The appearance and propagation of filaments in the private flux region in Mega Amp Spherical Tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harrison, J. R.; Fishpool, G. M.; Thornton, A. J.

2015-09-15

The transport of particles via intermittent filamentary structures in the private flux region (PFR) of plasmas in the MAST tokamak has been investigated using a fast framing camera recording visible light emission from the volume of the lower divertor, as well as Langmuir probes and IR thermography monitoring particle and power fluxes to plasma-facing surfaces in the divertor. The visible camera data suggest that, in the divertor volume, fluctuations in light emission above the X-point are strongest in the scrape-off layer (SOL). Conversely, in the region below the X-point, it is found that these fluctuations are strongest in the PFRmore » of the inner divertor leg. Detailed analysis of the appearance of these filaments in the camera data suggests that they are approximately circular, around 1–2 cm in diameter, but appear more elongated near the divertor target. The most probable toroidal quasi-mode number is between 2 and 3. These filaments eject plasma deeper into the private flux region, sometimes by the production of secondary filaments, moving at a speed of 0.5–1.0 km/s. Probe measurements at the inner divertor target suggest that the fluctuations in the particle flux to the inner target are strongest in the private flux region, and that the amplitude and distribution of these fluctuations are insensitive to the electron density of the core plasma, auxiliary heating and whether the plasma is single-null or double-null. It is found that the e-folding width of the time-average particle flux in the PFR decreases with increasing plasma current, but the fluctuations appear to be unaffected. At the outer divertor target, the fluctuations in particle and power fluxes are strongest in the SOL.« less

Tungsten migration in Alcator C-Mod: sputtering and melting

NASA Astrophysics Data System (ADS)

Wright, G. M.; Barnard, H.; Lipschultz, B.; Whyte, D. G.

2010-11-01

A row of bulk tungsten (W) tiles were installed near the typical outer strike-point location in the Alcator C-Mod divertor in 2007. In the 2009/2010 campaign, one of the W tiles mechanically failed resulting in significant W melting at that location. Post-campaign PIXE surface analysis has been used to observe tungsten (W) deposition and migration patterns in the divertor for the typical operations (sputtering only) and operation with melted components. For sputtering conditions, W deposition of up to 20 nm equivalent thickness is observed at various divertor surfaces indicating prompt re-deposition at the outer divertor, neutral and ion transport through the private-flux region and ion transport in the scrape off layer. For melting conditions, W deposition of up to 400 nm equivalent thickness is observed at some locations at the outer divertor. However, the toroidal distribution of W on the outer divertor is strongly non-uniform. There is no W deposition measured on the inner wall limiter. These results indicate that impurity migration is affected by the erosion mechanism and source, with the migration from melting being less predictable and uniform than from the sputtering case. Supported by USDoE award DE-SC00-02060.

Experimental study of heating scheme effect on the inner divertor power footprint widths in EAST lower single null discharges

NASA Astrophysics Data System (ADS)

Deng, G. Z.; Xu, J. C.; Liu, X.; Liu, X. J.; Liu, J. B.; Zhang, H.; Liu, S. C.; Chen, L.; Yan, N.; Feng, W.; Liu, H.; Xia, T. Y.; Zhang, B.; Shao, L. M.; Ming, T. F.; Xu, G. S.; Guo, H. Y.; Xu, X. Q.; Gao, X.; Wang, L.

2018-04-01

A comprehensive work of the effects of plasma current and heating schemes on divertor power footprint widths is carried out in the experimental advanced superconducting tokamak (EAST). The divertor power footprint widths, i.e., the scrape-off layer heat flux decay length λ q and the heat spreading S, are crucial physical and engineering parameters for fusion reactors. Strong inverse scaling of λ q and S with plasma current have been demonstrated for both neutral beam (NB) and lower hybrid wave (LHW) heated L-mode and H-mode plasmas at the inner divertor target. For plasmas heated by the combination of the two kinds of auxiliary heating schemes (NB and LHW), the divertor power widths tend to be larger in plasmas with higher ratio of LHW power. Comparison between experimental heat flux profiles at outer mid-plane (OMP) and divertor target for NB heated and LHW heated L-mode plasmas reveals that the magnetic topology changes induced by LHW may be the main reason to the wider divertor power widths in LHW heated discharges. The effect of heating schemes on divertor peak heat flux has also been investigated, and it is found that LHW heated discharges tend to have a lower divertor peak heat flux compared with NB heated discharges under similar input power. All these findings seem to suggest that plasmas with LHW auxiliary heating scheme are better heat exhaust scenarios for fusion reactors and should be the priorities for the design of next-step fusion reactors like China Fusion Engineering Test Reactor.

Investigation into the formation of the scrape-off layer density shoulder in JET ITER-like wall L-mode and H-mode plasmas

NASA Astrophysics Data System (ADS)

Wynn, A.; Lipschultz, B.; Cziegler, I.; Harrison, J.; Jaervinen, A.; Matthews, G. F.; Schmitz, J.; Tal, B.; Brix, M.; Guillemaut, C.; Frigione, D.; Huber, A.; Joffrin, E.; Kruzei, U.; Militello, F.; Nielsen, A.; Walkden, N. R.; Wiesen, S.; Contributors, JET

2018-05-01

The low temperature boundary layer plasma (scrape-off layer or SOL) between the hot core and the surrounding vessel determines the level of power loading, erosion and implantation of material surfaces, and thus the viability of tokamak-based fusion as an energy source. This study explores mechanisms affecting the formation of flattened density profiles, so-called ‘density shoulders’, in the low-field side (LFS) SOL, which modify ion and neutral fluxes to surfaces—and subsequent erosion. We find that increases in SOL parallel resistivity, Λdiv (=[L || ν eiΩi]/c sΩe), postulated to lead to shoulder growth through changes in SOL turbulence characteristics, correlates with increases in SOL shoulder amplitude, A s, only under a subset of conditions (D2-fuelled L-mode density scans with outer strike point on the horizontal target). Λdiv fails to correlate with A s for cases of N2 seeding or during sweeping of the strike point across the horizontal target. The limited correlation of Λdiv and A s is also found for H-mode discharges. Thus, while it may be necessary for Λdiv to be above a threshold of ~1 for shoulder formation and/or growth, another mechanism is required. More significantly, we find that in contrast to parallel resistivity, outer divertor recycling, as quantified by the total outer divertor Balmer D α emission, I-D α , does scale with A s where Λdiv does and even where Λdiv does not. Divertor recycling could lead to SOL density shoulder formation through: (a) reducing the parallel to the field flow (loss) of ions out of the SOL to the divertor; and (b) changes in radial electric fields which lead to E  ×  B poloidal flows as well as potentially affecting SOL turbulence birth characteristics. Thus, changes in divertor recycling may be the sole process involved in bringing about SOL density shoulders or it may be that it acts in tandem with parallel resistivity.

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

DOE PAGES

Frerichs, H.; Schmitz, O.; Waters, I.; ...

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 inter- action based onmore » the NSTX-U setup. An overview of different divertor con gurations 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.« less

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.

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

NASA Astrophysics Data System (ADS)

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

2017-12-01

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

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.

Advanced divertor configurations with large flux expansion

NASA Astrophysics Data System (ADS)

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

2013-07-01

Experimental studies of the novel snowflake divertor concept (D. Ryutov, Phys. Plasmas 14 (2007) 064502) performed in the NSTX and TCV tokamaks are reviewed in this paper. The snowflake divertor enables power sharing between divertor strike points, as well as the divertor plasma-wetted area, effective connection length and divertor volumetric power loss to increase beyond those in the standard divertor, potentially reducing heat flux and plasma temperature at the target. It also enables higher magnetic shear inside the separatrix, potentially affecting pedestal MHD stability. Experimental results from NSTX and TCV confirm the predicted properties of the snowflake divertor. In the NSTX, a large spherical tokamak with a compact divertor and lithium-coated graphite plasma-facing components (PFCs), the snowflake divertor operation led to reduced core and pedestal impurity concentration, as well as re-appearance of Type I ELMs that were suppressed in standard divertor H-mode discharges. In the divertor, an otherwise inaccessible partial detachment of the outer strike point with an up to 50% increase in divertor radiation and a peak divertor heat flux reduction from 3-7 MW/m2 to 0.5-1 MW/m2 was achieved. Impulsive heat fluxes due to Type-I ELMs were significantly dissipated in the high magnetic flux expansion region. In the TCV, a medium-size tokamak with graphite PFCs, several advantageous snowflake divertor features (cf. the standard divertor) have been demonstrated: an unchanged L-H power threshold, enhanced stability of the peeling-ballooning modes in the pedestal region (and generally an extended second stability region), as well as an H-mode pedestal regime with reduced (×2-3) Type I ELM frequency and slightly increased (20-30%) normalized ELM energy, resulting in a favorable average energy loss comparison to the standard divertor. In the divertor, ELM power partitioning between snowflake divertor strike points was demonstrated. The NSTX and TCV experiments are providing support for the snowflake divertor as a viable solution for the outstanding tokamak plasma-material interface issues.

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 D 2-seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D 2-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 q peak reduction factors (see standard divertor).« less

ELM induced divertor heat loads on TCV

NASA Astrophysics Data System (ADS)

Marki, J.; Pitts, R. A.; Horacek, J.; Tskhakaya, D.; TCV Team

2009-06-01

Results are presented for heat loads at the TCV outer divertor target during ELMing H-mode using a fast IR camera. Benefitting from a recent surface cleaning of the entire first wall graphite armour, a comparison of the transient thermal response of freshly cleaned and untreated tile surfaces (coated with thick co-deposited layers) has been performed. The latter routinely exhibit temperature transients exceeding those of the clean ones by a factor ˜3, even if co-deposition throughout the first days of operation following the cleaning process leads to the steady regrowth of thin layers. Filaments are occasionally observed during the ELM heat flux rise phase, showing a spatial structure consistent with energy release at discrete toroidal locations in the outer midplane vicinity and with individual filaments carrying ˜1% of the total ELM energy. The temporal waveform of the ELM heat load is found to be in good agreement with the collisionless free streaming particle model.

Controlling marginally detached divertor plasmas

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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 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 (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 » $$\

Quiescence near the X-point of MAST measured by high speed visible imaging

NASA Astrophysics Data System (ADS)

Walkden, N. R.; Harrison, J.; Silburn, S. A.; Farley, T.; Henderson, S. S.; Kirk, A.; Militello, F.; Thornton, A.; The MAST Team

2017-12-01

Using high speed imaging of the divertor volume, the region close to the X-point in MAST is shown to be quiescent. This is confirmed by three different analysis techniques and the quiescent X-point region (QXR) spans from the separatrix to the \\psiN = 1.02 flux surface. Local reductions to the atomic density and effects associated with the camera viewing geometry are ruled out as causes of the QXR, leaving quiescence in the local plasma conditions as being the most likely cause. The QXR is found to be ubiquitous across a significant operational space in MAST including L-mode and H-mode discharges across maximal ranges of 9.8×1019~m-2 in line integrated density, 0.36 MA in plasma current, 0.11 T in toroidal magnetic field and 3.2 MW in NBI power. When mapped to the divertor target the QXR occupies approximately an e-folding length of the heat-flux profile, containing  ∼60% of the total heat flux to the target, and also shows a tendency towards higher frequency shorter lived fluctuations in the ion-saturation current. This is consistent with short-lived divertor localised filamentary structures observed further down the outer divertor leg in the camera images, and suggests a complex multi-region picture of filamentary transport in the divertor.

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 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 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 (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 » $$\

Divertor heat flux mitigation in the National Spherical Torus Experimenta)

NASA Astrophysics Data System (ADS)

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.; NSTX Team

2009-02-01

Steady-state handling of divertor heat flux is a critical issue for both ITER and spherical torus-based devices with compact high power density divertors. Significant reduction of heat flux to the divertor plate has been achieved simultaneously with favorable core and pedestal confinement and stability properties in a highly shaped lower single null configuration in the National Spherical Torus Experiment (NSTX) [M. Ono 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 4-6MWm-2to0.5-2MWm-2 in small-ELM 0.8-1.0MA, 4-6MW neutral beam injection-heated H-mode discharges. A self-consistent picture of the outer strike point partial detachment was evident from divertor heat flux profiles and recombination, particle flux and neutral pressure measurements. Analytic scrape-off layer parallel transport models were used for interpretation of NSTX detachment experiments. The modeling showed that the observed peak heat flux reduction and detachment are possible with high radiated power and momentum loss fractions, achievable with divertor gas injection, and nearly impossible to achieve with main electron density, divertor neutral density or recombination increases alone.

Achievement of radiative feedback control for long-pulse operation on EAST

NASA Astrophysics Data System (ADS)

Wu, K.; Yuan, Q. P.; Xiao, B. J.; Wang, L.; Duan, Y. M.; Chen, J. B.; Zheng, X. W.; Liu, X. J.; Zhang, B.; Xu, J. C.; Luo, Z. P.; Zang, Q.; Li, Y. Y.; Feng, W.; Wu, J. H.; Yang, Z. S.; Zhang, L.; Luo, G.-N.; Gong, X. Z.; Hu, L. Q.; Hu, J. S.; Li, J.

2018-05-01

The active feedback control of radiated power to prevent divertor target plates overheating during long-pulse operation has been developed and implemented on EAST. The radiation control algorithm, with impurity seeding via a supersonic molecular beam injection (SMBI) system, has shown great success in both reliability and stability. By seeding a sequence of short neon (Ne) impurity pulses with the SMBI from the outer mid-plane, the radiated power of the bulk plasma can be well controlled, and the duration of radiative control (feedforward and feedback) is 4.5 s during a discharge of 10 s. Reliable control of the total radiated power of bulk plasma has been successfully achieved in long-pulse upper single null (USN) discharges with a tungsten divertor. The achieved control range of {{f}rad} is 20%–30% in L-mode regimes and 18%–36% in H-mode regimes. The temperature of the divertor target plates was maintained at a low level during the radiative control phase. The peak particle flux on the divertor target was decreased by feedforward Ne injection in the L-mode discharges, while the Ne pulses from the SMBI had no influence on the peak particle flux because of the very small injecting volume. It is shown that although the radiated power increased, no serious reduction of plasma-stored energy or confinement was observed during the control phase. The success of the radiation control algorithm and current experiments in radiated power control represents a significant advance for steady-state divertor radiation and heat flux control on EAST for near-future long-pulse operation.

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 n e, and ELM energy and divertor peak heat flux reduction, especially prominent in radiative D 2-seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D 2-seeded SF divertor at P SOL = 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 P SOL ≃9 MW case. In conclusion, the radiative SF divertor with carbon impurity provides a wider n e operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar q peak reduction factors (cf. standard divertor).« less

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 (B t) 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 (T e) and density (n e) across both divertor legs for individual discharges. The calculations show the same features of in/out plasma asymmetries as measured inmore » the experiment, with the normal B t direction (ion ∇B drift toward the X-point) having higher n e and lower T e 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

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.

Conceptual design study for heat exhaust management in the ARC fusion pilot plant

NASA Astrophysics Data System (ADS)

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

2017-10-01

The ARC pilot plant conceptual design study has been extended to explore solutions for managing heat exhaust resulting from 525 MW of fusion power in a compact (R 3.3 m) tokamak. Superconducting poloidal field coils are configured to produce double-null equilibria that support X-point target divertors while maintaining the original core plasma shape and toroidal field coil size. Long outer divertor legs are appended to the original vacuum vessel, providing both large surface areas for surface dissipation of radiative heat and significantly reduced neutron damage for divertor components. A molten salt FLiBe blanket adequately shields all superconductors and functions as a tritium breeder, with advanced neutronics calculations indicating a tritium breeding ratio of 1.08. In addition, FLiBe is used as the active coolant for the entire vessel. A tungsten swirl-tube cooling channel is implemented in the divertor, capable of exhausting 12 MW/m2, heat flux while keeping total FliBe pumping power below 1% of fusion power. Finally, three novel diagnostics are explored: Cherenkov radiation emitted in FLiBe to measure fusion reaction rate, microwave interferometry to measure divertor detachment front location, and IR imaging through the FLiBe blanket to monitor selected divertor ``hotspots.''

Particle and Power Exhaust in EAST

NASA Astrophysics Data System (ADS)

Wang, Liang; Ding, Fang; Yu, Yaowei; Gan, Kaifu; Liang, Yunfeng; Xu, Guosheng; Xiao, Bingjia; Sun, Youwen; Luo, Guangnan; Gong, Xianzu; Hu, Jiansheng; Li, Jiangang; Wan, Baonian; Maingi, Rajesh; Guo, Houyang; Garofalo, Andrea; EAST Team

2017-10-01

A total power injection up to 0.3GJ has been achieved in EAST long pulse USN operation with ITER-like water-cooling W-monoblock divertor, which has steady-state power exhaust capability of 10 MWm-2. The peak temperature of W target saturated at t = 12 s to the value T 500oC and a heat flux 3MWm-2was maintained. Great efforts to reduce heat flux and accommodate particle exhaust simultaneously have been made towards long pulse of 102s time scale. By exploiting the observation of Pfirsch-Schlüter flow direction in the SOL, the Bt direction with Bx ∇B away from the W divertor (more particles favor outer target in USN) was adopted along with optimizing the strike point location near the pumping slot, to facilitate particle and impurity exhaust with the top cryo-pump. By tailoring the 3D divertor footprint through edge magnetic topology change, the heat load was dispersed widely and thus peak heat flux and W sputtering was well controlled. Active feedback control of total radiative power with neon seeding was achieved within frad = 17-35%, exhibiting further potential for heat flux reduction with divertor and edge radiation. Other heat flux handling techniques, including quasi snowflake configuration, will also be presented.

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   ×  \

Modelling of mitigation of the power divertor loading for the EU DEMO through Ar injection

NASA Astrophysics Data System (ADS)

Subba, Fabio; Aho-Mantila, Leena; Coster, David; Maddaluno, Giorgio; Nallo, Giuseppe F.; Sieglin, Bernard; Wenninger, Ronald; Zanino, Roberto

2018-03-01

In this paper we present a computational study on the divertor heat load mitigation through impurity injection for the EU DEMO. The study is performed by means of the SOLPS5.1 code. The power crossing the separatrix is considered fixed and corresponding to H-mode operation, whereas the machine operating condition is defined by the outboard mid-plane upstream electron density and the impurity level. The selected impurity for this study is Ar, based on its high radiation efficiency at SOL characteristic temperatures. We consider a conventional vertical target geometry for the EU DEMO and monitor target conditions for different operational points, considering as acceptability criteria the target electron temperature (≤5 eV to provide sufficiently low W sputtering rate) and the peak heat flux (below 5-10 MW m-2 to guarantee safe steady-state cooling conditions). Our simulations suggest that, neglecting the radiated power deposition on the plate, it is possible to satisfy the desired constraints. However, this requires an upstream density of the order of at least 50% of the Greenwald limit and a sufficiently high argon fraction. Furthermore, if the radiated power deposition is taken into account, the peak heat flux on the outer plate could not be reduced below 15 MW m-2 in these simulations. As these simulations do not take into account neutron loading, they strongly indicate that the vertical target divertor solution with a radiative front distributed along the divertor leg has a very marginal operational space in an EU DEMO sized reactor.

Developing physics basis for the snowflake divertor in the DIII-D tokamak

DOE PAGES

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

2018-02-01

Recent DIII-D results demonstrate that the snowflake (SF) divertor geometry (cf. standard divertor) enables significant manipulation of divertor heat transport for heat spreading and reduction in attached and radiative divertor regimes, between and during edge localized modes (ELMs), while maintaining good H-mode confinement. Snowflake divertor configurations have been realized in the DIII-D tokamak for several seconds in H-mode discharges with heating power PNBImore » $$\\leqslant$$ 4-5 MW and a range of plasma currents Ip = 0.8-1.2 MA. In this work, inter-ELM transport and radiative SF divertor properties are studied. Significant impact of geometric properties on SOL and divertor plasma parameters, including increased poloidal magnetic flux expansion, divertor magnetic field line length and divertor volume, is confirmed. In the SF-minus configuration, heat deposition is affected by the geometry, and peak divertor heat fluxes are significantly reduced. In the SF-plus and near-exact SF configurations, divertor peak heat flux reduction and outer strike point heat flux profile broadening are observed. Inter-ELM sharing of power and particle fluxes between the main and additional snowflake divertor strike points has been demonstrated. The additional strike points typically receive up to 10-15% of total outer divertor power. Measurements of electron pressure and poloidal beta !p support the theoretically proposed churning mode that is driven by toroidal curvature and vertical pressure gradient in the weak poloidal field region. A comparison of the 4-4.5 MW NBI-heated H-mode plasmas with radiative SF divertor and the standard radiative divertor (both induced with additional gas puffing) shows a nearly complete power detachment and broader divertor radiated power distribution in the SF, as compared to a partial detachment and peaked localized radiation in the standard divertor. However, insignificant difference in the detachment onset w.r.t. density between the SF and the standard divertor was found. The results complement the initial SF divertor studies in the NSTX and DIII-D tokamaks and contribute to the physics basis of the SF divertor as a power exhaust concept for future tokamaks.« less

Developing physics basis for the snowflake divertor in the DIII-D tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

Recent DIII-D results demonstrate that the snowflake (SF) divertor geometry (cf. standard divertor) enables significant manipulation of divertor heat transport for heat spreading and reduction in attached and radiative divertor regimes, between and during edge localized modes (ELMs), while maintaining good H-mode confinement. Snowflake divertor configurations have been realized in the DIII-D tokamak for several seconds in H-mode discharges with heating power PNBImore » $$\\leqslant$$ 4-5 MW and a range of plasma currents Ip = 0.8-1.2 MA. In this work, inter-ELM transport and radiative SF divertor properties are studied. Significant impact of geometric properties on SOL and divertor plasma parameters, including increased poloidal magnetic flux expansion, divertor magnetic field line length and divertor volume, is confirmed. In the SF-minus configuration, heat deposition is affected by the geometry, and peak divertor heat fluxes are significantly reduced. In the SF-plus and near-exact SF configurations, divertor peak heat flux reduction and outer strike point heat flux profile broadening are observed. Inter-ELM sharing of power and particle fluxes between the main and additional snowflake divertor strike points has been demonstrated. The additional strike points typically receive up to 10-15% of total outer divertor power. Measurements of electron pressure and poloidal beta !p support the theoretically proposed churning mode that is driven by toroidal curvature and vertical pressure gradient in the weak poloidal field region. A comparison of the 4-4.5 MW NBI-heated H-mode plasmas with radiative SF divertor and the standard radiative divertor (both induced with additional gas puffing) shows a nearly complete power detachment and broader divertor radiated power distribution in the SF, as compared to a partial detachment and peaked localized radiation in the standard divertor. However, insignificant difference in the detachment onset w.r.t. density between the SF and the standard divertor was found. The results complement the initial SF divertor studies in the NSTX and DIII-D tokamaks and contribute to the physics basis of the SF divertor as a power exhaust concept for future tokamaks.« less

Analysis of drift effects on the tokamak power scrape-off width using SOLPS-ITER

NASA Astrophysics Data System (ADS)

Meier, E. T.; Goldston, R. J.; Kaveeva, E. G.; Makowski, M. A.; Mordijck, S.; Rozhansky, V. A.; Senichenkov, I. Yu; Voskoboynikov, S. P.

2016-12-01

SOLPS-ITER, a comprehensive 2D scrape-off layer modeling package, is used to examine the physical mechanisms that set the scrape-off width ({λq} ) for inter-ELM power exhaust. Guided by Goldston’s heuristic drift (HD) model, which shows remarkable quantitative agreement with experimental data, this research examines drift effects on {λq} in a DIII-D H-mode magnetic equilibrium. As a numerical expedient, a low target recycling coefficient of 0.9 is used in the simulations, resulting in outer target plasma that is sheath limited instead of conduction limited as in the experiment. Scrape-off layer (SOL) particle diffusivity (D SOL) is scanned from 1 to 0.1 m2 s-1. Across this diffusivity range, outer divertor heat flux is dominated by a narrow (˜3-4 mm when mapped to the outer midplane) electron convection channel associated with thermoelectric current through the SOL from outer to inner divertor. An order-unity up-down ion pressure asymmetry allows net ion drift flux across the separatrix, facilitated by an artificial mechanism that mimics the anomalous electron transport required for overall ambipolarity in the HD model. At {{D}\\text{SOL}}=0.1 m2 s-1, the density fall-off length is similar to the electron temperature fall-off length, as predicted by the HD model and as seen experimentally. This research represents a step toward a deeper understanding of the power scrape-off width, and serves as a basis for extending fluid modeling to more experimentally relevant, high-collisionality regimes.

Analysis of drift effects on the tokamak power scrape-off width using SOLPS-ITER

DOE PAGES

Meier, E. T.; Goldston, R. J.; Kaveeva, E. G.; ...

2016-11-02

SOLPS-ITER, a comprehensive 2D scrape-off layer modeling package, is used to examine the physical mechanisms that set the scrape-off width (more » $${{\\lambda}_{q}}$$ ) for inter-ELM power exhaust. Guided by Goldston's heuristic drift (HD) model, which shows remarkable quantitative agreement with experimental data, this research examines drift effects on $${{\\lambda}_{q}}$$ in a DIII-D H-mode magnetic equilibrium. As a numerical expedient, a low target recycling coefficient of 0.9 is used in the simulations, resulting in outer target plasma that is sheath limited instead of conduction limited as in the experiment. Scrape-off layer (SOL) particle diffusivity (D SOL) is scanned from 1 to 0.1 m2 s –1. Across this diffusivity range, outer divertor heat flux is dominated by a narrow (~3–4mm when mapped to the outer midplane) electron convection channel associated with thermoelectric current through the SOL from outer to inner divertor. An order-unity up–down ion pressure asymmetry allows net ion drift flux across the separatrix, facilitated by an artificial mechanism that mimics the anomalous electron transport required for overall ambipolarity in the HD model. At $${{D}_{\\text{SOL}}}=0.1$$ m2 s –1, the density fall-off length is similar to the electron temperature fall-off length, as predicted by the HD model and as seen experimentally. Furthermore, this research represents a step toward a deeper understanding of the power scrape-off width, and serves as a basis for extending fluid modeling to more experimentally relevant, high-collisionality regimes.« less

Developing physics basis for the snowflake divertor in the 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.; Ryutov, D. D.; Kolemen, E.; Groebner, R. J.; Hyatt, A. W.; Leonard, A. W.; Osborne, T. H.; Petrie, T. W.; Watkins, J.

2018-03-01

Recent DIII-D results demonstrate that the snowflake (SF) divertor geometry (see standard divertor) enables significant manipulation of divertor heat transport for heat spreading and reduction in attached and radiative divertor regimes, between and during edge localized modes (ELMs), while maintaining good H-mode confinement. Snowflake divertor configurations have been realized in the DIII-D tokamak for several seconds in H-mode discharges with heating power P_NBI ≤slant 4 -5 MW and a range of plasma currents I_p=0.8-1.2 MA. In this work, inter-ELM transport and radiative SF divertor properties are studied. Significant impact of geometric properties on SOL and divertor plasma parameters, including increased poloidal magnetic flux expansion, divertor magnetic field line length and divertor volume, is confirmed. In the SF-minus configuration, heat deposition is affected by the geometry, and peak divertor heat fluxes are significantly reduced. In the SF-plus and near-exact SF configurations, divertor peak heat flux reduction and outer strike point heat flux profile broadening are observed. Inter-ELM sharing of power and particle fluxes between the main and additional snowflake divertor strike points has been demonstrated. The additional strike points typically receive up to 10-15% of total outer divertor power. Measurements of electron pressure and poloidal beta βp support the theoretically proposed churning mode that is driven by toroidal curvature and vertical pressure gradient in the weak poloidal field region. A comparison of the 4-4.5 MW NBI-heated H-mode plasmas with radiative SF divertor and the standard radiative divertor (both induced with additional gas puffing) shows a nearly complete power detachment and broader divertor radiated power distribution in the SF, as compared to a partial detachment and peaked localized radiation in the standard divertor. However, insignificant difference in the detachment onset w.r.t. density between the SF and the standard divertor was found. The results complement the initial SF divertor studies conducted in high-power H-mode discharges in the NSTX and DIII-D tokamaks, and, along with snowflake divertor results from TCV and other tokamaks, contribute to the physics basis of the SF divertor as a power exhaust concept for future high power density tokamaks.

OEDGE modeling of plasma contamination efficiency of Ar puffing from different divertor locations in EAST

NASA Astrophysics Data System (ADS)

Pengfei, ZHANG; Ling, ZHANG; Zhenwei, WU; Zong, XU; Wei, GAO; Liang, WANG; Qingquan, YANG; Jichan, XU; Jianbin, LIU; Hao, QU; Yong, LIU; Juan, HUANG; Chengrui, WU; Yumei, HOU; Zhao, JIN; J, D. ELDER; Houyang, GUO

2018-04-01

Modeling with OEDGE was carried out to assess the initial and long-term plasma contamination efficiency of Ar puffing from different divertor locations, i.e. the inner divertor, the outer divertor and the dome, in the EAST superconducting tokamak for typical ohmic plasma conditions. It was found that the initial Ar contamination efficiency is dependent on the local plasma conditions at the different gas puff locations. However, it quickly approaches a similar steady state value for Ar recycling efficiency >0.9. OEDGE modeling shows that the final equilibrium Ar contamination efficiency is significantly lower for the more closed lower divertor than that for the upper divertor.

Influence of impurity seeding on the plasma radiation in the EAST tokamak

NASA Astrophysics Data System (ADS)

Liping, DONG; Yanmin, DUAN; Kaiyun, CHEN; Xiuda, YANG; Ling, ZHANG; Feng, XU; Jingbo, CHEN; Songtao, MAO; Zhenwei, WU; Liqun, HU

2018-04-01

Plasma radiation characteristics in EAST argon (Ar) gas and neon (Ne) gas seeding experiments are studied. The radiation profiles reconstructed from the fast bolometer measurement data by tomography method are compared with the ones got from the simulation program based on corona model. And the simulation results coincide roughly with the experimental data. For Ar seeding discharges, the substantial enhanced radiations can be generally observed in the edge areas at normalized radius ρ pol∼0.7–0.9, while the enhanced regions are more outer for Ne seeding discharges. The influence of seeded Ar gas on the core radiation is related to the injected position. In discharges with LSN divertor configuration, the Ar ions can permeate into the core region more easily when being injected from the opposite upper divertor ports. In USN divertor configuration, the W impurity sputtered from the upper divertor target plates are observed to be an important contributor to the increase of the core radiation no matter impurity seeding from any ports. The maximum radiated power fractions f rad (P rad/P heat) about 60%–70% have been achieved in the recent EAST experimental campaign in 2015–2016.

"Snowflake" divertor configuration in NSTX

NASA Astrophysics Data System (ADS)

Soukhanovskii, V. A.; Ahn, J.-W.; Bell, R. E.; Gates, D. A.; Gerhardt, S.; Kaita, R.; Kolemen, E.; Kugel, H. W.; Leblanc, B. P.; Maingi, R.; Maqueda, R.; McLean, A.; Menard, J. E.; Mueller, D. M.; Paul, S. F.; Raman, R.; Roquemore, A. L.; Ryutov, D. D.; Scott, H. A.

2011-08-01

Steady-state handling of divertor heat flux is a critical issue for present and future conventional and spherical tokamaks with compact high power density divertors. A novel "snowflake" divertor (SFD) configuration that takes advantage of magnetic properties of a second-order poloidal null has been predicted to have a larger plasma-wetted area and a larger divertor volume, in comparison with a standard first-order poloidal X-point divertor configuration. The SFD was obtained in 0.8 MA, 4-6 MW NBI-heated H-mode discharges in NSTX using two divertor magnetic coils. The SFD led to a partial detachment of the outer strike point even in low-collisionality scrape-off layer plasma obtained with lithium coatings in NSTX. Significant divertor peak heat flux reduction and impurity screening have been achieved simultaneously with good core confinement and MHD properties.

Numerical study of the current-convective instability driven by asymmetry of detachment in inner and outer divertors

NASA Astrophysics Data System (ADS)

Stepanenko, A. A.; Krasheninnikov, S. I.

2018-01-01

One of the possible mechanisms responsible for strong radiation fluctuations observed in recent experiments with detached plasmas at ASDEX Upgrade [Potzel et al., Nucl. Fusion 54, 013001 (2014)] can be related to the onset of the current-convective instability (CCI) driven by strong asymmetry of detachment in the inner and outer divertors of the tokamak [S. Krasheninnikov and A. Smolyakov, Phys. Plasmas 23, 092505 (2016)]. In this study, we present the physical model, used to simulate the CCI, and the first numerical results of modeling of the CCI dynamics in ASDEX Upgrade-like conditions. The simulation results provide frequency spectra of turbulent divertor plasma oscillations showing reasonably good agreement with the available experimental data.

Quantification of Chemical Erosion in the DIII-D Divertor

NASA Astrophysics Data System (ADS)

McLean, Adam

2009-11-01

Chemical erosion (CE) yield at the graphite divertor target in DIII-D was measured to be substantially lower in cold near-detached plasma conditions compared to well-attached ones, with major implications for ITER. Current estimates of tritium retention by co-deposition with hydrocarbons (HCs) in ITER place potentially severe restrictions on operation. However, calculations done to date have been based on excessively conservative assumptions, due to limited understanding of cold divertor plasmas (1-5eV) which bridge energy thresholds for complex atomic and molecular processes not present in attached conditions. Hydrocarbon injection through a unique porous graphite plate which realistically simulates secondary reactions of HCs with a graphite surface has been used to measure CE in-situ. For the first time in a divertor, measurements were made at extrinsic CH4 injection rates comparable to the expected intrinsic CE rate of C, with the resulting spectroscopic emissions separated from those of the intrinsic sources. Under cold plasma conditions the contribution of CE-produced C relative to total C sources in the divertor declined dramatically from ˜50% to <15%. Photon efficiencies for products from the breakup of injected CH4 were greater than previous measurements at higher puff rates, indicating the importance of minimizing perturbation to the local plasma. At 350K, the measured CE yield near the outer strike point was ˜2.6% in attachment dropping to only ˜0.5% in cold plasma; results are consistent with some theoretical predications and lab studies. Under full detachment, near total extinction of the CD band occurred, consistent with suppression of net C erosion. These findings have potentially major impact on projected target lifetime and tritium retention in future reactors, and for the PFC choice in ITER.

Characterization of chemical sputtering using the Mark II DiMES porous plug injector in attached and semi-detached divertor plasmas of DIII-D

NASA Astrophysics Data System (ADS)

McLean, A. G.; Davis, J. W.; Stangeby, P. C.; Allen, S. L.; Boedo, J. A.; Bray, B. D.; Brezinsek, S.; Brooks, N. H.; Fenstermacher, M. E.; Groth, M.; Haasz, A. A.; Hollmann, E. M.; Isler, R. C.; Lasnier, C. J.; Mu, Y.; Petrie, T. W.; Rudakov, D. L.; Watkins, J. G.; West, W. P.; Whyte, D. G.; Wong, C. P. C.

2009-06-01

An improved, self-contained gas injection system for the divertor material evaluation system (DiMES) on DIII-D has been employed for in situ study of chemical erosion in the tokamak divertor environment. To minimize perturbation to local plasma, the Mark II porous plug injector (PPI) releases methane through a porous graphite surface at the outer strike point at a rate precisely controlled by a micro-orifice flow restrictor to be approximately equal as that predicted for intrinsic chemical sputtering. Effective photon efficiencies resulting from CH 4 are found to be 58 ± 12 in an attached divertor ( ne ˜ 1.5 × 10 13/cm 3, Te ˜ 25 eV, Tsurf ˜ 450 K), and 94 ± 20 in a semi-detached cold divertor ( ne ˜ 6.0 × 10 13/cm 3, Te ˜ 2-3 eV, Tsurf ˜ 350 K). These values are significantly more than previous measurements in similar plasma conditions, indicating the importance of the injection rate and local re-erosion for the integrity of this analysis. The contribution of chemical versus physical sputtering to the source of C + at the target is assessed through simultaneous measurement of CII line, and CD plus CH-band emissions during release of CH 4 from the PPI, then compared with that seen in intrinsic sputtering.

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

NASA Astrophysics Data System (ADS)

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

2017-10-01

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

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.

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

NASA Astrophysics Data System (ADS)

Ogorodnikova, O. V.

2008-04-01

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

Particle-in-cell simulations of the plasma interaction with poloidal gaps in the ITER divertor outer vertical target

NASA Astrophysics Data System (ADS)

Komm, M.; Gunn, J. P.; Dejarnac, R.; Pánek, R.; Pitts, R. A.; Podolník, A.

2017-12-01

Predictive modelling of the heat flux distribution on ITER tungsten divertor monoblocks is a critical input to the design choice for component front surface shaping and for the understanding of power loading in the case of small-scale exposed edges. This paper presents results of particle-in-cell (PIC) simulations of plasma interaction in the vicinity of poloidal gaps between monoblocks in the high heat flux areas of the ITER outer vertical target. The main objective of the simulations is to assess the role of local electric fields which are accounted for in a related study using the ion orbit approach including only the Lorentz force (Gunn et al 2017 Nucl. Fusion 57 046025). Results of the PIC simulations demonstrate that even if in some cases the electric field plays a distinct role in determining the precise heat flux distribution, when heat diffusion into the bulk material is taken into account, the thermal responses calculated using the PIC or ion orbit approaches are very similar. This is a consequence of the small spatial scales over which the ion orbits distribute the power. The key result of this study is that the computationally much less intensive ion orbit approximation can be used with confidence in monoblock shaping design studies, thus validating the approach used in Gunn et al (2017 Nucl. Fusion 57 046025).

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.

Investigation of transient melting of tungsten by ELMs in ASDEX Upgrade

NASA Astrophysics Data System (ADS)

Krieger, K.; Sieglin, B.; Balden, M.; Coenen, J. W.; Göths, B.; Laggner, F.; de Marne, P.; Matthews, G. F.; Nille, D.; Rohde, V.; Dejarnac, R.; Faitsch, M.; Giannone, L.; Herrmann, A.; Horacek, J.; Komm, M.; Pitts, R. A.; Ratynskaia, S.; Thoren, E.; Tolias, P.; ASDEX-Upgrade Team; EUROfusion MST1 Team

2017-12-01

Repetitive melting of tungsten by power transients originating from edge localized modes (ELMs) has been studied in the tokamak experiment ASDEX Upgrade. Tungsten samples were exposed to H-mode discharges at the outer divertor target plate using the Divertor Manipulator II system. The exposed sample was designed with an elevated sloped surface inclined against the incident magnetic field to increase the projected parallel power flux to a level were transient melting by ELMs would occur. Sample exposure was controlled by moving the outer strike point to the sample location. As extension to previous melt studies in the new experiment both the current flow from the sample to vessel potential and the local surface temperature were measured with sufficient time resolution to resolve individual ELMs. The experiment provided for the first time a direct link of current flow and surface temperature during transient ELM events. This allows to further constrain the MEMOS melt motion code predictions and to improve the validation of its underlying model assumptions. Post exposure ex situ analysis of the retrieved samples confirms the decreased melt motion observed at shallower magnetic field line to surface angles compared to that at leading edges exposed to the parallel power flux.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lomanowski, B. A., E-mail: b.a.lomanowski@durham.ac.uk; Sharples, R. M.; Meigs, A. G.

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.

Alternative divertor target concepts for next step fusion devices

NASA Astrophysics Data System (ADS)

Mazul, I. V.

2016-12-01

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

Simulations of particle and heat fluxes in an ELMy H-mode discharge on EAST using BOUT++ code

NASA Astrophysics Data System (ADS)

Wu, Y. B.; Xia, T. Y.; Zhong, F. C.; Zheng, Z.; Liu, J. B.; team3, EAST

2018-05-01

In order to study the distribution and evolution of the transient particle and heat fluxes during edge-localized mode (ELM) bursts on the Experimental Advanced Superconducting Tokamak (EAST), the BOUT++ six-field two-fluid model is used to simulate the pedestal collapse. The profiles from the EAST H-mode discharge #56129 are used as the initial conditions. Linear analysis shows that the resistive ballooning mode and drift-Alfven wave are two dominant instabilities for the equilibrium, and play important roles in driving ELMs. The evolution of the density profile and the growing process of the heat flux at divertor targets during the burst of ELMs are reproduced. The time evolution of the poloidal structures of T e is well simulated, and the dominant mode in each stage of the ELM crash process is found. The studies show that during the nonlinear phase, the dominant mode is 5, and it changes to 0 when the nonlinear phase goes to saturation after the ELM crash. The time evolution of the radial electron heat flux, ion heat flux, and particle density flux at the outer midplane (OMP) are obtained, and the corresponding transport coefficients D r, χ ir, and χ er reach maximum around 0.3 ∼ 0.5 m2 s‑1 at ΨN = 0.9. The heat fluxes at outer target plates are several times larger than that at inner target plates, which is consistent with the experimental observations. The simulated profiles of ion saturation current density (j s) at the lower outboard (LO) divertor target are compared to those of experiments by Langmuir probes. The profiles near the strike point are similar, and the peak values of j s from simulation are very close to the measurements.

Nonlinear fluid simulation of particle and heat fluxes during burst of ELMs on DIII-D with BOUT++ code [Fluid Simulation of Particle and Heat Fluxes during Burst of ELMs on DIID with BOUT++ code

DOE PAGES

Xia, T. Y.; Xu, X. Q.

2015-09-01

In order to study the distribution and evolution of the transient particle and heat fluxes during edge-localized mode (ELM) bursts, a BOUT++ six-field two-fluid model based on the Braginskii equations with non-ideal physics effects is used to simulate pedestal collapse in divertor geometry. We used the profiles from the DIII-D H-mode discharge #144382 with fast target heat flux measurements as the initial conditions for the simulations. Moreover, a flux-limited parallel thermal conduction is used with three values of the flux-limiting coefficientmore » $${{\\alpha}_{j}}$$ , free streaming model with $${{\\alpha}_{j}}=1$$ , sheath-limit with $${{\\alpha}_{j}}=0.05$$ , and one value in between. The studies show that a 20 times increase in $${{\\alpha}_{j}}$$ leads to ~6 times increase in the heat flux amplitude to both the inner and outer targets, and the widths of the fluxes are also expanded. The sheath-limit model of flux-limiting coefficient is found to be the most appropriate one, which shows ELM sizes close to the measurements. The evolution of the density profile during the burst of ELMs of DIII-D discharge #144382 is simulated, and the collapse in width and depth of $${{n}_{\\text{e}}}$$ are reproduced at different time steps. The growing process of the profiles for the heat flux at divertor targets during the burst of ELMs measured by IRTV (infrared television) is also reproduced by this model. The widths of heat fluxes towards targets are a little narrower, and the peak amplitudes are twice the measurements possibly due to the lack of a model of divertor radiation which can effectively reduce the heat fluxes. The magnetic flutter combined with parallel thermal conduction is found to be able to increase the total heat loss by around 33% since the magnetic flutter terms provide the additional conductive heat transport in the radial direction. Finally, the heat flux profile at both the inner and outer targets is obviously broadened by magnetic flutter. The lobe structures near the X-point at LFS are both broadened and elongated due to the magnetic flutter.« less

Modeling study of radiation characteristics with different impurity species seeding in EAST

NASA Astrophysics Data System (ADS)

Liu, X. J.; Deng, G. Z.; Wang, L.; Liu, S. C.; Zhang, L.; Li, G. Q.; Gao, X.

2017-12-01

A critical issue for EAST and future tokamak machines such as ITER and China Fusion Engineering Testing Reactor is the handling of excessive heat load on the divertor target plates. As an effective means of actively reducing and controlling the power fluxes to the target plates, localized impurity (N, Ne, and Ar) gas puffing from the lower dome is investigated by using SOLPS5.0 for an L-mode discharge on EAST with double null configuration. The radiative efficiency and distribution of different impurities are compared. The effect of N, Ne, and Ar seeding on target power load, the power entering into scrape-off layer (SOL), Psep, and their concentration in SOL along the poloidal length and edge effective ion charge number (Zeff) which are closely related to core plasma performance are presented. The simulation results indicate that N, Ne, and Ar seeding can effectively reduce the peak heat load and electron temperature at divertor targets similarly. N seeding can reach the highest radiative loss fraction and both N and Ar strongly radiate power in the divertor region, while the radiative power inside the separatrix for Ar seeding is also significant. Ne radiates power mainly around the separatrix and X-point. Ne and Ar impurities' puffing results in a faster decrease of Psep than N seeding case; the reduction of Psep can eventually degrade the core performance of fusion plasma. Additionally, seeding with Ne has a totally larger concentration at the outer midplane and edge Zeff than those in N and Ar seeding cases; it suggests that N and Ar impurities are more acceptable than Ne in terms of fuel dilution for this discharge.

OEDGE Modeling of Collector Probe measurements in L-mode from the DIII-D tungsten ring campaign

NASA Astrophysics Data System (ADS)

Elder, J. D.; Stangeby, P. C.; Unterberg, Z.; Donovan, D.; Wampler, W. R.; Watkins, J.; Abrams, T.; McLean, A. G.

2017-10-01

During the tungsten ring campaign on DIII-D, a collector probe system with multiple diameter, dual-facing collector rods was inserted into the far scrape off layer (SOL) near the outer midplane to measure the plasma tungsten content. For most probes more tungsten was observed on the side connected along field lines to the inner divertor, with the larger probes showing largest divertor-facing asymmetries The OEDGE code is used to model the tungsten erosion, transport and deposition. It has been enhanced with (i) a peripheral particle transport and deposition model to record the impurity content in the peripheral region outside the regular mesh, and (ii) a collector probe model. The OEDGE results approximately reproduce both the divertor-facing asymmetries and the radial decay of each collector probe profile. The effect of changing impurity transport assumptions and wall location are examined. The measured divertor-facing asymmetries imply a higher tungsten density in the plasma upstream of the probe; this was expected theoretically from the effect of the parallel ion temperature gradient force driving upstream transport of tungsten from the outer divertor and was also found in the code analysis. Work supported by the US Department of Energy under DE-FC02-04ER54698, DE-NA0003525, DE-AC05-00OR22725, and DE-AC52-07NA27344.

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 P inj. Strike point sweeping allowed 2D divertor characterization using DTS optimized to measure T e 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. Measurementsmore » of n e, T e and p e 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

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.

Status of National Spherical Torus Experiment Liquid Lithium Divertor

NASA Astrophysics Data System (ADS)

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

2009-11-01

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

Modifications of W and Mo leading edges under plasma loads in DIII-D divertor

NASA Astrophysics Data System (ADS)

Rudakov, D. L.; Bykov, I.; Moyer, R. A.; Abrams, T.; Chrobak, C. P.; Guo, H. Y.; Stahl, B.; Thomas, D. M.; Barton, J. L.; Nygren, R. E.; Watkins, J. G.; Lasnier, C. J.; Litnovsky, Andrey; Stangeby, P. C.; Unterberg, E. A.

2017-10-01

Cracking and melting of W and Mo leading edges were observed in the lower divertor of DIII-D during experiments with intentionally misaligned W monoblocks (MBs) and in the course of the Metal Rings Campaign involving W-coated Mo tile inserts (TIs). MBs were exposed near the attached outer strike point during deuterium and helium L- and H-mode discharges using DiMES. Two of the MBs were misaligned by 0.3 mm and 1 mm, forming leading edges. Particulate ejection from a 1 mm leading edge was observed during the exposure, and evidence of melting and cracking was found post mortem. Two toroidal rings of TIs were installed in the lower outer divertor, the inner one at the floor and the outer one at the shelf. The floor TIs bowed during plasma exposure forming leading edges up to 1.2 mm high; about 40% of these edges experienced melting. Re-solidified melt layers up to 1 mm thick were observed, their shape being consistent with motion in the jx B direction with j driven by electron emission. Work supported by US DOE under DE-FC02-04ER54698, DE-FG02-07ER54917, DE-AC04-94AL85000, DE-AC52-07NA27344, and DE-AC05-00OR22725.

The effects of particle recycling on the divertor plasma: A particle-in-cell with Monte Carlo collision simulation

NASA Astrophysics Data System (ADS)

Chang, Mingyu; Sang, Chaofeng; Sun, Zhenyue; Hu, Wanpeng; Wang, Dezhen

2018-05-01

A Particle-In-Cell (PIC) with Monte Carlo Collision (MCC) model is applied to study the effects of particle recycling on divertor plasma in the present work. The simulation domain is the scrape-off layer of the tokamak in one-dimension along the magnetic field line. At the divertor plate, the reflected deuterium atoms (D) and thermally released deuterium molecules (D2) are considered. The collisions between the plasma particles (e and D+) and recycled neutral particles (D and D2) are described by the MCC method. It is found that the recycled neutral particles have a great impact on divertor plasma. The effects of different collisions on the plasma are simulated and discussed. Moreover, the impacts of target materials on the plasma are simulated by comparing the divertor with Carbon (C) and Tungsten (W) targets. The simulation results show that the energy and momentum losses of the C target are larger than those of the W target in the divertor region even without considering the impurity particles, whereas the W target has a more remarkable influence on the core plasma.

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.

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

Numerical exploration of non-axisymmetric divertor closure in the small angle slot (SAS) divertor at DIII-D

NASA Astrophysics Data System (ADS)

Frerichs, H.; Schmitz, O.; Covele, B.; Feng, Y.; Guo, H. Y.; Hill, D.

2018-05-01

Numerical simulations of toroidal asymmetries in a tightly baffled small angle slot (SAS) divertor on the DIII-D tokamak show that toroidal asymmetries in divertor closure result in (non-axisymmetric) local onset of detachment within a density window of 10-15% on top of the nominal threshold separatrix density. The SAS divertor is explored at DIII-D for improving access to cold, dissipative/detached divertor conditions. The narrow width of the slot divertor coupled with a small magnetic field line-to-target angle facilitates the buildup of neutral density, thereby increasing radiative and neutrals-related (atoms and molecules) losses in the divertor. Small changes in the strike point location can be expected to have a large impact on divertor conditions. The combination of misaligned slot structure and non-axisymmetric perturbations to the magnetic field configuration causes the strike point to move along the divertor target plate, possibly leaving the divertor slot at some locations. The latter extreme case essentially introduces an opening in the divertor slot from where recycling neutrals can easily escape, and thereby degrade the performance of the slot divertor. Such a strike point dislocation is approximated by a finite gap in the divertor baffle for which 3D edge plasma and neutral gas simulations are performed with the EMC3-EIRENE code.

Detachment experiments in new DIII-D upper divertor

NASA Astrophysics Data System (ADS)

Moser, A. L.; Leonard, A. W.; Groebner, R. J.; Guo, H.; Wang, H.; Watkins, J. G.; McLean, A. G.; Fenstermacher, M. E.; Shafer, M. W.; Briesemeister, A. R.; Hinson, E. T.

2017-10-01

Installation of the Small Angle Slot (SAS) in the upper divertor of DIII-D enables new studies of the effect of target and baffle geometry on divertor detachment. This structure provides a more-closed upper divertor as well as the SAS divertor itself. Initial SAS experiment results indicate that divertor detachment occurs at a lower line-averaged density than in the more-open, lower single null divertor configurations on DIII-D. In contrast, the increased divertor closure of the new installation did not reduce the upstream density required for detachment beyond that achieved with the previous upper divertor structure. Particle pumping in the upper divertor structure is found to produce a 10 % reduction in the pedestal density required for detachment compared to the case with no pumping. Comparisons focus on both the onset of detachment (measured by in-target Langmuir probes) as a function of upstream density, as well as the effect of the new divertor configurations on pedestal density profiles. Work supported by US DOE under DE-FC02-04ER54698, DE-AC05-00OR22725, DE-AC04-94AL85000, DE-AC52-07NA27344, and DE-SC00013911.

Reduction of Net Erosion of High-Z Divertor Surface by Local Redeposition in DIII-D

NASA Astrophysics Data System (ADS)

Stangeby, P. C.

2012-10-01

Utilizing the unique capability to expose material samples to well characterized diverted plasmas, recent DIII-D measurements have confirmed theoretical expectations of the relative net and gross erosion rates of molybdenum in the divertor region. Knowledge of these erosion rates is important for predicting first wall lifetime in future fusion devices. Theory suggests that the net erosion rate will be much less than gross erosion due to prompt local deposition of eroded ions by gyro-orbit motion, the strong E-field toward the target and friction with the fast plasma flow toward the target. However, experimental evidence to date has been contradictory. The results here, which are the most definitive to date, are consistent with the basic theoretical predictions. The net and gross erosion rates were measured utilizing 1-cm and 1-mm diameter Mo samples that are mounted on the DIII-D Divertor Material Evaluation System (DiMES) system and simultaneously exposed near the attached outer strike point of an L-mode plasma for 4 s. Due to the spatial extent of the re-deposition, the larger sample gives the net erosion while the smaller sample is indicative of the gross erosion. Post-mortem ion beam analysis (RBS) of the larger sample, indicates a 2.9 nm film thickness reduction (or 0.72 nm/s net erosion rate). Similar analysis of the smaller sample yields a 1.3 nm/s gross erosion rate, consistent with spectroscopic measurements of Mo I emission. The net to gross erosion ratio of 0.56 is consistent with calculations using a modeling package including REDEP/WBS and OEDGE codes. Using as input the measured plasma density and temperature profiles from divertor Langmuir probes, these codes estimate a net to gross erosion ratio of 0.46. Details of the modeling and implications for future devices will be discussed.

Experiments on transient melting of tungsten by ELMs in ASDEX Upgrade

NASA Astrophysics Data System (ADS)

Krieger, K.; Balden, M.; Coenen, J. W.; Laggner, F.; Matthews, G. F.; Nille, D.; Rohde, V.; Sieglin, B.; Giannone, L.; Göths, B.; Herrmann, A.; de Marne, P.; Pitts, R. A.; Potzel, S.; Vondracek, P.; ASDEX-Upgrade Team; EUROfusion MST1 Team

2018-02-01

Repetitive melting of tungsten by power transients originating from edge localized modes (ELMs) has been studied in ASDEX Upgrade. Tungsten samples were exposed to H-mode discharges at the outer divertor target plate using the divertor manipulator II (DIM-II) system (Herrmann et al 2015 Fusion Eng. Des. 98-9 1496-9). Designed as near replicas of the geometries used also in separate experiments on the JET tokamak (Coenen et al 2015 J. Nucl. Mater. 463 78-84 Coenen et al 2015 Nucl. Fusion 55 023010; Matthews et al 2016 Phys. Scr. T167 7), the samples featured a misaligned leading edge and a sloped ridge respectively. Both structures protrude above the default target plate surface thus receiving an increased fraction of the parallel power flux. Transient melting by ELMs was induced by moving the outer strike point to the sample location. The temporal evolution of the measured current flow from the samples to vessel potential confirmed transient melting. Current magnitude and dependency from surface temperature provided strong evidence for thermionic electron emission as main origin of the replacement current driving the melt motion. The different melt patterns observed after exposures at the two sample geometries support the thermionic electron emission model used in the MEMOS melt motion code, which assumes a strong decrease of the thermionic net current at shallow magnetic field to surface angles (Pitts et al 2017 Nucl. Mater. Energy 12 60-74). Post exposure ex situ analysis of the retrieved samples show recrystallization of tungsten at the exposed surface areas to a depth of up to several mm. The melt layer transport to less exposed surface areas leads to ratcheting pile up of re-solidified debris with zonal growth extending from the already enlarged grains at the surface.

