The Advanced Tokamak

The DIII-D Advanced Tokamak (AT) program in the DIII-D tokamak [J. L. Luxon, Plasma Physics and Controlled Fusion Research, 1986, Vol. I (International Atomic Energy Agency, Vienna, 1987), p. 159] is aimed at developing a scientific basis for steady-state, high-performance operation in future devices. This requires simultaneously achieving 100% noninductive operation with high self-driven bootstrap current fraction and toroidal beta. Recent progress in this area includes demonstration of 100% noninductive conditions with toroidal beta, βT=3.6%, normalized beta, βN=3.5, and confinement factor, H89=2.4 with the plasma current driven completely by bootstrap, neutral beam current drive, and electron cyclotron current drive (ECCD). The equilibrium reconstructions indicate that the noninductive current profile is well aligned, with little inductively driven current remaining anywhere in the plasma. The current balance calculation improved with beam ion redistribution that was supported by recent fast ion diagnostic measurements. The duration of this state is limited by pressure profile evolution, leading to magnetohydrodynamic (MHD) instabilities after about 1s or half of a current relaxation time (τCR). Stationary conditions are maintained in similar discharges (∼90% noninductive), limited only by the 2s duration (1τCR) of the present ECCD systems. By discussing parametric scans in a global parameter and profile databases, the need for low density and high beta are identified to achieve full noninductive operation and good current drive alignment. These experiments achieve the necessary fusion performance and bootstrap fraction to extrapolate to the fusion gain, Q=5 steady-state scenario in the International Thermonuclear Experimental Reactor (ITER) [R. Aymar et al., Fusion Energy Conference on Controlled Fusion and Plasma Physics, Sorrento, Italy (International Atomic Energy Agency, Vienna, 1987), paper IAEA-CN-77/OV-1]. The modeling tools that have been successfully employed to both plan and interpret the experiment are used to plan future DIII-D experiments with higher power and longer pulse ECCD and fast wave and co- and counterneutral beam injection in a pumped double-null configuration. The models predict our ability to control the current and pressure profiles to reach full noninductivity with increased beta, bootstrap fraction, and duration. The same modeling tools are applied to ITER, predicting favorable prospects for the success of the ITER steady-state scenario.

High-confinement (H-mode) operation is the choice for next-step tokamak devices based either on conventional or advanced tokamak physics. This choice, however, comes at a significant cost for both the conventional and advanced tokamaks because of the effects of edge localized modes (ELMs). ELMs can produce significant erosion in the divertor and can affect the beta limit and reduced core transport regions needed for advanced tokamak operation. Experimental results from DIII-D [J. L. Luxon et al., Plasma Physics and Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159] this year have demonstrated a new operating regime, the quiescent H-mode regime, which solves these problems. We have achieved quiescent H-mode operation that is ELM-free and yet has good density and impurity control. In addition, we have demonstrated that an internal transport barrier can be produced and maintained inside the H-mode edge barrier for long periods of time (>3.5 s or >25 energy confinement times τE), yielding a quiescent double barrier regime. By slowly ramping the input power, we have achieved βNH89=7 for up to 5 times the τE of 150 ms. The βNH89 values of 7 substantially exceed the value of 4 routinely achieved in the standard ELMing H mode. The key factors in creating the quiescent H-mode operation are neutral beam injection in the direction opposite to the plasma current (counter injection) plus cryopumping to reduce the density. Density and impurity control in the quiescent H mode is possible because of the presence of an edge magnetohydrodynamic (MHD) oscillation, the edge harmonic oscillation, which enhances the edge particle transport while leaving the energy transport unaffected.

Pellet delivery and survivability through curved guide tubes for fusion fueling and its implications for ITER

A high-repetition and high-energy transversely excited atmospheric carbon dioxide (CO2) laser for a collective Thomson scattering (CTS) diagnostic has been developed to establish a diagnostic method of confined α particles in burning plasmas. To excite a single-transverse and single-longitudinal mode, a continuous wave seed laser was injected to an unstable resonator with a cavity length of ∼4.4m. Pulse energy of 10J with a repetition rate of 10Hz has been achieved in the single-mode operation. Pulse energies of 18J with a repetition rate of 10Hz and 36J with single shot operation have also been achieved in the multimode operation. These results give a prospect for the CTS diagnostic on International Thermonuclear Experimental Reactor (ITER), which requires single-mode energy of 20J with repetition rate of 40Hz.

