. It is also shown that the magnetic shear present in the vicinity of the X-point is likely to confine the potential perturbations to the divertor region below the X-point, leaving the part of the SOL that is in direct contact with the core plasma intact. The current created by the biasing and the associated heating power are found to be modest.

  4. On the possibility of inducing strong plasma convection in the divertor of MAST

    NASA Astrophysics Data System (ADS)

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

    2001-10-01

    In this paper, a theory is developed to describe scrape-off layer (SOL) broadening by inducing convective cells through divertor plate biasing in a tokamak. The theory is applied to the Mega-Ampere Spherical Tokamak, where such experiments are planned in the near future. Criteria are derived for achieving strong broadening and for exciting shear-flow turbulence in the SOL, and these criteria are shown to be attainable in practice. It is also shown that the magnetic shear present in the vicinity of the X-point is likely to confine the potential perturbations to the divertor region below the X-point, leaving the part of the SOL that is in direct contact with the core plasma intact. The current created in the SOL by the biasing and the associated heating power are also calculated and are found to be modest.

  5. Edge plasma control by a local island divertor in the Compact Helical System

    SciTech Connect

    Komori, A.; Ohyabu, N.; Masuzaki, S.

    1997-12-31

    A local island divertor (LID) experiment was performed on the Compact Helical System (CHS) to demonstrate the principle of the LID. It was clearly demonstrated that the particle flow is controlled by adding a resonant perturbation field to the CHS magnetic configuration, and is guided to the back of an m/n = 1/1 island which is created by the perturbation field. The particles recycled there were pumped out with a pumping rate in the range from a few percent to about 10%. As a result, the line averaged core density was reduced by a factor of about 2 in comparison with non-LID discharges at the same gas puffing rate. In addition to the demonstration of these fundamental divertor functions, a modest improvement of energy confinement was observed, which could be attributed to the edge plasma control by the LID.

  6. Dynamic Responses of Tokamak Plasmas Induced by Externally Applied Rotating Helical Magnetic Field in Dynamic Ergodic Divertor

    NASA Astrophysics Data System (ADS)

    Kikuchi, Yusuke; Uesugi, Yoshihiko; Takamura, Shuichi; Budaev, Viatcheslav

    2002-11-01

    Dynamic Ergodic Divertor (DED) is an advanced concept for the control of the tokamak edge plasmas. In contrast to conventional Ergodic Divertor, an externally applied helical magnetic field rotates in the helical direction. This rotating helical field (RHF) may decrease the heat and particle flux onto the local target in the conventional divertor. In the DED experiment, it is also expected that RHF induces the edge plasma rotation due to the shielding current around the resonance surface. Penetration processes into tokamak plasmas and dynamic behavior of RHF have been investigated on HYBTOK-II tokamak. We have confirmed the attenuation of the radial component of RHF from the magnetic probe measurement. When the relative rotation velocity between the plasma and RHF is small around resonance surface, however, the radial component of RHF is amplified by the effect of re-distribution of the plasma current by island formation. We will analyze the penetration process of RHF by linear MHD theory.

  7. 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.

  8. Modeling of thermal effects on TIBER 2 (Thermal Ignition/Burn Experimental Reactor) divertor during plasma disruptions

    NASA Astrophysics Data System (ADS)

    Bruhn, M. L.; Perkins, L. J.

    1987-10-01

    Mapping the disruption power flow from the mid-plane of the TIBER Engineering Test Reactor to its divertor and calculating the resulting thermal effects are accomplished through the modification and coupling of three presently existing computer codes. The resulting computer code TADDPAK (Thermal Analysis Divertor during Disruption PAcKage) provides three-dimensional graphic presentations of time and positional dependent thermal effects on a poloidal cross section of the double-null-divertor configured reactor. These thermal effects include incident heat flux, surface temperature, vaporization rate, total vaporization, and melting depth. The dependence of these thermal effects on material choice, disruption pulse shape, and the characteristic thickness of the plasma scrape-off layer is determined through parametric analysis with TADDPAK. This computer code is designed to be a convenient, rapid, and user-friendly modeling tool which can be easily adapted to most tokamak double-null-divertor reactor designs.

  9. A divertor plasma configuration design method for tokamaks

    NASA Astrophysics Data System (ADS)

    Guo, Yong; Xiao, Bing-Jia; Liu, Lei; Yang, Fei; Wang, Yuehang; Qiu, Qinglai

    2016-11-01

    The efficient and safe operation of large fusion devices strongly relies on the plasma configuration inside the vacuum chamber. It is important to construct the proper plasma equilibrium with a desired plasma configuration. In order to construct the target configuration, a shape constraint module has been developed in the tokamak simulation code (TSC), which controls the poloidal flux and the magnetic field at several defined control points. It is used to construct the double null, lower single null, and quasi-snowflake configurations for the required target shape and calculate the required PF coils current. The flexibility and practicability of this method have been verified by the simulated results. Project supported by the National Magnetic Confinement Fusion Research Program of China (Grant Nos. 2014GB103000 and 2014GB110003), the National Natural Science Foundation of China (Grant Nos. 11305216, 11305209, and 11375191), and External Cooperation Program of BIC, Chinese Academy of Sciences (Grant No. GJHZ201303).

  10. Laser Induced Plasma Spectroscopy to Diagnose Impurities on a Tokamak Divertor

    NASA Astrophysics Data System (ADS)

    Kim, Minju; Cho, Min Sang; Cho, Byoung-Ick

    2015-11-01

    In order to monitor dust and impurity deposition on the plasma facing components (PFCs) of a fusion device, the Laser Induced Plasma Spectroscopy (LIPS) is considered. It is a powerful spectroscopic technique to measure emission lines from the excited atoms by means of the high power laser pulse, and could be applied to diagnose dust and impurities deposition on the PFCs. We have measured LIPS spectra for the inner-divertor tile from 2011 KSTAR campaign. Characteristic emission lines for several key elements, such as iron, chrome are identified. Using those lines, plasma conditions for various laser parameters and their temporal evolution are characterized. It will be also presented that the depth profiling for the deposited elements on a surface of graphite tile. This work is supported by the NRF (No. 2013M1A7A1A02043864), National Research Foundation of Korea (No. 2013R1A1A1007084) and the TBP research project of GIST. Laser Induced Plasma Spectroscopy to Diagnose Impurities on a Tokamak Divertor.

  11. Magnetic-divertor stabilization of an axisymmetric plasma with anisotropic temperature

    SciTech Connect

    Sasagawa, Y.; Katanuma, I.; Mizoguchi, Y.; Cho, T.; Pastukhov, V. P.

    2006-12-15

    Magnetohydrodynamic stabilization of an axisymmetric mirror plasma with a magnetic divertor is studied. An equation is found for the flute modes, which includes the stabilizing influence of ion temperature anisotropy and nonparaxial magnetic fields, as well as a finite ion Larmor radius. It is shown that if the density profile is sufficiently gentle, then the nonparaxial configuration can stabilize all modes as long as ion temperature is radially uniform. This can be demonstrated even when the density vanishes on the separatrix and even for small ion Larmor radii. It is found, however, that the ion temperature gradient makes the unstable region wider; high ion temperature is required to stabilize the flute mode.

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

    NASA Astrophysics Data System (ADS)

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

    2011-10-01

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

  13. 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.

  14. Vertical profiles and two-dimensional distributions of carbon line emissions from C2+-C5+ ions in attached and RMP-assisted detached plasmas of large helical device

    NASA Astrophysics Data System (ADS)

    Zhang, Hongming; Morita, Shigeru; Dai, Shuyu; Oishi, Tetsutarou; Goto, Motoshi; Huang, Xianli; Kawamura, Gakushi; Kobayashi, Masahiro; Liu, Yang; Murakami, Izumi; Narushima, Yoshiro

    2017-02-01

    In Large Helical Device (LHD), the detached plasma is obtained without external impurity gas feed by supplying an m/n = 1/1 resonant magnetic perturbation (RMP) field to a plasma with an outwardly shifted plasma axis position of Rax = 3.90 m where the magnetic resonance exists in the stochastic magnetic field layer outside the last closed flux surface. The plasma detachment is triggered by the appearance of an m/n = 1/1 island when the density, increased using hydrogen gas feed, exceeds a threshold density. The behavior of intrinsically existing impurities, in particular, carbon originating in the graphite divertor plates, is one of the important key issues to clarify the characteristic features of the RMP-assisted plasma detachment although the particle flux still remains on some divertor plates even in the detachment phase of the discharge. For this purpose, vertical profiles and two-dimensional (2-D) distributions of edge carbon emissions of CIII to CVI have been measured at extreme ultraviolet wavelength range, and the results are compared between attached and RMP-assisted detached plasmas. It is found that the CIII and CIV emissions located in the stochastic magnetic field layer are drastically increased near the m/n = 1/1 island O-point and in the vicinity of both inboard and outboard edge separatrix X-points during the RMP-assisted detachment, while those emissions are only enhanced in the vicinity of the outboard edge X-point in attached plasmas without RMP. The result clearly indicates a change in the magnetic field lines connecting to the divertor plates, which is caused by the growth of the m/n = 1/1 edge magnetic island. In contrast, the intensity of CVI emitted radially inside the magnetic island significantly decreases during the detachment, suggesting an enhancement of the edge impurity screening. The measured carbon distribution is analyzed with a three-dimensional edge plasma transport simulation code, EMC3-EIRENE, for the attached plasmas without

  15. 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.

  16. QED-1 device and measurements of gettering efficiency for a simulated divertor plasma

    SciTech Connect

    Owens, D.K.; Yamada, M.

    1980-03-01

    The QED-1 device at PPL has provided gettering efficiency data for neutralized hydrogen plasma on titanium. The hollow-anode arcjet produces a plasma column 1 cm in diameter with 10/sup 12/ < n/sub e/ < 10/sup 15/ cm/sup -3/ and T/sub i/ approx.< T/sub e/ = 3-10 eV, confined by an axial magnetic field of 1-6 kG. The gettering measurements are based on monitoring neutral gas density with respect to time in the divertor simulation chamber of QED-1. The present results indicate that the plasma particles lose their charge and most of their energy when they strike the neutralizer plate.

  17. 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

  18. Development of improved analysis of non-local electron parallel heat transport in divertor plasmas

    NASA Astrophysics Data System (ADS)

    Allais, Fabrice; Alouani Bibi, Fathallah; Kim, Chang-Geun; Matte, Jean-Pierre; Stotler, Daren P.

    2004-03-01

    Parallel electron heat transport in divertor plasmas is investigated. Our electron kinetic code "FPI" has been upgraded to take into account the hydrogen's atomic physics, including 30 energy levels in the computation. This required important improvements in the numerical algorithms in order to run the code within a reasonable time and compute the effects of each inelastic process. Their effects on non-local transport and the large enhancement of the effective (i.e. including ionization via excited states) ionization rates in the cold plasma due to nonlocal transport will be presented. A non-local electron heat flow formula [1] has been adapted and implemented in the "UEDGE" code. Simulations using it were compared to runs made with the more traditional flux limited heat diffusion formula. Considerable differences were seen in the temperature profiles. [1] F. Alouani Bibi and J.P. Matte, Phys. Rev. E 66, 066414 (2002)

  19. 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

  20. 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.

  1. 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

  2. Plasma flow and carbon production and circulation with the ergodic divertor of Tore Supra

    NASA Astrophysics Data System (ADS)

    Corre, Y.; Gunn, J.; Pégourié, B.; Guirlet, R.; DeMichelis, C.; Giannella, R.; Ghendrih, P.; Hogan, J.; Monier-Garbet, P.; Azéroual, A.; Escarguel, A.; Gauthier, E.

    2007-02-01

    This paper presents a detailed study of carbon production and transport from the ergodic divertor (ED) target plates to the plasma core in the Tore Supra tokamak. Adapted experimental and numerical modelling techniques have been used to describe each of the main phenomena in play. Edge electron density and temperature are measured with Langmuir probes. The C II, C III and Hα emission is measured with optical fibres and cameras. The background plasma flow is calculated consistently with the observed recycling pattern by the neutral transport code EDCOLL for the two magnetic connection schemes of interest (short or long connection lengths). 3D Monte-Carlo modelling of carbon near the neutralizer plate (BBQ code) shows that the transport of carbon ions is governed by the friction force in addition to the electric field. Finally, a simplified 3D test particle model is used to estimate the core penetration fraction of carbon. A high value is found for the carbon screening efficiency (fraction of particles that does not penetrate in the plasma core), in the range 95-97% depending on the edge plasma conditions. This value, combined with the calculated carbon influxes, yields the first quantitative estimate of the carbon core contamination during ED operation. The paper shows that the screening of carbon and core contamination are mainly dependent on the carbon source (partially controlled with the ED) and the plasma flow distribution in the laminar region (magnetic topology and particle drifts).

  3. Natural Divertor Spherical Tokamak Plasmas with bean shape and ergodic limiter

    NASA Astrophysics Data System (ADS)

    Ribeiro, Celso; Herrera, Julio; Chavez, Esteban; Tritz, Kevin

    2013-10-01

    The former spherical tokamak (ST) MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R < 0.14 m, a < 0.10 m, BT < 0.5T, Ip < 40 kA, 3 ms pulse) is being recommissioned in Costa Rica Institute of Technology. The main objectives of the MEDUSA-CR project are training and to clarify several issues in relevant physics for conventional and mainly STs, including beta studies in bean-shaped ST plasmas, transport, heating and current drive via Alfvén wave, and natural divertor STs with ergodic magnetic limiter. We report here improvements in the self-consistency of these equilibrium comparisons and a preliminary study of their MHD stability beta limits. VIE-ITCR, IAEA-CRP contract 17592, National Instruments of Costa Rica.

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

    NASA Astrophysics Data System (ADS)

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

    2001-03-01

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

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

    NASA Technical Reports Server (NTRS)

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

    2004-01-01

    Plasma volume recombination in the divertor, a process in which charged particles recombine to neutral atoms, contributes to plasma detachment and hence cooling at the divertor target region. Detachment has been observed at JET and other tokamaks and is known to occur at low electron temperatures (T(sub e)<1 eV) and at high electron density (n(sub e)>10(exp 20)/m(exp 3)). The ability to measure such low temperatures is therefore of interest for modelling the divertor. In present work we report development of a new spectroscopic technique for investigation of local electron density (n(sub e)) and temperature (T,) in the outer divertor at JET.

  6. 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.

  7. 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.

  8. 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.

  9. Be W alloy formation in static and divertor-plasma simulator experiments

    NASA Astrophysics Data System (ADS)

    Baldwin, M. J.; Doerner, R. P.; Nishijima, D.; Buchenauer, D.; Clift, W. M.; Causey, R. A.; Schmid, K.

    2007-06-01

    Collaborative Be-W interaction experiments conducted at UC-San Diego PISCES Laboratory, and Sandia National Laboratories, Livermore, CA (SNL/CA), are reported. In the divertor-plasma simulator PISCES-B, W targets are exposed to Be seeded D2 plasma in the temperature range 1070-1320 K. All reveal the formation of surface Be-W alloying. The alloy reaction rate is found to increase with surface temperature in the range 1023-1123 K in SNL vacuum-deposition phase formation experiments. In both sets of experiments the efficiency of surface alloying is found to depend on the availability of surface deposited Be. This availability is reduced by evaporation at high temperature, and also by plasma re-erosion in the case of PISCES-B targets. Surface analysis of targets using Auger electron spectroscopy (AES), wavelength dispersive X-ray spectroscopy (WDS), and X-ray diffraction (XRD) reveals Be12W as the dominant alloy composition where Be surface availability is optimal.

  10. SOLPS5.1 analysis of detachment with drifts and gas pumping effects in EAST

    NASA Astrophysics Data System (ADS)

    Du, Hailong; Sang, Chaofeng; Wang, Liang; Bonnin, Xavier; Guo, Houyang; Sun, Jizhong; Wang, Dezhen

    2016-08-01

    The aim of this paper is to estimate the effects of usual drifts and gas puffing/pumping locations on divertor detachment and Ar ion transport in the Experimental Advanced Superconducting Tokamak (EAST) by using the edge plasma code package SOLPS5.1. The simulated results reveal that which target plate first detaches depends strongly on the usual drifts, but not on the location of impurity gas puffing, which could be one of the possible explanations for the experimentally observed phenomenon (Chen et al 2013 Phys. Plasmas 20 022311) that the lower inner target first detached compared to the lower outer target with the lower outer gas puffing. The physics behind this phenomenon is that drifts not only can induce background ion flux, plasma density and temperature redistribution in the scrape-off layer (SOL) and divertor region, but also can change the Ar impurity force balance leading to Ar ions being dragged from bottom to top. Furthermore, the simulated results illustrate that the Ar ion transport in the SOL and divertor region is similar for different gas puffing locations including upstream and divertor region before partial detachment. However, the Ar ions penetrate into the core more easily, giving rise to more discharge disruption during complete detachment with upstream gas puffing than with divertor region puffing. Finally, we also estimate the effect of gas pumping on the detachment in order to realize long-pulse partial detachment in EAST. The results indicate that long-pulse partial detachment could be obtained by improving the pumping speed to match the puffing speed in case the excess Ar atoms accumulate in the core plasma during partial detachment in EAST.

  11. Overview of the Results on Divertor Heat Loads in RMP Controlled H-mode Plasmas on DIII-D

    SciTech Connect

    Jakubowski, M. W.; Evans, T. E.; Fenstermacher, M. E.; Groth, M.; Lasnier, C. J.; Leonard, A. W.; Schmitz, O.; Watkins, J. G.; Elch, T.; Wolf, R. C.; Baylor, L. B.; Boedo, J.A.; Burrell, K. H.; Frerichs, H.; DeGrassie, J. S.; Gohil, P.; Joseph, I.; Mordijck, S.; Lehnen, M.; Petty, C C.; Pinsker, R. I.; Reiter, D.; Rhodes, T. L.; Samm, U.; Schaffer, M. J.; Snyder, P. B.; Stoschus, H.; Unterberg, E. A.; West, W. P.

    2009-01-01

    n this paper the manipulation of power deposition on divertor targets at DIII-D by the application of resonant magnetic perturbations (RMPs) for suppression of large type-I edge localized modes (ELMs) is analysed. We discuss the modification of the ELM characteristics by the RMP applied. It is shown that the width of the deposition pattern in ELMy H-mode depends linearly on the ELM deposited energy, whereas in the RMP phase of the discharge those patterns are controlled by the externally induced magnetic perturbation. It was also found that the manipulation of heat transport due to the application of small, edge RMP depends on the plasma pedestal electron collisionality nu(e)*. We compare in this analysis RMP and no RMP phases with and without complete ELM suppression. At high nu(e)* > 0.5, the heat flux during the ELM suppressed phase is of the same order as the inter-ELM and the no-RMP phase. However, below this collisionality value, a slight increase in the total power flux to the divertor is observed during the RMP phase. This is most likely caused by a more negative potential at the divertor surface due to hot electrons reaching the divertor surface from the pedestal area along perturbed, open field lines.

