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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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ITPA Update

For information on the proposed agenda, see BPO forum link.

Coordinating Committee
	4th Meeting, ITER Site, France, December 9 - 11, 2013
Diagnostics Topical Group
	24th Meeting, San Diego, CA, USA, June 4 - 7, 2013
	BPO Forum: https://burningplasma.org/forum/index.php?showtopic=1247
Energetic Particle Physics Topical Group
	10th Meeting, Culham, UK, April 22 - 25, 2013 Agenda presently under discussion.
	Agenda presently under discussion.
	BPO Forum: https://burningplasma.org/forum/index.php?showtopic=1243
Integrated Operation Scenarios Topical Group
	10th Meeting, ITER Site, France, April 15 - 18, 2013
	Agenda includes "Review of ITER Control System," "Report on use of W in ITER, " etc.
MHD, Disruptions & Control Topical Group
	21st Meeting, Culham, UK, April 22 - 25, 2013
	Primary Topic: Disruptions
Pedestal & Edge Physics Topical Group
	24th Meeting, IPP Garching, Germany, April 22 - 24, 2013
Scrape-Off-Layer & Divertor Topical Group
	18th Meeting, Hefei, China, March 19 - 22, 2013
Transport & Confinement Topical Group
	10th Meeting, IPP Garching, Germany, April 22 - 25, 2013
	BPO Forum: https//burningplasma.org/forum/index.php?showtopic=1250

 

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USBPO Topical Group Highlights

(Editors note: The BPO Operations and Control Topical Group works to facilitate US efforts to understand and control the operation of existing and future fusion devices [leaders are Michael Walker and Egemen Kolemen]. This month's Research Highlight by S.A. Sabbagh, et al., describes the cutting edge of techniques to control resistive wall modes (RWMs) in which algorithms incorporate the entire conducting structure of the reactor in addition to the plasma response. As in last month’s eNews (Jan. 2013), the Topical Group Leaders have chosen a Research Highlight centered on disruption prevention or control, underscoring the intense US interest in that topic. -Ed.])

Controlling Global Instabilities for Continuous Tokamak Operation
S.A. Sabbagh¹, J.M. Bialek¹, S.P. Gerhardt² and O.N. Katsuro-Hopkins¹

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

Tokamak and spherical torus (ST) magnetic fusion devices designed to produce high levels of fusion reactions such as ITER [1], a nuclear Component Test Facility (e.g. ST-CTF [2]) or a pilot power plant (e.g. ST-Pilot [3]) are planned to be operated over timescales of many minutes to weeks. Therefore, large scale magnetohydrodynamic (MHD) instabilities driven by plasma pressure and current must be stabilized in these devices to avoid plasma disruption and termination on rapid (as low as microsecond) timescales. These fast terminations can cause significant electromagnetic forces and impulsive heat loads capable of damaging device structural elements and plasma facing components. Rapidly growing global kink-ballooning modes, which typically cause disruptions, can be stabilized in the presence of electrically conducting structure when rotating faster than the structure eddy current decay time (typically millisecond timescale). The field generated by these modes can couple strongly with the wall, causing the mode to rotate more slowly. These resistive wall modes (RWM) grow on the eddy current decay timescale [4,5] and can have complex shapes, due to a spectrum of basic eigenfunctions that form the mode (Figure 1:) [6]. Tokamak and ST experiments have demonstrated success in stabilizing the RWM both passively [6] and actively [7,8]. Research conducted to actively control global modes aims to eliminate disruptions that terminate the plasma due to these instabilities in fusion devices.

Figure 1: Reconstruction of experimentally unstable resistive wall mode with toroidal mode number components n = 1 - 3 (left: external view, right: internal view).