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

Structural investigation of re-deposited layers in JET

NASA Astrophysics Data System (ADS)

Likonen, J.; Vainonen-Ahlgren, E.; Khriachtchev, L.; Coad, J. P.; Rubel, M.; Renvall, T.; Arstila, K.; Hole, D. E.; Contributors to the EFDA-JET Work-programme

2008-07-01

JET Mk-II Gas Box divertor tiles exposed in 1998-2001 have been analysed with various ion beam techniques, secondary ion mass spectrometry (SIMS) and Raman spectroscopy. Inner divertor wall tiles removed in 2001 were covered with a duplex film. The inner layer was very rich in metallic impurities, with Be/C ˜ 1 and H-isotopes only present at low concentrations. The outer layer contained higher concentrations of D than normal for plasma-facing surfaces in JET (D/C ˜ 0.4), and Be/C ˜ 0.14. Raman and SIMS analyses show that the deposited films on inner divertor tiles are hydrogenated amorphous carbon with low sp 3 fractions. The deposits have polymeric structure and low density. Both Raman scattering and SIMS indicate that films on inner divertor wall Tiles 1 and 3, and on floor Tile 4 have some differences in the chemical structure of the deposited films

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soukhanovskii, V. A., E-mail: vlad@llnl.gov; 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 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 T{sub e}-dependent signal within a characteristic divertor detachment equilibration time of ∼10–15 ms is expected.« less

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 T e estimated spectroscopically in real time will be used as a control parameter. A vacuum ultraviolet spectrometer McPhersonmore » 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 e-dependent signal within a characteristic divertor detachment equilibration time of ~10–15 ms is expected.« less

Scrape off layer modelling studies for SST-I

NASA Astrophysics Data System (ADS)

Warrier, M.; Jaishankar, S.; Deshpande, S.; Coster, D.; Schneider, R.; Chaturvedi, S.; Srinivasan, R.; Braams, B. J.; SST Team

SOL modelling results for SST-1 (SST Team, Proceedings of the 16th IEEE/NPSS Symposium on Fusion Engineering, Champaign, IL, vol. II, 1995, p. 481) show a sheath limited flow regime. This is due to the low edge densities required by lower hybrid current drive (LHCD), coupled with high power input per unit volume. Coupled plasma-neutral transport studies using B2-Eirene [R. Schneider et al., J. Nucl. Mater. 196-198 (1992) 810] show significantly high charge exchange losses and radiated power from the core. It also shows that the heat flux to the inner divertor is higher than that to the outer divertor due to thinner inner SOL widths. The Monte-Carlo neutral transport code DEGAS [D. Heifitz et al., J. Comput. Phys. 46 (1982) 309] was used to optimise the baffle plate geometry and it was seen that a configuration where the baffle plate shields the main plasma from the divertor strike point results in reduced backflow of neutrals. The divertor erosion code DIVER (M. Warrier et al., SST Divertor Modelling Report, 1996-1997) was used to predict a steady state operating temperature for the SST divertor plate lying in the range 750-1000°C for which the erosion will be minimum.

Numerical exploration of non-axisymmetric divertor closure in the small angle slot (SAS) divertor at DIII-D

DOE PAGES

Frerichs, H.; Schmitz, O.; Covele, B.; ...

2018-02-28

Numerical simulations of toroidal asymmetries in a tightly baffled small angle slot (SAS) divertor on the DIII-D tokamak show that toroidal asymmetries in divertor closure result in (non-axisymmetric) local onset of detachment within a density window of 10-15% on top of the nominal threshold separatrix density. The SAS divertor is explored at DIII-D for improving access to cold, dissipative/detached divertor conditions. The narrow width of the slot divertor coupled with a small magnetic field line-to-target angle facilitates the buildup of neutral density, thereby increasing radiative and neutrals-related (atoms and molecules) losses in the divertor. Therefore, small changes in the strikemore » point location can be expected to have a large impact on diverter conditions. The combination of misaligned slot structure and non-axisymmetric perturbations to the magnetic field configuration causes the strike point to move along the divertor target plate, possibly leaving the diverter slot at some locations. The latter extreme case essentially introduces an opening in the divertor slot from where recycling neutrals can easily escape, and thereby degrade the performance of the slot divertor. Such a strike point dislocation is approximated by a finite gap in the divertor baffle for which three dimensional edge plasma and neutral gas simulations are performed with the EMC3-EIRENE code.« less

Numerical exploration of non-axisymmetric divertor closure in the small angle slot (SAS) divertor at DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Frerichs, H.; Schmitz, O.; Covele, B.

Numerical simulations of toroidal asymmetries in a tightly baffled small angle slot (SAS) divertor on the DIII-D tokamak show that toroidal asymmetries in divertor closure result in (non-axisymmetric) local onset of detachment within a density window of 10-15% on top of the nominal threshold separatrix density. The SAS divertor is explored at DIII-D for improving access to cold, dissipative/detached divertor conditions. The narrow width of the slot divertor coupled with a small magnetic field line-to-target angle facilitates the buildup of neutral density, thereby increasing radiative and neutrals-related (atoms and molecules) losses in the divertor. Therefore, small changes in the strikemore » point location can be expected to have a large impact on diverter conditions. The combination of misaligned slot structure and non-axisymmetric perturbations to the magnetic field configuration causes the strike point to move along the divertor target plate, possibly leaving the diverter slot at some locations. The latter extreme case essentially introduces an opening in the divertor slot from where recycling neutrals can easily escape, and thereby degrade the performance of the slot divertor. Such a strike point dislocation is approximated by a finite gap in the divertor baffle for which three dimensional edge plasma and neutral gas simulations are performed with the EMC3-EIRENE code.« less

Divertor target shape optimization in realistic edge plasma geometry

NASA Astrophysics Data System (ADS)

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

2014-07-01

Tokamak divertor design for next-step fusion reactors heavily relies on numerical simulations of the plasma edge. Currently, the design process is mainly done in a forward approach, where the designer is strongly guided by his experience and physical intuition in proposing divertor shapes, which are then thoroughly assessed by numerical computations. On the other hand, automated design methods based on optimization have proven very successful in the related field of aerodynamic design. By recasting design objectives and constraints into the framework of a mathematical optimization problem, efficient forward-adjoint based algorithms can be used to automatically compute the divertor shape which performs the best with respect to the selected edge plasma model and design criteria. In the past years, we have extended these methods to automated divertor target shape design, using somewhat simplified edge plasma models and geometries. In this paper, we build on and extend previous work to apply these shape optimization methods for the first time in more realistic, single null edge plasma and divertor geometry, as commonly used in current divertor design studies. In a case study with JET-like parameters, we show that the so-called one-shot method is very effective is solving divertor target design problems. Furthermore, by detailed shape sensitivity analysis we demonstrate that the development of the method already at the present state provides physically plausible trends, allowing to achieve a divertor design with an almost perfectly uniform power load for our particular choice of edge plasma model and design criteria.

Basic physical processes and reduced models for plasma detachment

NASA Astrophysics Data System (ADS)

Stangeby, P. C.

2018-04-01

The divertor of a tokamak reactor will have to satisfy a number of critical constraints, the first of which is that the divertor targets not fail due to excessive heating or sputter-erosion. This paramount constraint of target survival defines the operating window for the principal plasma properties at the divertor target, the density n t and temperature, T t. In particular T et < 10 eV is shown to be required. Code and experimental studies show that the pressure–momentum loss by the plasma that occurs along flux tubes in the edge, between the divertor entrance and target, (i) correlates strongly with T et, and (ii) begins to increase as T et falls below 10 eV, becoming very strong by 1 eV. The transition between the high-recycling regime and the detached divertor regime has therefore been defined here to occur when T et < 10 eV. Simple analytic models are developed (i) to relate (T t, n t) to the controlling conditions ‘upstream’ e.g. at the divertor entrance, and (ii) in turn to relate (T t, n t) to other important divertor quantities including (a) the required level of radiative cooling in the divertor, and (b) the ion flux to the target in the presence of volumetric loss of particles, momentum and power in the divertor. The 2 Point Model, 2PM, is a widely used analytic model for relating (T t, n t) to the controlling upstream conditions. The 2PM is derived here for various levels of complexity regarding the effects included. Analytic models of divertor detachment provide valuable insight and useful approximations, but more complete modeling requires the use of edge codes such as EDGE2D, SOLPS, SONIC, UEDGE, etc. Edge codes have grown to become quite sophisticated and now constitute, in effect, ‘code-experiments’ that—just as for actual experiments—can benefit from interpretation in terms of simple conceptual frameworks. 2 Point Model Formatting, 2PMF, of edge code output can provide such a conceptual framework. Methods of applying 2PMF are illustrated here with some examples.

Dynamic divertor control using resonant mixed toroidal harmonic magnetic fields during ELM suppression in DIII-D

NASA Astrophysics Data System (ADS)

Jia, M.; Sun, Y.; Paz-Soldan, C.; Nazikian, R.; Gu, S.; Liu, Y. Q.; Abrams, T.; Bykov, I.; Cui, L.; Evans, T.; Garofalo, A.; Guo, W.; Gong, X.; Lasnier, C.; Logan, N. C.; Makowski, M.; Orlov, D.; Wang, H. H.

2018-05-01

Experiments using Resonant Magnetic Perturbations (RMPs), with a rotating n = 2 toroidal harmonic combined with a stationary n = 3 toroidal harmonic, have validated predictions that divertor heat and particle flux can be dynamically controlled while maintaining Edge Localized Mode (ELM) suppression in the DIII-D tokamak. Here, n is the toroidal mode number. ELM suppression over one full cycle of a rotating n = 2 RMP that was mixed with a static n = 3 RMP field has been achieved. Prominent heat flux splitting on the outer divertor has been observed during ELM suppression by RMPs in low collisionality regime in DIII-D. Strong changes in the three dimensional heat and particle flux footprint in the divertor were observed during the application of the mixed toroidal harmonic magnetic perturbations. These results agree well with modeling of the edge magnetic field structure using the TOP2D code, which takes into account the plasma response from the MARS-F code. These results expand the potential effectiveness of the RMP ELM suppression technique for the simultaneous control of divertor heat and particle load required in ITER.

Plasma-wall interactions in ITER

NASA Astrophysics Data System (ADS)

Parker, R.; Janeschitz, G.; Pacher, H. D.; Post, D.; Chiocchio, S.; Federici, G.; Ladd, P.; Iter Joint Central Team; Home Teams

1997-02-01

This paper reviews the status of the design of the divertor and first-wall/shield, the main in-vessel components for ITER. Under nominal ignited conditions, 300 MW of alpha power will be produced and must be removed from the divertor and first-wall. Additional power from auxiliary sources up to the level of 100 MW must also be removed in the case of driven burns. In the ignited case, about 100 MW will be radiated to the first wall as bremsstrahlung. Allowing the remaining power to be conducted to the divertor target plates would result in excessive heat fluxes. The power handling strategy is to radiate an additional 100-150 MW in the SOL and the divertor channel via a combination of radiation from hydrogen, and intrinsic and seeded impurities. Vertical targets have been adopted for the baseline divertor configuration. This geometry promotes partial detachment, as found in present experiments and in the results of modelling runs for ITER conditions, and power densities on the target plates can be ≤ 5 MW/ m2. Such regimes promote relatively high pressure (> 1 Pa) in the divertor and even with a low helium enrichment factor of 0.2, the required pumping speed to pump helium is ≤ 50 m3/ s. An important physics question is the quality of core confinement in these attractive divertor regimes. In addition to power and particle handling issues, the effects of disruptions play a major role in the design and performance of in-vessel components. Both centered disruptions and VDE's produce stresses in the first-wall/shield modules, backplate and the divertor wings and cassettes that are near or even somewhat in excess of allowables for normal operation. Also plasma-wall contact from disruptions, including at the divertor target, together with material properties are major factors determining component lifetime. Considering the potential for impurity contamination and minimizing tritium inventory as well as thermomechanical performance, the present material selection calls for carbon divertor targets near the strike point, tungsten on the rest of the target and on the baffle where the charge-exchange flux could be high, and beryllium elsewhere. All three materials and relevant joining techniques are being developed in the R&D program and the final selection for the first assembly will be made at the end of the EDA.

ICRF-Induced Changes in Floating Potential and Ion Saturation Current in the EAST Divertor

NASA Astrophysics Data System (ADS)

Perkins, Rory; Hosea, Joel; Taylor, Gary; Bertelli, Nicola; Kramer, Gerrit; Qin, Chengming; Wang, Liang; Yang, Jichan; Zhang, Xinjun

2017-10-01

Injection of waves in the ion cyclotron range of frequencies (ICRF) into a tokamak can potentially raise the plasma potential via RF rectification. Probes are affected both by changes in plasma potential and also by RF-averaging of the probe characteristic, with the latter tending to drop the floating potential. We present the effect of ICRF heating on divertor Langmuir probes in the EAST experiment. Over a scan of the outer gap, probes connected to the antennas have increases in floating potential with ICRF, but probes in between the outer-vessel strike point and flux surface tangent to the antenna have decreased floating potential. This behaviour is investigated using field-line mapping. Preliminary results show that mdiplane gas puffing can suppress the strong influence of ICRF on the probes' floating potential.

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.

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.

Double-null divertor configuration discharge and disruptive heat flux simulation using TSC on EAST

NASA Astrophysics Data System (ADS)

Bo, SHI; Jinhong, YANG; Cheng, YANG; Desheng, CHENG; Hui, WANG; Hui, ZHANG; Haifei, DENG; Junli, QI; Xianzu, GONG; Weihua, WANG

2018-07-01

The tokamak simulation code (TSC) is employed to simulate the complete evolution of a disruptive discharge in the experimental advanced superconducting tokamak. The multiplication factor of the anomalous transport coefficient was adjusted to model the major disruptive discharge with double-null divertor configuration based on shot 61 916. The real-time feed-back control system for the plasma displacement was employed. Modeling results of the evolution of the poloidal field coil currents, the plasma current, the major radius, the plasma configuration all show agreement with experimental measurements. Results from the simulation show that during disruption, heat flux about 8 MW m‑2 flows to the upper divertor target plate and about 6 MW m‑2 flows to the lower divertor target plate. Computations predict that different amounts of heat fluxes on the divertor target plate could result by adjusting the multiplication factor of the anomalous transport coefficient. This shows that TSC has high flexibility and predictability.

Advanced simulation of mixed-material erosion/evolution and application to low and high-Z containing plasma facing components

NASA Astrophysics Data System (ADS)

Brooks, J. N.; Hassanein, A.; Sizyuk, T.

2013-07-01

Plasma interactions with mixed-material surfaces are being analyzed using advanced modeling of time-dependent surface evolution/erosion. Simulations use the REDEP/WBC erosion/redeposition code package coupled to the HEIGHTS package ITMC-DYN mixed-material formation/response code, with plasma parameter input from codes and data. We report here on analysis for a DIII-D Mo/C containing tokamak divertor. A DIII-D/DiMES probe experiment simulation predicts that sputtered molybdenum from a 1 cm diameter central spot quickly saturates (˜4 s) in the 5 cm diameter surrounding carbon probe surface, with subsequent re-sputtering and transport to off-probe divertor regions, and with high (˜50%) redeposition on the Mo spot. Predicted Mo content in the carbon agrees well with post-exposure probe data. We discuss implications and mixed-material analysis issues for Be/W mixing at the ITER outer divertor, and Li, C, Mo mixing at an NSTX divertor.

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.

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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

Divertor Heat Flux Reduction and Detachment in the National Spherical Torus eXperiment.

NASA Astrophysics Data System (ADS)

Soukhanovskii, Vsevolod

2007-11-01

Steady-state handling of the heat flux is a critical divertor issue for both the International Thermonuclear Experimental Reactor and spherical torus (ST) devices. Because of an inherently compact divertor, it was thought that ST-based devices might not be able to fully utilize radiative and dissipative divertor techniques based on induced power and momentum loss. However, initial experiments conducted in the National Spherical Torus Experiment in an open geometry horizontal carbon plate divertor using 0.8 MA 2-6 MW NBI-heated lower single null H-mode plasmas at the lower end of elongations κ=1.8-2.4 and triangularities δ=0.45-0.75 demonstrated that high divertor peak heat fluxes, up to 6-10 MW/ m^2, could be reduced by 50-75% using a high-recycling radiative divertor regime with D2 injection. Furthermore, similar reduction was obtained with a partially detached divertor (PDD) at high D2 injection rates, however, it was accompanied by an X-point MARFE that quickly led to confinement degradation. Another approach takes advantage of the ST relation between strong shaping and high performance, and utilizes the poloidal magnetic flux expansion in the divertor region. Up to 60 % reduction in divertor peak heat flux was achieved at similar levels of scrape-off layer power by varying plasma shaping and thereby increasing the outer strike point (OSP) poloidal flux expansion from 4-6 to 18-22. In recent experiments conducted in highly-shaped 1.0-1.2 MA 6 MW NBI heated H-mode plasmas with divertor D2 injection at rates up to 10^22 s-1, a PDD regime with OSP peak heat flux 0.5-1.5 MW/m^2 was obtained without noticeable confinement degradation. Calculations based on a two point scrape-off layer model with parameterized power and momentum losses show that the short parallel connection length at the OSP sets the upper limit on the radiative exhaust channel, and both the impurity radiation and large momentum sink achievable only at high divertor neutral pressures are required for detachment.

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

NASA Astrophysics Data System (ADS)

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

2017-12-01

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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

Modeling of combined effects of divertor closure and advanced magnetic configuration on detachment in DIII-D by SOLPS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Si, Hang; Guo, Houyang Y.; Covele, Brent

One of the major challenges facing the design and operation of next-step high-power steady-state fusion devices is to develop a divertor solution for handling power exhaust, while ensuring acceptable divertor target plate erosion, which necessitates access to divertor detachment at relative low main plasma densities compatible with current drive and high plasma confinement. Detailed modeling with SOLPS is carried out to examine the effect of divertor closure on detachment with the normal single null divertor (SD) configuration, as well as one of the advanced divertor configurations, such as x-divertor (XD) respectively. The SOLPS modeling for a high confinement plasma in DIII-D finds that increasing divertor closure with SD reduces the upstream separatrix density at the onset of detachment frommore » $$1.18\\times {{10}^{19}}\\,{{{\\rm m}}^{-3}}$$ to $$0.88\\times {{10}^{19}}\\,{{{\\rm m}}^{-3}}$$. Furthermore, coupling the divertor closure with XD further promotes the onset of divertor detachment at a still lower upstream separatrix density, down to the value of $$0.67\\times {{10}^{19}}\\,{{{\\rm m}}^{-3}}$$, thus, showing that divertor closure and advanced magnetic configuration can work synergistically to facilitate divertor detachment.« less

Modeling of combined effects of divertor closure and advanced magnetic configuration on detachment in DIII-D by SOLPS

DOE PAGES

Si, Hang; Guo, Houyang Y.; Covele, Brent; ...

2018-04-04

One of the major challenges facing the design and operation of next-step high-power steady-state fusion devices is to develop a divertor solution for handling power exhaust, while ensuring acceptable divertor target plate erosion, which necessitates access to divertor detachment at relative low main plasma densities compatible with current drive and high plasma confinement. Detailed modeling with SOLPS is carried out to examine the effect of divertor closure on detachment with the normal single null divertor (SD) configuration, as well as one of the advanced divertor configurations, such as x-divertor (XD) respectively. The SOLPS modeling for a high confinement plasma in DIII-D finds that increasing divertor closure with SD reduces the upstream separatrix density at the onset of detachment frommore » $$1.18\\times {{10}^{19}}\\,{{{\\rm m}}^{-3}}$$ to $$0.88\\times {{10}^{19}}\\,{{{\\rm m}}^{-3}}$$. Furthermore, coupling the divertor closure with XD further promotes the onset of divertor detachment at a still lower upstream separatrix density, down to the value of $$0.67\\times {{10}^{19}}\\,{{{\\rm m}}^{-3}}$$, thus, showing that divertor closure and advanced magnetic configuration can work synergistically to facilitate divertor detachment.« less

Modeling of combined effects of divertor closure and advanced magnetic configuration on detachment in DIII-D by SOLPS

NASA Astrophysics Data System (ADS)

Si, H.; Guo, H. Y.; Covele, B.; Leonard, A. W.; Watkins, J. G.; Thomas, D.; Ding, R.

2018-05-01

One of the major challenges facing the design and operation of next-step high-power steady-state fusion devices is to develop a divertor solution for handling power exhaust, while ensuring acceptable divertor target plate erosion, which necessitates access to divertor detachment at relative low main plasma densities compatible with current drive and high plasma confinement. Detailed modeling with SOLPS is carried out to examine the effect of divertor closure on detachment with the normal single null divertor (SD) configuration, as well as one of the advanced divertor configurations, such as x-divertor (XD) respectively. The SOLPS modeling for a high confinement plasma in DIII-D finds that increasing divertor closure with SD reduces the upstream separatrix density at the onset of detachment from 1.18× {{10}19} {{m}-3} to 0.88× {{10}19} {{m}-3} . Moreover, coupling the divertor closure with XD further promotes the onset of divertor detachment at a still lower upstream separatrix density, down to the value of 0.67× {{10}19} {{m}-3} , thus, showing that divertor closure and advanced magnetic configuration can work synergistically to facilitate divertor detachment.

Implementation of the 3D edge plasma code EMC3-EIRENE on NSTX

DOE PAGES

Lore, J. D.; Canik, J. M.; Feng, Y.; ...

2012-05-09

The 3D edge transport code EMC3-EIRENE has been applied for the first time to the NSTX spherical tokamak. A new disconnected double null grid has been developed to allow the simulation of plasma where the radial separation of the inner and outer separatrix is less than characteristic widths (e.g. heat flux width) at the midplane. Modelling results are presented for both an axisymmetric case and a case where 3D magnetic field is applied in an n = 3 configuration. In the vacuum approximation, the perturbed field consists of a wide region of destroyed flux surfaces and helical lobes which aremore » a mixture of long and short connection length field lines formed by the separatrix manifolds. This structure is reflected in coupled 3D plasma fluid (EMC3) and kinetic neutral particle (EIRENE) simulations. The helical lobes extending inside of the unperturbed separatrix are filled in by hot plasma from the core. The intersection of the lobes with the divertor results in a striated flux footprint pattern on the target plates. As a result, profiles of divertor heat and particle fluxes are compared with experimental data, and possible sources of discrepancy are discussed.« less

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

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 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 D 2- ion D + elastic collisions and neutral-neutral collisions on the divertor plasma solution are assessed.« less

DOE FES FY2017 Joint Research Target Fourth Quarter Milestone Report for theNational Spherical Torus Experiment Upgrade.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soukhanovskii, V. A.

2017-09-13

A successful high-performance plasma operation with a radiative divertor has been demonstrated on many tokamak devices, however, significant uncertainty remains in accurately modeling detachment thresholds, and in how detachment depends on divertor geometry. Whereas it was originally planned to perform dedicated divertor experiments on the National Spherical Tokamak Upgrade to address critical detachment and divertor geometry questions for this milestone, the experiments were deferred due to technical difficulties. Instead, existing NSTX divertor data was summarized and re-analyzed where applicable, and additional simulations were performed.

The influence of plasma-surface interaction on the performance of tungsten at the ITER divertor vertical targets

NASA Astrophysics Data System (ADS)

De Temmerman, G.; Hirai, T.; Pitts, R. A.

2018-04-01

The tungsten (W) material in the high heat flux regions of the ITER divertor will be exposed to high fluxes of low-energy particles (e.g. H, D, T, He, Ne and/or N). Combined with long-pulse operations, this implies fluences well in excess of the highest values reached in today’s tokamak experiments. Shaping of the individual monoblock top surface and tilting of the vertical targets for leading-edge protection lead to an increased surface heat flux, and thus increased surface temperature and a reduced margin to remain below the temperature at which recrystallization and grain growth begin. Significant morphology changes are known to occur on W after exposure to high fluences of low-energy particles, be it H or He. An analysis of the formation conditions of these morphology changes is made in relation to the conditions expected at the vertical targets during different phases of operations. It is concluded that both H and He-related effects can occur in ITER. In particular, the case of He-induced nanostructure (also known as ‘fuzz’) is reviewed. Fuzz formation appears possible over a limited region of the outer vertical target, the inner target being generally a net Be deposition area. A simple analysis of the fuzz growth rate including the effect of edge-localized modes (ELMs) and the reduced thermal conductivity of fuzz shows that the fuzz thickness is likely to be limited by the occurrence of annealing during ELM-induced thermal excursions. Not only the morphology, but the material mechanical and thermal properties can be modified by plasma exposure. A review of the existing literature is made, but the existing data are insufficient to conclude quantitatively on the importance and extent of these effects for ITER. As a consequence of the high surface temperatures in ITER, W recrystallization is an important effect to consider, since it leads to a decrease in material strength. An approach is proposed here to develop an operational budget for the W material, i.e. the time the divertor material can be operated at a given temperature before a significant fraction of the material is recrystallized. In general, while it is clear that significant surface damage can occur during ITER operations, the tolerable level of damage in terms of plasma operations currently remains unknown.

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.

Deposition Profile Analysis from DIII-D Metal Rings Campaign Outer-Midplane Collector Probe Diagnostic and Utilization of Enriched Isotopic Tungsten Tracer Particles

NASA Astrophysics Data System (ADS)

Donovan, D. C.; Duran, J.; Zamperini, S.; Lee, S.; Unterberg, E. A.; Wampler, W. R.; Rudakov, D. L.; Elder, D.; Stangeby, P. C.; Abrams, T.

2017-10-01

The DIII-D Metal Rings Campaign used isotopically-enriched, W-coated divertor tiles coupled with dual-facing midplane collector probes (CPs) in the far Scrape-off Layer (SOL). Inductively Coupled Plasma Mass Spectroscopy (ICP-MS) results are presented characterizing the isotopic ratios of deposited W on the CPs and which give quantitative information on the transport of W from specific poloidal locations within the lower outer divertor region having different isotopically-marked tiles. Rutherford Backscattering Spectrometry (RBS) of these CPs has provided areal densities of elemental W content. These resultant W deposition profiles were compared with DIVIMP modelling of the far-SOL to better understand impurity transport in the edge plasma. CPs were exposed for 37 distinct operating configurations (L-mode/H-mode, forward/reverse Bt, strikepoint position). Radial decay lengths (RDL) between 5 and 50 mm were observed with consistently narrower RDL and higher peak W content on the side of the probes connected along field lines to the inner divertor, indicating a concentration of W in the upstream plasma. Correlations are discussed between peak W content, RDL, and isotopic profiles on the CPs over a wide range of conditions. Work supported by US DOE under DE-AC05-00OR22725, DE-FG02-07ER54917, DE-FC02-04ER54698, DE-NA0003525.

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.

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

PubMed

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.

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.

Characteristics of the Secondary Divertor on DIII-D

NASA Astrophysics Data System (ADS)

Watkins, J. G.; Lasnier, C. J.; Leonard, A. W.; Evans, T. E.; Pitts, R.; Stangeby, P. C.; Boedo, J. A.; Moyer, R. A.; Rudakov, D. L.

2009-11-01

In order to address a concern that the ITER secondary divertor strike plates may be insufficiently robust to handle the incident pulses of particles and energy from ELMs, we performed dedicated studies of the secondary divertor plasma and scrape-off layer (SOL). Detailed measurements of the ELM energy and particle deposition footprint on the secondary divertor target plates were made with a fast IR camera and Langmuir probes and SOL profile and transport measurements were made with reciprocating probes. The secondary divertor and SOL conditions depended on changes in the magnetic balance and the core plasma density. Larger density resulted in smaller ELMs and the magnetic balance affected how many ELM particles coupled to the secondary SOL and divertor. Particularly striking are the images from a new fast IR camera that resolve ELM heat pulses and show spiral patterns with multiple peaks during ELMs in the secondary divertor.

Advanced Divertor Design and Application under Modern Superconducting Tokamak Constraints

NASA Astrophysics Data System (ADS)

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

2013-10-01

With current ITER projections already predicting divertor exhaust heat loads in the 5-10 MW/m2 range, i.e. at the maximum tolerance, it is clear that the divertor heat load problem will only be exacerbated for future superconducting tokamaks, as well as perhaps some modern tokamaks today. Thus, an advanced divertor, such as the X-Divertor (XD), Super-X Divertor (SXD), or Snowflake (SF) will become a virtual necessity to reduce incident heat flux at the target plates. Using the 2D magnetic equilibrium code CORSICA, we explore the possibilities of creating an advanced divertor for a next-generation superconducting tokamak (Ip = 15 MA, BT = 5.3 T, R = 6.2 m) under nominal engineering constraints. Advanced divertors were achieved with no in-vessel PF coils, PF current densities below 30 MA/m2, and vertical maintenance access, all of which are favorable conditions for tokamaks today. Both the XD and SF divertors are readily achievable while maintaining core plasma performance, and the advantages and disadvantages of each are discussed in turn. Some thought is given as to how the divertor cassette will need to be modified to accommodate advanced divertors. Work supported under US-DOE projects DE-FG02-04ER54742 and DE-FG02-04ER54754.

Characterizing Tungsten Sourcing and SOL Transport during the Metal Rings Campaign

NASA Astrophysics Data System (ADS)

Thomas, D. M.; Abrams, T.; Unterberg, E. A.; Donovan, D.; Elder, J. D.; Wampler, W. R.; DIII-D Team

2017-10-01

The Metal Rings Campaign on DIII-D utilized two isotopically and poloidally distinct toroidal arrays of tungsten coated inserts in the lower divertor to study W divertor erosion near the outer strike point (OSP) and divertor entrance and subsequent migration in a mixed-material (C-W) environment. In AT hybrid discharges (PAUX = 14 MW, H98 = 1.6, βN = 3.7) with rapid ELMs (fELM 200 Hz, δW/W 0.7%) W impurities are seen to reach the midplane predominantly from the OSP region rather than the divertor entrance (far-SOL). Conversely, in scenarios with less frequent larger ELMs (fELM 60 Hz, δW/W 3.6%), the W impurities are found to transport equally from the OSP and entrance region. ELM-resolved spectroscopic measurements of W sourcing indicate that large ELMs can source W at many times the inter ELM rate. The peak W erosion rate can shift radially outwards consistent with the ELM energy flux, thereby shifting the balance between strikepoint and far-SOL sources. Changes in the peak erosion locations between forward and reversed Bt discharges are consistent with ExB ion drift effects. Evidence for a near-SOL impurity buildup between the divertors driven by the parallel grad-Ti force is also seen. Work supported under USDOE Cooperative Agreement DE-FC02-04ER54698.

Characterization of edge turbulence in different states of divertor detachment using reflectometry in the ASDEX Upgrade tokamak

NASA Astrophysics Data System (ADS)

Nikolaeva, V.; Guimarais, L.; Manz, P.; Carralero, D.; Manso, M. E.; Stroth, U.; Silva, C.; Conway, G. D.; Seliunin, E.; Vicente, J.; Brida, D.; Aguiam, D.; Santos, J.; Silva, A.; ASDEX Upgrade team; MST1 team

2018-05-01

Transport in the scrape-off layer (SOL) depends on the state of divertor detachment. L-mode discharges were analyzed where the state of divertor detachment is varied through a density ramp-up. By means of reflectometry measurements at the low (LFS) and the high field side (HFS), midplane density fluctuations are studied for the first time in ASDEX Upgrade simultaneously at both sides of the tokamak. Radial density fluctuation profiles (δ {n}e/{n}e) increase with radius in both the HFS and the LFS. It is found that in the SOL density fluctuations at the LFS have about a factor of two larger amplitude than at the HFS in agreement with ballooned transport. Density fluctuations at the LFS show a modest variation with increasing background density resulting mainly from a rise of low frequency components. Experimental results are in good agreement with an enhanced convection of filaments at the LFS at the beginning of outer divertor detachment leading to a flatter SOL density profile. In this phase of the discharge, density fluctuations measured at the HFS far-SOL display a strong increase, which may be associated with the presence of faster filaments originated at the LFS.

Rapid change of blob structure in the outer scrape-off layer (SOL)

NASA Astrophysics Data System (ADS)

Cohen, R. H.

2005-10-01

Nonlinear structures (``blobs'') driven by the magnetic field curvature and highly elongated along the field lines may exist in the tokamak SOL.footnotetextS.I. Krasheninnikov. Phys. Lett. A 283, 368 (2001) The contact of the blob end with the divertor plate significantly affects the blob structure and velocity. However, the strong shearing of the flux-tube near the X-point makes impossible direct electrical contact of the blob in the upper SOL and the divertor, so that the sheath boundary condition (BC) has to be replaced by a BC imposed near the X point.footnotetextD. Ryutov, R.H. Cohen. Contr. Pl. Phys 44, 168 (2004) We show that, at larger distances from the separatrix, in the far SOL, the connection between the upper SOL and the divertor plate is re-established, and the sheath BC becomes again relevant. During the blob's outward radial motion, this event is reflected in a sudden change of its length, from the blob extending only to the X point to the blob extending down to the plate. Likewise, a blob initially existing only in the divertor leg becomes suddenly longer, and extends to the whole SOL.

Measurements of tungsten migration in the DIII-D divertor

NASA Astrophysics Data System (ADS)

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

2017-12-01

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

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.

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

Parallel Energy Transport in Detached DIII-D Divertor Plasmas

NASA Astrophysics Data System (ADS)

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

2017-10-01

A comparison of experiment and modeling of detached divertor plasmas is examined in the context of parallel energy transport. Experimental estimates of power carried by electron thermal conduction versus plasma convection are experimentally inferred from power balance measurements of radiated power and target plate heat flux combined with Thomson scattering measurements of the Te profile along the divertor leg. Experimental profiles of Te exhibit relatively low gradients with Te < 15 eV from the X-point to the target implying transport dominated by convection. In contrast, fluid modeling with SOLPS produces sharp Te gradients for Te > 3 eV, characteristic of transport dominated by electron conduction through the bulk of the divertor. This discrepancy with experimental transport dominated by convection and modeling by conduction has significant implications for the radiative capacity of divertor plasmas and may explain at least part of the difficulty for fluid modeling to obtain the experimentally observed radiative losses. Comparisons are also made for helium plasmas where the match between experiment and modeling is much better. Work supported by the US DOE under DE-FC02-04ER54698.

A study of X-divertor in NSTX-U with SOLPS simulations

NASA Astrophysics Data System (ADS)

Chen, Zhong-Ping; Kotschenreuther, Mike; Mahajan, Swadesh; Gerhardt, Stefan

2018-03-01

The X-divertor (XD) geometry in NSTX-U is demonstrated, via SOLPS simulations, to perform better than the standard divertor (SD); in particular, it allows detachment at a lower upstream density and stabilizes the detachment front near the target, away from the main X-point. Consequently a stable detached operation becomes possible—the localization near the plate allows a vast reduction of heat fluxes without degrading the core plasma. Indeed, it is confirmed by our simulation that at similar states of detachment the XD outperforms the SD by reducing the heat fluxes to the target and maintaining higher upstream temperatures, resulting in scrape-off layers that are more favorable for advanced tokamak operation. These advantages are attributed to the unique geometric characteristics of XD—poloidal flaring near the target.

Global transport of light elements boron and carbon in the full-W ASDEX Upgrade

NASA Astrophysics Data System (ADS)

ASDEX Upgrade Team; Hakola, A.; Likonen, J.; Koivuranta, S.; Krieger, K.; Mayer, M.; Neu, R.; Rohde, V.; Sugiyama, K.

2011-08-01

Transport of carbon and boron has been investigated in the full-W ASDEX Upgrade after experimental campaigns with (2008) and without (2007) boronizations. For this purpose, poloidal deposition profiles of the two elements on tungsten and graphite regions of lower-divertor tiles have been determined. Carbon is mainly deposited in the inner divertor - 80-90% of the determined 12C and 13C inventories on W - while boron shows a much more symmetric deposition profile. In the unboronized machine, the boron inventories are a factor of 10 smaller than in the boronized case and result from residual boron atoms left in the torus prior to the 2007 campaign. Both carbon and boron are deposited more efficiently and/or show less erosion on graphite than on tungsten, particularly in the outer divertor. For 13C, the difference is 10-100 in favor of graphite. This is most probably caused by a higher re-erosion from tungsten surfaces.

Evaluating Stellarator Divertor Designs with EMC3

NASA Astrophysics Data System (ADS)

Bader, Aaron; Anderson, D. T.; Feng, Y.; Hegna, C. C.; Talmadge, J. N.

2013-10-01

In this paper various improvements of stellarator divertor design are explored. Next step stellarator devices require innovative divertor solutions to handle heat flux loads and impurity control. One avenue is to enhance magnetic flux expansion near strike points, somewhat akin to the X-Divertor concept in Tokamaks. The effect of judiciously placed external coils on flux deposition is calculated for configurations based on the HSX stellarator. In addition, we attempt to optimize divertor plate location to facilitate the external coil placement. Alternate areas of focus involve altering edge island size to elucidate the driving physics in the edge. The 3-D nature of stellarators complicates design and necessitates analysis of new divertor structures with appropriate simulation tools. We evaluate the various configurations with the coupled codes EMC3-EIRENE, allowing us to benchmark configurations based on target heat flux, impurity behavior, radiated power, and transitions to high recycling and detached regimes. Work supported by DOE-SC0006103.

SOLPS simulations of X-divertor in NSTX-U

NASA Astrophysics Data System (ADS)

Chen, Zhongping; Kotschenreuther, Mike; Mahajan, Swadesh

2017-10-01

The X-divertor (XD) geometry in NSTX-U has demonstrated, in SOLPS simulations, a better performance than the standard divertor (SD) regarding detachment: achieving detachment with a lower upstream density and stabilizing the detachment front near the target. The benefits of such a localized front is that the power exhaust requirement can be satisfied without the radiation front encroaching on the core plasma. It is also found by our simulations that at similar states of detachment the XD outperforms the SD by reducing the heat fluxes to the target and maintaining higher upstream temperatures. These advantages are attributed to the unique geometric characteristics of XD - poloidal flaring near the target. The detailed physical mechanisms behind the better XD performance that is found in the simulations will be examined. Work supported by US DOE under DE-FG02-04ER54742 and SC 0012956.

Lesson from Tungsten Leading Edge Heat Load Analysis in KSTAR Divertor

NASA Astrophysics Data System (ADS)

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

2016-10-01

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

Innovative divertor concept development on DIII-D and EAST

DOE PAGES

Guo, H. Y.; Allen, S.; Canik, J.; ...

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.

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

NASA Astrophysics Data System (ADS)

Kawashima, H.; Hoshino, K.; Shimizu, K.; Takizuka, T.; Ide, S.; Sakurai, S.; Asakura, N.

2011-08-01

The JT-60SA divertor design has been established on the basis of engineering requirements and physics analysis. Heat and particle fluxes under the full input power of 41 MW can give severe heat loads on the divertor targets, while the allowable heat load is limited below 15 MW/m2. Dependence of the heat flux mitigation on a D2 gas-puff is evaluated by SONIC simulations for high density (ne_ave ˜ 1 × 1020 m-3) high current plasmas. It is found that the peak heat load 10 MW/m2 with dense (ned > 4 × 1020 m-3) and cold (Ted, Tid ⩽ 1 eV) divertor plasmas are obtained at a moderate gas-puff of Γpuff = 15 × 1021 s-1. Divertor plasmas are controlled from attached to detached condition using the divertor pump with pumping-speed below 100 m3/s. In full non-inductive current drive plasmas with low density (ne_ave ˜ 5 × 1019 m-3), the reduction of divertor heat load is achieved with the Ar injection.

Automated divertor target design by adjoint shape sensitivity analysis and a one-shot method

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dekeyser, W., E-mail: Wouter.Dekeyser@kuleuven.be; Reiter, D.; Baelmans, M.

As magnetic confinement fusion progresses towards the development of first reactor-scale devices, computational tokamak divertor design is a topic of high priority. Presently, edge plasma codes are used in a forward approach, where magnetic field and divertor geometry are manually adjusted to meet design requirements. Due to the complex edge plasma flows and large number of design variables, this method is computationally very demanding. On the other hand, efficient optimization-based design strategies have been developed in computational aerodynamics and fluid mechanics. Such an optimization approach to divertor target shape design is elaborated in the present paper. A general formulation ofmore » the design problems is given, and conditions characterizing the optimal designs are formulated. Using a continuous adjoint framework, design sensitivities can be computed at a cost of only two edge plasma simulations, independent of the number of design variables. Furthermore, by using a one-shot method the entire optimization problem can be solved at an equivalent cost of only a few forward simulations. The methodology is applied to target shape design for uniform power load, in simplified edge plasma geometry.« less

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

NASA Astrophysics Data System (ADS)

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

2007-06-01

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

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

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 flowmore » 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.« less

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 performance in terms of the parameters of discharges.

Divertor tungsten tile melting and its effect on core plasma performance

NASA Astrophysics Data System (ADS)

Lipschultz, B.; Coenen, J. W.; Barnard, H. S.; Howard, N. T.; Reinke, M. L.; Whyte, D. G.; Wright, G. M.

2012-12-01

For the 2007 and 2008 run campaigns, Alcator C-Mod operated with a full toroidal row of tungsten tiles in the high heat flux region of the outer divertor; tungsten levels in the core plasma were below measurement limits. An accidental creation of a tungsten leading edge in the 2009 campaign led to this study of a melting tungsten source: H-mode operation with strike point in the region of the melting tile was immediately impossible due to some fraction of tungsten droplets reaching the main plasma. Approximately 15 g of tungsten was lost from the tile over ˜100 discharges. Less than 1% of the evaporated tungsten was found re-deposited on surfaces, the rest is assumed to have become dust. The strong discharge variability of the tungsten reaching the core implies that the melt layer topology is always varying. There is no evidence of healing of the surface with repeated melting. Forces on the melted tungsten tend to lead to prominences that extend further into the plasma. A discussion of the implications of melting a divertor tungsten monoblock on the ITER plasma is presented.

Calculation of ion distribution functions and neoclassical transport in the edge of single-null divertor tokamaks

NASA Astrophysics Data System (ADS)

Rognlien, T. D.; Cohen, R. H.; Xu, X. Q.

2007-11-01

The ion distribution function in the H-mode pedestal region and outward across the magnetic separatrix is expected to have a substantial non-Maxwellian character owing to the large banana orbits and steep gradients in temperature and density. The 4D (2r,2v) version of the TEMPEST continuum gyrokinetic code is used with a Coulomb collision model to calculate the ion distribution in a single-null tokamak geometry throughout the pedestal/scrape-off-layer regions. The mean density, parallel velocity, and energy radial profiles are shown at various poloidal locations. The collisions cause neoclassical energy transport through the pedestal that is then lost to the divertor plates along the open field lines outside the separatrix. The resulting heat flux profiles at the inner and outer divertor plates are presented and discussed, including asymmetries that depend on the B-field direction. Of particular focus is the effect on ion profiles and fluxes of a radial electric field exhibiting a deep well just inside the separatrix, which reduces the width of the banana orbits by the well-known squeezing effect.

Material Transport in ASDEX Upgrade

NASA Astrophysics Data System (ADS)

Rohde, V.; Dux, R.; Mayer, M.; Neu, R.; PA~ 1/4 tterich, T.; Schneider, W.; ASDEX Upgrade-Team,

Today carbon is the most common first wall material in fusion experiments, whereas the first wall of the next step device will consist of a mixture of elements. Especially tungsten has been shown to be an alternative to low-Z materials. However, even with 40% of tungsten coated plasma facing components, carbon is still the dominant impurity at ASDEX Upgrade. A consistent picture of the carbon migration in ASDEX Upgrade has been achieved. Primary carbon sources are the protection limiters at the low field side of the main chamber. Eroded carbon is distributed all over the main chamber. So, the initially tungsten coated central column acts as the main carbon source during discharges, even though a considerable amount of tungsten surfaces persists. Carbon coverage of the central column can significantly change on a shot to shot basis. The divertor target plates act as a strong carbon sink. Deposits are found at the inner and outer divertor, which may be re-eroded forming precursors for layer production at remote areas. In ASDEX Upgrade, deposits on the subdivertor structure are formed by hydro-carbons with a high effective sticking coefficient. A parasitic plasma at these locations may enhance the surface loss probability by surface activation. At more remote areas, such as the pump ducts, a very small deposition is found. Non sticking hydro-carbons are effectively pumped by the cryopump and turbo molecular pumps.

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.

Upgrades toward high-heat flux, liquid lithium plasma-facing components in the NSTX-U

DOE PAGES

Jaworski, M. A.; Brooks, A.; Kaita, R.; ...

2016-08-08

Liquid metal plasma-facing components (PFCs) provide numerous potential advantages over solid-material components. One critique of the approach is the relatively less developed technologies associated with deploying these components in a fusion plasma-experiment. Exploration of the temperature limits of liquid lithium PFCs in a tokamak divertor and the corresponding consequences on core operation are a high priority informing the possibilities for future liquid lithium PFCs. An all-metal NSTX-U is envisioned to make direct comparison between all high-Z wall operation and liquid lithium PFCs in a single device. By executing the all-metal upgrades incrementally, scientific productivity will be maintained while enabling physicsmore » and engineering-science studies to further develop the solid- and liquid-metal components. Six major elements of a flowing liquid-metal divertor system are described and a three-step program for implementing this system is laid out. The upgrade steps involve the first high-Z divertor target upgrade in NSTX-U, pre-filled liquid metal targets and finally, an integrated, flowing liquid metal divertor target. As a result, two example issues are described where the engineering and physics experiments are shown to be closely related in examining the prospects for future liquid metal PFCs.« less

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

NASA Astrophysics Data System (ADS)

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

2017-08-01

The Alcator C-Mod team has recently developed a feedback system to measure and control surface heat flux in real-time. The system uses real-time measurements of surface heat flux from surface thermocouples and a pulse-width modulated piezo valve to inject low-Z impurities (typically N2) into the private flux region. It has been used in C-Mod to mitigate peak surface heat fluxes  >40 MW m-2 down to  <10 MW m-2 while maintaining excellent core confinement, H 98  >  1. While the system works quite well under relatively steady conditions, use of it during transients has revealed important limitations on feedback control of impurity seeding in conventional vertical target plate divertors. In some cases, the system is unable to avoid plasma reattachment to the divertor plate or the formation of a confinement-damaging x-point MARFE. This is due to the small operational window for mitigated heat flux in the parameters of incident plasma heat flux, plasma density, and impurity density as well as the relatively slow response of the impurity gas injection system compared to plasma transients. Given the severe consequences for failure of such a system to operate reliably in a reactor, there is substantial risk that the conventional vertical target plate divertor will not provide an adequately controllable system in reactor-class devices. These considerations motivate the need to develop passively stable, highly compliant divertor configurations and experimental facilities that can test such possible solutions.

Analysis of heat transfer and erosion effects on ITER divertor plasma facing components induced by slow high-power transients

DOE Office of Scientific and Technical Information (OSTI.GOV)

Federici, G.; Raffray, A.R.; Chiocchio, S.

1995-12-31

This paper presents the results of an analysis carried out to investigate the thermal response of ITER divertor plasma facing components (PFC`s) clad with Be, W, and CFC, to high-recycling, high-power thermal transients (i.e. 10--30 MW/m{sup 2}) which are anticipated to last up to a few seconds. The armour erosion and surface melting are estimated for the different plasma facing materials (PFM`s) together with the maximum heat flux to the coolant, and armour/heat-sink interface temperature. The analysis assumes that intense target evaporation will lead to high radiative power losses in the plasma in front of the target which self-protects themore » target. The cases analyzed clarify the influence of several key parameters such as the plasma heat flux to the target, the loss of the melt layer, the duration of the event, the thickness of the armour, and comparison is made with cases without vapor shielding. Finally, some implications for the performance and lifetime of divertor PFC`s clad with different PFM`s are discussed.« less

Dependence of the L-Mode scrape-off layer power fall-off length on the upper triangularity in TCV

NASA Astrophysics Data System (ADS)

Faitsch, M.; Maurizio, R.; Gallo, A.; Coda, S.; Eich, T.; Labit, B.; Merle, A.; Reimerdes, H.; Sieglin, B.; Theiler, C.; the Eurofusion MST1 Team; the TCV Team

2018-04-01

This paper reports on experimental observations on TCV with a scan in upper triangularity {δ }up}, including negative triangularity, focusing on the power fall-off length {λ }{{q}} in L-Mode. The upper triangularity is scanned from -0.28 to 0.47. Smaller {λ }{{q}}out} is measured at the outer divertor target for decreasing {δ }up} together with higher edge temperature {T}{{e},{edge}} leading to increased confinement. This effect is observed for both magnetic drift directions for discharges in deuterium and helium. In helium larger {λ }{{q}} values are observed compared to deuterium. The power fall-off length at the inner divertor target {λ }{{q}}in} has a non-monotonic behaviour with changing triangularity. The largest values are around {δ }up}=0. The ratio {λ }{{q}}in}/{λ }{{q}}out} increases for decreasing {δ }up} for positive triangularity and is approximately constant for negative triangularity. {λ }{{q}}out} is compared to available scaling laws. Partial agreement is only observed for a scaling law containing a proxy for {T}{{e},{edge}} at ASDEX Upgrade (Sieglin 2016 Plasma Phys. Control. Fusion 58 055015). Extending this scaling to TCV and using {T}{{e},{edge}} at {ρ }pol}=0.95 suggests that {λ }{{q}}out} is independent of machine size {λ }{{q}}{{L} - {Mode}} ({mm}) = 165\\cdot {B}pol}{({{T}})}-0.66\\cdot A{({{u}})}-0.15\\cdot {T}{{e},{edge}}{({eV})}-0.93\\cdot R{({{m}})}-0.03. Possible explanations for smaller {λ }{{q}}out} for decreasing {δ }up} is a reduction in turbulence or a direct effect of increasing {T}{{e},{edge}}.

Recent sheath physics studies on DIII-D

NASA Astrophysics Data System (ADS)

Watkins, J. G.; Labombard, B.; Stangeby, P. C.; Lasnier, C. J.; McLean, A. G.; Nygren, R. E.; Boedo, J. A.; Leonard, A. W.; Rudakov, D. L.

2015-08-01

A study to examine some current issues in the physics of the plasma sheath has been recently carried out in DIII-D low power Ohmic plasmas using both flush and domed Langmuir probes, divertor Thomson scattering (DTS), an infrared camera (IRTV), and a new calorimeter triple probe assembly mounted on the Divertor Materials Evaluation System (DIMES). The sheath power transmission factor was found to be consistent with the theoretically predicted value of 7 (±2) for low power plasmas. Using this factor, the three heat flux profiles derived from the LP, DTS, and calorimeter diagnostic measurements agree. Comparison of flush and domed Langmuir probes and divertor Thomson scattering indicates that proper interpretation of flush probe data to get target plate density and temperature is feasible and could potentially yield accurate measurements of target plate conditions where the probes are located.

Minimum magnetic curvature for resilient divertors using Compact Toroidal Hybrid geometry

NASA Astrophysics Data System (ADS)

Bader, A.; Hegna, C. C.; Cianciosa, M.; Hartwell, G. J.

2018-05-01

The properties of resilient divertors are explored using equilibria derived from Compact Toroidal Hybrid (CTH) geometries. Resilience is defined here as the robustness of the strike point patterns as the plasma geometry and/or plasma profiles are changed. The addition of plasma current in the CTH configurations significantly alters the shape of the last closed flux surface and the rotational transform profile, however, it does not alter the strike point pattern on the target plates, and hence has resilient divertor features. The limits of when a configuration transforms to a resilient configuration is then explored. New CTH-like configurations are generated that vary from a perfectly circular cross section to configurations with increasing amounts of toroidal shaping. It is found that even small amounts of toroidal shaping lead to strike point localization that is similar to the standard CTH configuration. These results show that only a small degree of three-dimensional shaping is necessary to produce a resilient divertor, implying that any highly shaped optimized stellarator will possess the resilient divertor property.

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.

Spectroscopic investigation of carbon migration with tungsten walls in ASDEX Upgrade

NASA Astrophysics Data System (ADS)

Kallenbach, A.; Dux, R.; Harhausen, J.; Maggi, C. F.; Neu, R.; Pütterich, T.; Rohde, V.; Schmid, K.; Wolfrum, E.; ASDEX Upgrade Team

2007-06-01

Spectroscopic measurements of carbon fluxes in the mainly tungsten-coated ASDEX Upgrade tokamak are analysed with a particle transport and migration code. The transport parameters for deuterium and carbon are calibrated against flux measurements for different experimental conditions. Additional information is obtained from the re-appearance time of carbon after a boronisation. The code reproduces the experimental finding that despite a 85% (2006 campaign) tungsten coverage of the primary PFCs, the carbon concentration in the core and edge plasma is reduced by about a factor 2 only compared to full carbon PFCs. This behaviour is explained by the strong main chamber recycling of carbon in comparison with the loss flux to the inner divertor. The quick recovery of the carbon level in the plasma after a boronisation is explained by carbon influx from the outer divertor.

3D nonlinear numerical simulation of the current-convective instability in detached diverter plasma

NASA Astrophysics Data System (ADS)

Stepanenko, Alexander; Krasheninnikov, Sergei

2017-10-01

One of the possible mechanisms responsible for strong radiation fluctuations observed in the recent experiments with detached plasmas at ASDEX Upgrade [Potzel et al., Nuclear Fusion, 2014] can be related to the onset of the current-convective instability (CCI) driven by strong asymmetry of detachment in the inner and outer tokamak divertors [Krasheninnikov and Smolyakov, PoP, 2016]. In this study we present the first results of 3D nonlinear numerical simulations of the CCI in divertor plasma for the conditions relevant to the AUG experiment. The general physical model used to simulate the CCI, qualitative estimates for the instability characteristic growth rate and transverse wavelengths derived for plasma, which is spatially inhomogeneous both across and along the magnetic field lines, are presented. The simulation results, demonstrating nonlinear dynamics of the CCI, provide the frequency spectra of turbulent divertor plasma fluctuations showing good agreement with the available experimental data. This material is based upon the work supported by the U.S. Department of Energy under Award No. DE-FG02-04ER54739 at UCSD and by the Russian Ministry of Education and Science Grant No. 14.Y26.31.0008 at MEPhI.

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.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Budaev, V. P., E-mail: budaev@mail.ru

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 ofmore » 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.« less

Modelling the detachment dependence on strike point location in the small angle slot divertor (SAS) with SOLPS

NASA Astrophysics Data System (ADS)

Casali, Livia; Covele, Brent; Guo, Houyang

2017-10-01

The new Small Angle Slot (SAS) divertor in DIII-D is characterized by a shallow-angle target enclosed by a slot structure about the strike point (SP). SOLPS modelling results of SAS have demonstrated divertor closure's utility in widening the range of acceptable densities for adequate heat handling. An extensive database of runs has been built to study the detachment dependence on SP location in SAS. Density scans show that lower Te at lower upstream density occur when the SP is at the critical location in the slot. The cooling front spreads across the entire target at higher densities, in agreement with experimental Langmuir probe measurements. A localized increase of the atomic and molecular density takes place near the SP, which reduces the target incident power density and facilitates detachment at lower upstream density. Systematic scans of variables such as power, transport, and viscosity have been carried out to assess the detachment sensitivity. Therein, a positive role of the viscosity is found. This work supported by DOE Contract Number DE-FC02-04ER54698.

Partial detachment of high power discharges in ASDEX Upgrade

NASA Astrophysics Data System (ADS)

Kallenbach, A.; Bernert, M.; Beurskens, M.; Casali, L.; Dunne, M.; Eich, T.; Giannone, L.; Herrmann, A.; Maraschek, M.; Potzel, S.; Reimold, F.; Rohde, V.; Schweinzer, J.; Viezzer, E.; Wischmeier, M.; the ASDEX Upgrade Team

2015-05-01

Detachment of high power discharges is obtained in ASDEX Upgrade by simultaneous feedback control of core radiation and divertor radiation or thermoelectric currents by the injection of radiating impurities. So far 2/3 of the ITER normalized heat flux Psep/R = 15 MW m-1 has been obtained in ASDEX Upgrade under partially detached conditions with a peak target heat flux well below 10 MW m-2. When the detachment is further pronounced towards lower peak heat flux at the target, substantial changes in edge localized mode (ELM) behaviour, density and radiation distribution occur. The time-averaged peak heat flux at both divertor targets can be reduced below 2 MW m-2, which offers an attractive DEMO divertor scenario with potential for simpler and cheaper technical solutions. Generally, pronounced detachment leads to a pedestal and core density rise by about 20-40%, moderate (<20%) confinement degradation and a reduction of ELM size. For AUG conditions, some operational challenges occur, like the density cut-off limit for X-2 electron cyclotron resonance heating, which is used for central tungsten control.

Divertor Coil Design and Implementation on Pegasus

NASA Astrophysics Data System (ADS)

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

2012-10-01

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

Minimum magnetic curvature for resilient divertors using Compact Toroidal Hybrid geometry

DOE PAGES

Bader, Aaron; Hegna, C. C.; Cianciosa, Mark R.; ...

