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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.

CONTENTS

Announcements
Director’s Corner	C.M. Greenfield
Research Highlight	Dan Brunner
Schedule of Burning Plasma Events
Contact and Contribution Information

 

NAS Burning Plasma committee

The NAS committee on “A Strategic Plan for U.S. Burning Plasma Research” held its fifth public meeting on Feb. 26-28 at General Atomics. Presentations to the committee can be found on the www.firefusionpower.org website. Additional public input submitted directly to the committee can be found here:

http://sites.nationalacademies.org/BPA/BPA_184701

 

ITER internships

An announcement for ITER internship opportunities for undergraduate and graduate students was recently posted. Please see the following link for a description: https://www.iter.org/jobs/internships

The proposed topics can be found here:

https://www.iter.org/doc/www/content/com/Lists/WebText_2014/Attachments/321/Call_for_topics-Internship_2018_V3.pdf

 

By C.M. Greenfield

A new record for a US Burning Plasma Organization webinar

Many of you joined us for our web seminar on February 13 to hear as Bernard Bigot, Director General of ITER, gave an update on the status and plans of the ITER project. We had about 90 sites connected to the Zoom video, including several with large rooms (about 40 in one conference room here at General Atomics). A conservative estimate is that we had an audience of at least 200 people from all over the United States (and a few from abroad) watching and listening to a talk given from France. And it worked quite well.

Regarding the talk itself, we saw that there is rapid progress at the ITER site and within each of the seven domestic agencies, with the entire project on track for a 2025 first plasma. Many of you have asked for Dr. Bigot’s slides; they have been posted at

/resources/ref/APS_Town_Hall_Videoconference+bb.pdf

Of course, we have been organizing seminars like this for some time, but the size of the audience for this month’s seminar encourages us that the technology works well. We are constantly looking for interesting and informative topics for the burning plasma community and your ideas are always welcome.

ITER construction progress

During my visits to the ITER site over the last two years, I have watched as the reinforced concrete “bioshield” that will surround the ITER tokamak gradually rose from the ground. Earlier this month the bioshield was completed and the scaffolding surrounding it was removed, revealing the cylindrical structure you see in the photo below.

For more pictures or the bioshield and the rest of the facility, please visit the ITER website at https://iter.org. Of particular interest is the 3-D interactive worksite tour at https://static.iter.org/com/360/. Here you can find tours from different dates, going back to mid-2016. The tours can be viewed either on a screen or with VR goggles such Google Cardboard.

Pedestal and Divertor/SOL (Leaders: Mike Jaworski & Jerry Hughes)

This month's research highlight by Dr. Dan Brunner of MIT describes novel feedback experiments on the Alcator C-Mod tokamak which successfully mitigated reactor relevant heat fluxes to divertor surfaces while maintaining good overall energy confinement. This exciting result is discussed in the context of additional heat removal challenges that lie ahead for reactors, and points toward critical research needed to address these challenges. Dr. Brunner published this research in Nuclear Fusion in 2017.

Practical challenges for feedback control of tokamak heat exhaust

 

D. Brunner†¹, S.M. Wolfe¹, B. LaBombard¹, A.Q. Kuang¹, B. Lipschultz², M.L. Reinke³, A. Hubbard¹, J.W. Hughes¹, R.T. Mumgaard¹, J.L. Terry¹, M.V. Umansky⁴,

and the Alcator C-Mod Team¹

¹MIT Plasma Science and Fusion Center, Cambridge MA 02139

²York Plasma Institute, University of York, Heslington, York YO10 5DQ UK

³Oak Ridge National Laboratory, Oak Ridge TN 37830

⁴Lawerence Livermore National Laboratory, Livermore CA 94550

 