Active control of these instabilities is becoming more sophisticated to further improve control reliability and to allow control systems to operate in difficult environments. For example, the plasma response can be modeled in various ways to predict mode dynamics and to determine the impact on magnetic control sensor measurements. Also, if it has an impact on control, the environment itself may be modeled. For example, the coils used to produce the magnetic field for RWM control in a device producing copious amounts of fusion power, such as ITER, must be shielded to avoid damage from the fusion neutrons. The electrically conducting structure of the device and the coil shielding carry currents generated by the plasma response and the applied control field, which contribute to the magnetic field that the control sensors measure, and so need to be included in the control model. Such a control system, including both the plasma response and the effect of conducting structure and plasma mode currents has been analyzed for ITER and found to be effective theoretically [9]. The controller uses a state-space approach and optimal control techniques.

Figure 2: Cut-away view of the three- dimensional conducting structure model of NSTX used in the RWM state space controller. Shown are the vacuum vessel, copper stabilizing plates, neutral beam port, magnetic control sensors (blue), and control coils (red).

A variant of such a system using state derivative feedback [10] has been implemented and used for resistive wall mode control in the National Spherical Torus Experiment (NSTX) [11]. A significant advantage of this approach over a simpler proportional gain controller is the inclusion of the state observer, which is used to correct the time evolution of the state vector by comparing a real-time model of the sensor signals with measurements. A linear quadratic regulator approach is used to determine the optimal controller gain. Along with the plasma response model [12], a three-dimensional model of the NSTX device conducting structure and magnetic control sensors (FIG. 2) is included in the controller. Control theory provides guidance on the number of states needed for the real-time RWM state-space controller (RWMSC) model to faithfully reproduce the measurements, which in this case is about 7. FIG. 3 shows how well the RWMSC model with 7 states compares to the measured sensor data. Three-dimensional details of the model can be important. An example of this is shown in FIG. 3, where frame (d) shows a significant improvement in the model/experiment comparison when the large neutral beam port is included in the model, versus when it is omitted [frame (c)].

Figure 3: Comparison of RWM sensor data (black) with RWMSC model (red) for (a) 2 states, and (b) 7 states; also (c) without, and (d) with the inclusion of the neutral beam port in the model (both with 7 states).

The fusion power produced in a tokamak scales as β_t²B₀⁴ (where β_t ≡ 2μ₀<p>/B₀² is the toroidal beta, <p> is the plasma pressure, and B₀ is the toroidal magnetic field). The normalized beta, β_N ≡ 10⁸<β_t>aB₀/I_p (where a is the plasma minor radius, and I_p is the plasma current), is a typical figure-of-merit for global mode stability. In addition, tokamak fusion plasmas designed to operate continuously have a high fraction of self-generated current (bootstrap current). Such plasmas tend to have broad current profiles, which is represented by low values of the plasma internal inductance, l_i (a measure of the current profile peakedness). Broader current profiles tend to decrease global mode stability with no stabilizing structure, and so the ratio β_N/l_i is also used as a stability figure-of-merit. The RWMSC on NSTX has been used to control instabilities in plasmas with high values of β_N reaching 6.4, and β_N/l_i reaching 13.4. These values are double the ideal MHD stability limit computed with no stabilizing structure for instabilities with toroidal mode number of unity. These plasmas have β_N and l_i in the range envisioned for ST-CTF and ST-Pilot devices. Future use of the RWMSC will concentrate on sustaining high stability parameters with very low disruption rate in the coming upgrade of NSTX [13], which will extend the plasma duration from about 1 second to over 5 seconds.

References

1. M. Shimada, et al., Nucl. Fusion 47, S1 (2007)
2. Y.K.M. Peng, et al., Plasma Phys. Control. Fusion 47, B263 (2005)
3. J.E. Menard, et al., Nucl. Fusion 51, 103014 (2011)
4. A. Bondeson, and D. J. Ward, Phys. Rev. Lett. 72, 2709 (1994) [
5. E.J. Strait et al., Phys. Rev. Lett. 74, 2483 (1995)
6. S.A. Sabbagh, et al., Nucl. Fusion 46 , 635 (2006)
7. E.J. Strait, et al., Phys. Plasmas 11, 2505 (2004)
8. S.A. Sabbagh, et al., Phys. Rev. Lett. 97, 045004 (2006)
9. O. Katsuro-Hopkins, et al., Nucl. Fusion 47, 1157 (2007)
10. T.H.S. Abdelaziz, and M. Valasek, in 16th IFAC World Congress, edited by P. Zitek (International Federation of Automatic Control, Czech Republic, 2005)
11. S.A. Sabbagh, et al., Proc. 24th Int. Conf. on Fusion Energy (San Diego, USA, 2012) (Vienna: IAEA) paper OV/3-1; S.A. Sabbagh, et al., “Overview of Physics Results from the National Spherical Torus Experiment”, submitted to Nuclear Fusion
12. A.H. Boozer, Phys. Plasmas 6, 3180 (1999)
13. J.E. Menard, et al., Nucl. Fusion 52, 083015 (2012)

