This chapter describes progress in burning plasma science since the United States joined the International Thermonuclear Experimental Reactor (ITER) partnership by way of international agreement.1 Since then, experiments using research facilities in the United States and in other nations have been highly productive. New ideas to control and sustain a burning plasma have been discovered, and theoretical and computational models developed in the United States have substantially improved the ability to control plasma stability, predict plasma confinement, and enhance fusion energy performance. Methods to control and mitigate transients and scenarios that will guide operation of ITER have been successfully tested. Confidence that ITER will succeed in achieving its scientific mission has substantially improved, and progress in burning plasma science and technology has motivated research beyond what will be achieved in ITER toward fusion electricity.
The following first describes progress in burning plasma science and technology that has resulted from the international effort to prepare for ITER operation. This progress combined with ITER construction progress (Figure 2.1) demonstrates a high level of readiness to undertake burning plasma experiments and the expectation that the burning plasma regime can be accessed for detailed study. Next, this chapter discusses the scientific and technical readiness to undertake research beyond ITER and address the potential for economical fusion electricity. Finally, this chapter concludes with a summary of three findings about progress and the further developments needed to progress beyond ITER toward fusion electricity.
In its 2004 report,2Burning Plasma: Bringing a Star to Earth, the Burning Plasma Assessment Committee of the National Research Council concluded that the global fusion science community is both scientifically and technically ready for proceeding with a burning plasma experiment. Scientific readiness was determined from empirical confinement predictions, knowledge of operational limits set by plasma stability, methods to mitigate abnormal events like plasma current disruptions, the ability to maintain plasma purity, methods to measure and characterize a burning plasma, and techniques to control a burning plasma. Technical readiness was determined by successful prototyping of ITER components, evidence of adequate component lifetime in a nuclear environment, tests of plasma-facing components and materials, initial analysis of the safe control of tritium, demonstrations of remote maintenance systems, and demonstration of the required fueling, heating, and current drive control.
Since the 2004 assessment, the ITER design was reviewed and updated in 2008.3 The updated ITER physics basis4 reflected progress from major experiments around the world. Scientists from the United States and from other nations significantly advanced the basic understanding of the processes expected in burning plasma, tested scenarios that will be used to study burning plasma, and achieved significant progress toward construction of the ITER facility through international cooperation. Experiments conducted using research facilities in the United States have discovered new ideas to control a burning plasma that can be tested in ITER. New tokomak experiments built with superconducting magnets in China and the Republic of Korea have demonstrated long-pulse plasma with high confinement
properties. U.S. researchers have used the National Spherical Torus Experiment (NSTX) to test innovative divertor configurations and understand plasma rotation and kinetic effects on the stabilization of instabilities. Using the DIII-D National Fusion Facility, fusion scientists have discovered and tested techniques to control edge localized modes, test disruption mitigation schemes for use on ITER, and experimentally verified computational models to help predict ITER operation. Record plasma pressure was achieved in the Alcator C-Mod experiment at the Massachusetts Institute of Technology. Experiments using Alcator C-Mod, the JET (Joint European Torus), and ASDEX-U (Axially Symmetric Divertor Experiment Upgrade) devices in Europe demonstrated use of a metallic first-wall in preparation for similar experiments in ITER. Theoretical and computational models developed in the United States have substantially improved the ability to control plasma stability, predict plasma confinement, and enhance fusion energy performance.
The ITER Organization and a team of international scientists developed the ITER Research Plan (IRP) in 2013.5 This plan was recently revised with the updated ITER schedule.6 The IRP presents a program logic for a sequence of plasma operations with hydrogen, helium, deuterium, and ultimately deuterium-tritium mixtures for fusion power production. Through dedicated experiments, research, and study by the international community, all of the science and technical criteria that established the readiness for burning plasma experiments in 2004 have progressed greatly.
In the next sections, scientific readiness for ITER operation is presented, including these main areas of research: (1) plasma confinement predictions, (2) plasma stability and operational boundaries, (3) energetic particle physics, (4) mitigations of transients and abnormal, events, and (5) technical progress for ITER construction and technology.
The understanding and optimization of plasma transport, and the ability to predict the plasma response to both collisional and turbulence-induced transport, have advanced considerably beyond the stage when only empirical scaling laws were employed to predict the performance of ITER. Physics-based optimization of confinement performance in both the core and edge of magnetically confined plasmas is one of the critical areas in present day burning plasma research.