2018-03-16

The properties of resilient divertors are explored using equilibria derived from Compact Toroidal Hybrid (CTH) geometries. Resilience is defined here as the robustness of the strike point patterns as the plasma geometry and/or plasma profiles are changed. The addition of plasma current in the CTH configurations significantly alters the shape of the last closed flux surface and the rotational transform profile, however, it does not alter the strike point pattern on the target plates, and hence has resilient divertor features. The limits of when a configuration transforms to a resilient configuration is then explored. New CTH-like configurations are generated thatmore » vary from a perfectly circular cross section to configurations with increasing amounts of toroidal shaping. It is found that even small amounts of toroidal shaping lead to strike point localization that is similar to the standard CTH configuration. Lastly, these results show that only a small degree of three-dimensional shaping is necessary to produce a resilient divertor, implying that any highly shaped optimized stellarator will possess the resilient divertor property.« less

Plasma power recycling at the divertor surface

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tang, Xian -Zhu; Guo, Zehua

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

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

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 coil current yield a reduction of the width of the divertor flux spreading to about 20-25 cm and cause increased peak heat fluxes back to values similar to those in the axisymmetric case. The dependencies of these features on the divertor recycling regime and the perpendicular transport assumptions, as well as toroidal averaged effects mimicking rotation of the RMP field, are discussed in the paper.

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

NASA Astrophysics Data System (ADS)

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

2008-10-01

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

Suppression of tritium retention in remote areas of ITER by nonperturbative reactive gas injection.

PubMed

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

2010-10-22

A technique based on reactive gas injection in the afterglow region of the divertor plasma is proposed for the suppression of tritium-carbon codeposits in remote areas of ITER when operated with carbon-based divertor targets. Experiments in a divertor simulator plasma device indicate that a 4  nm/min deposition can be suppressed by addition of 1  Pa·m³ s⁻¹ ammonia flow at 10 cm from the plasma. These results bolster the concept of nonperturbative scavenger injection for tritium inventory control in carbon-based fusion plasma devices, thus paving the way for ITER operation in the active phase under a carbon-dominated, plasma facing component background.

Manufacturing and testing of a prototypical divertor vertical target for ITER

NASA Astrophysics Data System (ADS)

Merola, M.; Plöchl, L.; Chappuis, Ph; Escourbiac, F.; Grattarola, M.; Smid, I.; Tivey, R.; Vieider, G.

2000-12-01

After an extensive R&D activity, a medium-scale divertor vertical target prototype has been manufactured by the EU Home Team. This component contains all the main features of the corresponding ITER divertor design and consists of two units with one cooling channel each, assembled together and having an overall length and width of about 600 and 50 mm, respectively. The upper part of the prototype has a tungsten macro-brush armour, whereas the lower part is covered by CFC monoblocks. A number of joining techniques were required to manufacture this component as well as an appreciable effort in the development of suitable non-destructive testing methods. The component was high heat flux tested in FE200 electron beam facility at Le Creusot, France. It endured 100 cycles at 5 MW/m 2, 1000 cycles at 10 MW/m 2 and more then 1000 cycles at 15-20 MW/m 2. The final critical heat flux test reached a value in excess of 30 MW/m 2.

EMC3-EIRENE modelling of toroidally-localized divertor gas injection experiments on Alcator C-Mod

DOE PAGES

Lore, Jeremy D.; Reinke, M. L.; LaBombard, Brian; ...

2014-09-30

Experiments on Alcator C-Mod with toroidally and poloidally localized divertor nitrogen injection have been modeled using the three-dimensional edge transport code EMC3-EIRENE to elucidate the mechanisms driving measured toroidal asymmetries. In these experiments five toroidally distributed gas injectors in the private flux region were sequentially activated in separate discharges resulting in clear evidence of toroidal asymmetries in radiated power and nitrogen line emission as well as a ~50% toroidal modulation in electron pressure at the divertor target. The pressure modulation is qualitatively reproduced by the modelling, with the simulation yielding a toroidal asymmetry in the heat flow to the outermore » strike point. Finally, toroidal variation in impurity line emission is qualitatively matched in the scrape-off layer above the strike point, however kinetic corrections and cross-field drifts are likely required to quantitatively reproduce impurity behavior in the private flux region and electron temperatures and densities directly in front of the target.« less

Measurements of plasma sheath heat flux in the Alcator C-Mod divertor

NASA Astrophysics Data System (ADS)

Brunner, Dan; Labombard, Brian; Terry, Jim; Reinke, Matt

2010-11-01

Heat flux is one of the most important parameters controlling the lifetime of first-wall components in fusion experiments and reactors. The sheath heat flux coefficient (γ) is a parameter relating heat flux (from a plasma to a material surface) to the electron temperature and ion saturation current. Being such a simple expression for a kinetic process, it is of great interest to plasma edge fluid modelers. Under the assumptions of equal ion and electron temperatures, no secondary electron emission, and no net current to the surface the value of γ is approximately 7 [1]. Alcator C-Mod provides a unique opportunity among today's experiments to measure reactor-relevant heat fluxes (100's of MW/m^2 parallel to the magnetic field) in reactor-like divertor geometry. Motivated by the DoE 2010 joint milestone to measure heat flux footprints, the lower outer divertor of Alcator has been instrumented with a suite of Langmuir probes, novel surface thermocouples, and calorimeters in tiles purposefully ramped to eliminate shadowing; all within view of an IR camera. Initial results indicate that the experimentally inferred values of γ are found to agree with simple theory in the sheath limited regime and diverges to lower values as the density increases.

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

NASA Astrophysics Data System (ADS)

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

2014-03-01

Progress in the definition of the requirements for edge localized mode (ELM) control and the application of ELM control methods both for high fusion performance DT operation and non-active low-current operation in ITER is described. Evaluation of the power fluxes for low plasma current H-modes in ITER shows that uncontrolled ELMs will not lead to damage to the tungsten (W) divertor target, unlike for high-current H-modes in which divertor damage by uncontrolled ELMs is expected. Despite the lack of divertor damage at lower currents, ELM control is found to be required in ITER under these conditions to prevent an excessive contamination of the plasma by W, which could eventually lead to an increased disruptivity. Modelling with the non-linear MHD code JOREK of the physics processes determining the flow of energy from the confined plasma onto the plasma-facing components during ELMs at the ITER scale shows that the relative contribution of conductive and convective losses is intrinsically linked to the magnitude of the ELM energy loss. Modelling of the triggering of ELMs by pellet injection for DIII-D and ITER has identified the minimum pellet size required to trigger ELMs and, from this, the required fuel throughput for the application of this technique to ITER is evaluated and shown to be compatible with the installed fuelling and tritium re-processing capabilities in ITER. The evaluation of the capabilities of the ELM control coil system in ITER for ELM suppression is carried out (in the vacuum approximation) and found to have a factor of ˜2 margin in terms of coil current to achieve its design criterion, although such a margin could be substantially reduced when plasma shielding effects are taken into account. The consequences for the spatial distribution of the power fluxes at the divertor of ELM control by three-dimensional (3D) fields are evaluated and found to lead to substantial toroidal asymmetries in zones of the divertor target away from the separatrix. Therefore, specifications for the rotation of the 3D perturbation applied for ELM control in order to avoid excessive localized erosion of the ITER divertor target are derived. It is shown that a rotation frequency in excess of 1 Hz for the whole toroidally asymmetric divertor power flux pattern is required (corresponding to n Hz frequency in the variation of currents in the coils, where n is the toroidal symmetry of the perturbation applied) in order to avoid unacceptable thermal cycling of the divertor target for the highest power fluxes and worst toroidal power flux asymmetries expected. The possible use of the in-vessel vertical stability coils for ELM control as a back-up to the main ELM control systems in ITER is described and the feasibility of its application to control ELMs in low plasma current H-modes, foreseen for initial ITER operation, is evaluated and found to be viable for plasma currents up to 5-10 MA depending on modelling assumptions.

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

DOE PAGES

Ahn, J. -W.; Briesemester, A. R.; Kobayashi, M.; ...

2017-06-22

Enhanced perpendicular heat and momentum transport induces parallel pressure loss leading to divertor detachment, which can be produced by the increase of density in 2D tokamaks. However, in the 3D configurations such as tokamaks with 3D fields and stellarators, the fraction of perpendicular transport can be higher even in a lower density regime, which could lead to the early transition to detachment without passing through the high-recycling regime. 3D fields applied to the limiter tokamak plasmas produce edge stochastic layers close to the last closed flux surface (LCFS), which can allow for enhanced perpendicular transport and indeed the absence ofmore » 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. In conclusion, work for optimal parameter window for best divertor operation scenario is needed particularly for the 3D divertor tokamak configuration.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ahn, J. -W.; Briesemester, A. R.; Kobayashi, M.

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 ofmore » 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. In conclusion, work for optimal parameter window for best divertor operation scenario is needed particularly for the 3D divertor tokamak configuration.« less

Dynamic power balance analysis in JET

NASA Astrophysics Data System (ADS)

Matthews, G. F.; Silburn, S. A.; Challis, C. D.; Eich, T.; Iglesias, D.; King, D.; Sieglin, B.; Contributors, JET

2017-12-01

The full scale realisation of nuclear fusion as an energy source requires a detailed understanding of power and energy balance in current experimental devices. In this we explore whether a global power balance model in which some of the calibration factors applied to the source or sink terms are fitted to the data can provide insight into possible causes of any discrepancies in power and energy balance seen in the JET tokamak. We show that the dynamics in the power balance can only be properly reproduced by including the changes in the thermal stored energy which therefore provides an additional opportunity to cross calibrate other terms in the power balance equation. Although the results are inconclusive with respect to the original goal of identifying the source of the discrepancies in the energy balance, we do find that with optimised parameters an extremely good prediction of the total power measured at the outer divertor target can be obtained over a wide range of pulses with time resolution up to ∼25 ms.

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

NASA Astrophysics Data System (ADS)

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

2017-12-01

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

Initial results from divertor heat-flux instrumentation on Alcator C-Mod

NASA Astrophysics Data System (ADS)

Labombard, B.; Brunner, D.; Payne, J.; Reinke, M.; Terry, J. L.; Hughes, J. W.; Lipschultz, B.; Whyte, D.

2009-11-01

Physics-based plasma transport models that can accurately simulate the heat-flux power widths observed in the tokamak boundary are lacking at the present time. Yet this quantity is of fundamental importance for ITER and most critically important for DEMO, a reactor similar to ITER but with ˜4 times the power exhaust. In order to improve our understanding, C-Mod, DIII-D and NSTX will aim experiments in FY10 towards characterizing the divertor ``footprint'' and its connection to conditions ``upstream'' in the boundary and core plasmas [2]. Standard IR-based heat-flux measurements are particularly difficult in C-Mod, due to its vertical-oriented divertor targets. To overcome this, a suite of embedded heat-flux sensor probes (tile thermocouples, calorimeters, surface thermocouples) combined with IR thermography was installed during the FY09 opening, along with a new divertor bolometer system. This paper will report on initial experiments aimed at unfolding the heat-flux dependencies on plasma operating conditions. [2] a proposed US DoE Joint Facilities Milestone.

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

NASA Astrophysics Data System (ADS)

Escourbiac, F.; Richou, M.; Guigon, R.; Constans, S.; Durocher, A.; Merola, M.; Schlosser, J.; Riccardi, B.; Grosman, A.

2009-12-01

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

Design of snowflake-diverted equilibria of CFETR

NASA Astrophysics Data System (ADS)

Hang, LI; Xiang, GAO; Guoqiang, LI; Zhengping, LUO; Damao, YAO; Yong, GUO

2018-03-01

The Chinese Fusion Engineering Test Reactor (CFETR) represents the next generation of full superconducting fusion reactors in China. Recently, CFETR was redesigned with a larger size and will be operated in two phases. To reduce the heat flux on the target plate, a snowflake (SF) divertor configuration is proposed. In this paper we show that by adding two dedicated poloidal field (PF) coils, the SF configuration can be achieved in both phases. The equilibria were calculated by TEQ code for a range of self-inductances l i3. The coil currents were calculated at some fiducial points in the flattop phase. The results indicate that the PF coil system has the ability to maintain a long flattop phase in 7.5 and 10 MA inductive scenarios for the single null divertor (SND) and SF divertor configurations. The properties of the SF configuration were also analyzed. The connection length and flux expansion of the SF divertor were both increased significantly over the SND.

Effects of two-dimensional magnetic uncertainties and three-dimensional error and perturbation fields on the Small Angle Slot divertor geometry and topology [Effects of two- and three-dimensional magnetic fields on the Small Angle Slot divertor magnetic topology

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trevisan, Gregorio L.; Lao, Lang L.; Evans, Todd E.

The Small Angle Slot (SAS) was recently installed on DIII-D as an advanced divertor, promising easier plasma detachment and lower temperatures across the whole target. A twofold study of the SAS magnetic topology is presented in this paper. On one hand, a twodimensional uncertainty quantification analysis is carried out through a Monte Carlo approach in order to understand the level of accuracy of two-dimensional equilibrium computations in reconstructing the strike point and angle onto the divertor. Under typical experimental conditions, the uncertainties are found to be roughly 6.8 mm and 0.56 deg, respectively. On the other hand, a three-dimensional “vacuum”more » analysis is carried out to understand the effects of typical external perturbation fields on the scrape-off layer topology. When the threedimensional I-coils are switched on, poloidally-localized lobes are found to appear, grow, and hit the SAS target, although barely, even for 5 kA; at the same time, the strike point modulation is found to be roughly 1.8 mm and thus negligible for most purposes. Furthermore, such results complement previous two-dimensional analyses in characterizing typical SAS equilibria and provide useful background information for planning and interpreting SAS experiments.« less

Effects of two-dimensional magnetic uncertainties and three-dimensional error and perturbation fields on the Small Angle Slot divertor geometry and topology [Effects of two- and three-dimensional magnetic fields on the Small Angle Slot divertor magnetic topology

DOE PAGES

Trevisan, Gregorio L.; Lao, Lang L.; Evans, Todd E.; ...

2018-01-04

The Small Angle Slot (SAS) was recently installed on DIII-D as an advanced divertor, promising easier plasma detachment and lower temperatures across the whole target. A twofold study of the SAS magnetic topology is presented in this paper. On one hand, a twodimensional uncertainty quantification analysis is carried out through a Monte Carlo approach in order to understand the level of accuracy of two-dimensional equilibrium computations in reconstructing the strike point and angle onto the divertor. Under typical experimental conditions, the uncertainties are found to be roughly 6.8 mm and 0.56 deg, respectively. On the other hand, a three-dimensional “vacuum”more » analysis is carried out to understand the effects of typical external perturbation fields on the scrape-off layer topology. When the threedimensional I-coils are switched on, poloidally-localized lobes are found to appear, grow, and hit the SAS target, although barely, even for 5 kA; at the same time, the strike point modulation is found to be roughly 1.8 mm and thus negligible for most purposes. Furthermore, such results complement previous two-dimensional analyses in characterizing typical SAS equilibria and provide useful background information for planning and interpreting SAS experiments.« less

Integration of uncooled scraper elements and its diagnostics into Wendelstein 7-X

DOE PAGES

Fellinger, Joris; Loesser, Doug; Neilson, Hutch; ...

2017-08-08

The modular stellarator Wendelstein 7-X in Greifswald (Germany) successfully started operation in 2015 with short pulse limiter plasmas. In 2017, the next operation phase (OP) OP1.2 will start once 10 uncooled test divertor units (TDU) with graphite armor will be installed. The TDUs allow for plasma pulses of 10 s with 8 MW heating. OP2, allowing for steady state operation, is planned for 2020 after the TDUs will be replaced by 10 water cooled CFC armored divertors. Due to the development of plasma currents like bootstrap currents in long pulse plasmas in OP2, the plasma could hit the edge ofmore » the divertor targets which has a reduced cooling capacity compared to the central part of the target tiles. To prevent overloading of these edges, a so-called scraper element can be positioned in front of the divertor, intersecting those strike lines that would otherwise hit the divertor edges. As a result, these edges are protected but as a drawback the pumping efficiency of neutrals is also reduced. As a test an uncooled scraper element with graphite tiles will be placed in two out of ten half modules in OP1.2. A decision to install ten water cooled scraper elements for OP2 is pending on the results of this test in OP1.2. To monitor the impact of the scraper element on the plasma, Langmuir probes are integrated in the plasma facing surface, and a neutral gas manometer measures the neutral density directly behind the plasma facing surface. Moreover, IR and VIS cameras observe the plasma facing surface and thermocouples monitor the temperatures of the graphite tiles and underlying support structure. This paper describes the integration of the scraper element and its diagnostics in Wendelstein 7-X.« less

Overview of ASDEX Upgrade results

NASA Astrophysics Data System (ADS)

Stroth, U.; Adamek, J.; Aho-Mantila, L.; Äkäslompolo, S.; Amdor, C.; Angioni, C.; Balden, M.; Bardin, S.; Barrera Orte, L.; Behler, K.; Belonohy, E.; Bergmann, A.; Bernert, M.; Bilato, R.; Birkenmeier, G.; Bobkov, V.; Boom, J.; Bottereau, C.; Bottino, A.; Braun, F.; Brezinsek, S.; Brochard, T.; Brüdgam, M.; Buhler, A.; Burckhart, A.; Casson, F. J.; Chankin, A.; Chapman, I.; Clairet, F.; Classen, I. G. J.; Coenen, J. W.; Conway, G. D.; Coster, D. P.; Curran, D.; da Silva, F.; de Marné, P.; D'Inca, R.; Douai, D.; Drube, R.; Dunne, M.; Dux, R.; Eich, T.; Eixenberger, H.; Endstrasser, N.; Engelhardt, K.; Esposito, B.; Fable, E.; Fischer, R.; Fünfgelder, H.; Fuchs, J. C.; Gál, K.; García Muñoz, M.; Geiger, B.; Giannone, L.; Görler, T.; da Graca, S.; Greuner, H.; Gruber, O.; Gude, A.; Guimarais, L.; Günter, S.; Haas, G.; Hakola, A. H.; Hangan, D.; Happel, T.; Härtl, T.; Hauff, T.; Heinemann, B.; Herrmann, A.; Hobirk, J.; Höhnle, H.; Hölzl, M.; Hopf, C.; Houben, A.; Igochine, V.; Ionita, C.; Janzer, A.; Jenko, F.; Kantor, M.; Käsemann, C.-P.; Kallenbach, A.; Kálvin, S.; Kantor, M.; Kappatou, A.; Kardaun, O.; Kasparek, W.; Kaufmann, M.; Kirk, A.; Klingshirn, H.-J.; Kocan, M.; Kocsis, G.; Konz, C.; Koslowski, R.; Krieger, K.; Kubic, M.; Kurki-Suonio, T.; Kurzan, B.; Lackner, K.; Lang, P. T.; Lauber, P.; Laux, M.; Lazaros, A.; Leipold, F.; Leuterer, F.; Lindig, S.; Lisgo, S.; Lohs, A.; Lunt, T.; Maier, H.; Makkonen, T.; Mank, K.; Manso, M.-E.; Maraschek, M.; Mayer, M.; McCarthy, P. J.; McDermott, R.; Mehlmann, F.; Meister, H.; Menchero, L.; Meo, F.; Merkel, P.; Merkel, R.; Mertens, V.; Merz, F.; Mlynek, A.; Monaco, F.; Müller, S.; Müller, H. W.; Münich, M.; Neu, G.; Neu, R.; Neuwirth, D.; Nocente, M.; Nold, B.; Noterdaeme, J.-M.; Pautasso, G.; Pereverzev, G.; Plöckl, B.; Podoba, Y.; Pompon, F.; Poli, E.; Polozhiy, K.; Potzel, S.; Püschel, M. J.; Pütterich, T.; Rathgeber, S. K.; Raupp, G.; Reich, M.; Reimold, F.; Ribeiro, T.; Riedl, R.; Rohde, V.; Rooij, G. v.; Roth, J.; Rott, M.; Ryter, F.; Salewski, M.; Santos, J.; Sauter, P.; Scarabosio, A.; Schall, G.; Schmid, K.; Schneider, P. A.; Schneider, W.; Schrittwieser, R.; Schubert, M.; Schweinzer, J.; Scott, B.; Sempf, M.; Sertoli, M.; Siccinio, M.; Sieglin, B.; Sigalov, A.; Silva, A.; Sommer, F.; Stäbler, A.; Stober, J.; Streibl, B.; Strumberger, E.; Sugiyama, K.; Suttrop, W.; Tala, T.; Tardini, G.; Teschke, M.; Tichmann, C.; Told, D.; Treutterer, W.; Tsalas, M.; Van Zeeland, M. A.; Varela, P.; Veres, G.; Vicente, J.; Vianello, N.; Vierle, T.; Viezzer, E.; Viola, B.; Vorpahl, C.; Wachowski, M.; Wagner, D.; Wauters, T.; Weller, A.; Wenninger, R.; Wieland, B.; Willensdorfer, M.; Wischmeier, M.; Wolfrum, E.; Würsching, E.; Yu, Q.; Zammuto, I.; Zasche, D.; Zehetbauer, T.; Zhang, Y.; Zilker, M.; Zohm, H.

2013-10-01

The medium size divertor tokamak ASDEX Upgrade (major and minor radii 1.65 m and 0.5 m, respectively, magnetic-field strength 2.5 T) possesses flexible shaping and versatile heating and current drive systems. Recently the technical capabilities were extended by increasing the electron cyclotron resonance heating (ECRH) power, by installing 2 × 8 internal magnetic perturbation coils, and by improving the ion cyclotron range of frequency compatibility with the tungsten wall. With the perturbation coils, reliable suppression of large type-I edge localized modes (ELMs) could be demonstrated in a wide operational window, which opens up above a critical plasma pedestal density. The pellet fuelling efficiency was observed to increase which gives access to H-mode discharges with peaked density profiles at line densities clearly exceeding the empirical Greenwald limit. Owing to the increased ECRH power of 4 MW, H-mode discharges could be studied in regimes with dominant electron heating and low plasma rotation velocities, i.e. under conditions particularly relevant for ITER. The ion-pressure gradient and the neoclassical radial electric field emerge as key parameters for the transition. Using the total simultaneously available heating power of 23 MW, high performance discharges have been carried out where feed-back controlled radiative cooling in the core and the divertor allowed the divertor peak power loads to be maintained below 5 MW m-2. Under attached divertor conditions, a multi-device scaling expression for the power-decay length was obtained which is independent of major radius and decreases with magnetic field resulting in a decay length of 1 mm for ITER. At higher densities and under partially detached conditions, however, a broadening of the decay length is observed. In discharges with density ramps up to the density limit, the divertor plasma shows a complex behaviour with a localized high-density region in the inner divertor before the outer divertor detaches. Turbulent transport is studied in the core and the scrape-off layer (SOL). Discharges over a wide parameter range exhibit a close link between core momentum and density transport. Consistent with gyro-kinetic calculations, the density gradient at half plasma radius determines the momentum transport through residual stress and thus the central toroidal rotation. In the SOL a close comparison of probe data with a gyro-fluid code showed excellent agreement and points to the dominance of drift waves. Intermittent structures from ELMs and from turbulence are shown to have high ion temperatures even at large distances outside the separatrix.

Time-resolved deposition in the remote region of the JET-ILW divertor: measurements and modelling

NASA Astrophysics Data System (ADS)

Catarino, N.; Widdowson, A.; Baron-Wiechec, A.; Coad, J. P.; Heinola, K.; Rubel, M.; Alves, E.; Contributors, JET

2017-12-01

One crucial requirement for the development of fusion power is to know where, and how much, impurities collect in the machine, and how much of the fuelling isotope tritium will be trapped therein. The most relevant information on this issue comes from the operation of the Joint European Tokamak (JET), which is the world’s largest operating tokamak and has the same interior plasma-facing materials as the next step machine, ITER. Much of the information gained so far has been from post-mortem analysis of samples collected after whole campaigns involving varied types of operation. This paper describes time-resolved measurements of the deposition rate using rotating collectors (RC) placed in remote areas of the JET divertor during the 2013-2014 campaign with the ITER-like Wall (ILW). These techniques allow the effects of different types of operation to be distinguished. Rotating collectors made of silicon discs housed behind an aperture are exposed to the plasma. Each time the magnetic field coils are ramped up for a discharge the disc rotates, providing a linear relationship between the exposed region and the discharge number. Post-mortem ion beam analyses provide information on the deposit composition as a function of the discharge number. The results show that the Be deposition average for the RC in the corners of the inner and outer divertor are 4.9 × 1016 cm-2 and 1.8 × 1017 cm-2, respectively, accumulated over an average of ˜25 pulses. Data from the rotating collector below Tile 5 in the central region of divertor indicate a Be deposition rate of 9.3 × 1015 cm-2, per ˜25 pulses.

Fast camera imaging of dust in the DIII-D tokamak

NASA Astrophysics Data System (ADS)

Yu, J. H.; Rudakov, D. L.; Pigarov, A. Yu.; Smirnov, R. D.; Brooks, N. H.; Muller, S. H.; West, W. P.

2009-06-01

Naturally occurring and injected dust particles are observed in the DIII-D tokamak in the outer midplane scrape-off-layer (SOL) using a visible fast-framing camera, and the size of dust particles is estimated using the observed particle lifetime and theoretical ablation rate of a carbon sphere. Using this method, the lower limit of detected dust radius is ˜3 μm and particles with inferred radius as large as ˜1 mm are observed. Dust particle 2D velocities range from approximately 10 to 300 m/s with velocities inversely correlated with dust size. Pre-characterized 2-4 μm diameter diamond dust particles are introduced at the lower divertor in an ELMing H-mode discharge using the divertor materials evaluation system (DiMES), and these particles are found to be at the lower size limit of detection using the camera with resolution of ˜0.2 cm 2 per pixel and exposure time of 330 μs.

Measurement and modeling of surface temperature dynamics of the NSTX liquid lithium divertor

NASA Astrophysics Data System (ADS)

McLean, A. G.; Gan, K. F.; Ahn, J.-W.; Gray, T. K.; Maingi, R.; Abrams, T.; Jaworski, M. A.; Kaita, R.; Kugel, H. W.; Nygren, R. E.; Skinner, C. H.; Soukhanovskii, V. A.

2013-07-01

Dual-band infrared (IR) measurements of the National Spherical Torus eXperiment (NSTX) Liquid Lithium Divertor (LLD) are reported that demonstrate liquid Li is more effective at removing plasma heat flux than Li-conditioned graphite. Extended dwell of the outer strike point (OSP) on the LLD caused an incrementally larger area to be heated above the Li melting point through the discharge leading to enhanced D retention and plasma confinement. Measurement of Tsurface near the OSP demonstrates a significant reduction of the LLD surface temperature compared to that of Li-coated graphite at the same major radius. Modeling of these data with a 2-D simulation of the LLD structure in the DFLUX code suggests that the structure of the LLD was successful at handling up to q⊥,peak = 5 MW/m2 inter-ELM and up to 10 MW/m2 during ELMs from its plasma-facing surface as intended, and provide an innovative method for inferring the Li layer thickness.

Exploring the engineering limit of heat flux of a W/RAFM divertor target for fusion reactors

NASA Astrophysics Data System (ADS)

Mao, X.; Fursdon, M.; Chang, X. B.; Zhang, J. W.; Liu, P.; Ellwood, G.; Qian, X. Y.; Qin, S. J.; Peng, X. B.; Barrett, T. R.; Liu, P.

2018-06-01

The design and development of a fusion reactor divertor plasma facing component (PFC) is one of the many challenging issues on the road to commercial use of fusion energy. The divertor PFC is expected to exhaust steady state heat loads in the region of 10 MW m‑2 while keeping temperatures and thermo-mechanical stresses in its structure within the allowable limits. For ITER (International Thermo-Nuclear Experimental Reactor) a water cooled W/CuCrZr divertor PFC concept has been developed. However, this concept is not necessarily assured for use in future fusion reactors mainly because the neutron radiation dose would be at least an order magnitude higher, resulting in limited thermo-mechanical performance and considerably more activated waste products. In the present study, a water cooled divertor PFC using reduced activation ferritic-martensitic (RAFM) steel as the heat sink pipe has been designed with pressurised water reactor-like cooling conditions (pressure of 15.5 MPa, velocity of 10–20 m s‑1 and temperature of 300 °C). The PFC is made up of a number of rectangular tungsten tiles, each with an inner circular hole (so-called monoblocks), joined onto a RAFM steel pipe with copper interlayers. The thermo-mechanical performance of the PFC has been studied in detail. The heat transfer coefficient between the RAFM pipe inner surface and the water was calculated using published correlations. Geometric parameters and water velocity were optimized with finite element (FE) thermal analysis, to achieve acceptable temperatures in the structure given the target exhaust heat load of 10 MW m‑2. Under this heat load and the optimised thermal design parameters, the structure of the PFC was further assessed by mechanical analysis. We find that under these conditions the RAFM steel pipe experiences cyclic plasticity, and fails the common linear elastic ratchetting (3 Sm) rule. Nevertheless, the designed W/RAFM divertor PFU can withstand 10 MW m‑2 heat load, albeit with a fatigue life of approximately 0.55 years based on the expected operation scenario of a prototype or test reactor. This study extends the state of knowledge of the technological limit of a divertor based on a RAFM steel pipe structure.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ahn, J. -W.; Briesemester, A. R.; Kobayashi, M.

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 ofmore » 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. In conclusion, work for optimal parameter window for best divertor operation scenario is needed particularly for the 3D divertor tokamak configuration.« less

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.

Feedback system for divertor impurity seeding based on real-time measurements of surface heat flux in the Alcator C-Mod tokamak

NASA Astrophysics Data System (ADS)

Brunner, D.; Burke, W.; Kuang, A. Q.; LaBombard, B.; Lipschultz, B.; Wolfe, S.

2016-02-01

Mitigation of the intense heat flux to the divertor is one of the outstanding problems in fusion energy. One technique that has shown promise is impurity seeding, i.e., the injection of low-Z gaseous impurities (typically N2 or Ne) to radiate and dissipate the power before it arrives to the divertor target plate. To this end, the Alcator C-Mod team has created a first-of-its-kind feedback system to control the injection of seed gas based on real-time surface heat flux measurements. Surface thermocouples provide real-time measurements of the surface temperature response to the plasma heat flux. The surface temperature measurements are inputted into an analog computer that "solves" the 1-D heat transport equation to deliver accurate, real-time signals of the surface heat flux. The surface heat flux signals are sent to the C-Mod digital plasma control system, which uses a proportional-integral-derivative (PID) algorithm to control the duty cycle demand to a pulse width modulated piezo valve, which in turn controls the injection of gas into the private flux region of the C-Mod divertor. This paper presents the design and implementation of this new feedback system as well as initial results using it to control divertor heat flux.

Feedback system for divertor impurity seeding based on real-time measurements of surface heat flux in the Alcator C-Mod tokamak.

PubMed

Brunner, D; Burke, W; Kuang, A Q; LaBombard, B; Lipschultz, B; Wolfe, S

2016-02-01

Mitigation of the intense heat flux to the divertor is one of the outstanding problems in fusion energy. One technique that has shown promise is impurity seeding, i.e., the injection of low-Z gaseous impurities (typically N2 or Ne) to radiate and dissipate the power before it arrives to the divertor target plate. To this end, the Alcator C-Mod team has created a first-of-its-kind feedback system to control the injection of seed gas based on real-time surface heat flux measurements. Surface thermocouples provide real-time measurements of the surface temperature response to the plasma heat flux. The surface temperature measurements are inputted into an analog computer that "solves" the 1-D heat transport equation to deliver accurate, real-time signals of the surface heat flux. The surface heat flux signals are sent to the C-Mod digital plasma control system, which uses a proportional-integral-derivative (PID) algorithm to control the duty cycle demand to a pulse width modulated piezo valve, which in turn controls the injection of gas into the private flux region of the C-Mod divertor. This paper presents the design and implementation of this new feedback system as well as initial results using it to control divertor heat flux.

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

NASA Astrophysics Data System (ADS)

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

2010-11-01

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

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

NASA Astrophysics Data System (ADS)

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

2017-10-01

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

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

NASA Astrophysics Data System (ADS)

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

2017-12-01

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murphy, Christopher; Nygren, R. E.; Chrobak, C P.

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

A fluid modeling perspective on the tokamak power scrape-off width using SOLPS-ITER

NASA Astrophysics Data System (ADS)

Meier, Eric

2016-10-01

SOLPS-ITER, a 2D fluid code, is used to conduct the first fluid modeling study of the physics behind the power scrape-off width (λq). When drift physics are activated in the code, λq is insensitive to changes in toroidal magnetic field (Bt), as predicted by the 0D heuristic drift (HD) model developed by Goldston. Using the HD model, which quantitatively agrees with regression analysis of a multi-tokamak database, λq in ITER is projected to be 1 mm instead of the previously assumed 4 mm, magnifying the challenge of maintaining the peak divertor target heat flux below the technological limit. These simulations, which use DIII-D H-mode experimental conditions as input, and reproduce the observed high-recycling, attached outer target plasma, allow insights into the scrape-off layer (SOL) physics that set λq. Independence of λq with respect to Bt suggests that SOLPS-ITER captures basic HD physics: the effect of Bt on the particle dwell time ( Bt) cancels with the effect on drift speed ( 1 /Bt), fixing the SOL plasma density width, and dictating λq. Scaling with plasma current (Ip), however, is much weaker than the roughly 1 /Ip dependence predicted by the HD model. Simulated net cross-separatrix particle flux due to magnetic drifts exceeds the anomalous particle transport, and a Pfirsch-Schluter-like SOL flow pattern is established. Up-down ion pressure asymmetry enables the net magnetic drift flux. Drifts establish in-out temperature asymmetry, and an associated thermoelectric current carries significant heat flux to the outer target. The density fall-off length in the SOL is similar to the electron temperature fall-off length, as observed experimentally. Finally, opportunities and challenges foreseen in ongoing work to extrapolate SOLPS-ITER and the HD model to ITER and future machines will be discussed. Supported by U.S. Department of Energy Contract DESC0010434.

Experimental validation of an analytical kinetic model for edge-localized modes in JET-ITER-like wall

NASA Astrophysics Data System (ADS)

Guillemaut, C.; Metzger, C.; Moulton, D.; Heinola, K.; O’Mullane, M.; Balboa, I.; Boom, J.; Matthews, G. F.; Silburn, S.; Solano, E. R.; contributors, JET

2018-06-01

The design and operation of future fusion devices relying on H-mode plasmas requires reliable modelling of edge-localized modes (ELMs) for precise prediction of divertor target conditions. An extensive experimental validation of simple analytical predictions of the time evolution of target plasma loads during ELMs has been carried out here in more than 70 JET-ITER-like wall H-mode experiments with a wide range of conditions. Comparisons of these analytical predictions with diagnostic measurements of target ion flux density, power density, impact energy and electron temperature during ELMs are presented in this paper and show excellent agreement. The analytical predictions tested here are made with the ‘free-streaming’ kinetic model (FSM) which describes ELMs as a quasi-neutral plasma bunch expanding along the magnetic field lines into the Scrape-Off Layer without collisions. Consequences of the FSM on energy reflection and deposition on divertor targets during ELMs are also discussed.

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.

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

NASA Astrophysics Data System (ADS)

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

2017-12-01

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

Progress in extrapolating divertor heat fluxes towards large fusion devices

NASA Astrophysics Data System (ADS)

Sieglin, B.; Faitsch, M.; Eich, T.; Herrmann, A.; Suttrop, W.; Collaborators, JET; the MST1 Team; the ASDEX Upgrade Team

2017-12-01

Heat load to the plasma facing components is one of the major challenges for the development and design of large fusion devices such as ITER. Nowadays fusion experiments can operate with heat load mitigation techniques, e.g. sweeping, impurity seeding, but do not generally require it. For large fusion devices however, heat load mitigation will be essential. This paper presents the current progress of the extrapolation of steady state and transient heat loads towards large fusion devices. For transient heat loads, so-called edge localized modes are considered a serious issue for the lifetime of divertor components. In this paper, the ITER operation at half field (2.65 T) and half current (7.5 MA) will be discussed considering the current material limit for the divertor peak energy fluence of 0.5 {MJ}/{{{m}}}2. Recent studies were successful in describing the observed energy fluence in the JET, MAST and ASDEX Upgrade using the pedestal pressure prior to the ELM crash. Extrapolating this towards ITER results in a more benign heat load compared to previous scalings. In the presence of magnetic perturbation, the axisymmetry is broken and a 2D heat flux pattern is induced on the divertor target, leading to local increase of the heat flux which is a concern for ITER. It is shown that for a moderate divertor broadening S/{λ }{{q}}> 0.5 the toroidal peaking of the heat flux disappears.

Numerical analyses of baseline JT-60SA design concepts with the COREDIV code

NASA Astrophysics Data System (ADS)

Zagórski, R.; Gałązka, K.; Ivanova-Stanik, I.; Stępniewski, W.; Garzotti, L.; Giruzzi, G.; Neu, R.; Romanelli, M.

2017-06-01

JT-60SA reference design scenarios at high (#3) and low (#2) density have been analyzed with the help of the self-consistent COREDIV code. Simulations results for a standard C wall and full W wall have been compared in terms of the influence of impurities, both intrinsic (C, W) and seeded (N, Ar, Ne, Kr), on the radiation losses and plasma parameters. For scenario #3 in a C environment, the regime of detachment on divertor plates can be achieved with N or Ne seeding, whereas for the low density and high power scenario (#2), the C and seeding impurity radiation does not effectively reduce power to the targets. In this case, only an increase of either average density or edge density together with Kr seeding might help to develop conditions with strong radiation losses and semi-detached conditions in the divertor. The calculations show that, in the case of a W divertor, the power load to the plate is mitigated by seeding and the central plasma dilution is smaller compared to the C divertor. For the high density case (#3) with Ne seeding, operation in full detachment mode is predicted. Ar seems to be an optimal choice for the low-density high-power scenario #2, showing a wide operating window, whereas Ne leads to high plasma dilution at high seeding levels albeit not achieving semi-detached conditions in the divertor.

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

PubMed

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

2012-10-01

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

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

NASA Astrophysics Data System (ADS)

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

2016-10-01

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

Neutral pressure behavior for diverted discharges in the Wendelstein 7-AS Stellarator

NASA Astrophysics Data System (ADS)

McCormick, K.; Grigull, P.; Burhenn, R.; Ehmler, H.; Feng, Y.; Giannone, L.; Haas, G.; Sardei, F.; NBI-, ECRH-; W7-AS Teams

2005-03-01

On the W7-AS stellarator, the subdivertor neutral pressure in an up-down divertor pair as well as at two points in the vicinity of a lower divertor module in the main chamber are measured. Results are presented for ι=5/9 island divertor discharges under conditions of normal confinement (NC) and the HDH-mode for: n˜0.1-4×1020 m-3, Pecrh = 0.5-1.5 MW, Pnbi = 2 MW, and H + and D + plasmas, with both normal- and reversed- Bt for H +. Subdivertor pressures are in the range 1-2 × 10 -3 mbar for HDH conditions. For plasma detachment at the target plates a strong up-down pressure asymmetry arises, with pup/ pdown ⩽ 5. The asymmetry reverses with reversed Bt. Main vessel pressures are a factor of 5-10 lower than the average subdivertor pressure for H +, with D + plasmas exhibiting still lower values.

Carbon Radiation Studies in the DIII-D Divertor with the Monte Carlo Impurity (MCI) Code

NASA Astrophysics Data System (ADS)

Evans, T. E.; Leonard, A. W.; West, W. P.; Finkenthal, D. F.; Fenstermacher, M. E.; Porter, G. D.; Chu, Y.

1998-11-01

Carbon sputtering and transport are modeled in the DIII--D divertor with the MCI code. Calculated 2-D radiation patterns are compared with measured radiation distributions. The results are particularly sensitive to Ti near the divertor target plates. For example, increasing the ion temperature from 8 eV to 20 eV in MCI raises P_rad^div from 1626 to 2862 kW. Although this presents difficulties in assessing which sputtering model best describes the plasma-surface interaction physics (because of experimental uncertainties in T_i), processes which either produce too much or too little radiated power compared to the measured value of 1718 kW can be eliminated. Based on this, the number of viable sputtering options has been reduced from 12 to 4. For the conditions studied, three of these options involve both physical and chemical sputtering, and one requires only physical sputtering.

Targeted Optimization of Quasi-Symmetric Stellarators

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hegna, Chris C.; Anderson, D. T.; Talmadge, J. N.

2016-10-06

The proposed research focuses on targeted areas of plasma physics dedicated to improving the stellarator concept. Research was pursued in the technical areas of edge/divertor physics in 3D configurations, magnetic island physics in stellarators, the role of 3D shaping on microinstabilities and turbulent transport and energetic ion confinement in stellarators.

Kinetic studies of divertor heat fluxes in Alcator C-Mod

NASA Astrophysics Data System (ADS)

Pankin, A. Y.; Bateman, G.; Kritz, A. H.; Rafiq, T.; Park, G. Y.; Chang, C. S.; Brunner, D.; Hughes, J. W.; Labombard, B.; Terry, J.

2010-11-01

The kinetic XGC0 code [C.S. Chang et al, Phys. Plasmas 11 (2004) 2649] is used to model the H- mode pedestal and SOL regions in Alcator C-Mod discharges. The self-consistent simulations in this study include kinetic neoclassical physics and anomalous transport models along with the ExB flow shear effects. The heat fluxes on the divertor plates are computed and the fluxes to the outer plate are compared with experimental observations. The dynamics of the radial electric field near the separatrix and in the SOL region are computed with the XGC0 code, and the effect of the anomalous transport on the heat fluxes in the SOL region is investigated. In particular, the particle and thermal diffusivities obtained in the analysis mode are compared with predictions from the theory-based anomalous transport models such as MMM95 [G. Bateman et al, Phys. Plasmas 5 (1998) 1793] and DRIBM [T. Rafiq et al, to appear in Phys. Plasmas (2010)]. It is found that there is a notable pinch effect in the inner separatrix region. Possible physical mechanisms for the particle and thermal pinches are discussed.

Erosion and Surface Morphology of Silicon Carbide Under Variable DIII-D Divertor Heat Fluxes

NASA Astrophysics Data System (ADS)

Bringuier, Stefan; Abrams, Tyler; Khalifa, Hesham; Thomas, Dan; Holland, Leo; Rudakov, Dmitry; Briesemeister, Alexis

2017-10-01

A SiC coating of 250 μm, deposited onto a graphite DiMES cap via chemical vapor deposition, was exposed to 80 s of H-mode plasma bombardment in the DIII-D outer divertor with steady-state heat fluxes up to 3 MW m-2 and transient loads due to ELMs typically peaking at 10 MW m-2. In-situ monitoring of Si I and Si II atomic spectral lines revealed the presence of significant neutral Si and Si+ impurity influx, which are used to determine quantitative erosion rates via the S/XB method. No visual macroscopic flaking or delamination of the SiC coating was observed, supporting the notion that SiC is thermal-mechanically robust and compatible with graphite substrates at elevated temperatures. Post-mortem profilometric analysis also indicates no pronounced change in surface roughness after plasma exposure. Finally, we investigate aspects of preferential sputtering and changes to surface composition exposure using scanning electron microscopy and Auger electron spectroscopy. Work supported under USDOE Cooperative Agreement DE-FC02-04ER54698.

Heat removal capability of divertor coaxial tube assembly

NASA Astrophysics Data System (ADS)

Shibui, Masanao; Nakahira, Masataka; Tada, Eisuke; Takatsu, Hideyuki

1994-05-01

To deal with high power flowing in the divertor region, an advanced divertor concept with gas target has been proposed for use in ITER/EDA. The concept uses a divertor channel to remove the radiated power while allowing neutrals to recirculate. Candidate channel wall designs include a tube array design where many coaxial tubes are arranged in the toroidal direction to make louver. The coaxial tube consists of a Be protection tube encases many supply tubes wound helically around a return tube. V-alloy and hardened Cu-alloy have been proposed for use in the supply and return tubes. Some coolants have also been proposed for the design including pressurized He and liquid metals, because these coolants are consistent with the selection of coolants for the blanket and also meet the requirement of high temperature operation. In the coaxial tube design, the coolant area is restricted and brittle Be material is used under severe thermal cyclings. Thus, to obtain the coaxial tube with sufficient safety margin for the expected fusion power excursion, it is essential to understand its applicability limit. The paper discusses heat removal capability of the coaxial tube and recommends some design modifications.

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

NASA Astrophysics Data System (ADS)

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

2017-10-01

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

Preliminary Results on the Heat Deposition on Divertor Plate using Low MN Map

NASA Astrophysics Data System (ADS)

Ali, Halima; Punjabi, Alkesh; Boozer, Allen

2003-10-01

The study of magnetic field line behavior and the closely related plasma behavior are important not only for their tokamak application but also for their application to other Hamiltonian, or near-Hamiltonian, systems. The behavior of field lines near a tokamak separatrix has been studied extensively using various approaches. Our approach is called method of maps. In this paper, we introduce an area-preserving map called Low MN map. We first derive the map from the general theory of maps /1/, and then use it to calculate the effects of m = 1, n = 1 perturbations on the stochastic layer and magnetic footprint in single-null divertor tokamaks. We show that there are self-similarities, singularities, and topological equivalences in the pattern of physical parameters that characterize the stochastic layer and the magnetic footprint. Preliminary results in the investigation on the heat distribution on the divertor plate indicate multiple peaked in heat flux profile distributed radially across the divertor target when the amplitude is 10-3. This, and other features, are in good agreement with experimental observations. This work is done under the DOE grant number DE-FG02-01ER54624. 1. A. Punjabi et al, J. Plasma Phys. 52, 91 (1994).

Modeling of rapid shutdown in the DIII-D tokamak by core deposition of high-Z material

DOE PAGES

Izzo, Valerie A.; Parks, Paul B.

2017-06-22

MHD modeling of shell-pellet injection for disruption mitigation is carried out under the assumption of idealized delivery of the radiating payload to the core, neglecting the physics of shell ablation. The shell pellet method is designed to produce an inside-out thermal quench in which core thermal heat is radiated while outer flux surfaces remain intact, protecting the divertor from large conducted heat loads. In the simulation, good outer surfaces remain until the thermal quench is nearly complete, and a high radiated energy fraction is achieved. As a result, when the outermost surfaces are destroyed, runaway electron test orbits indicate thatmore » the rate of runaway electron loss is very fast compared with prior massive gas injection simulations, which is attributed to the very different current profile evolution that occurs with central cooling.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lumsdaine, A.; Bjorholm, T.; Harris, J.

The Wendelstein 7-X stellarator is in final stages of commissioning, and will begin operation in late 2015. In the first phase, the machine will operate with a limiter, and will be restricted to low power and short pulse. But in 2019, plans are for an actively cooled divertor to be installed, and the machine will operate in steady state at full power. Recently, plasma simulations have indicated that, in this final operational phase, a bootstrap current will evolve in certain scenarios. This will cause the sensitive ends of the divertor target to be overloaded beyond their qualified limit. A highmore » heat flux scraper element (HHF-SE) has been proposed in order to take up some of the convective flux and reduce the load on the divertor. In order to examine whether the HHF-SE will be able to effectively reduce the plasma flux in the divertor region of concern, and to determine how the pumping effectiveness will be affected by such a component, it is planned to include a test divertor unit scraper element (TDU-SE) in 2017 during an earlier operational phase. Several U.S. fusion energy science laboratories have been involved in the design, analysis (structural and thermal finite element, as well as computational fluid dynamics), plasma simulation, planning, prototyping, and diagnostic development around the scraper element program (both TDU-SE and HHF-SE). As a result, this paper presents an overview of all of these activities and their current status.« less

Overview of design and analysis activities for the W7-X scraper element

DOE PAGES

Lumsdaine, A.; Bjorholm, T.; Harris, J.; ...

2016-08-18

The Wendelstein 7-X stellarator is in final stages of commissioning, and will begin operation in late 2015. In the first phase, the machine will operate with a limiter, and will be restricted to low power and short pulse. But in 2019, plans are for an actively cooled divertor to be installed, and the machine will operate in steady state at full power. Recently, plasma simulations have indicated that, in this final operational phase, a bootstrap current will evolve in certain scenarios. This will cause the sensitive ends of the divertor target to be overloaded beyond their qualified limit. A highmore » heat flux scraper element (HHF-SE) has been proposed in order to take up some of the convective flux and reduce the load on the divertor. In order to examine whether the HHF-SE will be able to effectively reduce the plasma flux in the divertor region of concern, and to determine how the pumping effectiveness will be affected by such a component, it is planned to include a test divertor unit scraper element (TDU-SE) in 2017 during an earlier operational phase. Several U.S. fusion energy science laboratories have been involved in the design, analysis (structural and thermal finite element, as well as computational fluid dynamics), plasma simulation, planning, prototyping, and diagnostic development around the scraper element program (both TDU-SE and HHF-SE). As a result, this paper presents an overview of all of these activities and their current status.« less

Transport simulations of linear plasma generators with the B2.5-Eirene and EMC3-Eirene codes

DOE PAGES

Rapp, Juergen; Owen, Larry W.; Bonnin, X.; ...

2014-12-20

Linear plasma generators are cost effective facilities to simulate divertor plasma conditions of present and future fusion reactors. For this research, the codes B2.5-Eirene and EMC3-Eirene were extensively used for design studies of the planned Material Plasma Exposure eXperiment (MPEX). Effects on the target plasma of the gas fueling and pumping locations, heating power, device length, magnetic configuration and transport model were studied with B2.5-Eirene. Effects of tilted or vertical targets were calculated with EMC3-Eirene and showed that spreading the incident flux over a larger area leads to lower density, higher temperature and off-axis profile peaking in front of themore » target. In conclusion, the simulations indicate that with sufficient heating power MPEX can reach target plasma conditions that are similar to those expected in the ITER divertor. B2.5-Eirene simulations of the MAGPIE experiment have been carried out in order to establish an additional benchmark with experimental data from a linear device with helicon wave heating.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Skovorodin, D. I., E-mail: dskovorodin@gmail.com; Arakcheev, A. S.; Pshenov, A. A.

2016-02-15

The erosion of divertor targets caused by high heat fluxes during transients is a serious threat to ITER operation, as it is going to be the main factor determining the divertor lifetime. Under the influence of extreme heat fluxes, the surface temperature of plasma facing components can reach some certain threshold, leading to an onset of intense material evaporation. The latter results in formation of cold dense vapor and secondary plasma cloud. This layer effectively absorbs the energy of the incident plasma flow, turning it into its own kinetic and internal energy and radiating it. This so called vapor shieldingmore » is a phenomenon that may help mitigating the erosion during transient events. In particular, the vapor shielding results in saturation of energy (per unit surface area) accumulated by the target during single pulse of heat load at some level E{sub max}. Matching this value is one of the possible tests to verify complicated numerical codes, developed to calculate the erosion rate during abnormal events in tokamaks. The paper presents three very different models of vapor shielding, demonstrating that E{sub max} depends strongly on the heat pulse duration, thermodynamic properties, and evaporation energy of the irradiated target material. While its dependence on the other shielding details such as radiation capabilities of material and dynamics of the vapor cloud is logarithmically weak. The reason for this is a strong (exponential) dependence of the target material evaporation rate, and therefore the “strength” of vapor shield on the target surface temperature. As a result, the influence of the vapor shielding phenomena details, such as radiation transport in the vapor cloud and evaporated material dynamics, on the E{sub max} is virtually completely masked by the strong dependence of the evaporation rate on the target surface temperature. However, the very same details define the amount of evaporated particles, needed to provide an effective shielding to the target, and, therefore, strongly influence resulting erosion rate. Thus, E{sub max} cannot be used for validation of shielding models and codes, aimed at the target material erosion calculations.« less

Snowflake divertor configuration studies for NSTX-Upgrade

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 controlmore » 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.« less

2D imaging of helium ion velocity in the DIII-D divertor

NASA Astrophysics Data System (ADS)

Samuell, C. M.; Porter, G. D.; Meyer, W. H.; Rognlien, T. D.; Allen, S. L.; Briesemeister, A.; Mclean, A. G.; Zeng, L.; Jaervinen, A. E.; Howard, J.

2018-05-01

Two-dimensional imaging of parallel ion velocities is compared to fluid modeling simulations to understand the role of ions in determining divertor conditions and benchmark the UEDGE fluid modeling code. Pure helium discharges are used so that spectroscopic He+ measurements represent the main-ion population at small electron temperatures. Electron temperatures and densities in the divertor match simulated values to within about 20%-30%, establishing the experiment/model match as being at least as good as those normally obtained in the more regularly simulated deuterium plasmas. He+ brightness (HeII) comparison indicates that the degree of detachment is captured well by UEDGE, principally due to the inclusion of E ×B drifts. Tomographically inverted Coherence Imaging Spectroscopy measurements are used to determine the He+ parallel velocities which display excellent agreement between the model and the experiment near the divertor target where He+ is predicted to be the main-ion species and where electron-dominated physics dictates the parallel momentum balance. Upstream near the X-point where He+ is a minority species and ion-dominated physics plays a more important role, there is an underestimation of the flow velocity magnitude by a factor of 2-3. These results indicate that more effort is required to be able to correctly predict ion momentum in these challenging regimes.

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

PubMed

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

2012-10-01

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

Ideal plasma response to vacuum magnetic fields with resonant magnetic perturbations in non-axisymmetric tokamaks

DOE PAGES

Kim, Kimin; Ahn, J. -W.; Scotti, F.; ...

2015-09-03

Ideal plasma shielding and amplification of resonant magnetic perturbations in non-axisymmetric tokamak is presented by field line tracing simulation with full ideal plasma response, compared to measurements of divertor lobe structures. Magnetic field line tracing simulations in NSTX with toroidal non-axisymmetry indicate the ideal plasma response can significantly shield/amplify and phase shift the vacuum resonant magnetic perturbations. Ideal plasma shielding for n = 3 mode is found to prevent magnetic islands from opening as consistently shown in the field line connection length profile and magnetic footprints on the divertor target. It is also found that the ideal plasma shielding modifiesmore » the degree of stochasticity but does not change the overall helical lobe structures of the vacuum field for n = 3. Furthermore, amplification of vacuum fields by the ideal plasma response is predicted for low toroidal mode n = 1, better reproducing measurements of strong striation of the field lines on the divertor plate in NSTX.« less

Bohm criterion and plasma particle/power exhaust to and recycling at the wall

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tang, Xianzhu; Guo, Zehua

The plasma particle and power exhaust to the divertor surface drives both particle and power recycling at the surface, which in return constrains the plasma density and temperature at the target and their profile further upstream. Both particle and power exhaust fluxes are mediated by the plasma sheath next to the divertor surface. In particular, the Bohm criterion constrains the ion exit flow speed, which enters directly into the particle flux and the kinetic flow energy component of the ion power flux, and indirectly into the electron power flux through the sheath potential drop. Here we give an overview onmore » how the Bohm speed is set in a general plasma and how it enters power exhaust and power recycling at the divertor surface, and the implication on the correct implementation of sheath boundary conditions in numerical codes. The cases of ideal and non-ideal Bohm speed are distinguished as a result of the physics discussion.« less

Bohm criterion and plasma particle/power exhaust to and recycling at the wall

DOE PAGES

Tang, Xianzhu; Guo, Zehua

2017-06-07

The plasma particle and power exhaust to the divertor surface drives both particle and power recycling at the surface, which in return constrains the plasma density and temperature at the target and their profile further upstream. Both particle and power exhaust fluxes are mediated by the plasma sheath next to the divertor surface. In particular, the Bohm criterion constrains the ion exit flow speed, which enters directly into the particle flux and the kinetic flow energy component of the ion power flux, and indirectly into the electron power flux through the sheath potential drop. Here we give an overview onmore » how the Bohm speed is set in a general plasma and how it enters power exhaust and power recycling at the divertor surface, and the implication on the correct implementation of sheath boundary conditions in numerical codes. The cases of ideal and non-ideal Bohm speed are distinguished as a result of the physics discussion.« less

Component with inspection-facilitating features

DOEpatents

Marra, John J; Zombo, Paul J

2014-02-11

A turbine airfoil can be formed with features to facilitate measurement of its wall thickness. An outer wall of the airfoil can include an outer surface and an inner surface. The outer surface of the airfoil can have an outer inspection target surface, and the inner surface of the airfoil can have an inner inspection target surface. The inner and outer target surfaces can define substantially flat regions in surfaces that are otherwise highly contoured. The inner and outer inspection target surfaces can be substantially aligned with each other. The inner and outer target surfaces can be substantially parallel to each other. As a result of these arrangements, a highly accurate measurement of wall thickness can be obtained. In one embodiment, the outer inspection target surface can be defined by an innermost surface of a groove formed in the outer surface of the outer wall of the airfoil.