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Announcements

Proposals Sought:

Physics and Engineering Design Support and Diagnostic Hall Instrumentation Development for ITER Low Field Side (LFS) Reflectometer (due by 4:00 pm EST on 3/14/2013) Full announcement on the U.S. Federal Business Opportunities website, https://www.fbo.gov/index?s=opportunity&mode=form&id=55a3684939b9b0da7f84d9fa543bdf03&tab=core&_cview=1 with more information on available at, http://procurement.pppl.gov/Procurement%20Opportunities.HTM

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Schedule of Burning Plasma Events

Click here to visit a list of previously concluded events.

2013
March 19 - 22, ITPA: 18th Scrape-Off-Layer & Divertor TG Meeting, Hefei, China April 15 - 18, ITPA: 10th Integrated Operation Scenarios TG Meeting, ITER April 22 - 24, ITPA: 24th Pedestal and Edge Physics TG Meeting, Garching, Germany April 22 - 25, ITPA: 10th Transport & Confinement TG Meeting, Garching, Germany April 29, DIII-D FY13 experimental campaign begins April 22 - 25, ITPA: 10th Energetic Particle Physics TG Meeting, Culham, United Kingdom April 22 - 25, ITPA: 21st MHD Disruptions and Control TG Meeting, Culham, United Kingdom June 4 - 7, ITPA: 24th Diagnostics TG 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: 4th CC/CTP Meeting, ITER December 11, 4th CTP Ex Com 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

 

Steady Dynamo, Stable Plasma

Innovative current drive research conducted on the Helicity Injected Torus-Steady Inductive device (HIT-SI, pictured above in 2004 prior to being enclosed in the vacuum chamber, and featuring then-graduate student V. Izzo*) at the University of Washington, has demonstrated a new dynamo method to form and sustain 55 kA of current in the confined plasma. This newly discovered steady state plasma sustainment mechanism involves driving the edge currents directly and imposing cross-field-current-coupling magnetic fluctuations using asymmetric injectors (in previous magnetic dynamo experiments, driving the plasma unstable produced the fluctuations). These injectors, the two handle-shaped coils shown in the upper inset model, alternately generate currents on either side of the central core. In HIT-SI the fluctuations are imposed by the injectors on a stable, and therefore better confined, plasma configuration. The lower inset shows the magnetic field lines resulting from the edge current drive and imposed fluctuations (gray), and the confining field maintained by dynamo action (color). The equilibrium can be efficiently sustained, and the current profile can, in principle, be controlled, both of which are necessary developments for controlled fusion. The next step, requiring a larger device, is to optimize the plasma confinement of this scenario such that it can be scaled up for fusion power production.

Contributed by T.R. Jarboe, University of Washington, Seattle, WA 98195
HIT-SI: http://www.aa.washington.edu/research/HITsi/index.html

T.R. Jarboe, B.S. Victor, B.A. Nelson, C.J. Hansen, C. Akcay, D.A. Ennis, N.K. Hicks, A.C. Hossack, G.J. Marklin and R.J. Smith, “Imposed-dynamo current drive,” Nucl. Fusion 52, 083017 (2012)
http://dx.doi.org/10.1088/0029-5515/52/8/083017

*Presently an Associate Research Scientist in the Center for Energy Research at UC San Diego.

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