A strength of the U.S. theory and simulation program is its close connection to experimental studies, which fosters the validation of the simulation tools and theory to existing observations. Examples include (1) multi-scale predictions of turbulent transport,7,8 which quantitatively match observations in experiments,9 (2) the impact of turbulence on neoclassical tearing modes (NTMs),10,11 and (3) the development of high-gain scenarios for ITER and other planned devices.12
While extensive validation against experimental databases of reduced models is still required, examples of the success of this reduced model development is seen in Figure 2.2, which shows the comparison between reduced model predictions and experimental observations for the plasma core (Figure 2.2a) and plasma edge (Figure 2.2b). These two models have been coupled to produce the core-edge optimized scenarios for burning plasma studies.13 The resulting agreement between experiment and theory has been determined through experimental observations of turbulence characteristics as well as through the expanded capabilities of comprehensive, physics-based numerical simulations.
An example of successful core-edge plasma optimization is to improve the performance of a burning plasma. This achievement was made possible by the recently developed capability to use physics-based core-edge coupled numerical simulations to predict plasma characteristics across the entire plasma. The predicted scenario was achieved through strong plasma shaping and by carefully controlling the trajectory of the plasma density to open access to a regime with stable, high density and pressure plasma edge.14,15 This optimization has led to the achievement of record pedestal pressures that are within approximately 10 percent of the ITER target16 and also led to equivalent fusion power gain, defined as Q, with deuterium-tritium (D-T) equivalent Q over 0.5 achieved on DIII-D.17 Another optimization led to another world record for the average plasma pressure achieved for any fusion device, as indicated in Figure 2.3. While the ITER target goal of
Q = 10 can be achieved with the expected performance in conventional operation, applying these optimization techniques to ITER plasmas can potentially enhance performance beyond the Q = 10 range.
Research has led to the understanding of zonal flows in plasma turbulence,18 and the role of single and multi-spatial scale effects in driving transport.19,20,21 Research has shed light on the structure and formation of the H-mode pedestal.22,23,24 In plasma stability, an understanding of the effects of and plasma response to applied
three-dimensional (3D) magnetic perturbations,25,26 as well as of the plasma behavior during disruptions27 and disruption mitigation,28 has been developed. Finally, comprehensive simulations29 have been used to understand processes controlling the scrape-off layer heat flux width, and they project to a more optimistic scenario for heat flux width on ITER than previously thought.30
Magnetohydrodynamic (MHD) instabilities in tokamaks are large-scale perturbations to the plasma that are driven by the plasma pressure and current, that can lead to degradation of plasma performance through enhanced radial transport of plasma and energetic particles. If not controlled, a larger-scale instability may lead to abrupt terminations of the plasma, called “disruptions.” Operational boundaries are defined by those plasma parameters where these instabilities can be avoided or controlled. For the ITER baseline scenario for achieving Q = 10, pressure-driven kink and resistive wall instabilities31 are avoided by operating below plasma pressure limits. However, the ITER baseline scenario is susceptible to NTM that are driven by plasma current density.32,33 Significant progress toward meeting the challenge posed by tearing instabilities has occurred for the ITER baseline stability by way of better understanding of the roles of plasma rotation, plasma collisionality, and the use of various control actuators, including cancelation of unwanted magnetic field errors34 and localized plasma current drive35 and optimized with advanced feedback and search algorithms.36
Active techniques include tailoring the plasma profiles with various heating and current drive schemes, as well as applying 3D magnetic field perturbations at the plasma edge.37 Achieving high plasma βN can also be aided by a judicious choice of plasma shaping and configuration. For instance, in the more “spherically shaped” NSTX plasmas, stable high-βN discharges were produced routinely with the aid of stabilization techniques.38 Stabilization studies in the mid-sized U.S. tokamaks DIII-D and National Spherical Torus Experiment-Upgrade (NSTX-U), along with further development of techniques to stabilize and control the plasma, including development of physics-based models of MHD instability thresholds, are necessary to inform decisions on an optimized shape and configuration for producing steady-state plasmas in a fusion pilot plant.
Steady-state operation with large self-induced plasma currents, called “bootstrap current” requires both high pressure and low plasma current. In order to accomplish this, both passive and active plasma control techniques need to be developed and incorporated into closed-loop feedback algorithms beyond their present state.
