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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.

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Director's Corner by C. M. Greenfield

Happy New Year

I wish you all a happy and successful New Year. 2012 was the embodiment of the old proverb/curse, “may you live in interesting times.” 2013 promises to continue to be interesting. Here’s hoping that’s a good thing.

Progress on ITER

2012 was an exciting year for ITER. The new Headquarters Building was completed, with ITER staff moving in during the last few months of the year, and the ITER Council inaugurating their new meeting room.

Elsewhere around the site, there are many signs of progress, including the electrical switchyard and the Poloidal Field Coil Assembly Building. In the coming months, the Tokamak Complex will begin construction, starting with the pouring of a basemat that will cover the seismic bearings that were completed during 2012. The Tokamak Complex, 80 meters high, 120 meters long and 80 meters wide, will be completed in time for assembly operations to begin in 2016.

The ITER site in December, 2012. The newly completed Headquarters Building is in the background. Photo © ITER Organization.

ITPA Coordinating Committee/CTP JEX/CTP ExComm

For the third time, the combined ITPA Coordinating Committee, IEA Implementing Agreement on Cooperation of Tokamak Programs (CTP) Joint Experiments, and CTP Executive Committee met together last month. It was the first time that this meeting was held in the new Council Room. During this meeting, we heard reports from the ITER Organization on progress on construction and a summary of research needs for ITER. We also heard reports on status and plans from each ITPA Topical Group, including a progress report on a cross-cutting task to support a decision on divertor materials expected later this year. In addition, overviews were given of the national programs of each ITER partner.

On the final day, the 3rd Executive Committee Meeting of the IEA Implementing Agreement for Cooperation on Tokamak Programmes (CTP) was held. During this meeting, the group reviewed and approved a long list of IAEA proposed joint experiments that will be taken up by many of the world’s fusion experiments.

Attendees at the ITPA Coordinating Committee and CTP Executive Committee meeting in the new ITER Council Room. Photo © ITER Organization.

What's coming up for the USBPO?

During the second half of 2012, many of us in the USBPO and in the community at large turned our attention to the FESAC Subpanel on MFE Priorities. Over the summer, the USBPO held two web seminars in support of the subpanel’s efforts, and provided web support for white papers and other material provided by the community. As the work of that subpanel nears completion, the USBPO will resume the periodic web seminars focused on ITPA activities and other topics (you can suggest topics). Also, we continue to have several task groups in different stages of organization, and we will work to increase their activities in 2013. In particular, the new task group on modes of participation in ITER (led by Rajesh Maingi and Mike Walker) will be ramping up its activities in the coming months.

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ITPA Report

Summary of the ITPA Coordinating Committee Meeting
S. Eckstrand
Office of Fusion Energy Science, U.S. Department of Energy

The 15th Meeting of the ITPA Coordinating Committee was held for the first time in the ITER Council Room in the new ITER Headquarters Building. Dr. Y. Kamada opened the meeting with a brief report on the overall status of the ITPA, including the organization of the working group to assess carbon and tungsten divertors, as per the request by STAC-13 in October 2012.

Director General O. Motojima gave an overview of the ITER Project, noting that the 24th IAEA Fusion Energy Conference marked the 5th anniversary of the establishment of the ITER Project. D. Campbell noted that the IO was planning an update of the ITER research plan in 2013. He then discussed the ITER physics R&D needs. The most urgent R&D needs are related to the design of the disruption mitigation system, the decision on starting up with a tungsten divertor target, and ELM control.

The seven topical groups all reported impressive progress during the past year. As in the past, there were a number of excellent papers presented at the IAEA meeting related to ITPA joint experiments. Overall, eleven joint experiments were completed in 2012, 50 joint experiments are continuing, and 6 new joint experiments were initiated. A few highlights from the topical group reports are summarized below.

Divertor and SOL

• Analysis of the database for midplane SOL heat flux channel width gives 1_q,ITER ~ 1 mm, but further experiments and analysis are needed to understand what determines this limit.
  
  
• Fuel retention with the ITER-like wall in JET is a factor of 10 lower than with the carbon wall.

Diagnostics

• With the ITER-like wall, carbon is not the main contributor in deposits on diagnostic mirrors; deposition with W PFCs as compared with earlier studies at the similar locations and elevated temperature.
  
  
• ECH stray/reflected power may pollute measurements and will be a major concern for potential diagnostic damage.

Edge Pedestal

• Initial experiments indicate L-H threshold is lower with tungsten than with carbon walls; however, pedestal confinement is lower with tungsten than with carbon. Confinement recovers with impurity seeding.
  
