Final Report of the Committee on a Strategic Plan for U.S. Burning Plasma Research (2019) / Chapter Skim
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Appendix G: Major Research Facilities of the United States and Other Nations
Appendix G: Major Research Facilities of the United States and Other Nations
Pages 221-239
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confinement research facilities; and the larger devices in Europe and Asia. All fusion research experiments with superconducting magnets are located outside the United States: Experimental Advanced Superconducting Tokamak (EAST, located in China)
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• Develop the physics basis for steady-state tokamak operation required for efficient power production; • Contribute substantially to the technical basis for a fusion nuclear science facility; and • Advance the fundamental understanding and predictive capability of fusion science. The DIII-D project commenced in 1986, and its technical capabilities have continually evolved so that DIII-D is presently a flexible device that can study con finement, stability, and divertor physics with a variety of heating and current drive
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This, in turn, allows for the development of the high-performance, advanced tokamak concept, which requires targeted simultaneous control of multiple plasma profiles both in the plasma core and at the edge. Near-term research on DIII-D addresses the development of plasma scenarios scalable to the high f usion gain ITER target.
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It has a maximum operating capacity of 2.2 T toroidal magnetic field and 3 MA plasma current, although it generally operates at lower currents, ≤2 MA. Eighteen field-shaping coils operated by a plasma control sys tem provide great flexibility in plasma shape, discharge evolution, and divertor configuration.
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A model was developed that explains the observed height and width of the pressure "pedestal" at the plasma edge when the tokamak operates in the "H-mode" confinement regime.7,8 Further analysis suggested that, by judicious choice of the plasma shape and discharge evolution, access to a higher pressure "super H-mode" was possible. Subsequent experiments accessed this higher performance regime, and produced plasmas with equivalent QDT of up to 0.6.
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Also because of the low toroidal field, the fast ion population that results from neutral beam injection in STs resides in a parameter space expected for α-heated plasmas at both conventional and low aspect ratio. These unique physics regimes, along with the compact nature of the ST, which leads to stringent requirements for developing
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NSTX had an aspect ratio of R/a = 0.85/0.68~1.25; operated with plasma currents and toroidal magnetic fields of up to 1.5 MA and 0.55 T, respectively; had pulse lengths of up to 1.5 s; and operated in either D+ of He++. NSTX was equipped with a three-source neutral beam capable of injecting 6 MW of D0 power at 90 keV, and up to 6 MW of high harmonic fast wave (HHFW)
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For instance, first-principles gyrokinetic simulations identified the microtearing mode, which is electromagnetic in nature and exists at high-β, as the microinstability respon sible for most of the energy loss from the plasma, which was through the electron channel.15 This mode becomes more stable as collisionality is reduced, consistent with the strong increase of global confinement time with decreasing ollisionality. c Theory development related to the fast ion-driven AE modes led to a deeper under standing of how these instabilities affect both the fast ion and thermal popula tions.16 This understanding led to the development of models of fast ion transport that have been applied successfully at low and conventional aspect ratio.
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NSTX-U is an excellent testbed for simulating α-particle physics applicable to burning plasmas and ITER. Neutral beam-heated NSTX-U plasmas will operate in the largest fast ion dynamic range of parameter space of any ST or conventional aspect ratio tokamaks, and in the regime expected for α-heated plasmas at both low and higher aspect ratio.
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The purpose of LTX-β is to develop the approach to using liquid lithium walls, and to study their effect on plasma performance. LTX used lithium coatings on a high-Z wall, and it exhibited flat electron temperature profiles and enhanced confinement without having the lithium dilute the core plasma or adiate power.22 LTX-β will extend the capabilities of LTX with 700 kW r of neutral beam heating and fueling, 100 kW of electron cyclotron heating/electron Bernstein waves (ECH/EBW)
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at the University of Illinois is a classical stellarator with R = 0.72 m and a = 0.19 m, with magnetic fields up to 0.5 T The main focus of HIDRA is to study plasma-material interactions, including liquid lithium science and technology.31 The Compact Toroidal Hybrid (CTH)
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. FIGURE G.5 Experimental Advanced Superconducting Tokamak (EAST)
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SOURCE: National Fusion Research Institute, Korea, "KSTAR Project," https://www.nfri.re.kr/eng/pageView/74. FIGURE G.7 JT60-SA is a joint international research project, involving Japan and Europe, and under construc tion in Naka, Japan.
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This work is carried out at three U.S. national laboratories (Princeton Plasma Physics Laboratory, PPPL; Oak Ridge National Laboratory, ORNL; and Los Alamos National Laboratory, LANL)
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Half of all articles appearing in Nuclear Fusion since 2016 reporting advancements in fusion simulation involved collaborating international co-authors. In the area of fusion technology and engineering science, the EUROfusion Work Package for Plasma Facing Components pays to use the PISCES-B facility at the University of California, San Diego, helping to identify first-wall materials for ITER and future fusion energy systems.
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al., 2017, Suppression of Alfvén modes on the National Spherical Torus Experiment Upgrade with outboard beam injection, Physical Review Letters 118:265001.
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Rognlien, 2017, Observation of flat electron temperature profiles in the Lithium Tokamak Experiment, Physical Review Letters 119:015001.
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Whyte, 2017, "Perspectives on a Restructured US Fusion Energy Research Program," presentation to the Workshop on U.S. Magnetic Fusion Research Strategic Directions on July 24.
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Team, 2017, Self-consistent modeling of CFETR baseline scenarios for steady-state operation, Plasma Physics and Controlled Fusion 59:075005.
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