The fraction of self-generated “bootstrap” current, fbs ≈ 0.02 q βN A1/2, is related to the plasma safety factor, q, the plasma aspect ratio, A, and the plasma pressure,
characterized by the normalized pressure parameter, βN.39 Shown in Figure 2.4, the fraction of self-generated “bootstrap” current reaches 100 percent when βN ≈ 25/q for an ITER-like aspect ratio A = 3. High bootstrap current fraction is a goal for efficient steady-state operational scenarios that require minimal injected current drive power. The TCV tokamak in Switzerland achieved 100 percent bootstrap current in 2008.40 Passive control techniques use the stabilizing influence of image currents in conductors that are proximate to the plasma.41,42
Burning plasmas will be self-heated by the α-particles (He++) produced from fusion reactions. To date, only Tokamak Fusion Test Reactor and JET operated with significant amounts of tritium to allow study of energetic α-particles. Both devices produced modest levels of fusion power, and the fraction of α-particle heating power to heating power supplied externally was low, between 10 and 15
percent. Burning plasma experiments in ITER are expected to achieve a fraction of α-particle heating power exceeding 60 percent.
Other fusion experiments without significant heating from fusion reactions do not generate an α-particle population; however, the energetic particle (EP) population created from neutral beam heating can serve as a proxy for the α-particles. Some present-day fusion experiments can produce an EP population that resides in the same “phase space” as that for the α-population expected in both ITER and a burning plasma produced in a more spherical configuration (see Figure 2.5). The EP phase-space is characterized by the fraction of energetic particle energy relative to the total plasma energy and by the speed of the EP motion relative to an Alfvén wave.
The dynamics of the EP population are governed by classical collisional processes in MHD-quiescent plasmas. A sufficient EP population, however, can drive instabilities, which can then interact with the EP population and affect performance through loss or redistribution of the energetic particles, leading to reduced plasma heating.
The past decade has seen significant progress in both understanding and mitigating the EP-driven instabilities in both stellarators and tokamaks.43,44 Theory and numerical simulation advances have led to the development of models, validated
by experiment, that predict the existence and characteristics of EP-driven modes, as well as their resulting interactions with the EP population itself.45,46,47,48,49 This has allowed for accurate descriptions of their ability to heat the plasma and drive current. Presently, reduced models of the modes and their interactions are under development, and, along with reduced models of plasma transport and stability, will lead to a comprehensive physics-based capability for developing the operational scenarios necessary for ITER to achieve its Q = 10 target. Furthermore, this model development will leverage off of the success in mitigating EP-driven instabilities through “phase space engineering,” in which neutral beam50 and radio frequency wave51 actuators are used to tailor the EP phase space in order to control the detrimental consequences of EP instabilities without compromising heating and current drive performance. These same actuators are planned for ITER.
While the tools and understanding that have developed over the past decade are sufficient for developing scenarios in which ITER can achieve its Q = 10 goal, only the study of the α-particle population generated by these burning plasmas will help ensure extrapolation from relative short-pulse fusion power production to one where α-particle modes can remain stable in long-pulse, steady-state operation.
Transients, or abnormal events, are phenomena that are short-lived and that release plasma energy at levels that can damage plasma-facing components. These phenomena include major disruptions and more localized plasma edge instabilities called edge localized modes (ELMs).
Major disruptions are due to crossing into unstable operational space in current, density or β, or to technical system failures, and result in most of the plasma and magnetic field energy being released within thousandths of a second. The disruption can also result in a beam of relativistic electrons, which can cause highly localized damage to the reactor inner wall. While present devices are capable of handling these disruption loads, at least an order of magnitude increase in plasma and field energy content is estimated for ITER and future reactors, and therefore it is essential to develop reliable methods to avoid or mitigate disruptions to avoid damaging vessel components, to allow achievement of high fusion power density, and to ensure continuous operation in a reactor.
Significant progress in disruption control has been made on three fronts: prediction, avoidance, and mitigation. Real-time descriptions of conditions that can lead to a disruption have been developed within empirical frameworks,52,53,54 and these have led to predictability at or above the 95 percent level. For early operations in ITER, predictability higher than 80 percent is required, although 98 percent disruption predictability is required at ITER’s full operational capabilities.55 Recent advances have included developing machine learning/neural network
approaches,56,57,58 real-time monitoring of plasma stability59,60 and incorporating real-time stability calculations from reduced, physics-based MHD stability models to warn of an approach to stability limits.
If a disruption cannot be controlled or avoided, it would need to be mitigated, and progress to do this has been made using injection of gas, liquid, and solid, which can cause high enough radiation to decrease the plasma energy content and provide high enough collisionality to inhibit the formation of runaway electron beams.61,62,63 Shattered pellet injection, successfully developed on DIII-D,64 has been adopted as the mitigation technique-of-choice for ITER. Demonstrations of post-disruption runaway electrons using shattered pellet injection and massive gas injection in the DIII-D experiment are shown in Figure 2.6. After injection of pure (100 percent) neon or argon atoms, total dissipation of fully avalanched runaway electron beams is achieved.65 Novel methods, like shell-pellet injection,66 have been demonstrated. Other methods to use radio frequency waves to dissipate runaway electrons have been proposed67 that will be tested on DIII-D and could potentially be employed on ITER.