  
• Planning a strong effort on ELM physics and ELM control/mitigation in 2013.

Energetic Particles

• Comparison between modeling and experimental results indicates fast ion transport by small scale turbulence will be insignificant in ITER.
  
  
• Completed work on benchmarking of linear codes used to calculate Alfvén Eigenmode stability.

Integrated Operation Scenarios

• Recommend poloidal steering of the equatorial EC launchers in ITER to give greater control over driven current.
  
  
• Results from the ITPA database indicate the collisionality scaling may be very important for projection of present results to ITER; plan to initiate experiments on AUG, DIII-D, and JET.

MHD Stability

• Recent success in NTM stabilization with real-time EC mirror steering.
  
  
• JET disruptions with ILW show less radiated energy, slower current quench.
  
  
• Significant commitment needed from international facilities to determine critical R&D needs for ITER’s disruption mitigation system.

Transport and Confinement

• Various experiments show LOC/SOC, nonlocality, V_tor reversal and poloidal asymmetry all linked with coincident n_crit.
  
  
• Density pumpout, associated with externally applied magnetic perturbations is a well established phenomena; planning a dedicated study to understand impact of RMPs on transport and confinement.

There was a separate report on the ITPA evaluation of carbon and tungsten as divertor target materials.

• STAC-13 has requested evaluation of carbon and tungsten with an integrated list of:
  • Key physics and technological Items,
  • Criteria for each item needed for judgment,
  • Achievements and remaining issues for each item,
  • and comments (and recommendation) from ITPA to IO
    
    
• Report to be completed by May 2013 for STAC-14.

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USBPO Topical Group Highlights

(Editors note: The BPO MHD and Macroscopic Plasma Physics Topical Group covers research across a wide area of plasma stability and tokamak performance [leaders are François Waelbroeck and Bob Granetz]. This month's Research Highlight by S.P. Gerhardt, et al., describes the experimental mapping of disruption likelihood across the high-performance operating space of the National Spherical Torus Experiment. Understanding disruptions, and learning to minimize or avoid them, is of particular importance to burning plasma experiments in which the plasma energy transferred to machine components during a disruption has the potential to significantly damage the reactor. In addition, the US has the responsibility of supplying a disruption mitigation system for ITER. -Ed.)

Operating Regimes with Minimal Disruptivity at High-β in the National Spherical Torus Experiment
S.P. Gerhardt¹, J.W. Berkery², and S.A. Sabbagh²

¹ Princeton Plasma Physics Laboratory
² Columbia University, Department of Applied Physics and Mathematics

The fusion power generated by a spherical torus (ST) or tokamak scales as β_T²B_T⁴, implying that it is desirable to maximize the toroidal β, β_t = 2_μ0<p>/B_T,0². However, MHD stability theory and experimental results have shown that the maximum stable β_T scales with the toroidal field and plasma current as [1], where β_N is a constant in the range of 2-8. Hence, for fixed field B_T and current I_P, maximizing the fusion power is tantamount to optimizing the configuration for the highest possible β_N limit. Furthermore, the economic viability of the tokamak and ST configurations is greatly improved if they can operate with minimal external power used to drive the required plasma current. The self-generated current driven by the bootstrap effect scales as qI_P β_N, indicating again the important of maximizing the β_N limit. Both the plasma boundary shaping and the internal profiles can impact the β_N limit. For instance, a more “D” shaped plasma boundary, known as having high “triangularity” δ, will have a higher β_N limit than one with an circular boundary; pressure profiles that are “broad”, placing the maximum pressure gradient near the plasma boundary, also have higher β_N limits.

The desire for high-β configurations, however, may appear to be at odds with a second requirement for the tokamak system: that disruptions be minimized or avoided. A disruption [2] is the catastrophic loss of plasma confinement, followed by the rapid quenching of the plasma current. Interestingly, research at NSTX has shown that minimal disruptivity often occurs in the highest heating power and β_N regimes accessed in the device, provided certain criterion are met.

Figure 1: Disruptivity as a function of βN and a) shape factor or b) the internal inductance. The “N” on the colorbar indicates that no disruptions were observed in these bins.

“Disruptivity” is defined as the number of disruptions when the plasma is in a given portion of parameter space, divided by the total amount of discharge time the plasma is in that part of parameter space [3]. A database sampling the equilibrium properties of all beam-heated discharges in the 2006-2010 run campaigns has been created, in order to determine regimes of minimal disruptivity in NSTX [4].