Conceptual design of the FAST load assembly

An in depth understanding of the collective modes that can be excited in a wide range of high-energy plasmas is necessary to advance nuclear fusion research in parallel with other fields that include space and astrophysics in particular. Important achievements are shown to have resulted from implementing programs based on this reality, maintaining a tight connection with different areas of investigations. This involves the undertaking of a plurality of experimental approaches aimed at understanding the physics of fusion burning plasmas. At present, the most advanced among these is the Ignitor experiment involving international cooperation, that is designed to investigate burning plasma regimes near ignition for the first time.

The Foundation for Research and Technology–Hellas in Greece has named Raymond E. Goldstein, an associate professor of physics and of applied mathematics at the University of Arizona, as the winner of the Pnevmatikos Award in Nonlinear Science for 2000. The foundation recognized “his contributions to the understanding of pattern formation involving the dynamics of filaments, interfaces and surfaces, combining powerful mathematical methods with penetrating physical arguments, numerical computations, and experiments to clarify nonlinear phenomena in a wide variety of physical and biological systems.” The award will be presented next month at an international conference on Crete.At the American Physical Society March meeting in Seattle, Taekjip Ha, an assistant professor of physics at the University of Illinois at Urbana-Champaign, received the 2001 Outstanding Young Researcher Award from the Association of Korean Physicists in America. The association acknowledged Ha for “his truly outstanding scholarly and pioneering research on the biological physics in the field of single-molecule fluorescence.” Dirk Rischke joined the University of Frankfurt in Germany earlier this year as the chair for theoretical heavy-ion physics. He previously was a RIKEN–BNL Research Fellow with the Brookhaven National Laboratory theory group. Rischke succeeds Walter Greiner, who will retire in three years. Until that time, Rischke and Greiner will be working together to ensure a smooth transition.At the annual Institute of Electrical and Electronics Engineers honors ceremony next month in Jersey City, New Jersey, Herwig Kogelnik will receive the IEEE’s highest award, the 2001 IEEE Medal of Honor. Kogelnik is adjunct photonics systems research vice president at Lucent Technologies’ Bell Labs in Holmdel, New Jersey. According to the citation, he is being acknowledged for “fundamental contributions to the science and technology of lasers and optoelectronics, and for leadership in research and development of photonics and lightwave communication systems.”On the occasion of his retirement in March from MIT’s Plasma Science and Fusion Center, Dieter J. Sigmar, senior research scientist, received the US Department of Energy’s Distinguished Associate Award. He was honored, in part, for his “contributions to our understanding of plasma confinement, the physics of burning plasmas, and the role of the plasma edge,” according to the citation, which also noted that his “commitment to international fusion collaboration and stimulation of theory programs around the world has been critical to our progress in fusion science research.” Sigmar plans to continue working on transport theory in tokamak plasmas with his MIT colleagues and following developments worldwide in the physics of burning plasmas. © 2001 American Institute of Physics.

The DIII-D tokamak research program is carried out by General Atomics (GA) for the U.S. Department of Energy (DOE). The DIII-D is the most flexible tokamak in the world. The primary goal of the DIII-D tokamak research program is to provide data to develop a conceptual physics blueprint for a commercially attractive electrical demonstration plant (DEMO) that would open a path to fusion power commercialization. In doing so, the DIII-D program provides physics and technology R&D outputs to aid the Tokamak Physics Experiment (TPX) and the International Thermonuclear Experimental Reactor (ITER). Specific DIII-D objectives include the steady-state sustainment of plasma current as well as demonstrating techniques for microwave heating, divertor heat removal, fuel exhaust and tokamak plasma control. The DIII-D program is addressing these objectives in an integrated fashion with high beta and with good confinement. The long-range plan is organized into two major thrusts; the development of an advanced divertor and the development of advanced tokamak concepts. These two thrusts have a common goal: an improved DEMO reactor with lower cost and smaller size than the present DEMO which can be extrapolated from the conventional ITER operational scenario. In order to prepare for the long-range program, in FY93 the DIII-D research program concentrated on three major areas: Divertor and Boundary Physics, Advanced Tokamak Studies, and Tokamak Physics. The major goals of the Divertor and Boundary Physics studies are the control of impurities, efficient heat removal and understanding the strong role that the edge plasma plays in the global energy confinement of the plasma. The advanced tokamak studies initiated the investigation into new techniques for improving energy confinement, controlling particle fueling and increasing plasma beta. The major goal of the Tokamak Physics Studies is the understanding of energy and particle transport in a reactor relevant plasma.