ELMs are features of H-mode plasmas in which the high edge pressure and pressure gradients cause intermittent edge instabilities that can release significant amounts of energy to the plasma facing components (PFCs). The accompanying impurity influx from these material surfaces can lead to degradation of plasma performance. As with disruptions, present day devices are capable of handling the peak and averaged heat loads from ELMs, but even a 0.3 percent (~1 MJ) loss of thermal energy in ITER can lead to PFC damage.
A variety of mitigation methods have been developed and tested successfully on present-day devices. Three-dimensional edge magnetic fields have been applied
to both mitigate and suppress ELMs68,69,70,71 (see Figure 2.7). Repetitive injection of solid pellets was used to control the frequency of ELMs.72,73,74,75 Deposition of lithium on the walls has been used to reduce the frequency of and ultimately eliminate ELMs.76,77 Additionally, there are naturally occurring ELM-free regimes in which edge instabilities flush out impurities but preserve plasma performance,78,79 and these regimes can be used as a basis for the optimized core-edge coupled operating scenarios in steady-state devices.
In addition to the scientific progress in the preparations for burning plasma operation in ITER, progress in fusion technology and engineering science has been significant. The fabrication of superconducting magnets (Figure 2.8), the ITER cryostat, and vacuum components is in progress. ITER has been licensed as a first-of-a-kind basic nuclear fusion facility. And ITER construction is now more than halfway completed to first plasma. Additionally, since the 2004 Burning Plasma report,80 fusion technology advances have been driven by ITER research needs and by next-step goals to fully enable the fusion energy system. Key contributions from the U.S. fusion technology program in recent years have led to successful progress in blanket research,81 tritium and fuel cycle research,82 fusion safety and environmental aspects,83 remote handling approaches,84 fusion materials science,85 superconducting magnets,86 and fusion energy systems studies.87 These contribu-
tions have resulted from joint international projects in support of ITER and from tasks directed by U.S. researchers. Accomplishments include (but are not limited to) the development of vacuum and gas species management,88,89 tritium fusion fuel cycle systems,90 pellet injection for fueling and disruption mitigation,91 and the manufacture of the ITER central solenoid.92
International research progress preparing for burning plasma study on ITER has also increased the state of readiness to undertake research beyond ITER leading toward the construction of follow-on devices that demonstrate fusion power production and the potential for economical fusion electricity. In Europe, Japan, South Korea, and China, research beyond ITER is directed to develop the interconnected science and technology needed to design and construct a device to demonstrate fusion power. Just as ITER has provided a research focus for international fusion researchers to advance burning plasma science, a strategy for the accompanying research and technology programs needed to progress beyond ITER to a commercial fusion reactor guides national research and innovation programs, helps to engage industrial partners, and sets national priorities.
While temperatures in the core of the magnetically confined plasma can be up to hundreds of millions of degrees Kelvin, the plasma cools as it is transported radially from core to edge through plasma transport, radiation, and other processes. Nevertheless, the edge plasma temperature can still be in the tens of millions of degrees Kelvin range, and is much greater than the temperature that can be withstood by surrounding walls, for which there is an upper limit of ~1,500 to 3,000 K for most solid material to avoid melting.93 Even at lower temperatures, the plasma-material interactions (PMIs) between the hot plasma and solid surfaces can cause drastic changes in the integrity of the wall material, notably through physical and chemical erosion. This can introduce impurities into the plasma, which can degrade plasma performance and fusion gain.94,95,96
There has been considerable progress over the last decade in mitigating heat fluxes escaping the plasma by various means that increase the area over which the heat is deposited. One such method is through plasma “detachment” from the divertor targets, obtained by additional gas fueling in the divertor region, which increases density and lowers the temperature of the plasma in the divertor region primarily through radiation.97,98 Detachment can lead to factors of several decrease in the heat flux deposition without impacting the performance of the core plasma.
While detachment in a conventional divertor alone is estimated to reduce heat loading in ITER to an acceptable level by radiating 60-70 percent of the escaping heat flux, a next-step burning plasma may have heating and fusion powers greater than those expected in ITER. With a conventional divertor, up to 90 percent of the heat exhaust would have to be radiated away to avoid material surface damage; at these levels, core plasma performance could be severely affected.99 Studies of the compatibility of innovative divertor designs with divertor plasma detachment, which can significantly relax the radiated power requirement, are needed.