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soukhanovskii, V. A.

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

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

NASA Astrophysics Data System (ADS)

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

2013-10-01

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

ERO modelling of tungsten erosion and re-deposition in EAST L mode discharges

NASA Astrophysics Data System (ADS)

Xie, H.; Ding, R.; Kirschner, A.; Chen, J. L.; Ding, F.; Mao, H. M.; Feng, W.; Borodin, D.; Wang, L.

2017-09-01

Tungsten erosion and re-deposition at the upper outer divertor of the Experimental Advanced Superconducting Tokamak has been modelled using the 3D Monte Carlo code ERO. The measured divertor plasma condition in attached L mode discharges with upper single null configuration has been used to build the background plasma in the simulations. The tungsten gross erosion rate is mainly determined by carbon impurity in the background plasma. Increasing carbon concentration can first increase and afterwards suppress the tungsten erosion rate. Taking into account the material mixing surface model, the influence of eroded particles returning to the surface on sputtering has been studied. Sputtering by eroded particles returning to the surface can significantly enhance the gross erosion by reduction of the carbon ratio within the surface interaction layer and by increasing the erosion rate due to sputtering by both eroded tungsten and carbon particles. Modelling indicates that carbon deposition occurs on the dome plate and part of the vertical plate close to the dome plate, whereas tungsten net erosion occurs on most of the vertical plate. The modelling results are in reasonable agreement with the experimental WI spectroscopy.

Plasma detachment in divertor tokamaks

NASA Astrophysics Data System (ADS)

Leonard, A. W.

2018-04-01

Observations of divertor plasma detachment in tokamaks are reviewed. Plasma detachment is characterized in terms of transport and dissipation of power, momentum and particle flux along the open field lines from the midplane to the divertor. Asymmetries in detachment onset and other characteristics between the inboard and outboard divertor plasmas is found to be primarily driven by plasma E× B drifts. The effect of divertor plate geometry and magnetic configuration on divertor detachment is summarized. Control of divertor detachment has progressed with a development of a number of diagnostics to characterize the detached state in real-time. Finally the compatibility of detached divertor operation with high performance core plasmas is examined.

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

NASA Astrophysics Data System (ADS)

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

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 trackings 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.

Plasma detachment in divertor tokamaks

DOE PAGES

Leonard, A. W.

2018-02-07

In this study, observations of divertor plasma detachment in tokamaks are reviewed. Plasma detachment is characterized in terms of transport and dissipation of power, momentum and particle flux along the open field lines from the midplane to the divertor. Asymmetries in detachment onset and other characteristics between the inboard and outboard divertor plasmas is found to be primarily driven by plasmamore » $$\\vec{E}$$ x $$\\vec{B}$$ drifts. The effect of divertor plate geometry and magnetic configuration on divertor detachment is summarized. Control of divertor detachment has progressed with a development of a number of diagnostics to characterize the detached state in real-time. Finally the compatibility of detached divertor operation with high performance core plasmas is examined.« less

Plasma detachment in divertor tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Leonard, A. W.

In this study, observations of divertor plasma detachment in tokamaks are reviewed. Plasma detachment is characterized in terms of transport and dissipation of power, momentum and particle flux along the open field lines from the midplane to the divertor. Asymmetries in detachment onset and other characteristics between the inboard and outboard divertor plasmas is found to be primarily driven by plasmamore » $$\\vec{E}$$ x $$\\vec{B}$$ drifts. The effect of divertor plate geometry and magnetic configuration on divertor detachment is summarized. Control of divertor detachment has progressed with a development of a number of diagnostics to characterize the detached state in real-time. Finally the compatibility of detached divertor operation with high performance core plasmas is examined.« less

Divertor scenario development for NSTX Upgrade

NASA Astrophysics Data System (ADS)

Soukhanovskii, V. A.; McLean, A. G.; Meier, E. T.; Rognlien, T. D.; Ryutov, D. D.; Bell, R. E.; Diallo, A.; Gerhardt, S. P.; Kaita, R.; Kolemen, E.; Leblanc, B. P.; Menard, J. E.; Podesta, M.; Scotti, F.

2012-10-01

In the NSTX-U tokamak, initial plans for divertor plasma-facing components (PFCs) include lithium and boron coated graphite, with a staged transition to molybdenum. Steady-state peak divertor heat fluxes are projected to reach 20-30 MW/m^2 in 2 MA, 12 MW NBI-heated discharges of up to 5 s duration, thus challenging PFC thermal limits. Based on the recent NSTX divertor experiments and modeling with edge transport code UEDGE, a favorable basis for divertor power handling in NSTX-U is developed. The snowflake divertor geometry and feedback-controlled divertor impurity seeding applied to the lower and upper divertors are presently envisioned. In the NSTX snowflake experiments with lithium-coated graphite PFCs, the peak divertor heat fluxes from Type I ELMs and between ELMs were significantly reduced due to geometry effects, increased volumetric losses and null-point convective redistribution between strike points. H-mode core confinement was maintained at H98(y,2)<=1 albeit the radiative detachment. Additional CD4 seeding demonstrated potential for a further increase of divertor radiation.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reichle, R.; Beaumont, B.; Boilson, D.

2012-10-15

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

Development and application of W/Cu flat-type plasma facing components at ASIPP

NASA Astrophysics Data System (ADS)

Li, Q.; Zhao, S. X.; Sun, Z. X.; Xu, Y.; Li, B.; Wei, R.; Wang, W. J.; Qin, S. G.; Shi, Y. L.; Xie, C. Y.; Wang, J. C.; Wang, X. L.; Missirlian, M.; Guilhem, D.; Liu, G. H.; Yang, Z. S.; Luo, G.-N.

2017-12-01

W/Cu flat-type plasma facing components (PFCs) were widely used in divertor of fusion device because of its advantages, such as low cost, light in weight and good machinability. However, it is very difficult to manufacture them due to the large mismatch between the thermo-mechanical properties of W and Cu. Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) has successfully developed W/Cu flat-type PFCs for EAST W/Cu divertor project by hot isostatic pressing (HIP) technology. This paper presents the development and application of W/Cu flat-type PFCs at ASIPP. The optimized manufacturing process is to cast pure copper onto the rear side of W tiles at temperature of 1200 °C firstly, and then to HIP the W/Cu tiles onto CuCrZr heat sink at temperature of 600 °C, pressure of 150 MPa and duration of 3 h. W/Cu flat-type testing mock-up for EAST survived 1000 cycles at heat load of 5 MW m-2 in high heat flux tests. And then ASIPP prepared two mock-ups for CEA’s tungsten environment in steady-state tokamak (WEST) project. One mock-up withstood successfully 302 cycles of 20 MW m-2, which are far beyond the design requirement. Since 2014, W/Cu flat-type PFCs were wildly used in EAST upper divertor as baffle and dome components which showed excellent performance in 2015 and 2016 campaigns. Given the success in EAST upper divertor, W/Cu flat-type concept is as well applied in the design of actively cooled Langmuir probes which will be mounted onto EAST divertor targets soon.

A new scaling for divertor detachment

NASA Astrophysics Data System (ADS)

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

2017-05-01

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 P sep B/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-like 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.

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

Estimation of the tritium retention in ITER tungsten divertor target using macroscopic rate equations simulations

NASA Astrophysics Data System (ADS)

Hodille, E. A.; Bernard, E.; Markelj, S.; Mougenot, J.; Becquart, C. S.; Bisson, R.; Grisolia, C.

2017-12-01

Based on macroscopic rate equation simulations of tritium migration in an actively cooled tungsten (W) plasma facing component (PFC) using the code MHIMS (migration of hydrogen isotopes in metals), an estimation has been made of the tritium retention in ITER W divertor target during a non-uniform exponential distribution of particle fluxes. Two grades of materials are considered to be exposed to tritium ions: an undamaged W and a damaged W exposed to fast fusion neutrons. Due to strong temperature gradient in the PFC, Soret effect’s impacts on tritium retention is also evaluated for both cases. Thanks to the simulation, the evolutions of the tritium retention and the tritium migration depth are obtained as a function of the implanted flux and the number of cycles. From these evolutions, extrapolation laws are built to estimate the number of cycles needed for tritium to permeate from the implantation zone to the cooled surface and to quantify the corresponding retention of tritium throughout the W PFC.

Numerical exploration of non-axisymmetric divertor closure in the small angle slot (SAS) divertor at DIII-D

NASA Astrophysics Data System (ADS)

Frerichs, Heinke; Schmitz, Oliver; Covele, Brent; Guo, Houyang; Hill, David; Feng, Yuhe

2017-10-01

In the Small Angle Slot (SAS) divertor in DIII-D, the combination of misaligned slot structure and non-axisymmetric perturbations to the magnetic field causes the strike point to vary radially along the divertor slot and even leave it at some toroidal locations. This effect essentially introduces an opening in the divertor slot from where recycling neutrals can easily escape, and thereby degrade performance of the slot divertor. This effect has been approximated by a finite gap in the divertor baffle. Simulations with EMC3-EIRENE show that a toroidally localized loss of divertor closure can result in non-axisymmetric divertor densities and temperatures. This introduces a density window of 10-15% on top of the nominal threshold separatrix density during which a non-axisymmetric onset of local detachment occurs, initially leaving the gap and up to 60 deg beyond that still attached. Conversely, the impact of such toroidally localized divertor perturbations on the toroidal symmetry of midplane separatrix conditions is small. This work has been funded by the U.S. Department of Energy under Early Career Award Grant DE-SC0013911, and Grant DE-FC02-04ER54698.

Comparative divertor-transport study for helical devices

NASA Astrophysics Data System (ADS)

Feng, Y.; Kobayashi, M.; Sardei, F.; Masuzaki, S.; Kisslinger, J.; Morisaki, T.; Grigull, P.; Yamada, H.; McCormick, K.; Ohyabu, N.; König, R.; Yamada, I.; Giannone, L.; Narihara, K.; Wenzel, U.; Morita, S.; Thomsen, H.; Miyazawa, J.; Hildebrandt, D.; Watanabe, T.; Wagner, F.; Ashikawa, N.; Ida, K.; Komori, A.; Motojima, O.; Nakamura, Y.; Peterson, B. J.; Sato, K.; Shoji, M.; Tamura, N.; Tokitani, M.; LHD experimental Group

2009-09-01

Using the island divertors (IDs) of W7-AS and W7-X and the helical divertor (HD) of LHD as examples, the paper presents a comparative divertor transport study for three typical helical devices of different machine sizes following two distinct divertor concepts, aiming at identifying common physics issues/effects for mutual validation and combined studies. Based on EMC3/EIRENE simulations supported by experimental results, the paper first reviews and compares the essential transport features of the W7-AS ID and the LHD HD in order to build a base and framework for a predictive study of W7-X. The fundamental role of low-order magnetic islands in both divertor concepts is emphasized. Preliminary EMC3/EIRENE simulation results for W7-X are presented and discussed with respect to W7-AS and LHD in order to show how the individual field and divertor topologies affect the divertor transport and performance. For instance, a high recycling regime, which is absent from W7-AS and LHD, is predicted to exist for W7-X. The paper focuses on identifying and understanding the role of divertors for high density plasma operations in helical devices. In this regard, special attention is paid to investigating the divertor function for controlling intrinsic impurities. Impurity transport behaviour and wall-sputtering processes of CX-neutrals are studied under different divertor plasma conditions. A divertor retention effect on intrinsic impurities at high SOL collisonalities is predicted for all the three devices. The required SOL plasma conditions and the underlying mechanisms are analysed in detail. Numerical results are discussed in conjunction with the experimental observations for high density divertor plasmas in W7-AS and LHD. Different SOL transport regimes are numerically identified for the standard divertor configuration of W7-X and the possible consequences on high density plasmas are assessed. All the EMC3-EIRENE simulations presented in this paper are based on vacuum fields and comparisons with local diagnostics are made for low-ß plasmas.

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

NASA Astrophysics Data System (ADS)

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

2015-05-01

The MIT Plasma Science and Fusion Center and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX)—a tokamak specifically designed to address critical gaps in the world fusion research programme on the pathway to next-step devices: fusion nuclear science facility (FNSF), fusion pilot plant (FPP) and/or demonstration power plant (DEMO). This high-field (⩾6.5 T, 1.5 MA), high power density facility (P/S ˜ 1.5 MW m-2) will test innovative divertor ideas, including an ‘X-point target divertor’ concept, at the required performance parameters—reactor-level boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region—while simultaneously producing high-performance core plasma conditions that are prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fuelling from external heating and current drive systems. Equally important, the experimental platform will test innovative concepts for lower hybrid current drive and ion cyclotron range of frequency actuators with the unprecedented ability to deploy launch structures both on the low-magnetic-field side and the high-magnetic-field side—the latter being a location where energetic plasma-material interactions can be controlled and favourable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination—advanced divertors, advanced RF actuators, reactor-prototypical core plasma conditions—will enable ADX to explore enhanced core confinement physics, such as made possible by reversed central shear, using only the types of external drive systems that are considered viable for a fusion power plant. Such an integrated demonstration of high-performance core-divertor operation with steady-state sustainment would pave the way towards an attractive pilot plant, as envisioned in the ARC concept (affordable, robust, compact) (Sorbom et al 2015 Fusion Eng. Des. submitted (arXiv:1409.3540)) that makes use of high-temperature superconductor technology—a high-field (9.25 T) tokamak the size of the Joint European Torus that produces 270 MW of net electricity.

Development of Radiated Power Diagnostics for NSTX-U

NASA Astrophysics Data System (ADS)

Reinke, Matthew; van Eden, G. G.; Lovell, Jack; Peterson, Byron; Gray, Travis; Chandra, Rian; Stratton, Brent; Ellis, Robert; NSTX-U Team

2016-10-01

New tools to measure radiated power in NSTX-U are under development to support a range of core and boundary physics research. Multiple resistive bolometer pinhole cameras are being built and calibrated to support FY17 operations, all utilizing standard Au-foil sensors from IPT-Albrecht. The radiation in the lower divertor will be measured using two, 8 channel arrays viewing both vertically and radially to enable estimates of the 2D radiation structure. The core radiation will be measured using a 24 channel array viewing tangentially near the midplane, observing the full cross-section from the inner to outer limiter. This enables characterization of the centrifugally-driven in/out radiation asymmetry expected from mid-Z and high-Z impurities in highly rotating NSTX-U plasmas. All sensors utilize novel FPGA-based BOLO8BLF analyzers from D-tAcq Solutions. Resistive bolometer measurements are complemented by an InfraRed Video Bolometer (IRVB) which measures the temperature change of radiation absorber using an IR camera. A prototype IRVB system viewing the lower divertor was installed on NSTX-U for FY16 operations. Initial results from the plasma and benchtop testing are used to demonstrate the relative advantages between IRVB and resistive bolometers. Supported in Part by DE-AC05-00OR22725 & DE-AC02-09CH11466.

Development and qualification of a bulk tungsten divertor row for JET

NASA Astrophysics Data System (ADS)

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

2009-06-01

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

Understanding tungsten divertor sourcing and SOL transport using multiple poloidally-localized sources in DIII-D ELM-y H-mode discharges

NASA Astrophysics Data System (ADS)

Unterberg, Ea; Donovan, D.; Barton, J.; Wampler, Wr; Abrams, T.; Thomas, Dm; Petrie, T.; Guo, Hy; Stangeby, Pg; Elder, Jd; Rudakov, D.; Grierson, B.; Victor, B.

2017-10-01

Experiments using metal inserts with novel isotopically-enriched tungsten coatings at the outer divertor strike point (OSP) have provided unique insight into the ELM-induced sourcing, main-SOL transport, and core accumulation control mechanisms of W for a range of operating conditions. This experimental approach has used a multi-head, dual-facing collector probe (CP) at the outboard midplane, as well as W-I and core W spectroscopy. Using the CP system, the total amount of W deposited relative to source measurements shows a clear dependence on ELM size, ELM frequency, and strike point location, with large ELMs depositing significantly more W on the CP from the far-SOL source. Additionally, high spatial ( 1mm) and ELM resolved spectroscopic measurements of W sourcing indicate shifts in the peak erosion rate. Furthermore, high performance discharges with rapid ELMs show core W concentrations of few 10-5, and the CP deposition profile indicates W is predominantly transported to the midplane from the OSP rather than from the far-SOL region. The low central W concentration is shown to be due to flattening of the main plasma density profile, presumably by on-axis electron cyclotron heating. Work supported under USDOE Cooperative Agreement DE-FC02-04ER54698.

A review of direct experimental measurements of detachment

DOE PAGES

Boedo, J.; McLean, A. G.; Rudakov, D. L.; ...

2018-02-22

Detached divertor plasmas feature strong radial and parallel gradients of density, temperature, electric fields and flow over the divertor volume and therefore, sampling the divertor plasma directly provides crucial knowledge to the interpretation and modeling efforts. Here, we review the contribution of diagnostics that directly sample the plasma to the advancement of knowledge of the physics of detachment and detached divertors, such as the characteristics of the various regimes, discovery and quantification of drifts and identification of convection of heat and particles. We focus on wall probes, scanning probes, retarding field analyzers and Thomson Scattering (TS) in the divertor regionmore » and also include the contribution of measurements away from the divertor that provide insight on how divertor detachment affects core, edge or pedestal conditions. Wall probes are critical as they can be installed in closed volumes of difficult access to other diagnostics and measure plasma parameters at the divertor structures, which define the plasma boundary conditions and where detachment effects are more likely to be strongest.« less

A review of direct experimental measurements of detachment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boedo, J.; McLean, A. G.; Rudakov, D. L.

Detached divertor plasmas feature strong radial and parallel gradients of density, temperature, electric fields and flow over the divertor volume and therefore, sampling the divertor plasma directly provides crucial knowledge to the interpretation and modeling efforts. Here, we review the contribution of diagnostics that directly sample the plasma to the advancement of knowledge of the physics of detachment and detached divertors, such as the characteristics of the various regimes, discovery and quantification of drifts and identification of convection of heat and particles. We focus on wall probes, scanning probes, retarding field analyzers and Thomson Scattering (TS) in the divertor regionmore » and also include the contribution of measurements away from the divertor that provide insight on how divertor detachment affects core, edge or pedestal conditions. Wall probes are critical as they can be installed in closed volumes of difficult access to other diagnostics and measure plasma parameters at the divertor structures, which define the plasma boundary conditions and where detachment effects are more likely to be strongest.« less

A super-cusp divertor configuration for tokamaks

NASA Astrophysics Data System (ADS)

Ryutov, D. D.

2015-10-01

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

A review of direct experimental measurements of detachment

NASA Astrophysics Data System (ADS)

Boedo, J.; McLean, A. G.; Rudakov, D. L.; Watkins, J. G.

2018-04-01

Detached divertor plasmas feature strong radial and parallel gradients of density, temperature, electric fields and flow over the divertor volume and therefore, sampling the divertor plasma directly provides crucial knowledge to the interpretation and modeling efforts. We review the contribution of diagnostics that directly sample the plasma to the advancement of knowledge of the physics of detachment and detached divertors, such as the characteristics of the various regimes, discovery and quantification of drifts and identification of convection of heat and particles. We focus on wall probes, scanning probes, retarding field analyzers and Thomson scattering in the divertor region and also include the contribution of measurements away from the divertor that provide insight on how divertor detachment affects core, edge or pedestal conditions. Wall probes are critical as they can be installed in closed volumes of difficult access to other diagnostics and measure plasma parameters at the divertor structures, which define the plasma boundary conditions and where detachment effects are more likely to be strongest.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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

Advantages and Challenges of Radiative Liquid Lithium Divertor

NASA Astrophysics Data System (ADS)

Ono, Masayuki

2017-10-01

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

Spectroscopic determinations of carbon fluxes, sources, and shielding in the DIII-D divertors

NASA Astrophysics Data System (ADS)

Isler, R. C.; Colchin, R. J.; Brooks, N. H.; Evans, T. E.; West, W. P.; Whyte, D. G.

2001-10-01

The most important mechanisms for eroding plasma-facing components (PFCs) and introducing carbon into tokamak divertors are believed to be physical sputtering, chemical sputtering, sublimation, and radiation enhance sublimation (RES). The relative importance of these processes has been investigated by analyzing the spectral emission rates and the effective temperatures of CI, CD, and C2 under several operating conditions in the DIII-D tokamak [Plasma Physics Controlled Nuclear Fusion Research, 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159; Proceedings of the 18th IEEE/NPSS Symposium on Fusion Engineering, Albuquerque (Institute of Electrical and Electronic Engineers, Piscataway, 1999), p. 515]. Discrimination of chemical sputtering from physical sputtering is accomplished by quantitatively relating the fraction of CI influxes expected from dissociation of hydrocarbons to the measured CD and C2 influxes. Characteristics of sublimation are studied from carbon test samples heated to surface temperatures exceeding 2000 K. The shielding efficiency of carbon produced at the divertor target is assessed from comparison of fluxes of neutral atoms and ions; approximately 95% of the primary influx appears to be redeposited before being transported far enough upstream to fuel the core plasma.

Spectroscopic determinations of carbon fluxes, sources, and shielding in the DIII-D divertors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Isler, R. C.; Colchin, R. J.; Brooks, N. H.

2001-10-01

The most important mechanisms for eroding plasma-facing components (PFCs) and introducing carbon into tokamak divertors are believed to be physical sputtering, chemical sputtering, sublimation, and radiation enhance sublimation (RES). The relative importance of these processes has been investigated by analyzing the spectral emission rates and the effective temperatures of CI, CD, and C{sub 2} under several operating conditions in the DIII-D tokamak [Plasma Physics Controlled Nuclear Fusion Research, 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159; Proceedings of the 18th IEEE/NPSS Symposium on Fusion Engineering, Albuquerque (Institute of Electrical and Electronic Engineers, Piscataway, 1999), p. 515]. Discriminationmore » of chemical sputtering from physical sputtering is accomplished by quantitatively relating the fraction of CI influxes expected from dissociation of hydrocarbons to the measured CD and C{sub 2} influxes. Characteristics of sublimation are studied from carbon test samples heated to surface temperatures exceeding 2000 K. The shielding efficiency of carbon produced at the divertor target is assessed from comparison of fluxes of neutral atoms and ions; approximately 95% of the primary influx appears to be redeposited before being transported far enough upstream to fuel the core plasma.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

1995-09-01

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

DiMES Tests of W Leading Edge Power Loading in DIII-D

NASA Astrophysics Data System (ADS)

Nygren, R. E.; Watkins, J. G.; Rudakov, D. L.; Lasnier, C. J.; Pitts, R. A.; Stangeby, P. C.

2015-11-01

In a transient melt experiment in JET, the power to a ~1-mm-high leading edge on a W lamella in the bulk-W outer divertor was lower than expected from the geometry by factors of 5 and 2 for L-mode and H-mode discharges, respectively. We checked this surprising result in DIII-D with 3 W blocks (10 mm square) mounted radially side-by-side in DiMES with leading edges of 0.0, 0.3, 1.0 mm, single null L-mode plasmas, OSP just outside ``0.0'' block, limited scans (NBI+ECH), B-field incident at 1.5° or 2.5°, and viewed, as in JET, from above with 0.2mm/pixel resolution IRTV. Langmuir probes measured parallel power to the target. We compared probe and IR data with a detailed thermal model of the blocks and concluded provisionally that we did not reproduce the power deficit found in JET. Blurred IR images complicated fitting of temperature distributions from the thermal model. We plan an experiment with both L- and H-mode He plasmas before the APS meeting. Supported by US DOE under DE-AC04-94AL85000, 44500007360, DE-AC52-07NA27344, and DE-FC02-04ER54698.

The ARIES Advanced and Conservative Tokamak Power Plant Study

DOE PAGES

Kessel, C. E; Tillak, M. S; Najmabadi, F.; ...

2015-12-22

Tokamak power plants are studied with advanced and conservative design philosophies to identify the impacts on the resulting designs and to provide guidance to critical research needs. Incorporating updated physics understanding and using more sophisticated engineering and physics analysis, the tokamak configurations have developed a more credible basis compared with older studies. The advanced configuration assumes a self-cooled lead lithium blanket concept with SiC composite structural material with 58% thermal conversion efficiency. This plasma has a major radius of 6.25 m, a toroidal field of 6.0 T, a q₉₅ of 4.5, aᵦ total N of 5.75, an H98 of 1.65,more » an n/n Gr of 1.0, and a peak divertor heat flux of 13.7 MW/m² . The conservative configuration assumes a dual-coolant lead lithium blanket concept with reduced activation ferritic martensitic steel structural material and helium coolant, achieving a thermal conversion efficiency of 45%. The plasma has a major radius of 9.75 m, a toroidal field of 8.75 T, a q₉₅ of 8.0, aᵦ total N of 2.5, an H₉₈ of 1.25, an n/n Gr of 1.3, and a peak divertor heat flux of 10 MW/m² . The divertor heat flux treatment with a narrow power scrape off width has driven the plasmas to larger major radius. Edge and divertor plasma simulations are targeting a basis for high radiated power fraction in the divertor, which is necessary for solutions to keep the peak heat flux in the range 10 to 15 MW/m² . Combinations of the advanced and conservative approaches show intermediate sizes. A new systems code using a database approach has been used and shows that the operating point is really an operating zone with some range of plasma and engineering parameters and very similar costs of electricity. Other papers in this issue provide more detailed discussion of the work summarized here.« less

The ARIES Advanced And Conservative Tokamak (ACT) Power Plant Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kessel, C. E.; Poli, F. M.; Ghantous, K.

2014-03-05

Tokamak power plants are studied with advanced and conservative design philosophies in order to identify the impacts on the resulting designs and to provide guidance to critical research needs. Incorporating updated physics understanding, and using more sophisticated engineering and physics analysis, the tokamak configurations have developed a more credible basis compared to older studies. The advanced configuration assumes a self-cooled lead lithium (SCLL) blanket concept with SiC composite structural material with 58% thermal conversion efficiency. This plasma has a major radius of 6.25 m, a toroidal field of 6.0 T, a q95 of 4.5, a βN total of 5.75, Hmore » 98 of 1.65, n/nGr of 1.0, and peak divertor heat flux of 13.7 MW/m 2. The conservative configuration assumes a dual coolant lead lithium (DCLL) blanket concept with ferritic steel structural material and helium coolant, achieving a thermal conversion efficiency of 45%. The plasma major radius is 9.75 m, a toroidal field of 8.75 T, a q95 of 8.0, a βN total of 2.5, H 98 of 1.25, n/n Gr of 1.3, and peak divertor heat flux of 10 MW/m 2. The divertor heat flux treatment with a narrow power scrape-off width has driven the plasmas to larger major radius. Edge and divertor plasma simulations are targeting a basis for high radiated power fraction in the divertor, which is necessary for solutions to keep the peak heat flux in the range of 10-15 MW/m 2. Combinations of the advanced and conservative approaches show intermediate sizes. A new systems code using a database approach has been used and shows that the operating point is really an operating zone with some range of plasma and engineering parameters and very similar costs of electricity. Papers in this issue provide more detailed discussion of the work summarized here.« less

Vacuum System and Modeling for the Materials Plasma Exposure Experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lumsdaine, Arnold; Meitner, Steve; Graves, Van

Understanding the science of plasma-material interactions (PMI) is essential for the future development of fusion facilities. The design of divertors and first walls for the next generation of long-pulse fusion facilities, such as a Fusion Nuclear Science Facility (FNSF) or a DEMO, requires significant PMI research and development. In order to meet this need, a new linear plasma facility, the Materials Plasma Exposure Experiment (MPEX) is proposed, which will produce divertor relevant plasma conditions for these next generation facilities. The device will be capable of handling low activation irradiated samples and be able to remove and replace samples without breakingmore » vacuum. A Target Exchange Chamber (TEC) which can be disconnected from the high field environment in order to perform in-situ diagnostics is planned for the facility as well. The vacuum system for MPEX must be carefully designed in order to meet the requirements of the different heating systems, and to provide conditions at the target similar to those expected in a divertor. An automated coupling-decoupling (“autocoupler”) system is designed to create a high vacuum seal, and will allow the TEC to be disconnected without breaking vacuum in either the TEC or the primary plasma materials interaction chamber. This autocoupler, which can be actuated remotely in the presence of the high magnetic fields, has been designed and prototyped, and shows robustness in a variety of conditions. The vacuum system has been modeled using a simplified finite element analysis, and indicates that the design goals for the pressures in key regions of the facility are achievable.« less

Vacuum System and Modeling for the Materials Plasma Exposure Experiment

DOE PAGES

Lumsdaine, Arnold; Meitner, Steve; Graves, Van; ...

2017-08-07

Understanding the science of plasma-material interactions (PMI) is essential for the future development of fusion facilities. The design of divertors and first walls for the next generation of long-pulse fusion facilities, such as a Fusion Nuclear Science Facility (FNSF) or a DEMO, requires significant PMI research and development. In order to meet this need, a new linear plasma facility, the Materials Plasma Exposure Experiment (MPEX) is proposed, which will produce divertor relevant plasma conditions for these next generation facilities. The device will be capable of handling low activation irradiated samples and be able to remove and replace samples without breakingmore » vacuum. A Target Exchange Chamber (TEC) which can be disconnected from the high field environment in order to perform in-situ diagnostics is planned for the facility as well. The vacuum system for MPEX must be carefully designed in order to meet the requirements of the different heating systems, and to provide conditions at the target similar to those expected in a divertor. An automated coupling-decoupling (“autocoupler”) system is designed to create a high vacuum seal, and will allow the TEC to be disconnected without breaking vacuum in either the TEC or the primary plasma materials interaction chamber. This autocoupler, which can be actuated remotely in the presence of the high magnetic fields, has been designed and prototyped, and shows robustness in a variety of conditions. The vacuum system has been modeled using a simplified finite element analysis, and indicates that the design goals for the pressures in key regions of the facility are achievable.« less

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.

A super-cusp divertor configuration for tokamaks

DOE PAGES

Ryutov, D. D.

2015-08-26

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

On heat loading, novel divertors, and fusion reactors

NASA Astrophysics Data System (ADS)

Kotschenreuther, M.; Valanju, P. M.; Mahajan, S. M.; Wiley, J. C.

2007-07-01

The limited thermal power handling capacity of the standard divertors (used in current as well as projected tokamaks) is likely to force extremely high (˜90%) radiation fractions frad in tokamak fusion reactors that have heating powers considerably larger than ITER [D. J. Campbell, Phys. Plasmas 8, 2041 (2001)]. Such enormous values of necessary frad could have serious and debilitating consequences on the core confinement, stability, and dependability for a fusion power reactor, especially in reactors with Internal Transport Barriers. A new class of divertors, called X-divertors (XD), which considerably enhance the divertor thermal capacity through a flaring of the field lines only near the divertor plates, may be necessary and sufficient to overcome these problems and lead to a dependable fusion power reactor with acceptable economics. X-divertors will lower the bar on the necessary confinement to bring it in the range of the present experimental results. Its ability to reduce the radiative burden imparts the X-divertor with a key advantage. Lower radiation demands allow sharply peaked density profiles that enhance the bootstrap fraction creating the possibility for a highly increased beta for the same beta normal discharges. The X-divertor emerges as a beta-enhancer capable of raising it by up to roughly a factor of 2.

The flush-mounted rail Langmuir probe array designed for the Alcator C-Mod vertical target plate divertor

NASA Astrophysics Data System (ADS)

Kuang, A. Q.; Brunner, D.; LaBombard, B.; Leccacorvi, R.; Vieira, R.

2018-04-01

An array of flush-mounted and toroidally elongated Langmuir probes (henceforth called rail probes) have been specifically designed for the Alcator C-Mod's vertical target plate divertor and operated over multiple campaigns. The "flush" geometry enables the tungsten electrodes to survive high heat flux conditions in which traditional "proud" tungsten electrodes suffer damage from melting. The toroidally elongated rail-like geometry reduces the influence of sheath expansion, which is an important effect to consider in the design and interpretation of flush-mounted Langmuir probes. The new rail probes successfully operated during C-Mod's FY2015 and FY2016 experimental campaigns with no evidence of damage, despite being regularly subjected to heat flux densities parallel to the magnetic field exceeding ˜1 GW m-2 for short periods of time. A comparison between rail and proud probe data indicates that sheath expansion effects were successfully mitigated by the rail design, extending the use of these Langmuir probes to incident magnetic field line angles as low as 0.5°.

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.

Response to "Comment on `Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake' " [Phys. Plasmas 21, 054701 (2014)

NASA Astrophysics Data System (ADS)

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

2014-05-01

Relying on coil positions relative to the plasma, the "Comment on `Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake' " [Phys. Plasmas 21, 054701 (2014)], emphasizes a criterion for divertor characterization that was critiqued to be ill posed [M. Kotschenreuther et al., Phys. Plasmas 20, 102507 (2013)]. We find that no substantive physical differences flow from this criteria. However, using these criteria, the successful NSTX experiment by Ryutov et al. [Phys. Plasmas 21, 054701 (2014)] has the coil configuration of an X-divertor (XD), rather than a snowflake (SF). On completing the divertor index (DI) versus distance graph for this NSTX shot (which had an inexplicably missing region), we find that the DI is like an XD for most of the outboard wetted divertor plate. Further, the "proximity condition," used to define an SF [M. Kotschenreuther et al., Phys. Plasmas 20, 102507 (2013)], does not have a substantive physics basis to override metrics based on flux expansion and line length. Finally, if the criteria of the comment are important, then the results of NSTX-like experiments could have questionable applicability to reactors.

Theory of Advanced Magnetic Divertors

NASA Astrophysics Data System (ADS)

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

2013-10-01

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

GITR Simulation of Helium Exposed Tungsten Erosion and Redistribution in PISCES-A

NASA Astrophysics Data System (ADS)

Younkin, T. R.; Green, D. L.; Doerner, R. P.; Nishijima, D.; Drobny, J.; Canik, J. M.; Wirth, B. D.

2017-10-01

The extreme heat, charged particle, and neutron flux / fluence to plasma facing materials in magnetically confined fusion devices has motivated research to understand, predict, and mitigate the associated detrimental effects. Of relevance to the ITER divertor is the helium interaction with the tungsten divertor, the resulting erosion and migration of impurities. The linear plasma device PISCES A has performed dedicated experiments for high (4x10-22 m-2s-1) and low (4x10-21 m-2s-1) flux, 250 eV He exposed tungsten targets to assess the net and gross erosion of tungsten and volumetric transport. The temperature of the target was held between 400 and 600 degrees C. We present results of the erosion / migration / re-deposition of W during the experiment from the GITR (Global Impurity Transport) code coupled to materials response models. In particular, the modeled and experimental W I emission spectroscopy data for the 429.4 nm wavelength and net erosion through target and collector mass difference measurements are compared. Overall, the predictions are in good agreement with experiments. This material is supported by the US DOE, Office of Science, Office of Fusion Energy Sciences and Office of Advanced Scientific Computing Research through the SciDAC program on Plasma-Surface Interactions.

Drug discovery strategies to outer membrane targets in Gram-negative pathogens.

PubMed

Brown, Dean G

2016-12-15

This review will cover selected recent examples of drug discovery strategies which target the outer membrane (OM) of Gram-negative bacteria either by disruption of outer membrane function or by inhibition of essential gene products necessary for outer membrane assembly. Significant advances in pathway elucidation, structural biology and molecular inhibitor designs have created new opportunities for drug discovery within this target-class space. Copyright © 2016 Elsevier Ltd. All rights reserved.

ELM mitigation techniques

NASA Astrophysics Data System (ADS)

Evans, T. E.

2013-07-01

Large edge-localized mode (ELM) control techniques must be developed to help ensure the success of burning and ignited fusion plasma devices such as tokamaks and stellarators. In full performance ITER tokamak discharges, with QDT = 10, the energy released by a single ELM could reach ˜30 MJ which is expected to result in an energy density of 10-15 MJ/m2on the divertor targets. This will exceed the estimated divertor ablation limit by a factor of 20-30. A worldwide research program is underway to develop various types of ELM control techniques in preparation for ITER H-mode plasma operations. An overview of the ELM control techniques currently being developed is discussed along with the requirements for applying these techniques to plasmas in ITER. Particular emphasis is given to the primary approaches, pellet pacing and resonant magnetic perturbation fields, currently being considered for ITER.

Particle beam injection system

DOEpatents

Jassby, Daniel L.; Kulsrud, Russell M.

1977-01-01

This invention provides a poloidal divertor for stacking counterstreaming ion beams to provide high intensity colliding beams. To this end, method and apparatus are provided that inject high energy, high velocity, ordered, atomic deuterium and tritium beams into a lower energy, toroidal, thermal equilibrium, neutral, target plasma column that is magnetically confined along an endless magnetic axis in a strong restoring force magnetic field having helical field lines to produce counterstreaming deuteron and triton beams that are received bent, stacked and transported along the endless axis, while a poloidal divertor removes thermal ions and electrons all along the axis to increase the density of the counterstreaming ion beams and the reaction products resulting therefrom. By balancing the stacking and removal, colliding, strong focused particle beams, reaction products and reactions are produced that convert one form of energy into another form of energy.

Protein targeting and integration signal for the chloroplastic outer envelope membrane.

PubMed Central

Li, H M; Chen, L J

1996-01-01

Most proteins in chloroplasts are encoded by the nuclear genome and synthesized in the cytosol. With the exception of most quter envelope membrane proteins, nuclear-encoded chloroplastic proteins are synthesized with N-terminal extensions that contain the chloroplast targeting information of these proteins. Most outer membrane proteins, however, are synthesized without extensions in the cytosol. Therefore, it is not clear where the chloroplastic outer membrane targeting information resides within these polypeptides. We have analyzed a chloroplastic outer membrane protein, OEP14 (outer envelope membrane protein of 14 kD, previously named OM14), and localized its outer membrane targeting and integration signal to the first 30 amino acids of the protein. This signal consists of a positively charged N-terminal portion followed by a hydrophobic core, bearing resemblance to the signal peptides of proteins targeted to the endoplasmic reticulum. However, a chimeric protein containing this signal fused to a passenger protein did not integrate into the endoplasmic reticulum membrane. Furthermore, membrane topology analysis indicated that the signal inserts into the chloroplastic outer membrane in an orientation opposite to that predicted by the "positive inside" rule. PMID:8953775

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

DOE PAGES

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.

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 C 2+ 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 Dopplermore » Coherence Imaging System (CIS), measuring the toroidal velocity of the C 2+ 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

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

NASA Astrophysics Data System (ADS)

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

2017-07-01

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

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

Access to edge scenarios for testing a scraper element in early operation phases of Wendelstein 7-X

DOE PAGES

Holbe, H.; Pedersen, T. Sunn; Geiger, J.; ...

2016-01-29

The edge topology of magnetic fusion devices is decisive for the control of the plasma exhaust. In Wendelstein 7-X, the island divertor concept will be used, for which the edge topology can change significantly as the internal currents in a plasma discharge evolve towards steady-state. Consequently, the device has been optimized to minimize such internal currents, in particular the bootstrap current [1]. Nonetheless, there are predicted pulse scenarios where effects of the remaining internal currents could potentially lead to overload of plasma-facing components. These internal currents are predicted to evolve on long time scales (tens of seconds) so their effectsmore » on the edge topology and the divertor heat loads may not be experimentally accessible in the first years of W7-X operation, where only relatively short pulses are possible. However, we show here that for at least one important long-pulse divertor operation issue, relevant physics experiments can be performed already in short-pulse operation, through judicious adjustment of the edge topology by the use of the existing coil sets. The specific issue studied here is a potential overload of the divertor element edges. This overload might be mitigated by the installation of an extra set of plasma-facing components, so-called scraper elements, as suggested in earlier publications. It is shown here that by a targeted control of edge topology, the effectiveness of such scraper elements can be tested already with uncooled test-scraper elements in short-pulse operation. Furthermore, this will allow an early and well-informed decision on whether long-pulse-capable (actively cooled) scraper elements should be built and installed.« less

Reactor application of an improved bundle divertor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yang, T.F.; Ruck, G.W.; Lee, A.Y.

1978-11-01

A Bundle Divertor was chosen as the impurity control and plasma exhaust system for the beam driven Demonstration Tokamak Hybrid Reactor - DTHR. In the context of a preconceptual design study of the reactor and associated facility a bundle divertor concept was developed and integrated into the reactor system. The overall system was found feasible and scalable for reactors with intermediate torodial field strengths on axis. The important design characteristics are: the overall average current density of the divertor coils is 0.73 kA for each tesla of toroidal field on axis; the divertor windings are made from super-conducting cables supportedmore » by steel structures and are designed to be maintainable; the particle collection assembly and auxiliary cryosorption vacuum pump are dual systems designed such that they can be reactivated alterntively to allow for continuous reactor operation; and the power requirement for energizing and operating the divertor is about 5 MW.« less

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

PubMed

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

2016-08-01

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xu, J. C.; Jia, M. N.; Feng, W.

2016-08-15

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

A New Scaling for Divertor Detachment

NASA Astrophysics Data System (ADS)

Goldston, Robert

2017-10-01

The ITER design and future fusion power plant designs depend on divertor detachment, whether partial, pronounced or complete, both to limit heat flux to plasma-facing components and to limit surface erosion due to sputtering. Generally the parallel heat flux, estimated as proportional to Psep / R or Psep B / R , is used as a proxy for the difficulty of achieving detachment. Here we argue that the impurity cooling required for detachment is strongly dependent on the upstream separatrix density, which is limited by Greenwald scaling. Taking this into account self-consistently, along with the Heuristic Drift (HD) model for the SOL width, and using a Lengyel radiation model that includes non-coronal effects, we find that the relative impurity concentration, cz ≡nz /ne , required for detachment scales dominantly as cz Psep /Bp(nsep /nGW) 2 . The absence of any explicit favorable size scaling is concerning, as Psep must increase by an order of magnitude from present experiments to an economic fusion power system, while increases in the poloidal magnetic field strength are limited by magnet technology and MHD stability. This result should not be surprising, as it follows from the simplest scaling, Psep czne2VSOL , taking into account the Greenwald density limit and the HD SOL volume scaling. Reinke has combined a similar approach with the requirement to maintain H-mode, which sets a lower limit on Psep, and also arrives at an incentive for high field and disincentive for large size. These results should be challenged by comparison with 2D divertor codes and with measurements on existing experiments. In particular measurements are required for extrinsic divertor impurity concentration over a range of power and density conditions far from the regime where detachment can be achieved with deuterium puffing and intrinsic impurities alone. Nonetheless, these results suggest that higher magnetic field, stronger shaping, double-null operation, `advanced' divertor magnetic and baffle configurations, as well as lithium vapor targets merit greater attention. This work supported by the US DOE under contract DE-AC02-09CH11466.

Impurity seeding for tokamak power exhaust: from present devices via ITER to DEMO

NASA Astrophysics Data System (ADS)

Kallenbach, A.; Bernert, M.; Dux, R.; Casali, L.; Eich, T.; Giannone, L.; Herrmann, A.; McDermott, R.; Mlynek, A.; Müller, H. W.; Reimold, F.; Schweinzer, J.; Sertoli, M.; Tardini, G.; Treutterer, W.; Viezzer, E.; Wenninger, R.; Wischmeier, M.; the ASDEX Upgrade Team

2013-12-01

A future fusion reactor is expected to have all-metal plasma facing materials (PFMs) to ensure low erosion rates, low tritium retention and stability against high neutron fluences. As a consequence, intrinsic radiation losses in the plasma edge and divertor are low in comparison to devices with carbon PFMs. To avoid localized overheating in the divertor, intrinsic low-Z and medium-Z impurities have to be inserted into the plasma to convert a major part of the power flux into radiation and to facilitate partial divertor detachment. For burning plasma conditions in ITER, which operates not far above the L-H threshold power, a high divertor radiation level will be mandatory to avoid thermal overload of divertor components. Moreover, in a prototype reactor, DEMO, a high main plasma radiation level will be required in addition for dissipation of the much higher alpha heating power. For divertor plasma conditions in present day tokamaks and in ITER, nitrogen appears most suitable regarding its radiative characteristics. If elevated main chamber radiation is desired as well, argon is the best candidate for the simultaneous enhancement of core and divertor radiation, provided sufficient divertor compression can be obtained. The parameter Psep/R, the power flux through the separatrix normalized by the major radius, is suggested as a suitable scaling (for a given electron density) for the extrapolation of present day divertor conditions to larger devices. The scaling for main chamber radiation from small to large devices has a higher, more favourable dependence of about Prad,main/R2. Krypton provides the smallest fuel dilution for DEMO conditions, but has a more centrally peaked radiation profile compared to argon. For investigation of the different effects of main chamber and divertor radiation and for optimization of their distribution, a double radiative feedback system has been implemented in ASDEX Upgrade (AUG). About half the ITER/DEMO values of Psep/R have been achieved so far, and close to DEMO values of Prad,main/R2, albeit at lower Psep/R. Further increase of this parameter may be achieved by increasing the neutral pressure or improving the divertor geometry.

Macroscopic erosion of divertor and first wall armour in future tokamaks

NASA Astrophysics Data System (ADS)

Würz, H.; Bazylev, B.; Landman, I.; Pestchanyi, S.; Safronov, V.

2002-12-01

Sputtering, evaporation and macroscopic erosion determine the lifetime of the 'in vessel' armour materials CFC, tungsten and beryllium presently under discussion for future tokamaks. For CFC armour macroscopic erosion means brittle destruction and dust formation whereas for metallic armour melt layer erosion by melt motion and droplet splashing. Available results on macroscopic erosion from hot plasma and e-beam simulation experiments and from tokamaks are critically evaluated and a comprehensive discussion of experimental and numerical macroscopic erosion and its extrapolation to future tokamaks is given. Shielding of divertor armour materials by their own vapor exists during plasma disruptions. The evolving plasma shield protects the armour from high heat loads, absorbs the incoming energy and reradiates it volumetrically thus reducing drastically the deposited energy. As a result, vertical target erosion by vaporization turns out to be of the order of a few microns per disruption event and macroscopic erosion becomes the dominant erosion source.

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

NASA Astrophysics Data System (ADS)

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

2017-12-01

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

ELM-induced transient tungsten melting in the JET divertor

NASA Astrophysics Data System (ADS)

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

2015-02-01

The original goals of the JET ITER-like wall included the study of the impact of an all W divertor on plasma operation (Coenen et al 2013 Nucl. Fusion 53 073043) and fuel retention (Brezinsek et al 2013 Nucl. Fusion 53 083023). ITER has recently decided to install a full-tungsten (W) divertor from the start of operations. One of the key inputs required in support of this decision was the study of the possibility of W melting and melt splashing during transients. Damage of this type can lead to modifications of surface topology which could lead to higher disruption frequency or compromise subsequent plasma operation. Although every effort will be made to avoid leading edges, ITER plasma stored energies are sufficient that transients can drive shallow melting on the top surfaces of components. JET is able to produce ELMs large enough to allow access to transient melting in a regime of relevance to ITER. Transient W melt experiments were performed in JET using a dedicated divertor module and a sequence of IP = 3.0 MA/BT = 2.9 T H-mode pulses with an input power of PIN = 23 MW, a stored energy of ˜6 MJ and regular type I ELMs at ΔWELM = 0.3 MJ and fELM ˜ 30 Hz. By moving the outer strike point onto a dedicated leading edge in the W divertor the base temperature was raised within ˜1 s to a level allowing transient, ELM-driven melting during the subsequent 0.5 s. Such ELMs (δW ˜ 300 kJ per ELM) are comparable to mitigated ELMs expected in ITER (Pitts et al 2011 J. Nucl. Mater. 415 (Suppl.) S957-64). Although significant material losses in terms of ejections into the plasma were not observed, there is indirect evidence that some small droplets (˜80 µm) were released. Almost 1 mm (˜6 mm3) of W was moved by ˜150 ELMs within 7 subsequent discharges. The impact on the main plasma parameters was minor and no disruptions occurred. The W-melt gradually moved along the leading edge towards the high-field side, driven by j × B forces. The evaporation rate determined from spectroscopy is 100 times less than expected from steady state melting and is thus consistent only with transient melting during the individual ELMs. Analysis of IR data and spectroscopy together with modelling using the MEMOS code Bazylev et al 2009 J. Nucl. Mater. 390-391 810-13 point to transient melting as the main process. 3D MEMOS simulations on the consequences of multiple ELMs on damage of tungsten castellated armour have been performed. These experiments provide the first experimental evidence for the absence of significant melt splashing at transient events resembling mitigated ELMs on ITER and establish a key experimental benchmark for the MEMOS code.

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

NASA Astrophysics Data System (ADS)

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

2013-10-01

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

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

NASA Astrophysics Data System (ADS)

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

2017-12-01

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

Divertor electron temperature and impurity diffusion measurements with a spectrally resolved imaging radiometer.

PubMed

Clayton, D J; Jaworski, M A; Kumar, D; Stutman, D; Finkenthal, M; Tritz, K

2012-10-01

A divertor imaging radiometer (DIR) diagnostic is being studied to measure spatially and spectrally resolved radiated power P(rad)(λ) in the tokamak divertor. A dual transmission grating design, with extreme ultraviolet (~20-200 Å) and vacuum ultraviolet (~200-2000 Å) gratings placed side-by-side, can produce coarse spectral resolution over a broad wavelength range covering emission from impurities over a wide temperature range. The DIR can thus be used to evaluate the separate P(rad) contributions from different ion species and charge states. Additionally, synthetic spectra from divertor simulations can be fit to P(rad)(λ) measurements, providing a powerful code validation tool that can also be used to estimate electron divertor temperature and impurity transport.

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

NASA Astrophysics Data System (ADS)

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

2017-12-01

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

Measurements and modeling of intra-ELM tungsten sourcing and transport in DIII-D

NASA Astrophysics Data System (ADS)

Abrams, T.; Leonard, A. W.; Thomas, D. M.; McLean, A. G.; Makowski, M. A.; Wang, H. Q.; Unterberg, E. A.; Briesemeister, A. R.; Rudakov, D. L.; Bykov, I.; Donovan, D.

2017-10-01

Intra-ELM tungsten erosion profiles in the DIII-D divertor, acquired via W I spectroscopy with high temporal and spatial resolution, are consistent with SDTrim.SP sputtering modeling using measured ion saturation currents and impact energies during ELMs as input and an ad-hoc 2% C2+ impurity flux. The W sputtering profile peaks close to the OSP both during and between ELMs in the favorable BT direction. In reverse BT the W source peaks close to the OSP between ELMs but strongly broadens and shifts outboard during ELMs, heuristically consistent with radially outward ion transport via ExB drifts. Ion impact energies during ELMs (inferred taking the ratio of divertor heat flux to the ion saturation current) are found to be approximately equal to Te,ped, lower than the 4*Te,ped value predicted by the Fundamenski/Moulton free streaming model. These impact energies imply both D main ions and C impurities contribute strongly to W sputtering during ELMs on DIII-D. This work represents progress towards a predictive model to link upstream conditions (i.e., pedestal height) and SOL impurity levels to the ELM-induced W impurity source at both the strike-point and far-target regions in the ITER divertor. Correlations between ELM size/frequency and SOL W fluxes measured via a midplane deposition probe will also be presented. Work supported by US DOE under DE-FC02-04ER54698.

Towards a Lithium Radiative / Vapor-Box Divertor

NASA Astrophysics Data System (ADS)

Goldston, Robert; Constantin, Marius; Jaworski, Michael; Myers, Rachel; Ono, Masayuki; Schwartz, Jacob; Scotti, Filippo; Qu, Zhaonan

2014-10-01

Recent research has indicated that the peak perpendicular heat flux on reactor divertor targets will be hundreds of MW/m2 in the absence of dissipation and/or spatial spreading. Thus we are attracted to both enhanced radiative cooling and continuous vapor shielding. Lithium particle lifetimes <=100 micro-sec enhance radiation efficiency at T < 10 eV, while lithium charge-exchange with neutral hydrogen may enhance radiative efficiency for T > 10 eV and n0/ni > 0.1. We are examining if the latter mechanism plays a role in the narrowing of the heat-flux footprint in lithiated NSTX discharges. In parallel we are investigating the possibility of immersing a reactor divertor leg in a channel of lithium vapor. If we approximate the vapor channel as in local equilibrium with lithium-wetted walls ranging from 300 oC at the entrance point to 950 oC 10m downstream in the parallel direction, we find that the vapor can both balance reactor levels of upstream plasma pressure and stop energetic ions and electrons with energies up to at least 25 keV, as might be produced in ELMs. Each 10 l/sec of lithium evaporated deep in the channel and recondensed in cooler regions spreads 100 MW over a much wider area than the original strike point. This work supported by US DOE Contract DE-AC02-09CH11466.

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-C 1 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 differencemore » between 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

Edge simulations in ELMy H-mode discharges of EAST tokamak

NASA Astrophysics Data System (ADS)

Xia, T. Y.; Huang, Y. Q.; Xu, X. Q.; Wu, Y. B.; Wang, L.; Zheng, Z.; Liu, J. B.; Zang, Q.; Li, Y. Y.; Zhao, D.

2017-10-01

Simulations of ELM crash followed by a coherent mode, leading to transient divertor heat flux on EAST are achieved by the six-field two-fluid model in BOUT + + . Three EAST ELMy H-mode discharges with different pedestal structure, geometry and plasma current Ip are studied. The ELM-driven crash of the profiles in pedestal is reproduced, and the footprints of ELM filaments on targets are comparable with the measurements from divertor probes. A coherent mode is also found in the edge region in all the simulations after the ELM crash. The frequency and poloidal wave number are in the range of the edge coherent mode (ECM) on EAST. The magnetic fluctuations of the mode are smaller than the electric field fluctuations. The detailed comparisons between simulated mode structures with measurements will be reported. Statistical analysis on the simulated turbulent fluctuations shows that both the turbulent and blobby electron anomalous transport can pump the pedestal energy out into SOL, and then flow to divertors. The similar trend of the heat flux width with Ip is obtained in the simulations. The effects of the SOL current driven by LHW on ELMs will be discussed in this paper. This work was performed under the auspices of the US DOE by LLNL under contract DE-AC52-07NA27344. It was supported by the China NSF 11405215 and 11675217.

Honeycomblike large area LaB6 plasma source for Multi-Purpose Plasma facility

NASA Astrophysics Data System (ADS)

Woo, Hyun-Jong; Chung, Kyu-Sun; You, Hyun-Jong; Lee, Myoung-Jae; Lho, Taihyeop; Choh, Kwon Kook; Yoon, Jung-Sik; Jung, Yong Ho; Lee, Bongju; Yoo, Suk Jae; Kwon, Myeon

2007-10-01

A Multi-Purpose Plasma (MP2) facility has been renovated from Hanbit mirror device [Kwon et al., Nucl. Fusion 43, 686 (2003)] by adopting the same philosophy of diversified plasma simulator (DiPS) [Chung et al., Contrib. Plasma Phys. 46, 354 (2006)] by installing two plasma sources: LaB6 (dc) and helicon (rf) plasma sources; and making three distinct simulators: divertor plasma simulator, space propulsion simulator, and astrophysics simulator. During the first renovation stage, a honeycomblike large area LaB6 (HLA-LaB6) cathode was developed for the divertor plasma simulator to improve the resistance against the thermal shock fragility for large and high density plasma generation. A HLA-LaB6 cathode is composed of the one inner cathode with 4in. diameter and the six outer cathodes with 2in. diameter along with separate graphite heaters. The first plasma is generated with Ar gas and its properties are measured by the electric probes with various discharge currents and magnetic field configurations. Plasma density at the middle of central cell reaches up to 2.6×1012 cm-3, while the electron temperature remains around 3-3.5eV at the low discharge current of less than 45A, and the magnetic field intensity of 870G. Unique features of electric property of heaters, plasma density profiles, is explained comparing with those of single LaB6 cathode with 4in. diameter in DiPS.

