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CONTENTS
USBPO Topical Group Highlights G. Wallace, S. Shiraiwa, R. Parker, and the C-Mod LH Team
ITPA Report Image of the Month Upcoming Burning Plasma-related Events
Dear Burning Plasma Aficionados:
This newsletter provides a short update on U.S. Burning Plasma Organization activities. E-News is also available online. Comments on articles in the newsletter may be sent to the editor (David Pace) or assistant editor (Amadeo Gonzales). Thank you for your interest in Burning Plasma research in the U.S.!
USBPO Topical Group Highlights
Editors note: The BPO Integrated Scenarios Topical Group works to facilitate U.S. efforts to understand, improve, and predict the behavior of whole-device operation [leaders are Stefan Gerhardt and Chris Holcomb]. This month's Research Highlight by G. Wallace, et al., describes experimental research using lower hybrid waves to drive plasma current in the Alcator C-Mod tokamak. In addition to experiments, this research also includes significant modeling and simulation efforts, and fusion technology development. Efficient non-inductive current drive is a demonstration goal for ITER, and it will certainly be important for electricity-producing reactors.
Lower Hybrid Current Drive Exeriments in Alcator C-Mod
Greg Wallace, Syun'ichi Shiraiwa, Ron Parker, and the C-Mod LH Team
MIT Plasma Science and Fusion Center, Cambridge, MA 02139, USA
In addition to its highest level scientific objective to obtain an energy gain, Q, of 10, the mission of ITER is also to demonstrate "steady state through current drive at Q > 5" [1]. This objective is motivated by the need for a fusion reactor to operate in essentially steady-state conditions, where the lifetime of the components would not be unduly limited by considerations of thermo-mechanical fatigue.
The conventional tokamak is an inherently pulsed device as can be appreciated by the fact that the toroidal current needed to provide equilibrium and confinement is generated through mutual induction as in a transformer, with an electromagnet as the primary coil and the plasma itself as the secondary. By using the integral form of Faraday’s Law, the loop voltage, Vloop, induced around a closed toroidal contour in the plasma is:
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where ψ is the magnetic flux through the area enclosed by the contour. Vloop must be held constant to create a steady state toroidal current through induction, however it is not possible to maintain a constant ∂ψ/∂t indefinitely as this would require a power supply with no current limit, as well as a primary coil which could withstand an infinite current.
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Figure 1: The LHCD antenna and molybdenum protection limiters installed on Alcator C-Mod |
The limitations of a pulsed tokamak can be overcome by the addition of non-inductive current drive mechanisms. Current can be driven non-inductively through the application of high power neutral beams and RF waves, and even by the plasma itself due to radial gradients of the temperature and density (the so-called bootstrap current). Lower Hybrid Current Drive (LHCD) utilizes waves at RF frequencies between the ion cyclotron and electron cyclotron frequencies [2]. The optimum LHCD frequency for typical burning plasma parameters (toroidal magnetic field, BT ~ 5 T, and electron density, ne ~ 1020 m-3) is ~5 GHz. Lower Hybrid (LH) waves are launched preferentially in one toroidal direction by a phased array of waveguides. The waves deposit momentum and energy directly to electrons satisfying the Landau resonance condition ω/k|| ≈ ve||, where ω/k|| (~ c/2) is the parallel phase velocity of the wave and ve|| is the parallel velocity of the electron. Here “parallel” refers to the direction parallel to the magnetic field in the plasma. Wave damping becomes strong when ω/k|| ~ 3vte. Adjusting the launched k|| spectrum changes the temperature, and thus the radius, at which the LH waves will damp and drive current. These electrons, which are weakly relativistic and relatively collisionless, can drive a significant current even if the LH driven fast electrons are few in number. As compared to other non-inductive current drive mechanisms, LHCD has a relatively high engineering current drive efficiency.