Collective Thomson scattering (CTS) diagnostic based on a pulsed CO2 laser (wavelength 10.6 μm) has been developed to establish a diagnostic method of confined α-particles in burning plasmas. A high-repetition and high-energy transversely excited atmospheric (TEA) laser has been developed as a source of the CTS diagnostic. In order to obtain single-mode output, which is needed for CTS diagnostic, seed laser is injected into the cavity with unstable resonator. Pulse energy of 17 J with a repetition rate of 15 Hz has been achieved in a single-mode operation. This result gives a prospect for the CTS diagnostic on International Thermonuclear Experimental Reactor (ITER), which requires energy of 20 J with repetition rate of 40 Hz. Proof-of-principle test will be carried out in the JT-60U tokamak by using the newly developed laser. Preliminary consideration of the CTS diagnostic in the tandem mirror GAMMA 10 shows that axial profiles of ion temperature will be obtained using a circumferential collection mirror of scattered power.

A study is presented for the stability of alpha-particle driven shear Alfven Eigenmodes (AE) for the normal parameters of the three major burning plasma proposals, ITER (International Thermonuclear Experimental Reactor), FIRE (Fusion Ignition Research Experiment), and IGNITOR (Ignited Torus). A study of the JET (Joint European Torus) plasma, where fusion alphas were generated in tritium experiments, is also included to attempt experimental validation of the numerical predictions. An analytic assessment of Toroidal AE (TAE) stability is first presented, where the alpha particle beta due to the fusion reaction rate and electron drag is simply and accurately estimated in 7-20 keV plasma temperature regime. In this assessment the hot particle drive is balanced against ion-Landau damping of the background deuterons and electron collision effects and stability boundaries are determined. Then two numerical studies of AE instability are presented. In one the High-n stability code HINST is used . This code is capable of predicting instabilities of low and moderately high frequency Alfven modes. HINST computes the non-perturbative solution of the Alfven eigenmodes including effects of ion finite Larmor radius, orbit width, trapped electrons etc. The stability calculations are repeated using the global code NOVAK. We show that for these tokamaks the spectrum of the least stable AE modes are TAE that appear at medium-/high-n numbers. In HINST TAEs are locally unstable due to the alphas pressure gradient in all the devices under the consideration except IGNITOR. However, NOVAK calculations show that the global mode structure enhances the damping mechanisms and produces stability in all configurations considered here. A serious question remains whether the perturbation theory used in NOVAK overestimates the stability predictions, so that it is premature to conclude that the nominal operation of all three proposals are stable to AEs. In addition NBI ions produce a strong stabilizing effect for JET. However, in ITER the beam energies needed to penetrate to the core must be high so that a diamagnetic drift frequency comparable to that of the alpha particles is produced by the beam ions which induces a destabilizing effect.

Figure 1 shows, schematically, the tokamak device, whose intent is to magnetically confine a high-temperature plasma and produce energy through nuclear fusion. The confined plasma, shown in pink, is toroidal in shape. For several decades now, scientists have been studying and improving these devices, working toward a successful fusion ignition—creating a burning plasma—and sustaining it to generate more power than it took to create it. The next big experimental step will be the International Thermonuclear Experimental Reactor, a twenty billion dollar burning plasma device under construction in France, involving the partnership of seven governments. The human in Figure 2 illustrates the scale of ITER. The magnetic field in the tokamak confines the plasma—ions, electrons, and their heat and momentum. Various kinds of instabilities in the plasma work against that confinement. Turbulent fluctuations can cause transport of particles and energy across the magnetic field lines, toward the outside of the plasma, where it is lost. Loss of plasma and energy, of course, works against confinement and successful fusion. This microturbulence is something the fusion community needs to understand and control. In past and present tokamaks, measurements have shown that the transport of energy and particles caused by microturbulence depends on the size of the tokamak—larger means more transport. This is Bohm scaling. However, theoretical arguments predict that beyond a certain size, which ITER will be beyond, the size dependence goes away, leading to relatively smaller turbulent losses and thus better confinement. This is GyroBohm scaling. The key target of this ESP project is to simulate microturbulent transport for devices of different sizes up through ITER size, and validate and understand GyroBohm scaling.