Innovative divertor designs use more complex magnetic topologies to spread the heat flux over a wider area on the target plates. This is done by increasing the field line length, which allows for additional cross-field transport of the heat, making the angle at which the field lines impinge on the surface shallower, and increasing the broadening of the magnetic field lines “tubes.” Innovative configurations that have shown heat flux reductions on a variety of devices include the Snowflake,100,101,102 the X-divertor, and the Small-Angle Slot divertor.103,104 The Super-X105 divertor is closely related to the X-divertor, and both will be studied on MAST-U106 tokamak. Also related to the Super-X is the X-point target divertor concept.107 These various innovative divertor configurations are shown in Figure 2.9.
Liquid metal (LM) surfaces are a potentially transformative solution to the heat flux challenge for all magnetic configurations, as the damage and/or erosion that can occur in solid PFCs are eliminated since the liquid metal walls are continu-
ally replenished with new, clean surfaces contacting the plasma. Liquid metals are renewable, and they return to equilibrium after perturbations. LMs can handle heat fluxes up to factors of several over the upper limits for solid walls.108 One of the leading candidates for LM walls is lithium, which, when coated on solid walls through evaporation, led to improved confinement109,110 and suppressed or mitigated ELMs.111,112 Liquid lithium surface research and development on tokamaks is in the early stage,113,114 and the challenges involve their design, stability in the presence of magnetic fields, retention of tritium (an issue for both liquid and solid walls),115 and impact on plasma performance.116,117
Among the key challenges for obtaining a steady-state burning plasma beyond ITER is the capability to drive the plasma current noninductively. This is of fundamental importance in tokamaks since tokamak confinement improves with the poloidal magnetic field that is produced by the plasma current. Efficient steady state operation in a tokamak requires operation in plasma regimes that yield significant self-driven (or “bootstrap”) plasma current.118 It is important that the self-driven current profile leads to a scenario that is stable to large-scale MHD instabilities.
The physics of externally produced noninductive current drive (CD) from neutral beam (NB) and radio frequency (RF) methods such as electron cyclotron (EC), lower hybrid (LH), and fast waves (FW) is now well understood and are documented in the ITER physics basis.119,120 Recently, current drive by high frequency “helicon” waves has been proposed.121 NB-CD and EC-CD are most robust and free from accessibility issues. EC-CD is particularly attractive since it can achieve the highest transmission power density, and the space necessary for power injection is smallest. Local current drive by EC-CD can suppress internal MHD modes and the local EC-CD is also effective for local pressure profile modification to further enhance plasma performance. The challenges facing EC-CD include the relatively high cost of sources, the possible need for new source development if high fields are used, and the lower current drive efficiency for off-axis current generation which may be necessary for high bootstrap fraction. In contrast, LH-CD already has readily available sources at lower costs and the off-axis current drive efficiency is perhaps the highest of any of the RF options. On the other hand, LH-CD has important challenges because of the need for a close-to-the-plasma-edge launching structure.
Establishment of a high-bootstrap-current-fraction, fully noninductive tokamak plasma at sufficiently high normalized-beta using reactor-relevant and efficient current drive systems is a critical research subject for developing steady-state scenarios and being able to operate in the reduced tritium breeding blanket space of a compact configuration.
Perhaps the greatest challenge faced in a burning plasma regime is the simultaneous solution to all of the above issues to achieve the overall goal of a fusion power system capable of uninterrupted operation. Measuring progress on multiple design metrics is not trivial. A number of the issues, including self-driven “bootstrap” current, external current drive and plasma heating systems, plasma-material interactions and power handling, and robust control and mitigation of transients, are to some extent measured by the normalized fusion gain (or triple product) and the duration of uninterrupted plasma confinement pulses. (See Figure 4.5 in Chapter 4.)
Burning plasmas must simultaneously achieve a high triple product while eventually being sustained for months of steady-state operation. Until now, short pulse experiments built with copper magnets have operated for several seconds and have achieved conditions equivalent to “scientific breakeven,” or Q ~ 1, when extrapolated to operation with deuterium-tritium fuel. Longer pulse studies using experiments with superconducting magnets have been performed,122,123,124,125 but these research devices are not large enough to operate at the high pressures needed for fusion gain. High-power-density scenarios capable of steady-state should be explored in actual reactor-like conditions, providing scientists the opportunity to study coupled electron-ion turbulence, super-Alfvénic ion distributions, and high opacity plasma edge. The highly nonlinear interactions between different phenomena in fusion equivalent regimes, defined at reactor relevant integration parameters, have not yet been examined.