†brunner@psfc.mit.edu

Mitigation and control of extreme power exhaust is one of the great challenges that tokamaks must overcome to provide fusion power. Reactor-class tokamaks (e.g., ITER, Demo, ARIES, ARC) will have parallel heat flux densities impinging on their divertor targets in excess of 5-10 GW/m² [1], if left unmitigated. Presently, the state-of-the-art with regard to heat flux mitigation is the vertical target plate divertor, which is planned for ITER. In this geometry, magnetic field lines strike the divertor target at oblique angles (down to ~1°), reducing incident heat flux densities by up to a factor of ~50. Additional reduction is attained by volumetric dissipation, such as low-Z impurity line radiation and plasma-neutral interactions. These processes can be enhanced by employing a closed divertor geometry and by injecting low-Z impurity ‘seed’ gases into the divertor. Alcator C-Mod was the first tokamak to employ a closed, vertical target plate geometry and, being the only tokamak in the world capable of approaching reactor-level conditions (1–2 GW/m², unmitigated), tested this divertor configuration extensively – with both feed forward injection of low Z impurities and most recently with feedback control of impurity injection, using a first-of-a-kind direct measurement of divertor target plate heat fluxes [2].

One of the key results of feed forward impurity injection experiments in C-Mod divertor [3–6] was the demonstration that greater that 90% of the tokamak exhaust power could be dissipated while still maintaining good core plasma confinement, H₉₈ > 1. But this required a careful shot-by-shot tuning of the impurity seed gas programming to obtain both a highly dissipative, partially detached divertor state and an optimized pedestal/core plasma response.

For a reactor, it is imperative to develop and demonstrate a reliable and robust scheme for fast, real-time feedback control of the divertor heat flux – at no time should unmitigated heat fluxes be allowed to shine through to the divertor target. To this end, the Alcator Team investigated the use of surface thermocouples as a means for direct, real-time measurements of surface heat fluxes [2,7,8]. Such sensors have distinct advantages over IR based systems; there is no need to correct for a loosely-bound surface layers [9] nor the need to re-calibrate the optical system and account for time-evolving surface emissivities [10]. The C-Mod digital plasma control system was configured to input the real-time surface heat flux measurements and to output a demand signal to adjust the injection rate of impurity gas into the private flux region of the divertor via a fast-acting piezo valve. The close-loop response time of the system was ~100­–200 ms, set largely by the flow rate of gas down the delivery tube (~2 m length) and volumes behind the divertor and surrounding the plasma. This is perhaps the shortest close-loop response time that one can practically achieve in a tokamak. Reactor-scale devices will have longer time responses due to the longer distances traveled and larger volumes to fill. For example, the valve for ITER’s gas system is ~20 m away from the plasma. The gas takes 500 ms to travel the length of the tube and another 500 ms before it reaches ~2/3 of the maximum flow rate [11,12].

Using the feedback system in EDA H-mode plasmas (steady-state, ELM-free [13]), the C-Mod team was able to reproduce the performance demonstrated earlier in feed-forward experiments: nearly complete mitigation of plasma surface heat flux while maintaining excellent core confinement. In an example case, Figure 1, the peak surface heat flux was mitigated from ~50 MW/m² (~1 GW/m² parallel to the magnetic field) down to ~10 MW/m² while maintaining good core confinement H₉₈~1.05.

Fig. 1. Feedback control of N₂ impurity seeding during an EDA H-mode. Peak surface heat flux was controlled down to <10 MW/m² while maintaining H₉₈>1.

Although demonstrating feedback control under relatively steady conditions is a notable achievement, obtaining feedback control and mitigation of transient divertor heat fluxes is ultimately required and much more challenging. The challenge stems from the very small operational window available to a conventional vertical target plate divertor [14,15]. This in turn requires a very fast and precise heat flux detection and feedback response. The operational window of a divertor is defined as the combination of incident plasma heat flux, plasma density, and impurity seed density in which sufficient heat flux mitigation can be obtained without adversely affecting the core plasma. Too much mitigation can cause the region of cold, high-density, radiating divertor plasma to intrude onto closed flux surfaces near the x-point region, forming an ‘x-point MARFE’ [16,17]. Core plasma confinement tends to degrade with x-point MARFEs, with H₉₈ values typically dropping to ~0.7–0.9 [17,18]. For ITER, lowering H₉₈ from 1.0 to 0.8 would reduce Q from 10 to 5 [19].