Honeycomblike large area LaB6 plasma source for Multi-Purpose Plasma facility.

PubMed

Woo, Hyun-Jong; Chung, Kyu-Sun; You, Hyun-Jong; Lee, Myoung-Jae; Lho, Taihyeop; Choh, Kwon Kook; Yoon, Jung-Sik; Jung, Yong Ho; Lee, Bongju; Yoo, Suk Jae; Kwon, Myeon

2007-10-01

A Multi-Purpose Plasma (MP(2)) facility has been renovated from Hanbit mirror device [Kwon et al., Nucl. Fusion 43, 686 (2003)] by adopting the same philosophy of diversified plasma simulator (DiPS) [Chung et al., Contrib. Plasma Phys. 46, 354 (2006)] by installing two plasma sources: LaB(6) (dc) and helicon (rf) plasma sources; and making three distinct simulators: divertor plasma simulator, space propulsion simulator, and astrophysics simulator. During the first renovation stage, a honeycomblike large area LaB(6) (HLA-LaB(6)) cathode was developed for the divertor plasma simulator to improve the resistance against the thermal shock fragility for large and high density plasma generation. A HLA-LaB(6) cathode is composed of the one inner cathode with 4 in. diameter and the six outer cathodes with 2 in. diameter along with separate graphite heaters. The first plasma is generated with Ar gas and its properties are measured by the electric probes with various discharge currents and magnetic field configurations. Plasma density at the middle of central cell reaches up to 2.6 x 10(12) cm(-3), while the electron temperature remains around 3-3.5 eV at the low discharge current of less than 45 A, and the magnetic field intensity of 870 G. Unique features of electric property of heaters, plasma density profiles, is explained comparing with those of single LaB(6) cathode with 4 in. diameter in DiPS.

EDGE2D-EIRENE modelling of near SOL E r: possible impact on the H-mode power threshold

NASA Astrophysics Data System (ADS)

Chankin, A. V.; Delabie, E.; Corrigan, G.; Harting, D.; Maggi, C. F.; Meyer, H.; Contributors, JET

2017-04-01

Recent EDGE2D-EIRENE simulations of JET plasmas showed a significant difference between radial electric field (E r) profiles across the separatrix in two divertor configurations, with the outer strike point on the horizontal target (HT) and vertical target (VT) (Chankin et al 2016 Nucl. Mater. Energy, doi: 10.1016/j.nme.2016.10.004). Under conditions (input power, plasma density) where the HT plasma went into the H-mode, a large positive E r spike in the near scrape-off layer (SOL) was seen in the code output, leading to a very large E × B shear across the separatrix over a narrow region of a fraction of a cm width. No such E r feature was obtained in the code solution for the VT configuration, where the H-mode power threshold was found to be twice as high as in the HT configuration. It was hypothesised that the large E × B shear across the separatrix in the HT configuration could be responsible for the turbulence suppression leading to an earlier (at lower input power) L-H transition compared to the VT configuration. In the present work these ideas are extended to cover some other experimental observations on the H-mode power threshold variation with parameters which typically are not included in the multi-machine H-mode power threshold scalings, namely: ion mass dependence (isotope H-D-T exchange), dependence on the ion ∇B drift direction, and dependence on the wall material composition (ITER-like wall versus carbon wall in JET). In all these cases EDGE2D-EIRENE modelling shows larger positive E r spikes in the near SOL under conditions where the H-mode power threshold is lower, at least in the HT configuration.

Gas turbine row #1 steam cooled vane

DOEpatents

Cunha, Frank J.

2000-01-01

A design for a vane segment having a closed-loop steam cooling system is provided. The vane segment comprises an outer shroud, an inner shroud and an airfoil, each component having a target surface on the inside surface of its walls. A plurality of rectangular waffle structures are provided on the target surface to enhance heat transfer between each component and cooling steam. Channel systems are provided in the shrouds to improve the flow of steam through the shrouds. Insert legs located in cavities in the airfoil are also provided. Each insert leg comprises outer channels located on a perimeter of the leg, each outer channel having an outer wall and impingement holes on the outer wall for producing impingement jets of cooling steam to contact the airfoil's target surface. Each insert leg further comprises a plurality of substantially rectangular-shaped ribs located on the outer wall and a plurality of openings located between outer channels of the leg to minimize cross flow degradation.

Optimization of a bundle divertor for FED

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hively, L.M.; Rothe, K.E.; Minkoff, M.

1982-01-01

Optimal double-T bundle divertor configurations have been obtained for the Fusion Engineering Device (FED). On-axis ripple is minimized, while satisfying a series of engineering constraints. The ensuing non-linear optimization problem is solved via a sequence of quadratic programming subproblems, using the VMCON algorithm. The resulting divertor designs are substantially improved over previous configurations.

Modelling of Divertor Detachment in MAST Upgrade

NASA Astrophysics Data System (ADS)

Moulton, David; Carr, Matthew; Harrison, James; Meakins, Alex

2017-10-01

MAST Upgrade will have extensive capabilities to explore the benefits of alternative divertor configurations such as the conventional, Super-X, x divertor, snowflake and variants in a single device with closed divertors. Initial experiments will concentrate on exploring the Super-X and conventional configurations, in terms of power and particle loads to divertor surfaces, access to detachment and its control. Simulations have been carried out with the SOLPS5.0 code validated against MAST experiments. The simulations predict that the Super-X configuration has significant advantages over the conventional, such as lower detachment threshold (2-3x lower in terms of upstream density and 4x higher in terms of PSOL). Synthetic spectroscopy diagnostics from these simulations have been created using the Raysect ray tracing code to produce synthetic filtered camera images, spectra and foil bolometer data. Forward modelling of the current set of divertor diagnostics will be presented, together with a discussion of future diagnostics and analysis to improve estimates of the plasma conditions. Work supported by the RCUK Energy Programme [Grant Number EP/P012450/1] and EURATOM.

Thermal Fatigue Study on the Divertor Plate Materials

NASA Astrophysics Data System (ADS)

Wu, Ji-hong; Zhang, Fu; Xu, Zeng-yu; Yan, Jian-cheng

2002-10-01

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

McLean, A. G., E-mail: mclean@fusion.gat.com; Soukhanovskii, V. A.; Allen, S. L.

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 onmore » NSTX-U, the divertor TS system would provide an invaluable diagnostic for the boundary program to characterize the edge plasma.« less

Tritium analysis of divertor tiles used in JET ITER-like wall campaigns by means of β-ray induced x-ray spectrometry

NASA Astrophysics Data System (ADS)

Hatano, Y.; Yumizuru, K.; Koivuranta, S.; Likonen, J.; Hara, M.; Matsuyama, M.; Masuzaki, S.; Tokitani, M.; Asakura, N.; Isobe, K.; Hayashi, T.; Baron-Wiechec, A.; Widdowson, A.; contributors, JET

2017-12-01

Energy spectra of β-ray induced x-rays from divertor tiles used in ITER-like wall campaigns of the Joint European Torus were measured to examine tritium (T) penetration into tungsten (W) layers. The penetration depth of T evaluated from the intensity ratio of W(Lα) x-rays to W(Mα) x-rays showed clear correlation with poloidal position; the penetration depth at the upper divertor region reached several micrometers, while that at the lower divertor region was less than 500 nm. The deep penetration at the upper part was ascribed to the implantation of high energy T produced by DD fusion reactions. The poloidal distribution of total x-ray intensity indicated higher T retention in the inboard side than the outboard side of the divertor region.

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€.

X-Divertors on ITER - with no hardware changes

NASA Astrophysics Data System (ADS)

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

2014-10-01

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

Radioisotope Electric Propulsion for Fast Outer Planetary Orbiters

NASA Technical Reports Server (NTRS)

Oleson, Steven; Benson, Scott; Gefert, Leon; Patterson, Michael; Schreiber, Jeffrey

2002-01-01

Recent interest in outer planetary targets by the Office of Space Science has spurred the search for technology options to enable relatively quick missions to outer planetary targets. Several options are being explored including solar electric propelled stages combined with aerocapture at the target and nuclear electric propulsion. Another option uses radioisotope powered electric thrusters to reach the outer planets. Past work looked at using this technology to provide faster flybys. A better use for this technology is for outer planet orbiters. Combined with medium class launch vehicles and a new direct trajectory these small, sub-kilowatt ion thrusters and Stirling radioisotope generators were found to allow missions as fast as 5 to 12 years for objects from Saturn to Pluto, respectively. Key to the development is light spacecraft and science payload technologies.

Damage of target edges in brush-like geometry in the course of ELM-like plasma pulses in QSPA Kh-50

NASA Astrophysics Data System (ADS)

Makhlaj, V. A.; Garkusha, I. E.; Aksenov, N. N.; Bazylev, B.; Byrka, O. V.; Chebotarev, V. V.; Landman, I.; Herashchenko, S. S.; Staltsov, V. V.

2015-08-01

Castellated edges of macro-brush armour elements of ITER divertor can be a source of molten/solid dust particles which are injected into the plasma. The targets that combined in brush-like geometry have been irradiated under different inclination angles in QSPA Kh-50. The cubic brushes element has typical size of 1 cm. The titanium was used to investigate dynamics of mountains' formation. The onset of dust particles ejection from the exposed castellated targets has been studied. Formation of resolidified bridges through the gaps of brush-like targets due to the melt motion is studied in dynamics. With following plasma impacts such resolidified bridges became additional source of dust.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ono, M.; Jaworski, M. A.; Kaita, R.

Steady-state fusion reactor operation presents major divertor technology challenges, including high divertor heat flux both steady-state and transients. In addition to those issues, there are unresolved issues of long term dust accumulation and associated tritium inventory and safety issues. It has been suggested that radiative liquid lithium divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues while potentially improving the reactor plasma performance. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-freemore » core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid lithium divertor (RLLD) concept and its variant, the active liquid lithium divertor concept (ARLLD), taking advantage of the enhanced Li radiation in relatively poorly confined divertor plasmas. It was estimated that only a few moles/sec of lithium injection would be needed to significantly reduce the divertor heat flux in a tokamak fusion power plant. By operating at lower temperatures ≤ 500°C than the first wall ~ 600 – 700°C, the LL-covered divertor chamber wall surfaces can serve as an effective particle pump, as impurities generally migrate toward lower temperature LL divertor surfaces. To maintain the LL purity, a closed LL loop system with a modest circulating capacity of ~ 1 liter/second (l/sec) is envisioned to sustain the steady-state operation of a 1 GW-electric class fusion power plant. By running the Li loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to outside where the dust / impurities are removed by relatively simple filter and cold/hot trap systems. Using a cold trap system, it can recover in tritium (T) in real time from LL at a rate of ~ 0.5 g / sec needed to sustain the fusion reaction while minimizing the T inventory issue. With an expected T fraction of ≤ 0.7 %, an acceptable level of T inventory can be achieved. In NSTX-U, preparations are now underway to elucidate the physics of Li plasma interactions with a number of Li application tools and Li radiation spectroscopic instruments. The NSTX-U Li evaporator which provides Li coating over the lower divertor plate, can offer important information on the RLLD concept, and the Li granule injector will test some of the key physics issue on the ARLLD concept. A LL-loop is also being prepared off line for prototyping future use on NSTX-U.« less

Investigation of Tokamak Solid Divertor Target Options.

DTIC Science & Technology

1981-05-26

but materials are not known which could operate at the high resulting wall temperatures . Mist- steam flows would also demand a relatively high ...flux P = coolant density = bulk coolant viscosity w = coolant viscosity at average wall temperature = units conversion At high heat loads and moderate...therefore, the inner wall temperature will be over 300 OF, posing a high temp- erature materials challenge. E. Swirl and Mixed Flow Schemes Extensive work

Impact of target material on D and D2 recycling in DIII-D ELMy H-mode discharges

NASA Astrophysics Data System (ADS)

Bykov, Igor; Hollmann, Eric; Rudakov, Dmitry; Moyer, Richard; Boedo, Jose; Din, Rui; Wang, Huiqian; Unterberg, Ezekeal; Briesemeister, Alexis; Chrobak, Christopher; Abrams, Tyler; Watkins, Jon; Lasnier, Charles; McLean, Adam

2017-10-01

DIII-D operation with W divertor inserts shows molecular recycling flux (measured by Fulcher-a spectroscopy) is reduced between ELMs in comparison with a C divertor where the flux is dominated by D2 molecules (>=90%). This effect is partly explained by the higher reflection probability of atomic D on W. During ELMs, the molecular fraction drops by factor >2 on both C and W targets. To study the effect of higher ion impact energy (Eimp) on transient D re-emission during ELMs we have applied fast electrostatic bias to a DiMES probe equipped with a W and C sample set. A 50% increase of Eimp from 150 eV due to biasing led to transient increase of atomic D re-emission flux on both targets. Similar increase of the D2 flux was only seen on C. Thus, the ratios of atomic and molecular fluxes on C varied in a similar way to those measured during ELMs. This variation in molecular recycling fraction with material has implications for the dynamics of density pedestal recovery between ELMs, the overall global particle balance of the system, and possibly the overall detachment onset conditions transiently due to the ELM particle influx. Supported by the US DOE under DE-FG02-07ER54917, DE-FG02-04ER54758, DE-FC02-04ER54698, DE-FG03-95ER54309, and DE-FG02-04ER54762.

Morphological and structural effects on tungsten targets produced by fusion plasma pulses from a table top plasma focus

NASA Astrophysics Data System (ADS)

Inestrosa-Izurieta, M. J.; Ramos-Moore, E.; Soto, L.

2015-09-01

A table top plasma focus device operating at hundreds of joules was used to simulate an equivalent damage factor than the obtained on the divertor in tokamak experiments. Using the ejected plasma produced after the pinch disruptions, the effects on tungsten targets from 50 cumulative plasma shocks with power fluxes per shot between 2.6 and 9200 kW cm-2 and with a duration time in the order of tens of nanoseconds (damage factor in the order of 100-103 (W cm-2)s1/2) were studied. Morphological analysis shows an increasing appearance of cracked surfaces with holes, fissures and defects, suggesting a potential progression of stress effects and a fast heat load that melts the surface, ending in thermal contractions that recrystallize the surface of the target. A structural analysis demonstrates a compressive stress development and suggests that part of the energy is released in the melting of the surface in case of a plasma shock with a power flux of 9.2 MW cm-2, 75 ns duration pulse, 2.5   ×   103 (W cm-2)s1/2 damage factor. How to increase the damage factor by one order of magnitude up to the expected value from type I ELMs on the ITER divertor, i.e. 104 (W cm-2)s1/2 is discussed.

Experimental results of near real-time protection system for plasma facing components in Wendelstein 7-X at GLADIS

NASA Astrophysics Data System (ADS)

Ali, A.; Jakubowski, M.; Greuner, H.; Böswirth, B.; Moncada, V.; Sitjes, A. Puig; Neu, R.; Pedersen, T. S.; the W7-X Team

2017-12-01

One of the aims of stellarator Wendelstein 7-X (W7-X), is to investigate steady state operation, for which power exhaust is an important issue. The predominant fraction of the energy lost from the confined plasma region will be absorbed by an island divertors, which is designed for 10 {{MWm}}-2 steady state operation. In order to protect the divertor targets from overheating, 10 state-of-the-art infrared endoscopes will be installed at W7-X. In this work, we present the experimental results obtained at the high heat flux test facility GLADIS (Garching LArge DIvertor Sample test facility in IPP Garching) [1] during tests of a new plasma facing components (PFCs) protection algorithm designed for W7-X. The GLADIS device is equipped with two ion beams that can generate a heat load in the range from 3 MWm-2 to 55 MWm-2. The algorithms developed at W7-X to detect defects and hot spots are based on the analysis of surface temperature evolution and are adapted to work in near real-time. The aim of this work was to test the near real-time algorithms in conditions close to those expected in W7-X. The experiments were performed on W7-X pre-series tiles to detect CFC/Cu delaminations. For detection of surface layers, carbon fiber composite (CFC) blocks from the divertor of the Wendelstein 7-AS stellarator were used to observe temporal behavior of fully developed surface layers. These layers of re-deposited materials, like carbon, boron, oxygen and iron, were formed during the W7-AS operation. A detailed analysis of the composition and their thermal response to high heat fluxes (HHF) are described in [2]. The experiments indicate that the automatic detection of critical events works according to W7-X PFC protection requirements.

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.

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.

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.

ELM elimination with lithium aerosol injection in upper-single null discharges using the tungsten divertor in EAST

NASA Astrophysics Data System (ADS)

Sun, Z.; Maingi, R.; Hu, J.; Lunsford, R.; Diallo, A.; Tritz, K.; Osborne, T.; Canik, J.; Zuo, G.; Wang, L.; Xu, G.; Gong, X.; EAST Team Team

2017-10-01

A reproducible, fully non-inductive H-mode regime devoid of large ELMs has been achieved by continuous Li injection in EAST into the upper `ITER-like' tungsten divertor, extending previous results on the graphite divertor. These discharges did not suffer from density or impurity accumulation, and maintained constant core radiated power. The new results extend the energy confinement multiplier H98(y,2) 1.2, as compared to H98(y,2) 0.75 previously on the graphite divertor. The observed ELM elimination is correlated with a decrease in particle recycling, as expected from the strong Li coating before the experiment, and real-time Li aerosol injection. In addition, core W concentration was reduced during the Li injection. ELM elimination is likely related to the reduced recycling and density /temperature profile changes. A low-n electromagnetic coherent mode (MCM) at 40kHz became stronger in amplitude and also more coherent. The MCM shows strong magnetic fluctuations as measured by fast Mirnov coils, but weak density fluctuations. As compared to the graphite divertor, Li injection into the tungsten divertor eliminated ELMs at twice the previous auxiliary heating power, and reduced pedestal collisionality.

Experimental investigations of castellated monoblock structures in TEXTOR

NASA Astrophysics Data System (ADS)

Litnovsky, A.; Philipps, V.; Wienhold, P.; Sergienko, G.; Emmoth, B.; Rubel, M.; Breuer, U.; Wessel, E.

2005-03-01

To insure the thermo-mechanical durability of ITER it is planned to manufacture the castellated armour of the divertor i.e. to split the armour into cells [W. Daener et al., Fusion Eng. Des. 61&62 (2002) 61]. This will cause an increase of the surface area and may lead to carbon deposition and tritium accumulation in the gaps in between cells. To investigate the processes of deposition and fuel accumulation in gaps, a castellated test-limiter was exposed to the SOL plasma of TEXTOR. The geometry of castellation used was the same as proposed for the vertical divertor target in ITER [W. Daener et al., Fusion Eng. Des. 61&62 (2002) 61]. After exposure the limiter was investigated with various surface diagnostic techniques. Deposited layers containing carbon, hydrogen, deuterium and boron were found both on top plasma-facing surfaces and in the gaps. The amount of deuterium in the gaps was at least 30% of that found on the top surfaces.

Effects of ELMs on ITER divertor armour materials

NASA Astrophysics Data System (ADS)

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

2007-06-01

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Roquemore, A; Maingi, R; Lasnier, C

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 smallmore » 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.« less

Progress towards modeling tokamak boundary plasma turbulence and understanding its role in setting divertor heat flux widths

NASA Astrophysics Data System (ADS)

Chen, B.; Xu, X. Q.; Xia, T. Y.; Li, N. M.; Porkolab, M.; Edlund, E.; LaBombard, B.; Terry, J.; Hughes, J. W.; Ye, M. Y.; Wan, Y. X.

2018-05-01

The heat flux distributions on divertor targets in H-mode plasmas are serious concerns for future devices. We seek to simulate the tokamak boundary plasma turbulence and heat transport in the edge localized mode-suppressed regimes. The improved BOUT++ model shows that not only Ip but also the radial electric field Er plays an important role on the turbulence behavior and sets the heat flux width. Instead of calculating Er from the pressure gradient term (diamagnetic Er), it is calculated from the plasma transport equations with the sheath potential in the scrape-off layer and the plasma density and temperature profiles inside the separatrix from the experiment. The simulation results with the new Er model have better agreement with the experiment than using the diamagnetic Er model: (1) The electromagnetic turbulence in enhanced Dα H-mode shows the characteristics of quasi-coherent modes (QCMs) and broadband turbulence. The mode spectra are in agreement with the phase contrast imaging data and almost has no change in comparison to the cases which use the diamagnetic Er model; (2) the self-consistent boundary Er is needed for the turbulence simulations to get the consistent heat flux width with the experiment; (3) the frequencies of the QCMs are proportional to Er, while the divertor heat flux widths are inversely proportional to Er; and (4) the BOUT++ turbulence simulations yield a similar heat flux width to the experimental Eich scaling law and the prediction from the Goldston heuristic drift model.

Edge turbulence and divertor heat flux width simulations of Alcator C-Mod discharges using an electromagnetic two-fluid model

NASA Astrophysics Data System (ADS)

Chen, B.; Xu, X. Q.; Xia, T. Y.; Porkolab, M.; Edlund, E.; LaBombard, B.; Terry, J.; Hughes, J. W.; Mao, S. F.; Ye, M. Y.; Wan, Y. X.

2017-11-01

The BOUT++ code has been exploited in order to improve the understanding of the role of turbulent modes in controlling edge transport and resulting scaling of the scrape-off layer (SOL) heat flux width. For the C-Mod enhanced D_α (EDA) H-mode discharges, BOUT++ six-field two-fluid nonlinear simulations show a reasonable agreement of upstream turbulence and divertor target heat flux behavior: (a) the simulated quasi-coherent modes show consistent characteristics of the frequency versus poloidal wave number spectra of the electromagnetic fluctuations when compared with experimental measurements: frequencies are around 60-120 kHz (experiment: about 70-110 kHz), k_θ are around 2.0 cm-1 which is similar to the phase contrast imaging data; (b) linear spectrum analysis is consistent with the nonlinear phase relationship calculation which indicates the dominance of resistive-ballooning modes and drift-Alfven wave instabilities; (c) the SOL heat flux width λq versus current I p scaling is reproduced by turbulent transport: the simulations yield similar λq to experimental measurements within a factor of 2. However the magnitudes of divertor heat fluxes can be varied, depending on the physics models, sources and sinks, sheath boundary conditions, or flux limiting coefficient; (d) Simple estimate by the ‘2-point model’ for λq is consistent with simulation. Moreover, blobby turbulent spreading is confirmed for these relatively high B p shots.

Liquid lithium applications for solving challenging fusion reactor issues and NSTX-U contributions

DOE PAGES

Ono, M.; Jaworski, M. A.; Kaita, R.; ...

2016-08-05

Steady-state fusion reactor operation presents major divertor technology challenges, including high divertor heat flux both steady-state and transients. In addition to those issues, there are unresolved issues of long term dust accumulation and associated tritium inventory and safety issues. It has been suggested that radiative liquid lithium divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues while potentially improving the reactor plasma performance. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-freemore » core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid lithium divertor (RLLD) concept and its variant, the active liquid lithium divertor concept (ARLLD), taking advantage of the enhanced Li radiation in relatively poorly confined divertor plasmas. It was estimated that only a few moles/sec of lithium injection would be needed to significantly reduce the divertor heat flux in a tokamak fusion power plant. By operating at lower temperatures ≤ 500°C than the first wall ~ 600 – 700°C, the LL-covered divertor chamber wall surfaces can serve as an effective particle pump, as impurities generally migrate toward lower temperature LL divertor surfaces. To maintain the LL purity, a closed LL loop system with a modest circulating capacity of ~ 1 liter/second (l/sec) is envisioned to sustain the steady-state operation of a 1 GW-electric class fusion power plant. By running the Li loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to outside where the dust / impurities are removed by relatively simple filter and cold/hot trap systems. Using a cold trap system, it can recover in tritium (T) in real time from LL at a rate of ~ 0.5 g / sec needed to sustain the fusion reaction while minimizing the T inventory issue. With an expected T fraction of ≤ 0.7 %, an acceptable level of T inventory can be achieved. In NSTX-U, preparations are now underway to elucidate the physics of Li plasma interactions with a number of Li application tools and Li radiation spectroscopic instruments. The NSTX-U Li evaporator which provides Li coating over the lower divertor plate, can offer important information on the RLLD concept, and the Li granule injector will test some of the key physics issue on the ARLLD concept. A LL-loop is also being prepared off line for prototyping future use on NSTX-U.« less

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

ELM elimination with Li powder injection in EAST discharges using the tungsten upper divertor

DOE PAGES

Maingi, R.; Hu, J. S.; Sun, Z.; ...

2018-01-05

Here, we report the first successful use of lithium (Li) to eliminate edge-localized modes (ELMs) with tungsten divertor plasma-facing components in the EAST device. Li powder injected into the scrape-off layer of the tungsten upper divertor successfully eliminated ELMs for 3–5 s in EAST. The ELM elimination became progressively more effective in consecutive discharges at constant lithium delivery rates, and the divertor D α baseline emission was reduced, both signatures of improved wall conditioning. A modest decrease in stored energy and normalized energy confinement was also observed, but the confinement relative to H98 remained well above 1, extending the previousmore » ELM elimination results via Li injection into the lower carbon divertor in EAST. These results can be compared with recent observations with lithium pellets in ASDEX-Upgrade that failed to mitigate ELMs, highlighting one comparative advantage of continuous powder injection for real-time ELM elimination.« less

ELM elimination with Li powder injection in EAST discharges using the tungsten upper divertor

NASA Astrophysics Data System (ADS)

Maingi, R.; Hu, J. S.; Sun, Z.; Tritz, K.; Zuo, G. Z.; Xu, W.; Huang, M.; Meng, X. C.; Canik, J. M.; Diallo, A.; Lunsford, R.; Mansfield, D. K.; Osborne, T. H.; Gong, X. Z.; Wang, Y. F.; Li, Y. Y.; EAST Team

2018-02-01

We report the first successful use of lithium (Li) to eliminate edge-localized modes (ELMs) with tungsten divertor plasma-facing components in the EAST device. Li powder injected into the scrape-off layer of the tungsten upper divertor successfully eliminated ELMs for 3-5 s in EAST. The ELM elimination became progressively more effective in consecutive discharges at constant lithium delivery rates, and the divertor D α baseline emission was reduced, both signatures of improved wall conditioning. A modest decrease in stored energy and normalized energy confinement was also observed, but the confinement relative to H98 remained well above 1, extending the previous ELM elimination results via Li injection into the lower carbon divertor in EAST (Hu et al 2015 Phys. Rev. Lett. 114 055001). These results can be compared with recent observations with lithium pellets in ASDEX-Upgrade that failed to mitigate ELMs (Lang et al 2017 Nucl. Fusion 57 016030), highlighting one comparative advantage of continuous powder injection for real-time ELM elimination.

ELM elimination with Li powder injection in EAST discharges using the tungsten upper divertor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maingi, R.; Hu, J. S.; Sun, Z.

Here, we report the first successful use of lithium (Li) to eliminate edge-localized modes (ELMs) with tungsten divertor plasma-facing components in the EAST device. Li powder injected into the scrape-off layer of the tungsten upper divertor successfully eliminated ELMs for 3–5 s in EAST. The ELM elimination became progressively more effective in consecutive discharges at constant lithium delivery rates, and the divertor D α baseline emission was reduced, both signatures of improved wall conditioning. A modest decrease in stored energy and normalized energy confinement was also observed, but the confinement relative to H98 remained well above 1, extending the previousmore » ELM elimination results via Li injection into the lower carbon divertor in EAST. These results can be compared with recent observations with lithium pellets in ASDEX-Upgrade that failed to mitigate ELMs, highlighting one comparative advantage of continuous powder injection for real-time ELM elimination.« less

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

NASA Astrophysics Data System (ADS)

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

2017-12-01

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

Simple Map with Low MN Perturbation for a Single-Null Divertor Tokamak with Constant Width of Stochastic Layer

NASA Astrophysics Data System (ADS)

Verma, Arun; Smith, Terry; Punjabi, Alkesh; Boozer, Allen

1996-11-01

In this work, we investigate the effects of low MN perturbations in a single-null divertor tokamak with stochastic scrape-off layer. The unperturbed magnetic topology of a single-null divertor tokamak is represented by Simple Map (Punjabi A, Verma A and Boozer A, Phys Rev Lett), 69, 3322 (1992) and J Plasma Phys, 52, 91 (1994). We choose the combinations of the map parameter k, and the strength of the low MN perturbation such that the width of stochastic layer remains unchanged. We give detailed results on the effects of low MN perturbation on the magnetic topology of the stochastic layer and on the footprint of field lines on the divertor plate given the constraint of constant width of the stochastic layer. The low MN perturbations occur naturally and therefore their effects are of considerable importance in tokamak divertor physics. This work is supported by US DOE OFES. Use of CRAY at HU and at NERSC is gratefully acknowledged.

On Heat Loading, Novel Divertors, and Fusion Reactors

NASA Astrophysics Data System (ADS)

Kotschenreuther, Mike

2006-10-01

A new magnetic divertor geometry has been proposed to solve reactor heat exhaust problems, which are far more severe for a reactor than for ITER. Using reactor-compatible coils to generate an extra X-point downstream from the main X-point, the new X-divertor (XD) is shown to greatly expand magnetic flux at the divertor plates. As a result, the heat is distributed over a larger area and the line length is greatly increased. The heat-flux limitations of a standard divertor (SD) force a high core radiation fraction (fRad) in most reactor designs that necessarily have a several times higher ratio of heating power to radius (P/R) than ITER. It is argued that such high values of fRad will probably have serious deleterious consequences on the core confinement and stability of a burning plasma. Operation with internal transport barriers (ITBs) does not appear to overcome this problem. By reducing the core fRad within an acceptable range, the X-divertor is shown to substantially lower the core confinement requirement for a fusion reactor. As a bonus, the XD also enables the use of liquid metals by reducing the MHD drag. A possible series of experiments for an efficient and attractive path to practical fusion power is suggested.

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.

Upgraded divertor Thomson scattering system on DIII-D.

PubMed

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 (T e in the range of 0.5 eV-2 keV, n e in the range of 5 × 10 18 -1 × 10 21 m 3 ) for both low T e in detachment and high T e measurement up beyond the separatrix.

DOE Office of Scientific and Technical Information (OSTI.GOV)

M. Ono, M. Jaworski, R. Kaita, C. N. Skinner, J.P. Allain, R. Maingi, F. Scotti, V.A. Soukhanovskii, and the NSTX-U Team

Developing a reactor compatible divertor and managing the associated plasma material interaction (PMI) has been identified as a high priority research area for magnetic confinement fusion. Accordingly on NSTXU, the PMI research has received a strong emphasis. With ~ 15 MW of auxiliary heating power, NSTX-U will be able to test the PMI physics with the peak divertor plasma facing component (PFC) heat loads of up to 40-60 MW/m2 . To support the PMI research, a comprehensive set of PMI diagnostic tools are being implemented. The snow-flake configuration can produce exceptionally high divertor flux expansion of up to ~ 50.more » Combined with the radiative divertor concept, the snow-flake configuration has reduced the divertor heat flux by an order of magnitude in NSTX. Another area of active PMI investigation is the effect of divertor lithium coating (both in solid and liquid phases). The overall NSTX lithium PFC coating results suggest exciting opportunities for future magnetic confinement research including significant electron energy confinement improvements, Hmode power threshold reduction, the control of Edge Localized Modes (ELMs), and high heat flux handling. To support the NSTX-U/PPPL PMI research, there are also a number of associated PMI facilities implemented at PPPL/Princeton University including the Liquid Lithium R&D facility, Lithium Tokamak Experiment, and Laboratories for Materials Characterization and Surface Chemistry.« less

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.

Micro-cone targets for producing high energy and low divergence particle beams

DOEpatents

Le Galloudec, Nathalie

2013-09-10

The present invention relates to micro-cone targets for producing high energy and low divergence particle beams. In one embodiment, the micro-cone target includes a substantially cone-shaped body including an outer surface, an inner surface, a generally flat and round, open-ended base, and a tip defining an apex. The cone-shaped body tapers along its length from the generally flat and round, open-ended base to the tip defining the apex. In addition, the outer surface and the inner surface connect the base to the tip, and the tip curves inwardly to define an outer surface that is concave, which is bounded by a rim formed at a juncture where the outer surface meets the tip.

A Spherical Torus Nuclear Fusion Reactor Space Propulsion Vehicle Concept for Fast Interplanetary Travel

NASA Technical Reports Server (NTRS)

Williams, Craig H.; Borowski, Stanley K.; Dudzinski, Leonard A.; Juhasz, Albert J.

1998-01-01

A conceptual vehicle design enabling fast outer solar system travel was produced predicated on a small aspect ratio spherical torus nuclear fusion reactor. Initial requirements were for a human mission to Saturn with a greater than 5% payload mass fraction and a one way trip time of less than one year. Analysis revealed that the vehicle could deliver a 108 mt crew habitat payload to Saturn rendezvous in 235 days, with an initial mass in low Earth orbit of 2,941 mt. Engineering conceptual design, analysis, and assessment was performed on all ma or systems including payload, central truss, nuclear reactor (including divertor and fuel injector), power conversion (including turbine, compressor, alternator, radiator, recuperator, and conditioning), magnetic nozzle, neutral beam injector, tankage, start/re-start reactor and battery, refrigeration, communications, reaction control, and in-space operations. Detailed assessment was done on reactor operations, including plasma characteristics, power balance, power utilization, and component design.

Fine metal dust particles on the wall probes from JET-ILW

NASA Astrophysics Data System (ADS)

Fortuna-Zaleśna, E.; Grzonka, J.; Moon, Sunwoo; Rubel, M.; Petersson, P.; Widdowson, A.; Contributors, JET

2017-12-01

Collection and ex situ studies of dust generated in controlled fusion devices during plasma operation are regularly carried out after experimental campaigns. Herewith results of the dust survey performed in JET after the second phase of operation with the metal ITER-like wall (2013-2014) are presented. For the first-time-ever particles deposited on silicon plates acting as dust collectors installed in the inner and outer divertor have been examined. The emphasis is on analysing metal particles (Be and W) with the aim to determine their composition, size and surface topography. The most important is the identification of beryllium dust in the form of droplets (both splashes and spherical particles), flakes of co-deposits and small fragments of Be tiles. Tungsten and nickel rich (from Inconel) particles are also identified. Nitrogen from plasma edge cooling has been detected in all types of particles. They are categorized and the origin of various constituents is discussed.

Present status of liquid metal research for a fusion reactor

NASA Astrophysics Data System (ADS)

Tabarés, Francisco L.

2016-01-01

Although the use of solid materials as targets of divertor plasmas in magnetic fusion research is accepted as the standard solution for the very challenging issue of power and particle handling in a fusion reactor, a generalized feeling that the present options chosen for ITER will not represent the best choice for a reactor is growing up. The problems found for tungsten, the present selection for the divertor target of ITER, in laboratory tests and in hot plasma fusion devices suggest so. Even in the absence of the strong neutron irradiation expected in a reactor, issues like surface melting, droplet ejection, surface cracking, dust generation, etc., call for alternative solutions in a long pulse, high efficient fusion energy-producing continuous machine. Fortunately enough, decades of research on plasma facing materials based on liquid metals (LMs) have produced a wealth of appealing ideas that could find practical application in the route to the realization of a commercial fusion power plant. The options presently available, although in a different degree of maturity, range from full coverage of the inner wall of the device with liquid metals, so that power and particle exhaust together with neutron shielding could be provided, to more conservative combinations of liquid metal films and conventional solid targets basically representing a sort of high performance, evaporative coating for the alleviation of the surface degradation issues found so far. In this work, an updated review of worldwide activities on LM research is presented, together with some open issues still remaining and some proposals based on simple physical considerations leading to the optimization of the most conservative alternatives.

Various divertor biasing configurations and improved divertor performance with biasing on Tokamak de Varennes (TdeV)*

DOE Office of Scientific and Technical Information (OSTI.GOV)

Decoste, R.; Lachambre, J.; Abel, G.

1994-05-01

Electrically insulated divertor plates are used on TdeV (Tokamak de Varennes) [18[ital th] [ital EPS] [ital Conference] [ital on] [ital Controlled] [ital Fusion] [ital and] [ital Plasma] [ital Physics] Berlin (European Physical Society, Petit-Lancy, 1991), Vol. 15C, Part I, pp. 1--141] to produce various biasing configurations, which can be decomposed into two basic modes. Plasma biasing, with a radial electric field [ital E][sub [ital r

Upgraded divertor Thomson scattering system on DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Glass, F., E-mail: glassf@fusion.gat.com; Carlstrom, T. N.; Du, D.

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, beforemore » 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.« less

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.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dr. Ricardo Maqueda; Dr. Fred M. Levinton

Nova Photonics, Inc. has a collaborative effort at the National Spherical Torus Experiment (NSTX). This collaboration, based on fast imaging of visible phenomena, has provided key insights on edge turbulence, intermittency, and edge phenomena such as edge localized modes (ELMs) and multi-faceted axisymmetric radiation from the edge (MARFE). Studies have been performed in all these areas. The edge turbulence/intermittency studies make use of the Gas Puff Imaging diagnostic developed by the Principal Investigator (Ricardo Maqueda) together with colleagues from PPPL. This effort is part of the International Tokamak Physics Activity (ITPA) edge, scrape-off layer and divertor group joint activity (DSOL-15:more » Inter-machine comparison of blob characteristics). The edge turbulence/blob study has been extended from the current location near the midplane of the device to the lower divertor region of NSTX. The goal of this effort was to study turbulence born blobs in the vicinity of the X-point region and their circuit closure on divertor sheaths or high density regions in the divertor. In the area of ELMs and MARFEs we have studied and characterized the mode structure and evolution of the ELM types observed in NSTX, as well as the study of the observed interaction between MARFEs and ELMs. This interaction could have substantial implications for future devices where radiative divertor regions are required to maintain detachment from the divertor plasma facing components.« less

Geographic Variations and Time Trends in Cancer Treatments in Taiwan.

PubMed

Hsu, Jason C; Chang, Sheng-Mao; Lu, Christine Y

2017-08-02

Targeted therapies have become important treatment options for cancer care in many countries. This study aimed to examine recent trends in utilization of antineoplastic drugs, particularly the use of targeted therapies for treatment of cancer, by geographic region in Taiwan (northern, midwestern, southern, and eastern regions and the outer islands). This was a retrospective observational study of antineoplastic agents using 2009-2012 quarterly claims data from Taiwan's National Health Insurance Research Database. Yearly market shares by prescription volume and costs for targeted therapies among total antineoplastic agents by region were estimated. We used multivariate regression model and ANOVA to examine variations in utilization of targeted therapies between geographic regions and used ARIMA models to estimate longitudinal trends. Population-adjusted use and costs of antineoplastic drugs (including targeted therapies) were highest in the southern region of Taiwan and lowest in the outer islands. We found a 4-fold difference in use of antineoplastic drugs and a 49-fold difference in use of targeted therapies between regions if the outer islands were included. There were minimal differences in use of antineoplastic drugs between other regions with about a 2-fold difference in use of targeted therapies. Without considering the outer islands, the market share by prescription volume and costs of targeted therapies increased almost 2-fold (1.84-1.90) and 1.5-fold (1.26-1.61) respectively between 2009 and 2012. Furthermore, region was not significantly associated with use of antineoplastic agents or use of targeted therapies after adjusting for confounders. Region was associated with costs of antineoplastic agents but it was not associated with costs of targeted therapies after confounding adjustments. Use of antineoplastic drugs overall and use of targeted therapies for treatment of cancer varied somewhat between regions in Taiwan; use was notably low in the outer islands. Strategies might be needed to ensure access to cancer care in each region as economic burden of cancer care increase due to growing use of targeted therapies.

Outer membrane vesicles from Neisseria gonorrhoeae target PorB to mitochondria and induce apoptosis

PubMed Central

Elgass, Kirstin D.; Gabriel, Kipros; Dougan, Gordon; Lithgow, Trevor; Heinz, Eva

2018-01-01

Neisseria gonorrhoeae causes the sexually transmitted disease gonorrhoea by evading innate immunity. Colonizing the mucosa of the reproductive tract depends on the bacterial outer membrane porin, PorB, which is essential for ion and nutrient uptake. PorB is also targeted to host mitochondria and regulates apoptosis pathways to promote infections. How PorB traffics from the outer membrane of N. gonorrhoeae to mitochondria and whether it modulates innate immune cells, such as macrophages, remains unclear. Here, we show that N. gonorrhoeae secretes PorB via outer membrane vesicles (OMVs). Purified OMVs contained primarily outer membrane proteins including oligomeric PorB. The porin was targeted to mitochondria of macrophages after exposure to purified OMVs and wild type N. gonorrhoeae. This was associated with loss of mitochondrial membrane potential, release of cytochrome c, activation of apoptotic caspases and cell death in a time-dependent manner. Consistent with this, OMV-induced macrophage death was prevented with the pan-caspase inhibitor, Q-VD-PH. This shows that N. gonorrhoeae utilizes OMVs to target PorB to mitochondria and to induce apoptosis in macrophages, thus affecting innate immunity. PMID:29601598

Integrated simulations of H-mode operation in ITER including core fuelling, divertor detachment and ELM control

NASA Astrophysics Data System (ADS)

Polevoi, A. R.; Loarte, A.; Dux, R.; Eich, T.; Fable, E.; Coster, D.; Maruyama, S.; Medvedev, S. Yu.; Köchl, F.; Zhogolev, V. E.

2018-05-01

ELM mitigation to avoid melting of the tungsten (W) divertor is one of the main factors affecting plasma fuelling and detachment control at full current for high Q operation in ITER. Here we derive the ITER operational space, where ELM mitigation to avoid melting of the W divertor monoblocks top surface is not required and appropriate control of W sources and radiation in the main plasma can be ensured through ELM control by pellet pacing. We apply the experimental scaling that relates the maximum ELM energy density deposited at the divertor with the pedestal parameters and this eliminates the uncertainty related with the ELM wetted area for energy deposition at the divertor and enables the definition of the ITER operating space through global plasma parameters. Our evaluation is thus based on this empirical scaling for ELM power loads together with the scaling for the pedestal pressure limit based on predictions from stability codes. In particular, our analysis has revealed that for the pedestal pressure predicted by the EPED1  +  SOLPS scaling, ELM mitigation to avoid melting of the W divertor monoblocks top surface may not be required for 2.65 T H-modes with normalized pedestal densities (to the Greenwald limit) larger than 0.5 to a level of current of 6.5–7.5 MA, which depends on assumptions on the divertor power flux during ELMs and between ELMs that expand the range of experimental uncertainties. The pellet and gas fuelling requirements compatible with control of plasma detachment, core plasma tungsten accumulation and H-mode operation (including post-ELM W transient radiation) have been assessed by 1.5D transport simulations for a range of assumptions regarding W re-deposition at the divertor including the most conservative assumption of zero prompt re-deposition. With such conservative assumptions, the post-ELM W transient radiation imposes a very stringent limit on ELM energy losses and the associated minimum required ELM frequency. Depending on W transport assumptions during the ELM, a maximum ELM frequency is also identified above which core tungsten accumulation takes place.

Modeling carbon production and transport during ELMs in DIII-D

NASA Astrophysics Data System (ADS)

Hogan, J.; Wade, M.; Coster, D.; Lasnier, C.

2004-11-01

Large-scale Type I ELM events could provide a significant C source in ITER, and C production rates depend on incident D flux density and surface temperature, quantities which can vary significantly during an ELM event. Recent progress on DIII-D has improved opportunities for code comparison. Fast time-scale measurements of divertor CIII evolution [1] and fast edge CER measurements of C profile evolution during low-density DIII-D LSN ELMy H-modes (type I) [2] have been modeled using the solps5.0/Eirene99 coupled edge code and time dependent thermal analysis codes. An ELM model based on characteristics of MHD peeling-ballooning modes reproduces the pedestal evolution. Qualitative agreement for the CIII evolution during an ELM event is found using the Roth et al annealing model for chemical sputtering and the sensitivity to other models is described. Significant ELM-to-ELM variations in observed maximum divertor target IR temperature during nominally identical ELMs are investigated with models for C emission from micron-scale dust particles. [1] M Groth, M Fenstermacher et al J Nucl Mater 2003, [2] M Wade, K Burrell et al PSI-16

Overview of the EUROfusion Medium Size Tokamak scientific program

NASA Astrophysics Data System (ADS)

Bernert, Matthias; Bolzonella, Tommaso; Coda, Stefano; Hakola, Antti; Meyer, Hendrik; Eurofusion Mst1 Team; Tcv Team; Mast-U Team; ASDEX Upgrade Team

2017-10-01

Under the EUROfusion MST1 program, coordinated experiments are conducted at three European medium sized tokamaks (ASDEX Upgrade, TCV and MAST-U). It complements the JET program for preparing a safe and efficient operation for ITER and DEMO. Work under MST1 benefits from cross-machine comparisons but also makes use of the unique capabilities of each device. For the 2017/2018 campaign 25 topic areas were defined targeting three main objectives: 1) Development towards an edge and wall compatible H-mode scenario with small or no ELMs. 2) Investigation of disruptions in order to achieve better predictions and improve avoidance or mitigation schemes. 3) Exploring conventional and alternative divertor configurations for future high P/R scenarios. This contribution will give an overview of the work done under MST1 exemplified by the highlight results for each top objective from the last campaigns, such as evaluation of natural small ELM scenarios, runaway mitigation and control, assessment of detachment in alternative divertor configurations and highly radiative scenarios. See author list of ``H. Meyer et al. 2017 Nucl. Fusion 57, 102014''.

Tungsten dust impact on ITER-like plasma edge

DOE PAGES

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

2015-01-12

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

Measurements of W Erosion using UV Emission from DIII-D and CTH

NASA Astrophysics Data System (ADS)

Johnson, Curtis; Ennis, David; Loch, Stuart; Balance, Connor; Victor, Brian; Allen, Steve; Samuell, Cameron; Abrams, Tyler; Unterberg, Ezekial

2017-10-01

of Plasma Facing Components (PFCs) will play a critical role in establishing the performance of reactor-relevant fusion devices, particularly for tungsten (W) divertor targets. Erosion can be diagnosed from spectral line emission together with atomic coefficients representing the `ionizations per photon' (S/XB). Emission from W I is most intense in the UV region. Thus, UV survey spectrometers (200-400 nm) are used to diagnose W PFCs erosion in the DIII-D divertor and from a W tipped probe in the CTH experiment. Nineteen W emission lines in the UV region are identified between the two experiments, allowing for multiple S/XB erosion measurements. Initial W erosion measurements are compared to erosion using the 400.9 nm W I line. Complete UV spectra will be presented and compared to synthetic spectra for varying plasma conditions. Analysis of the metastable states impact on the S/XB will be presented as well as possible electron temperature and density diagnosis from W I line ratios. Work supported by USDOE Grants DE-SC0015877 & DE-FC02-04ER54698.

Optimizing stellarator coil winding surfaces with Regcoil

NASA Astrophysics Data System (ADS)

Bader, Aaron; Landreman, Matt; Anderson, David; Hegna, Chris

2017-10-01

We show initial attempts at optimizing a coil winding surface using the Regcoil code [1] for selected quasi helically symmetric equilibria. We implement a generic optimization scheme which allows for variation of the winding surface to allow for improved diagnostic access and allow for flexible divertor solutions. Regcoil and similar coil-solving algorithms require a user-input winding surface, on which the coils lie. Simple winding surfaces created by uniformly expanding the plasma boundary may not be ideal. Engineering constraints on reactor design require a coil-plasma separation sufficient for the introduction of neutron shielding and a tritium generating blanket. This distance can be the limiting factor in determining reactor size. Furthermore, expanding coils in other regions, where possible, can be useful for diagnostic and maintenance access along with providing sufficient room for a divertor. We minimize a target function that includes as constraints, the minimum coil-plasma distance, the winding surface volume, and the normal magnetic field on the plasma boundary. Results are presented for two quasi-symmetric equilibria at different aspect ratios. Work supported by the US DOE under Grant DE-FG02-93ER54222.

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.

Divertor with a third-order null of the poloidal field

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

2013-09-15

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

Initial development of the DIII–D snowflake divertor control

NASA Astrophysics Data System (ADS)

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

2018-06-01

Simultaneous control of two proximate magnetic field nulls in the divertor region is demonstrated on DIII–D to enable plasma operations in an advanced magnetic configuration known as the snowflake divertor (SFD). The SFD is characterized by a second-order poloidal field null, created by merging two first-order nulls of the standard divertor configuration. The snowflake configuration has many magnetic properties, such as high poloidal flux expansion, large plasma-wetted area, and additional strike points, that are advantageous for divertor heat flux management in future fusion reactors. However, the magnetic configuration of the SFD is highly-sensitive to changes in currents within the plasma and external coils and therefore requires complex magnetic control. The first real-time snowflake detection and control system on DIII–D has been implemented in order to stabilize the configuration. The control algorithm calculates the position of the two nulls in real-time by locally-expanding the Grad–Shafranov equation in the divertor region. A linear relation between variations in the poloidal field coil currents and changes in the null locations is then analytically derived. This formulation allows for simultaneous control of multiple coils to achieve a desired SFD configuration. It is shown that the control enabled various snowflake configurations on DIII–D in scenarios such as the double-null advanced tokamak. The SFD resulted in a 2.5×  reduction in the peak heat flux for many energy confinement times (2–3 s) without any adverse effects on core plasma performance.

Sensitivity of WallDYN material migration modeling to uncertainties in mixed-material surface binding energies

DOE PAGES

Nichols, J. H.; Jaworski, M. A.; Schmid, K.

2017-03-09

The WallDYN package has recently been applied to a number of tokamaks to self-consistently model the evolution of mixed-material plasma facing surfaces. A key component of the WallDYN model is the concentration-dependent surface sputtering rate, calculated using SDTRIM.SP. This modeled sputtering rate is strongly influenced by the surface binding energies (SBEs) of the constituent materials, which are well known for pure elements but often are poorly constrained for mixed-materials. This work examines the sensitivity of WallDYN surface evolution calculations to different models for mixed-material SBEs, focusing on the carbon/lithium/oxygen/deuterium system present in NSTX. A realistic plasma background is reconstructed frommore » a high density, H-mode NSTX discharge, featuring an attached outer strike point with local density and temperature of 4 × 10 20 m -3 and 4 eV, respectively. It is found that various mixed-material SBE models lead to significant qualitative and quantitative changes in the surface evolution profile at the outer divertor, with the highest leverage parameter being the C-Li binding model. Uncertainties of order 50%, appearing on time scales relevant to tokamak experiments, highlight the importance of choosing an appropriate mixed-material sputtering representation when modeling the surface evolution of plasma facing components. Lastly, these results are generalized to other fusion-relevant materials with different ranges of SBEs.« less

Sensitivity of WallDYN material migration modeling to uncertainties in mixed-material surface binding energies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nichols, J. H.; Jaworski, M. A.; Schmid, K.

The WallDYN package has recently been applied to a number of tokamaks to self-consistently model the evolution of mixed-material plasma facing surfaces. A key component of the WallDYN model is the concentration-dependent surface sputtering rate, calculated using SDTRIM.SP. This modeled sputtering rate is strongly influenced by the surface binding energies (SBEs) of the constituent materials, which are well known for pure elements but often are poorly constrained for mixed-materials. This work examines the sensitivity of WallDYN surface evolution calculations to different models for mixed-material SBEs, focusing on the carbon/lithium/oxygen/deuterium system present in NSTX. A realistic plasma background is reconstructed frommore » a high density, H-mode NSTX discharge, featuring an attached outer strike point with local density and temperature of 4 × 10 20 m -3 and 4 eV, respectively. It is found that various mixed-material SBE models lead to significant qualitative and quantitative changes in the surface evolution profile at the outer divertor, with the highest leverage parameter being the C-Li binding model. Uncertainties of order 50%, appearing on time scales relevant to tokamak experiments, highlight the importance of choosing an appropriate mixed-material sputtering representation when modeling the surface evolution of plasma facing components. Lastly, these results are generalized to other fusion-relevant materials with different ranges of SBEs.« less

Power exhaust scenarios and control for projected high-power NSTX-U operation

NASA Astrophysics Data System (ADS)

Menard, Jonathan; Gerhardt, S. P.; Myers, C. E.; Reinke, M. L.; Brooks, A.; Mardenfeld, M.; NSTX Upgrade Team

2017-10-01

An important goal of the NSTX Upgrade (NSTX-U) research program is to characterize energy confinement in the low-aspect-ratio spherical tokamak configuration over a significantly expanded range of plasma current, toroidal field, and heating power, while increasing flattop durations up to 5 seconds. However, the narrowing of the scrape-off layer at higher current combined with an improved understanding of expected halo-current loads has motivated a significant re-design of NSTX-U plasma facing components in the high-heat-flux regions of the divertor. In order to reduce the expected divertor heat flux to acceptable levels, a combination of mitigation techniques will be used: increased divertor poloidal flux expansion, increased divertor radiation, and controlled strike-point sweeping. The machine requirements for these various mitigation techniques are studied here using a newly implemented reduced heat-flux model. Systematic equilibrium scans are used to quantify the required divertor coil currents and to verify vertical stability for a range of plasma shapes. Free-boundary control schemes to constrain the strike-point location and field-line angle-of-incidence will also be discussed. Work supported by DOE contract DE-AC02- 09CH11466.

A mechanism for large divertor plasma energy loss via lithium radiation in tokamaks

NASA Astrophysics Data System (ADS)

Rognlien, T. D.; Meier, E. T.; Soukhanovskii, V. A.

2012-10-01

Lithium has been used as a wall-conditioning element in a number of tokamaks over the years, including TFTR, FTU, and NSTX, where core plasma energy confinement and particle control are often found to improve following such conditioning. Here the possible role of Li in providing substantial energy loss for divertor plasmas via line radiation is reported. A multi-charge-state 2D UEDGE fluid model is used where the hydrogenic and Li ions and neutrals are each evolved as separate species and separate equations are solved for the electron and ion temperatures. It is shown that a sufficient level of Li neutrals evolving from the divertor surface via sputtering or evaporation can induce energy detachment of the divertor plasma, yielding a strongly radiating zone near the divertor where ionization and recombination from/to neutral Li can radiate most of the power flowing into the scrape-off layer while maintaining low core contamination. A local peaking of Li emissivity for electron temperatures near 1 eV appears to play an important role in the detachment of the mixed deuterium/Li plasma. Evidence of such behavior from NSTX discharges will be discussed.

Plasma transport in a simulated magnetic-divertor configuration

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 tomore » 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.« less

Initial development of the DIII–D snowflake divertor control

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kolemen, Egemen; Vail, P. J.; Makowski, M. A.

Simultaneous control of two proximate magnetic field nulls in the divertor region is demonstrated on DIII–D to enable plasma operations in an advanced magnetic configuration known as the snowflake divertor (SFD). The SFD is characterized by a second-order poloidal field null, created by merging two first-order nulls of the standard divertor configuration. The snowflake configuration has many magnetic properties, such as high poloidal flux expansion, large plasma-wetted area, and additional strike points, that are advantageous for divertor heat flux management in future fusion reactors. However, the magnetic configuration of the SFD is highly-sensitive to changes in currents within the plasmamore » and external coils and therefore requires complex magnetic control. The first real-time snowflake detection and control system on DIII–D has been implemented in order to stabilize the configuration. The control algorithm calculates the position of the two nulls in real-time by locally-expanding the Grad–Shafranov equation in the divertor region. A linear relation between variations in the poloidal field coil currents and changes in the null locations is then analytically derived. This formulation allows for simultaneous control of multiple coils to achieve a desired SFD configuration. It is shown that the control enabled various snowflake configurations on DIII–D in scenarios such as the double-null advanced tokamak. In conclusion, the SFD resulted in a 2.5×reduction in the peak heat flux for many energy confinement times (2–3s) without any adverse effects on core plasma performance.« less

Initial development of the DIII–D snowflake divertor control

DOE PAGES

Kolemen, Egemen; Vail, P. J.; Makowski, M. A.; ...