  12. The symmetric quartic map for trajectories of magnetic field lines in elongated divertor tokamak plasmas

    SciTech Connect

    Jones, Morgin; Wadi, Hasina; Ali, Halima; Punjabi, Alkesh

    2009-04-15

    The coordinates of the area-preserving map equations for integration of magnetic field line trajectories in divertor tokamaks can be any coordinates for which a transformation to ({psi}{sub t},{theta},{phi}) coordinates exists [A. Punjabi, H. Ali, T. Evans, and A. Boozer, Phys. Lett. A 364, 140 (2007)]. {psi}{sub t} is toroidal magnetic flux, {theta} is poloidal angle, and {phi} is toroidal angle. This freedom is exploited to construct the symmetric quartic map such that the only parameter that determines magnetic geometry is the elongation of the separatrix surface. The poloidal flux inside the separatrix, the safety factor as a function of normalized minor radius, and the magnetic perturbation from the symplectic discretization are all held constant, and only the elongation is {kappa} varied. The width of stochastic layer, the area, and the fractal dimension of the magnetic footprint and the average radial diffusion coefficient of magnetic field lines from the stochastic layer; and how these quantities scale with {kappa} is calculated. The symmetric quartic map gives the correct scalings which are consistent with the scalings of coordinates with {kappa}. The effects of m=1, n={+-}1 internal perturbation with the amplitude that is expected to occur in tokamaks are calculated by adding a term [H. Ali, A. Punjabi, A. H. Boozer, and T. Evans, Phys. Plasmas 11, 1908 (2004)] to the symmetric quartic map. In this case, the width of stochastic layer scales as 0.35 power of {kappa}. The area of the footprint is roughly constant. The average radial diffusion coefficient of field lines near the X-point scales linearly with {kappa}. The low mn perturbation changes the quasisymmetric structure of the footprint, and reorganizes it into a single, large scale, asymmetric structure. The symmetric quartic map is combined with the dipole map [A. Punjabi, H. Ali, and A. H. Boozer, Phys. Plasmas 10, 3992 (2003)] to calculate the effects of magnetic perturbation from a current

  13. Development of Improved Analysis of 1-D Non-Local Electron Parallel Heat Transport in Divertor Plasmas

    NASA Astrophysics Data System (ADS)

    Kim, Chang-Geun; Alouani Bibi, Fathallah; Matte, Jean-Pierre; Rognlien, Thomas D.; Stotler, Daren P.

    2003-10-01

    Improved analysis of 1-D non-local electron parallel heat transport in divertor plasmas is investigated. In the divertor of NSTX, as well as in other magnetic confinement devices, there are steep temperature gradients parallel to the direction of the magnetic field, especially near the neutralizer plates. Strong temperature gradients modify the electrons' thermal transport. A recently developed delocalization formula [1] is implemented in the "UEDGE" fluid edge plasma simulation code. UEDGE simulations are performed using this non-local electron heat flow formula (without any external heating source), and are compared to runs made with the more traditional flux limited heat diffusion formula. . [1] F. Alouani Bibi and J.P. Matte, Phys. Rev. E 66, 066414 (2002)

  14. 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.

  15. Be I and Be II spectroscopy in divertor plasma relevant conditions

    NASA Astrophysics Data System (ADS)

    Nishijima, D.; Doerner, R. P.; Seraydarian, R. P.

    2013-07-01

    Intensity ratios of various Be I and Be II lines measured in Be-seeded D and He plasmas in the PISCES-B linear divertor plasma simulator are compared with the corresponding ratios of the photon emissivity coefficient, PEC, calculated by ADAS. Agreement of measured intensity ratios with calculated PEC ratios is satisfactory within a factor of ˜2 for both Be I and Be II. It is proposed that a Be I line ratio of 234.8 nm/265.0 nm and a Be II line ratio of 467.3 nm/313.1 nm can be used to estimate the electron temperature, while a 265.0 nm/332.1 nm Be I line ratio is sensitive to the electron density. Further, S/XB values of a Be I line at 457.3 nm were experimentally determined from a ratio of the sputtered Be flux to the emission intensity. Measured values are systematically lower than calculated ADAS values, which may be explained by the increased sputtering yield of redeposited Be atoms.

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

    NASA Astrophysics Data System (ADS)

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

    2013-10-01

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

  17. 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.

  18. Full toroidal imaging of non-axisymmetric plasma material interaction in the National Spherical Torus Experiment divertor

    SciTech Connect

    Scotti, Filippo; Roquemore, A. L.; Soukhanovskii, V. A.

    2012-10-01

    A pair of two dimensional fast cameras with a wide angle view (allowing a full radial and toroidal coverage of the lower divertor) was installed in the National Spherical Torus Experiment in order to monitor non-axisymmetric effects. A custom polar remapping procedure and an absolute photometric calibration enabled the easier visualization and quantitative analysis of non-axisymmetric plasma material interaction (e.g., strike point splitting due to application of 3D fields and effects of toroidally asymmetric plasma facing components).

  19. Conceptual design of a divertor Thomson scattering diagnostic for NSTX-U

    SciTech Connect

    McLean, A. G. Soukhanovskii, V. A.; Allen, S. L.; Carlstrom, T. N.; LeBlanc, B. P.; Ono, M.; Stratton, B. C.

    2014-11-15

    A conceptual design for a divertor Thomson scattering (DTS) diagnostic has been developed for the NSTX-U device to operate in parallel with the existing multipoint Thomson scattering system. Higher projected peak heat flux in NSTX-U will necessitate application of advanced magnetics geometries and divertor detachment. Interpretation and modeling of these divertor scenarios will depend heavily on local measurement of electron temperature, T{sub e}, and density, n{sub e}, which DTS provides in a passive manner. The DTS design for NSTX-U adopts major elements from the successful DIII-D DTS system including 7-channel polychromators measuring T{sub e} to 0.5 eV. If implemented on NSTX-U, the divertor TS system would provide an invaluable diagnostic for the boundary program to characterize the edge plasma.

  20. PREFACE: Light element atom, molecule and radical behaviour in the divertor and edge plasma regions

    NASA Astrophysics Data System (ADS)

    Braams, Bastiaan J.; Chung, Hyun-Kung

    2015-01-01

    This volume of Journal of Physics: Conference Series contains contributions by participants in an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on "Light element atom, molecule and radical behaviour in the divertor and edge plasma regions" (in magnetic fusion devices). Light elements are the dominant impurity species in fusion experiments and in the near-wall plasma they occur as atoms or ions and also as hydrides and other molecules and molecular ions. Hydrogen (H or D, and T in a reactor) is the dominant species in fusion experiments, but all light elements He - O and Ne are of interest for various reasons. Helium is a product of the D+T fusion reaction and is introduced in experiments for transport studies. Lithium is used for wall coating and also as a beam diagnostic material. Beryllium is foreseen as a wall material for the ITER experiment and is used on the Joint European Torus (JET) experiment. Boron may be used as a coating material for the vessel walls. Carbon (graphite or carbon-fiber composite) is often used as the target material for wall regions subject to high heat load. Nitrogen may be used as a buffer gas for edge plasma cooling. Oxygen is a common impurity in experiments due to residual water vapor. Finally, neon is another choice as a buffer gas. Data for collisional and radiative processes involving these species are important for plasma modelling and for diagnostics. The participants in the CRP met 3 times over the years 2009-2013 for a research coordination meeting. Reports and presentation materials for these meetings are available through the web page on coordinated research projects of the (IAEA) Atomic and Molecular Data Unit [1]. Some of the numerical data generated in the course of the CRP is available through the ALADDIN database [2]. The IAEA takes the opportunity to thank the participants in the CRP for their dedicated efforts in the course of the CRP and for their contributions to this volume. The IAEA

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

    NASA Astrophysics Data System (ADS)

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

    2015-10-01

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

  2. SOL-Divertor Plasma Simulation in the KSTAR Tokamak with a Neutral Transport Model Using the TEP method

    NASA Astrophysics Data System (ADS)

    Han, Hyun-Sun; Kim, Ki Min; Hong, Sang Hee

    2007-11-01

    A two-dimensional numerical simulation has been carried out to analyze the transport phenomena of plasma and neutrals in Scrape- off Layer (SOL) and divertor region of the Korea Superconducting Tokamak Advanced Research (KSTAR) tokamak. In this numerical work, the transmission and escape probability (TEP) method is used for a neutral transport model by adapting the GTNEUT [1] code, which is coupled with a plasma transport model based on the Braginskii's fluid formulation. Prior to combining the GTNEUT code with the plasma transport one, preliminary tests are conducted by comparison with a Monte Carlo method to check the numerical accuracy and efficiency of the neutral model. A performance improvement of computing time is achieved during the coupling processes by pre-computing the various transmission coefficients and setting up an interpolation lookup table. As results of the simulation, plasma density and temperature distributions in the SOL-divertor region are calculated for the baseline operation of the KSTAR tokamak. This integrated modeling method could be extended to the simulation of more complicated edge transport for the advanced tokamak operation including impurity transport. [1] J. Mandrekas, Comput. Phys. Comm., 161, 36 (2004)

  3. 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.

  4. 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.

  5. Electron pressure balance in the SOL through the transition to detachment

    DOE PAGES

    McLean, A. G.; Leonard, A. W.; Makowski, M. A.; ...

    2015-02-07

    Upgrades to core and divertor Thomson scattering (DTS) diagnostics at DIII-D have provided measurements of electron pressure profiles in the lower divertor from attached- to fully-detached divertor plasma conditions. Detailed, multistep sequences of discharges with increasing line-averaged density were run at several levels of Pinj. Strike point sweeping allowed 2D divertor characterization using DTS optimized to measure Te down to 0.5 eV. The ionization front at the onset of detachment is found to move upwards in a controlled manner consistent with the indication that scrape-off layer parallel power flux is converted from conducted to convective heat transport. Measurements of ne,more » Te and pe in the divertor versus Lparallel demonstrate a rapid transition from Te ≥ 15 eV to ≤3 eV occurring both at the outer strike point and upstream of the X-point. Furthermore, these observations provide a strong benchmark for ongoing modeling of divertor detachment for existing and future tokamak devices.« less

  6. Electron pressure balance in the SOL through the transition to detachment

    SciTech Connect

    McLean, A. G.; Leonard, A. W.; Makowski, M. A.; Groth, M.; Allen, S. L.; Boedo, J. A.; Bray, B. D.; Briesemeister, A. R.; Carlstrom, T. N.; Eldon, D.; Fenstermacher, M. E.; Hill, D. N.; Lasnier, C. J.; Liu, C.; Osborne, T. H.; Petrie, T. W.; Soukhanovskii, V. A.; Stangeby, P. C.; Tsui, C.; Unterberg, E. A.; Watkins, J. G.

    2015-02-07

    Upgrades to core and divertor Thomson scattering (DTS) diagnostics at DIII-D have provided measurements of electron pressure profiles in the lower divertor from attached- to fully-detached divertor plasma conditions. Detailed, multistep sequences of discharges with increasing line-averaged density were run at several levels of Pinj. Strike point sweeping allowed 2D divertor characterization using DTS optimized to measure Te down to 0.5 eV. The ionization front at the onset of detachment is found to move upwards in a controlled manner consistent with the indication that scrape-off layer parallel power flux is converted from conducted to convective heat transport. Measurements of ne, Te and pe in the divertor versus Lparallel demonstrate a rapid transition from Te ≥ 15 eV to ≤3 eV occurring both at the outer strike point and upstream of the X-point. Furthermore, these observations provide a strong benchmark for ongoing modeling of divertor detachment for existing and future tokamak devices.

  7. Electron pressure balance in the SOL through the transition to detachment

    NASA Astrophysics Data System (ADS)

    McLean, A. G.; Leonard, A. W.; Makowski, M. A.; Groth, M.; Allen, S. L.; Boedo, J. A.; Bray, B. D.; Briesemeister, A. R.; Carlstrom, T. N.; Eldon, D.; Fenstermacher, M. E.; Hill, D. N.; Lasnier, C. J.; Liu, C.; Osborne, T. H.; Petrie, T. W.; Soukhanovskii, V. A.; Stangeby, P. C.; Tsui, C.; Unterberg, E. A.; Watkins, J. G.

    2015-08-01

    Upgrades to core and divertor Thomson scattering (DTS) diagnostics at DIII-D have provided measurements of electron pressure profiles in the lower divertor from attached- to fully-detached divertor plasma conditions. Detailed, multistep sequences of discharges with increasing line-averaged density were run at several levels of Pinj. Strike point sweeping allowed 2D divertor characterization using DTS optimized to measure Te down to 0.5 eV. The ionization front at the onset of detachment is found to move upwards in a controlled manner consistent with the indication that scrape-off layer parallel power flux is converted from conducted to convective heat transport. Measurements of ne, Te and pe in the divertor versus Lparallel demonstrate a rapid transition from Te ⩾ 15 eV to ⩽3 eV occurring both at the outer strike point and upstream of the X-point. These observations provide a strong benchmark for ongoing modeling of divertor detachment for existing and future tokamak devices.

  8. Effect of Magnetic Field Gradient on Plasma Detachment Induced by Breaking of Adiabatic Plasma Expansion

    NASA Astrophysics Data System (ADS)

    Chung, K. S.; Kim, June Young; Chung, Kyoung-Jae; Hwang, Y. S.

    2016-10-01

    A magnetic field gradient that is a variation in the magnetic field around the ion flow has been investigated as a primary parameter for ion detachment in the magnetic nozzle geometries. Some scale lengths of magnetic field are controlled by two solenoid coils outside the diffusion chamber of a ECR-driven linear plasma device. The axial and radial profiles of the plasma potential and electron temperature are measured by a Langmuir probe array for the various magnetic field configurations in the downstream. The local adiabaticity, strong constant magnetic moment, is satisfied with a linear relationship between the change in effective electron temperature and the change in plasma potential in the low magnetic field gradient. Whereas, with an increasing non-homogeneity of the magnetic field in the direction of the flow, the breaking of adiabatic plasma expansion is identified to measure the nonlinear process which is the variation for an adiabatic exponent. Such the loss of adiabaticity is also explained in terms of non-adiabaticity parameter i.e. degree of demagnetization. This research was supported by National R&D Program through the National Research Foundation of Korea (NRF) funded by the Ministry of Science, ICT & Future Planning (Nos. 2014M1A7A1A02030165 and 2014M1A7A1A03045367).

  9. Magnetic Nozzles for Plasma Thrusters: Acceleration, Thrust, and Detachment Mechanisms

    DTIC Science & Technology

    2011-10-01

    In the unmagnetized case the plasma is accelerated diffusively and some ion streamlines go backwards to the left dielectric wall , where ions are...processes related to the plasma wall interaction, virtual cathode considerations and anomalous diffusion. In this work, anomalous diffusion and virtual...demagnetized, which allows the development of the electric force and ion acceleration there, and increases the plasma flux to the wall . For β0 > 3 − 4

  10. Optimizing stability, transport, and divertor operation through plasma shaping for steady-state scenario development in DIII-D

    SciTech Connect

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

    2008-12-19

    Recent studies on the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] have elucidated key aspects of the dependence of stability, confinement, and density control on the plasma magnetic configuration, leading to the demonstration of nearly noninductive operation for >1 s with pressure 30% above the ideal no-wall stability limit. Achieving fully noninductive tokamak operation requires high pressure, good confinement, and density control through divertor pumping. Plasma geometry affects all of these. Ideal magnetohydrodynamics modeling of external kink stability suggests that it may be optimized by adjusting the shape parameter known as squareness ({zeta}). Optimizing kink stability leads to an increase in the maximum stable pressure. Experiments confirm that stability varies strongly with {zeta}, in agreement with the modeling. Optimization of kink stability via {zeta} is concurrent with an increase in the H-mode edge pressure pedestal stability. Global energy confinement is optimized at the lowest {zeta} tested, with increased pedestal pressure and lower core transport. Adjusting the magnetic divertor balance about a double-null configuration optimizes density control for improved noninductive auxiliary current drive. The best density control is obtained with a slight imbalance toward the divertor opposite the ion grad(B) drift direction, consistent with modeling of these effects. These optimizations have been combined to achieve noninductive current fractions near unity for over 1 s with normalized pressure of 3.5<{beta}{sub N}<3.9, bootstrap current fraction of >65%, and a normalized confinement factor of H{sub 98(y,2)}{approx}1.5.

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

    NASA Astrophysics Data System (ADS)

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

    2009-05-01

    Recent studies on the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] have elucidated key aspects of the dependence of stability, confinement, and density control on the plasma magnetic configuration, leading to the demonstration of nearly noninductive operation for >1 s with pressure 30% above the ideal no-wall stability limit. Achieving fully noninductive tokamak operation requires high pressure, good confinement, and density control through divertor pumping. Plasma geometry affects all of these. Ideal magnetohydrodynamics modeling of external kink stability suggests that it may be optimized by adjusting the shape parameter known as squareness (ζ). Optimizing kink stability leads to an increase in the maximum stable pressure. Experiments confirm that stability varies strongly with ζ, in agreement with the modeling. Optimization of kink stability via ζ is concurrent with an increase in the H-mode edge pressure pedestal stability. Global energy confinement is optimized at the lowest ζ tested, with increased pedestal pressure and lower core transport. Adjusting the magnetic divertor balance about a double-null configuration optimizes density control for improved noninductive auxiliary current drive. The best density control is obtained with a slight imbalance toward the divertor opposite the ion grad(B) drift direction, consistent with modeling of these effects. These optimizations have been combined to achieve noninductive current fractions near unity for over 1 s with normalized pressure of 3.5<βN<3.9, bootstrap current fraction of >65%, and a normalized confinement factor of H98(y ,2)≈1.5.

  12. 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.

  13. 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.

  14. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2016-11-01

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

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

    SciTech Connect

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

    2016-11-15

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

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

    DOE PAGES

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

    2016-08-04

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

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

    PubMed

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

    2016-11-01

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

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

    SciTech Connect

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

    2016-08-04

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

  20. Dependence of divertor heat flux widths on heating power, flux expansion, and plasma current in the NSTX

    SciTech Connect

    Maingi, Rajesh; Soukhanovskii, V. A.; Ahn, J.W.

