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.
Announcements Director’s Corner C.M. Greenfield Research Highlight F. Turco, et al. Schedule of Burning Plasma Events Contact and Contribution Information
Job Posting: Program Manager within the Office of Fusion Energy Sciences
Deadline to Apply: July 10, 2015
The focus of this position will be to serve as a recognized scientific authority and expert in magnetic confinement of high-temperature plasmas and the operation of large toroidal magnetic fusion science experimental facilities and other technical areas, and as such involves the responsibility to plan, coordinate, implement, and evaluate research programs in plasma and fusion science on a national and international level.
The position will also involve planning, budgeting, justifying, and allocating funds among a variety of research programs; preparing analytical documents; giving oral presentations; selecting experts to review proposals; planning, directing, and evaluating complex research programs, projects, and policies; traveling to conferences; and identifying pioneering and strategic research needs.
by C.M. Greenfield
USBPO Task Group on Modes of Research Participation in ITER
This task group has completed its work. The final report, entitled “Recommendations for ITER Experimental Operation, U.S. Team Formation and Participation,” was recently approved by the USBPO Council and is posted at https://burningplasma.org (look for “Recent Posts” on the right side of the page, or follow this direct link, Modes of Participation in ITER).
The report includes two major chapters. The first “Recommendations for ITER Research Program Planning and Experiment Execution,” considers how the ITER research program could best be carried out, from formulation of experimental proposals to execution and analysis of the results. The group considered experience and practices on US experiments as well as ways in which ITER will be different as an international project. The second, “U.S. Team formation and participation in ITER,” considers how the US research team should be formed and managed, and the capabilities which will be needed to effectively carry out our research. Draft conclusions were shared with the USBPO membership in a web seminar last fall, and feedback included in this final report. It is hoped that it will provide useful input, from the perspective of US scientists, to the ITER Organization and to FES in their preparation for the all-important research phase of ITER.
The task group, led by Rajesh Maingi, has very worked hard over an extended period. I would like to thank them for their efforts on behalf of the entire USBPO leadership and the community. Members included M. Greenwald, D. Hillis, A. Hubbard, J. Hughes, S. Kaye, G. McKee, D. Thomas, M. Van Zeeland, and M. Walker, who co-led the first phase of the task.
As you know, the US Burning Plasma Organization is organized into ten topical groups whose leaders are appointed for two-year terms. We are beginning work to identify candidates to fill six positions (shown in red in the table; five leaders’ terms are ending and one deputy leader is stepping down).
The USBPO bylaws state
The Director is responsible for appointing the Topical Group Leader and Deputy Leader(s) for each Topical Group, with the advice of the Council and in consultation with the OFES program manager for the USBPO. The candidates for these positions will be developed on the basis of nominations by the membership of the respective Topical Group, discussion with the Research Committee members, and community input as appropriate. If a Topical Group Leader or Deputy Leader resigns before the normal end of his or her term, the Director will appoint a replacement for the remainder of the term, following the same procedure as for regular appointments. The new Leader or Deputy Leader is eligible to serve a second, full term.
At this point we are looking for nominations (you may nominate yourself) to fill these six positions. We will give preference to candidates who are active (or are willing and able to become active) in the appropriate ITPA topical groups. Also, topical group leaders and deputy leaders make up the USBPO Research Committee, which holds one-hour long videoconferences approximately once every three weeks.
If you would like to make a nomination, please contact either me or the outgoing leader of the appropriate topical group.
This month marks ten years since the selection of the ITER site in Saint-Paul-lez-Durance. Although ITER still has a long way to go before it can achieve QDT = 10, it has already come a long way (see photos, below). The ITER Council met last week to discuss plans for the future, with a revised project plan being prepared under the leadership of new Director General Bernard Bigot. The new plan recognizes the serious accumulated delays, and integrates the scope, cost and schedule for the Project going forward. The plan will be further discussed at the next Council meeting in November 2015. During their meeting, the Council acknowledged actions already taken under the new Director General to strengthen leadership and better integrate the central team and domestic agencies.
|Figure 1: The ITER site in 2007, as site preparation began (left) and in April, 2015 (right; Photos © ITER Organization).|
Integrated Scenarios Topical Group, Leaders: C. Holcomb and F. Poli
This month’s Highlight focuses on the so-called “hybrid” operating mode for tokamaks. Recent work on DIII-D has shown that this is a viable candidate for long pulse or steady-state operation in ITER and other next-step tokamaks.