2018-04-11

Simultaneous control of two proximate magnetic field nulls in the divertor region is demonstrated on DIII–D to enable plasma operations in an advanced magnetic configuration known as the snowflake divertor (SFD). The SFD is characterized by a second-order poloidal field null, created by merging two first-order nulls of the standard divertor configuration. The snowflake configuration has many magnetic properties, such as high poloidal flux expansion, large plasma-wetted area, and additional strike points, that are advantageous for divertor heat flux management in future fusion reactors. However, the magnetic configuration of the SFD is highly-sensitive to changes in currents within the plasmamore » and external coils and therefore requires complex magnetic control. The first real-time snowflake detection and control system on DIII–D has been implemented in order to stabilize the configuration. The control algorithm calculates the position of the two nulls in real-time by locally-expanding the Grad–Shafranov equation in the divertor region. A linear relation between variations in the poloidal field coil currents and changes in the null locations is then analytically derived. This formulation allows for simultaneous control of multiple coils to achieve a desired SFD configuration. It is shown that the control enabled various snowflake configurations on DIII–D in scenarios such as the double-null advanced tokamak. In conclusion, the SFD resulted in a 2.5×reduction in the peak heat flux for many energy confinement times (2–3s) without any adverse effects on core plasma performance.« less

Targeting and assembly of components of the TOC protein import complex at the chloroplast outer envelope membrane

PubMed Central

Richardson, Lynn G. L.; Paila, Yamuna D.; Siman, Steven R.; Chen, Yi; Smith, Matthew D.; Schnell, Danny J.

2014-01-01

The translocon at the outer envelope membrane of chloroplasts (TOC) initiates the import of thousands of nuclear encoded preproteins required for chloroplast biogenesis and function. The multimeric TOC complex contains two GTP-regulated receptors, Toc34 and Toc159, which recognize the transit peptides of preproteins and initiate protein import through a β–barrel membrane channel, Toc75. Different isoforms of Toc34 and Toc159 assemble with Toc75 to form structurally and functionally diverse translocons, and the composition and levels of TOC translocons is required for the import of specific subsets of coordinately expressed proteins during plant growth and development. Consequently, the proper assembly of the TOC complexes is key to ensuring organelle homeostasis. This review will focus on our current knowledge of the targeting and assembly of TOC components to form functional translocons at the outer membrane. Our analyses reveal that the targeting of TOC components involves elements common to the targeting of other outer membrane proteins, but also include unique features that appear to have evolved to specifically facilitate assembly of the import apparatus. PMID:24966864

Targeting and assembly of components of the TOC protein import complex at the chloroplast outer envelope membrane.

PubMed

Richardson, Lynn G L; Paila, Yamuna D; Siman, Steven R; Chen, Yi; Smith, Matthew D; Schnell, Danny J

2014-01-01

The translocon at the outer envelope membrane of chloroplasts (TOC) initiates the import of thousands of nuclear encoded preproteins required for chloroplast biogenesis and function. The multimeric TOC complex contains two GTP-regulated receptors, Toc34 and Toc159, which recognize the transit peptides of preproteins and initiate protein import through a β-barrel membrane channel, Toc75. Different isoforms of Toc34 and Toc159 assemble with Toc75 to form structurally and functionally diverse translocons, and the composition and levels of TOC translocons is required for the import of specific subsets of coordinately expressed proteins during plant growth and development. Consequently, the proper assembly of the TOC complexes is key to ensuring organelle homeostasis. This review will focus on our current knowledge of the targeting and assembly of TOC components to form functional translocons at the outer membrane. Our analyses reveal that the targeting of TOC components involves elements common to the targeting of other outer membrane proteins, but also include unique features that appear to have evolved to specifically facilitate assembly of the import apparatus.

Conceptual design of a pre-loaded liquid lithium divertor target for NSTX-U

DOE PAGES

Rindt, P.; Lopes Cardozo, N. J.; van Dommelen, J. A. W.; ...

2016-09-03

In this study, a conceptual design for a pre-filled liquid lithium divertor target for the National Spherical Torus Experiment Upgrade (NSTX-U) is presented. The design is aimed at facilitating experiments with high lithium flux from the plasma facing components (PFCs) in NSTX-U and investigating the potential of capillary based liquid lithium components. In the design, lithium is supplied from a reservoir in the PFC to the plasma facing surface via capillary action in a wicking structure. This working principle is also demonstrated experimentally. Next, a titanium zirconium molybdenum (TZM) prototype design is presented, required to withstand a steady state heatmore » flux peaking at 10 MW m –2 for 5 s and edge localized modes depositing (130 kJ in 2 ms at 10 Hz). The main challenge is to sufficiently reduce the thermal stresses. This is achieved by dividing the surface into brushes and filling the slots in between with liquid lithium. The principle of using this liquid “interlayer” allows for thermal expansion and simultaneously heat conduction, and could be used to significantly reduce the demands to solids in future PFCs. Lithium flow to the surface is analyzed using a novel analytical model, ideally suited for design purposes. Thermal stresses in the PFC are analyzed using the finite element method. As a result, the requirements are met, and thus a prototype will be manufactured for physical testing.« less

Conceptual design of a pre-loaded liquid lithium divertor target for NSTX-U

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rindt, P.; Lopes Cardozo, N. J.; van Dommelen, J. A. W.

In this study, a conceptual design for a pre-filled liquid lithium divertor target for the National Spherical Torus Experiment Upgrade (NSTX-U) is presented. The design is aimed at facilitating experiments with high lithium flux from the plasma facing components (PFCs) in NSTX-U and investigating the potential of capillary based liquid lithium components. In the design, lithium is supplied from a reservoir in the PFC to the plasma facing surface via capillary action in a wicking structure. This working principle is also demonstrated experimentally. Next, a titanium zirconium molybdenum (TZM) prototype design is presented, required to withstand a steady state heatmore » flux peaking at 10 MW m –2 for 5 s and edge localized modes depositing (130 kJ in 2 ms at 10 Hz). The main challenge is to sufficiently reduce the thermal stresses. This is achieved by dividing the surface into brushes and filling the slots in between with liquid lithium. The principle of using this liquid “interlayer” allows for thermal expansion and simultaneously heat conduction, and could be used to significantly reduce the demands to solids in future PFCs. Lithium flow to the surface is analyzed using a novel analytical model, ideally suited for design purposes. Thermal stresses in the PFC are analyzed using the finite element method. As a result, the requirements are met, and thus a prototype will be manufactured for physical testing.« less

Coupled two-dimensional edge plasma and neutral gas modeling of tokamak scrape-off-layers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maingi, Rajesh

1992-08-01

The objective of this study is to devise a detailed description of the tokamak scrape-off-layer (SOL), which includes the best available models of both the plasma and neutral species and the strong coupling between the two in many SOL regimes. A good estimate of both particle flux and heat flux profiles at the limiter/divertor target plates is desired. Peak heat flux is one of the limiting factors in determining the survival probability of plasma-facing-components at high power levels. Plate particle flux affects the neutral flux to the pump, which determines the particle exhaust rate. A technique which couples a two-dimensionalmore » (2-D) plasma and a 2-D neutral transport code has been developed (coupled code technique), but this procedure requires large amounts of computer time. Relevant physics has been added to an existing two-neutral-species model which takes the SOL plasma/neutral coupling into account in a simple manner (molecular physics model), and this model is compared with the coupled code technique mentioned above. The molecular physics model is benchmarked against experimental data from a divertor tokamak (DIII-D), and a similar model (single-species model) is benchmarked against data from a pump-limiter tokamak (Tore Supra). The models are then used to examine two key issues: free-streaming-limits (ion energy conduction and momentum flux) and the effects of the non-orthogonal geometry of magnetic flux surfaces and target plates on edge plasma parameter profiles.« less

Experimentally-based ExB drifts in the DIII-D divertor and SOL calculated from integration of Ohm's law using Thomson scattering measurements of T e and n e

DOE PAGES

Stangeby, Peter C.; Elder, J. David; McLean, Adam G.; ...

2017-03-27

We calculated the 2D spatial distributions of cross field drift velocities from 2D Thomson scattering measurements of T e and n e in the divertor and SOL of DIII-D. In contrast with the method that has been used on DIII-D where the 2D distribution of plasma potential V plasma is obtained from measurements of the probe floating potential of reciprocating probes, the present method does not require insertion of a probe into the plasma and can therefore be used in high power discharges. Furthermore, the 2D spatial distribution of V plasma is calculated from Ohm’s Law for the parallel electricmore » field E || along each flux tube, E || s || = -1.71dT e/ds || - T e/n edn e/ds ||, where the Thomson scattering values of T e and n e are used. To within a constant of integration, V plasma is obtained by integrating E || along the flux-tubes (field lines); the constant is obtained for each flux tube using the sheath drop at the target calculated from the characteristic of Langmuir probes built into the divertor tiles. The 2D distributions of E./01/2 = -dV4/ds./01/2, E452510/2 = -dV4/ds452510/2, v789 452510/2 = E./01/2/B and v789 ./01/2 = E452510/2/B are then calculated as well as the particle drift flux densities Γ789 452510/2 = nv789 452510/2 and Γ789 ./01/2 = nv789 ./01/2 for electrons, fuel ions and impurity ions, using the appropriate values of particle density, n.« less

Tungsten and beryllium armour development for the JET ITER-like wall project

NASA Astrophysics Data System (ADS)

Maier, H.; Hirai, T.; Rubel, M.; Neu, R.; Mertens, Ph.; Greuner, H.; Hopf, Ch.; Matthews, G. F.; Neubauer, O.; Piazza, G.; Gauthier, E.; Likonen, J.; Mitteau, R.; Maddaluno, G.; Riccardi, B.; Philipps, V.; Ruset, C.; Lungu, C. P.; Uytdenhouwen, I.; EFDA contributors, JET

2007-03-01

For the ITER-like wall project at JET the present main chamber CFC tiles will be exchanged with Be tiles and in parallel a fully tungsten-clad divertor will be prepared. Therefore three R&D programmes were initiated: Be coatings on Inconel as well as Be erosion markers were developed for the first wall of the main chamber. High heat flux screening and cyclic loading tests carried out on the Be coatings on Inconel showed excellent performance, above the required power and energy density. For the divertor a conceptual design for a bulk W horizontal target plate was investigated, with the emphasis on minimizing electromagnetic forces. The design consisted of stacks of W lamellae of 6 mm width that were insulated in the toroidal direction. High heat flux tests of a test module were performed with an electron beam at an absorbed power density up to 9 MW m-2 for more than 150 pulses and finally with increasing power loads leading to surface temperatures in excess of 3000 °C. No macroscopic failure occurred during the test while SEM showed the development of micro-cracks on the loaded surface. For all other divertor parts R&D was performed to provide the technology to coat the 2-directional CFC material used at JET with thin tungsten coatings. The W-coated CFC tiles were subjected to heat loads with power densities ranging up to 23.5 MW m-2 and exposed to cyclic heat loading for 200 pulses at 10.5 MW m-2. All coatings developed cracks perpendicular to the CFC fibres due to the stronger contraction of the coating upon cool-down after the heat pulses.

Experimentally-based ExB drifts in the DIII-D divertor and SOL calculated from integration of Ohm's law using Thomson scattering measurements of T e and n e

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stangeby, Peter C.; Elder, J. David; McLean, Adam G.

We calculated the 2D spatial distributions of cross field drift velocities from 2D Thomson scattering measurements of T e and n e in the divertor and SOL of DIII-D. In contrast with the method that has been used on DIII-D where the 2D distribution of plasma potential V plasma is obtained from measurements of the probe floating potential of reciprocating probes, the present method does not require insertion of a probe into the plasma and can therefore be used in high power discharges. Furthermore, the 2D spatial distribution of V plasma is calculated from Ohm’s Law for the parallel electricmore » field E || along each flux tube, E || s || = -1.71dT e/ds || - T e/n edn e/ds ||, where the Thomson scattering values of T e and n e are used. To within a constant of integration, V plasma is obtained by integrating E || along the flux-tubes (field lines); the constant is obtained for each flux tube using the sheath drop at the target calculated from the characteristic of Langmuir probes built into the divertor tiles. The 2D distributions of E./01/2 = -dV4/ds./01/2, E452510/2 = -dV4/ds452510/2, v789 452510/2 = E./01/2/B and v789 ./01/2 = E452510/2/B are then calculated as well as the particle drift flux densities Γ789 452510/2 = nv789 452510/2 and Γ789 ./01/2 = nv789 ./01/2 for electrons, fuel ions and impurity ions, using the appropriate values of particle density, n.« less

Twenty Years of Research on the Alcator C-Mod Tokamak

NASA Astrophysics Data System (ADS)

Greenwald, Martin

2013-10-01

Alcator C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since its start in 1993, contributing data that extended tests of critical physical models into new parameter ranges and into new regimes. Using only RF for heating and current drive with innovative launching structures, C-Mod operates routinely at very high power densities. Research highlights include direct experimental observation of ICRF mode-conversion, ICRF flow drive, demonstration of Lower-Hybrid current drive at ITER-like densities and fields and, using a set of powerful new diagnostics, extensive validation of advanced RF codes. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing components--an approach adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated the critical role of cross-field transport in divertor operation, edge flows and the tokamak density limit. C-Mod developed the I-mode and EDA H-mode regimes which have high performance without large ELMs and with pedestal transport self-regulated by short-wavelength electromagnetic waves. C-Mod has carried out pioneering studies of intrinsic rotation and found that self-generated flow shear can be strong enough to significantly modify transport. C-Mod made the first quantitative link between pedestal temperature and H-mode performance, showing that the observed self-similar temperature profiles were consistent with critical-gradient-length theories and followed up with quantitative tests of nonlinear gyrokinetic models. Disruption studies on C-Mod provided the first observation of non-axisymmetric halo currents and non-axisymmetric radiation in mitigated disruptions. Work supported by U.S. DoE

Physics basis for an advanced physics and advanced technology tokamak power plant configuration: ARIES-ACT1

DOE PAGES

Kessel, C. E.; Poli, F. M.; Ghantous, K.; ...

2015-01-01

Here, the advanced physics and advanced technology tokamak power plant ARIES-ACT1 has a major radius of 6.25 m at an aspect ratio of 4.0, toroidal field of 6.0 T, strong shaping with elongation of 2.2, and triangularity of 0.63. The broadest pressure cases reached wall-stabilized β N ~ 5.75, limited by n = 3 external kink mode requiring a conducting shell at b/a = 0.3, requiring plasma rotation, feedback, and/or kinetic stabilization. The medium pressure peaking case reaches β N = 5.28 with B T = 6.75, while the peaked pressure case reaches β N < 5.15. Fast particle magnetohydrodynamicmore » stability shows that the alpha particles are unstable, but this leads to redistribution to larger minor radius rather than loss from the plasma. Edge and divertor plasma modeling shows that 75% of the power to the divertor can be radiated with an ITER-like divertor geometry, while >95% can be radiated in a stable detached mode with an orthogonal target and wide slot geometry. The bootstrap current fraction is 91% with a q95 of 4.5, requiring ~1.1 MA of external current drive. This current is supplied with 5 MW of ion cyclotron radio frequency/fast wave and 40 MW of lower hybrid current drive. Electron cyclotron is most effective for safety factor control over ρ~0.2 to 0.6 with 20 MW. The pedestal density is ~0.9×10 20/m 3, and the temperature is ~4.4 keV. The H98 factor is 1.65, n/n Gr = 1.0, and the ratio of net power to threshold power is 2.8 to 3.0 in the flattop.« less

Effect of the resonant magnetic perturbation on the plasma parameters in COMPASS tokamak’s divertor region

NASA Astrophysics Data System (ADS)

Dimitrova, M.; Cahyna, P.; Peterka, M.; Hasan, E.; Popov, Tsv K.; Ivanova, P.; Vasileva, E.; Panek, R.; Cavalier, J.; Seidl, J.; Markovic, T.; Havlicek, J.; Dejarnac, R.; Weinzettl, V.; Hacek, P.; Tomes, M.; the COMPASS Team; the EUROfusion MST1 Team

2018-02-01

The resonant magnetic perturbation (RMP) has proven to be a useful way to suppress edge-localized modes that under certain conditions can damage the device by the large power fluxes carried from the bulk plasma to the wall. The effect of RMP on the L-mode plasma parameters in the divertor region of the COMPASS tokamak was studied using the array of 39 Langmuir probes embedded into the divertor target. The current-voltage (IV) probe characteristics were processed by the first-derivative probe technique to obtain the plasma potential and the electron energy distribution function (EEDF) which was approximated by a bi-Maxwellian EEDF with a low-energy (4-6 eV) fraction and a high-energy (11-35 eV) one, the both factions having similar electron density. Clear splitting was observed during the RMP pulse in the low-field-side scrape-off-layer profiles of the floating potential U fl and the ion saturation current density J sat; these two quantities were obtained both by direct continuous measurement and by evaluation of the IV characteristics of probes with swept bias. The negative peaks of U fl induced by RMP spatially overlaps with the local minima of J sat (and n e) rather than with its local maxima which is partly caused by the spatial variation of the plasma potential and partly by the changed shape of the EEDF. The effective temperature of the whole EEDF is not correlated with the negative peaks of U fl, and the profile of the parallel power flux density shows secondary maxima due to RMP which mimic those of J sat.

Overview of progress in European medium sized tokamaks towards an integrated plasma-edge/wall solution

NASA Astrophysics Data System (ADS)

Meyer, H.; Eich, T.; Beurskens, M.; Coda, S.; Hakola, A.; Martin, P.; Adamek, J.; Agostini, M.; Aguiam, D.; Ahn, J.; Aho-Mantila, L.; Akers, R.; Albanese, R.; Aledda, R.; Alessi, E.; Allan, S.; Alves, D.; Ambrosino, R.; Amicucci, L.; Anand, H.; Anastassiou, G.; Andrèbe, Y.; Angioni, C.; Apruzzese, G.; Ariola, M.; Arnichand, H.; Arter, W.; Baciero, A.; Barnes, M.; Barrera, L.; Behn, R.; Bencze, A.; Bernardo, J.; Bernert, M.; Bettini, P.; Bilková, P.; Bin, W.; Birkenmeier, G.; Bizarro, J. P. S.; Blanchard, P.; Blanken, T.; Bluteau, M.; Bobkov, V.; Bogar, O.; Böhm, P.; Bolzonella, T.; Boncagni, L.; Botrugno, A.; Bottereau, C.; Bouquey, F.; Bourdelle, C.; Brémond, S.; Brezinsek, S.; Brida, D.; Brochard, F.; Buchanan, J.; Bufferand, H.; Buratti, P.; Cahyna, P.; Calabrò, G.; Camenen, Y.; Caniello, R.; Cannas, B.; Canton, A.; Cardinali, A.; Carnevale, D.; Carr, M.; Carralero, D.; Carvalho, P.; Casali, L.; Castaldo, C.; Castejón, F.; Castro, R.; Causa, F.; Cavazzana, R.; Cavedon, M.; Cecconello, M.; Ceccuzzi, S.; Cesario, R.; Challis, C. D.; Chapman, I. T.; Chapman, S.; Chernyshova, M.; Choi, D.; Cianfarani, C.; Ciraolo, G.; Citrin, J.; Clairet, F.; Classen, I.; Coelho, R.; Coenen, J. W.; Colas, L.; Conway, G.; Corre, Y.; Costea, S.; Crisanti, F.; Cruz, N.; Cseh, G.; Czarnecka, A.; D'Arcangelo, O.; De Angeli, M.; De Masi, G.; De Temmerman, G.; De Tommasi, G.; Decker, J.; Delogu, R. S.; Dendy, R.; Denner, P.; Di Troia, C.; Dimitrova, M.; D'Inca, R.; Dorić, V.; Douai, D.; Drenik, A.; Dudson, B.; Dunai, D.; Dunne, M.; Duval, B. P.; Easy, L.; Elmore, S.; Erdös, B.; Esposito, B.; Fable, E.; Faitsch, M.; Fanni, A.; Fedorczak, N.; Felici, F.; Ferreira, J.; Février, O.; Ficker, O.; Fietz, S.; Figini, L.; Figueiredo, A.; Fil, A.; Fishpool, G.; Fitzgerald, M.; Fontana, M.; Ford, O.; Frassinetti, L.; Fridström, R.; Frigione, D.; Fuchert, G.; Fuchs, C.; Furno Palumbo, M.; Futatani, S.; Gabellieri, L.; Gałązka, K.; Galdon-Quiroga, J.; Galeani, S.; Gallart, D.; Gallo, A.; Galperti, C.; Gao, Y.; Garavaglia, S.; Garcia, J.; Garcia-Carrasco, A.; Garcia-Lopez, J.; Garcia-Munoz, M.; Gardarein, J.-L.; Garzotti, L.; Gaspar, J.; Gauthier, E.; Geelen, P.; Geiger, B.; Ghendrih, P.; Ghezzi, F.; Giacomelli, L.; Giannone, L.; Giovannozzi, E.; Giroud, C.; Gleason González, C.; Gobbin, M.; Goodman, T. P.; Gorini, G.; Gospodarczyk, M.; Granucci, G.; Gruber, M.; Gude, A.; Guimarais, L.; Guirlet, R.; Gunn, J.; Hacek, P.; Hacquin, S.; Hall, S.; Ham, C.; Happel, T.; Harrison, J.; Harting, D.; Hauer, V.; Havlickova, E.; Hellsten, T.; Helou, W.; Henderson, S.; Hennequin, P.; Heyn, M.; Hnat, B.; Hölzl, M.; Hogeweij, D.; Honoré, C.; Hopf, C.; Horáček, J.; Hornung, G.; Horváth, L.; Huang, Z.; Huber, A.; Igitkhanov, J.; Igochine, V.; Imrisek, M.; Innocente, P.; Ionita-Schrittwieser, C.; Isliker, H.; Ivanova-Stanik, I.; Jacobsen, A. S.; Jacquet, P.; Jakubowski, M.; Jardin, A.; Jaulmes, F.; Jenko, F.; Jensen, T.; Jeppe Miki Busk, O.; Jessen, M.; Joffrin, E.; Jones, O.; Jonsson, T.; Kallenbach, A.; Kallinikos, N.; Kálvin, S.; Kappatou, A.; Karhunen, J.; Karpushov, A.; Kasilov, S.; Kasprowicz, G.; Kendl, A.; Kernbichler, W.; Kim, D.; Kirk, A.; Kjer, S.; Klimek, I.; Kocsis, G.; Kogut, D.; Komm, M.; Korsholm, S. B.; Koslowski, H. R.; Koubiti, M.; Kovacic, J.; Kovarik, K.; Krawczyk, N.; Krbec, J.; Krieger, K.; Krivska, A.; Kube, R.; Kudlacek, O.; Kurki-Suonio, T.; Labit, B.; Laggner, F. M.; Laguardia, L.; Lahtinen, A.; Lalousis, P.; Lang, P.; Lauber, P.; Lazányi, N.; Lazaros, A.; Le, H. B.; Lebschy, A.; Leddy, J.; Lefévre, L.; Lehnen, M.; Leipold, F.; Lessig, A.; Leyland, M.; Li, L.; Liang, Y.; Lipschultz, B.; Liu, Y. Q.; Loarer, T.; Loarte, A.; Loewenhoff, T.; Lomanowski, B.; Loschiavo, V. P.; Lunt, T.; Lupelli, I.; Lux, H.; Lyssoivan, A.; Madsen, J.; Maget, P.; Maggi, C.; Maggiora, R.; Magnussen, M. L.; Mailloux, J.; Maljaars, B.; Malygin, A.; Mantica, P.; Mantsinen, M.; Maraschek, M.; Marchand, B.; Marconato, N.; Marini, C.; Marinucci, M.; Markovic, T.; Marocco, D.; Marrelli, L.; Martin, Y.; Solis, J. R. Martin; Martitsch, A.; Mastrostefano, S.; Mattei, M.; Matthews, G.; Mavridis, M.; Mayoral, M.-L.; Mazon, D.; McCarthy, P.; McAdams, R.; McArdle, G.; McCarthy, P.; McClements, K.; McDermott, R.; McMillan, B.; Meisl, G.; Merle, A.; Meyer, O.; Milanesio, D.; Militello, F.; Miron, I. G.; Mitosinkova, K.; Mlynar, J.; Mlynek, A.; Molina, D.; Molina, P.; Monakhov, I.; Morales, J.; Moreau, D.; Morel, P.; Moret, J.-M.; Moro, A.; Moulton, D.; Müller, H. W.; Nabais, F.; Nardon, E.; Naulin, V.; Nemes-Czopf, A.; Nespoli, F.; Neu, R.; Nielsen, A. H.; Nielsen, S. K.; Nikolaeva, V.; Nimb, S.; Nocente, M.; Nouailletas, R.; Nowak, S.; Oberkofler, M.; Oberparleiter, M.; Ochoukov, R.; Odstrčil, T.; Olsen, J.; Omotani, J.; O'Mullane, M. G.; Orain, F.; Osterman, N.; Paccagnella, R.; Pamela, S.; Pangione, L.; Panjan, M.; Papp, G.; Papřok, R.; Parail, V.; Parra, F. I.; Pau, A.; Pautasso, G.; Pehkonen, S.-P.; Pereira, A.; Perelli Cippo, E.; Pericoli Ridolfini, V.; Peterka, M.; Petersson, P.; Petrzilka, V.; Piovesan, P.; Piron, C.; Pironti, A.; Pisano, F.; Pisokas, T.; Pitts, R.; Ploumistakis, I.; Plyusnin, V.; Pokol, G.; Poljak, D.; Pölöskei, P.; Popovic, Z.; Pór, G.; Porte, L.; Potzel, S.; Predebon, I.; Preynas, M.; Primc, G.; Pucella, G.; Puiatti, M. E.; Pütterich, T.; Rack, M.; Ramogida, G.; Rapson, C.; Rasmussen, J. Juul; Rasmussen, J.; Rattá, G. A.; Ratynskaia, S.; Ravera, G.; Réfy, D.; Reich, M.; Reimerdes, H.; Reimold, F.; Reinke, M.; Reiser, D.; Resnik, M.; Reux, C.; Ripamonti, D.; Rittich, D.; Riva, G.; Rodriguez-Ramos, M.; Rohde, V.; Rosato, J.; Ryter, F.; Saarelma, S.; Sabot, R.; Saint-Laurent, F.; Salewski, M.; Salmi, A.; Samaddar, D.; Sanchis-Sanchez, L.; Santos, J.; Sauter, O.; Scannell, R.; Scheffer, M.; Schneider, M.; Schneider, B.; Schneider, P.; Schneller, M.; Schrittwieser, R.; Schubert, M.; Schweinzer, J.; Seidl, J.; Sertoli, M.; Šesnić, S.; Shabbir, A.; Shalpegin, A.; Shanahan, B.; Sharapov, S.; Sheikh, U.; Sias, G.; Sieglin, B.; Silva, C.; Silva, A.; Silva Fuglister, M.; Simpson, J.; Snicker, A.; Sommariva, C.; Sozzi, C.; Spagnolo, S.; Spizzo, G.; Spolaore, M.; Stange, T.; Stejner Pedersen, M.; Stepanov, I.; Stober, J.; Strand, P.; Šušnjara, A.; Suttrop, W.; Szepesi, T.; Tál, B.; Tala, T.; Tamain, P.; Tardini, G.; Tardocchi, M.; Teplukhina, A.; Terranova, D.; Testa, D.; Theiler, C.; Thornton, A.; Tolias, P.; Tophøj, L.; Treutterer, W.; Trevisan, G. L.; Tripsky, M.; Tsironis, C.; Tsui, C.; Tudisco, O.; Uccello, A.; Urban, J.; Valisa, M.; Vallejos, P.; Valovic, M.; Van den Brand, H.; Vanovac, B.; Varoutis, S.; Vartanian, S.; Vega, J.; Verdoolaege, G.; Verhaegh, K.; Vermare, L.; Vianello, N.; Vicente, J.; Viezzer, E.; Vignitchouk, L.; Vijvers, W. A. J.; Villone, F.; Viola, B.; Vlahos, L.; Voitsekhovitch, I.; Vondráček, P.; Vu, N. M. T.; Wagner, D.; Walkden, N.; Wang, N.; Wauters, T.; Weiland, M.; Weinzettl, V.; Westerhof, E.; Wiesenberger, M.; Willensdorfer, M.; Wischmeier, M.; Wodniak, I.; Wolfrum, E.; Yadykin, D.; Zagórski, R.; Zammuto, I.; Zanca, P.; Zaplotnik, R.; Zestanakis, P.; Zhang, W.; Zoletnik, S.; Zuin, M.; ASDEX Upgrade, the; MAST; TCV Teams

2017-10-01

Integrating the plasma core performance with an edge and scrape-off layer (SOL) that leads to tolerable heat and particle loads on the wall is a major challenge. The new European medium size tokamak task force (EU-MST) coordinates research on ASDEX Upgrade (AUG), MAST and TCV. This multi-machine approach within EU-MST, covering a wide parameter range, is instrumental to progress in the field, as ITER and DEMO core/pedestal and SOL parameters are not achievable simultaneously in present day devices. A two prong approach is adopted. On the one hand, scenarios with tolerable transient heat and particle loads, including active edge localised mode (ELM) control are developed. On the other hand, divertor solutions including advanced magnetic configurations are studied. Considerable progress has been made on both approaches, in particular in the fields of: ELM control with resonant magnetic perturbations (RMP), small ELM regimes, detachment onset and control, as well as filamentary scrape-off-layer transport. For example full ELM suppression has now been achieved on AUG at low collisionality with n  =  2 RMP maintaining good confinement {{H}\\text{H≤ft(98,\\text{y}2\\right)}}≈ 0.95 . Advances have been made with respect to detachment onset and control. Studies in advanced divertor configurations (Snowflake, Super-X and X-point target divertor) shed new light on SOL physics. Cross field filamentary transport has been characterised in a wide parameter regime on AUG, MAST and TCV progressing the theoretical and experimental understanding crucial for predicting first wall loads in ITER and DEMO. Conditions in the SOL also play a crucial role for ELM stability and access to small ELM regimes. In the future we will refer to the author list of the paper as the EUROfusion MST1 Team.

Developing the science and technology for the Material Plasma Exposure eXperiment

NASA Astrophysics Data System (ADS)

Rapp, J.; Biewer, T. M.; Bigelow, T. S.; Caneses, J. F.; Caughman, J. B. O.; Diem, S. J.; Goulding, R. H.; Isler, R. C.; Lumsdaine, A.; Beers, C. J.; Bjorholm, T.; Bradley, C.; Canik, J. M.; Donovan, D.; Duckworth, R. C.; Ellis, R. J.; Graves, V.; Giuliano, D.; Green, D. L.; Hillis, D. L.; Howard, R. H.; Kafle, N.; Katoh, Y.; Lasa, A.; Lessard, T.; Martin, E. H.; Meitner, S. J.; Luo, G.-N.; McGinnis, W. D.; Owen, L. W.; Ray, H. B.; Shaw, G. C.; Showers, M.; Varma, V.; the MPEX Team

2017-11-01

Linear plasma generators are cost effective facilities to simulate divertor plasma conditions of present and future fusion reactors. They are used to address important R&D gaps in the science of plasma material interactions and towards viable plasma facing components for fusion reactors. Next generation plasma generators have to be able to access the plasma conditions expected on the divertor targets in ITER and future devices. The steady-state linear plasma device MPEX will address this regime with electron temperatures of 1-10 eV and electron densities of 1021{\\text{}}-1020 m-3 . The resulting heat fluxes are about 10 MW m-2 . MPEX is designed to deliver those plasma conditions with a novel Radio Frequency plasma source able to produce high density plasmas and heat electron and ions separately with electron Bernstein wave (EBW) heating and ion cyclotron resonance heating with a total installed power of 800 kW. The linear device Proto-MPEX, forerunner of MPEX consisting of 12 water-cooled copper coils, has been operational since May 2014. Its helicon antenna (100 kW, 13.56 MHz) and EC heating systems (200 kW, 28 GHz) have been commissioned and 14 MW m-2 was delivered on target. Furthermore, electron temperatures of about 20 eV have been achieved in combined helicon and ECH heating schemes at low electron densities. Overdense heating with EBW was achieved at low heating powers. The operational space of the density production by the helicon antenna was pushed up to 1.1 × 1020 m-3 at high magnetic fields of 1.0 T at the target. The experimental results from Proto-MPEX will be used for code validation to enable predictions of the source and heating performance for MPEX. MPEX, in its last phase, will be capable to expose neutron-irradiated samples. In this concept, targets will be irradiated in ORNL’s High Flux Isotope Reactor and then subsequently exposed to fusion reactor relevant plasmas in MPEX.

Safety characteristics of the monolithic CFC divertor

NASA Astrophysics Data System (ADS)

Zucchetti, M.; Merola, M.; Matera, R.

1994-09-01

The main distinguishing feature of the monolithic CFC divertor is the use of a single material, a carbon fibre reinforced carbon, for the protective armour, the heat sink and the cooling channels. This removes joint interface problems which are one of the most important concerns related to the reference solutions of the ITER CDA divertor. An activation analysis of the different coolant options for this concept is presented. It turns out that neither short-term nor long-term activation are a concern for any coolants investigated. Therefore the proposed concept proves to be attractive from a safety stand-point also.

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

NASA Astrophysics Data System (ADS)

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

2015-09-01

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

Wide-angle ITER-prototype tangential infrared and visible viewing system for DIII-D.

PubMed

Lasnier, C J; Allen, S L; Ellis, R E; Fenstermacher, M E; McLean, A G; Meyer, W H; Morris, K; Seppala, L G; Crabtree, K; Van Zeeland, M A

2014-11-01

An imaging system with a wide-angle tangential view of the full poloidal cross-section of the tokamak in simultaneous infrared and visible light has been installed on DIII-D. The optical train includes three polished stainless steel mirrors in vacuum, which view the tokamak through an aperture in the first mirror, similar to the design concept proposed for ITER. A dichroic beam splitter outside the vacuum separates visible and infrared (IR) light. Spatial calibration is accomplished by warping a CAD-rendered image to align with landmarks in a data image. The IR camera provides scrape-off layer heat flux profile deposition features in diverted and inner-wall-limited plasmas, such as heat flux reduction in pumped radiative divertor shots. Demonstration of the system to date includes observation of fast-ion losses to the outer wall during neutral beam injection, and shows reduced peak wall heat loading with disruption mitigation by injection of a massive gas puff.

Comparison of the numerical modelling and experimental measurements of DIII-D separatrix displacements during H-modes with resonant magnetic perturbations

DOE PAGES

Orlov, Dmitry M.; Moyer, Richard A.; Evans, Todd E.; ...

2014-08-15

Numerical modeling of the plasma boundary position and its displacement due to external magnetic perturbations in DIII-D low-collisionality H-mode discharges is presented. The results of the vacuum model are compared to the experimental measurements for boundary displacements including Thomson scattering electron temperature T e, charge exchange recombination spectroscopy, beam emission spectroscopy, soft x-ray, and divertor Langmuir probe measurements. Magnetically perturbed discharges with toroidal mode number n=2 and n=3 are studied. It is shown that the vacuum model predictions agree well with the measurements above and below the midplane, and disagree at the outer midplane in discharges where significant kink amplificationmore » is present. Lastly, the role of the plasma response is studied using the two-fluid MHD code M3D-C 1, and the results are compared to the vacuum model showing that the plasma response model underestimates the boundary displacements.« less

Study of the impact of resonant magnetic perturbation fields on gross tungsten erosion using DiMES samples in DIII-D

NASA Astrophysics Data System (ADS)

Hinson, E. T.; Schmitz, O.; Frerichs, H.; Abrams, T.; Briesemeister, A.; Rudakov, D. L.; Unterberg, E. A.; Wampler, W. R.; Watkins, J. G.; Wang, H. Q.

2017-12-01

An experiment was conducted in DIII-D to compare gross tungsten (W) erosion on samples exposed to outer strike point (OSP) sweeps in L-mode plasmas for three conditions. These included two phases of resonant magnetic perturbations (RMPs), and a set with no perturbations. Upon RMP application, lobe structures indicative of strike point splitting of the OSP were evident in divertor camera data and on Langmuir probes. Gross W erosion flux, {{{Γ }}}{{W}}, inferred spectroscopically using the S/XB method applied to the 400.9 nm W-I line, was generally in the range {{{Γ }}}{{W}}/{{{Γ }}}{{D}+,\\perp }=2× {10}-4 referenced to incident deuterium ion flux {{{Γ }}}{{D}+,\\perp }, and was increased in the RMP cases by no more than 30% of the level observed in unperturbed discharges. A large reduction in gross erosion (50%) was observed in the private flux region at the W sample for one specific toroidal phase of the RMP field.

Comparison of tungsten nano-tendrils grown in Alcator C-Mod and linear plasma devices

NASA Astrophysics Data System (ADS)

Wright, G. M.; Brunner, D.; Baldwin, M. J.; Bystrov, K.; Doerner, R. P.; Labombard, B.; Lipschultz, B.; De Temmerman, G.; Terry, J. L.; Whyte, D. G.; Woller, K. B.

2013-07-01

Growth of tungsten nano-tendrils ("fuzz") has been observed for the first time in the divertor region of a high-power density tokamak experiment. After 14 consecutive helium L-mode discharges in Alcator C-Mod, the tip of a tungsten Langmuir probe at the outer strike point was fully covered with a layer of nano-tendrils. The depth of the W fuzz layer (600 ± 150 nm) is consistent with an empirical growth formula from the PISCES experiment. Re-creating the C-Mod exposures as closely as possible in Pilot-PSI experiment can produce nearly-identical nano-tendril morphology and layer thickness at surface temperatures that agree with uncertainties with the C-Mod W probe temperature data. Helium concentrations in W fuzz layers are measured at 1-4 at.%, which is lower than expected for the observed sub-surface voids to be filled with several GPa of helium pressure. This possibly indicates that the void formation is not pressure driven.

Wide-angle ITER-prototype tangential infrared and visible viewing system for DIII-D

DOE PAGES

Lasnier, Charles J.; Allen, Steve L.; Ellis, Ronald E.; ...

2014-08-26

An imaging system with a wide-angle tangential view of the full poloidal cross-section of the tokamak in simultaneous infrared and visible light has been installed on DIII-D. The optical train includes three polished stainless steel mirrors in vacuum, which view the tokamak through an aperture in the first mirror, similar to the design concept proposed for ITER. A dichroic beam splitter outside the vacuum separates visible and infrared (IR) light. Spatial calibration is accomplished by warping a CAD-rendered image to align with landmarks in a data image. The IR camera provides scrape-off layer heat flux profile deposition features in divertedmore » and inner-wall-limited plasmas, such as heat flux reduction in pumped radiative divertor shots. As a result, demonstration of the system to date includes observation of fast-ion losses to the outer wall during neutral beam injection, and shows reduced peak wall heat loading with disruption mitigation by injection of a massive gas puff.« less

Endobronchial Photoacoustic Microscopy for Staging of Lung Cancer

DTIC Science & Technology

2016-08-01

acoustic lens: (1) Three hairs were buried at different depths within the background phantom with 4mm distance between each hair . The advantage of this...and carried out tests to demonstrate this advantage using human hair as micro-scale targets (Figure 4). The targets were buried in a background with...the signal from the hair targets. The 30MHz transducer has outer diameter 11mm, and was equipped with corresponding lens whose apertures fit its outer

The outer epidermis of Avena and maize coleoptiles is not a unique target for auxin in elongation growth

NASA Technical Reports Server (NTRS)

Cleland, R. E.

1991-01-01

A controversy exists as to whether or not the outer epidermis in coleoptiles is a unique target for auxin in elongation growth. The following evidence indicates that the outer epidermis is not the only auxin-responsive cell layer in either Avena sativa L. or Zea mays L. coleoptiles. Coleoptile sections from which the epidermis has been removed by peeling elongate in response to auxin. The magnitude of the response is similar to that of intact sections provided the incubation solution contains both auxin and sucrose. The amount of elongation is independent of the amount of epidermis removed. Sections of oat coleoptiles from which the epidermis has been removed from one side are nearly straight after 22 h in auxin and sucrose, despite extensive growth of the sections. These data indicate that the outer epidermis is not a unique target for auxin in elongation growth, at least in Avena and maize coleoptiles.

Constrained ripple optimization of Tokamak bundle divertors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hively, L.M.; Rome, J.A.; Lynch, V.E.

1983-02-01

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

HSX as an example of a resilient non-resonant divertor

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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

HSX as an example of a resilient non-resonant divertor

DOE PAGES

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

2017-03-16

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

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

NASA Astrophysics Data System (ADS)

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

2017-10-01

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

All-reflective optical target illumination system with high numerical aperture

DOEpatents

Sigler, Robert D.

1978-01-01

An all-reflective optical system for providing illumination of a target focal region at high numerical aperture from a pair of co-axially, confluent collimated light beams. A target cavity is defined by a pair of opposed inner ellipsoidal reflectors having respective first focal points within a target region and second focal points at a vertex opening in the opposing reflector. Outwardly of each inner reflector is the opposed combination of a spherical reflector, and an outer generally ellipsoidal reflector having an aberrated first focal point coincident with the focus of the opposing spherical reflector and a second focal point coincident with the second focal point of the opposing inner ellipsoidal reflector through a vertex opening in the spherical reflector. The confluent collimated beams are incident through vertex openings in the outer ellipsoidal reflectors onto respective opposing spherical reflectors. Each beam is reflected by the associated spherical reflector onto the opposing outer ellipsoidal reflector and focused thereby onto the opposing inner ellipsoidal reflector, and then onto the target region.

Armour Materials for the ITER Plasma Facing Components

NASA Astrophysics Data System (ADS)

Barabash, V.; Federici, G.; Matera, R.; Raffray, A. R.; ITER Home Teams,

The selection of the armour materials for the Plasma Facing Components (PFCs) of the International Thermonuclear Experimental Reactor (ITER) is a trade-off between multiple requirements derived from the unique features of a burning fusion plasma environment. The factors that affect the selection come primarily from the requirements of plasma performance (e.g., minimise impurity contamination in the confined plasma), engineering integrity, component lifetime (e.g., withstand thermal stresses, acceptable erosion, etc.) and safety (minimise tritium and radioactive dust inventories). The current selection in ITER is to use beryllium on the first-wall, upper baffle and on the port limiter surfaces, carbon fibre composites near the strike points of the divertor vertical target and tungsten elsewhere in the divertor and lower baffle modules. This paper provides the background for this selection vis-à-vis the operating parameters expected during normal and off-normal conditions. The reasons for the selection of the specific grades of armour materials are also described. The effects of the neutron irradiation on the properties of Be, W and carbon fibre composites at the expected ITER conditions are briefly reviewed. Critical issues are discussed together with the necessary future R&D.

ITER in-vessel system design and performance

NASA Astrophysics Data System (ADS)

Parker, R. R.

2000-03-01

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

Prototyping phase of the high heat flux scraper element of Wendelstein 7-X

DOE PAGES

Boscary, Jean; Greuner, Henri; Ehrke, G.; ...

2016-03-24

The water-cooled high heat flux scraper element aims to reduce excessive heat loads on the target element ends of the actively cooled divertor of Wendelstein 7-X. Its purpose is to intercept some of the plasma fluxes both upstream and downstream before they reach the divertor surface. The scraper element has 24 identical plasma facing components (PFCs) divided into 6 modules. One module has 4 PFCs hydraulically connected in series by 2 water boxes. A PFC, 247 mm long and 28 mm wide, has 13 monoblocks made of CFC NB31 bonded by hot isostatic pressing onto a CuCrZr cooling tube equippedmore » with a copper twisted tape. 4 full-scale prototypes of PFCs have been successfully tested in the GLADIS facility up to 20 MW/m 2. The difference observed between measured and calculated surface temperatures is probably due to the inhomogeneity of CFC properties. The design of the water box prototypes has been detailed to allow the junction between the cooling pipe of the PFCs and the water boxes by internal orbital welding. In conclusion, the prototypes are presently under fabrication.« less

Design of an advanced bundle divertor for the Demonstration Tokamak Hybrid Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yang, T.F.; Lee, A.Y.; Ruck, G.W.

1979-01-25

The conclusion of this work is that a bundle divertor, using an improved method of designing the magnetic field configuration, is feasible for the Demonstration Tokamak Hybrid Reactor (DTHR) investigated by Westinghouse. The most significant achievement of this design is the reduction in current density (1 kA/cm/sup 2/) in the divertor coils in comparison to the overall averaged current densities per tesla of field to be nulled for DITE (25 kA/cm/sup 2/) and for ISX-B/sup 2/ (11 kA/cm/sup 2/). Therefore, superconducting magnets can be built into the tight space available with a sound mechanical structure.

Plasma confinement at JET

NASA Astrophysics Data System (ADS)

Nunes, I.; JET Contributors

2016-01-01

Operation with a Be/W wall at JET (JET-ILW) has an impact on scenario development and energy confinement with respect to the carbon wall (JET-C). The main differences observed were (1) strong accumulation of W in the plasma core and (2) the need to mitigate the divertor target temperature to avoid W sputtering by Be and other low Z impurities and (3) a decrease of plasma energy confinement. A major difference is observed on the pedestal pressure, namely a reduction of the pedestal temperature which, due to profile stiffness the plasma core temperature is also reduced leading to a degradation of the global confinement. This effect is more pronounced in low β N scenarios. At high β N, the impact of the wall on the plasma energy confinement is mitigated by the weaker plasma energy degradation with power relative to the IPB98(y, 2) scaling calculated empirically for a CFC first wall. The smaller tolerable impurity concentration for tungsten (<10-5) compared to that of carbon requires the use of electron heating methods to prevent W accumulation in the plasma core region as well as gas puffing to avoid W entering the plasma core by ELM flushing and reduction of the W source by decreasing the target temperature. W source and the target temperature can also be controlled by impurity seeding. Nitrogen and Neon have been used and with both gases the reduction of the W source and the target temperature is observed. Whilst more experiments with Neon are necessary to assess its impact on energy confinement, a partial increase of plasma energy confinement is observed with Nitrogen, through the increase of edge temperature. The challenge for scenario development at JET is to extend the pulse length curtailed by its transient behavior (W accumulation or MHD), but more importantly by the divertor target temperature limits. Re-optimisation of the scenarios to mitigate the effect of the change of wall materials maintaining high global energy confinement similar to JET-C is underway and JET has successfully achieved H 98(y,2)  =  1 for plasma currents up to 2.5 MA at moderate β N.

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

DOE PAGES

Perkins, R. J.; Hosea, J. C.; Jaworski, M. A.; ...

2015-04-13

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

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

NASA Astrophysics Data System (ADS)

Domalapally, Phani; Di Caro, Marco

2018-05-01

Among the options considered for cooling of the Plasma facing components of the DEMO reactor, water cooling is a conservative option because of its high heat removal capability. In this work a two-dimensional transient thermal hydraulic code is developed to support the design of the divertor for the projected DEMO reactor with water as a coolant. The mathematical model accounts for transient 2D heat conduction in the divertor section. Temperature-dependent properties are used for more accurate analysis. Correlations for single phase flow forced convection, partially developed subcooled nucleate boiling, fully developed subcooled nucleate boiling and film boiling are used to calculate the heat transfer coefficients on the channel side considering the swirl flow, wherein different correlations found in the literature are compared against each other. Correlation for the Critical Heat Flux is used to estimate its limit for a given flow conditions. This paper then investigates the results of the parametric analysis performed, whereby flow velocity, diameter of the coolant channel, thickness of the coolant pipe, thickness of the armor material, inlet temperature and operating pressure affect the behavior of the divertor under steady or transient heat fluxes. This code will help in understanding the basic parameterś effect on the behavior of the divertor, to achieve a better design from a thermal hydraulic point of view.

Low-Outgassing Photogrammetry Targets for Use in Outer Space

NASA Technical Reports Server (NTRS)

Gross, Jason N.; Sampler, Henry; Reed, Benjamin B.

2011-01-01

A short document discusses an investigation of materials for photogrammetry targets for highly sensitive optical scientific instruments to be operated in outer space and in an outer-space-environment- simulating thermal vacuum chamber on Earth. A key consideration in the selection of photogrammetry-target materials for vacuum environments is the need to prevent contamination that could degrade the optical responses of the instruments. Therefore, in addition to the high levels and uniformity of reflectivity required of photogrammetry-target materials suitable for use in air, the materials sought must exhibit minimal outgassing. Commercially available photogrammetry targets were found to outgas excessively under the thermal and vacuum conditions of interest; this finding prompted the investigators to consider optically equivalent or superior, lower-outgassing alternative target materials. The document lists several materials found to satisfy the requirements, but does not state explicitly whether the materials can be used individually or must be combined in the proper sequence into layered target structures. The materials in question are an aluminized polyimide tape, an acrylic pressure- sensitive adhesive, a 500-A-thick layer of vapor-deposited aluminum, and spherical barium titanate glass beads having various diameters from 20 to 63 microns..

On the targeting and membrane assembly of the Escherichia coli outer membrane porin, PhoE.

PubMed

Phoenix, D A

1996-12-01

Within gram-negative bacteria such as Escherichia coli, the outer membrane porins provide a relatively non-specific uptake route which is utilised by a wide range of solutes including many antibiotics. Understanding the targeting and membrane assembly of these proteins is therefore of importance and this mini review aims to discuss this process in light of present knowledge.

Final Progress Report The U.S. Department of Energy Research Grant No. DE-SC0008660 Plasma Surface Interactions: Bridging from the Surface to the Micron Frontier through Leadership Class Computing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Krasheninnikov, Sergei; Smirnov, Roman; Guterl, Jerome

The choice of material for the plasma facing components (PFC), in particular, for divertor targets, is one of the main issues for future tokamak reactors. There are two major requirements for the PFC’s material: acceptable level of tritium retention and durability in a harsh environment of fusion grade plasma. Based on these criteria, some years ago it was decided that tungsten is an acceptable material for divertor targets in ITER. However, further experimental studies reveal that the irradiation of tungsten even with low energetic (well below sputtering threshold!) He containing plasma causes significant modification of surface morphology, formation of themore » layer of He nano-bubbles (in the temperature range T<1000 K), “fuzz” (for 1000 K2000 K) (e.g. see Fig. 1). Recall that He, being an ash of D-T fusion reactions, is an inherent impurity in fusion plasma. The goals of the UCSD Applied Plasma Theory Group was: i) investigate the mechanisms of the formation of He nano-bubble layer and fuzz growth under He irradiation, as well as the physics of transport of hydrogen species in tungsten lattice, and ii) develop physics understanding of the models suitable for the incorporation into the Xolotl-PSI code based on the reaction-diffusion approach, which is the flagship of the whole SciDAC project [8], which can guide both numerical simulations and experimental studies. Here we just highlight our major accomplishments.« less

The snowflake divertor

NASA Astrophysics Data System (ADS)

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

2015-11-01

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.

Toroidally symmetric plasma vortex at tokamak divertor null point

DOE PAGES

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

2016-03-09

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

Divertor sheath power studies in DIII-D using fixed Langmuir probes and three-dimensional modeling of tile heat flows

NASA Astrophysics Data System (ADS)

Donovan, D.; Nygren, R.; Buchenauer, D.; Watkins, J.; Rudakov, D.; Leonard, A.; Wong, C. P. C.; Makowski, M.

2014-04-01

Experimental results are presented from the three-Langmuir probe (LP) diagnostic head of the divertor material evaluation system (DiMES) on DIII-D that confirm the size of the projected current collection area of the LPs, which is essential for properly measuring ion saturation current density (Jsat) and the sheath power transmission factor (SPTF). Also using the 3-LP DiMES head, the hypothesis that collisional effects on plasma density occurring in the magnetic sheath of the tile are responsible for a lower than expected SPTF is tested and deemed not to have a significant impact on the SPTF. Three-dimensional thermal modeling of wall tiles is presented that accounts for lateral heat conduction, temperature dependence of tile material properties and radiative heat loss from the tile surface. This modeling was developed to be used in the analysis of temperature profiles of the divertor embedded thermocouple (TC) array to obtain more accurate interpretations of TC temperature profiles to infer divertor surface heat flux than have previously been accomplished using more basic one-dimensional methods.

Fast plasma shutdown by killer pellet injection in JT-60U with reduced heat flux on the divertor plate and avoiding runaway electron generation

NASA Astrophysics Data System (ADS)

Yoshino, R.; Kondoh, T.; Neyatani, Y.; Itami, K.; Kawano, Y.; Isei, N.

1997-02-01

A killer pellet is an impurity pellet that is injected into a tokamak plasma in order to terminate a discharge without causing serious damage to the tokamak machine. In JT-60U neon ice pellets have been injected into OH and NB heated plasmas and fast plasma shutdowns have been demonstrated without large vertical displacement. The heat pulse on the divertor plate has been greatly reduced by killer pellet injection (KPI), but a low-power heat flux tail with a long time duration is observed. The total energy on the divertor plate increases with longer heat flux tail, so it has been reduced by shortening the tail. Runaway electron (RE) generation has been observed just after KPI and/or in the later phase of the plasma current quench. However, RE generation has been avoided when large magnetic perturbations are excited. These experimental results clearly show that KPI is a credible fast shutdown method avoiding large vertical displacement, reducing heat flux on the divertor plate, and avoiding (or minimizing) RE generation.

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.

FLIT: Flowing LIquid metal Torus

NASA Astrophysics Data System (ADS)

Kolemen, Egemen; Majeski, Richard; Maingi, Rajesh; Hvasta, Michael

2017-10-01

The design and construction of FLIT, Flowing LIquid Torus, at PPPL is presented. FLIT focuses on a liquid metal divertor system suitable for implementation and testing in present-day fusion systems, such as NSTX-U. It is designed as a proof-of-concept fast-flowing liquid metal divertor that can handle heat flux of 10 MW/m2 without an additional cooling system. The 72 cm wide by 107 cm tall torus system consisting of 12 rectangular coils that give 1 Tesla magnetic field in the center and it can operate for greater than 10 seconds at this field. Initially, 30 gallons Galinstan (Ga-In-Sn) will be recirculated using 6 jxB pumps and flow velocities of up to 10 m/s will be achieved on the fully annular divertor plate. FLIT is designed as a flexible machine that will allow experimental testing of various liquid metal injection techniques, study of flow instabilities, and their control in order to prove the feasibility of liquid metal divertor concept for fusion reactors. FLIT: Flowing LIquid metal Torus. This work is supported by the US DOE Contract No. DE-AC02-09CH11466.

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

NASA Astrophysics Data System (ADS)

Deng, Guozhong; Liu, Xiaoju; Wang, Liang; Liu, Shaocheng; Xu, Jichan; Feng, Wei; Liu, Jianbin; Liu, Huan; Gao, Xiang

2017-04-01

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

Overview of decade-long development of plasma-facing components at ASIPP

NASA Astrophysics Data System (ADS)

Luo, G.-N.; Liu, G. H.; Li, Q.; Qin, S. G.; Wang, W. J.; Shi, Y. L.; Xie, C. Y.; Chen, Z. M.; Missirlian, M.; Guilhem, D.; Richou, M.; Hirai, T.; Escourbiac, F.; Yao, D. M.; Chen, J. L.; Wang, T. J.; Bucalossi, J.; Merola, M.; Li, J. G.; EAST Team

2017-06-01

The first EAST (Experimental Advanced Superconducting Tokamak) plasma ignited in 2006 with non-actively cooled steel plates as the plasma-facing materials and components (PFMCs) which were then upgraded into full graphite tiles bolted onto water-cooled copper heat sinks in 2008. The first wall was changed further into molybdenum alloy in 2012, while keeping the graphite for both the upper and lower divertors. With the rapid increase in heating and current driving power in EAST, the W/Cu divertor project was launched around the end of 2012, aiming at achieving actively cooled full W/Cu-PFCs for the upper divertor, with heat removal capability up to 10 MW m-2. The W/Cu upper divertor was finished in the spring of 2014, consisting of 80 cassette bodies toroidally assembled. Commissioning of the EAST upper W/Cu divertor in 2014 was unsatisfactory and then several practical measures were implemented to improve the design, welding quality and reliability, which helped us achieve successful commissioning in the 2015 Spring Campaign. In collaboration with the IO and CEA teams, we have demonstrated our technological capability to remove heat loads of 5000 cycles at 10 MW m-2 and 1000 cycles at 20 MW m-2 for the small scale monoblock mockups, and surprisingly over 300 cycles at 20 MW m-2 for the flat-tile ones. The experience and lessons we learned from batch production and commissioning are undoubtedly valuable for ITER (International Thermonuclear Experimental Reactor) engineering validation and tungsten-related plasma physics.