    2011-01-01

    We report the dependence of the lower divertor surface heat flux profiles, measured from infrared thermography and mapped magnetically to the mid-plane on loss power into the scrape-off layer (P{sub LOSS}), plasma current (I{sub p}), and magnetic flux expansion (f{sub exp}), as well as initial results with lithium wall conditioning in NSTX. Here we extend previous studies [R. Maingi et al., J. Nucl. Mater. 363-365 (2007) 196-200] to higher triangularity similar to 0.7 and higher I{sub p} {le} 1.2 MA. First we note that the mid-plane heat flux width mapped to the mid-plane, {lambda}{sub q}{sup mid} is largely independent of P{sub LOSS} for P{sub LOSS} {ge} 4 MW. {lambda}{sub q}{sup mid} is also found to be relatively independent of f{sub exp}; peak heat flux is strongly reduced as f{sub exp} is increased, as expected. Finally, {lambda}{sub q}{sup mid} is shown to strongly contract with increasing I{sub p} such that {lambda}{sub q}{sup mid} {alpha} I{sub p}{sup -1.6} with a peak divertor heat flux of q{sub div,peak} similar to 15 MW/m{sup 2} when I{sub p} = 1.2 MA and P{sub LOSS} similar to 6 MW. These relationships are then used to predict the divertor heat flux for the planned NSTX-Upgrade, with heating power between 10 and 15 MW, B{sub t} = 1.01 and I{sub p}= 2.0 MA for 5 s.

  1. 3D plasma fluid simulations in divertor tokamaks. Final technical report, 1993--1995

    SciTech Connect

    Strauss, H.R.

    1995-08-01

    The main accomplishment of this grant was the development of a finite element time dependent magnetofluid code, FEMHD. The code is nonlinear and three dimensional. In the poloidal plane, the elemental cells of the mesh are triangles, which offer both simplicity and adaptability. In the third, toroidal, direction, there is an option of a standard staggered finite difference mesh, or Fourier transforms. The FEMHD code runs on several platforms, including Crays, UNIX workstations, and a parallel version runs on an IBM SP1. Several problems have been considered with the unstructured mesh FEMHD code. They are (1) MHD simulations in divertor tokamaks; (2) simulations of ELM-like ballooning modes in divertor tokamaks; and (3) reconnection and singular MHD equilibria.

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

    NASA Astrophysics Data System (ADS)

    Boedo, J.; Rudakov, D.

    2016-10-01

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

  3. Formation of Non-Monotonic Potential Structure in the Detached Plasma

    NASA Astrophysics Data System (ADS)

    Ishiguro, Seiji; Pianpanit, Theerasarn; Hasegawa, Hiroki

    2016-10-01

    Plasma detachment has been investigated by means of PIC simulation which includes plasma-neutral collision and Coulomb collision. In our previous study, we have shown that a strong gradient in temperature appears in front of the target plate in the case that high density and low temperature neutral gas is introduced. It is observed that a potential hill is created in the neutral gas region where ions lose energy due to the elastic and charge exchange collision and, as a result, the ion density increases. This potential structure traps the low energy electrons and may play a role in the development of plasma detachment state. This work is supported by NIFS Collaboration Research Programs NIFS14KNXN279 and NIFS14KNSS059 and the NIFS/NINS project of Formation of International Scientific Base and Network.

  4. Modeling of ultra-high recycling divertors with the PLANET code

    SciTech Connect

    Petravic, M.

    1993-07-01

    The handling of power carried by the charged particles into the scrape-off layer of a tokamak reactor remains a major obstacle for its continuous and reliable operation. Ways of reducing this power through radiation have been studied numerically using fluid models for both the plasma and neutral gas. A new model for the combined plasma and neutral gas 2-D transport capable of simultaneously representing regions of fully-ionized plasma, partially ionized plasma, and pure neutral gas has been assembled and implemented in the PLANET code. Divertor plasma temperatures of just below 1 eV have been achieved in a pure hydrogen plasma, resulting in an ionization-free region together with ionization and recombination fronts detached from the material walls. In this regime energy reaches the walls almost exclusively in the form of radiation which, in principle, solves the divertor heat load problems.

  5. Determination of the plasma potential and the EEDF by Langmuir probes in the divertor region of COMPASS tokamak

    NASA Astrophysics Data System (ADS)

    Ivanova, P.; Dimitrova, M.; Vasileva, E.; Popov, Tsv K.; Dejarnac, R.; Stockel, J.; Imríšfsek, M.; Hacek, P.; Panek, R.

    2016-10-01

    This paper reports experimental data obtained on the COMPASS tokamak by an array of Langmuir probes embedded in the divertor tiles. The measured current-voltage probe characteristics were processed by the recently published first-derivative probe technique for precise determination of the plasma potential and the electron energy distribution function (EEDF). The measurements were performed during L-mode hydrogen and deuterium plasma with a toroidal magnetic field Bt = 1.15 T, plasma current Ip = 180 kA and average electron density ne = 8*1019 m-3. The spatial profile of the electron temperatures shows that in the vicinity of the inner and outer strike points in hydrogen plasma the EEDF can be approximated by a bi-Maxwellian distribution, with a dominating low-energy electron population (4 - 7 eV) and a minority of electrons with higher energies (12 - 18 eV). In the private flux region between the two strike points, the EEDF is found to be Maxwellian with temperatures in the range of 7 - 9 eV. In the case of deuterium plasma under similar discharge conditions, the EEDFs in the vicinity of the inner and outer strike points, as well as in the private flux region between the two strike points, are found to be bi-Maxwellian.

  6. 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.

  7. 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

  8. 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

  9. OEDGE modeling of DIII-D density scan discharges leading to detachment

    NASA Astrophysics Data System (ADS)

    Elder, J. D.; Stangeby, P. C.; Bray, B. D.; Brooks, N.; Leonard, A. W.; McLean, A. G.; Unterberg, E. A.; Watkins, J. G.

    2015-08-01

    The OEDGE code is used to model the outer divertor plasma for discharges from a density scan experiment on DIII-D with the objective of assessing EIRENE and ADAS hydrogenic emission atomic physics data for Dα, Dβ and Dγ for values of Te and ne characteristic of the range of divertor plasma conditions from attached to weakly detached. Confidence in these values is essential to spectroscopic interpretation of any experiment or modeling effort. Good agreement between experiment and calculated emissions is found for both EIRENE and ADAS calculated emission profiles, confirming their reliability for plasma conditions down to ∼1 eV. For the cold dense plasma conditions characteristic of detachment, it is found that the calculated emissions are especially sensitive to Te.

  10. Modeling of the edge plasma of MAST Upgrade with a Super-X divertor including drifts and an edge transport barrier

    NASA Astrophysics Data System (ADS)

    Rozhansky, V.; Molchanov, P.; Veselova, I.; Voskoboynikov, S.; Kirk, A.; Fishpool, G.; Boerner, P.; Reiter, D.; Coster, D.

    2013-03-01

    The Super-X divertor edge plasma of the future MAST Upgrade tokamak (Lisgo et al 2009 36th EPS Plasma Physics Conf. 33E O4-046, Katramados et al 2011 Fusion Eng. Des. 86 1595, Morris 2012 IEEE Trans. Plasma Sci. 40 682) was simulated with the B2SOLPS5.2 transport code including, for the first time, the effects of drifts due to electric field and magnetic field gradients. The previous simulations did not contain drift effects as well as an account of the external transport barrier. The expected reduction in temperatures and heat flux densities at the low-field side divertors was obtained in the simulations. However, an account of \\vec{{E}}\\times \\vec{{B}} drifts and parallel currents led to an up-down asymmetry of the power to the plates in the connected double null configuration, which was not observed in the simulations without drifts.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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€.

  18. Development of Numerical Tools for the Investigation of Plasma Detachment from Magnetic Nozzles

    NASA Technical Reports Server (NTRS)

    Sankaran, Kamesh; Polzin, Kurt A.

    2007-01-01

    A multidimensional numerical simulation framework aimed at investigating the process of plasma detachment from a magnetic nozzle is introduced. An existing numerical code based on a magnetohydrodynamic formulation of the plasma flow equations that accounts for various dispersive and dissipative processes in plasmas was significantly enhanced to allow for the modeling of axisymmetric domains containing three.dimensiunai momentum and magnetic flux vectors. A separate magnetostatic solver was used to simulate the applied magnetic field topologies found in various nozzle experiments. Numerical results from a magnetic diffusion test problem in which all three components of the magnetic field were present exhibit excellent quantitative agreement with the analytical solution, and the lack of numerical instabilities due to fluctuations in the value of del(raised dot)B indicate that the conservative MHD framework with dissipative effects is well-suited for multi-dimensional analysis of magnetic nozzles. Further studies will focus on modeling literature experiments both for the purpose of code validation and to extract physical insight regarding the mechanisms driving detachment.

  19. M3D-C1 simulations of the plasma response to RMPs in NSTX-U single-null and snowflake divertor configurations

    DOE PAGES

    Canal, G. P.; Ferraro, N. M.; Evans, T. E.; ...

    2017-04-20

    Here in this work, single- and two-fluid resistive magnetohydrodynamic calculations of the plasma response to n = 3 magnetic perturbations in single-null (SN) and snowflake (SF) divertor configurations are compared with those based on the vacuum approach. The calculations are performed using the code M3D-C1 and are based on simulated NSTX-U plasmas. Significantly different plasma responses were found from these calculations, with the difference between the single- and two-fluid plasma responses being caused mainly by the different screening mechanism intrinsic to each of these models. Although different plasma responses were obtained from these different plasma models, no significant difference betweenmore » the SN and SF plasma responses were found. However, due to their different equilibrium properties, magnetic perturbations cause the SF configuration to develop additional and longer magnetic lobes in the null-point region than the SN, regardless of the plasma model used. The intersection of these longer and additional lobes with the divertor plates are expected to cause more striations in the particle and heat flux target profiles. In addition, the results indicate that the size of the magnetic lobes, in both single-null and snowflake configurations, are more sensitive to resonant magnetic perturbations than to non-resonant magnetic perturbations.« less

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

    NASA Astrophysics Data System (ADS)

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

    2017-08-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2017-04-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2017-07-01

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

  3. Two Contemporary Problems in Magnetized Plasmas: The ion-ion hybrid resonator and MHD stability in a snowflake divertor

    NASA Astrophysics Data System (ADS)

    Farmer, William Anthony

    The first part of the dissertation investigates the effects of multiple-ions on the propagation of shear Alfven waves. It is shown that the presence of a second ion-species allows for the formation of an ion-ion hybrid resonator in the presence of a magnetic well. A full-wave description is shown to explain the measured eigenfrequencies and spatial form of the resonator modes identified in experiments in the Large Plasma Device (LAPD) at UCLA. However, it is determined that neither electron collisions or radial convection of the mode due to coupling to either the compressional or ion-Bernstein wave can explain the observed dissipation. Ray tracing studies for shear Alfven waves are performed in various magnetic geometries of contemporary interest. In a tokamak, it is found that the hybrid resonator can exist in the cold-plasma regime, but that ion-temperature effects combined with curvature effects cause the wave reflection point to shift towards the cyclotron frequency of the heavier ion. A one-dimensional WKB model is applied to a tokamak geometry for conditions corresponding to a burning fusion plasma to characterize the resonator. Instability due to fusion-born alpha particles is assessed. An approximate form of the global eigenmode is considered. It is identified that magnetic field shear combined with large ion temperature can cause coupling to an ion-Bernstein wave, which can limit the instability. Finally, the radiation pattern of shear Alfven waves generated by a burst of charged particles in the presence of two-ion species is considered. The spectral content and spatial patterns of the radiated waves are determined. The second part of the dissertation considers the MHD stability of the plasma near a divertor in a tokamak. Two types of modes are considered: a ballooning mode and an axisymmetric, quasi-flute mode. Instability thresholds are derived for both modes and numerically evaluated for parameters relevant to recent experiments. This is done to

  4. Measurements of flows in the DIII-D divertor by Mach probes

    SciTech Connect

    Boedo, J.A.; Lehmer, R.; Moyer, R.A.; Watkins, J.G.; Porter, G.D.; Evans, T.E.; Leonard, A.W.; Schaffer, M.J.

    1998-06-01

    First measurements of Mach number of background plasma in the DIII-D divertor are presented in conjunction with temperature T{sub e} and density n{sub e} using a fast scanning probe array. To validate the probe measurements, the authors compared the T{sub e}, n{sub e} and J{sub sat} data to Thomson scattering data and find good overall agreement in attached discharges and some discrepancy for T{sub e} and n{sub e} in detached discharges. The discrepancy is mostly due to the effect of large fluctuations present during detached plasmas on the probe characteristic; the particle flux is accurately measured in every case. A composite 2-D map of measured flows is presented for an ELMing H-mode discharge and they focus on some of the details. They have also documented the temperature, density and Mach number in the private flux region of the divertor and the vicinity of the X-point, which are important transition regions that have been little studied or modeled. Background parallel plasma flows and electric fields in the divertor region show a complex structure.

  5. 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.

  6. 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.

  7. Moving Divertor Plates in a Tokamak

    SciTech Connect

    S.J. Zweben, H. Zhang

    2009-02-12

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

  8. 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.

  9. 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.

  10. 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.

  11. Testing of plasma facing materials for divertors in the JET neutral-beam test bed

    NASA Astrophysics Data System (ADS)

    Falter, Hans D.; Ciric, Dragoslav; Massmann, P.; Cardella, Antonio; Akiba, Masato

    1993-11-01

    Flat Beryllium tiles brazed to actively cooled CuCrZr withstand power densities up to 17 MW/m2. Tiles with 2 mm thickness have been cycled for 1000 pulses at power densities between 12 and 14 MW/m2. Flat CFC tiles brazed to OFHC copper failed at power densities of 13 MW/m2 with long (10 s) pulses. The fault developed from an area which showed signs of overheating from the start of the test and led to a complete detachment of one tile. For shorter pulse duration (2.5 s) higher power densities (17 MW/m2) could be applied. CFC cubes brazed around a central cooling pipe (monoblocks) have been tested to 15 MW/m2. The power density was limited by the peak surface temperature of the tiles, which exceeded 1500 - 2000 degree(s)C. None of the tiles failed during the test, but there were considerable temperature variations from tile to tile.

  12. 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

  13. 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.

  14. A 250 GHz microwave interferometer for divertor experiments on DIII-D

    SciTech Connect

    James, R.A.; Nilson, D.G.; Stever, R.D.; Hill, D.N.; Casper, T.A.

    1994-01-31

    A new 250 GHz, two-frequency microwave interferometer system has been developed to diagnose divertor plasmas on DIII-D. This diagnostic will measure the line-averaged density across both the inner and outer, lower divertor legs. With a cut-off density of over 7 {times} 10{sup 14} cm{sup {minus}3}, temporal measurements of ELMs, MARFs and plasma detachment are expected. The outer leg system will use a double pass method while the inner leg system will be single pass. Two special 3D carbon composite tiles are used, one to protect the microwave antennas mounted directly under the strike point and the other as the outer leg reflecting surface. Performance, design constraints, and the thermalmechanical design of the 3D carbon composite tiles are discussed.

  15. Experimental determination of plasma detachment from the diverging magnetic nozzle of the VASIMR VX-200 Electric Thruster

    NASA Astrophysics Data System (ADS)

    Olsen, Christopher; Squire, Jared; Longmier, Benjamin; Ballenger, Maxwell; Cassady, Leonard; Carter, Mark; Ilin, Andrew; Cloutier, Paul; Bering, Edgar; Giambusso, Matthew; Ad Astra Rocket Company Team; Rice University Collaboration; University of Houston Collaboration

    2011-10-01

    Theories of magnetized plasma detachment in an expanding magnetic field have been lacking detailed experimental evidence. Recent experiments using a 200 kW class electric rocket (VX-200), run at 100 kW using argon and a peak magnetic field of 2 T, produced ion energies greater than 100 eV with a flux of 2x1022 ions/s in a 150 m3 vacuum facility. Ion-neutral charge exchange effects were reduced and the resultant data show evidence of plasma detachment in a diverging magnetic field on a scale length of 2 m. The detachment is confirmed using multiple plasma diagnostics and magnetic nozzle topologies. Spatial maps of the data are compared to simulations from a particle detachment model, ParTraj, as well as MHD detachment theory. ParTraj, when compared to experiment, is shown to be more consistent in describing the data. Unless the MHD models are modified to incorporation two-fluid effects, single fluid MHD theory is inconsistent with the observations.

  16. 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

  17. 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).

  18. 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.

  19. Espisodic detachment of Martian crustal magnetic fields leading to bulk atmospheric plasma escape

    SciTech Connect

    Brain, D A; Baker, A H; Briggs, J; Eastwood, J P; Halekas, J S; Phan, T

    2009-06-02

    We present an analysis of magnetic field and suprathermal electron measurements from the Mars Global Surveyor (MGS) spacecraft that reveals isolated magnetic structures filled with Martian atmospheric plasma located downstream from strong crustal magnetic fields with respect to the flowing solar wind. The structures are characterized by magnetic field enhancements and rotations characteristic of magnetic flux ropes, and characteristic ionospheric electron energy distributions with angular distributions distinct from surrounding regions. These observations indicate that significant amounts of atmosphere are intermittently being carried away from Mars by a bulk removal process: the top portions of crustal field loops are stretched through interaction with the solar wind and detach via magnetic reconnection. This process occurs frequently and may account for as much as 10% of the total present-day ion escape from Mars.

  20. Thirty-minute plasma sustainment by real-time magnetic-axis swing for effective divertor-load-dispersion in the Large Helical Devicea)

    NASA Astrophysics Data System (ADS)

    Mutoh, T.; Masuzaki, S.; Kumazawa, R.; Seki, T.; Saito, K.; Nakamura, Y.; Kubo, S.; Takeiri, Y.; Shimozuma, T.; Yoshimura, Y.; Igami, H.; Ohkubo, K.; Watanabe, T.; Ogawa, H.; Miyazawa, J.; Shoji, M.; Ashikawa, N.; Nishimura, K.; Sakamoto, M.; Osakabe, M.; Tsumori, K.; Ikeda, K.; Chikaraishi, H.; Funaba, H.; Morita, S.; Goto, M.; Tokuzawa, T.; Takeuchi, N.; Shimpo, F.; Nomura, G.; Takahashi, C.; Yokota, M.; Zhao, Y. P.; Kwak, J. G.; Yamada, H.; Kawahata, K.; Ohyabu, N.; Kaneko, O.; Ida, K.; Nagayama, Y.; Noda, N.; Komori, A.; Sudo, S.; Motojima, O.; LHD Experiment Group

    2006-05-01

    Achieving steady-state plasma operation at high plasma temperatures is one of the important goals of worldwide magnetic fusion research. A high temperature of approximately 2keV, and steady-state plasma-sustainment operation of the Large Helical Device (LHD) [O. Motojima, K. Akaishi, H. Chikaraishi et al., Nucl. Fusion 40, 599 (2000)] is reported. High-temperature plasmas were created and maintained for more than 30min with a world record injected heating power of 1.3GJ. The three-dimensional heat-deposition profile of the LHD helical divertor was modified and during long-pulse discharges it effectively dispersed the heat load using a magnetic-axis swing technique developed at the LHD. A sweep of only 3cm of the major radius of the magnetic axis position (less than 1% of the major radius of the LHD) was enough to disperse the divertor heat load. The modification of the heat-load profile was explained well by field-line tracing. The steady-state plasma was heated and sustained mainly by hydrogen minority ion heating using ion cyclotron range of frequencies. The operation lasted until a sudden increase of radiation loss occurred, presumably because of wall metal flakes dropping into the plasma. The sustained line-averaged electron density was approximately 0.7-0.8×1019m-3. The average input power was 680kW, and the plasma duration was 31min 45s. This successful long operation shows that the heliotron configuration has a high potential as a steady-state fusion reactor.