As noted above, strong shaping of the plasma boundary can raise the β_N limit. A common means of capturing the degree of boundary shaping is the shape parameter S = q₉₅I_P/αB_T∝(l+K²)ƒ(A,k,δ)/A, where A is the aspect ratio and κ is the elongation of the boundary [5]. This parameter measures how effective boundary shaping is at raising the edge safety factor for given values of I_P and B_T, with increasing values of elongation, triangularity, and inverse aspect ratio raising S. Figure 1(a) shows the disruptivity data, sorted by β_N and S. The data show a clear trend of reduced disruptivity with strong shaping of the plasma boundary, with the lowest disruptivity in regions of strong shaping and high-β_N.

The β_N limit depends on the shape of the profiles within the plasma boundary. A key profile parameter is the internal inductance l_i = l²_P∭B²_PdV/V(µ₀I_P)², where l_p is the poloidal circumference of the boundary and V is the volume. This parameter measures the peaking of the current profile, with high l_i corresponding to configurations with the current more peaked on the magnetic axis. Equilibria dominated by the self-generated bootstrap current tend to have broad current profiles and low internal inductance. The low l_i facilitates strong shaping of the plasma cross-section, seen above to be beneficial in raising the β_N limit. However, reducing the internal inductance generally reduces the n=1 β_N limit of the configuration [6]. Fortuitously, the inclusion of nearby conducting structures, as is the case in NSTX, can recover higher global stability limits. Hence, we see in Fig. 1b) that a region of minimum disruptivity occurs with low internal inductance and high β_N. Indeed, with this conducting wall, some low-li cases in NSTX have β_N values up to twice the computed so called “no-wall β_N limits”, using the full available heating power. Finally, additional analysis shows that the strong toroidal rotation and broad pressure profiles contribute to reducing the disruptivity at high-β_N [4].

Figure 2: Plot of the pulse average bN, as a function of the IP flat-top duration, showing that high bN, and strong shaping facilitates the longest pulses in NSTX.

Another means of observing these trends in the data is shown in Fig. 2. This figure shows the value of β_N, averaged over the I_P flat-top, as a function of the flat-top duration, and color coded by the shape parameter S. The longest discharges in the NSTX databases occurred with strong shaping, and at high-β_N, consistent with the disruptivity being minimal in this portion of parameter space.

The reasons for these trends can be observed in a number of sources. In the presence of the stabilizing conducting wall as in NSTX, a new branch of the kink known as the resistive wall mode (RWM) can become unstable; this mode is a key limiting instability in NSTX [7]. A sensitive measure of RWM stability is “resonant field amplification” (RFA) [8]. In plasmas with β_N beneath the marginal stability point for the RWM, any n=1 externally applied fields can be amplified by the stable RWM, with the amplification increasing as the plasma approaches the marginal stability point. Hence, applying an external n=1 field and measuring the plasma response can provide a direct assessment of the global stability. Figure 3 shows the measured RFA from 20 high-performance discharges as a function of β_N/l_i [9]. We observe that the RFA is minimal at low-β_N or with peaked current profiles (high-l_i), has a peaked value at intermediate values of β_N/l_i , and then decreases with high-β_N and low-l_i, consistent with the disruptivity analysis. Further analysis [10,11] shows that this increase in stability at higher values of β_N/l_i is due to kinetic stabilization effects on the resistive wall mode, which tend to be strong in rapidly rotating NSTX plasmas.

Figure 3: Resonant field amplification as a function of βN/li.

A second cause of reduced disruptivity at high β_N and strong shaping is the reduced prevalence of core n=1 modes [4,12-14]. These modes typically manifest themselves in NSTX H-modes as coupled m/n=2/1 islands + 1/1 kinks; they result in a dramatic reduction in the plasma rotation, and often lock to the wall, leading to disruption. The broad current profiles at high-β_N and low-l_i assist in keeping q_min elevated, avoiding the instability conditions for these core n=1 modes.

Having established a recipe for minimal disruptivity, it is reasonable to ask why discharges in NSTX ever disrupted? It has been difficult to maintain the optimal conditions throughout the duration of the discharge in NSTX. As noted above, the plasma often evolved to have an unstable current profile. Furthermore, the evolution of the plasma rotation, profile parameters, boundary shape, or divertor parameters often resulted in the plasma deviating from the optimal state. A key goal of the NSTX-Upgrade [15] research program will involve developing profile [11], boundary, and divertor control techniques to maintain these high-performance, minimal disruptivity configurations.

