USBPO Mission Statement:

Advance the scientific understanding of burning plasmas and ensure the greatest benefit from a burning plasma experiment by coordinating relevant U.S. fusion research with broad community participation.

Contact and Contribution Information

This newsletter provides a monthly update on U.S. Burning Plasma Organization activities. Topical Group Highlight articles are selected by the Leader and Deputy Leader of those groups (http://burningplasma.org/groups.html). ITPA Reports are solicited by the Editor based on recently held meetings. Announcements, Upcoming Burning Plasma Events, and all comments may be sent to the Editor. Suggestions for the Image of the Month may be sent to the Editor. The images should be photos, as opposed to data plots, though combined graphics are welcome. The goal is to highlight U.S. fusion resources through interesting visualizations.

Become a member of the U.S. Burning Plasma Organization by signing up for a topical group:
burningplasma.org/jointopical

Editor: David Pace (pacedc@fusion.gat.com)
Assistant Editor: Amadeo Gonzales (agonzales@austin.utexas.edu)

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USBPO Topical Group Highlights

[The BPO Plasma-Wave Interactions Topical Group works to facilitate U.S. efforts to understand radio frequency plasma heating and current drive in existing and future magnetic fusion devices via experiments and simulations (leaders are Gary Taylor and David Green). This month's Research Highlight by R.J. Perkins, et al., describes experiment and modeling results from NSTX that quantify the deposition of high-harmonic fast-wave energy into the scrape-off layer and plasma facing components. Understanding this loss mechanism is necessary to optimize fast-wave heating in the ITER regime. The NSTX Upgrade (NSTX-U) device mentioned in this Research Highlight is discussed in more detail as our Image of the Month. -Ed.]

Towards Efficient Fast-wave Heating in NSTX through Understanding Direct Scrape-off Layer Losses of High-harmonic FW Power
R.J. Perkins^1*, D.L. Green², J.C. Hosea¹, G.J. Kramer¹, G. Taylor¹, and the NSTX team

¹ Princeton Plasma Physics Laboratory, Princeton, NJ
² Oak Ridge National Laboratory, Oak Ridge, TN
* E-mail: rperkins@pppl.gov

20 MW of fast-wave heating are planned for the International Thermonuclear Experimental Reactor (ITER) [1], but the amount of power that actually couples to the core plasma depends, among other issues, on the scrape-off layer (SOL) conditions. Indeed, high-harmonic fast-wave (HHFW) heating experiments on the National Spherical Torus eXperiment (NSTX) show that a significant loss of power can occur directly in the scrape off layer (SOL) [2-5], leading to bright streaks emanating from the antenna region and terminating on the upper and lower divertor in bright and hot spirals (Fig. 1) [3,4]. In some cases, only approximately 40% of the RF power coupled from the antenna is observed in the core plasma [2]; meanwhile, an infrared (IR) camera measures a significant RF-produced heat flux to the divertor region under the spirals, up to ~ 2 MW/m² for an RF coupled power of 1.8 MW [5]. Understanding these losses is important not only for efficient fast-wave heating on NSTX-U but also for verifying advanced RF codes that include the SOL region [6] that will be used to predict fast-wave heating performance in ITER. The remarkable properties of these losses described below offer insight into the underlying physical mechanisms and provide guidance for future experiments on NSTX-U.

Figure 1: Strong interactions of the HHFW power and the edge plasma are clearly seen in this midplane visible-light image of the spirals on the upper and lower divertor for an ELM-free H-mode plasma with PRF = 1.8 MW. The conditions for shot 130621 are: ƒANT = -90°, PNB = 2 MW, IP = 1 MA, Bƒ = 0.55 T.

The HHFW power lost in the SOL flows mainly along the magnetic field, as is evidenced by comparing field-line mapping to experimental measurements of RF-produced effects in the divertor regions [7]. The SPIRAL code [8] is used to track field lines from the midplane in front of the antenna to where they strike the divertor region. Figure 2 plots these strike points on the lower divertor floor for a dense set of field lines that originate at many different major radii in the SOL (R_SOL) at the midplane; the color-coding denotes the R_SOL of the field line from which each strike point originated. The set of strike points forms a spiral as is observed experimentally, and the strike-point positions can be compared with measurements of RF-induced effects such as camera images, Langmuir probes, and current-sensing tiles [7]. One such comparison is shown in Fig. 3, where data for the heat flux incident on the lower divertor is plotted along with the computed strike points for four shots in a magnetic pitch scan (pitch values quoted at antenna). The IR cameras are located at Bay I [10] (the sight line for the IR data is indicated in Fig 2). The first and second passes of the spiral are apparent for all shots except 141888, in which only the second pass is visible. The second pass of the spiral moves inward radially with increasing magnetic pitch, and the strike points track this movement over the gap in the vessel floor. For the low pitch case, the spiral does not reach Bay I on the first pass as indicated by the absence of strike points at large radii, and there is correspondingly no RF-produced heat flux in the IR data. These results demonstrate that much of the RF losses to the divertor region follow SOL field lines.