  1. 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.

  2. Recent progress in R&D on tungsten alloys for divertor structural and plasma facing materials

    SciTech Connect

    Wurster, S.; Baluc, N.; Battabyal, M.; Crosby, T.; Du, J.; Garcia-Rosales, C.; Hasegawa, Akira; Hoffmann, A.; Kimura, A.; Kurishita, Hiroaki; Kurtz, Richard J.; Li, H.; Noh, S.; Reiser, J.; Riesch, J.; Rieth, Michael; Setyawan, Wahyu; Walter, M.; You, J. H.; Pippan, R.

    2013-03-13

    Tungsten materials are candidates for plasma facing components for ITER and DEMO because of their superior thermophysical properties. Knowledge and strategies to improve properties of tungsten-based materials are still under development, as they are not a common structural material such as steel. Consequently, several activities have started in Europe, Japan, USA and China. Research is directed towards manufacturing of new materials based on alloying, microstructure stabilizing and composite formation involving improved processing steps. Beside experimental analyses, work also focuses on computational treatment of open questions, supporting the development of better tungsten materials. Assuming the availability of an ideal material that is ready to use, there remain the questions of inherent safety, the joining of tungsten to steel and the influence of radiation damage. These are topics of increasing interest when the material comes to application.

  3. Thermographic observation of the divertor target plates in the stellarators W7-AS and W7-X

    NASA Astrophysics Data System (ADS)

    Hildebrandt, D.; Gadelmeier, F.; Grigull, P.; McCormick, K.; Naujoks, D.; Sünder, D.; W7-AS Team

    2003-03-01

    Thermography is applied on the stellarator W7-AS to monitor the thermal load of the recently installed divertor targets. A three dimensional numerical code was developed to evaluate power fluxes arriving at the targets from the measured temporal evolution of the surface temperature distribution. Values of the thermal conductivity of the used CFC-target material for all three directions are required for this evaluation and determined by observing the propagation of controlled heat pulses applied by an infrared laser. The evaluation of the thermographic measurements during plasma operation shows characteristic spatial and temporal features of the arrived heat fluxes. Significant features in high density regimes like plasma detachment from the divertor target plates or strongly enhanced localised plasma radiation (MARFE) has been observed by the installed infrared cameras. The implications of these observations for the thermographic system for W7-X are shortly addressed.

  4. Detachment-induced electron production in the early afterglow of pulsed cc-rf oxygen plasmas

    SciTech Connect

    Kuellig, C.; Dittmann, K.; Meichsner, J.

    2012-07-15

    Line integrated electron densities are measured by 160.28 GHz Gaussian beam microwave interferometry in a 10 Hz pulsed (50% duty cycle) cc-rf oxygen discharge, operating at 13.56 MHz. Depending on the processing parameters, the oxygen rf discharge displays two different operation modes regarding its electronegativity. For higher rf power with negative self-bias voltage above -220 V, the oxygen discharge acts as electropositive plasma (n{sub -}/n{sub e} Much-Less-Than 1), whereas at lower rf power and self-bias voltage the plasma becomes strongly electronegative (n{sub -}/n{sub e}>2). In the latter mode, a significant electron density increase is measured in the early afterglow (<100 {mu}s) within a pressure range from 20 to 100 Pa. By use of a simple rate equation model, the temporal behavior of the electron density could be reproduced for both modes of electronegativity. The electron production in the early afterglow is mainly caused due to the detachment of negative atomic oxygen ions by metastable oxygen molecules.

  5. 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.

  6. 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

  7. Numerical simulations used for a validity check on the laser induced photo-detachment diagnostic method in electronegative plasmas

    SciTech Connect

    Oudini, N.; Taccogna, F.; Aanesland, A.

    2014-06-15

    Laser photo-detachment is used as a method to measure or determine the negative ion density and temperature in electronegative plasmas. In essence, the method consists of producing an electropositive channel (negative ion free region) via pulsed laser photo-detachment within an electronegative plasma bulk. Electrostatic probes placed in this channel measure the change in the electron density. A second pulse might be used to track the negative ion recovery. From this, the negative ion density and temperature can be determined. We study the formation and relaxation of the electropositive channel via a two-dimensional Particle-In-Cell/Mote Carlo collision model. The simulation is mainly carried out in a Hydrogen plasma with an electronegativity of α = 1, with a parametric study for α up to 20. The temporal and spatial evolution of the plasma potential and the electron densities shows the formation of a double layer (DL) confining the photo-detached electrons within the electropositive channel. This DL evolves into two fronts that move in the opposite directions inside and outside of the laser spot region. As a consequence, within the laser spot region, the background and photo-detached electron energy distribution function relaxes/thermalizes via collisionless effects such as Fermi acceleration and Landau damping. Moreover, the simulations show that collisional effects and the DL electric field strength might play a non-negligible role in the negative ion recovery within the laser spot region, leading to a two-temperature negative ion distribution. The latter result might have important effects in the determination of the negative ion density and temperature from laser photo detachment diagnostic.

  8. M3D-C1 simulations of the plasma response to n = 3 magnetic perturbations applied to the NSTX-U snowflake divertor

    NASA Astrophysics Data System (ADS)

    Canal, G. P.; Ferraro, N. M.; Evans, T. E.; Osborne, T. H.; Menard, J. E.; Ahn, J.-W.; Maingi, R.; Wingen, A.; Ciro, D.; Frerichs, H.; Schmitz, O.; Soukhanoviskii, V.; Waters, I.

    2016-10-01

    Single- and two-fluid resistive magnetohydrodynamic simulations, performed with the code M3D-C1, are used to investigate the effect of n = 3 magnetic perturbations on the SF divertor configuration. The calculations are based on simulated NSTX-U plasmas and the results show that additional and longer magnetic lobes are created in the null-point region of the SF configuration, compared to those in the conventional single-null. The intersection of these additional and longer lobes with the divertor plates are expected to cause more striations in the particle and heat flux target profiles. In addition, the results indicate that the size of the magnetic lobes, in both single-null and SF configurations, are more sensitive to resonant than to non-resonant magnetic perturbations. The results also suggest that lower values of current in non-axisymmetric control coils close enough to the primary x-point would be required to suppress edge localized modes in plasmas with the SF configuration. This work has been supported by the US Department of Energy, Office of Science, Office of Fusion Energy Science under DOE Award DE-SC0012706.

  9. Two-dimensional full particle simulation of the flow patterns in the scrape-off-layer plasma for upper- and lower-null point divertor configurations in tokamaks

    NASA Astrophysics Data System (ADS)

    Takizuka, T.; Shimizu, K.; Hayashi, N.; Hosokawa, M.; Yagi, M.

    2009-07-01

    The plasma flow in the scrape-off-layer (SOL) plays an important role in particle control in magnetic fusion reactors. The flow is expected to expel helium ashes and to retain impurities in the divertor region, if it is directed towards the divertor plate. It has been experimentally observed, however, that the flow direction is sometimes opposite; from the outer plate side to the SOL middle side in the outer SOL region of tokamaks. In order to study these SOL flow patterns by fully taking account of the kinetic effects, a full particle code, PARASOL, is applied to a tokamak plasma with the upper-null point (UN) or lower-null point (LN) divertor configuration for the downward ion ∇B drift. PARASOL simulations for the medium aspect ratio (A = 5.5) reveal the variation of the flow pattern. For the UN case with the ion ∇B drift away from the null point, the flow velocity Vpar parallel to the magnetic field is formed almost in-out symmetrically. In the inner SOL region Vpar is directed to the inner divertor plate and in the outer SOL Vpar is directed to the outer plate. The stagnation point (Vpar = 0) is located symmetrically at the bottom. On the other hand for the LN case with the ion ∇B drift towards the null point, Vpar in the outer SOL region has a backward flow pattern. The stagnation point moves below the mid-plane of the outer SOL and Vpar in the mid-plane outer SOL is directed to the inner plate. These simulation results are very similar to the experimental results. Simulations are carried out by changing the aspect ratio and by artificially cutting the electric field. It is found that the banana motion of trapped ions is very important for the formation of the flow pattern in addition to the self-consistent electric field. The trapped-ion effects can be stronger than the electric-field effects for the standard tokamaks with A < 5.

  10. 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.

  11. Molecular-level mechanisms of nanoparticle detachment in laser-induced plasma shock waves

    SciTech Connect

    Zhou Dong; Cetinkaya, Cetin

    2006-04-24

    Detachment and detachment mechanisms of nanoparticles from flat surfaces subjected to shock waves are investigated by employing molecular gas dynamic simulations using the direct simulation Monte Carlo method and experimental transient pressure data. Two mechanisms for nanoparticle detachment based on rolling moment resistance of the adhesion bond and the elastic restitution effect are introduced. As a result of present simulations, it is computationally demonstrated that the pulsed laser-induced shock waves can generate sufficient rolling moments to detach sub-100-nm particles and initiate removal. The transient moment exerted on a 60 nm polystyrene latex particle on a silicon substrate is presented and discussed.

  12. Molecular-level mechanisms of nanoparticle detachment in laser-induced plasma shock waves

    NASA Astrophysics Data System (ADS)

    Zhou, Dong; Cetinkaya, Cetin

    2006-04-01

    Detachment and detachment mechanisms of nanoparticles from flat surfaces subjected to shock waves are investigated by employing molecular gas dynamic simulations using the direct simulation Monte Carlo method and experimental transient pressure data. Two mechanisms for nanoparticle detachment based on rolling moment resistance of the adhesion bond and the elastic restitution effect are introduced. As a result of present simulations, it is computationally demonstrated that the pulsed laser-induced shock waves can generate sufficient rolling moments to detach sub-100-nm particles and initiate removal. The transient moment exerted on a 60nm polystyrene latex particle on a silicon substrate is presented and discussed.

  13. On the Measurement of Electron Temperature by Single Langmuir Probes in High Recycling Divertors

    NASA Astrophysics Data System (ADS)

    Pitts, Richard; Horacek, Jan; Loarte, Alberto

    2000-10-01

    Under high recycling and detached conditions, divertor Langmuir probes often yield a significantly higher value of Te than expected. The influence of plasma turbulence and the effect of fast electrons/plasma collisionality are two reasons why this might occur. We concentrate on these two candidates, with particular reference to observations on the TCV tokamak. A systematic study of the effects of noise on simulated probe characteristics at low T_e, shows that the asymmetric, exponential nature of the characteristic favours electron collection such that fluctuations in Vf alone actually tend to reduce the derived Te from that which would otherwise be found. We have also studied the effects of correlated density and potential fluctuations, finding no effect on the fitted T_e. The sheath potential fall energetically filters electrons such that at high densities, the probe measured Te may be characteristic of hotter, more distant zones in the plasma. We use model parallel field profiles of Te and ne generated from B2-Eirene simulations of TCV discharges as input to the analytic theory of Wesson [1] to show how a divertor plate measurement of Te in TCV can exceed the expected value by factors of up to 6 as detachment is approached. [1] J. A. Wesson, Plasma Phys. and Contr. Fusion 37 (1995) 1459

  14. Effect of changes in separatrix magnetic geometry on divertor behaviour in DIII-D

    NASA Astrophysics Data System (ADS)

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

    2013-11-01

    Results and interpretation of recent experiments on DIII-D designed to evaluate divertor geometries favourable for radiative heat dispersal are presented. Two approaches examined here involved lengthening the parallel connection in the scrape-off layer, L‖, and increasing the radius of the outer divertor separatrix strike point, ROSP, with the goal of reducing target temperature, TTAR, and increasing target density, nTAR. From one-dimensional (1D) two-point modelling based on conducted parallel heat flux, it is expected that: n_{TAR} \\propto R_{OSP}^{2} L_{\\parallel}^{6/7} n_{SEP}^{3} and T_{TAR} \\propto R_{OSP}^{-2} L_{\\parallel}^{{-4}/7} n_{SEP}^{-2} , where nSEP is the midplane separatrix density. These scalings suggest that conditions conducive to a radiative divertor solution can be achieved at low nSEP by increasing either ROSP or L‖. Our data are consistent with the above L‖ scalings. On the other hand, the observed dependence of nTAR and TTAR on ROSP displayed a more complex behaviour, under certain conditions deviating from the above scalings. Our analysis indicates that deviations from the ROSP scaling were due to the presence of convected heat flux, driven by escaping neutrals, in the more open configurations of the larger ROSP cases. A comparison of ‘open’ versus ‘closed’ divertor configurations for the H-mode plasmas in this study show that the ‘closed’ case provides at least 30% reduction in the peaked heat flux at common density with the ‘open’ case and partial divertor detachment at lower plasma density.

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

    DOE PAGES

    Briesemeister, Alexis R.; Ahn, Joon -Wook; Canik, John M.; ...

    2017-06-07

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

  16. 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.

  17. The use of carbon fiber composites in divertor target plate tiles and structures

    SciTech Connect

    Altmann, H.; Deksnis, E.; Fanthome, J.; Froger, C.; Lowry, C.; Peacock, A.; Pick, M.A.

    1995-12-31

    Carbon-Carbon fiber composite (CFC), in the form of target plate tiles, is a standard material for intercepting conducted power in divertors. Its good thermal conductivity and ability to withstand high temperatures without melting give it an advantage over other materials for plasma facing components,. Supporting structures are generally made from stainless steels or Inconels. In a radiating divertor, designed to operate with detached plasmas, heat is deposited by radiation on to both plasma facing components and supporting structures. Consequently CFC has been chosen as a structural material to overcome the limitations of metal structures. The various material properties such as thermal conductivity, strength, rigidity and anisotropic behavior need to be taken into account for the design. Tests have been carried out on a range of CFC materials from different manufacturers to determine these properties. Thermal behavior has been studied through power load tests in the JET Neutral Beam Test Bed. Structural properties have been measured through mechanical tests on prototype assemblies. Clamping and joining methods have been developed for the assembly of CFC plates to form structurally and thermally robust tile carriers for a radiating divertor.

  18. 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.

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

    NASA Astrophysics Data System (ADS)

    Rognlien, Thomas

    2014-10-01

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

  20. 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.

  1. Modeling of detachment experiments at DIII-D

    DOE PAGES

    Canik, John M.; Briesemeister, Alexis R.; Lasnier, C. J.; ...

    2014-11-26

    Edge fluid–plasma/kinetic–neutral modeling of well-diagnosed DIII-D experiments is performed in order to document in detail how well certain aspects of experimental measurements are reproduced within the model as the transition to detachment is approached. Results indicate, that at high densities near detachment onset, the poloidal temperature profile produced in the simulations agrees well with that measured in experiment. However, matching the heat flux in the model requires a significant increase in the radiated power compared to what is predicted using standard chemical sputtering rates. Lastly, these results suggest that the model is adequate to predict the divertor temperature, provided thatmore » the discrepancy in radiated power level can be resolved.« less

  2. Modeling of detachment experiments at DIII-D

    SciTech Connect

    Canik, John M.; Briesemeister, Alexis R.; Lasnier, C. J.; Leonard, A. W.; Lore, J. D.; McLean, A. G.; Watkins, J. G.

    2014-11-26

    Edge fluid–plasma/kinetic–neutral modeling of well-diagnosed DIII-D experiments is performed in order to document in detail how well certain aspects of experimental measurements are reproduced within the model as the transition to detachment is approached. Results indicate, that at high densities near detachment onset, the poloidal temperature profile produced in the simulations agrees well with that measured in experiment. However, matching the heat flux in the model requires a significant increase in the radiated power compared to what is predicted using standard chemical sputtering rates. Lastly, these results suggest that the model is adequate to predict the divertor temperature, provided that the discrepancy in radiated power level can be resolved.

  3. The role of parallel and poloidal heat flux in setting the detachment threshold in DIII-D

    NASA Astrophysics Data System (ADS)

    Hill, D. N.; Allen, S. L.; Lasnier, C. J.; McLean, A. G.; Petrie, T. W.; Leonard, A. W.; Groth, M.

    2014-10-01

    Experimental results show that the threshold density for divertor detachment is reduced even as the parallel scrape-off-layer (SOL) heat flux (q| |) is more than doubled, contrary to expectation. The work is part of a systematic study to identify the physics basis for obtaining detached divertors in future high power burning plasma experiments, consistent with requirements for high confinement steady-state operation. Parallel heat flux [PSOL * (Btor /Bpol) / 2 πRλq ; λq is the SOL width] is independent of poloidal flux expansion and is commonly used to quantify the divertor heat flux challenge. In these experiments, the parallel heat flux was varied either by changing the heating power (thereby PSOL), plasma current (the SOL width), or toroidal field (the projection of PSOL onto Btor). The data point to poloidal-field physics effects (e.g., neutral penetration field, line length, and impurity radiation volume) playing a dominant role in setting the detachment threshold. Comparison with 2D simulation will be shown. Work supported by the US DOE under DE-AC52-07NA27344 and DE-FC02-04ER54698.

  4. An improved model to analyze Langmuir probe assisted photo-detachment signal for measuring electronegative plasma parameters

    NASA Astrophysics Data System (ADS)

    Sirse, Nishant; Oudini, Noureddine; Bendib, Abderrezeg; Ellingboe, Albert R.

    2016-09-01

    A diagnostic technique for measuring negative ion parameters based on Langmuir probe assisted laser photo-detachment relies on a theoretical model which relates the rise in the electron saturation current to electronegativity in the plasma. The existing model depend on various assumptions and neglect electrostatic potential barrier formed between the laser column (electropositive column) and the surrounding electronegative plasma in order to prevent the outward flow of electrons from the electropositive plasma column. These assumptions leads to erroneous estimation of the plasma electronegativity. In the present work, we present an analytical model to analyze Langmuir probe assisted photo-detachment signal in order to improve the accuracy of measured electronegativity and extended this technique for measuring electron temperature and charged species density. The analytical model is validated using both experiments and particle-in-cell simulation. The results shows improved accuracy in the measured parameters when compared to existing model. This work was supported by the Korea Institute for the Advancement of Technology and Ministry of Knowledge Economy (L-2010-1438-000), Republic of Korea, Enterprise Ireland and the European Regional Development Fund (ERDF) under NSRF 2007-2013.

  5. Vitreous Detachment

    MedlinePlus

    ... of vision. Treatment How does vitreous detachment affect vision? Although a vitreous detachment does not threaten sight, ... hole or detached retina can lead to permanent vision loss in the affected eye. Those who experience ...