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This makes LHCD a very desirable means for driving non-inductive current, but it is not without its drawbacks. Unresolved physics and engineering issues such as compatibility with ion cyclotron resonance heating (ICRH), anomalously low current drive efficiency at high density, and survivability of the LHCD antenna in a steady state environment must be understood and overcome if it is to be applied to a reactor.
The LHCD system on Alcator C-Mod operates with similar key parameters for current drive as foreseen for ITER and DEMO: ƒLHCD = 4.6 GHz, BT = 3-8 T, ne = 0.5-5x1020 m-3 [3]. Figure 1 shows the LHCD antenna installed on C-Mod. The antenna consists of a 16x4 phased array of active waveguides with a column of 4 passive waveguides on each side of the antenna. Molybdenum limiters on either side protect the antenna from the plasma. The launched k|| spectrum can be adjusted during a discharge by varying the phase difference between adjacent columns. Fully non-inductive discharges on C-Mod (plasma current, Ip ~500 kA and central density, n0 ~ 0.7x1019 m-3) are used to study plasma transport and magnetohydrodynamic (MHD) stability under conditions of zero loop voltage and enhanced magnetic shear. The short current relaxation timescale (τCR ~ 0.15 - 0.20 s) of these discharges allows the current profile to reach a steady state equilibrium during the available LH pulse lengths of 0.5 - 2.0 s. Figure 2 shows a non-inductive discharge exhibiting an internal transport barrier (ITB) in the electron temperature profile. The ITB is triggered by shear reversal in the safety factor profile, which occurs when the current profile in the plasma relaxes to a new equilibrium. This ITB is terminated by a 2/1 MHD tearing mode. Additional power injected at somewhat higher phase velocity is predicted to move the ITB outward in minor radius and raise the central safety factor, improving the MHD stability and enhancing both the self-generated bootstrap current and confinement. Verifying this prediction awaits installation of a second antenna and an additional MW of RF power, now in the fabrication stage.
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Figure 2: A non-inductive discharge with LHCD on Alcator C-Mod. An electron temperature ITB is present in the shaded time window. Panels show (a) plasma current and LHCD power (b) loop voltage (c) Thompson scattering core Te (d) core soft X-rays (e) soft X-ray profile showing increase in amplitude of a rotating 2/1 island during the ITB phase. |
The LH research program on C-Mod is also focused on extending LHCD to higher densities (ne ~ 1.5x1020 m-3). These higher density plasmas allow access to so-called advanced non-inductive scenarios in which the plasma current is sustained by roughly equal combinations of microwave current drive (in this case LHCD) and bootstrap current. The higher density is desirable for increasing the bootstrap current, which as mentioned above relies on gradients in the plasma temperature and density for self-generation. C-Mod experiments and simulations show that absorption of LH waves in the outer edges of the plasma at high density can reduce current drive efficiency through a variety of mechanisms, including collisional absorption [4,5], upshifts in the wave k|| [6], and parametric instability [7,8]. Modeling of these discharges indicates that the edge losses can be mitigated by moving from the present C-Mod multi-pass absorption regime, in which the waves pass from the core through the edge and back to the core several times before fully damping, to a strong single pass absorption regime. A new LH antenna design under construction aims to increase single-pass absorption through the use of a velocity-space synergy between waves launched from a conventional LH antenna at the mid-plane and the new antenna located off mid-plane [9]. Figure 3 shows the predicted driven current with and without velocity space synergy for a discharge with a line-average density of 1.4x1020 m-3. When velocity space synergy is included, a maximum current of 0.3 MA is obtained at n|| ≡ ck||/ω = -2 from the mid-plane antenna and n|| = -3 from the off mid-plane antenna, while the maximum current without synergy would be 0.2 MA.
Single-pass absorption in the core of a burning plasma, such as ITER or DEMO, is expected to be very strong due to the high electron temperature, and absorption of LH waves in the plasma periphery should be small. The experiments at C-Mod will provide an important verification of the model predictions regarding the role of single-pass absorption in LHCD at high density. Increasing the amount of current driven at high density will also allow for deeper exploration of the physics of these advanced non-inductive scenarios on C-Mod.