References

1. E.J. Strait, Phys. Plasmas 1, 1415 (1994)
2. T.C. Hender, et al., Nucl. Fusion 47, S128 (2007)
3. P.C. deVries, et al., Nucl. Fusion 49, 055011 (2009)
4. S.P. Gerhardt, et al, “Disruptions, Disruptivity, and Safer Operating Windows in the High-β Spherical Torus NSTX”, submitted to Nuclear Fusion (2012)
5. E.A. Lazarus, et al., Phys. Fluids B 4, 3644 (1992)
6. W. Howl, et al., Phys. Plasmas B 4, 1724 (1992)
7. S.A. Sabbagh, et al., Nuclear Fusion 46, 635 (2006)
8. H. Reimerdes, et al., Nucl. Fusion 45, 368 (2005)
9. J.W. Berkery, et al., Proc. 24th Int. Conf. on Fusion Energy (San Diego, USA, 2012) (Vienna: IAEA) paper EX/P8-07
10. S.A. Sabbagh, et al., Nucl. Fusion 50, 025020 (2010)
11. J.W. Berkery, et al., Phys. Rev. Lett. 104, 035003 (2010)
12. S.P. Gerhardt, et al., Nucl. Fusion 49, 032003 (2009)
13. S.P. Gerhardt, et al., Nucl. Fusion 51, 073031 (2011)
14. J.A. Breslau, et al, Nuclear Fusion 51, 033004 (2011)
15. J.M. Menard, et al., Nucl. Fusion 52, 083015 (2012)

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Schedule of Burning Plasma Events

Click here to visit a list of previously concluded events.

2013 Events
March 19 - 22, ITPA DSOL Topical Group Meeting, Hefei, China April 15 - 18, ITPA IOS Topical Group Meeting, ITER April 22 - 24, ITPA Joint Meeting of PED and T&C Topical Groups, Garching, Germany April 29, DIII-D FY13 experimental campaign begins April 22 - 25, ITPA Joint Meeting of MHD and EP Topical Groups, Culham, United Kingdom June 4 - 7, ITPA Diag Topical Group Meeting, San Diego, United States July 1 - 5, EPS Conference on Plasma Physics, Espoo, Finland October 7 - 9, ITPA PED Topical Group Meeting, Japan November 11 - 15, APS DPP Meeting, Denver, United States December 9 - 11, ITPA CC and CTP Meeting, ITER
2014 NSTX-U commissioning operations begin	2020 November, First plasma at ITER
2019 First plasma at JT-60SA	2027 March, Beginning of full DT-operation at ITER

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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

Full archive of eNews at the USBPO website:
burningplasma.org/enews.html

Editor: David Pace (pace@psfc.mit.edu)
Assistant Editor: Amadeo Gonzales (agonzales@austin.utexas.edu)

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Image of the Month

 

Massive Magnetics Modernization

Research at the DIII-D tokamak is leading the development and understanding of the physics behind ELM-suppression and improved plasma performance due to the application of magnetic perturbations. ITER will also employ a set of externally controlled coils for ELM-suppression. These perturbations must be studied in three-dimensions, which has motivated a massive upgrade of the magnetic probe diagnostic at DIII-D. In this photograph, Dr. J. King (ORISE) is shown inside the DIII-D vessel during work to install 110 new probes that will provide a 3D measurement of the magnetic field equilibrium and fluctuations, along with toroidal and poloidal mode numbers of electromagnetic instabilities. The inset photographs show an example saddle loop (upper) and pick-up coil (lower). Coaxial probe wires are shown coiled up in the background awaiting routing to an exit port (a total in-vessel wire length of ~5 km is added for this upgrade). The laptop screen in the foreground displays real-time measurements of probe positions made with an articulated arm. This measurement data confirms the accuracy of the probe positions and alignment to better than 500 microns. This new diagnostic system will produce approximately 8 GB of data per shot (pared down to 1 GB stored), greatly enhancing experimental observations of applied magnetic perturbations. This work supported in part by the US Department of Energy under DE-FC02-04ER54698 and DE-AC05-06OR23100.

Contributed by J.D. King¹ and E.J. Strait^2
1Oak Ridge Associated Universities, Oak Ridge, TN 37831
²General Atomics, PO Box 85608, San Diego, CA 92186

DIII-D: https://fusion.gat.com/global/DIII-D

Click here to visit a Directory of Other Plasma Events

Please contact the administrator with additions and corrections.

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