Too little mitigation can cause reattachment of the plasma. For a solid metal, immediate damage could result due to melting, which causes both surface deformation and loss of material as well as recrystallization. Long term damage due to erosion and energetic helium bombardment can result as well. Reattachment and exposure to unmitigated heat fluxes can cause permanent melt damage to the divertor plate in well under 1 s (see Figure 1 in [2]). Even a small amount of melt damage and surface deformation in high-heat flux regions is likely unacceptable, since it can run-away – accelerating the damage and finally disallowing any plasma operation with significant heat fluxes contacting the melt-damaged surface [20]. These considerations have led researchers to considered self-renewing liquid metal surfaces [21] to avoid melt damage and erosion issues.

Despite the direct surface heat flux measurement technique and fast closed-loop impurity seeding response of the C-Mod feedback system [2], we have found that such control schemes appear to have an important, fundamental limitation: they cannot possibly respond fast enough to many of the most important transients (e.g., intrinsic impurity injections, H-L transitions). This means that the use of impurity seeding with standard vertical target plate divertor geometry is probably not an acceptable approach for a reactor. Improved sensors, actuators, and control systems are unlikely to change this conclusion.

Fortunately, new ideas for divertor geometries and divertor plasma physics operational regimes – which have been yet to be explored experimentally – show potential for handling extreme heat fluxes and fast transients in tokamaks. These ideas include a variety of ‘advanced’ magnetic divertor geometries proposed over the last ~10 years. For example, simulations with the 2D plasma-neutral fluid code UEDGE suggest that the scrape-off layer power window for obtaining a passively-stable, fully detached divertor condition for an X-point target divertor (as proposed for the divertor test tokamak ADX [1]) spans the range of ~0.5–7.0 MW [22,23]. This is a factor of ~10 improvement in heat flux handling and detachment power window compared to the original Alcator C-Mod vertical target plate divertor. In response to variation in power exhaust, the divertor detachment thermal front simply changes its equilibrium location in the leg – moving closer to the divertor target at high power and closer to the core plasma at low power, yet all the while remaining fully detached. The combined effects of long-leg magnetic geometry, enhanced gas-plasma interactions, presence of a secondary x-point in the leg and cross-field transport to the divertor leg sidewalls (akin to main-chamber recycling) are all found to contribute to this behavior. These are promising results towards a controllable solution. MAST-U, having the first tightly baffled, long-legged divertor, will provide an important first test of this physics, but a divertor test tokamak will be needed to test it at reactor-level conditions and retire this key risk.

Such a robust divertor response to heat flux potentially allows a slow time response impurity seeding method, as explored in the Alcator C-Mod experiments, to be an acceptable feedback control system. The nominal position of the radiation front could be maintained ~midway in the divertor leg. Large power transients would be accommodated passively by the divertor. A heat flux transient that is faster than the time response of the impurity seeding system would simply move the front position within the leg. The new front position would then be readjusted back to the ~midway point by the seeding feedback system, on its own, slower time scale. Similarly, active control of gas pressures in the divertor may be used to affect detachment front location feedback control, such as via dynamic control of divertor bypass leakage, similar to a technique used on Alcator C-Mod [24].

Acknowledgements

Thanks are extended to the entire Alcator C-Mod team for their dedication and efforts, which enabled this research. This research was performed at the Alcator C-Mod tokamak, a DOE Office of Science user facility, supported by DOE Contract No. DE-FC02-99ER54512-CMOD. The research by B. Lipschultz was funded in part by both the Wolfson Foundation and UK Royal Society through a Royal Society Wolfson Research Merit Award as well as the RCUK Energy Programme [Grant No. EP/ I501045].

References

[1] B. LaBombard et al., Nucl. Fusion 55 53020 (2015).

[2] D. Brunner et al., Nucl. Fusion 57 86030 (2017).

[3] A. Loarte et al., Phys. Plasmas 18 56105 (2011).

[4] J.W. Hughes et al., Nucl. Fusion 51 83007 (2011).

[5] M.L. Reinke et al., J. Nucl. Mater. 415 S340 (2011).

