Divertor Heat Flux Mitigation in High-Performance H-mode Plasmas in the National Spherical Torus Experiment PDF Download

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Languages : en
Pages : 10

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Experiments conducted in high-performance 1.0-1.2 MA 6 MW NBI-heated H-mode plasmas with a high flux expansion radiative divertor in NSTX demonstrate that significant divertor peak heat flux reduction and access to detachment may be facilitated naturally in a highly-shaped spherical torus (ST) configuration. Improved plasma performance with high [beta]{sub p} = 15-25%, a high bootstrap current fraction f{sub BS} = 45-50%, longer plasma pulses, and an H-mode regime with smaller ELMs has been achieved in the lower single null configuration with higher-end elongation 2.2-2.4 and triangularity 0.6-0.8. Divertor peak heat fluxes were reduced from 6-12 MW/m2 to 0.5-2 MW/m2 in ELMy H-mode discharges using high magnetic flux expansion and partial detachment of the outer strike point at several D2 injection rates, while good core confinement and pedestal characteristics were maintained. The partially detached divertor regime was characterized by a 30-60% increase in divertor plasma radiation, a peak heat flux reduction by up to 70%, measured in a 10 cm radial zone, a five-fold increase in divertor neutral pressure, and a significant volume recombination rate increase.

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Languages : en
Pages : 23

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Experiments conducted in high-performance 1.0 MA and 1.2 MA 6 MW NBI-heated H-mode discharges with a high magnetic flux expansion radiative divertor in NSTX demonstrate that significant divertor peak heat flux reduction and access to detachment may be facilitated naturally in a highly-shaped spherical torus (ST) configuration. Improved plasma performance with high [beta]{sub t} = 15-25%, a high bootstrap current fraction f{sub BS} = 45-50%, longer plasma pulses, and an H-mode regime with smaller ELMs has been achieved in the strongly-shaped lower single null configuration with elongation [kappa] = 2.2-2.4 and triangularity [delta] = 0.6-0.8. Divertor peak heat fluxes were reduced from 6-12 MW/m2 to 0.5-2 MW/m2 in ELMy H-mode discharges using the inherently high magnetic flux expansion f{sub m} = 16-25 and the partial detachment of the outer strike point at several D2 injection rates. A good core confinement and pedestal characteristics were maintained, while the core carbon concentration and the associated Z{sub eff} were reduced. The partially detached divertor regime was characterized by an increase in divertor radiated power, a reduction of ion flux to the plate, and a large neutral compression ratio. Spectroscopic measurements indicated a formation of a high-density, low temperature region adjacent to the outer strike point, where substantial increases in the volume recombination rate and CII, CIII emission rates was measured.

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Languages : en
Pages : 6

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Steady-state handling of divertor heat flux is a critical issue for both the International Thermonuclear Experimental Reactor and spherical torus (ST) based devices with compact high power density divertors. The ST compact divertor with a small plasma volume, a small plasma-wetted area, and a short parallel connection length can reduce the operating space of heat flux dissipation techniques based on induced edge and/or scrape-off layer (SOL) power and momentum loss, such as the radiative and dissipative divertors and radiative mantles. Access to these regimes is studied in the National Spherical Torus Experiment (NSTX) with an open geometry horizontal carbon plate divertor in 2-6 MW NBI-heated H-mode plasmas in a lower single null (LSN) configuration in a range of elongations [kappa] = 1.8-2.4 and triangularities [delta]= 0.40-0.75. Experiments conducted in a lower end [kappa]H".8-2.0 and [delta]H"0.4-0.5 LSN shape using deuterium injection in the divertor region have achieved the outer strike point (OSP) peak heat flux reduction from 4-6 MW/m2 to a manageable level of 1-2 MW/m2. However, only the high-recycling radiative divertor (RD) regime was found to be compatible with good performance and H-mode confinement. A partially detached divertor (PDD) could only be obtained at a high D2 injection rate that led to an X-point MARFE formation and confinement degradation. Also in the low [kappa]H"2, [delta]H"0.45 shape, peak heat flux q{sub pk} and heat flux width [lambda]{sub q} scaling studies have been conducted. Similar to tokamak divertor studies, q{sub pk} was found to be a strong function of input power PNBI and plasma current Ip, and the heat flux midplane scale length [lambda]{sub q} was found to be large as compared with simple SOL models. In this paper, we report on the first experiments to assess steady-state divertor heat flux amelioration in highly shaped plasmas in NSTX.