  6. Simulation of Edge Plasmas in DIII-D Double-Null Configurations

    SciTech Connect

    Rensink, M E; Lasnier, C J; Petrie, T W; Porter, G D; Rognlien, T D

    2001-09-06

    We present fluid model simulation results for the edge plasma in the DIII-D tokamak with unbalanced double-null magnetic configurations, including cross field drifts. Input parameters are typical of low-power operation in DIII-D. For high-recycling the plasma tends to be detached from all divertor plates. Midplane plasma and electric field profiles are relatively insensitive to the magnetic imbalance. Divertor heat flux profiles exhibit sharp peaks due to cross-field drifts when the ion grad-B drift direction is away from the x-point toward the magnetic axis.

  7. 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.

  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. 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.

  10. 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.

  11. 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.

  12. 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.

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

    SciTech Connect

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

    1996-05-01

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

  14. 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.

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

    SciTech Connect

    Minjeaud, Sebastian; Pasquetti, Richard

    2016-09-15

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

  16. Comparison of H-mode plasmas in JET-ILW and JET-C with and without nitrogen seeding

    NASA Astrophysics Data System (ADS)

    Jaervinen, A. E.; Giroud, C.; Groth, M.; Belo, P.; Brezinsek, S.; Beurskens, M.; Corrigan, G.; Devaux, S.; Drewelow, P.; Harting, D.; Huber, A.; Jachmich, S.; Lawson, K.; Lipschultz, B.; Maddison, G.; Maggi, C.; Marchetto, C.; Marsen, S.; Matthews, G. F.; Meigs, A. G.; Moulton, D.; Sieglin, B.; Stamp, M. F.; Wiesen, S.; Contributors, JET

    2016-04-01

    In high confinement mode, highly shaped plasmas with edge localized modes in JET, and for heating power of 15-17 MW, the edge fluid code EDGE2D-EIRENE predicts transition to detachment assisted by nitrogen at the low field side (LFS) target when more than 50% of the power crossing the separatrix between ELMs is radiated in the divertor chamber, i.e. ~4 MW. This is observed both in the ITER-like wall (JET-ILW) and in the carbon wall (JET-C) configurations and is consistent with experimental observations within their uncertainty. In these conditions, peak heat fluxes below 1 MW m-2 are measured at the LFS target and predicted for both wall configurations. When the JET-C configuration is replaced with the JET-ILW, a factor of two reduction in the divertor radiated power and 25-50% increase in the peak and total power deposited to the LFS divertor plate is predicted by EDGE2D-EIRENE for unseeded plasmas similar to experimental observations. At the detachment threshold, EDGE2D-EIRENE shows that nitrogen radiates more than 80% of the total divertor radiation in JET-ILW with beryllium contributing less than a few %. With JET-C, nitrogen radiates more than 70% with carbon providing less than 20% of the total radiation. Therefore, the lower intrinsic divertor radiation with JET-ILW is compensated by stronger nitrogen radiation contribution in simulations leading to detachment at similar total divertor radiation fractions. 20-100% higher deuterium molecular fraction in the divertor recycling fluxes is predicted with light JET-C materials when compared to heavy tungsten. EDGE2D-EIRENE simulations indicate that the stronger molecular contribution can reduce the divertor peak power deposition in high recycling conditions by 10-20% due to enhanced power dissipation by molecular interaction.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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

  2. 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.

  3. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2013-09-01

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

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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

  10. 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.

  11. 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.

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

    SciTech Connect

    Zhao, M. L.; Chen, Y. P.; Li, G. Q.; Luo, Z. P.; Guo, H. Y.; 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.

  13. 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.

  14. 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.

  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. 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.

  17. 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).

  18. 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.

  19. Characterisation of the deuterium recycling at the W divertor target plates in JET during steady-state plasma conditions and ELMs

    NASA Astrophysics Data System (ADS)

    Brezinsek, S.; Wiesen, S.; Harting, D.; Guillemaut, C.; Webster, A. J.; Heinola, K.; Meigs, A. G.; Rack, M.; Gao, Y.; Sergienko, G.; Philipps, V.; Stamp, M. F.; Jachmich, S.; Contributors, JET

    2016-02-01

    Experiments in the JET tokamak equipped with the ITER-like wall (ILW) revealed that the inner and outer target plate at the location of the strike points represent after one year of operation intact tungsten (W) surfaces without any beryllium (Be) surface coverage. The dynamics of near-surface retention, implantation, desorption and recycling of deuterium (D) in the divertor of plasma discharges are determined by W target plates. As the W plasma-facing components (PFCs) are not actively cooled, the surface temperature (T surface) is increasing with plasma exposure, varying the balance between these processes in addition to the impinging deuteron fluxes and energies. The dynamic behaviour on a slow time scale of seconds was quantified in a series of identical L-mode discharges (JET Pulse Number (JPN) \\# 81938-73) by intra-shot gas analysis providing the reduction of deuterium retention in W PFCs by 1/3 at a base temperature (T base) range at the outer target plate between 65 °C and 150 °C equivalent to a T surface span of 150 °C and 420 °C. The associated recycling and molecular D desorption during the discharge varies only at lowest temperatures moderately, whereas desorption between discharges rises significantly with increasing T base. The retention measurements represent the sum of inner and outer divertor interaction at comparable T surface. The dynamic behaviour on a fast time scale of ms was studied in a series of identical H-mode discharges (JPN \\#83623-83974) and coherent edge-localized mode (ELM) averaging. High energetic ELMs of about 3 keV are impacting on the W PFCs with fluxes of 3× {10}23 {{{D}}}+ {{{s}}}-1{{{m}}}-2 which is about four times higher than inter-ELM ion fluxes with an impact energy of about E im = 200 eV. This intra-ELM ion flux is associated with a high heat flux of about 60 MW m-2 to the outer target plate which causes T surface rise by Δ T = 100 K per ELM covering finally the range between 160 °C and 1400 °C during the flat

  20. Retinal detachment.

    PubMed

    Cavallerano, A A

    1992-01-01

    Retinal detachment is a separation of the neurosensory retina from the retinal pigment epithelium. Most retinal detachments are rhegmatogenous, and identification of risk factors and predisposing lesions are important aspects of patient management. Retinal detachment is relatively rare, but can pose a significant threat to vision if there is macular involvement. Prompt diagnosis combined with patient education and appropriate intervention often can avert irrevocable visual impairment. This paper presents an overview of the categories of retinal detachment, discusses the pathogenesis of the various types of detachment, and provides recommendations for primary care of patients with predisposing factors and high-risk characteristics.

  1. 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.

  2. DiMES Studies of Temperature Dependence of Carbon Erosion and Re-Deposition in the DIII-D Divertor

    SciTech Connect

    Rudakov, D; Jacob, W; Krieger, K; Litnovsky, A; Philipps, V; West, W; Wong, C; Allen, S; Bastasz, R; Boedo, J; Brooks, N; Boivin, R; De Temmerman, G; Fenstermacher, M; Groth, M; Hollmann, E; Lasnier, C; McLean, A; Moyer, R; Stangeby, P; Wampler, W; Watkins, J; Wienhold, P; Whaley, J

    2006-10-02

    A strong effect of a moderately elevated surface temperature on net carbon deposition and deuterium co-deposition in the DIII-D divertor was observed under detached conditions. A DiMES sample with a gap 2 mm wide and 18 mm deep was exposed to lower-single-null (LSN) L-mode plasmas first at room temperature, and then at 200 C. At the elevated temperature, deuterium co-deposition in the gap was reduced by an order of magnitude. At the plasma-facing surface of the heated sample net carbon erosion was measured at a rate of 3 nm/s, whereas without heating net deposition is normally observed under detachment. In a related experiment three sets of molybdenum mirrors recessed 2 cm below the divertor floor were exposed to identical LSN ELMy H-mode discharges. The first set of mirrors exposed at ambient temperature exhibited net carbon deposition at a rate of up to 3.7 nm/s and suffered a significant drop in reflectivity. In contrast, two other mirror sets exposed at elevated temperatures between 90 C and 175 C exhibited virtually no carbon deposition.

  3. DiMES Studies of Temperature Dependence of Carbon Erosion and Re-Deposition in the DIII-D Divertor

    SciTech Connect

    Rudakov, D L; Jacob, W; Krieger, K; Litnovsky, A; Philipps, V; West, W P; Wong, C C; Allen, S L; Bastasz, R J; Boedo, J A; Brooks, N H; Boivin, R L; De Temmerman, G; Fenstermacher, M E; Groth, M; Hollmann, E M; Lasnier, C J; McLean, A G; Moyer, R A; Stangeby, P C; Wampler, W R; Watkins, J G; Wienhold, P; Whaley, J

    2007-03-15

    A strong effect of a moderately elevated surface temperature on net carbon deposition and deuterium co-deposition in the DIII-D divertor was observed under detached conditions. A DiMES sample with a gap 2 mm wide and 18 mm deep was exposed to lower-single-null (LSN) L-mode plasmas first at room temperature, and then at 200 C. At the elevated temperature, deuterium co-deposition in the gap was reduced by an order of magnitude. At the plasma-facing surface of the heated sample net carbon erosion was measured at a rate of 3 nm/s, whereas without heating net deposition is normally observed under detachment. In a related experiment three sets of molybdenum mirrors recessed 2 cm below the divertor floor were exposed to identical LSN ELMy H-mode discharges. The first set of mirrors exposed at ambient temperature exhibited net carbon deposition at a rate of up to 3.7 nm/s and suffered a significant drop in reflectivity. In contrast, two other mirror sets exposed at elevated temperatures between 90 C and 175 C exhibited practically no carbon deposition.

  4. 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.

  5. Erosion/re-deposition modeling in an ITER divertor-like high-density, low-temperature plasma beam

    NASA Astrophysics Data System (ADS)

    van Swaaij, G. A.; Kirschner, A.; Borodin, D.; Goedheer, W. J.; Bystrov, K.; De Temmerman, G.

    2014-09-01

    Transport of hydrocarbon impurities in a high-density (>1020 m-3), low-temperature (<2 eV) plasma beam was studied with the ERO code. The high ion density and low temperature cause strong Coulomb collisionality between plasma ions and impurity ions. The collisionality is so strong that ions typically do not complete their Larmor orbits. The high collisionality causes impurity entrainment: impurity ions quickly acquire a velocity close to the plasma flow velocity. This causes a relatively high surface impact energy: the calculated mean impact energy of CHx was 8.1 eV in a plasma with Te = 0.7 eV. Simulation results were compared to an a-C : H erosion experiment in the linear plasma generator Pilot-PSI. The large uncertainties in literature values for the sticking probability of hydrocarbon radicals are shown to cause a serious uncertainty in the calculated re-deposition pattern. In contrast, the radial electric field component perpendicular to the axial magnetic field lines did not have a major effect on the redeposition profile.

  6. 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.

  7. Active Control of Power Exhaust in Strongly Heated ASDEX Upgrade Plasmas

    NASA Astrophysics Data System (ADS)

    Dux, Ralph; Kallenbach, Arne; Bernert, Matthias; Eich, Thomas; Fuchs, Christoph; Giannone, Louis; Herrmann, Albrecht; Schweinzer, Josef; Treutterer, Wolfgang

    2012-10-01

    Due to the absence of carbon as an intrinsic low-Z radiator, and tight limits for the acceptable power load on the divertor target, ITER will rely on impurity seeding for radiative power dissipation and for generation of partial detachment. The injection of more than one radiating species is required to optimise the power removal in the main plasma and in the divertor region, i.e. a low-Z species for radiation in the divertor and a medium-Z species for radiation in the outer core plasma. In ASDEX Upgrade, a set of robust sensors, which is suitable to feedback control the radiated power in the main chamber and the divertor as well as the electron temperature at the target, has been developed. Different feedback schemes were applied in H-mode discharges with a maximum heating power of up to 23,W, i.e. at ITER values of P/R (power per major radius) to control all combinations of power flux into the divertor region, power flux onto the target or electron temperature at the target through injection of nitrogen as the divertor radiator and argon as the main chamber radiator. Even at the highest heating powers the peak heat flux density at the target is kept at benign values. The control schemes and the plasma behaviour in these discharges will be discussed.

  8. 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.

  9. 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

  10. 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.

  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. 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.

  13. 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.

  14. Two Contemporary Problems in Magnetized Plasmas: the ion-ion hybrid resonator and MHD stability in a snowflake divertor

    SciTech Connect

    Farmer, William Anthony

    2014-01-01

    The rst part of the dissertation investigates the e ects of multiple-ions on the propagation of shear Alfv en waves. It is shown that the presence of a second ion-species allows for the formation of an ion-ion hybrid resonator in the presence of a magnetic well. A fullwave description is shown to explain the measured eigenfrequencies and spatial form of the resonator modes identi ed in experiments in the Large Plasma Device (LAPD) at UCLA. However, it is determined that neither electron collisions or radial convection of the mode due to coupling to either the compressional or ion-Bernstein wave can explain the observed dissipation.

  15. 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.

  16. 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

  17. Dissociative recombination and electron-impact de-excitation in CH photon emission under ITER divertor-relevant plasma conditions

    NASA Astrophysics Data System (ADS)

    van Swaaij, G. A.; Bystrov, K.; Borodin, D.; Kirschner, A.; van der Vegt, L. B.; van Rooij, G. J.; De Temmerman, G.; Goedheer, W. J.

    2012-09-01

    For understanding carbon erosion and redeposition in nuclear fusion devices, it is important to understand the transport and chemical break-up of hydrocarbon molecules in edge plasmas, often diagnosed by emission of the CH A 2Δ-X 2Π Gerö band around 430 nm. The CH A-level can be excited either by electron-impact (EI) or by dissociative recombination (DR) of hydrocarbon ions. These processes were included in the 3D Monte Carlo impurity transport code ERO. A series of methane injection experiments was performed in the high-density, low-temperature linear plasma generator Pilot-PSI, and simulated emission intensity profiles were benchmarked against these experiments. It was confirmed that excitation by DR dominates at Te < 1.5 eV. The results indicate that the fraction of DR events that lead to a CH radical in the A-level and consequent photon emission is at least 10%. Additionally, quenching of the excited CH radicals by EI de-excitation was included in the modeling. This quenching is shown to be significant: depending on the electron density, it reduces the effective CH emission by a factor of 1.4 at ne = 1.3 × 1020 m-3, to 2.8 at ne = 9.3 × 1020 m-3. Its inclusion significantly improved agreement between experiment and modeling.

  18. 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.

  19. High heat flux testing of divertor plasma facing materials and components using the HHF test facility at IPR

    NASA Astrophysics Data System (ADS)

    Patil, Yashashri; Khirwadkar, S. S.; Belsare, Sunil; Swamy, Rajamannar; Tripathi, Sudhir; Bhope, Kedar; Kanpara, Shailesh

    2016-02-01

    The High Heat Flux Test Facility (HHFTF) was designed and established recently at Institute for Plasma Research (IPR) in India for testing heat removal capability and operational life time of plasma facing materials and components of the ITER-like tokamak. The HHFTF is equipped with various diagnostics such as IR cameras and IR-pyrometers for surface temperature measurements, coolant water calorimetry for absorbed power measurements and thermocouples for bulk temperature measurements. The HHFTF is capable of simulating steady state heat load of several MW m-2 as well as short transient heat loads of MJ m-2. This paper presents the current status of the HHFTF at IPR and high heat flux tests performed on the curved tungsten monoblock type of test mock-ups as well as transient heat flux tests carried out on pure tungsten materials using the HHFTF. Curved tungsten monoblock type of test mock-ups were fabricated using hot radial pressing (HRP) technique. Two curved tungsten monoblock type test mock-ups successfully sustained absorbed heat flux up to 14 MW m-2 with thermal cycles of 30 s ON and 30 s OFF duration. Transient high heat flux tests or thermal shock tests were carried out on pure tungsten hot-rolled plate material (Make:PLANSEE) with incident power density of 0.49 GW m-2 for 20 milliseconds ON and 1000 milliseconds OFF time. A total of 6000 thermal shock cycles were completed on pure tungsten material. Experimental results were compared with mathematical simulations carried out using COMSOL Multiphysics for transient high heat flux tests.

  20. BBQ Modeling of Recycling from the Tore Supra Ergodic Divertor Neutraliser

    NASA Astrophysics Data System (ADS)

    Giannella, R.; Guirlet, R.; Demichelis, C.; Hogan, J.; Cherigier, L.

    1998-11-01

    Generation and recycling of carbon and hydrocarbon impurities, and recycling of neon at the Tore Supra pumped ergodic divertor have been analyzed using the BBQ 3-D scrape-off layer transport code. Code results are compared with spectroscopic observations from fibres located on the neutralizer plates, and background plasma conditions used in the code are constrained with data from langmuir probes embedded in the plates. The sensitivity of neon recycling to assumed reflection coefficients has been studied. A detailed 3-D geometry model for the neutralizer, including all 4 plates, and recycling from the notches between plates, has been prepared. A version of the code describing deuterium processes is being developed to study conditions during the onset of detachment at high density

  1. 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.

  2. Divertor design for the Tokamak Physics Experiment

    SciTech Connect

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

    1994-05-01

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

  3. Divertor 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. Retinal detachment

    MedlinePlus

    ... of the retina responsible for sharp, detailed vision. Symptoms Symptoms of detached retina can include: Bright flashes of ... right away or within a short time after diagnosis. Some types of surgery can be done in ...

  9. Retinal Detachment

    MedlinePlus

    ... immediately. Treatment How is retinal detachment treated? Small holes and tears are treated with laser surgery or ... laser surgery tiny burns are made around the hole to “weld” the retina back into place. Cryopexy ...

  10. Retinal Detachment

    MedlinePlus

    The retina is a layer of tissue in the back of your eye that senses light and sends images ... for reading, driving, and seeing fine detail. A retinal detachment lifts or pulls the retina from its ...

  11. 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.

  12. 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.

  13. 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}.