The C-Mod research program also addresses technology issues LHCD antennas will face operating in a reactor environment. Lower hybrid waves require that the electron density be in excess of the cutoff density (2.7x1017 m-3 at 4.6 GHz) for propagation. Both LHCD and ICRH are likely to be critical auxiliary systems on a reactor, and the two must operate together to provide the necessary heating and current drive. The use of ICRH antennas that are magnetically connected to the LH antenna can reduce the density in front of the antenna to below that critical cutoff, which results in very high reflected power. A feedback controlled fast ferrite stub tuner (FFT) under development at C-Mod will provide dynamic load matching to many plasma conditions. The FFT will redirect reflected power back to the plasma, increasing the overall coupling efficiency and net power during combined operation with ICRH. The FFT will also allow the antenna to operate effectively farther from the plasma edge where the plasma density, and consequent heat flux on the antenna, is lower, thus increasing the reliability and lifetime of the highly engineered launching structure.
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Figure 3: GENRAY/CQL3D [10] prediction of total LH driven current for the mid-plane (LH2) + off mid-plane (LH3) antennas with (top) and without (bottom) velocity space synergy. Colors indicate driven current in MA. |
Recent simulations [11] have shown that achieving ITER's steady-state goal may require substantial off-axis current drive produced by 20-40 MW of LHCD. Alcator C-Mod’s unique parameters provide a crucial test-bed for engineering and physics issues associated with the use of LHCD on ITER. The C-Mod research effort on LHCD is laying the groundwork needed for validated simulations which will allow reliable predictions of the performance of lower hybrid current drive systems in ITER, and its ability to access steady state regimes at Q > 5.
This work was supported by US Department of Energy awards DE-FC02-99ER54512 and DE-AC02-09CH11466.
References
- Y. Shimomura, et al., Nucl. Fusion 41 309 (2001)
- N.J. Fisch, Rev. Mod. Phys. 59, 175–234 (1987)
- P.T. Bonoli, et al., Fus. Sci. Tech. 51, 401-436 (2007)
- G.M. Wallace, et al., Phys. Plasmas 17, 082508 (2010)
- E. Barbato, Nucl. Fusion 51, 103032 (2011)
- O. Meneghini, PhD Thesis, Massachusetts Institute of Technology, (2011); G.M. Wallace, et al., Phys. Plasmas 19, 062505 (2012)
- R. Cesario, et al., Nature Commun. 1, Article number: 55 (2010)
- S.G. Baek, et al., submitted to Plasma Phys. Contr. Fusion (2012)
- G.M. Wallace, et al., submitted to Nucl. Fusion (2012)
- A. P. Smirnov and R. Harvey, Bull. Am. Phys. Soc., 40, 1837 (1995); R. W. Harvey and M. McCoy, in Proceedings of the IAEA Tech. Comm. Meeting on Simulation and Modeling of Thermonuclear Plasmas, Montreal, (1992)
- F. Poli, et al., submitted to Phys. of Plasmas (2012)a S
ITPA Report
Summary of the 9th Meeting of the ITPA Topical Group on MHD Stability
E. Strait1, P. Martin2, and Y. Gribov3
1 General Atomics, San Diego, California, United States
2 Consorzio RFX - Associazione ENEA-Euratom sulla fusione - Padova, Italy
3 ITER Organization, 13115 St. Paul Lez Durance, France
The ITPA Topical Group on MHD Stability held its ninth meeting in San Diego, USA during October 15-17, 2012, immediately following the IAEA Fusion Energy Conference. The meeting was hosted by the General Atomics. There were approximately 51 participants in the MHD group meeting, including 2 from the ITER IO, 2 from Japan, 10 from the EU, 2 from India, 1 from Korea, 27 from the US, 4 from China and 3 from Russia. These totals include several remote participants from the EU and US.