[6] J.D. Lore et al., Phys. Plasmas 22 56106 (2015).

[7] D. Brunner et al., Rev. Sci. Instrum. 83, 033501 (2012).

[8] D. Brunner et al., Rev. Sci. Instrum. 87, 023504 (2016).

[9] J. Marki et al., J. Nucl. Mater. 365 382 (2007).

[10] J.L. Terry et al., Review of Scientific Instruments 81, 2 (2010).

[11] T. Jiang et al., IEEE Trans. Plasma Sci. 40 788 (2012).

[12] X. Bonnin et al., Nucl. Mater. Energy (in press).

[13] M. Greenwald et al., Phys. Plasmas 6 1943 (1999).

[14] B. Lipschultz et al., Fusion Sci. Technol. 51 369 (2007).

[15] B. Lipschultz et al., Nucl. Fusion 56 56007 (2016).

[16] J.A. Goetz et al., Phys. Plasmas 6 1899 (1999).

[17] M. Bernert et al., Nucl. Mater. Energy (in press).

[18] N. Asakura et al., Nucl. Fusion 49 115010 (2009).

[19] V. Mukhovatov et al., Nucl. Fusion 43 942 (2003).

[20] B. Lipschultz et al., Nucl. Fusion 52 123002 (2012).

[21] M.A. Jaworski et al., Plasma Phys. Control. Fusion 55 124040 (2013).

[22] M.V. Umansky et al., Nucl. Mater. Energy 12 918 (2017).

[23] M.V. Umansky et al., Phys. Plasmas 24 56112 (2017).

[24] C.S. Pitcher et al., Phys. Plasmas 7 1894 (2000).

2018

March 5-9	ITPA MHD, Disruptions & Control Topical Group meeting	Naka, Japan
April 4-6	ITPA Pedestal & Edge Physics Topical Group meeting	Stockholm, Sweden
April 9-11	ITPA Transport & Confinement Topical Group meeting	Daejeon, South Korea
April 9-11	ITPA Integrated Operating Scenarios Topical Group meeting	Daejeon, South Korea
April 16-19	High Temperature Plasma Diagnostics (HTPD) conference	San Diego, CA
April 23-25	Sherwood Theory Conference	Auburn, AL
April 23-26	ITPA Diagnostics Topical Group meeting	San Diego, CA
May 8-11	US Transport Task Force (US-TTF) meeting	San Diego, CA
May 23-25	ITPA Energetic Particles Topical Group meeting	ITER HQ, France
June 17-22	International Conference on Plasma Surface Interactions (PSI)	Princeton, NJ
June 24-28	2018 IEEE International Conference on Plasma Science (ICOPS)	Denver, CO
July 2-6	EPS Conference on Plasma Physics	Prague, Czech Rep.
Sept 11-14	EU Transport Task Force (EU-TTF) meeting	Seville, Spain
October 22-27	IAEA Fusion Energy Conference	Gandhinagar, Gujarat, India
November 5-9	60th Annual Meeting of the APS Division of Plasma Physics	Portland, OR
November 11-15	ANS 23rd Topical Meeting on the Technology of Fusion Energy (TOFE)	Orlando, FL
November 12-18	2nd Asia-Pacific Conference on Plasma Physics	Kanazawa, Japan
December 4-6	ITPA Coordinating Committee & CTP ExComm	ITER HQ, France

2019

JET DT-campaign (https://www.euro-fusion.org/newsletter/jet-full-throttle/)
October 21-25	61st Annual Meeting of the APS Division of Plasma Physics	Fort Lauderdale, Florida, USA

2020

JT60-SA First Plasma (http://jt60sa.org/)

 

This newsletter provides a monthly update on U.S. Burning Plasma Organization activities. The USBPO operates under the auspices of the U.S. Department of Energy, Fusion Energy Sciences (FES) division. All comments, including suggestions for content, may be sent to the Editor. Correspondence may also be submitted through the USBPO Website Feedback Form.

Become a member of the U.S. Burning Plasma Organization by signing up for a topical group.

Editor: Walter Guttenfelder (wgutten@pppl.gov)

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