  14. Retinal detachment

    PubMed Central

    2010-01-01

    Introduction Rhegmatogenous retinal detachment (RRD) is the most common form of retinal detachment, where a retinal "break" allows the ingress of fluid from the vitreous cavity to the subretinal space, resulting in retinal separation. It occurs in about 1 in 10,000 people a year. Methods and outcomes We conducted a systematic review and aimed to answer the following clinical questions: What are the effects of interventions to prevent progression from retinal breaks or lattice degeneration to retinal detachment? What are the effects of different surgical interventions in people with rhegmatogenous retinal detachment? What are the effects of interventions to treat proliferative vitreoretinopathy occurring as a complication of retinal detachment or previous treatment for retinal detachment? We searched: Medline, Embase, The Cochrane Library, and other important databases up to June 2010 (Clinical Evidence reviews are updated periodically; please check our website for the most up-to-date version of this review). We included harms alerts from relevant organisations such as the US Food and Drug Administration (FDA) and the UK Medicines and Healthcare products Regulatory Agency (MHRA). Results We found 21 systematic reviews, RCTs, or observational studies that met our inclusion criteria. We performed a GRADE evaluation of the quality of evidence for interventions. Conclusions In this systematic review, we present information relating to the effectiveness and safety of the following interventions: corticosteroids, cryotherapy, daunorubicin, fluorouracil plus low molecular weight heparin, laser photocoagulation, pneumatic retinopexy, scleral buckling, short-acting or long-acting gas tamponade, silicone oil tamponade, and vitrectomy. PMID:21406128

  15. Retinal detachment

    PubMed Central

    2009-01-01

    Introduction Rhegmatogenous retinal detachment (RRD) is the most common form of retinal detachment, where a retinal "break" allows the ingress of fluid from the vitreous cavity to the subretinal space, resulting in retinal separation. It occurs in about 1 in 10,000 people a year. Methods and outcomes We conducted a systematic review and aimed to answer the following clinical questions: What are the effects of interventions to prevent progression from retinal breaks or lattice degeneration to retinal detachment? What are the effects of different surgical interventions in people with rhegmatogenous retinal detachment? What are the effects of interventions to treat proliferative vitreoretinopathy occurring as a complication of retinal detachment or previous treatment for retinal detachment? We searched: Medline, Embase, The Cochrane Library, and other important databases up to March 2009 (Clinical Evidence reviews are updated periodically; please check our website for the most up-to-date version of this review). We included harms alerts from relevant organisations such as the US Food and Drug Administration (FDA) and the UK Medicines and Healthcare products Regulatory Agency (MHRA). Results We found 20 systematic reviews, RCTs, or observational studies that met our inclusion criteria. We performed a GRADE evaluation of the quality of evidence for interventions. Conclusions In this systematic review, we present information relating to the effectiveness and safety of the following interventions: corticosteroids; cryotherapy; daunorubicin; fluorouracil plus low-molecular-weight heparin; laser photocoagulation; pneumatic retinopexy; scleral buckling; short-acting or long-acting gas tamponade; silicone oil tamponade; and vitrectomy. PMID:19450333

  16. The Lithium Vapor Box Divertor

    NASA Astrophysics Data System (ADS)

    Goldston, Robert; Hakim, Ammar; Hammett, Gregory; Jaworski, Michael; Myers, Rachel; Schwartz, Jacob

    2015-11-01

    Projections of scrape-off layer width to a demonstration power plant suggest an immense parallel heat flux, of order 12 GW/m2, which will necessitate nearly fully detached operation. Building on earlier work by Nagayama et al. and by Ono et al., we propose to use a series of differentially pumped boxes filled with lithium vapor to isolate the buffering vapor from the main plasma chamber, allowing stable detachment. This powerful differential pumping is only available for condensable vapors, not conventional gases. We demonstrate the properties of such a system through conservation laws for vapor mass and enthalpy, and then include plasma entrainment and ultimately an estimate of radiated power. We find that full detachment should be achievable with little leakage of lithium to the main plasma chamber. We also present progress towards solving the Navier-Stokes equation numerically for the chain of vapor boxes, including self-consistent wall boundary conditions and fully-developed shocks, as well as concepts for an initial experimental demonstration-of-concept. This work supported by DOE Contract No. DE-AC02-09CH11466.

  17. 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.

  18. 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.

  19. Retinal detachment

    PubMed Central

    2014-01-01

    Introduction Rhegmatogenous retinal detachment (RRD) is the most common form of retinal detachment, where a retinal 'break' allows the ingress of fluid from the vitreous cavity to the subretinal space, resulting in retinal separation. It occurs in about 1 in 10,000 people a year. Methods and outcomes We conducted a systematic review and aimed to answer the following clinical questions: What are the effects of different surgical interventions in people with rhegmatogenous retinal detachment? What are the effects of interventions to treat proliferative vitreoretinopathy occurring as a complication of retinal detachment or previous treatment for retinal detachment? We searched: Medline, Embase, The Cochrane Library, and other important databases up to September 2013 (Clinical Evidence reviews are updated periodically; please check our website for the most up-to-date version of this review). We included harms alerts from relevant organisations such as the US Food and Drug Administration (FDA) and the UK Medicines and Healthcare products Regulatory Agency (MHRA). Results We found 14 studies that met our inclusion criteria. We performed a GRADE evaluation of the quality of evidence for interventions. Conclusions In this systematic review, we present information relating to the effectiveness and safety of the following interventions: corticosteroids, daunorubicin, fluorouracil plus low molecular weight heparin, pneumatic retinopexy, scleral buckling, short-acting or long-acting gas tamponade, silicone oil tamponade, and vitrectomy. PMID:24807890

  20. 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.

  1. 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.

  2. 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.

  3. ExB CIRCULATION AT THE TOKAMAK DIVERTOR X-POINT

    SciTech Connect

    M.J. SCHAFFER; B.D. BRAY; J.A.BOEDO; T.N. CARLSTROM; R.J.COLCHIN; J.G. WATKINS

    2000-11-01

    Detailed measurements in two dimensions by probes and Thomson scattering reveal unexpected local electric potential and electron pressure (p{sub e}) maxima near the divertor X-point in L-mode plasmas in the DIII-D tokamak [J.L. Luxon and L.G. Davis, Fusion Technol. 8, 441 (1985)]. The potential drives E x B circulation about the X-point, thereby exchanging plasma between closed and open magnetic surfaces at rates that can be comparable to the total cross-separatrix transport. The potential is consistent with the classical parallel Ohm's law. A simple model is proposed to explain the pressure and potential hills in low power, nearly detached plasmas. Recent two-dimensional edge transport modeling with plasma drifts also shows X-point pressure and potential hills but by a different mechanism. These experimental and theoretical results demonstrate that low power tokamak plasmas can be far from poloidal uniformity in a boundary layer just inside the separatrix. Additional data, though preliminary and incomplete, suggest that E x B circulation across the separatrix might be a common feature of low confinement behavior.

  4. 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

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  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. Scrape-off layer plasma modeling for the DIII-D tokamak

    NASA Astrophysics Data System (ADS)

    Porter, G. D.; Rognlien, T. D.; Allen, S. L.

    1994-09-01

    The behavior of the scrape-off layer (SOL) region in tokamaks is believed to play an important role determining the overall device performance. In addition, control of the exhaust power has become one of the most important issues in the design of future devices such as ITER and TPX. This paper presents the results of application of 2-D fluid models to the DII-D tokamak, and research into the importance of processes which are inadequately treated in the fluid models. Comparison of measured and simulated profiles of SOL plasma parameters suggest the physics model contained in the UEDGE code is sufficient to simulate plasmas which are attached to the divertor plates. Experimental evidence suggests the presence of enhanced plasma recombination and momentum removal leading to the existence of detached plasma states. UEDGE simulation of these plasmas obtains a bifurcation to a low temperature plasma at the divertor, but the plasma remains attached. Understanding the physics of this detachment is important for the design of future devices. Analytic studies of the behavior of SOL plasmas enhance our understanding beyond that achieved with fluid modeling. Analysis of the effect of drifts on sheath structure suggest these drifts may play a role in the detachment process. Analysis of the turbulent-transport equations indicate a bifurcation which is qualitatively similar to the experimentally different behavior of the L- and H-mode SOL. Electrostatic simulations of conducting wall modes suggest possible control of the SOL width by biasing.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 3D effects of edge magnetic field configuration on divertor/scrape-off layer transport and optimization possibilities for a future reactor

    NASA Astrophysics Data System (ADS)

    Kobayashi, M.; Xu, Y.; Ida, K.; Corre, Y.; Feng, Y.; Schmitz, O.; Frerichs, H.; Tabares, F. L.; Evans, T. E.; Coenen, J. W.; Liang, Y.; Bader, A.; Itoh, K.; Yamada, H.; Ghendrih, Ph.; Ciraolo, G.; Tafalla, D.; Lopez-Fraguas, A.; Guo, H. Y.; Cui, Z. Y.; Reiter, D.; Asakura, N.; Wenzel, U.; Morita, S.; Ohno, N.; Peterson, B. J.; Masuzaki, S.

    2015-10-01

    This paper assesses the three-dimensional (3D) effects of the edge magnetic field structure on divertor/scrape-off layer transport, based on an inter-machine comparison of experimental data and on the recent progress of 3D edge transport simulation. The 3D effects are elucidated as a consequence of competition between transports parallel (\\parallel ) and perpendicular (\\bot ) to the magnetic field, in open field lines cut by divertor plates, or in magnetic islands. The competition has strong impacts on divertor functions, such as determination of the divertor density regime, impurity screening and detachment control. The effects of magnetic perturbation on the edge electric field and turbulent transport are also discussed. Parameterization to measure the 3D effects on the edge transport is attempted for the individual divertor functions. Based on the suggested key parameters, an operation domain of the 3D divertor configuration is discussed for future devices.

  19. 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

  20. Overview of Stellarator Divertor Studies: Final Report of LDRD Project 01-ERD-069

    SciTech Connect

    Fenstermacher, M E; Rognlien, T D; Koniges, A; Unmansky, M; Hill, D N

    2003-01-21

    A summary is given of the work carried out under the LDRD project 01-ERD-069 entitled Stellarator Divertor Studies. This project has contributed to the development of a three-dimensional edge-plasma modeling and divertor diagnostic design capabilities at LLNL. Results are demonstrated by sample calculations and diagnostic possibilities for the edge plasma of the proposed U.S. National Compact Stellarator Experiment device. Details of the work are contained in accompanying LLNL reports that have been accepted for publication.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2017-09-01

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

  6. Characterizing the transition from high recycling to partial detachment

    NASA Astrophysics Data System (ADS)

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

    2014-10-01

    Experiments at DIII-D have explored the transition from the high recycling to the partially detached divertor condition in L- and H-mode with an unprecedented level of detail. Improved divertor and core Thomson scattering diagnostics were coupled with high resolution spectroscopic studies of molecular and neutral emissions. 2-D Te and ne profiles of the outer leg reveal the earliest indications of formation of the detachment front at the target plate, reducing local Te at the outer strike point from 8-10 eV to 2-3 eV with a marginal (< 10 %) increase in upstream. These data help guide and expose any missing physics in simulations of detachment onset using state-of-the-art boundary codes, and in predictions for operation with a partial detached divertor in future devices. This work supported in part by the US Department of Energy under DE-AC52-07NA27344 and DE-FC02-04ER54698.

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

    NASA Astrophysics Data System (ADS)

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

    2017-06-01

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

  8. 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.

  9. 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.

  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. 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.

  12. Recent Progress in the NSTX/NSTX-U Lithium Program and Prospects for Reactor-Relevant Liquid-Lithium Based Divertor Development

    SciTech Connect

    M. Ono, et al.

    2012-10-27

    Developing a reactor compatible divertor has been identified as a particularly challenging technology problem for magnetic confinement fusion. While tungsten has been identified as the most attractive solid divertor material, the NSTX/NSTX-U lithium (Li) program is investigating the viability of liquid lithium (LL) as a potential reactor compatible divertor plasma facing component (PFC) . In the near term, operation in NSTX-U is projected to provide reactor-like divertor heat loads < 40 MW/m^2 for 5 s. During the most recent NSTX campaign, ~ 0.85 kg of Li was evaporated onto the NSTX PFCs where a ~50% reduction in heat load on the Liquid Lithium Divertor (LLD) was observed, attributable to enhanced divertor bolometric radiation. This reduced divertor heat flux through radiation observed in the NSTX LLD experiment is consistent with the results from other lithium experiments and calculations. These results motivate an LL-based closed radiative divertor concept proposed here for NSTX-U and fusion reactors. With an LL coating, the Li is evaporated from the divertor strike point surface due to the intense heat. The evaporated Li is readily ionized by the plasma due to its low ionization energies, and the ionized Li ions can radiate strongly, resulting in a significant reduction in the divertor heat flux. Due to the rapid plasma transport in divertor plasma, the radiation values can be significantly enhanced up to ~ 11 MJ/cc of LL. This radiative process has the desired function of spreading the focused divertor heat load to the entire divertor chamber facilitating the divertor heat removal. The LL divertor surface can also provide a "sacrificial" surface to protect the substrate solid material from transient high heat flux such as the ones caused by the ELMs. The closed radiative LLD concept has the advantages of providing some degree of partition in terms of plasma disruption forces on the LL, Li particle divertor retention, and strong divertor pumping action from the

  13. 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.

  14. 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.

  15. Investigation of Main-Chamber and Divertor Recycling in DIII-D Using Tangentially Viewing CID Cameras

    SciTech Connect

    Groth, M; Porter, G D; Petrie, T W; Fenstermacher, M E; Brooks, N H

    2003-06-16

    Measurements of the D{sub {alpha}} emission profiles from the divertor and main chamber region in DIII-D, performed in low-density L-mode, and low and high-density ELMy H-mode plasmas imply that core plasma fueling occurs through the divertor channel. Emission profiles of carbon, combined with UEDGE modeling of the L-mode plasmas, also suggests that chemical sputtering of carbon from the flux surface adjacent to the inner divertor walls, and temperature gradient forces in the scrape-off layer, determine the carbon content of the inner main chamber scrape-off layer.

  16. 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.

  17. 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.

  18. Modeling of advanced divertor configuration on experimental advanced superconducting tokamak by SOLPS5.0/B2.5-Eirene

    SciTech Connect

    Si, H. Xu, G. S.; Xiao, B. J.; Luo, Z. P.; Guo, Y.; Wang, L.; Ding, R.; Guo, H. Y.

    2016-03-15

    Heat exhaust is one of the most challenging issues to be addressed for tokamak magnetic confinement fusion research. Detailed modeling with SOLPS5.0/B2.5-Eirene code package is carried out to examine an alternative advanced divertor configuration, i.e., quasi snowflake (QSF), for long pulse operation in EAST. Comparison is also made with the lower single null (LSN) divertor configuration. SOLPS predicts that the quasi snowflake configuration significantly reduces the peak heat flux at the lower divertor outer target, by a factor of 2–3, owing to the magnetic flux expansion. Furthermore, the density threshold for detachment is much lower for QSF, compared to LSN under the same upstream conditions. This indicates that QSF provides a promising tool for controlling heat flux at divertor target while maintaining a lower separatrix density, which is highly desirable for current drive, thus greatly facilitating long-pulse operation in EAST.

  19. Modeling of advanced divertor configuration on experimental advanced superconducting tokamak by SOLPS5.0/B2.5-Eirene

    NASA Astrophysics Data System (ADS)

    Si, H.; Guo, H. Y.; Xu, G. S.; Xiao, B. J.; Luo, Z. P.; Guo, Y.; Wang, L.; Ding, R.

    2016-03-01

    Heat exhaust is one of the most challenging issues to be addressed for tokamak magnetic confinement fusion research. Detailed modeling with SOLPS5.0/B2.5-Eirene code package is carried out to examine an alternative advanced divertor configuration, i.e., quasi snowflake (QSF), for long pulse operation in EAST. Comparison is also made with the lower single null (LSN) divertor configuration. SOLPS predicts that the quasi snowflake configuration significantly reduces the peak heat flux at the lower divertor outer target, by a factor of 2-3, owing to the magnetic flux expansion. Furthermore, the density threshold for detachment is much lower for QSF, compared to LSN under the same upstream conditions. This indicates that QSF provides a promising tool for controlling heat flux at divertor target while maintaining a lower separatrix density, which is highly desirable for current drive, thus greatly facilitating long-pulse operation in EAST.

  20. 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.

  1. Initial Results of Local Island Divertor Experiments in the Large Helical Device

    SciTech Connect

    Komori, Akio; Morisaki, Tomohiro; Masuzaki, Suguru

    2004-07-15

    A local island divertor (LID) experiment has begun in the Large Helical Device (LHD) to demonstrate improved plasma confinement, and fundamental LID functions were demonstrated in the sixth experimental campaign in 2002-2003. It was clearly shown that when an m/n = 1/1 island is generated by adding a resonant perturbation field to the LHD magnetic configuration, the particle flow is guided along the island separatrix to the backside of the island, where carbon plates are located on a divertor head. The particles recycled there are pumped out efficiently so that the line-averaged core plasma density is reduced by a factor of {approx}2 at the same gas puff rate, compared with non-LID discharges. Obvious improvement of the global plasma confinement was, however, not observed yet, because the discharge could not be optimized, due to a large amount of outgas from the divertor head to the core plasma. The size of the divertor head was found to be larger than the optimum one; hence, the core plasma impacted slightly on the core plasma-facing portion of the divertor head with which the core plasma was not expected to collide.

  2. 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.

  3. Interpretations of the impact of cross-field drifts on divertor flows in DIII-D with UEDGE

    DOE PAGES

    Jaervinen, Aaro E.; Allen, Steve L.; Groth, Mathias; ...

    2017-01-27

    Simulations using the multi-fluid code UEDGE indicates that, in low confinement (Lmode) plasmas in DIII-D, recycling driven flows dominate poloidal particle flows in the divertor, whereas E×B drift flows dominate the radial particle flows. In contrast, in high confinement (H-mode) conditions E×B drift flows dominate both poloidal and radial particle flows in the divertor. UEDGE indicates that the toroidal C2+ flow velocities in the divertor plasma are entrained within 30% to the background deuterium flow in both Land H-mode plasmas in the plasma region where the CIII 465 nm emission is measured. Therefore, UEDGE indicates that the Carbon Doppler Coherencemore » Imaging System (CIS), measuring the toroidal velocity of the C2+ ions, can provide insight to the deuterium flows in the divertor. Parallel-to-B velocity dominates the toroidal divertor flow; direct drift impact being less than 1%. Toroidal divertor flow is predicted to reverse when the magnetic field is reversed. This is explained by the parallel-B flow towards the nearest divertor plate corresponding to opposite toroidal directions in opposite toroidal field configurations. Due to strong poloidal E×B flows in H-mode, net poloidal particle transport can be in opposite direction than the poloidal component of the parallel-B plasma flow.« less

  4. The ratio of specific heats for postshock plasmas of a detached bow shock - An MHD model. [in solar wind-earth interaction

    NASA Technical Reports Server (NTRS)

    Chao, J. K.; Wiskerchen, M. J.

    1974-01-01

    The empirical relationship between the standoff distance of a detached bow shock (generated by the flow of a supersonic gas past an impenetrable obstacle), the size of the obstacle, the Mach number of the gas, and the ratio of specific heats has been generalized to include the magnetic field. The value of the ratio of specific heats (gamma-prime) in the postshock plasma has been calculated in terms of the preshock Alfvenic and sonic Mach numbers and orientation of the magnetic field. The empirical relationship is further generalized by taking into consideration the normal momentum and energy flux due to waves and/or turbulence and/or heat flow in association with high Mach number shocks. The computed value of gamma prime is substantially modified in comparison with that given by the MHD or the gas dynamic model. For this generalized model the computed gamma prime can be considered to be a more precise thermodynamic quantity, since the macroscopic parameters of the plasma have been separated out. Application of this empirical relationship to the earth's bow shock has been given.