A Hybrid Scenario to Achieve Fusion in Steady-State in ITER and FNSF
F. Turco1, C.C. Petty2, T.C. Luce2, T.N. Carlstrom2, F. Carpanese3, W. Solomon4, C.T. Holcomb5, J.R. Ferron2
1Columbia University, New York, New York USA
2General Atomics, P.O. Box 85608, San Diego, California 92186-5608 USA
3Politecnico di Milano, Dipartimento di Energia, via Ponzio 34/3, I-20133 Milano, Italy
4Princeton Plasma Physics Laboratory, Princeton, New Jersey USA
5Lawrence Livermore National Laboratory, Livermore, California USA
Correspondence email: email@example.com
New experiments have demonstrated the potential of the steady-state hybrid scenario to become a useful advanced tokamak (AT) regime for ITER and FNSF, with qmin ≥ 1 and projecting to fusion gain Q=5.5 for the ITER 9MA conditions, and Q=3.5 for FNSF operation. Fully-non inductive hybrid operation has been achieved in DIII-D, at 1 MA of plasma current, in MHD stable discharges up to βN=3.7, for 3s at peak performance (~1.5 τR in DIII-D).
The ITER and FNSF steady-state missions require plasmas with long duration and fully non-inductive conditions (fNI=1) at fusion gain Q=5 and Q≤5, respectively. Extrapolation to these conditions from the current scenarios requires demonstration discharges in present machines and validation of the models used for the extrapolation. Since the presently available means for heating and current generation have fairly low efficiency, simultaneously meeting the constraints of high fusion gain and fully-noninductive current drive (CD) entails the maximization of the self-generated bootstrap current fraction (fBS) and good confinement. The magnitude of the bootstrap current is proportional to the plasma pressure, therefore, operation at high-normalized pressure, βN [βN=β(%)·a(m)·BT(T)/IP(MA)], where β is the ratio of the plasma pressure to the magnetic field pressure, a the plasma minor radius, BT the toroidal magnetic field and IP the plasma current), is required to maximize the bootstrap current generation and the performance of the scenario. Finally, in order to achieve fully-noninductive conditions, the sum of the non-inductive sources (external CD and bootstrap) have to align to the total current profile that gives the q-profile desired for the scenario. For this to be stationary, these conditions have to be achieved without Ohmic current (which is inherently transient), therefore, the non-inductive sources, in the final desired condition, have to be positioned in a way that substitutes the Ohmic current over the whole radius. For the high-qmin scenario foreseen for ITER Q=5 operation, this requires a significant part of the current to be moved from the center to mid-radius, which has proven challenging and potentially power expensive.
The hybrid scenario, obtained in several machines [1-3], has the attractive characteristic of a self-organized current profile, which derives from a “flux pumping” mechanism transferring part of the central CD to an off-axis position. This is believed to be caused by the presence of a saturated, benign m3/n=2 or m=4/n=3 (m is the poloidal mode number and n is the toroidal mode number) tearing mode, located at ρ≈0:25-0:4, which only slightly degrades the confinement. For this reason, the issue of aligning the external current sources to the total current does not apply to the hybrid plasmas, which eliminates one of the challenges faced by some candidates for steady-state operation. All the non-inductive current can be driven centrally around ρ~0, where Te and Ti are large, and the current drive efficiency is highest. Despite all the central current drive, the flux pumping mechanism produces a q-profile that remains slightly above 1 for the duration of the discharges. This eliminates the m=1/n=1 internal kink mode, and no core instabilities appear even after several current diffusion times.
|Figure 1: Time histories of 3 repeat hybrid shots, showing the high βN~3:6 and zero loop voltage achieved for ~3 s of βN flattop.|
|Figure 2: Evolution of the surface loop voltage with the value of βP, for all the shots in the SS-Hybrid scenario database.|
|Figure 3: Calculated ideal βN limits (color coded as in the color bar) with varying squareness (x and y axes), for several values of the outer gap (distance from the wall) [stacked surfaces].|
The DIII-D plasmas that reached steady-state conditions were designed using 11 MW of Neutral Beam Injection (NBI) power and 2.5 - 3.3 MW of Electron Cyclotron Heating (ECH) power to sustain a range of βN=3.2-3.7 for the duration of the available heating pulse . A good example of this type of plasmas is illustrated in Fig. 1, where we show the time history of the main parameters for 3 similar fully-noninductive hybrids. The loop voltage (a proxy for the amount of residual Ohmic current) is driven to zero by decreasing the plasma current slightly from 1.1 MA to 1.0 MA and from there by increasing the ECH power by ~800 kW, from 2.4 MW to 3.2 MW. The path to fully noninductive conditions is shown in Fig. 2, where the measured surface loop voltage at the edge is represented against the values of poloidal $beta; (βP) for all discharges in the database. While all these cases are passively stable to ideal and resistive modes, at βN≥4 the discharges are limited by the appearance of 2/1 tearing modes that do not trigger a disruption, but expel ~50% of the stored energy.