The scientific program of the MHD group meeting covered key issues of MHD stability and non-axisymmetric plasmas, including sawtooth stability, tearing mode stability, resistive wall mode stability, 3D equilibrium, error fields, and disruption avoidance and mitigation. A joint session with the Divertor/SOL topical group discussed disruptions with tungsten versus carbon divertors. Two joint sessions with the Energetic Particles topical group focused on runaway electron physics and MHD instabilities with energetic particles.
It has been proposed to transfer responsibility for the topic of runaway electrons from the Energetic Particles group to the MHD Stability group, because this topic is very closely related to the issues of disruptions and disruption mitigation, which are within the scope of the MHD group. This proposal was previously circulated by e-mail to the members of both groups, and was agreed at the joint session on this topic. The proposal will now be submitted to the ITPA Coordinating Committee.
The MHD TG also reviewed the progress of the various joint experiments and working groups. (A list of the groups is shown below.) A draft preliminary report on radiation asymmetry during massive gas injection has been prepared by working group WG-8, and is now going through clearance by the participating institutions. It is proposed to close three of the joint experiments: MDC-4: NTM physics - aspect ratio comparison, MDC-5: Sawtooth control methods for NTM suppression, and MDC-14: Rotation effects on neoclassical tearing modes. Possible new joint experiments to extend these studies will be considered during the coming months.
The urgent R&D needs for disruption mitigation in ITER were considered in two sessions of open discussion. The priority of these issues is determined in part by their urgency for the design of ITER’s disruption mitigation system. The topical group will continue to consider these issues, with the aim of articulating a proposed plan for R&D to address them.
The next meeting of the TG will be held on April 22-25, 2013 at the Culham Science Center, Abingdon, UK, in conjunction with a meeting of the Energetic Particles Topical Group.
Joint Experiments and Working Groups of the MHD Stability Topical Group
Joint Experiments
MDC-1 Disruption mitigation by massive gas jets
MDC-2 Joint experiments on resistive wall mode physics
MDC-4 NTM physics - aspect ratio comparison (*)
MDC-5 Sawtooth control methods for NTM suppression (*)
MDC-8 Current drive prevention/stabilization of NTMs
MDC-14 Rotation effects on neoclassical tearing modes (*)
MDC-15 Disruption database development
MDC-16 Runaway electron generation, confinement, and loss
MDC-17 Active disruption avoidance
MDC-18 Evaluation of axisymmetric control aspects for ITER (*) Closed at the end of 2012.
Working Groups
WG-7 Resistive Wall Mode feedback control
WG-8 Radiation asymmetry during MGI
WG-9 Criteria for error field correction
WG-10 Halo current modeling
WG-11 Control of locked modes
WG-12 3D distortion of the plasma boundary
Joint Activities
JA-1 Joint Theoretical Activity on Shear Flow Effects for NTMs.
Image of the Month
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A Fisheye View Inside Alcator
written by M. Greenwald
The Alcator C-Mod tokamak, which operates at ITER plasma densities (1020 cm-3) and magnetic field (5.3 T) features a high-Z metal first wall and reactor level power and particle fluxes, enabling burning plasma relevant plasma-wall studies. Two ion cyclotron resonance heating (ICRH) antennas are also visible on either side of the central column. The one on the right is a novel "field-aligned" design in which the current straps are aligned perpendicular to the total magnetic field (in both antennas the Faraday screen is aligned with the total magnetic field). The field-aligned antenna results in a lower E|| and produces fewer impurities from the interaction of RF waves with the first wall. [This view is taken from the position of the LH-launcher discussed in this month’s Research Highlight - Ed.]
Full Resolution: http://upload.wikimedia.org/wikipedia/commons/c/c7/Alcator_C-Mod_Tokamak_Interior.jpg
Alcator C-Mod: http://www.psfc.mit.edu/research/alcator/facility/index.htm
(Suggestions for the BPO eNews 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.)
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