  5. Divertor Target Heat Load Reduction by Electrical Biasing, and Application to COMPASS-D

    SciTech Connect

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

    2001-03-07

    A toroidally-asymmetric potential structure in the scrape-off layer (SOL) plasma may be formed by toroidally distributed electrical biasing of the divertor target tiles. The resulting ExB convective motions should increase the plasma radial transport in the SOL and thereby reduce the heat load at the divertor [1]. In this paper we develop theoretical modeling and describe the implementation of this concept to the COMPASS-D divertor. We show that strong magnetic shear near the X-point should cause significant squeezing of the convective cells preventing convection from penetrating above the X-point. This should result in reduced heat load at the divertor target without increasing the radial transport in the portion of the SOL in direct contact with the core plasma, potentially avoiding any confinement degradation. implementation of divertor biasing is in hand on COMPASS-D involving insulation of, and modifications to, the present divertor tiles. Calculations based on measured edge parameters suggest that modest currents {approx} 8 A/tile are required, at up to 150V, to drive the convection. A technical test is preceeding full bias experiments.

  6. Divertor target heat load reduction by electrical biasing, and application to COMPASS-D

    NASA Astrophysics Data System (ADS)

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

    2001-03-01

    A toroidally asymmetric potential structure in the scrape-off layer (SOL) plasma may be formed by toroidally distributed electrical biasing of the divertor target tiles. The resulting E× B convective motions should increase the plasma radial transport in the SOL and thereby reduce the heat load at the divertor [R.H. Cohen, D.D. Ryutov, Nucl. Fus. 37 (1997) 621]. In this paper, we develop theoretical modelling and describe the implementation of this concept to the COMPASS-D divertor. We show that a strong magnetic shear near the X-point should cause significant squeezing of the convective cells preventing convection from penetrating above the X-point. This should result in reduced heat load at the divertor target without increasing the radial transport in the portion of the SOL in direct contact with the core plasma, potentially avoiding any confinement degradation. Implementation of divertor biasing is in hand on COMPASS-D involving insulation of, and modifications to, the present divertor tiles. Calculations based on measured edge parameters suggest that modest currents ˜8 A/tile are required, at up to 150 V, to drive the convection. A technical test is preceding full bias experiments.

  7. 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.

  8. 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.

  9. Analytical model of particle and heat flux collection by dust immersed in dense magnetized plasmas

    NASA Astrophysics Data System (ADS)

    Vignitchouk, L.; Ratynskaia, S.; Tolias, P.

    2017-10-01

    A comprehensive analytical description is presented for the particle and heat fluxes collected by dust in dense magnetized plasmas. Compared to the widely used orbital motion limited theory, the suppression of cross-field transport leads to a strong reduction of the electron fluxes, while ion collection is inhibited by thin-sheath effects and the formation of a potential overshoot along the field lines. As a result, the incoming heat flux loses its sensitivity to the floating potential, thereby diminishing the importance of electron emission processes in dust survivability. Numerical simulations implementing the new model for ITER-like detached divertor plasmas predict a drastic enhancement of the dust lifetime.

  10. 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.

  11. 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.

  12. 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

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. U.S. Assessment of advanced limiter-divertor plasma-facing systems (ALPS) design, analysis, and R and D needs.

    SciTech Connect

    Mattas, R. F.

    1998-06-17

    The purpose of the ALPS program is to identify and evaluate advanced limiter/diverter systems that will enhance the attractiveness of fusion power. The highest priority goals at present are achieving high power density, up to 50 MW/m{sup 2}, and showing compatibility of plasma-facing surfaces with plasma operation. Personnel representing a wide range of disciplines from a number of institutions are engaged in the program, where an evaluation phase of the program is planned for three years. Successful identification of promising concepts in the evaluation phase should lead to an R and D phase that includes proof-of-principle experiments.

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

    NASA Astrophysics Data System (ADS)

    Whyte, Dennis; ADX Team

    2015-11-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2016-10-01

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

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

    NASA Astrophysics Data System (ADS)

    Rimza, Sandeep; Khirwadkar, Samir; Velusamy, Karupanna

    2017-04-01

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

  1. Flute instability and the associated radial transport in the tandem mirror with a divertor mirror cell

    SciTech Connect

    Katanuma, I.; Yagi, K.; Haraguchi, Y.; Ichioka, N.; Masaki, S.; Ichimura, M.; Imai, T.

    2010-11-15

    The flute instability and the associated radial transport are investigated in the tandem mirror with a divertor mirror cell (the GAMMA10 A-divertor) with help of computer simulation, where GAMMA10 is introduced [Inutake et al., Phys. Rev. Lett. 55, 939 (1985)]. The basic equations used in the simulation were derived on the assumption of an axisymmetric magnetic field. So the high plasma pressure in a nonaxisymmetric minimum-B anchor mirror cell, which is important for the flute mode stability, is taken into account by redefining the specific volume of a magnetic field line. It is found that the flute modes are stabilized by the minimum-B magnetic field even with a divertor mirror although its stabilizing effects are weaker than that without the divertor mirror. The flute instability enhances the radial transport by intermittently repeating the growing up and down of the Fourier amplitude of the flute instability in time.

  2. Analysis of pumping requirement for exhausting duct in close vicinity of divertor in Tokamak Reactor

    SciTech Connect

    Saito, S.; Abe, T.; Fujisawa, N.; Sugihara, M.; Veda, K.

    1983-11-01

    An improved method for Monte Carlo simulation is described to calculate the neutral-particle transport in a divertor throat and to evaluate the helium removal efficiency from a burning plasma. The required pumping speed for the helium removal is discussed with special emphasis placed on the effects of long exhausting duct and of scrape-off plasma variables. The analysis for International Tokamak Reactor (INTOR) single null divertor suggests a possibility that the pumping requirement for INTOR could be drastically eased--e.g., <10/sup 4/ l/s, for the high scrape-off plasma density of the order of 10/sup 13/ cm/sup -3/.

  3. 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.

  4. 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.

  5. Behavior of 23S metastable state He atoms in low-temperature recombining plasmas

    NASA Astrophysics Data System (ADS)

    Kajita, Shin; Tsujihara, Tadashi; Aramaki, Mitsutoshi; van der Meiden, Hennie; Oshima, Hiroshi; Ohno, Noriyasu; Tanaka, Hirohiko; Yasuhara, Ryo; Akiyama, Tsuyoshi; Fujii, Keisuke; Shikama, Taiichi

    2017-07-01

    We measured the electron density and temperature using laser Thomson scattering and metastable state (23S) of He atoms by laser absorption spectroscopy in the detached recombining plasmas in the divertor simulator NAGDIS-II. Using the measured electron density and temperature combined with the particle trajectory trace simulation, we discussed the behavior of the metastable state He atoms based on comparisons with the experimental results. It is shown that the metastable state atoms are mainly produced in the peripheral region of the plasma column, where the temperature is lower than the central part, and diffused in the vacuum vessel. It was shown that the 0D model is not valid and the transport of the metastable states is to be taken into account for the population distribution of He atoms in the detached plasmas.

  6. Thermal Analysis to Calculate the Vessel Temperature and Stress in Alcator C-Mod Due to the Divertor Upgrade

    SciTech Connect

    Han Zhang, Peter H. Titus, Robert Ellis, Soren Harrison and Rui Vieira

    2012-08-29

    Alcator C-Mod is planning an upgrade to its outer divertor. The upgrade is intended to correct the existing outer divertor alignment with the plasma, and to operate at elevated temperatures. Higher temperature operation will allow study of edge physics behavior at reactor relevant temperatures. The outer divertor and tiles will be capable of operating at 600oC. Longer pulse length, together with the plasma and RF heat of 9MW, and the inclusion of heater elements within the outer divertor produces radiative energy which makes the sustained operation much more difficult than before. An ANSYS model based on ref. 1 was built for the global thermal analysis of C-Mod. It models the radiative surfaces inside the vessel and between the components, and also includes plasma energy deposition. Different geometries have been simulated and compared. Results show that steady state operation with the divertor at 600oC is possible with no damage to major vessel internal components. The differential temperature between inner divertor structure, or "girdle" and inner vessel wall is ~70oC. This differential temperature is limited by the capacity of the studs that hold the inner divertor backing plates to the vessel wall. At a 70oC temperature differential the stress on the studs is within allowable limits. The thermal model was then used for a stress pass to quantify vessel shell stresses where thermal gradients are significant.

  7. Nanoparticle removal using laser induced plasma (LIP) technique and study of detachment modes based on molecular dynamics simulations

    NASA Astrophysics Data System (ADS)

    Peri, M. D. Murthy

    Nanoparticle contamination is a major problem in many industries. In the semiconductor industry, as the device (integrated circuit) size shrinks with each technological node (DRAM half-pitch), the feature size that has to be fabricated shrinks. Consequently, the minimum tolerable particle defect size also reduces to sub-100 nm level. In order to attain the stringent small size features, Extreme Ultraviolet Lithography (EUVL) technique is being explored in the semiconductor industry. As the EUVL masks are reflective and pellicle free, the cleaning techniques employed to remove the native particle defects must be more effective for the removal of the sub-100 nm particles without any substrate damage. The effectiveness of Laser Induced Plasma (LIP) technique, considered as a next generation cleaning method, for removal of 30 nm PSL particles from silicon substrate was previously demonstrated by our group. In the current study, the removal of 100 nm PSL particles from photomask and 300 nm PSL particles from 500 nm patterns was investigated. It was observed that the patterns were damaged which could be attributed to the radiation heating of the plasma, and this necessitated pressure amplification techniques to amplify the transient pressure and minimize the risk of damage. As a potential solution, shocktubes were designed and transient pressure measurements were carried out in air medium. Also, plasma was generated in water, in order to take advantage of the density of the medium, to generate stronger shocks and consequently higher pressure. The performance of the shocktubes was characterized based on their pressure amplification factor. The shocktubes resulted in a pressure amplification factor of 11 in air. The particle removal experiments with shocktubes on 150 nm patterns were performed and no damage to the patterns was observed. However, there were particle adders due to the ablation of the shocktube material. Molecular Dynamics (MD) simulations were initiated and

  8. ELM PARTICLE AND ENERGY TRANSPORT IN THE SOL AND DIVERTOR OF DIII-D

    SciTech Connect

    FENSTERMACHER,ME; LEONARD,AW; SNYDER,PB; BOEDO,JA; COLCHIN,RJ; GROEBNER,RJ; GRAY,DS; GROTH,M; HOLLMANN,E; LASNIER,CJ; OSBORNE,TH; PETRIE,TW; RUDAKOV,DL; TAKAHASHI,H; WATKINS,JG; ZENG,L

    2003-04-01

    A271 ELM PARTICLE AND ENERGY TRANSPORT IN THE SOL AND DIVERTOR OF DIII-D. Results from a series of dedicated experiments measuring the effect of particle and energy pulses from Type-I Edge Localized Modes (ELMs) in the DIII-D scrape-off layer (SOL) and divertor are compared with a simple model of ELM propagation in the boundary plasma. The simple model asserts that the propagation of ELM particle and energy perturbations is dominated by ion parallel convection along SOL fields lines and the recovery from the ELM perturbation is determined by recycling physics. Time scales associated with the initial changes of boundary plasma parameters are expected to be on the order of the ion transit time from the outer midplane, where the ELM instability is initiated, to the divertor targets. To test the model, the ion convection velocity is changed in the experiment by varying the plasma density. At moderate to high density, n{sub e}/n{sub Gr} = 0.5-0.8, the delays in the response of the boundary plasma to the midplane ELM pulses, the density dependence of those delays and other observations are consistent with the model. However, at the lowest densities, n{sub e}/n{sub Gr} {approx} 0.35, small delays between the response sin the two divertors, and changes in the response of the pedestal thermal energy to ELM events, indicate that additional factors including electron conduction in the SOL, the pre-ELM condition of the divertor plasma, and the ratio of ELM instability duration to SOL transit time, may be playing a role. The results show that understanding the response of the SOL and divertor plasmas to ELMs, for various pre-ELM conditions, is just as important to predicting the effect of ELM pulses on the target surfaces of future devices as is predicting the characteristics of the ELM perturbation of the core plasma.

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

    NASA Astrophysics Data System (ADS)

    Bak, J. G.; Kim, H. S.; Bae, M. K.; Juhn, J. W.; Seo, D. C.; Bang, E. N.; Shim, S. B.; Chung, K. S.; Lee, H. J.; Hong, S. H.

    2015-08-01

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

  10. 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.

  11. The super X divertor (SXD) and a compact fusion neutron source (CFNS)

    SciTech Connect

    Kotschenreuther, M.; Valanju, P.; Mahajan, S.; Zheng, L. J.; Pearlstein, L. D.; Bulmer, R. H.; Canik, John; Maingi, R.

    2010-01-01

    A new magnetic geometry, the super X divertor (SXD), is invented to solve severe heat exhaust problems in high power density fusion plasmas. SXD divertor plates are moved to the largest major radii inside the TF coils, increasing the wetted area by 2-3 and the line length by 2-5. Two-dimensional fluid simulations with SOLPS (Schneider et al 2006 SOLPS 2-D edge calculation code Contrib. Plasma Phys. 46) show a several-fold decrease in divertor heat flux and plasma temperature at the plate. A small high power density tokamak using SXD is proposed, for either (1) useful fusion applications using conservative physics, such as a component test facility (CTF) or fission fusion hybrid, or (2) to develop more advanced physics modes for a pure fusion reactor in an integrated fusion environment.

  12. Transport of Aluminum impurities in Helium Plasma

    NASA Astrophysics Data System (ADS)

    Hajjar, Rima; Hollmann, Eric; Krasheninnikov, Sergei; Doerner, Russell

    2013-10-01

    Impurity radiation losses at the edge of fusion devices are crucial for establishing detached divertor regimes in ITER and future tokamak reactors, despite the problem they cause in reducing plasma efficiency. Complex parallel and cross-field impurity transport suggest a rather fluid description when treating edge dynamics, leading somehow to marginal simulation results of the impurity transport problem. A kinetic description accounting for impurity/plasma collisions should be used instead, generating more details on the collision dynamics, while the relatively high mass difference between colliding particles leads to major simplifications in the physics of the problem. Modeling of Aluminum injection and entrainment into steady-state Helium plasma is presented. Multiple ionization and radial losses are included and numerical results are then compared to experimental data obtained from PISCES machine. Work was supported in part by the DOE Grant DE-FG02-04ER54739 at UCSD.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. A Fast Visible Camera Divertor-Imaging Diagnostic on DIII-D

    SciTech Connect

    Roquemore, A; Maingi, R; Lasnier, C; Nishino, N; Evans, T; Fenstermacher, M; Nagy, A

    2007-06-19

    In recent campaigns, the Photron Ultima SE fast framing camera has proven to be a powerful diagnostic when applied to imaging divertor phenomena on the National Spherical Torus Experiment (NSTX). Active areas of NSTX divertor research addressed with the fast camera include identification of types of EDGE Localized Modes (ELMs)[1], dust migration, impurity behavior and a number of phenomena related to turbulence. To compare such edge and divertor phenomena in low and high aspect ratio plasmas, a multi-institutional collaboration was developed for fast visible imaging on NSTX and DIII-D. More specifically, the collaboration was proposed to compare the NSTX small type V ELM regime [2] and the residual ELMs observed during Type I ELM suppression with external magnetic perturbations on DIII-D[3]. As part of the collaboration effort, the Photron camera was installed recently on DIII-D with a tangential view similar to the view implemented on NSTX, enabling a direct comparison between the two machines. The rapid implementation was facilitated by utilization of the existing optics that coupled the visible spectral output from the divertor vacuum ultraviolet UVTV system, which has a view similar to the view developed for the divertor tangential TV camera [4]. A remote controlled filter wheel was implemented, as was the radiation shield required for the DIII-D installation. The installation and initial operation of the camera are described in this paper, and the first images from the DIII-D divertor are presented.

  18. 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.

  19. Extension of operation regimes and investigation of three-dimensional currentless plasmas in the Large Helical Device

    NASA Astrophysics Data System (ADS)

    Kaneko, O.; Yamada, H.; Inagaki, S.; Jakubowski, M.; Kajita, S.; Kitajima, S.; Kobayashi; Koga, K.; Morisaki, T.; Morita, S.; Mutoh, T.; Sakakibara, S.; Suzuki, Y.; Takahashi, H.; Tanaka, K.; Toi, K.; Yoshimura, Y.; Akiyama, T.; Asahi, Y.; Ashikawa, N.; Chikaraishi, H.; Cooper, A.; Darrow, D. S.; Drapiko, E.; Drewelow, P.; Du, X.; Ejiri, A.; Emoto, M.; Evans, T.; Ezumi, N.; Fujii, K.; Fukuda, T.; Funaba, H.; Furukawa, M.; Gates, D. A.; Goto, M.; Goto, T.; Guttenfelder, W.; Hamaguchi, S.; Hasuo, M.; Hino, T.; Hirooka, Y.; Ichiguchi, K.; Ida, K.; Idei, H.; Ido, T.; Igami, H.; Ikeda, K.; Imagawa, S.; Imai, T.; Isobe, M.; Itagaki, M.; Ito, T.; Itoh, K.; Itoh, S.; Iwamoto, A.; Kamiya, K.; Kariya, T.; Kasahara, H.; Kasuya, N.; Kato, D.; Kato, T.; Kawahata, K.; Koike, F.; Kubo, S.; Kumazawa, R.; Kuwahara, D.; Lazerson, S.; Lee, H.; Masuzaki, S.; Matsuoka, S.; Matsuura, H.; Matsuyama, A.; Michael, C.; Mikkelsen, D.; Mitarai, O.; Mito, T.; Miyazawa, J.; Motojima, G.; Mukai, K.; Murakami, A.; Murakami, I.; Murakami, S.; Muroga, T.; Muto, S.; Nagaoka, K.; Nagasaki, K.; Nagayama, Y.; Nakajima, N.; Nakamura, H.; Nakamura, Y.; Nakanishi, H.; Nakano, H.; Nakano, T.; Narihara, K.; Narushima, Y.; Nishimura, K.; Nishimura, S.; Nishiura, M.; Nunami, Y. M.; Obana, T.; Ogawa, K.; Ohdachi, S.; Ohno, N.; Ohyabu, N.; Oishi, T.; Okamoto, M.; Okamoto, A.; Osakabe, M.; Oya, Y.; Ozaki, T.; Pablant, N.; Peterson, B. J.; Sagara, A.; Saito, K.; Sakamoto, R.; Sakaue, H.; Sasao, M.; Sato, K.; Sato, M.; Sawada, K.; Seki, R.; Seki, T.; Sergeev, V.; Sharapov, S.; Sharov, I.; Shimizu, A.; Shimozuma, T.; Shiratani, M.; Shoji, M.; Sudo, S.; Sugama, H.; Suzuki, C.; Takahata, K.; Takeiri, Y.; Takemura, Y.; Takeuchi, M.; Tamura, H.; Tamura, N.; Tanaka, H.; Tanaka, T.; Tingfeng, M.; Todo, Y.; Tokitani, M.; Tokunaga, K.; Tokuzawa, T.; Tsuchiya, H.; Tsumori, K.; Ueda, Y.; Vyacheslavov, L.; Watanabe, K. Y.; Watanabe, T.; Watanabe, T. H.; Wieland, B.; Yamada, I.; Yamada, S.; Yamamoto, S.; Yanagi, N.; Yasuhara, R.; Yokoyama, M.; Yoshida, N.; Yoshimura, S.; Yoshinaga, T.; Yoshinuma, M.; Komori, A.