A critical integration aspect upon reaching fusion equivalent regimes is to reconcile the core and the edge. Because the plasma collisionality parameter, ν* scales strongly with the ratio of plasma density to plasma pressure, ν* ~ n3/P2, the parameters of both the divertor and the plasma core cannot simultaneously operate at reactor relevant parameters unless absolute reactor-relevant plasma pressure, P, is achieved. Because the plasma core and edge strongly interact, the core-edge-divertor interactions are altered by increasing neutral opacity as reactor-like densities are approached. As the fusion equivalent regime is approached, the divertor, scrape-off layer, and pedestal become increasingly opaque, and pedestal profiles become more strongly dependent on transport and pinch effects. Divertor and plasma confinement are linked, and integrated solutions require both regions to be in the relevant density regimes.
Another major issue is integrating the design and operation of tokamaks with requirements and technology imposed by future reactor engineering constraints. Progress toward this integration has been slow since plasma science and device performance are each a necessary first step. As devices approach the burning plasma regime it is appropriate to embrace design choices that are compatible with
a future compact pilot plant. Notably, this includes operating with metal walls to minimize tritium retention, optimization of the geometry/topology of magnets, and measurement/control systems that can operate in a nuclear environment. In addition, the tritium breeding ratio and corresponding blanket design should be compatible with overall plasma performance.
The study of very long pulse, toroidal magnetic confinement is under way with major research facilities using superconducting magnets. EAST (China), KSTAR (Republic of Korea), WEST (formerly Tore-Supra, in France) are superconducting tokamaks, as is also ITER and the JT-60SA experiment under construction in Japan. The two superconducting stellarator facilities are the Large Helical Device (LHD) in Japan and the Wendelstein 7-X (W7-X) device in Germany. Because tokamaks and stellarators have strong magnetic fields and toroidal geometries, the fusion science and technologies of tokamaks and stellarators are similar. The fundamental dynamics of plasma confinement are described with the same methods; scientists produce and diagnose confined plasma using the same technologies; and tokamaks and stellarators are challenged to achieve and sustain the same fusion equivalent conditions.
Stellarators and tokamaks differ by the degree by which the magnetic field is three dimensional.126 Tokamaks, like ITER, use relatively small three-dimensional magnetic field perturbations to control plasma instabilities, such as ELMs, and influence plasma profiles and flows (see Figure 2.7). As the degree of three-dimensional magnetic field increases significantly, the usual confinement properties that arise from symmetry are broken and particle confinement requires carefully designed three-dimensional magnetic fields to avoid rapid particle loss. Careful design is also necessary for the three-dimensional magnetic fields used to control tokamak plasmas,127 and this further motivates ongoing research linking the science of strong three-dimensional magnetic fields optimized for stellarators and the weaker three-dimensional magnetic fields needed to optimize performance of tokamaks, like ITER.128
Understanding transport in stellarators (Figure 2.10) has led to the design of a new class of stellarators that have recently commenced operating. Transport in early stellarators was dominated by collisional, or neoclassical, processes. Neoclassical effects cause the cross-field transport of otherwise trapped particles as well as large impurity fluxes, which can dilute the plasma and compromise performance. However, advances in theory and in numerical tools have led to identifying stellarator configurations with various optimizations of the magnetic field that predict a reduction in neoclassical transport to levels comparable to those in tokamaks.129,130,131 The development of optimization codes have success-
fully designed stellarators with reduced neoclassical transport.132 A major new facility with such optimization, W7-X in Germany,133 recently began operating, and experiments have confirmed the three-dimensional magnetic field optimization.134 A prime goal of W7-X is to test fusion magnetic confinement for one such optimization scheme and the balance between neoclassical and turbulence-driven transport. The experiments will serve to validate results of several gyrokinetic codes, including the first fully global and physically comprehensive turbulence codes, which have recently been developed.135
Some MHD and energetic particles instabilities are predicted to behave differently in stellarators than in tokamaks. Understanding these differences will help to better control a burning plasma and to better predict fusion performance. For example, some stellarator configurations, like that used in W7-X, minimize the toroidal plasma current, and plasma current induced MHD instabilities can potentially be avoided. Other stellarator configurations, like the National Compact Stellarator Experiment,136 are predicted to reduce the toroidal plasma current by more than half an equivalently sized tokamak. Because relatively few stellarators have been built and studied, stellarators provide new tests of our understanding of plasma stability limits. Although disruptions are absent in existing low current stellarators,137 because high-pressure plasma confined within a quasi-symmetric stellarator is predicted to generate significant bootstrap current, current-driven instabilities and plasma current disruptions may need active methods for controlling the plasma profiles and mitigating the effects of transients just as in tokamaks. MHD-stability studies have been ongoing in LHD138 and will be undertaken in the W7-X and Compact Toroidal Hybrid.139
Stellarators also provide opportunities to investigate plasma heat flux issues and to better understand how to design an optimized divertor. Because the stellarator
configuration is intrinsically three dimensional, the design, maintenance, and replacement of a stellarator divertor is more complicated than in a tokamak while also providing an opportunity to validate three-dimensional models of the plasma edge and the plasma interaction with the first wall. There are several types of divertors that are being assessed on stellarators, including the helical divertor140,141 and the island divertor.142,143 Stellarator divertors have not been tested to the extent that they have been tested on tokamaks. Nevertheless, these challenges are common to both stellarators and tokamaks and include understanding the plasma boundary, including the effect of edge magnetic field shear, active control requirements, and assessing compatibility of divertor solutions in a high-performance integrated core-edge coupled regime.144
For many years, the U.S. effort in basic theory, simulation, and modeling of fusion plasmas has been extremely strong, with many U.S. scientists being recognized internationally as world leaders in their respective fields. U.S. fusion researchers have been instrumental in driving recent, and seminal, progress in several diverse areas. Both analytic theory and reduced models and high-fidelity physics simulations development comprise this impressive set of accomplishments.