Figure 2: Computed field-line strike points on the lower divertor floor using EFIT02 [9] for different magnetic field pitch (shots 141888 and 141899 at t = 0.355 s). As the magnetic pitch increases, the spiral rotates toroidally across the IR sight line for Fig. 3. The HHFW antenna is centered at Bay D and spans 90o. [kANT= -8m-1 (-90°), D2, PRF = 1.4 MW starting at t = 0.25 s, PNB = 2 MW]	Figure 3: IR data from Bay I shows the first and second passes of the spiral; the radii of the heat peaks generally agrees with the field-line mapping at Bay I. The IR data is taken at the times shown, and EFIT02 equilibrium fits are taken at 355 ms.

Another remarkable property of the edge losses studied here is that they are spread across the width of the SOL as opposed to being localized near the antenna. As shown in Fig. 4, the lower-divertor heat flux, when mapped back to the midplane in front of the antenna using the field-line mapping, shows the lost-power flux is relatively large both close to the antenna and also again near the LCFS (details can be found in Ref. 11). The mapping procedure in Fig. 4 incorporates flux expansion and projection effects. This demonstrates that the underlying loss mechanisms are not localized to the antenna components.

Figure 4: The black curve is the RF-produced heat flux to the lower divertor at Bay I; the red points are the corresponding values of the lost RF power flux at the midplane (shot 135333). These points correspond to the different passes of the spiral across Bay I; a complete reconstruction of the midplane profile would require IR coverage of the entire lower divertor.

We hypothesize that theses losses are due to fast-wave propagation in the SOL, as evidenced by the relationship between heating efficiency, edge density, and the onset density for perpendicular fast-wave propagation [2,3], and by the resemblance of the loss profiles obtained in Fig. 4 to a radial standing-wave pattern in a cavity. This hypothesis will be tested by direct measurements of RF fields in the SOL during future fast-wave heating experiments on NSTX-U. While the losses to the divertor region along field lines discussed here are observed in a spherical torus with relatively high magnetic pitch, field-aligned RF-effects in the SOL have also been observed under conventional-tokamak conditions on Alcator C-Mod [12] and indicate these effects are common to fast-wave heating systems. The results obtained on NSTX provide a clear test for verifying advanced RF codes that treat the SOL region and that in turn can be used to theoretically understand the process(s) underlying the SOL power flow. Such a verification will also validate the use of advanced RF codes to predict and possibly minimize SOL power losses to the divertor region on ITER.

* This work is supported by USDOE Contract No. DE-AC02-09CH11466

References

1. Swain D W and Goulding R H 2007 Fusion Eng. Des. 82 603
2. Hosea J C et al. 2008 Phys. Plasmas 15 056104
3. Phillips C K et al. 2009 Nucl. Fusion 49 075015
4. Hosea J C et al. 2009 AIP Conf. Proceedings 1187 105
5. Taylor G et al. 2010 Phys. Plasmas 17 056114
6. Green D L et al. 2011 Phys. Rev. Lett. 107 145001
7. Perkins R J et al. 2012 Phys. Rev. Lett. 109 045001
8. Kramer G et al. 2013 Plasma Phys. Contr. F. 55 025013
9. Sabbagh S et al. 2001 Nucl. Fusion 41 1601
10. Mastrovito D et al. 2003 Rev. Sci. Instrum. 74 5090
11. Perkins R J et al. 2013 Nucl. Fusion submitted
12. Wukitch S et al. 2007 AIP Conf. Proc. 933 75

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ITPA Update

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Schedule of Burning Plasma Events

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Image of the Month

A Powerful Upgrade

The National Spherical Torus eXperiment (NSTX) is undergoing a major upgrade that will provide access to new parameter regimes (e.g., from low to high beta at low collisionality) that are important for both burning plasma physics, and for the design of a US Fusion Nuclear Science Facility (FNSF). Compared to NSTX, the NSTX Upgrade (NSTX-U: R₀ = 0.94 m, a = 0.62 m,) device will access up to a factor of two higher plasma current (2 MA), toroidal field (1 T) and NBI heating power (15 MW) with total auxiliary heating power up to 19 MW. A major goal of NSTX-U is to access 100% non-inductive operation at plasma performance applicable to a future FNSF. Much of this improved performance will result from the fabrication of a more robust center-stack. The photos on the left side of this month’s image depict the new technologies associated with the toroidal field (TF) bundle. The upper left photo shows a single inner TF conductor being insulated, while the lower left photo shows a recently completed inner TF coil quadrant. The quadrants are being fabricated using a vacuum pressure impregnation technique with the same cyanate-ester epoxy (CTD-425) that is planned for use in the ITER TF coils. Another major element of NSTX-U is the installation of a second neutral beam injector (NBI) that was previously used on TFTR. A worker provides scale for the photo in the bottom right, in which the second NBI housing is seen as it is lowered into position adjacent to the tokamak. This new beamline takes advantage of a major retrofit to the vacuum vessel that provides an increased beam tangency radius of up to R_TAN = 1.3 m for off-axis current drive and increased current drive efficiency. Finally, the present configuration of the NSTX-U test cell is shown in the upper right of the image. The NSTX-U project is currently on schedule with first plasma expected in 2014 and a full experimental campaign in 2015.

Contributed by J. Menard and M. Ono
Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543-0451, USA
NSTX-U: http://nstx-u.pppl.gov/
Overview Paper: J.E. Menard, et al., Nucl. Fusion 52, 083015 (2012)

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