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Languages : en
Pages : 41

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Steady-state handling of divertor heat flux is a critical issue for both ITER and spherical torus-based devices with compact high power density divertors. Significant reduction of heat flux to the divertor plate has been achieved simultaneously with favorable core and pedestal confinement and stability properties in a highly-shaped lower single null configuration in the National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40, 557 2000] using high magnetic flux expansion at the divertor strike point and the radiative divertor technique. A partial detachment of the outer strike point was achieved with divertor deuterium injection leading to peak flux reduction from 4-6 MW m−2 to 0.5-2 MW m−2 in small-ELM 0.8-1.0 MA, 4-6 MW neutral beam injection-heated H-mode discharges. A self-consistent picture of 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.

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Languages : en
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The National Spherical Torus Experiment [Ono et al., Nucl. Fusion, 44, 452 (2004)] is targeting long pulse high performance, noninductive sustained operations at low aspect ratio, and the demonstration of nonsolenoidal startup and current rampup. The modeling of these plasmas provides a framework for experimental planning and identifies the tools to access these regimes. Simulations based on neutral beam injection (NBI)-heated plasmas are made to understand the impact of various modifications and identify the requirements for (1) high elongation and triangularity, (2) density control to optimize the current drive, (3) plasma rotation and/or feedback stabilization to operate above the no-wall limit, and (4) electron Bernstein waves (EBW) for off-axis heating/current drive (H/CD). Integrated scenarios are constructed to provide the transport evolution and H/CD source modeling, supported by rf and stability analyses. Important factors include the energy confinement, Zeff, early heating/H mode, broadening of the NBI-driven current profile, and maintaining q(0) and qmin>1.0. Simulations show that noninductive sustained plasmas can be reached at IP=800 kA, BT=0.5 T, 2.5, N5, 15%, fNI=92%, and q(0)>1.0 with NBI H/CD, density control, and similar global energy confinement to experiments. The noninductive sustained high plasmas can be reached at IP=1.0 MA, BT=0.35 T, 2.5, N9, 43%, fNI=100%, and q(0)>1.5 with NBI H/CD and 3.0 MW of EBW H/CD, density control, and 25% higher global energy confinement than experiments. A scenario for nonsolenoidal plasma current rampup is developed using high harmonic fast wave H/CD in the early low IP and low Te phase, followed by NBI H/CD to continue the current ramp, reaching a maximum of 480 kA after 3.4 s.

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Languages : en
Pages : 39

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In this paper we present a summary of recent DIII-D divertor physics activity and plans for future divertor upgrades. During the past year, DIII-D experimental effort was focused on areas of active heat and particle control and divertor target erosion studies. Using the DIII-D Advanced Divertor system we have succeeded for the first time to control the plasma density and demonstrate helium exhaust in H-mode plasmas. Divertor heat flux control by means of D2 gas puffing and impurity injection were studied separately and in, both cases up to a factor of five reduction of the divertor peak heat flux was observed. Using the DiMES sample transfer system we have obtained erosion data on various material samples in well diagnosed plasmas and compared the results with predictions of numerical models.

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Languages : en
Pages : 18

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This paper presents an overview of recent results from divertor physics studies in DIII-D. Heat flux measurements at input power levels up to 20 MW show that steady divertor heat loads of up to 4 MW/m2 are obtained in H-mode discharges with ELMs. No carbon blooms are observed. The heat flux profile is highly peaked at the outside strike point in single-null discharges, and is up/down asymmetric in double-null discharges. Recent experiments with gas injection below the X-point have demonstrated a factor of two reduction in the peak divertor heat flux for H-mode plasmas at these power levels. These heat flux data, along with measurements of the n{sub e} and T{sub e} profiles at the divertor are being used to help interpret the first reported measurements of the erosion profile for a set of graphite divertor tiles exposed to several months of high power tokamak operation. We have now modified the divertor hardware in order to carry out experiments with divertor biasing, baffling, and pumping. 26 refs., 8 figs.

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Languages : en
Pages : 6

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Author: S. Allen
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Languages : en
Pages : 6

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

Author: G. Porter
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Languages : en
Pages : 17

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Identification of divertor operating regimes is of particular importance for heat and particle control optimization in high performance plasmas of a spherical torus, because of the magnetic geometry effects and compactness of the divertor region. Recent measurements of radiated power, heat and particle fluxes in lower single null and double null plasmas with 0.8 - 6 MW NBI heating suggest that the inner divertor is detached at {bar n} {sub e} {