Theory and simulation offer important new opportunities for accelerating progress toward the objective of economical fusion energy development by incorporating recent advances in theoretical understanding, validated physics models, computing infrastructure, and diagnosis of experiments.145,146 As described in the Report of the 2015 Workshop on Integrated Simulations for Magnetic Fusion Energy Sciences,147 what is needed is to comprehensively and self-consistently advance the many complex, nonlinear, and multi-scale plasma descriptions into an integrated suite of whole device modeling (WDM) capabilities.148 The long-term goal of the project is to have a complete and comprehensive application that will include all the important physics components required to simulate a full toroidal discharge. Such a predictive modeling capability is required—for example, for the interpretive analysis of existing devices as well as for minimizing risks and qualifying operating scenarios for next-step burning plasma experiments. In all future burning plasma facilities, the optimization of fusion performance and control scenarios will require predictive WDM with a quantified, validated uncertainty, as it will not be feasible to determine operational limits by running trial discharges.
The following research elements are critical for realization of a successful WDM capability:
- Continued efforts to better understand and distill the physics of gap areas in fusion theory. These gaps include understanding transients, the plasma
boundary, fully noninductive operation, and optimization of the toroidal magnetic configuration.
- Increased development of and support for modular WDM frameworks. A sustainable path forward will require support both for the most mission-critical legacy tools and for development and expansion of the newer efforts that can more effectively utilize leadership-class computing resources and execute next-generation work.
- Increased connection to experiment through validation. This will require the development and extensive use of tools that fulfill validation hierarchies and compute associated metrics, as well as an accurate and comprehensive set of diagnostics. Such an approach will require expertise in large-scale data management and analysis, including machine-learning strategies, for both leadership-class code output and the experimental observations against which they will be tested.
While present models can help develop operational scenarios for ITER and increase confidence in achieving its goals, a validated WDM capability would provide the confidence to explore the extreme parameter regimes of fusion reactors, informing decisions and lowering the future risks of building full-scale fusion power plants, and thus accelerating the fusion program.
In addition to projects in support of ITER and domestic Department of Energy (DOE) Office of Fusion Energy Sciences research activities, the fusion program can leverage technology developments in areas that have significant potential to advance fusion energy development. In order to capture the impact of such advances and identify ground breaking research opportunities for advancing fusion energy, DOE’s Fusion Energy Sciences Advisory Committee (FESAC) was recently charged “to identify the most promising transformative enabling capabilities (TEC) for the U.S. to pursue that could promote efficient advance toward fusion energy,” The 2018 FESAC report149 identified four areas of transformative enabling capabilities where, building from significant progress in certain areas of research, the United States has a strategic opportunity to develop transformative technologies to enable fusion energy. The four top-tier TECs identified by the panel were: advanced algorithms, high critical temperature superconductors, advanced materials, and novel technologies for tritium fuel cycle control.