    2013-10-01

    The progress of physical understanding as well as parameter improvement of net-current-free helical plasma is reported for the Large Helical Device since the last Fusion Energy Conference in Daejeon in 2010. The second low-energy neutral beam line was installed, and the central ion temperature has exceeded 7 keV, which was obtained by carbon pellet injection. Transport analysis of the high-Ti plasmas shows that the ion-thermal conductivity and viscosity decreased after the pellet injection although the improvement does not last long. The effort has been focused on the optimization of plasma edge conditions to extend the operation regime towards higher ion temperature and more stable high density and high beta. For this purpose a portion of the open helical divertors are being modified to the baffle-structured closed ones aimed at active control of the edge plasma. It is compared with the open case that the neutral pressure in the closed helical divertor increased by ten times as predicted by modelling. Studies of physics in a three-dimensional geometry are highlighted in the topics related to the response to a resonant magnetic perturbation at the plasma periphery such as edge-localized-mode mitigation and divertor detachment. Novel approaches of non-local and non-diffusive transport have also been advanced.

  20. 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.

  1. 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.

  2. Posterior vitreous detachment.

    PubMed

    Thimons, J J

    1992-01-01

    Posterior vitreous detachment is an expected consequence of aging, but it can also be the initiating cause of a retinal detachment. To understand the mechanism of posterior vitreous detachment and its sequelae, it is necessary to appreciate the anatomy of the vitreous, its development, and the pathogenesis of vitreous degeneration. This paper is a discussion of these considerations, the types of complications that may result from vitreous detachment, the proper examination of patients who present with the symptoms of vitreous detachment, and appropriate patient management.

  3. Structural studies of deposited layers on JET MkII-SRP inner divertor tiles

    NASA Astrophysics Data System (ADS)

    Likonen, J.; Coad, J. P.; Vainonen-Ahlgren, E.; Renvall, T.; Hole, D. E.; Rubel, M.; Widdowson, A.; JET-EFDA Contributors

    2007-06-01

    Deposited layers formed on JET inner divertor tiles during 1998-2004 and 2001-2004 campaigns have been investigated using secondary ion mass spectrometry (SIMS), Rutherford Backscattering (RBS) and optical microscopy. The thickness of the deposit decreases from the top of vertical tile 1 to the bottom and then increases on vertical tile 3 reaching ∼60 μm. There are even thicker deposits on the small sloping section of the floor tile 4 that can be accessed by the plasma at the inner divertor legs. Deposited films on divertor inner wall tiles are enriched in Be indicating chemical erosion of C and a multi-step transport of C to the shadowed area on floor tile 4. The films have generally a layered and globular structure in the areas with plasma contact.

  4. Dynamics of slab detachment

    NASA Astrophysics Data System (ADS)

    Duretz, T.; Schmalholz, S. M.; Gerya, T. V.

    2012-04-01

    Our study investigates the dynamics of slab detachment and evaluates the amount of time necessary for slabs to detach. We combine both the results of two-dimensional numerical modeling with the prediction of a one-dimensional analytical solution for viscous necking under gravity. This tidy suggest that the dominant deformation mechanisms leading to slab detachment is viscous necking, independently of the depth of slab detachment. Localised simple shear may also occur when the slab dip is moderate, especially in the colder parts of the slab. Brittle fracturing, or breaking, plays a minor role during the slab detachment process. 2D thermo-mechanical simulations indicate that the duration of slab detachment is short (< 4 Ma) and can occur in less than 0.5 Ma. No simple correlation between the slab detachment depth and duration was found. Our results suggest that deep slab detachments (> 250 km) can also occur within a short time (< 1 Ma). On the other hand, slab detachments taking place between 35 and 250 km depth may last less than 2 Ma. This aspect has implications for geodynamic interpretations using slab detachment as explanation for processes such as melting, exhumation or surface uplift.

  5. Assessment of the effect of parallel temperature gradients in the JET SOL on Te measured by divertor target Langmuir probes

    NASA Astrophysics Data System (ADS)

    Ďuran, I.; Ješko, K.; Fuchs, V.; Groth, M.; Guillemaut, C.; Gunn, J. P.; Horacek, J.; Pitts, R. A.; Tskhakaya, D.

    2015-08-01

    Higher than expected electron temperatures (Te) are often measured by divertor Langmuir probes (LP) in high recycling and detached regimes in JET and other tokamaks. As a possible mechanism to explain this discrepancy, we investigate the effect of penetration of fast, almost collisionless electrons connecting the hot upstream scrape-off layer (SOL) region to the divertor targets in JET. We simulate the electron velocity distribution function (EVDF) near the divertor targets using a simple 1D kinetic model using parallel SOL profiles from EDGE2D-EIRENE simulations. The resulting EVDF is used to construct synthetic LP current-voltage (IV) characteristics and evaluation of Te is performed in the same way as for experimental data. Results indicate that the process does not explain the anomalously high Te values estimated from the target probe measurements if the EDGE2D-EIRENE simulated parallel profiles are a good representation of reality.

  6. A tangentially viewing visible TV system for the DIII-D divertor

    SciTech Connect

    Fenstermacher, M.E.; Meyer, W.H.; Wood, R.D.

    1996-02-01

    A video camera system has been installed on the DIII-D tokamak for 2-D spatial studies of line emission in the lower divertor region. The system views the divertor tangentially from an outer port at approximately the height of the X-point. At the tangency plane the entire divertor from inner wall to outside the DIII-D bias ring is viewed with spatial resolution of approximately 1 cm. The image contains information from approximately 90 degrees of toroidal angle. In a recent upgrade, remotely controllable filter changers were added which have produced images from nominally identical shots using a series of spectral lines. Software was developed to calculate the response function matrix using distributed computing techniques and assuming toroidal symmetry. Standard sparse matrix algorithms are then used to invert the 3-D images onto a poloidal plane. Spatial resolution of the inverted images is 2 cm; higher resolution simply increases the size of the response function matrix. Initial results from a series of experiments with multiple identical shots show that the emission from CII and CIII, which appears along the inner scrape-off layer above and below the X-point during ELMing H-mode, moves outward and becomes localized near the X-point in Partially Detached Divertor (PDD) operation.

  7. Engineering design of cryocondensation pumps for the DIII-D Radiative Divertor Program

    SciTech Connect

    Bozek, A.S.; Baxi, C.B.; Del Bene, J.V.; Laughon, G.J.; Reis, E.E.; Shatoff, H.D.; Smith, J.P.

    1995-10-01

    A new double-null, slotted divertor configuration will be installed for the DIII-D Radiative Divertor Program at General Atomics in late 1996. Four cryocondensation pumps, three new and one existing, will be part of this new divertor. The purpose of the pumps is to provide plasma density control and to limit the impurities entering the plasma core by providing pumping at each divertor strike point. The three new pumps are based on the design of the existing pump, installed in 1992 as part of the Advanced Divertor Program. The new pumps require geometry modifications to the original design. Therefore, extensive modal and dynamic analyses were performed to determine the behavior of these pumps and their helium and nitrogen feed lines during disruption events. Thermal and fluid analyses were also performed to characterize the helium two-phase flow regime in the pumps and their feedlines. A flow testing program was completed to test the change in geometry of the pump feed lines with respect to helium flow stability. The results were compared to the helium thermal and fluid analyses to verify predicted flow regimes and flow stability.

  8. Poloidal divertor experiment with applied E vector x B vector/B/sup 2/ drift

    SciTech Connect

    Strait, E J

    1980-05-01

    It has been proposed that the E vector x B vector/B/sup 2/ drift arising from an externally applied electric field could be used in a tokamak or other toroidal device to remove plasma and impurities from the region near the wall and to reduce the amount of plasma striking the wall, either assisting or replacing a conventional magnetic field divertor. A poloidal magnetic divertor (without pumping chamber) was added to the Wisconsin Levitated Toroidal Octupole, and the octupole was operated with a tokamak-like magnetic field configuration (q = 0.7). A radial electric field was applied in the scrape-off zone, causing an E vector x B vector/B/sup 2/ drift with a large poloidal component. This reduced plasma flux reaching the wall of the toroid by up to a factor of 5 beyond the effect of the magnetic divertor, for divertor configurations with both high and low magnetic mirror ratios, in good agreement with a simple theoretical model. Plasma density and density scale length were also reduced in the scrape-off zone, in qualitative agreement with the model. This was not accompanied by any new instabilities in the scrape-off zone, nor by any appreciable degradation of confinement of the central plasma.

  9. Density fluctuations at high density in the ergodic divertor configuration of Tore Supra

    NASA Astrophysics Data System (ADS)

    Devynck, P.; Gunn, J.; Ghendrih, Ph.; Garbet, X.; Antar, G.; Beyer, P.; Boucher, C.; Honore, C.; Gervais, F.; Hennequin, P.; Quémeneur, A.; Truc, A.

    2001-03-01

    The effect of the ergodic divertor on the plasma edge in Tore Supra is to enhance the perpendicular transport through ergodization of the magnetic field lines [Ph. Ghendrih et al., Contrib. Plasma Phys. 32 (3&4) (1992) 179]. Nevertheless, the hot spots observed on the divertor plates during ergodic divertor operation indicate that the cross-field transport driven by the fluctuations is still playing an important role, although measurements by CO 2 laser scattering and reflectometry show a decrease of the turbulence level [J. Payan, X. Garbet, J.H. Chatenet et al., Nucl. Fusion 35 (1995) 1357; P. Beyer, X. Garbet, P. Ghendrih, Phys. Plasmas 5 (12) (1998) 4271]. In order to gain more understanding, fluctuation level and poloidal velocity have been measured with a reciprocating Langmuir probe biased to collect the ion saturation current ( jsat) and with a CO 2 laser scattering diagnostic. Though the relative fluctuation level behaves as previously observed at low density, a new interesting result is that this picture is gradually modified when the density is increased. Both diagnostics observe an increase of δn/ n with density in the ergodic region, which is not the usual behavior observed in limiter configuration. This increase is detected on both sides of the Er inversion radius and is therefore also affecting the plasma bulk. Finally, the confinement time is found to follow an L-mode law at all densities indicating that the ergodic divertor does not change the global confinement properties of the plasma.

  10. Descemet membrane detachment.

    PubMed

    Mackool, R J; Holtz, S J

    1977-03-01

    Four eyes of three patients had extensive postoperative Descemet membrane (DM) detachment. Blood was present just anterior to the DM in three of the four eyes and later converted to and persisted as pigment. Haziness of the cornea at the level of the DM could be seen with reattachment. Detachments of the DM are classified as planar when there is 1 mm or less separation of the DM from its overlying stroma in all areas. Nonplanar DM detachments exceed 1 mm of separation. Planar detachments have a much better prognosis than nonplanar detachments do, with or without descemetopexy. Repair of DM detachments, when necessary, should include air injection, with the lease possible instrumentation of the DM.

  11. Retinal detachment repair

    MedlinePlus

    Scleral buckling; Vitrectomy; Pneumatic retinopexy; Laser retinopexy; Rhegmatogenous retinal detachment repair ... it meets the hole in the retina. Scleral buckling can be done using numbing medicine while you ...

  12. 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.}

  13. 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.

  14. Fabrication and installation of the DIII-D radiative divertor structures

    SciTech Connect

    Hollerbach, M.A.; Smith, J.P.

    1997-11-01

    Phase 1A of the Radiative Divertor Program (RDP) is now installed in the DIII-D tokamak located at General Atomics. This hardware was added to enhance both the Divertor and Advanced Tokamak research elements of the DIII-D program. This installation consists of a divertor baffle enveloping a cryocondensation pump at the upper outer divertor target of DIII-D. The divertor baffle consists of two toroidally continuous Inconel 625 water-cooled rings and a toroidal array of discontinuous radiatively-cooled plates. The water-cooled rings are each comprised of four quadrants, mechanically formed, chem.-milled, and resistance and TIG welded Inconel 625 panels. The supports attaching the panels to the vessel wall are designed to accommodate the differential thermal expansion between the rings and vessel during bake and to react the electromagnetic loads induced during disruptions. They are made from either Inconel 625 or Inconel 718 depending on the stress levels predicted in Finite Element Analysis. Gas seals are designed to limit the leakage from the baffle chamber back to the core plasma to 2,500 {ell}/s and incorporate plasma sprayed alumina to minimize currents flowing through them. The bulk of the water-cooled ring fabrication was performed by a vendor, however, the final machining of penetrations in the conical ring for diagnostic access was performed in-house using a unique machining configuration. This configuration, and the machining of the diagnostic cutouts is described. Graphite tiles were machined from ATJ graphite to form a smooth plasma-facing surface. The installation of all divertor components required only four weeks.

  15. 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.

  16. Deposition of 13C tracer in the JET MkII-HD divertor

    NASA Astrophysics Data System (ADS)

    Likonen, J.; Airila, M.; Alves, E.; Barradas, N.; Brezinsek, S.; Coad, J. P.; Devaux, S.; Groth, M.; Grünhagen, S.; Hakola, A.; Jachmich, S.; Koivuranta, S.; Makkonen, T.; Rubel, M.; Strachan, J.; Stamp, M.; Widdowson, A.; EFDA contributors, JET-

    2011-12-01

    Migration of 13C has been investigated at JET by injecting 13C-labelled methane at the outer divertor base at the end of the 2009 campaign. The 13C deposition profiles on carbon fibre composite divertor tiles were measured by secondary ion mass spectrometry and Rutherford backscattering techniques. 13C was mainly deposited near the puffing location on the outer divertor base tiles. High amounts of 13C were also found at the outer vertical target: at the bottom of the lower and at the top of the upper plates. Thirty-three percent of puffed 13CH4 was instantly pumped out by the divertor cryopump, which is close to the pump duct entrance. Global 13C transport in the torus was modelled by the EDGE2D/EIRENE and DIVIMP codes, and local 13C migration in the vicinity of the injection location by the ERO code. The DIVIMP and EDGE2D simulations show strong prompt deposition of 13C directly adjacent to the injection point as well as in the far scrape-off layer (SOL) along both the inner and outer divertor targets. In addition, the measured 13C deposition along the outer divertor wall tiles is qualitatively reproduced. However, EDGE2D and DIVIMP do not predict any deposition along the divertor surfaces facing the private plasma on the inner floor tile and inboard of the outer strike point on tile 5. The ERO calculations also indicate that most of the deposition occurs close to the injection location on the vertical face of the LBSRP tile and the horizontal part of tile 6.

  17. Cyclic heat load testing of improved CFC/Cu bonding for the W 7-X divertor targets

    NASA Astrophysics Data System (ADS)

    Greuner, H.; Böswirth, B.; Boscary, J.; Chaudhuri, P.; Schlosser, J.; Friedrich, T.; Plankensteiner, A.; Tivey, R.

    2009-04-01

    Extensive high heat flux cycling testing of pre-series targets was performed in the neutral beam facility GLADIS to establish the industrial process for the manufacturing of 890 targets, which will be needed for the installation of the WENDELSTEIN 7-X divertor. The targets are manufactured of flat tiles of CFC NB31 as plasma facing material bonded by an Active Metal Casting copper interlayer onto a water-cooled CuCrZr structure. Based on the results of the 3D thermo-mechanical FEM analysis of the CFC/Cu interface, an additional set of 17 full-scale pre-series elements including three design variations was manufactured by PLANSEE SE. The insertion of an additional plastically compliant copper interlayer between the cooling structure and the Active Metal Casting interlayer showed the best results. No critical tile detachment was observed during >5000 cycles at 10 MW/m 2. These results demonstrated the sufficient life time of the component for the expected heat load in operation.

  18. 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.

  19. Modeling of Detached Solidification

    NASA Technical Reports Server (NTRS)

    Regel, Liya L.; Wilcox, William R.; Popov, Dmitri

    1997-01-01

    Our long term goal is to develop techniques to achieve detached solidification reliably and reproducibly, in order to produce crystals with fewer defects. To achieve this goal it is necessary to understand thoroughly the physics of detached solidification. It was the primary objective of the current project to make progress toward this complete understanding. 'Me products of this grant are attached. These include 4 papers and a preliminary survey of the observations of detached solidification in space. We have successfully modeled steady state detached solidification, examined the stability of detachment, and determined the influence of buoyancy-driven convection under different conditions. Directional solidification in microgravity has often led to ingots that grew with little or no contact with the ampoule wall. When this occurred, crystallographic perfection was usually greatly improved -- often by several orders of magnitude. Indeed, under the Soviet microgravity program the major objective was to achieve detached solidification with its resulting improvement in perfection and properties. Unfortunately, until recently the true mechanisms underlying detached solidification were unknown. As a consequence, flight experiments yielded erratic results. Within the past three years, we have developed a new theoretical model that explains many of the flight results. This model gives rise to predictions of the conditions required to yield detached solidification.

  20. Observation of Non-Maxwellian Electron Distributions in th e NSTX Divertor

    SciTech Connect

    M.A. Jaworski, et. al.

    2013-03-07

    The scrape-off layer plasma at the tokamak region is characterized by open field lines and often contains large variations in plasma properties along these field-lines. Proper characterization of local plasma conditions is critical to assessing plasma-material interaction processes occuring at the target. Langmuir probes are frequently employed in tokamak divertors but are challenging to interpretation. A kinetic interpretation for Langmuir probes in NSTX has yielded non-Maxwellian electron distributions in the divertor characterized by cool bulk populations and energetic tail populations with temperatures of 2-4 times the bulk. Spectroscopic analysis and modeling confirms the bulk plasma temperature and density which can only be obtained with the kinetic interpretation

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

    NASA Astrophysics Data System (ADS)

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

    2017-03-01

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

  2. Impurity diagnosis of a KSTAR graphite divertor tile using laser induced breakdown spectroscopy technique

    NASA Astrophysics Data System (ADS)

    Kim, Minju; Cho, Min Sang; Cho, Byoung Ick

    2017-04-01

    Laser induced breakdown spectroscopy (LIBS) has been tested to diagnose impurity elements on a Korea Superconducting Tokamak Advanced Research (KSTAR) divertor tile. Spectral lines of various impurity elements such as iron, chromium, and nickel were detected from the divertor surface. The variation of spectra with consecutive laser pulses demonstrates the potential for depth profiling analysis for the deposited impurity layer. The LIBS plasma parameters have been qualitatively determined from analysis of the relative line intensities and linewidths for each element. The va