Modeling of the new hybrid database indicates that these plasmas are operated at 80-90% of the βN MHD limit with an ideal wall  (the latter defines the upper limit that the pressure is allowed to reach in a toroidal configuration with a perfectly conducting boundary). In order to operate stably at higher βN, one approach is to modify the equilibrium to increase the ideal limits: pushing the ideal limit to higher βN values allows us to have a high βN scenario, while operating in the region where the classical tearing stability index, Δ' varies slowly and has reduced values. A comprehensive modelling study has been performed, to map the ideal stability space around the plasma shape used in the hybrid shots. By modifying the outer section of the shape and decreasing the distance from the outer wall, the ideal with-wall MHD limits can be substantially increased, up to βN~6:5. This is illustrated in Fig. 3, where the βN limit calculated for an experimental equilibrium has been modelled for different values of the upper (y-axis) and lower (x-axis) outer squareness (which measures how close to a square each quadrant of the shape is), for a series of decreasing outer gap shapes (stacked colored surfaces). There is a region of maximum stability for a pair of upper-lower squareness values, for each value of outer gap: these stability maps will be used to guide the next sets of hybrid experiments, aimed at producing discharges with higher ideal limits, and higher stable βN.
In order to determine the feasibility of the hybrid scenario in burning plasma devices, one can follow a 0-D model [5,6], where the input parameters, such as the geometry, the density, BT, IP, are scaled up to the values of the desired machine, and the output parameters are determined self-consistently in the model, which yields the expected confinement, the temperature and the fusion, alpha and current drive power (Pfus, Pα, and PCD, respectively). To cover both the lower-single null shape ITER scenario and the double-null shape FNSF plasmas, we scaled two fully-noninductive hybrid discharges whose characteristics are listed in Table 1. For the ITER Q=5 mission, the present temperature of 5.25 keV (averaged between Te and Ti) scales to Q=5.6, while in the case of FNSF, the Fusion Development Facility (FDF) 6.7 MA proposed scenario , the present double null discharges scale to Q=3.5. This is very encouraging from the point of view of the confinement. However, taking into account the alpha power, which drives minimal current, the power available for current drive is lower than the power necessary to drive all the current (PHeat-Pα
CD in both the ITER and FNSF scalings, Table 1). Besides the solution to reduce the high density fraction, this study illustrates that better current drive systems may be necessary to achieve fully-noninductive conditions in future machines for the operating scenarios discussed here. Moreover, it is important to note that these projections have been performed based on model hybrid discharges with all co-IP current and torque injection. This results in a high external torque of ~8.8 Nm and high toroidal rotation. It is expected that ITER may have much lower rotation, due to its larger volume; therefore, it will be necessary to extend the all-co-NBI steady-state hybrid scenario to low torque conditions, which could project to lower confinement than the present plasmas. These examples show that the hybrid scenario is a very promising alternative scenario that has very good MHD stability and reaches the values of fusion gain required by the ITER and FNSF steady-state missions.
|Table 1: Single parameter scalings for a DN and a SN DIII-D hybrid discharge.|
- T.C. Luce et al., Nucl. Fusion 54, 013015 (2014)
- A.C.C. Sips et al., Plasma Phys. Control. Fusion 44, B69 (2002)
- E. Joffrin et al., Nucl. Fusion 45, 626 (2005)
- F. Turco et al., Phys. Plasmas 22, 056113 (2015)
- V.S. Chan et al., Fus. Sci. Tech. 57, 66 (2010)
- T.C. Luce et al., Plasma Phys. Control. Fusion 50, 043001 (2008)
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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.
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Editor: David Pace (firstname.lastname@example.org)