Simulated nuclear reactor fuel assembly

DOEpatents

Berta, V.T.

1993-04-06

An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end.

Simulated nuclear reactor fuel assembly

DOEpatents

Berta, Victor T.

1993-01-01

An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end.

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 N 2, Ne and Ar have been performed in DIII-D to assess the impact of the different impurities on divertor detachment and confinement. Seeding with N 2, 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 powermore » to the 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 H 98 reducing from 1.1 to below 0.7. As for Ne seeding, H 98 near 0.8 can be maintained during the detachment phase with the pedestal temperature being reduced by about 50%. In contrast, in the N 2 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 N 2 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

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

NASA Astrophysics Data System (ADS)

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

2017-07-01

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

Specific targeting of proteins to outer envelope membranes of endosymbiotic organelles, chloroplasts, and mitochondria

PubMed Central

Lee, Junho; Kim, Dae Heon; Hwang, Inhwan

2014-01-01

Chloroplasts and mitochondria are endosymbiotic organelles thought to be derived from endosymbiotic bacteria. In present-day eukaryotic cells, these two organelles play pivotal roles in photosynthesis and ATP production. In addition to these major activities, numerous reactions, and cellular processes that are crucial for normal cellular functions occur in chloroplasts and mitochondria. To function properly, these organelles constantly communicate with the surrounding cellular compartments. This communication includes the import of proteins, the exchange of metabolites and ions, and interactions with other organelles, all of which heavily depend on membrane proteins localized to the outer envelope membranes. Therefore, correct and efficient targeting of these membrane proteins, which are encoded by the nuclear genome and translated in the cytosol, is critically important for organellar function. In this review, we summarize the current knowledge of the mechanisms of protein targeting to the outer membranes of mitochondria and chloroplasts in two different directions, as well as targeting signals and cytosolic factors. PMID:24808904

Liquid lithium loop system to solve challenging technology issues for fusion power plant

DOE PAGES

Ono, Masayuki; Majeski, Richard P.; Jaworski, Michael A.; ...

2017-07-12

Here, steady-state fusion power plant designs present major divertor technology challenges, including high divertor heat flux both in steady-state and during transients. In addition to these concerns, there are the unresolved technology issues of long term dust accumulation and associated tritium inventory and safety issues. It has been suggested that radiation-based liquid lithium (LL) divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues, while potentially improving reactor plasma performance. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peakmore » heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid lithium divertor (RLLD) concept and its variant, the active liquid lithium divertor concept (ARLLD), taking advantage of the enhanced or non-coronal Li radiation in relatively poorly confined divertor plasmas. To maintain the LL purity in a 1 GW-electric class fusion power plant, a closed LL loop system with a modest circulating capacity of ~ 1 liter/second (l/sec) is envisioned. We examined two key technology issues: 1) dust or solid particle removal and 2) real time recovery of tritium from LL while keeping the tritium inventory level to an acceptable level. By running the LL-loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to the outside where the dust / impurities can be removed by relatively simple dust filter, cold trap and/or centrifugal separation systems. With ~ 1 l/sec LL flow, even a small 0.1% dust content by weight (or 0.5 g per sec) suggests that the LL-loop could carry away nearly 16 tons of dust per year. In a 1 GW-electric (or ~ 3 GW fusion power) fusion power plant, about 0.5 g / sec of tritium is needed to maintain the fusion fuel cycle assuming ~ 1 % fusion burn efficiency. It appears feasible to recover tritium (T) in real time from LL while maintaining an acceptable T inventory level. Laboratory tests are being conducted to investigate T recovery feasibility with the surface cold trap (SCT) concept.« less

Liquid lithium loop system to solve challenging technology issues for fusion power plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ono, Masayuki; Majeski, Richard P.; Jaworski, Michael A.

Here, steady-state fusion power plant designs present major divertor technology challenges, including high divertor heat flux both in steady-state and during transients. In addition to these concerns, there are the unresolved technology issues of long term dust accumulation and associated tritium inventory and safety issues. It has been suggested that radiation-based liquid lithium (LL) divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues, while potentially improving reactor plasma performance. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peakmore » heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid lithium divertor (RLLD) concept and its variant, the active liquid lithium divertor concept (ARLLD), taking advantage of the enhanced or non-coronal Li radiation in relatively poorly confined divertor plasmas. To maintain the LL purity in a 1 GW-electric class fusion power plant, a closed LL loop system with a modest circulating capacity of ~ 1 liter/second (l/sec) is envisioned. We examined two key technology issues: 1) dust or solid particle removal and 2) real time recovery of tritium from LL while keeping the tritium inventory level to an acceptable level. By running the LL-loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to the outside where the dust / impurities can be removed by relatively simple dust filter, cold trap and/or centrifugal separation systems. With ~ 1 l/sec LL flow, even a small 0.1% dust content by weight (or 0.5 g per sec) suggests that the LL-loop could carry away nearly 16 tons of dust per year. In a 1 GW-electric (or ~ 3 GW fusion power) fusion power plant, about 0.5 g / sec of tritium is needed to maintain the fusion fuel cycle assuming ~ 1 % fusion burn efficiency. It appears feasible to recover tritium (T) in real time from LL while maintaining an acceptable T inventory level. Laboratory tests are being conducted to investigate T recovery feasibility with the surface cold trap (SCT) concept.« less

Liquid lithium loop system to solve challenging technology issues for fusion power plant

NASA Astrophysics Data System (ADS)

Ono, M.; Majeski, R.; Jaworski, M. A.; Hirooka, Y.; Kaita, R.; Gray, T. K.; Maingi, R.; Skinner, C. H.; Christenson, M.; Ruzic, D. N.

2017-11-01

Steady-state fusion power plant designs present major divertor technology challenges, including high divertor heat flux both in steady-state and during transients. In addition to these concerns, there are the unresolved technology issues of long term dust accumulation and associated tritium inventory and safety issues. It has been suggested that radiation-based liquid lithium (LL) divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues, while potentially improving reactor plasma performance. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid lithium divertor concept and its variant, the active liquid lithium divertor concept, taking advantage of the enhanced or non-coronal Li radiation in relatively poorly confined divertor plasmas. To maintain the LL purity in a 1 GW-electric class fusion power plant, a closed LL loop system with a modest circulating capacity of ~1 l s-1 is envisioned. We examined two key technology issues: (1) dust or solid particle removal and (2) real time recovery of tritium from LL while keeping the tritium inventory level to an acceptable level. By running the LL-loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to the outside where the dust/impurities can be removed by relatively simple dust filter, cold trap and/or centrifugal separation systems. With ~1 l s-1 LL flow, even a small 0.1% dust content by weight (or 0.5 g s-1) suggests that the LL-loop could carry away nearly 16 tons of dust per year. In a 1 GW-electric (or ~3 GW fusion power) fusion power plant, about 0.5 g s-1 of tritium is needed to maintain the fusion fuel cycle assuming ~1% fusion burn efficiency. It appears feasible to recover tritium (T) in real time from LL while maintaining an acceptable T inventory level. Laboratory tests are being conducted to investigate T recovery feasibility with the surface cold trap concept.

Lithium As Plasma Facing Component for Magnetic Fusion Research

DOE Office of Scientific and Technical Information (OSTI.GOV)

Masayuki Ono

The use of lithium in magnetic fusion confinement experiments started in the 1990's in order to improve tokamak plasma performance as a low-recycling plasma-facing component (PFC). Lithium is the lightest alkali metal and it is highly chemically reactive with relevant ion species in fusion plasmas including hydrogen, deuterium, tritium, carbon, and oxygen. Because of the reactive properties, lithium can provide strong pumping for those ions. It was indeed a spectacular success in TFTR where a very small amount (~ 0.02 gram) of lithium coating of the PFCs resulted in the fusion power output to improve by nearly a factor ofmore » two. The plasma confinement also improved by a factor of two. This success was attributed to the reduced recycling of cold gas surrounding the fusion plasma due to highly reactive lithium on the wall. The plasma confinement and performance improvements have since been confirmed in a large number of fusion devices with various magnetic configurations including CDX-U/LTX (US), CPD (Japan), HT-7 (China), EAST (China), FTU (Italy), NSTX (US), T-10, T-11M (Russia), TJ-II (Spain), and RFX (Italy). Additionally, lithium was shown to broaden the plasma pressure profile in NSTX, which is advantageous in achieving high performance H-mode operation for tokamak reactors. It is also noted that even with significant applications (up to 1,000 grams in NSTX) of lithium on PFCs, very little contamination (< 0.1%) of lithium fraction in main fusion plasma core was observed even during high confinement modes. The lithium therefore appears to be a highly desirable material to be used as a plasma PFC material from the magnetic fusion plasma performance and operational point of view. An exciting development in recent years is the growing realization of lithium as a potential solution to solve the exceptionally challenging need to handle the fusion reactor divertor heat flux, which could reach 60 MW/m2 . By placing the liquid lithium (LL) surface in the path of the main divertor heat flux (divertor strike point), the lithium is evaporated from the surface. The evaporated lithium is quickly ionized by the plasma and the ionized lithium ions can provide a strongly radiative layer of plasma ("radiative mantle"), thus could significantly reduce the heat flux to the divertor strike point surfaces, thus protecting the divertor surface. The protective effects of LL have been observed in many experiments and test stands. As a possible reactor divertor candidate, a closed LL divertor system is described. Finally, it is noted that the lithium applications as a PFC can be quite flexible and broad. The lithium application should be quite compatible with various divertor configurations, and it can be also applied to protecting the presently envisioned tungsten based solid PFC surfaces such as the ones for ITER. Lithium based PFCs therefore have the exciting prospect of providing a cost effective flexible means to improve the fusion reactor performance, while providing a practical solution to the highly challenging divertor heat handling issue confronting the steadystate magnetic fusion reactors.« less

Fuel assembly for the production of tritium in light water reactors

DOEpatents

Cawley, W.E.; Trapp, T.J.

1983-06-10

A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

Fuel assembly for the production of tritium in light water reactors

DOEpatents

Cawley, William E.; Trapp, Turner J.

1985-01-01

A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

Quantitative comparisons between experimentally measured 2D carbon radiation and Monte Carlo impurity (MCI) code simulations

NASA Astrophysics Data System (ADS)

Evans, T. E.; Finkenthal, D. F.; Fenstermacher, M. E.; Leonard, A. W.; Porter, G. D.; West, W. P.

Experimentally measured carbon line emissions and total radiated power distributions from the DIII-D divertor and scrape-off layer (SOL) are compared to those calculated with the Monte Carlo impurity (MCI) model. A UEDGE [T.D. Rognlien et al., J. Nucl. Mater. 196-198 (1992) 347] background plasma is used in MCI with the Roth and Garcia-Rosales (RG-R) chemical sputtering model [J. Roth, C. García-Rosales, Nucl. Fusion 36 (1992) 196] and/or one of six physical sputtering models. While results from these simulations do not reproduce all of the features seen in the experimentally measured radiation patterns, the total radiated power calculated in MCI is in relatively good agreement with that measured by the DIII-D bolometric system when the Smith78 [D.L. Smith, J. Nucl. Mater. 75 (1978) 20] physical sputtering model is coupled to RG-R chemical sputtering in an unaltered UEDGE plasma. Alternatively, MCI simulations done with UEDGE background ion temperatures along the divertor target plates adjusted to better match those measured in the experiment resulted in three physical sputtering models which when coupled to the RG-R model gave a total radiated power that was within 10% of measured value.

Two conceptual designs of helical fusion reactor FFHR-d1A based on ITER technologies and challenging ideas

NASA Astrophysics Data System (ADS)

Sagara, A.; Miyazawa, J.; Tamura, H.; Tanaka, T.; Goto, T.; Yanagi, N.; Sakamoto, R.; Masuzaki, S.; Ohtani, H.; The FFHR Design Group

2017-08-01

The Fusion Engineering Research Project (FERP) at the National Institute for Fusion Science (NIFS) is conducting conceptual design activities for the LHD-type helical fusion reactor FFHR-d1A. This paper newly defines two design options, ‘basic’ and ‘challenging.’ Conservative technologies, including those that will be demonstrated in ITER, are chosen in the basic option in which two helical coils are made of continuously wound cable-in-conduit superconductors of Nb3Sn strands, the divertor is composed of water-cooled tungsten monoblocks, and the blanket is composed of water-cooled ceramic breeders. In contrast, new ideas that would possibly be beneficial for making the reactor design more attractive are boldly included in the challenging option in which the helical coils are wound by connecting high-temperature REBCO superconductors using mechanical joints, the divertor is composed of a shower of molten tin jets, and the blanket is composed of molten salt FLiNaBe including Ti powers to increase hydrogen solubility. The main targets of the challenging option are early construction and easy maintenance of a large and three-dimensionally complicated helical structure, high thermal efficiency, and, in particular, realistic feasibility of the helical reactor.

Control of three dimensional particle flux to divertor using rotating RMP in the EAST tokamak

NASA Astrophysics Data System (ADS)

Jia, M.; Sun, Y.; Liang, Y.; Wang, L.; Xu, J.; Gu, S.; Lyu, B.; Wang, H. H.; Yang, X.; Zhong, F.; Chu, N.; Feng, W.; He, K.; Liu, Y. Q.; Qian, J.; Shi, T.; Shen, B.

2018-04-01

Controlling the steady state particle and heat flux impinging on the plasma facing components, as one of the main concerns of future fusion reactors, is still necessary when the transient power loads induced by edge localized modes (ELMs) have been eliminated by resonant magnetic perturbations (RMPs) in high confinement tokamak experiments. This is especially true for long pulse operation. One promising solution is to use the rotating perturbed field. Recently rotating and differential phase scans of n  =  1 and 2 RMP fields have been operated for the first time in EAST discharges. The particle flux patterns on the divertor targets change synchronously with both rotating and phasing RMP fields as predicted by the modeled magnetic footprint patterns. The modeling with plasma response, which is calculated by MARS-F, is also carried out. The plasma response shows amplifying or screening effect to n  =  2 perturbations with different spectra. This changes the field line penetration depth rather than the general footprint shape. This has been verified by experimental observations on EAST. These experiments motivate further study of reducing both transient and steady state local power load and particle flux with the help of rotating RMPs in long pulse operation.

Dust Studies in DIII-D and TEXTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rudakov, D L; Litnovsky, A; West, W P

2009-02-17

Studies of naturally occurring and artificially introduced carbon dust are conducted in DIII-D and TEXTOR. In DIII-D, dust does not present operational concerns except immediately after entry vents. Submicron sized dust is routinely observed using Mie scattering from a Nd:Yag laser. The source is strongly correlated with the presence of Type I edge localized modes (ELMs). Larger size (0.005-1 mm diameter) dust is observed by optical imaging, showing elevated dust levels after entry vents. Inverse dependence of the dust velocity on the inferred dust size is found from the imaging data. Direct heating of the dust particles by the neutralmore » beam injection (NBI) and acceleration of dust particles by the plasma flows are observed. Energetic plasma disruptions produce significant amounts of dust. Large flakes or debris falling into the plasma may result in a disruption. Migration of pre-characterized carbon dust is studied in DIII-D and TEXTOR by introducing micron-size dust in plasma discharges. In DIII-D, a sample holder filled with {approx}30 mg of dust is introduced in the lower divertor and exposed to high-power ELMing H-mode discharges with strike points swept across the divertor floor. After a brief exposure ({approx}0.1 s) at the outer strike point, part of the dust is injected into the plasma, raising the core carbon density by a factor of 2-3 and resulting in a twofold increase of the radiated power. In TEXTOR, instrumented dust holders with 1-45 mg of dust are exposed in the scrape-off layer 0-2 cm radially outside of the last closed flux surface in discharges heated with neutral beam injection (NBI) power of 1.4 MW. At the given configuration of the launch, the dust did not penetrate the core plasma and only moderately perturbed the edge plasma, as evidenced by an increase of the edge carbon content.« less

Developing the science and technology for the Material Plasma Exposure eXperiment

DOE PAGES

Rapp, J.; Biewer, T. M.; Bigelow, T. S.; ...

2017-07-27

Linear plasma generators are cost effective facilities to simulate divertor plasma conditions of present and future fusion reactors. They are used to address important R&D gaps in the science of plasma material interactions and towards viable plasma facing components for fusion reactors. Next generation plasma generators have to be able to access the plasma conditions expected on the divertor targets in ITER and future devices. The steady-state linear plasma device MPEX will address this regime with electron temperatures of 1–10 eV and electron densities ofmore » $$10^{21}{\\text{}}\\!-\\!10^{20}$$ $${\\rm m}^{-3}$$. The resulting heat fluxes are about 10 MW $${\\rm m}^{-2}$$ . MPEX is designed to deliver those plasma conditions with a novel Radio Frequency plasma source able to produce high density plasmas and heat electron and ions separately with electron Bernstein wave (EBW) heating and ion cyclotron resonance heating with a total installed power of 800 kW. The linear device Proto-MPEX, forerunner of MPEX consisting of 12 water-cooled copper coils, has been operational since May 2014. Its helicon antenna (100 kW, 13.56 MHz) and EC heating systems (200 kW, 28 GHz) have been commissioned and 14 MW $${\\rm m}^{-2}$$ was delivered on target. Furthermore, electron temperatures of about 20 eV have been achieved in combined helicon and ECH heating schemes at low electron densities. Overdense heating with EBW was achieved at low heating powers. The operational space of the density production by the helicon antenna was pushed up to $$1.1 \\times 10^{20}$$ $${\\rm m}^{-3}$$ at high magnetic fields of 1.0 T at the target. Finally, the experimental results from Proto-MPEX will be used for code validation to enable predictions of the source and heating performance for MPEX. MPEX, in its last phase, will be capable to expose neutron-irradiated samples. In this concept, targets will be irradiated in ORNL's High Flux Isotope Reactor and then subsequently exposed to fusion reactor relevant plasmas in MPEX.« less

Developing the science and technology for the Material Plasma Exposure eXperiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rapp, J.; Biewer, T. M.; Bigelow, T. S.

Linear plasma generators are cost effective facilities to simulate divertor plasma conditions of present and future fusion reactors. They are used to address important R&D gaps in the science of plasma material interactions and towards viable plasma facing components for fusion reactors. Next generation plasma generators have to be able to access the plasma conditions expected on the divertor targets in ITER and future devices. The steady-state linear plasma device MPEX will address this regime with electron temperatures of 1–10 eV and electron densities ofmore » $$10^{21}{\\text{}}\\!-\\!10^{20}$$ $${\\rm m}^{-3}$$. The resulting heat fluxes are about 10 MW $${\\rm m}^{-2}$$ . MPEX is designed to deliver those plasma conditions with a novel Radio Frequency plasma source able to produce high density plasmas and heat electron and ions separately with electron Bernstein wave (EBW) heating and ion cyclotron resonance heating with a total installed power of 800 kW. The linear device Proto-MPEX, forerunner of MPEX consisting of 12 water-cooled copper coils, has been operational since May 2014. Its helicon antenna (100 kW, 13.56 MHz) and EC heating systems (200 kW, 28 GHz) have been commissioned and 14 MW $${\\rm m}^{-2}$$ was delivered on target. Furthermore, electron temperatures of about 20 eV have been achieved in combined helicon and ECH heating schemes at low electron densities. Overdense heating with EBW was achieved at low heating powers. The operational space of the density production by the helicon antenna was pushed up to $$1.1 \\times 10^{20}$$ $${\\rm m}^{-3}$$ at high magnetic fields of 1.0 T at the target. Finally, the experimental results from Proto-MPEX will be used for code validation to enable predictions of the source and heating performance for MPEX. MPEX, in its last phase, will be capable to expose neutron-irradiated samples. In this concept, targets will be irradiated in ORNL's High Flux Isotope Reactor and then subsequently exposed to fusion reactor relevant plasmas in MPEX.« less

All-reflective optical target illumination system with high numerical aperture

DOEpatents

Thomas, Carlton E.; Sigler, Robert D.; Hoeger, John G.

1979-01-01

An all-reflective optical system for providing illumination of a target focal region at high numerical aperture from a pair of confluent collimated light beams. The collimated beams are each incident upon an associated concave eccentric pupil paraboloidal reflective surface, and thereby each focused through an opening in an associated outer ellipsoidal reflective surface onto a plane reflector. Each beam is reflected by its associated plane reflector onto the opposing concave surface of the outer ellipsoids to be focused through an opening in the plane surface onto an opposing inner concave ellipsoidal reflective surface, and thence onto the target region.

High-resolution disruption halo current measurements using Langmuir probes in Alcator C-Mod

NASA Astrophysics Data System (ADS)

Tinguely, R. A.; Granetz, R. S.; Berg, A.; Kuang, A. Q.; Brunner, D.; LaBombard, B.

2018-01-01

Halo currents generated during disruptions on Alcator C-Mod have been measured with Langmuir ‘rail’ probes. These rail probes are embedded in a lower outboard divertor module in a closely-spaced vertical (poloidal) array. The dense array provides detailed resolution of the spatial dependence (~1 cm spacing) of the halo current distribution in the plasma scrape-off region with high time resolution (400 kHz digitization rate). As the plasma limits on the outboard divertor plate, the contact point is clearly discernible in the halo current data (as an inversion of current) and moves vertically down the divertor plate on many disruptions. These data are consistent with filament reconstructions of the plasma boundary, from which the edge safety factor of the disrupting plasma can be calculated. Additionally, the halo current ‘footprint’ on the divertor plate is obtained and related to the halo flux width. The voltage driving halo current and the effective resistance of the plasma region through which the halo current flows to reach the probes are also investigated. Estimations of the sheath resistance and halo region resistivity and temperature are given. This information could prove useful for modeling halo current dynamics.

Low temperature tungsten spectroscopy on a Penning Ionization Discharge

NASA Astrophysics Data System (ADS)

Kumar, Deepak; Englesbe, Alexander; Stutman, Dan; Finkenthal, Michael

2011-10-01

Complete Tungsten divertor operation is being planned on many tokamaks including Tore Supra and ITER. Thus, low temperature tungsten spectroscopy is important for aiding the divertor diagnostics on larger machines. A Penning Ionization Discharge (PID) at the Johns Hopkins University produces steady state plasmas with Te ~ 2 eV, ne ~1013 cm-3 and a fast electron fraction at ~ 10 s eV. Similar bi-Maxwellian distributions, but with slightly higher electron temperatures, are found in the divertor plasmas of tokamaks. The two significant populating mechanisms for higher charge states in the PID are: (a) collisional excitation from bulk electrons, and (b) inner shell ionization from the fast electrons. The PID is diagnosed in a wide wavelength range - XUV, VUV and visible, to differentiate the two populating mechanisms. W is introduced in the PID by the sputtering of cathodes made of CuW alloy. Spectral emission from significantly higher charge states of W (up to W IV) has been observed in the experiment. This poster will describe results indicating the populating mechanism of W ions and also describe plans on upgrading the experiment to achieve higher temperatures which are closer to the divertor conditions. Supported by USDOE.

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

NASA Astrophysics Data System (ADS)

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

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

Overview of the JET results

NASA Astrophysics Data System (ADS)

Romanelli, F.; JET Contributors,

2015-10-01

Since the installation of an ITER-like wall, the JET programme has focused on the consolidation of ITER design choices and the preparation for ITER operation, with a specific emphasis given to the bulk tungsten melt experiment, which has been crucial for the final decision on the material choice for the day-one tungsten divertor in ITER. Integrated scenarios have been progressed with the re-establishment of long-pulse, high-confinement H-modes by optimizing the magnetic configuration and the use of ICRH to avoid tungsten impurity accumulation. Stationary discharges with detached divertor conditions and small edge localized modes have been demonstrated by nitrogen seeding. The differences in confinement and pedestal behaviour before and after the ITER-like wall installation have been better characterized towards the development of high fusion yield scenarios in DT. Post-mortem analyses of the plasma-facing components have confirmed the previously reported low fuel retention obtained by gas balance and shown that the pattern of deposition within the divertor has changed significantly with respect to the JET carbon wall campaigns due to the absence of thermally activated chemical erosion of beryllium in contrast to carbon. Transport to remote areas is almost absent and two orders of magnitude less material is found in the divertor.

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

NASA Astrophysics Data System (ADS)

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

2017-07-01

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

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

NASA Astrophysics Data System (ADS)

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

2012-05-01

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

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

DOE PAGES

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

2017-06-07

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

Effects of low and high mode number tearing modes in divertor tokamaks

NASA Astrophysics Data System (ADS)

Punjabi, Alkesh; Ali, Halima; Boozer, Allen; Evans, Todd

2007-08-01

The topological effects of magnetic perturbations on a divertor tokamak, such as DIII-D, are studied using field-line maps that were developed by Punjabi et al. [A. Punjabi, A. Verma, and A. Boozer, Phys. Rev. Lett. 69, 3322 (1992)]. The studies consider both long-wavelength perturbations, such as those of m =1, n =1 tearing modes, and localized perturbations, which are represented as a magnetic dipole. The parameters of the dipole map are set using DIII-D data from shot 115467 in which the C-coils were activated [J. L. Luxon and L. E. Davis, Fusion Technol. 8, 441 (1985)]. The long-wavelength perturbations alter the structure of the interception of magnetic field lines with the divertor plates, but the interception is in sharp lines. The dipole perturbations cause a spreading of the interception of the field lines with the divertor plates, which alleviates problems associated with heat deposition. Magnetic field lines are the trajectories of a one-and-a-half degree of freedom Hamiltonian, which strongly constrains the topological features of the lines. Although the field line maps that we use do not accurately represent the trajectories through ordinary space of individual field lines, they do represent their topological structure.

Two-point modeling of SOL losses of HHFW power in NSTX

NASA Astrophysics Data System (ADS)

Kish, Ayden; Perkins, Rory; Ahn, Joon-Wook; Diallo, Ahmed; Gray, Travis; Hosea, Joel; Jaworski, Michael; Kramer, Gerrit; Leblanc, Benoit; Sabbagh, Steve

2017-10-01

High-harmonic fast-wave (HHFW) heating is a heating and current-drive scheme on the National Spherical Torus eXperiment (NSTX) complimentary to neutral beam injection. Previous experiments suggest that a significant fraction, up to 50%, of the HHFW power is promptly lost to the scrape-off layer (SOL). Research indicates that the lost power reaches the divertor via wave propagation and is converted to a heat flux at the divertor through RF rectification rather than heating the SOL plasma at the midplane. This counter-intuitive hypothesis is investigated using a simplified two-point model, relating plasma parameters at the divertor to those at the midplane. Taking measurements at the divertor region of NSTX as input, this two-point model is used to predict midplane parameters, using the predicted heat flux as an indicator of power input to the SOL. These predictions are compared to measurements at the midplane to evaluate the extent to which they are consistent with experiment. This work was made possible by funding from the Department of Energy for the Summer Undergraduate Laboratory Internship (SULI) program. This work is supported by the US DOE Contract No. DE-AC02-09CH11466.

Predictive modelling of JT-60SA high-beta steady-state plasma with impurity accumulation

NASA Astrophysics Data System (ADS)

Hayashi, N.; Hoshino, K.; Honda, M.; Ide, S.

2018-06-01

The integrated modelling code TOPICS has been extended to include core impurity transport, and applied to predictive modelling of JT-60SA high-beta steady-state plasma with the accumulation of impurity seeded to reduce the divertor heat load. In the modelling, models and conditions are selected for a conservative prediction, which considers a lower bound of plasma performance with the maximum accumulation of impurity. The conservative prediction shows the compatibility of impurity seeding with core plasma with high-beta (β N  >  3.5) and full current drive conditions, i.e. when Ar seeding reduces the divertor heat load below 10 MW m‑2, its accumulation in the core is so moderate that the core plasma performance can be recovered by additional heating within the machine capability to compensate for Ar radiation. Due to the strong dependence of accumulation on the pedestal density gradient, high separatrix density is important for the low accumulation as well as the low divertor heat load. The conservative prediction also shows that JT-60SA has enough capability to explore the divertor heat load control by impurity seeding in high-beta steady-state plasmas.

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

DOE PAGES

Clementson, Joel; Lennartsson, Thomas; Beiersdorfer, Peter

2015-09-09

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

The near infrared imaging system for the real-time protection of the JET ITER-like wall

NASA Astrophysics Data System (ADS)

Huber, A.; Kinna, D.; Huber, V.; Arnoux, G.; Balboa, I.; Balorin, C.; Carman, P.; Carvalho, P.; Collins, S.; Conway, N.; McCullen, P.; Jachmich, S.; Jouve, M.; Linsmeier, Ch; Lomanowski, B.; Lomas, P. J.; Lowry, C. G.; Maggi, C. F.; Matthews, G. F.; May-Smith, T.; Meigs, A.; Mertens, Ph; Nunes, I.; Price, M.; Puglia, P.; Riccardo, V.; Rimini, F. G.; Sergienko, G.; Tsalas, M.; Zastrow, K.-D.; contributors, JET

2017-12-01

This paper describes the design, implementation and operation of the near infrared (NIR) imaging diagnostic system of the JET ITER-like wall (JET-ILW) plasma experiment and its integration into the existing JET protection architecture. The imaging system comprises four wide-angle views, four tangential divertor views, and two top views of the divertor covering 66% of the first wall and up to 43% of the divertor. The operation temperature ranges which must be observed by the NIR protection cameras are, for the materials used on JET: Be 700 °C-1400 °C W coating 700 °C-1370 °C W bulk 700 °C-1400 °C. The Real-Time Protection system operates routinely since 2011 and successfully demonstrated its capability to avoid the overheating of the main chamber beryllium wall as well as of the divertor W and W-coated carbon fibre composite (CFC) tiles. During this period, less than 0.5% of the terminated discharges were aborted by a malfunction of the system. About 2%-3% of the discharges were terminated due to the detection of actual hot spots.

Predictions of VRF on a Langmuir Probe under the RF Heating Spiral on the Divertor Floor on NSTX-U

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hosea, J C; Perkins, R J; Jaworski, M A

RF heating deposition spirals are observed on the divertor plates on NSTX as shown in for a NB plus RF heating case. It has been shown that the RF spiral is tracked quite well by the spiral mapping of the strike points on the divertor plate of magnetic field lines passing in front of the high harmonic fast wave (HHFW) antenna on NSTX. Indeed, both current instrumented tiles and Langmuir probes respond to the spiral when it is positioned over them. In particular, a positive increment in tile current (collection of electrons) is obtained when the spiral is over themore » tile. This current can be due to RF rectification and/or RF heating of the scrape off layer (SOL) plasma along the magnetic field lines passing in front of the the HHFW antenna. It is important to determine quantitatively the relative contributions of these processes. Here we explore the properties of the characteristics of probes on the lower divertor plate to determine the likelyhood that the primary cause of the RF heat deposition is RF rectification.« less

Gunshot residue patterns on skin in angled contact and near contact gunshot wounds.

PubMed

Plattner, T; Kneubuehl, B; Thali, M; Zollinger, U

2003-12-17

The goal of this study was the reproduction of shape and pattern of gunshot residues in near contact and contact gunshot wounds by a series of experimental gunshots on a skin and soft tissue model. The aim was to investigate the shape and direction of soot deposits with regard to the muzzle according to different muzzle-target angles, firing distances, type of ammunition and weapon and barrel length. Based on a review of the literature and on the results of the experiments the authors could make the following statements of gunshot residues in angled contact and close contact gunshot: (1) gunshot residues on the target surface can be differentiated in a "inner" and "outer powder soot zone"; (2) the outer powder soot zone is much less visible than the inner powder soot zone and may lack on human skin; (3) with increasing muzzle target distance both inner and outer powder soot halo increase in size and decrease in density; (4) in angled shots the inner powder soot halo shows an eccentric, elliptic shape which points towards the muzzle, regardless of ammunition, calibre and barrel length; (5) the outer powder soot points away from the muzzle in angled contact and close contact shots.

Injected mass deposition thresholds for lithium granule instigated triggering of edge localized modes on EAST

NASA Astrophysics Data System (ADS)

Lunsford, R.; Sun, Z.; Maingi, R.; Hu, J. S.; Mansfield, D.; Xu, W.; Zuo, G. Z.; Diallo, A.; Osborne, T.; Tritz, K.; Canik, J.; Huang, M.; Meng, X. C.; Gong, X. Z.; Wan, B. N.; Li, J. G.; the EAST Team

2018-03-01

The ability of an injected lithium granule to promptly trigger an edge localized mode (ELM) has been established in multiple experiments. By horizontally injecting granules ranging in diameter from 200 microns to 1 mm in diameter into the low field side of EAST H-mode discharges we have determined that granules with diameter  >600 microns are successful in triggering ELMs more than 95% of the time. It was also demonstrated that below 600 microns the triggering efficiency decreased roughly with granule size. Granules were radially injected from the outer midplane with velocities ~80 m s-1 into EAST upper single null discharges with an ITER like tungsten monoblock divertor. These granules were individually tracked throughout their injection cycle in order to determine their efficacy at triggering an ELM. For those granules of sufficient size, ELM triggering was a prompt response to granule injection. By simulating the granule injection with an experimentally benchmarked neutral gas shielding (NGS) model, the ablatant mass deposition required to promptly trigger an ELM is calculated and the fractional mass deposition is determined.

Power handling of a segmented bulk W tile for JET under realistic plasma scenarios

NASA Astrophysics Data System (ADS)

Jet-Efda Contributors Mertens, Ph.; Coenen, J. W.; Eich, T.; Huber, A.; Jachmich, S.; Nicolai, D.; Riccardo, V.; Senik, K.; Samm, U.

2011-08-01

A solid tungsten divertor row has been designed for JET in the frame of the ITER-like Wall project (ILW). The plasma-facing tiles are segmented in four stacks of tungsten lamellae oriented in the toroidal direction. Earlier estimations of the expected tile performance were carried out mostly for engineering purposes, to compare the permissible heat load with the power density of 7 MW/m2 originally specified for the ILW as a uniform load for 10 s.The global thermal model developed for the W modules delivers results for more realistic plasma footprints: the poloidal extension of the outer strike point was reduced from the full lamella width of 62 mm to ⩾15 mm. Model validation is given by the experimental exposure of a 1:1 prototype stack in the ion beam facility MARION (incidence ˜6°, load E ⩽ 66 MJ/m2 on the wetted surface). Spreading the deposited energy by appropriate sweeping over one or several stacks in the torus is beneficial for the tungsten lamellae and for the support structure.

Parameter dependences of the separatrix density in nitrogen seeded ASDEX Upgrade H-mode discharges

NASA Astrophysics Data System (ADS)

Kallenbach, A.; Sun, H. J.; Eich, T.; Carralero, D.; Hobirk, J.; Scarabosio, A.; Siccinio, M.; ASDEX Upgrade Team; EUROfusion MST1 Team

2018-04-01

The upstream separatrix electron density is an important interface parameter for core performance and divertor power exhaust. It has been measured in ASDEX Upgrade H-mode discharges by means of Thomson scattering using a self-consistent estimate of the upstream electron temperature under the assumption of Spitzer-Härm electron conduction. Its dependence on various plasma parameters has been tested for different plasma conditions in H-mode. The leading parameter determining n e,sep was found to be the neutral divertor pressure, which can be considered as an engineering parameter since it is determined mainly by the gas puff rate and the pumping speed. The experimentally found parameter dependence of n e,sep, which is dominated by the divertor neutral pressure, could be approximately reconciled by 2-point modelling.

Thermal strain measurement of EAST tungsten divertor component with bare fiber Bragg grating sensors

NASA Astrophysics Data System (ADS)

Wang, Xingli; Wang, Wanjing; Wang, Jichao; Wei, Ran; Sun, Zhaoxuan; Li, Qiang; Xie, Chunyi; Luo, Guang-Nan

2017-12-01

Fiber Bragg Gratings (FBGs) have been widely used in the sensor field to monitor temperature and strain. However, the weak mechanical property of optical fibers and insufficient heat-resistant property of general optic-fiber sensors have prevented it from being widely used, such as in some extreme engineering situations. In this work, a bare FBG sensor system had been introduced to measure thermal strain of an Experimental Advanced Superconducting Tokamak tungsten divertor component under baking condition. This strain measurement system had withstood as high temperature as 210 °C and finished the measurement experiment successfully. Meaningful measurement results had been obtained and analyzed, which showed the applicability of such a bare fiber grating sensor system and as well contributed to studying on tungsten divertor's thermal strain conditions.

Failure study of helium-cooled tungsten divertor plasma-facing units tested at DEMO relevant steady-state heat loads

NASA Astrophysics Data System (ADS)

Ritz, G.; Hirai, T.; Norajitra, P.; Reiser, J.; Giniyatulin, R.; Makhankov, A.; Mazul, I.; Pintsuk, G.; Linke, J.

2009-12-01

Tungsten was selected as armor material for the helium-cooled divertor in future DEMO-type fusion reactors and fusion power plants. After realizing the design and testing of them under cyclic thermal loads of up to ~14 MW m-2, the tungsten divertor plasma-facing units were examined by metallography; they revealed failures such as cracks at the thermal loaded and as-machined surfaces, as well as degradation of the brazing layers. Furthermore, in order to optimize the machining processes, the quality of tungsten surfaces prepared by turning, milling and using a diamond cutting wheel were examined. This paper presents a metallographic examination of the tungsten plasma-facing units as well as technical studies and the characterization on machining of tungsten and alternative brazing joints.

Inhibition of Gene Expression in Escherichia coli by Antisense Phosphorodiamidate Morpholino Oligomers

PubMed Central

Geller, B. L.; Deere, J. D.; Stein, D. A.; Kroeker, A. D.; Moulton, H. M.; Iversen, P. L.

2003-01-01

Antisense phosphorodiamidate morpholino oligomers (PMOs) were tested for the ability to inhibit gene expression in Escherichia coli. PMOs targeted to either a myc-luciferase reporter gene product or 16S rRNA did not inhibit luciferase expression or growth. However, in a strain with defective lipopolysaccharide (lpxA mutant), which has a leaky outer membrane, PMOs targeted to the myc-luciferase or acyl carrier protein (acpP) mRNA significantly inhibited their targets in a dose-dependent response. A significant improvement was made by covalently joining the peptide (KFF)3KC to the end of PMOs. In strains with an intact outer membrane, (KFF)3KC-myc PMO inhibited luciferase expression by 63%. A second (KFF)3KC-PMO conjugate targeted to lacI mRNA induced β-galactosidase in a dose-dependent response. The end of the PMO to which (KFF)3KC is attached affected the efficiency of target inhibition but in various ways depending on the PMO. Another peptide-lacI PMO conjugate was synthesized with the cationic peptide CRRRQRRKKR and was found not to induce β-galactosidase. We conclude that the outer membrane of E. coli inhibits entry of PMOs and that (KFF)3KC-PMO conjugates are transported across both membranes and specifically inhibit expression of their genetic targets. PMID:14506035

Particle simulations on transport control in divertors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kashiwagi, Mieko; Ido, Shunji

1995-04-01

Particle orbit simulations are carried out to study the reflection of He ions recycled from a tokamak divertor by RF electric fields, which have the frequency close to ion cyclotron resonance frequency (ICRF). The performance of particle reflection and the requirement to the intensity of RF fields are studied. The control of He recycling by ICRF fields is found to be available. 4 refs., 4 figs.

DOE Office of Scientific and Technical Information (OSTI.GOV)

M. Ono; Jaworski, M.; Kaita, R.

Developing a reactor compatible divertor and managing the associated plasma material interaction (PMI) has been identified as a high priority research area for magnetic confinement fusion. Accordingly on NSTX-U, the PMI research has received a strong emphasis. Moreover, with ˜15 MW of auxiliary heating power, NSTX-U will be able to test the PMI physics with the peak divertor plasma facing component (PFC) heat loads of up to 40-60 MW/m 2.

Experiments and numerical modeling of fast flowing liquid metal thin films under spatially varying magnetic field conditions

NASA Astrophysics Data System (ADS)

Narula, Manmeet Singh

Innovative concepts using fast flowing thin films of liquid metals (like lithium) have been proposed for the protection of the divertor surface in magnetic fusion devices. However, concerns exist about the possibility of establishing the required flow of liquid metal thin films because of the presence of strong magnetic fields which can cause flow disrupting MHD effects. A plan is underway to design liquid lithium based divertor protection concepts for NSTX, a small spherical torus experiment at Princeton. Of these, a promising concept is the use of modularized fast flowing liquid lithium film zones, as the divertor (called the NSTX liquid surface module concept or NSTX LSM). The dynamic response of the liquid metal film flow in a spatially varying magnetic field configuration is still unknown and it is suspected that some unpredicted effects might be lurking. The primary goal of the research work being reported in this dissertation is to provide qualitative and quantitative information on the liquid metal film flow dynamics under spatially varying magnetic field conditions, typical of the divertor region of a magnetic fusion device. The liquid metal film flow dynamics have been studied through a synergic experimental and numerical modeling effort. The Magneto Thermofluid Omnibus Research (MTOR) facility at UCLA has been used to design several experiments to study the MHD interaction of liquid gallium films under a scaled NSTX outboard divertor magnetic field environment. A 3D multi-material, free surface MHD modeling capability is under development in collaboration with HyPerComp Inc., an SBIR vendor. This numerical code called HIMAG provides a unique capability to model the equations of incompressible MHD with a free surface. Some parts of this modeling capability have been developed in this research work, in the form of subroutines for HIMAG. Extensive code debugging and benchmarking exercise has also been carried out. Finally, HIMAG has been used to study the MHD interaction of fast flowing liquid metal films under various divertor relevant magnetic field configurations through numerical modeling exercises.

Experimental and analytical studies of high heat flux components for fusion experimental reactor

NASA Astrophysics Data System (ADS)

Araki, Masanori

1993-03-01

In this report, the experimental and analytical results concerning the development of plasma facing components of ITER are described. With respect to developing high heat removal structures for the divertor plates, an externally-finned swirl tube was developed based on the results of critical heat flux (CHF) experiments on various tube structures. As the result, the burnout heat flux, which also indicates incident CHF, of 41 (+/-) 1 MW/sq m was achieved in the externally-finned swirl tube. The applicability of existing CHF correlations based on uniform heating conditions was evaluated by comparing the CHF experimental data with the smooth and the externally-finned tubes under one-sided heating condition. As the results, experimentally determined CHF data for straight tube show good agreement, for the externally-finned tube, no existing correlations are available for prediction of the CHF. With respect to the evaluation of the bonds between carbon-based material and heat sink metal, results of brazing tests were compared with the analytical results by three dimensional model with temperature-dependent thermal and mechanical properties. Analytical results showed that residual stresses from brazing can be estimated by the analytical three directional stress values instead of the equivalent stress value applied. In the analytical study on the separatrix sweeping for effectively reducing surface heat fluxes on the divertor plate, thermal response of the divertor plate was analyzed under ITER relevant heat flux conditions and has been tested. As the result, it has been demonstrated that application of the sweeping technique is very effective for improvement in the power handling capability of the divertor plate and that the divertor mock-up has withstood a large number of additional cyclic heat loads.

The Simple Map for a Single-null Divertor Tokamak: How to Find the Footprint of Field lines

NASA Astrophysics Data System (ADS)

Figgins, Montoya; Ali, Halima; Punjabi, Alkesh

2000-10-01

We are working with the Simple Map^1 to find the footprint of field lines on the diverter plate in a single-null tokamak. Footprint of a field line is the position of the line when it escapes across the divertor plate. The Simple Map represents the magnetic field in a single-null divertor tokamak. The path of a field line is given by the equations: X_n+1=X_n-kY_n(1-Y_n) and Y_n+1=Y_n+kX_n+1. In order to find the footprint, we must first find the last good surface which is Y=0.997135768 and X=0. The value of k is fixed at 0.6. The starting values X0 are fixed at X_0=0. We use 10,000 points between the last good surface and the X-point. The X-point is located at (0,1). We also use the Continuous Analog of the Simple Map given by the equations: X(φ)=X_0-kY0 (1-Y_0)φ and Y(φ)=Y_0+kX(φ)φ. This will tell us what the (φ,X) is which represents the field lines crossing the divertor plate. The divertor plate is located at Y=1. When graphed, the footprint of field lines looks like the rings of Saturn. This work is supported by US DOES OFES. Ms. Montoya Figgins is HU CFRT Summer Fusion High School Scholar from E. E. Smith High School in North Carolina. She is supported by NASA under its NASA SHARP Plus Program. 1. Punjabi A, Verma A, and Boozer A, Phys Rev Lett, 69, 3322 (1992) and J Plasma Phys, 52, 91 (1994)

Striation pattern of target particle and heat fluxes in three dimensional simulations for DIII-D [On the striation pattern of target particle and heat fluxes in three dimensional simulations for DIII-D

DOE PAGES

Frerichs, H.; Schmitz, Oliver; Reiter, D.; ...

2014-02-04

The application of resonant magnetic perturbations (RMPs) results in a non-axisymmetric striation pattern of magnetic field lines from the plasma interior which intersect the divertor targets. The impact on related particle and heat fluxes is investigated by three dimensional computer simulations for two different recycling conditions (controlled via neutral gas pumping). It is demonstrated that a mismatch between the particle and heat flux striation pattern, as is repeatedly observed in ITER similar shape H-mode plasmas at DIII-D, can be reproduced by the simulations for high recycling conditions at the onset of partial detachment. Finally, these results indicate that a detailedmore » knowledge of the particle and energy balance is at least as important for realistic simulations as the consideration of a change in the magnetic field structure by plasma response effects.« less

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

PubMed

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

2012-10-01

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

Analysis of the plasma-wall interaction in the Heliotron E device

NASA Astrophysics Data System (ADS)

Motojima, O.; Mizuuchi, T.; Besshou, S.; Iiyoshi, A.; Uo, K.; Yamashina, T.; Mohri, M.; Satake, T.; Hashiba, M.; Amemiya, S.; Miwa, H.

1984-12-01

The plasma-wall interaction (PWI) of the currentless plasmas with temperature To, Tio ≤ 1.1 keV, density N¯e = (2-10)× 1013/cm3, and volume-averaged beta value of β$¯≤ 2% was investigated. We have observed that PWI took place mainly where the divertor field line intersected the chamber wall (called divertor traces). Boundary plasmas were measured with electrostatic probes, which showed the presence of the divertor region with the parameters in the range of Ned = 1010-1011/cm3 and Ted = 10-50 eV. Surface analysis techniques (ESCA, AES, and RBS) were applied to analyze the surface probes (Si, graphite and stainless steel) and the test pieces (SiC, TiC, and stainless steel), which were irradiated by plasmas for short and long times respectively.

Studies of relationships among outer solar system small bodies and related objects

NASA Technical Reports Server (NTRS)

Hartmann, William K.

1991-01-01

This program involves telescopic observations of colorimetry, spectroscopy, and photometry of small bodies of the solar system, emphasizing possible relationships among outer solar system asteroids, comets, and certain satellites. Earth approacher targets of opportunity and lab spectroscopic studies are included.

Numerical studies of edge localized instabilities in tokamaks

NASA Astrophysics Data System (ADS)

Wilson, H. R.; Snyder, P. B.; Huysmans, G. T. A.; Miller, R. L.

2002-04-01

A new computational tool, edge localized instabilities in tokamaks equilibria (ELITE), has been developed to help our understanding of short wavelength instabilities close to the edge of tokamak plasmas. Such instabilities may be responsible for the edge localized modes observed in high confinement H-mode regimes, which are a serious concern for next step tokamaks because of the high transient power loads which they can impose on divertor target plates. ELITE uses physical insight gained from analytic studies of peeling and ballooning modes to provide an efficient way of calculating the edge ideal magnetohydrodynamic stability properties of tokamaks. This paper describes the theoretical formalism which forms the basis for the code.

Studies of Be migration in the JET tokamak using AMS with 10Be marker

NASA Astrophysics Data System (ADS)

Bykov, I.; Bergsåker, H.; Possnert, G.; Zhou, Y.; Heinola, K.; Pettersson, J.; Conroy, S.; Likonen, J.; Petersson, P.; Widdowson, A.

2016-03-01

The JET tokamak is operated with beryllium limiter tiles in the main chamber and tungsten coated carbon fiber composite tiles and solid W tiles in the divertor. One important issue is how wall materials are migrating during plasma operation. To study beryllium redistribution in the main chamber and in the divertor, a 10Be enriched limiter tile was installed prior to plasma operations in 2011-2012. Methods to take surface samples have been developed, an abrasive method for bulk Be tiles in the main chamber, which permits reuse of the tiles, and leaching with hot HCl to remove all Be deposited at W coated surfaces in the divertor. Quantitative analysis of the total amount of Be in cm2 sized samples was made with inductively coupled plasma atomic emission spectroscopy (ICP-AES). The 10Be/9Be ratio in the samples was measured with accelerator mass spectrometry (AMS). The experimental setup and methods are described in detail, including sample preparation, measures to eliminate contributions in AMS from the 10B isobar, possible activation due to plasma generated neutrons and effects of diffusive isotope mixing. For the first time marker concentrations are measured in the divertor deposits. They are in the range 0.4-1.2% of the source concentration, with moderate poloidal variation.

In-Vessel Tritium Retention and Removal in ITER-FEAT

NASA Astrophysics Data System (ADS)

Federici, G.; Brooks, J. N.; Iseli, M.; Wu, C. H.

Erosion of the divertor and first-wall plasma-facing components, tritium uptake in the re-deposited films, and direct implantation in the armour material surfaces surrounding the plasma, represent crucial physical issues that affect the design of future fusion devices. In this paper we present the derivation, and discuss the results, of current predictions of tritium inventory in ITER-FEAT due to co-deposition and implantation and their attendant uncertainties. The current armour materials proposed for ITER-FEAT are beryllium on the first-wall, carbon-fibre-composites on the divertor plate near the separatrix strike points, to withstand the high thermal loads expected during off-normal events, e.g., disruptions, and tungsten elsewhere in the divertor. Tritium co-deposition with chemically eroded carbon in the divertor, and possibly with some Be eroded from the first-wall, is expected to represent the dominant mechanism of in-vessel tritium retention in ITER-FEAT. This demands efficient in-situ methods of mitigation and retrieval to avoid frequent outages due to the reaching of precautionary operating limits set by safety considerations (e.g., ˜350 g of in-vessel co-deposited tritium) and for fuel economy reasons. Priority areas where further R&D work is required to narrow the remaining uncertainties are also briefly discussed.

Overview of Compact Toroidal Hybrid research program progress and plans

NASA Astrophysics Data System (ADS)

Maurer, David; Ennis, David; Hanson, James; Hartwell, Gregory; Herfindal, Jeffrey; Knowlton, Stephen; Ma, Xingxing; Pandya, Mihir; Roberds, Nicholas; Ross, Kevin; Traverso, Peter

2016-10-01

disruptive behavior on the level of applied 3D magnetic shaping; (2) test and advance the V3FIT reconstruction code and NIMROD modeling of CTH; and (3) study the implementation of an island divertor. Progress towards these goals and other developments are summarized. The disruptive density limit exceeds the Greenwald limit as the vacuum transform is increased, but a threshold for avoidance is not observed. Low- q disruptions, with 1.1 < q (a) <2.0, cease to occur if the vacuum transform is raised above 0.07. Application of vacuum transform can reduce and eliminate the vertical drift of elongated discharges that would otherwise be vertically unstable. Reconstructions using external magnetics give accurate estimates for quantities near the plasma boundary, and internal diagnostics have been implemented to extend the range of accuracy into the plasma core. Sawtooth behavior has been reproducibly modified with external transform and NIMROD is used to model these observations and reproduces experimental trends. An island divertor design has begun with connection length studies to model energy deposition on divertor plates located in an edge 1/3 island as well as the study of a non-resonant divertor configuration. This work is supported by U.S. Department of Energy Grant No. DE-FG02-00ER54610.

Symmetric Simple Map with Dipole Map for a Single-Null Divertor Tokamak

NASA Astrophysics Data System (ADS)

Ali, Halima; Watson, Michael; Punjabi, Alkesh; Boozer, Allen

1996-11-01

This investigation focuses on the effects of an externally placed dipole coil on the magnetic topology of a single-null divertor tokamak with a stochastic scrape-off layer using the Method of Maps (Punjabi A, Verma A and Boozer A, Phys Rev Lett), 69, 3322 (1992) and J Plasma Phys, 52, 91 (1994). The unperturbed magnetic topology is represented by the Symmetric Simple Map (Ali H, Watson M, Mayer C, Punjabi A and Boozer A, Bull Am Phys Soc), 40, 1855 (1995). The effect of dipole perturbation is repesented by the Dipole Map (Ali H, Watson M, Punjabi A and Boozer A, Sherwood Mtg), paper 1C20 (1996). A single dipole coil is placed across from the X-point below the last good surface. The strength of the dipole perturbation and the distance of the coil from the last good surface are varied. We observe that the dipole perturbation causes spatially intermittent chaos. This has significant implications for radiative divertor concepts as well for impurity control. We also present the detailed results on the effects of the dipole coil on the properties of the stochastic layer and the footprint of the field lines on the divertor plate. This work is supported by the US DOE OFES.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

1994-12-31

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

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

NASA Astrophysics Data System (ADS)

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

2017-12-01

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

Lunar Dust Monitor to BE Onboard the Next Japanese Lunar Mission SELENE-2

NASA Astrophysics Data System (ADS)

Ohashi, Hideo

The next Japanese lunar mission SELENE-2, after a successful mission Kaguya (a project named SELENE), is planned to be launched in mid 2010s and is consisted of a lander, a rover, and an orbiter, as a transmitting satellite to the earth. A dust particle detector LDM (Lunar Dust Monitor) is proposed to be onboard the orbiter. The LDM is an impact ionization detector with dimensions 25 cm × 25 cm × 30 cm, and it has a sensor part (LDM-S, upper module) and an electronics part (LDM-E, lower module). The LDM-S has a large target (gold-plated Al) of 400 cm2 , to which a high voltage of +500 V is applied. The LDM-S also has two meshed grids parallel to the target. The grids are etched stainless steel with 90% transparency: the inner grid is 2 cm apart from the target and the outer grid is 15 cm from the target. When a charged dust particle passes through the outer and inner grids, it induces an electric signal on the grids separated by a certain time interval, determined by the velocity of the incident particle and the distance between the outer and inner grids. By measuring the time interval, we can calculate the velocity of the particle, with the ambiguity of its trajectory to the target. When the incident particle impacts on the target, plasma gas of electrons and ions is generated. The electrons of the plasma are collected by the target and the ions are accelerated toward the inner grids as a result of the electric field. Some of the ions drift through the inner grid and reach the outer grid. The outer and inner grids and the target are connected to charge-sensitive amplifiers, which convert charge signals induced by the electrons and ions to voltage signals that are fed to a following flash ADC driven with 10 MHz. The waveforms from two grids and the target can be stored and be sent back to ground for data analysis. We can deduce the mass and velocity information of the incident dust particle from the recorded waveforms. The orbiter of SELENE-2 is planned to be in operation for one year or more, and the LDM will observe circumlunar dust for as long as possible. We report scientific importance of dust measurement around the Moon, and current status of LDM in this conference.

Retention of ferrofluid aggregates at the target site during magnetic drug targeting

NASA Astrophysics Data System (ADS)

Asfer, Mohammed; Saroj, Sunil Kumar; Panigrahi, Pradipta Kumar

2017-08-01

The present study reports the retention dynamics of a ferrofluid aggregate localized at the target site inside a glass capillary (500 × 500 μm2 square cross section) against a bulk flow of DI water (Re = 0.16 and 0.016) during the process of magnetic drug targeting (MDT). The dispersion dynamics of iron oxide nanoparticles (IONPs) into bulk flow for different initial size of aggregate at the target site is reported using the brightfield visualization technique. The flow field around the aggregate during the retention is evaluated using the μPIV technique. IONPs at the outer boundary experience a higher shear force as compared to the magnetic force, resulting in dispersion of IONPs into the bulk flow downstream to the aggregate. The blockage effect and the roughness of the outer boundary of the aggregate resulting from chain like clustering of IONPs contribute to the flow recirculation at the downstream region of the aggregate. The entrapment of seeding particles inside the chain like clusters of IONPs at the outer boundary of the aggregate reduces the degree of roughness resulting in a streamlined aggregate at the target site at later time. The effect of blockage, structure of the aggregate, and disturbed flow such as recirculation around the aggregate are the primary factors, which must be investigated for the effectiveness of the MDT process for in vivo applications.