The next major step in magnetic fusion studies will be the construction of a burning plasma (BP) experiment where the goals will be to achieve and understand the plasma behavior with the internal heating provided by the fusion-generated alpha- particles (α-particles). Two devices with these physics goals have been proposed, the International Thermonuclear Experimental Reactor (ITER)1 and the Fusion Ignition Research Experiment (FIRE)2. Extensive conceptual design work for the instrumentation to try to meet the physics demands has been done for these devices, especially ITER3, and the overall requirements are reviewed in this book.KeywordsGood Spatial ResolutionInternational Thermonuclear Experimental ReactorNeutral BeamTokamak PlasmaBurning PlasmaThese keywords were added by machine and not by the authors. This process is experimental and the keywords may be updated as the learning algorithm improves.

The DIII-D tokamak research program is managed by General Atomics (GA) for the US Department of Energy (DOE). Major program participants include GA, Lawrence Livermore National Laboratory (LLNL), Oak Ridge National Laboratory (ORNL), and the University of California together with several other national laboratories and universities. The DIII-D is a moderate sized tokamak with great flexibility and extremely capable subsystems. The primary goal of the DIII-D tokamak research program is to provide data for development of a conceptual physics blueprint for a commercially attractive fusion power plant. In so doing, the DIII-D program provides physics and technology R&D output to aid the International Thermonuclear Experimental Reactor (ITER) and the Princeton Tokamak Physics Experiment (TPX) projects. Specific DIII-D objectives include the achievement of steady-state plasma current as well as the demonstration of techniques for radio frequency heating, divertor heat removal, particle exhaust and tokamak plasma control. The DIII-D program is addressing these objectives in an integrated fashion in plasmas with high beta and with high confinement. The long-range plan is organized with two principal elements, the development of an advanced divertor and the development of advanced tokamak concepts. These two elements have a common goal: an improved demonstration reactor (DEMO) with lower cost and smaller size than present DEMO concepts. In order to prepare for this long-range development, in FY94 the DIII-D research program concentrated on three major areas: Divertor and Boundary Physics, Advanced Tokamak studies, and Tokamak Physics.

This paper describes the conceptual design of a midsize tokamak as a fast track to the investigation of burning plasmas. It is shown that it could reach large values of energy gain (≥ 10) with only a modest improvement in confinement over the scaling that was used for designing the International Thermonuclear Experimental Reactor (ITER). This can be achieved by operating in a low plasma recycling regime that experiments indicate can lead to improved plasma confinement. The possibility of reaching the necessary conditions of low recycling using a different magnetic divertor from those currently employed in present experiments is discussed.

On October 24, 2021, the Chinese government unveiled a guiding document about state-level work to achieve carbon peaking and carbon neutrality goals under a new development philosophy, laying out key targets and measures for upcoming decades. The documment, titled “Working Guidance For Carbon Dioxide Peaking and Carbon Neutrality In Fully And Faithful Implementation Of The New Development Philosophy.” stated that China's carbon dioxide emissions per unit GDP would have dropped by 18% from the level in 2020 by 2025 and by more than 68% from the level in 2005 by 2030. Fusion energy, as an environment-friendly and intrinsically safe energy source with an abundant fuel supply, is considered an effective low-carbon ultimate solution. Fusion research has been conducted for 60 years, and has been demonstrated that magnetic confinement controlled fusion is very effective magnetic confinement fusion research has progressed significantly over the years. Generally, it is believed in the magnetic confinement fusion plasma community that the next pivotal step in magnetic confinement fusion plasma research is to create burning plasma, where the dominant method of heating plasmas is alpha particle self-sustaining heating. The research aims to understand the underlying physics of the confinement, heating, and instability of burning plasma and to explore the technologies associated with the power-producing fusion reactor. In future burning plasmas, multiple mode-number instabilities are very important but complicated physics issues because of the employment of strong power auxiliary heating schemes. Compared with the time-consuming large-scale computer simulation, the machine learning can bring an advantage for the investigations concerning multiple mode-number instabilities, especially in predictions compared to the time-consuming large-scale computer simulation.