Effects of Gravity-Assist Timing on Outer-Planet Missions Using Solar-Electric Propulsion

NASA Technical Reports Server (NTRS)

Woo, Byoungsam; Coverstone, Victoria L.; Cupples, Michael

2004-01-01

Missions to the outer planets for spacecraft with a solar-electric propulsion system (SEPS) and that utilize a single Venus gravity assist are investigated. The trajectories maximize the delivered mass to the target planet for a range of flight times. A comparison of the trajectory characteristics (delivered mass, launch energy and onboard propulsive energy) is made for various Venus gravity assist opportunities. Methods to estimate the delivered mass to the outer planets are developed.

Reactor plasma facing component designs based on liquid metal concepts supported in porous systems

NASA Astrophysics Data System (ADS)

Tabarés, F. L.; Oyarzabal, E.; Martin-Rojo, A. B.; Tafalla, D.; de Castro, A.; Soleto, A.

2017-01-01

The use of liquid metals (LMs) as plasma facing components in fusion devices was proposed as early as 1970 for a field reversed concept and inertial fusion reactors. The idea was extensively developed during the APEX Project, at the turn of the century, and it is the subject at present of the biennial International Symposium on Lithium Applications (ISLA), whose fourth meeting took place in Granada, Spain at the end of September 2015. While liquid metal flowing concepts were specially addressed in USA research projects, the idea of embedding the metal in a capillary porous system (CPS) was put forwards by Russian teams in the 1990s, thus opening the possibility of static concepts. Since then, many ideas and accompanying experimental tests in fusion devices and laboratories have been produced, involving a large fraction of countries within the international fusion community. Within the EUROFusion Roadmap, these activities are encompassed into the working programs of the plasma facing components (PFC) and divertor tokamak test (DTT) packages. In this paper, a review of the state of the art in concepts based on the CPS set-up for a fusion reactor divertor target, aimed at preventing the ejection of the liquid metal by electro-magnetic (EM) forces generated under plasma operation, is described and required R+D activities on the topic, including ongoing work at CIEMAT specifically oriented to filling the remaining gaps, are stressed.

Discharge start-up and ramp-up development for NSTX-U and MAST-U

NASA Astrophysics Data System (ADS)

Battaglia, D. J.; Boyer, M. D.; Gerhardt, S. P.; Menard, J. E.; Mueller, D.; Cunningham, G.; Kirk, A.; Kogan, L.; McArdle, G.; Pangione, L.; Thornton, A. J.; Ren, E.

2017-10-01

A collaborative modeling effort is underway to develop robust inductive start-up and ramp-up scenarios for NSTX-U and MAST-U. These complementary spherical tokamak devices aim to generate the physics basis for achieving steady-state, high-beta and high-confinement plasma discharges with a self-consistent solution for managing the divertor heat flux. High-performance discharges in these devices require sufficient plasma elongation (κ = 2.4 - 2.8) to maximize the bootstrap and beam-driven current drive, increase MHD stability at high Ip and high βN, and realize advanced divertor geometries such as the snowflake and super-X. Achieving the target elongation on NSTX-U is enabled by an L-H transition in the current ramp-up that slows the current diffusion and maintains a low internal inductance (li <= 0.8). Modeling focuses on developing scenarios that achieve a suitable field null for breakdown and discharge conditions conducive to an early L-H transition while maintaining vertical and MHD stability, with appropriate margin for variation in experimental conditions. The toroidal currents induced in conducting structures and the specifications of the real-time control and power supply systems are unique constraints for the two devices. Work Supported by U.S. DOE Contract No. DE-AC02-09CH11466 and the RCUK Energy Programme [Grant Number EP/P012450/1].

The pattern of parallel edge plasma flows due to pressure gradients, recycling, and resonant magnetic perturbations in DIII-D

DOE PAGES

Frerichs, H.; Schmitz, Oliver; Evans, Todd; ...

2015-07-13

High resolution plasma transport simulations with the EMC3-EIRENE code have been performed to address the parallel plasma flow structure in the boundary of a poloidal divertor configuration with non-axisymmetric perturbations at DIII-D. Simulation results show that a checkerboard pattern of flows with alternating direction is generated inside the separatrix. This pattern is aligned with the position of the main resonances (i.e. where the safety factor is equal to rational values q = m/n for a perturbation field with base mode number n): m pairs of alternating forward and backward flow channel exist for each resonance. The poloidal oscillations are alignedmore » with the subharmonic Melnikov function, which indicates that the plasma flow is generated by parallel pressure gradients along perturbed field lines. Lastly, an additional scrape-off layer-like domain is introduced by the perturbed separatrix which guides field lines from the interior to the divertor targets, resulting in an enhanced outward flow that is consistent with the experimentally observed particle pump-out effect. However, while the lobe structure of the perturbed separatrix is very well reflected in the temperature profile, the same lobes can appear to be smaller in the flow profile due to a competition between high upstream pressure and downstream particle sources driving flows in opposite directions.« less

Overview of Alcator C-Mod Research

NASA Astrophysics Data System (ADS)

White, A. E.

2017-10-01

Alcator C-Mod, a compact (R =0.68m, a =0.21m), high magnetic field, Bt <= 8T, tokamak accesses a variety of naturally ELM-suppressed high confinement regimes that feature extreme power density into the divertor, q|| <= 3 GW/m2, with SOL heat flux widths λq <0.5mm, exceeding conditions expected in ITER and approaching those foreseen in power plants. The unique parameter range provides much of the physics basis of a high-field, compact tokamak reactor. Research spans the topics of core transport and turbulence, RF heating and current drive, pedestal physics, scrape-off layer, divertor and plasma wall interactions. In the last experimental campaign, Super H-mode was explored and featured the highest pedestal pressures ever recorded, pped 90 kPa (90% of ITER target), consistent with EPED predictions. Optimization of naturally ELM-suppressed EDA H-modes accessed the highest volume averaged pressures ever achieved (〈p〉>2 atm), with pped 60 kPa. The SOL heat flux width has been measured at Bpol = 1.25T, confirming the Eich scaling over a broader poloidal field range than before. Multi-channel transport studies focus on the relationship between momentum transport and heat transport with perturbative experiments and new multi-scale gyrokinetic simulation validation techniques were developed. U.S. Department of Energy Grant No. DE-FC02-99ER54512.

Effect of ELMs on deuterium-loaded-tungsten plasma facing components

NASA Astrophysics Data System (ADS)

Umstadter, K. R.; Rudakov, D. L.; Wampler, W.; Watkins, J. G.; Wong, C. P. C.

2011-08-01

Prior heat pulse testing of plasma facing components (PFCs) has been completed in vacuum environments without the presence of background plasma. Edge localized modes (ELMs) will not be this kind of isolated event and one should know the effect of a plasma background during these transients. Heat-pulse experiments have been conducted in the PISCES-A device utilizing laser heating in a divertor-like plasma background. Initial results indicate that the erosion of PFCs is enhanced as compared to heat pulse or plasma only tests. To determine if the enhanced erosion effect is a phenomena only witnessed in the laboratory PISCES device, tungsten and graphite samples were exposed to plasmas in the lower divertor of the DIII-D tokamak using the Divertor Material Evaluation System (DiMES). Mass loss analysis indicates that materials that contain significant deuterium prior to experiencing a transient heating event will erode faster than those that have no or little retained deuterium.

Performance of V-4Cr-4Ti material exposed to DIII-D tokamak environment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tsai, H.; Chung, H.M.; Smith, D.L.

1997-04-01

Test specimens made with the 832665 heat of V-4Cr-4Ti alloy were exposed in the DIII-D tokamak environment to support the installation of components made of a V-4Cr-4Ti alloy in the radiative divertor of the DIII-D. Some of the tests were conducted with the Divertor Materials Evaluation System (DiMES) to study the short-term effects of postvent bakeout, when concentrations of gaseous impurities in the DIII-D chamber are the highest. Other specimens were mounted next to the chamber wall behind the divertor baffle plate, to study the effects of longer-term exposures. By design, none of the specimens directly interacted with the plasma.more » Preliminary results from testing the exposed specimens indicate only minor degradation of mechanical properties. Additional testing and microstructural characterization are in progress.« less

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

NASA Astrophysics Data System (ADS)

You, Jeong-Ha

2005-01-01

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

Long-term fuel retention in JET ITER-like wall

NASA Astrophysics Data System (ADS)

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

2016-02-01

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

High density operation for reactor-relevant power exhaust

NASA Astrophysics Data System (ADS)

Wischmeier, M.; ASDEX Upgrade Team; Jet Efda Contributors

2015-08-01

With increasing size of a tokamak device and associated fusion power gain an increasing power flux density towards the divertor needs to be handled. A solution for handling this power flux is crucial for a safe and economic operation. Using purely geometric arguments in an ITER-like divertor this power flux can be reduced by approximately a factor 100. Based on a conservative extrapolation of current technology for an integrated engineering approach to remove power deposited on plasma facing components a further reduction of the power flux density via volumetric processes in the plasma by up to a factor of 50 is required. Our current ability to interpret existing power exhaust scenarios using numerical transport codes is analyzed and an operational scenario as a potential solution for ITER like divertors under high density and highly radiating reactor-relevant conditions is presented. Alternative concepts for risk mitigation as well as strategies for moving forward are outlined.

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

We conducted physics experiments: record normalized {Beta} = 4.9 achieved in VH-mode, {Beta} limits of ITER-like configurations evaluated, FWCD commissioning. The tokamak vessel was opened to atmosphere for six weeks and a number of key diagnostics for understanding the divertor were installed. The DIII-D Advisory Committee met in January to review the DIII-D program and plan. They commended us for recent progress and supported the vanadium divertor design. The U.S./Japan DIII-D steering committee met and recommended extending the agreement to the year 2000. The field work proposal for FY 96/97 was presented in Washington on March 29, 1995. A reviewmore » of the DIII-D plan to install vanadium structural components as part of the new radiative divertor modification was held in Washington 31, 1995 and the panel endorsed the plans. Preliminary plans were developed with PPPL for collaborations in FY96,« less

Tokamak reactor for treating fertile material or waste nuclear by-products

DOEpatents

Kotschenreuther, Michael T.; Mahajan, Swadesh M.; Valanju, Prashant M.

2012-10-02

Disclosed is a tokamak reactor. The reactor includes a first toroidal chamber, current carrying conductors, at least one divertor plate within the first toroidal chamber and a second chamber adjacent to the first toroidal chamber surrounded by a section that insulates the reactor from neutrons. The current carrying conductors are configured to confine a core plasma within enclosed walls of the first toroidal chamber such that the core plasma has an elongation of 1.5 to 4 and produce within the first toroidal chamber at least one stagnation point at a perpendicular distance from an equatorial plane through the core plasma that is greater than the plasma minor radius. The at least one divertor plate and current carrying conductors are configured relative to one another such that the current carrying conductors expand the open magnetic field lines at the divertor plate.

Applications of Collisional Radiative Modeling of Helium and Deuterium for Image Tomography Diagnostic of Te, Ne, and ND in the DIII-D Tokamak

NASA Astrophysics Data System (ADS)

Munoz Burgos, J. M.; Brooks, N. H.; Fenstermacher, M. E.; Meyer, W. H.; Unterberg, E. A.; Schmitz, O.; Loch, S. D.; Balance, C. P.

2011-10-01

We apply new atomic modeling techniques to helium and deuterium for diagnostics in the divertor and scrape-off layer regions. Analysis of tomographically inverted images is useful for validating detachment prediction models and power balances in the divertor. We apply tomographic image inversion from fast tangential cameras of helium and Dα emission at the divertor in order to obtain 2D profiles of Te, Ne, and ND (neutral ion density profiles). The accuracy of the atomic models for He I will be cross-checked against Thomson scattering measurements of Te and Ne. This work summarizes several current developments and applications of atomic modeling into diagnostic at the DIII-D tokamak. Supported in part by the US DOE under DE-AC05-06OR23100, DE-FC02-04ER54698, DE-AC52-07NA27344, and DE-AC05-00OR22725.

Design and Test of Wendelstein 7-X Water-Cooled Divertor Scraper

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boscary, J.; Greuner, Henri; Ehrke, Gunnar

Heat load calculations have indicated the possible overloading of the ends of the water-cooled divertor facing the pumping gap beyond their technological limit. The intention of the scraper is the interception of some of the plasma fluxes both upstream and downstream before they reach the divertor surface. The scraper is divided into six modules of four plasma facing components (PFCs); each module has four PFCs hydraulically connected in series by two water boxes (inlet and outlet). A full-scale prototype of one module has been manufactured. Development activities have been carried out to connect the water boxes to the cooling pipesmore » of the PFCs by tungsten inert gas internal orbital welding. This prototype was successfully tested in the GLADIS facility with 17 MW/m2 for 500 cycles. The results of these activities have confirmed the possible technological basis for a fabrication of the water-cooled scraper.« less

Characterization of the targeting signal in mitochondrial β-barrel proteins

PubMed Central

Jores, Tobias; Klinger, Anna; Groß, Lucia E.; Kawano, Shin; Flinner, Nadine; Duchardt-Ferner, Elke; Wöhnert, Jens; Kalbacher, Hubert; Endo, Toshiya; Schleiff, Enrico; Rapaport, Doron

2016-01-01

Mitochondrial β-barrel proteins are synthesized on cytosolic ribosomes and must be specifically targeted to the organelle before their integration into the mitochondrial outer membrane. The signal that assures such precise targeting and its recognition by the organelle remained obscure. In the present study we show that a specialized β-hairpin motif is this long searched for signal. We demonstrate that a synthetic β-hairpin peptide competes with the import of mitochondrial β-barrel proteins and that proteins harbouring a β-hairpin peptide fused to passenger domains are targeted to mitochondria. Furthermore, a β-hairpin motif from mitochondrial proteins targets chloroplast β-barrel proteins to mitochondria. The mitochondrial targeting depends on the hydrophobicity of the β-hairpin motif. Finally, this motif interacts with the mitochondrial import receptor Tom20. Collectively, we reveal that β-barrel proteins are targeted to mitochondria by a dedicated β-hairpin element, and this motif is recognized at the organelle surface by the outer membrane translocase. PMID:27345737

Outer Membrane Targeting of Passenger Proteins by the Vacuolating Cytotoxin Autotransporter of Helicobacter pylori

PubMed Central

Fischer, Wolfgang; Buhrdorf, Renate; Gerland, Elke; Haas, Rainer

2001-01-01

Helicobacter pylori produces a number of proteins associated with the outer membrane, including adhesins and the vacuolating cytotoxin. These proteins are supposed to integrate into the outer membrane by β-barrel structures, characteristic of the family of autotransporter proteins. By using the SOMPES (shuttle vector-based outer membrane protein expression) system for outer membrane protein production, we were able to functionally express in H. pylori the cholera toxin B subunit genetically fused to the C-terminal VacA domain. We demonstrate that the fusion protein is translocated to the H. pylori outer membrane and that the CtxB domain is exposed on the H. pylori surface. Thus, we provide the first experimental evidence that the C-terminal β-domain of VacA can transport a foreign passenger protein to the H. pylori surface and hence acts as a functional autotransporter. PMID:11598049

Upgrades of edge, divertor and scrape-off layer diagnostics of W7-X for OP1.2

DOE PAGES

Hathiramani, D.; Ali, A.; Anda, G.; ...

2018-02-07

In this work, Wendelstein 7-X (W7-X) is the world’s largest superconducting nuclear fusion experiment of the optimized stellarator type. In the first Operation Phase (OP1.1) helium and hydrogen plasmas were studied in limiter configuration. The heating energy was limited to 4 MJ and the main purpose of that campaign was the integral commissioning of the machine and diagnostics, which was achieved very successfully. Already from the beginning a comprehensive set of diagnostics was available to study the plasma. On the path towards high-power, high-performance plasmas, W7-X will be stepwise upgraded from an inertially cooled (OP1.2, limited to 80 MJ) tomore » an actively cooled island divertor (OP2, 10 MW steady-state plasma operation). The machine is prepared for OP1.2 with 10 inertially cooled divertor units, and the experimental campaign has started recently.The paper describes a subset of diagnostics which will be available for OP1.2 to study the plasma edge, divertor and scrape-off layer physics including those already available for OP1.1, plus modifications, upgrades and new systems. In conclusion, the focus of this summary will be on technical and engineering aspects, like feasibility and assembly but also on reliability, thermal loads and shielding against magnetic fields.« less

Fast wave power flow along SOL field lines in NSTX

NASA Astrophysics Data System (ADS)

Perkins, R. J.; Bell, R. E.; Diallo, A.; Gerhardt, S.; Hosea, J. C.; Jaworski, M. A.; Leblanc, B. P.; Kramer, G. J.; Phillips, C. K.; Roquemore, L.; Taylor, G.; Wilson, J. R.; Ahn, J.-W.; Gray, T. K.; Green, D. L.; McLean, A.; Maingi, R.; Ryan, P. M.; Jaeger, E. F.; Sabbagh, S.

2012-10-01

On NSTX, a major loss of high-harmonic fast wave (HHFW) power can occur along open field lines passing in front of the antenna over the width of the scrape-off layer (SOL). Up to 60% of the RF power can be lost and at least partially deposited in bright spirals on the divertor floor and ceiling [1,2]. The flow of HHFW power from the antenna region to the divertor is mostly aligned along the SOL magnetic field [3], which explains the pattern of heat deposition as measured with infrared (IR) cameras. By tracing field lines from the divertor back to the midplane, the IR data can be used to estimate the profile of HHFW power coupled to SOL field lines. We hypothesize that surface waves are being excited in the SOL, and these results should benchmark advanced simulations of the RF power deposition in the SOL (e.g., [4]). Minimizing this loss is critical optimal high-power long-pulse ICRF heating on ITER while guarding against excessive divertor erosion.[4pt] [1] J.C. Hosea et al., AIP Conf Proceedings 1187 (2009) 105. [0pt] [2] G. Taylor et al., Phys. Plasmas 17 (2010) 056114. [0pt] [3] R.J. Perkins et al., to appear in Phys. Rev. Lett. [0pt] [4] D.L. Green et al., Phys. Rev. Lett. 107 (2011) 145001.

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

TEMPEST simulations of the plasma transport in a single-null tokamak geometry

NASA Astrophysics Data System (ADS)

Xu, X. Q.; Bodi, K.; Cohen, R. H.; Krasheninnikov, S.; Rognlien, T. D.

2010-06-01

We present edge kinetic ion transport simulations of tokamak plasmas in magnetic divertor geometry using the fully nonlinear (full-f) continuum code TEMPEST. Besides neoclassical transport, a term for divergence of anomalous kinetic radial flux is added to mock up the effect of turbulent transport. To study the relative roles of neoclassical and anomalous transport, TEMPEST simulations were carried out for plasma transport and flow dynamics in a single-null tokamak geometry, including the pedestal region that extends across the separatrix into the scrape-off layer and private flux region. A series of TEMPEST simulations were conducted to investigate the transition of midplane pedestal heat flux and flow from the neoclassical to the turbulent limit and the transition of divertor heat flux and flow from the kinetic to the fluid regime via an anomalous transport scan and a density scan. The TEMPEST simulation results demonstrate that turbulent transport (as modelled by large diffusion) plays a similar role to collisional decorrelation of particle orbits and that the large turbulent transport (large diffusion) leads to an apparent Maxwellianization of the particle distribution. We also show the transition of parallel heat flux and flow at the entrance to the divertor plates from the fluid to the kinetic regime. For an absorbing divertor plate boundary condition, a non-half-Maxwellian is found due to the balance between upstream radial anomalous transport and energetic ion endloss.

Characterization of mechanisms of quinolone resistance in Pseudomonas aeruginosa strains isolated in vitro and in vivo during experimental endocarditis.

PubMed Central

Chamberland, S; Bayer, A S; Schollaardt, T; Wong, S A; Bryan, L E

1989-01-01

Mechanisms of resistance to quinolones were characterized in Pseudomonas aeruginosa strains isolated after Tn5 insertional mutagenesis and in resistant strains that emerged during pefloxacin therapy of experimental aortic endocarditis. Quinolone resistance achieved in in vitro-selected mutants Qr-1 and Qr-2 was associated with cross-resistance to several groups of antimicrobial agents, including beta-lactams, tetracycline, and chloramphenicol. A significant reduction of norfloxacin uptake was also observed. After ether permeabilization of the cells, DNA synthesis of these two isolates was as susceptible to norfloxacin as DNA synthesis of the parent strain (PAO1). These results indicate that alteration of outer membrane permeability is the primary determinant of resistance in these isolates. This altered cell permeability was correlated with reduction of outer membrane protein G (25.5 kilodaltons) and loss of a 40-kilodalton outer membrane protein in strain Qr-1. Resistance to quinolones that emerged during experimental endocarditis therapy was associated with both modification of outer membrane permeability (decreased uptake of norfloxacin) and decreased susceptibility of DNA synthesis to norfloxacin. Resistance was limited to quinolones and chloramphenicol. For these strains, norfloxacin inhibitory doses (50%) for DNA synthesis were identical to the drug MICs, suggesting that despite the identification of a permeability change, perhaps due to changes of lipopolysaccharide, the alteration of the quinolone intracellular target(s) susceptibility constitutes the primary determinant of resistance. Also, two distinct levels of norfloxacin resistance of DNA synthesis were found in these isolates, indicating that at least two distinct alterations of the drug target(s) are possible in P. aeruginosa. Images PMID:2502066

Progress towards modeling tokamak boundary plasma turbulence and understanding its role in setting divertor heat flux widths

NASA Astrophysics Data System (ADS)

Chen, Bin

2017-10-01

QCMs (quasi-coherent modes) are well characterized in the edge of Alcator C-Mod, when operating in the Enhanced Dα (EDA) H-mode, a promising alternative regime for ELM (edge localized modes) suppressed operation. To improve the understanding of the physics behind the QCMs, three typical C-Mod EDA H-Mode discharges are simulated by BOUT + + using a six-field two-fluid model (based on the Braginskii equations). The simulated characteristics of the frequency versus wave number spectra of the modes is in reasonable agreement with phase contrast imaging data. The key simulation results are: 1) Linear spectrum analysis and the nonlinear phase relationship indicate the dominance of resistive-ballooning modes and drift-Alfven wave instabilities; 2) QCMs originate inside the separatrix; (3) magnetic flutter causes the mode spreading into the SOL; 4) the boundary electric field Er changes the turbulent characteristics of the QCMs and controls edge transport and the divertor heat flux width; 5) the magnitude of the divertor heat flux depends on the physics models, such as sources and sinks, sheath boundary conditions, and parallel heat flux limiting coefficient. The BOUT + + simulations have also been performed for inter-ELM periods of DIII-D and EAST discharges, and similar quasi-coherent modes have been found. The parallel electron heat fluxes projected onto the target from these BOUT + + simulations follow the experimental heat flux width scaling, in particular the inverse dependence of the width on the poloidal magnetic field with an outlier. Further turbulence statistics analysis shows that the blobs are generated near the pedestal peak gradient region inside the separatrix and contribute to the transport of the particle and heat in the SOL region. To understand the Goldston heuristic drift-based model, results will also be presented from self-consistent transport simulations with the electric and magnetic drifts in BOUT + + and with the sheath potential included in the SOL. Work supported by LLNL under Contract DE-AC52-07NA27344. This work was also supported by US DOE Grant DE-FC02-99ER54512, using Alcator C-Mod, a DOE Office of Science User Facility, and under the auspices of the CSC (No. 201506340019).

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

NASA Astrophysics Data System (ADS)

Brezinsek, S.; JET-EFDA contributors

2015-08-01

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

Dynamic ELM and divertor control using resonant toroidal multi-mode magnetic fields in DIII-D and EAST

NASA Astrophysics Data System (ADS)

Sun, Youwen

2017-10-01

A rotating n = 2 Resonant Magnetic Perturbation (RMP) field combined with a stationary n = 3 RMP field has validated predictions that access to ELM suppression can be improved, while divertor heat and particle flux can also be dynamically controlled in DIII-D. Recent observations in the EAST tokamak indicate that edge magnetic topology changes, due to nonlinear plasma response to magnetic perturbations, play a critical role in accessing ELM suppression. MARS-F code MHD simulations, which include the plasma response to the RMP, indicate the nonlinear transition to ELM suppression is optimized by configuring the RMP coils to drive maximal edge stochasticity. Consequently, mixed toroidal multi-mode RMP fields, which produce more densely packed islands over a range of additional rational surfaces, improve access to ELM suppression, and further spread heat loading on the divertor. Beneficial effects of this multi-harmonic spectrum on ELM suppression have been validated in DIII-D. Here, the threshold current required for ELM suppression with a mixed n spectrum, where part of the n = 3 RMP field is replaced by an n = 2 field, is smaller than the case with pure n = 3 field. An important further benefit of this multi-mode approach is that significant changes of 3D particle flux footprint profiles on the divertor are found in the experiment during the application of a rotating n = 2 RMP field superimposed on a static n = 3 RMP field. This result was predicted by modeling studies of the edge magnetic field structure using the TOP2D code which takes into account plasma response from MARS-F code. These results expand physics understanding and potential effectiveness of the technique for reliably controlling ELMs and divertor power/particle loading distributions in future burning plasma devices such as ITER. Work supported by USDOE under DE-FC02-04ER54698 and NNSF of China under 11475224.

Multishell inertial confinement fusion target

DOEpatents

Holland, James R.; Del Vecchio, Robert M.

1984-01-01

A method of fabricating multishell fuel targets for inertial confinement fusion usage. Sacrificial hemispherical molds encapsulate a concentric fuel pellet which is positioned by fiber nets stretched tautly across each hemispherical mold section. The fiber ends of the net protrude outwardly beyond the mold surfaces. The joint between the sacrificial hemispheres is smoothed. A ceramic or glass cover is then deposited about the finished mold surfaces to produce an inner spherical surface having continuously smooth surface configuration. The sacrificial mold is removed by gaseous reaction accomplished through the porous ceramic cover prior to enclosing of the outer sphere by addition of an outer coating. The multishell target comprises the inner fuel pellet concentrically arranged within a surrounding coated cover or shell by fiber nets imbedded within the cover material.

Multishell inertial confinement fusion target

DOEpatents

Holland, James R.; Del Vecchio, Robert M.

1987-01-01

A method of fabricating multishell fuel targets for inertial confinement fusion usage. Sacrificial hemispherical molds encapsulate a concentric fuel pellet which is positioned by fiber nets stretched tautly across each hemispherical mold section. The fiber ends of the net protrude outwardly beyond the mold surfaces. The joint between the sacrificial hemispheres is smoothed. A ceramic or glass cover is then deposited about the finished mold surfaces to produce an inner spherical surface having continuously smooth surface configuration. The sacrificial mold is removed by gaseous reactions accomplished through the porous ceramic cover prior to enclosing of the outer sphere by addition of an outer coating. The multishell target comprises the inner fuel pellet concentrically arranged within a surrounding coated cover or shell by fiber nets imbedded within the cover material.

Visible and infrared investigations of planet-crossing asteroids and outer solar system objects

NASA Technical Reports Server (NTRS)

Tholen, David J.

1991-01-01

The project is supporting lightcurve photometry, colorimetry, thermal radiometry, and astrometry of selected asteroids. Targets include the planet-crossing population, particularly Earth approachers, which are believed to be the immediate source of terrestrial meteorites, future spacecraft targets, and those objects in the outer belt, primarily the Hilda and Trojan populations, that are dynamically isolated from the main asteroid belt. Goals include the determination of population statistics for the planet-crossing objects, the characterization of spacecraft targets to assist in encounter planning and subsequent interpretation of the data, a comparison of the collisional evolution of dynamically isolated Hilda and Trojan populations with the main belt, and the determination of the mechanism driving the activity of the distant object 2060 Chiron.

Mission Steering Profiles of Outer Planetary Orbiters Using Radioisotope Electric Propulsion

NASA Technical Reports Server (NTRS)

Fiehler, Douglas; Oleson, Steven

2004-01-01

Radioisotope Electric Propulsion (REP) has the potential to enable small spacecraft to orbit outer planetary targets with trip times comparable to flyby missions. The ability to transition from a flyby to an orbiter mission lies in the availability of continuous low power electric propulsion along the entire trajectory. The electric propulsion system s role is to add and remove energy from the spacecraft s trajectory to bring it in and out of a heliocentric hyperbolic escape trajectory for the outermost target bodies. Energy is added and the trajectory is reshaped to rendezvous with the closer-in target bodies. Sample REP trajectories will be presented for missions ranging for distances from Jupiter orbit to the Pluto-Kuiper Belt.

Overview of innovative PMI research on NSTX-U and associated PMI facilities at PPPL

DOE PAGES

M. Ono; Jaworski, M.; Kaita, R.; ...

2013-05-01

Developing a reactor compatible divertor and managing the associated plasma material interaction (PMI) has been identified as a high priority research area for magnetic confinement fusion. Accordingly on NSTX-U, the PMI research has received a strong emphasis. Moreover, with ˜15 MW of auxiliary heating power, NSTX-U will be able to test the PMI physics with the peak divertor plasma facing component (PFC) heat loads of up to 40-60 MW/m 2.

Dust remobilization tests in DIII-D divertor

NASA Astrophysics Data System (ADS)

Bykov, I.; Rudakov, D.; Moyer, R.; Ratynskaia, S.; Tolias, P.; Deangeli, M.; McLean, A.; Bystrov, K.

2015-11-01

Accumulation of dust on hot surfaces is a safety concern for ITER operation. We studied the life cycle of pre-deposited dust under ITER-relevant conditions by exposing W samples with W, C and Al (surrogate for Be) dust at the outer strike point (OSP) in a few ELMy H-mode discharges using DiMES. The maxima in the dust ejection rate correspond to ELM crashes under both attached and detached OSP conditions, as confirmed by a fast camera monitoring DiMES. SEM mapping of dust before and after exposures shows that >95 % of C and <5 % of metal dust gets remobilized in a few shots. In discharges with detached OSP, remaining Al particles melt and fuse together, forming larger spherical grains. At elevated heat flux with attached OSP, they melt, destruct and fuse with W substrate, which is not thermally affected. In this mode W grains partly melt and adjacent particles can weld together, forming larger asymmetric agglomerates with increased adhesion to the surface. We show that these results are consistent with recent observations from Pilot-PSI. Work supported by the US DOE under DE-FC02-04ER54698, DE-FG02-07ER54917 and DE-AC52-07NA27344.

A Peptidomimetic Antibiotic Targets Outer Membrane Proteins and Disrupts Selectively the Outer Membrane in Escherichia coli*

PubMed Central

Urfer, Matthias; Bogdanovic, Jasmina; Lo Monte, Fabio; Moehle, Kerstin; Zerbe, Katja; Omasits, Ulrich; Ahrens, Christian H.; Pessi, Gabriella; Eberl, Leo; Robinson, John A.

2016-01-01

Increasing antibacterial resistance presents a major challenge in antibiotic discovery. One attractive target in Gram-negative bacteria is the unique asymmetric outer membrane (OM), which acts as a permeability barrier that protects the cell from external stresses, such as the presence of antibiotics. We describe a novel β-hairpin macrocyclic peptide JB-95 with potent antimicrobial activity against Escherichia coli. This peptide exhibits no cellular lytic activity, but electron microscopy and fluorescence studies reveal an ability to selectively disrupt the OM but not the inner membrane of E. coli. The selective targeting of the OM probably occurs through interactions of JB-95 with selected β-barrel OM proteins, including BamA and LptD as shown by photolabeling experiments. Membrane proteomic studies reveal rapid depletion of many β-barrel OM proteins from JB-95-treated E. coli, consistent with induction of a membrane stress response and/or direct inhibition of the Bam folding machine. The results suggest that lethal disruption of the OM by JB-95 occurs through a novel mechanism of action at key interaction sites within clusters of β-barrel proteins in the OM. These findings open new avenues for developing antibiotics that specifically target β-barrel proteins and the integrity of the Gram-negative OM. PMID:26627837

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

NASA Astrophysics Data System (ADS)

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

2017-01-01

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

Characterized the pattern of the material deposition in the HL-2A tokamak

NASA Astrophysics Data System (ADS)

Cai, Laizhong; Wang, Jianbao; Wu, Ting; Zeng, Xiaoxiao; Hai, Ran; Ding, Hongbin

2017-03-01

Since the divertor geometry of a tokamak has a strong impact on the material erosion and deposition on the wall and HL-2A has a unique divertor configuration, it is necessary to investigate the material deposition pattern in HL-2A although a few results on other tokamaks have already been published. In this paper, tiles retrieved from the vessel are analyzed ex-situ by SIMS, SEM and laser-induced breakdown spectroscopy (LIBS). And deposition behind the lower divertor is in-situ measured by a quartz crystal microbalance (QMB). The deposition in HL-2A displays a complex pattern and clear localization characteristic. The thickness of the deposition layer varies in the range of 0-4μm. And in-situ diagnostic of QMB indicates that the average thickness of the deposition layer per pulse is over ten nanometers. In addition, the results imply that Si, Fe and D have different behaviors during the material deposition in HL-2A.

Powder Injection Molding for mass production of He-cooled divertor parts

NASA Astrophysics Data System (ADS)

Antusch, S.; Norajitra, P.; Piotter, V.; Ritzhaupt-Kleissl, H.-J.

2011-10-01

A He-cooled divertor for future fusion power plants has been developed at KIT. Tungsten and tungsten alloys are presently considered the most promising materials for functional and structural divertor components. The advantages of tungsten materials lie, e.g. in the high melting point, and low activation, the disadvantages are high hardness and brittleness. The machinig of tungsten, e.g. milling, is very complex and cost-intensive. Powder Injection Molding (PIM) is a method for cost effective mass production of near-net-shape parts with high precision. The complete W-PIM process route is outlined and, results of product examination discussed. A binary tungsten powder feedstock with a grain size distribution in the range 0.7-1.7 μm FSSS, and a solid load of 50 vol.% was developed. After heat treatment, the successfully finished samples showed promising results, i.e. 97.6% theoretical density, a grain size of approximately 5 μm, and a hardness of 457 HV0.1.

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

NASA Astrophysics Data System (ADS)

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

2016-02-01

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

Characterizing low-Z erosion and deposition in the DIII-D divertor using aluminum

DOE PAGES

Chrobak, Chris P.; Doerner, R. P.; Stangeby, Peter C.; ...

2017-01-28

Here, we present measurements and modeling of aluminum erosion and redeposition experiments in separate helium and deuterium low power, low density L-mode plasmas at the outer divertor strike point of DIII-D to provide a low-Z material benchmark dataset for tokamak erosion-deposition modeling codes. Coatings of Al ~100nm thick were applied to ideal (smooth) and realistic (rough) surfaces and exposed to repeat plasma discharges using the DiMES probe. Redeposition and re-erosion in all cases was primarily in the downstream toroidal field direction, evident from both in-situ spectroscopic and post-mortem non spectroscopic measurements. The gross Al erosion yield estimated from both Hemore » and D plasma exposures was ~40-70% of the expected erosion yield based on theoretical physical sputtering yields. However, the multi-step redeposition and re-erosion process, and hence the measured net erosion yield and material migration, was found to be influenced by the surface roughness and/or porosity. On rough surfaces, the fraction of the eroded Al coating found redeposited outside the original coating area was 25x higher than on smooth surfaces. The amount of Al found redeposited on the rough substrate was in fact proportional to the net eroded Al, suggesting an accumulation of deposited Al in surface pores and other areas shadowed from re-erosion. In order to determine the fraction and distribution of eroded Al returning to the surface, a simple model for erosion and redeposition was developed and fitted to the measurements. The model presented here reproduces many of the observed results in these experiments by using theoretically calculated sputtering yields, calculating surface composition changes and erosion rates in time, assuming a spatial distribution function for redepositing atoms, and accounting for deposit trapping in pores. The results of the model fits reveal that total redeposition fraction increases with higher plasma temperature (~30% for 15-18eV plasmas, and ~45% for 25-30eV plasmas), and that 50% of the atoms redepositing on rough surfaces accumulated in shadowed areas.« less

Identifying the location of the OMP separatrix in DIII-D using power accounting

DOE PAGES

Stangeby, Peter C.; Canik, John M.; Elder, J. D.; ...

2015-08-07

In order to identify reliable scalings for the scrape-off layer (SOL) power width it is necessary to know the location of the separatrix in divertor tokamaks as accurately as possible, specifically its location at the outside midplane (OMP) the standard reference location. Two methods are described which use power accounting to improve the accuracy of identifying the location of the OMP separatrix. The first uses the infrared-measured deposited power profile at the outer target as the primary input, the 'more » $$P_{{\\rm SOL}}^{{\\rm exhaust}}$$ method'. The second uses the measured power input to the SOL, obtained by subtracting the power radiated from inside the separatrix from the total heating power, the ' $$P_{{\\rm SOL}}^{{\\rm input}}$$ method'. Furthermore, these two power accounting methods are illustrated with the examples of 21 H-mode DIII-D discharges. High spatial resolution Thomson scattering measured profiles of ne and Te for the main SOL near the OMP are also used as primary input to the analysis; only between-edge localized mode data are used here. The Thomson profiles are used to calculate the electron parallel conducted heat flux profiles which are then matched to the measured $$P_{{\\rm SOL}}^{{\\rm exhaust}}$$ and $$P_{{\\rm SOL}}^{{\\rm input}}$$ by adjusting the location of the OMP separatrix relative to that of the Thomson data. For these attached discharges, it is found that the values of $$R_{{\\rm sep}}^{{\\rm omp}}$$ given by the two power accounting methods agree to within ~1 mm of each other and also to within ~1 mm of the values given by the 'standard DIII-D method' described by Porter et al (1998 Phys. Plasmas 5 1410). Lastly, the shifted $$R_{{\\rm sep}}^{{\\rm omp}}$$ results in only modest changes to the values of ne and Te at the OMP separatrix relative to the 'standard' values, increasing $$n_{{\\rm e}}^{{\\rm sep}}$$ by 8% and $$T_{{\\rm e}}^{{\\rm sep}}$$ by 20%.« less

Time-to-burnout data for a prototypical ITER divertor tube during a simulated loss of flow accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

The Loss of Flow Accident (LOFA) is a serious safety concern for the International Thermonuclear Experimental Reactor (ITER) as it has been suggested that greater than 100 seconds are necessary to safely shutdown the plasma when ITER is operating at full power. In this experiment, the thermal response of a prototypical ITER divertor tube during a simulated LOFA was studied. The divertor tube was fabricated from oxygen-free high-conductivity copper to have a square geometry with a circular coolant channel. The coolant channel inner diameter was 0.77 cm, the heated length was 4.0 cm, and the heated width was 1.6 cm.more » The mockup did not feature any flow enhancement techniques, i.e., swirl tape, helical coils, or internal fins. One-sided surface heating of the mockup was accomplished through the use of the 30 kW Sandia Electron Beam Test System. After reaching steady state temperatures in the mockup, as determined by two Type-K thermocouples installed 0.5 mm beneath the heated surface, the coolant pump was manually tripped off and the coolant flow allowed to naturally coast down. Electron beam heating continued after the pump trip until the divertor tube`s heated surface exhibited the high temperature transient normally indicative of rapidly approaching burnout. Experimental data showed that time-to-burnout increases proportionally with increasing inlet velocity and decreases proportionally with increasing incident heat flux.« less

The lithium vapor box divertor

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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 almore » 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.« less

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.

Non-solenoidal Startup with High-Field-Side Local Helicity Injection on the Pegasus ST

NASA Astrophysics Data System (ADS)

Perry, J. M.; Bodner, G. M.; Bongard, M. W.; Burke, M. G.; Fonck, R. J.; Pachicano, J. L.; Pierren, C.; Richner, N. J.; Rodriguez Sanchez, C.; Schlossberg, D. J.; Reusch, J. A.; Weberski, J. D.

2017-10-01

Local Helicity Injection (LHI) is a non-solenoidal startup technique utilizing electron current injectors at the plasma edge to initiate a tokamak-like plasma at high Ip . Recent experiments on Pegasus explore the inherent tradeoffs between high-field-side (HFS) injection in the lower divertor region and low-field-side (LFS) injection at the outboard midplane. Trade-offs include the relative current drive contributions of HI and poloidal induction, and the magnetic geometry required for relaxation to a tokamak-like state. HFS injection using a set of two increased-area injectors (Ainj = 4 cm2, Vinj 1.5 kV, and Iinj 8 kA) in the lower divertor is demonstrated over the full range of toroidal field available on Pegasus (BT 0 <= 0.15 T). Increased PMI on both the injectors and the lower divertor plates was observed during HFS injection, and was substantively mitigated through optimization of injector geometry and placement of local limiters to reduce scrape-off density in the divertor region. Ip up to 200 kA is achieved with LHI as the dominant current drive, consistent with expectations from helicity balance. To date, experiments support Ip increasing linearly with helicity injection rate. The high normalized current (IN >= 10) attainable with LHI and the favorable stability of the ultra-low aspect ratio, low-li LHI-driven plasmas allow access to high βt-up to 100 % , as indicated by kinetically-constrained equilibrium reconstructions. Work supported by US DOE Grant DE-FG02-96ER54375.

TEMPEST Simulations of the Plasma Transport in a Single-Null Tokamak Geometry

DOE PAGES

X. Q. Xu; Bodi, K.; Cohen, R. H.; ...

2010-05-28

We present edge kinetic ion transport simulations of tokamak plasmas in magnetic divertor geometry using the fully nonlinear (full-f) continuum code TEMPEST. Besides neoclassical transport, a term for divergence of anomalous kinetic radial flux is added to mock up the effect of turbulent transport. In order to study the relative roles of neoclassical and anomalous transport, TEMPEST simulations were carried out for plasma transport and flow dynamics in a single-null tokamak geometry, including the pedestal region that extends across the separatrix into the scrape-off layer and private flux region. In a series of TEMPEST simulations were conducted to investigate themore » transition of midplane pedestal heat flux and flow from the neoclassical to the turbulent limit and the transition of divertor heat flux and flow from the kinetic to the fluid regime via an anomalous transport scan and a density scan. The TEMPEST simulation results demonstrate that turbulent transport (as modelled by large diffusion) plays a similar role to collisional decorrelation of particle orbits and that the large turbulent transport (large diffusion) leads to an apparent Maxwellianization of the particle distribution. Moreover, we show the transition of parallel heat flux and flow at the entrance to the divertor plates from the fluid to the kinetic regime. For an absorbing divertor plate boundary condition, a non-half-Maxwellian is found due to the balance between upstream radial anomalous transport and energetic ion endloss.« less

TEMPEST Simulations of the Plasma Transport in a Single-Null Tokamak Geometry

DOE Office of Scientific and Technical Information (OSTI.GOV)

X. Q. Xu; Bodi, K.; Cohen, R. H.

We present edge kinetic ion transport simulations of tokamak plasmas in magnetic divertor geometry using the fully nonlinear (full-f) continuum code TEMPEST. Besides neoclassical transport, a term for divergence of anomalous kinetic radial flux is added to mock up the effect of turbulent transport. In order to study the relative roles of neoclassical and anomalous transport, TEMPEST simulations were carried out for plasma transport and flow dynamics in a single-null tokamak geometry, including the pedestal region that extends across the separatrix into the scrape-off layer and private flux region. In a series of TEMPEST simulations were conducted to investigate themore » transition of midplane pedestal heat flux and flow from the neoclassical to the turbulent limit and the transition of divertor heat flux and flow from the kinetic to the fluid regime via an anomalous transport scan and a density scan. The TEMPEST simulation results demonstrate that turbulent transport (as modelled by large diffusion) plays a similar role to collisional decorrelation of particle orbits and that the large turbulent transport (large diffusion) leads to an apparent Maxwellianization of the particle distribution. Moreover, we show the transition of parallel heat flux and flow at the entrance to the divertor plates from the fluid to the kinetic regime. For an absorbing divertor plate boundary condition, a non-half-Maxwellian is found due to the balance between upstream radial anomalous transport and energetic ion endloss.« less

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 almore » 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.« less

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

NASA Astrophysics Data System (ADS)

Smirnov, Roman; Krasheninnikov, Sergei

2017-10-01

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

APT Blanket Thermal Analyses of Top Horizontal Row 1 Modules

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shadday, M.A.

1999-09-20

The Accelerator Production of Tritium (APT) cavity flood system (CFS) is designed to be the primary safeguard for the integrity of the blanket modules during loss of coolant accidents (LOCAs). For certain large break LOCAs the CFS also provides backup for the residual heat removal systems (RHRs) in cooling the target assemblies. In the unlikely event that the internal flow passages in a blanket module or target assembly dryout, decay heat in the metal structures will be dissipated to the CFS through the module or assembly walls (i.e., rung outer walls). The target assemblies consist of tungsten targets encased inmore » steel conduits, and they can safely sustain high metal temperatures. Under internally dry conditions, the cavity flood fluid will cool the target assemblies with vigorous nucleate boiling on the external surfaces. However, the metal structures in the blanket modules consist of lead cladded in aluminum, and they have a long-term exposure temperature limit currently set to 150 degrees C. Simultaneous LOCAs in both the target and blanket heat removal systems (HRS) could result in dryout of the target ladders, as well as the horizontal blanket modules above the target. The cavity flood coolant would boil on the outside surfaces of the target ladder rungs, and the resultant steam could reduce the effectiveness of convection heat transfer from the blanket modules to the cavity flood coolant. A two-part analysis was conducted to ascertain if the cavity flood system can adequately cool the blanket modules above the targets, even when boiling is occurring on the outer surfaces of the target ladder rungs. The first part of the analysis was to model transient thermal conduction in the front top horizontal row 1 module (i.e. top horizontal modules nearest the incoming beam), while varying parametrically the convection heat transfer coefficient (htc) for the external surfaces exposed to the cavity flood flow. This part of the analysis demonstrated that the module could adequately conduct heat to the outer module surfaces, given reasonable values for the convection heat transfer coefficients. The second part of the analysis consisted of two-phase flow modeling of the natural circulation of the cavity flood fluid past the top modules. Slots in the top shield allow the cavity flood fluid to circulate. The required width for these slots, to prevent steam from backing up and blanketing the outer surfaces of the top modules, was determined.« less

Preparation of erosion and deposition investigations on plasma facing components in Wendelstein 7-X

NASA Astrophysics Data System (ADS)

Dhard, C. P.; Balden, M.; Braeuer, T.; Brezinsek, S.; Coenen, J. W.; Dudek, A.; Ehrke, G.; Hathiramani, D.; Klose, S.; König, R.; Laux, M.; Linsmeier, Ch; Manhard, A.; Masuzaki, S.; Mayer, M.; Motojima, G.; Naujoks, D.; Neu, R.; Neubauer, O.; Rack, M.; Ruset, C.; Schwarz-Selinger, T.; Pedersen, T. Sunn; Tokitani, M.; Unterberg, B.; Yajima, M.; W7-X Team1, The

2017-12-01

In the Wendelstein 7-X stellarator with its twisted magnetic geometry the investigation of plasma wall interaction processes in 3D plasma configurations is an important research subject. For the upcoming operation phase i.e. OP1.2, three different types of material probes have been installed within the plasma vessel for the erosion/deposition investigations in selected areas with largely different expected heat load levels, namely, ≤10 MW m-2 at the test divertor units (TDU), ≤500 kW m-2 at the baffles, heat shields and toroidal closures and ≤100 kW m-2 at the stainless steel wall panels. These include 18 exchangeable target elements at TDU, about 30 000 screw heads at graphite tiles and 44 wafer probes on wall panels, coated with marker layers. The layer thicknesses, surface morphologies and the impurity contents were pre-characterized by different techniques and subjected to various qualification tests. The positions of these probes were fixed based on the strike line locations on the divertor predicted by field line diffusion and EMC3/EIRENE modeling calculations for the OP1.2 plasma configurations and availability of locations on panels in direct view of the plasma. After the first half of the operation phase i.e. OP1.2a the probes will be removed to determine the erosion/deposition pattern by post-mortem analysis and replaced by a new set for the second half of the operation phase, OP1.2b.

Facilitative glucose transporter Glut1 is actively excluded from rod outer segments.

PubMed

Gospe, Sidney M; Baker, Sheila A; Arshavsky, Vadim Y

2010-11-01

Photoreceptors are among the most metabolically active cells in the body, relying on both oxidative phosphorylation and glycolysis to satisfy their high energy needs. Local glycolysis is thought to be particularly crucial in supporting the function of the photoreceptor's light-sensitive outer segment compartment, which is devoid of mitochondria. Accordingly, it has been commonly accepted that the facilitative glucose transporter Glut1 responsible for glucose entry into photoreceptors is localized in part to the outer segment plasma membrane. However, we now demonstrate that Glut1 is entirely absent from the rod outer segment and is actively excluded from this compartment by targeting information present in its cytosolic C-terminal tail. Our data indicate that glucose metabolized in the outer segment must first enter through other parts of the photoreceptor cell. Consequently, the entire energy supply of the outer segment is dependent on diffusion of energy-rich substrates through the thin connecting cilium that links this compartment to the rest of the cell.

Method for sputtering with low frequency alternating current

DOEpatents

Timberlake, John R.

1996-01-01

Low frequency alternating current sputtering is provided by connecting a low frequency alternating current source to a high voltage transformer having outer taps and a center tap for stepping up the voltage of the alternating current. The center tap of the transformer is connected to a vacuum vessel containing argon or helium gas. Target electrodes, in close proximity to each other, and containing material with which the substrates will be coated, are connected to the outer taps of the transformer. With an applied potential, the gas will ionize and sputtering from the target electrodes onto the substrate will then result. The target electrodes can be copper or boron, and the substrate can be stainless steel, aluminum, or titanium. Copper coatings produced are used in place of nickel and/or copper striking.

Method for sputtering with low frequency alternating current

DOEpatents

Timberlake, J.R.

1996-04-30

Low frequency alternating current sputtering is provided by connecting a low frequency alternating current source to a high voltage transformer having outer taps and a center tap for stepping up the voltage of the alternating current. The center tap of the transformer is connected to a vacuum vessel containing argon or helium gas. Target electrodes, in close proximity to each other, and containing material with which the substrates will be coated, are connected to the outer taps of the transformer. With an applied potential, the gas will ionize and sputtering from the target electrodes onto the substrate will then result. The target electrodes can be copper or boron, and the substrate can be stainless steel, aluminum, or titanium. Copper coatings produced are used in place of nickel and/or copper striking. 6 figs.

NIF Double Shell outer/inner shell collision experiments

NASA Astrophysics Data System (ADS)

Merritt, E. C.; Loomis, E. N.; Wilson, D. C.; Cardenas, T.; Montgomery, D. S.; Daughton, W. S.; Dodd, E. S.; Desjardins, T.; Renner, D. B.; Palaniyappan, S.; Batha, S. H.; Khan, S. F.; Smalyuk, V.; Ping, Y.; Amendt, P.; Schoff, M.; Hoppe, M.

2017-10-01

Double shell capsules are a potential low convergence path to substantial alpha-heating and ignition on NIF, since they are predicted to ignite and burn at relatively low temperatures via volume ignition. Current LANL NIF double shell designs consist of a low-Z ablator, low-density foam cushion, and high-Z inner shell with liquid DT fill. Central to the Double Shell concept is kinetic energy transfer from the outer to inner shell via collision. The collision determines maximum energy available for compression and implosion shape of the fuel. We present results of a NIF shape-transfer study: two experiments comparing shape and trajectory of the outer and inner shells at post-collision times. An outer-shell-only target shot measured the no-impact shell conditions, while an `imaging' double shell shot measured shell conditions with impact. The `imaging' target uses a low-Z inner shell and is designed to perform in similar collision physics space to a high-Z double shell but can be radiographed at 16keV, near the viable 2DConA BL energy limit. Work conducted under the auspices of the U.S. DOE by LANL under contract DE-AC52-06NA25396.

The mitochondrial outer membrane protein MDI promotes local protein synthesis and mtDNA replication.

PubMed

Zhang, Yi; Chen, Yong; Gucek, Marjan; Xu, Hong

2016-05-17

Early embryonic development features rapid nuclear DNA replication cycles, but lacks mtDNA replication. To meet the high-energy demands of embryogenesis, mature oocytes are furnished with vast amounts of mitochondria and mtDNA However, the cellular machinery driving massive mtDNA replication in ovaries remains unknown. Here, we describe a Drosophila AKAP protein, MDI that recruits a translation stimulator, La-related protein (Larp), to the mitochondrial outer membrane in ovaries. The MDI-Larp complex promotes the synthesis of a subset of nuclear-encoded mitochondrial proteins by cytosolic ribosomes on the mitochondrial surface. MDI-Larp's targets include mtDNA replication factors, mitochondrial ribosomal proteins, and electron-transport chain subunits. Lack of MDI abolishes mtDNA replication in ovaries, which leads to mtDNA deficiency in mature eggs. Targeting Larp to the mitochondrial outer membrane independently of MDI restores local protein synthesis and rescues the phenotypes of mdi mutant flies. Our work suggests that a selective translational boost by the MDI-Larp complex on the outer mitochondrial membrane might be essential for mtDNA replication and mitochondrial biogenesis during oogenesis. Published 2016. This article is a U.S. Government work and is in the public domain in the USA.

Halo current diagnostic system of experimental advanced superconducting tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen, D. L.; Shen, B.; Sun, Y.

2015-10-15

The design, calibration, and installation of disruption halo current sensors for the Experimental Advanced Superconducting Tokamak are described in this article. All the sensors are Rogowski coils that surround conducting structures, and all the signals are analog integrated. Coils with two different cross-section sizes have been fabricated, and their mutual inductances are calibrated. Sensors have been installed to measure halo currents in several different parts of both the upper divertor (tungsten) and lower divertor (graphite) at several toroidal locations. Initial measurements from disruptions show that the halo current diagnostics are working well.

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

PubMed

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).

Method for fabricating .sup.99 Mo production targets using low enriched uranium, .sup.99 Mo production targets comprising low enriched uranium

DOEpatents

Wiencek, Thomas C.; Matos, James E.; Hofman, Gerard L.

1997-01-01

A radioisotope production target and a method for fabricating a radioisotope production target is provided, wherein the target comprises an inner cylinder, a foil of fissionable material circumferentially contacting the outer surface of the inner cylinder, and an outer hollow cylinder adapted to receive the substantially foil-covered inner cylinder and compress tightly against the foil to provide good mechanical contact therewith. The method for fabricating a primary target for the production of fission products comprises preparing a first substrate to receive a foil of fissionable material so as to allow for later removal of the foil from the first substrate, preparing a second substrate to receive the foil so as to allow for later removal of the foil from the second substrate; attaching the first substrate to the second substrate such that the foil is sandwiched between the first substrate and second substrate to prevent foil exposure to ambient atmosphere, and compressing the exposed surfaces of the first and second substrate to assure snug mechanical contact between the foil, the first substrate and the second substrate.

Method for fabricating .sup.99 Mo production targets using low enriched uranium, .sup.99 Mo production targets comprising low enriched uranium

DOEpatents

Wiencek, Thomas C [Orland Park, IL; Matos, James E [Oak Park, IL; Hofman, Gerard L [Downers Grove, IL

2000-12-12

A radioisotope production target and a method for fabricating a radioisotope production target is provided, wherein the target comprises an inner cylinder, a foil of fissionable material circumferentially contacting the outer surface of the inner cylinder, and an outer hollow cylinder adapted to receive the substantially foil-covered inner cylinder and compress tightly against the foil to provide good mechanical contact therewith. The method for fabricating a primary target for the production of fission products comprises preparing a first substrate to receive a foil of fissionable material so as to allow for later removal of the foil from the first substrate, preparing a second substrate to receive the foil so as to allow for later removal of the foil from the second substrate; attaching the first substrate to the second substrate such that the foil is sandwiched between the first substrate and second substrate to prevent foil exposure to ambient atmosphere, and compressing the exposed surfaces of the first and second substrate to assure snug mechanical contact between the foil, the first substrate and the second substrate.