Advanced control systems are required to control a burning plasma during the normal evolution of the plasma and also during transient events that need to be either predicted and avoided or mitigated. Advancements in feedback algorithms and intelligent systems may significantly improve the reliability of fusion power and perhaps enable operation at optimized operating points whose achievement and sustainment are impossible without high performance feedback control. As explained in the TEC Report, the area of advanced algorithms includes the related fields of mathematical control, machine learning, artificial intelligence, integrated data analysis, and other algorithm-based research and development. The importance of advanced algorithms is illustrated in Figure 2.11, which shows the variables and control functions in a fusion plasma “control operating map” and a chart illustrating the control function interactions for ITER’s disruption mitigation system. As understanding of burning plasma operation improves, advanced control algorithms will support and accelerate the pace of physics understanding, enable the experimental realization of theoretically predicted operating scenarios, and build mathematical synergies with advances in other high-performance computing capabilities that will enable improved physics understanding. Machine learning and mathematical control can also help to bridge gaps in knowledge when these exist, for example to enable effective control of fusion plasmas with imperfect understanding of the plasma state.150
Advances in higher field superconductors present a major opportunity to enhance the performance and feasibility of fusion reactor designs. High-field, high-temperature superconductors would enable a new generation of relatively compact fusion experiments and power plants, dramatically speeding the development path and improving the overall attractiveness of fusion energy.151 High-temperature superconductor technology could provide high energy gain and power density in much smaller devices, together with operational robustness and steady state potential needed for successful fusion energy. Strong magnetic fields are critical to the success of magnetic fusion as a source of energy. Achieving higher magnetic field strength extends the allowable plasma properties to higher plasma density, higher plasma current, and higher plasma pressure while retaining the same dimensionless scaling parameters found at lower magnetic field strength. This extended range of plasma parameters from high-field magnets allows more compact tokamak devices that may provide a lower cost path to future fusion reactors. ITER’s superconducting magnet system will be the largest ever made and is designed to operate with the highest practical magnetic field strength for large toroidal field
coils made of niobium-tin superconductors and consistent with the strength of steel.152 New developments of rare-earth barium-copper-oxide high-temperature superconductors (Figure 2.12) may lead to larger magnetic field strength153 and improved maintenance that potentially improve the prospects for economical magnetic fusion energy.154
The behavior and integrity of materials in a fusion system are of great importance to the long-term viability of fusion energy.155 The flux of energetic neutrons to the vessel and structural materials poses a serious materials problem that will require substantial testing, some of which may be done on a burning plasma experiment. The high energy neutrons from the D-T fusion reaction generate between 50 to 100 times higher helium to dpa ratio in materials such as ferritic steels than does fission reactor irradiation. Burning plasma experiments will thus aid in the development of high-heat-flux components and will serve as testbeds in which to evaluate the performance of the components in a reactor-like fusion environment.
The heat loads on components in a burning plasma experiment will be comparable to those expected in a reactor and will require the application of state-
of-the-art high-heat-flux technology using materials that satisfy requirements of tritium retention, safety, structural integrity, lifetime, and plasma compatibility.156
Since the 2004 Burning Plasma Assessment report, the United States has made significant advances in fusion materials studies, including contributing to the qualification of reduced activation ferritic martensitic steels for the European demonstration fusion reactor,157 nanostructured158 and oxide dispersed strengthened steels,159 all aspects of SiC/SiC technology,160 and new understanding of tungsten161 and tungsten composites162 as fusion plasma-facing materials. Linear plasma simulators allow for long-duration study of material evolution under fusion-relevant plasma flux, but they are not useful to test integrated plasma-material effects expected in fusion divertors. In the United States, linear plasma simulators include the PISCES facility at University of California, San Diego,163 the Tritium Plasma Experiment at Idaho National Laboratory (INL), and the Material Plasma Exposure Experiment being developed at Oak Ridge National Laboratory (ONRL).164
Qualification of materials to operate safely in a fusion environment is of critical importance for moving forward. Material effects have to be strongly connected to any consideration of availability, reliability, and maintainability of the fusion core. All of these have a large impact on the three overarching factors of plant economics, public perception, and ability to license.
Looking toward future opportunities, new material designs and processes will enable the realization of resilient components that are essential to survive the harsh fusion environment and to optimize the reactor’s performance. As reported by the 2018 U.S. DOE FESAC Committee on Transformative Enabling Capabilities, “Advances in novel synthesis, manufacturing and materials design are providing for some of the most promising transformation enabling technologies in PMI and nuclear fusion materials to enable fusion energy for the future.”165 The novel features enabled by advanced manufacturing and additive manufacturing include complex geometries and transitional structures, often with materials or constituents including hard-to-machine refractory metals; the potential for local control of material microstructure; rapid design-build-test iteration cycles; and exploration of materials and structures for containing and delivering liquid metals. The DOE’s first Manufacturing Demonstration Facility166 is located at ORNL and is developing new manufacturing technologies beneficial for fusion energy technologies. These include advanced metals manufacturing: electron beam melting, laser-blown metal powder deposition, laser sintering, and metal laser melting. Figure 2.13 shows an example of selective laser melting for complex manufacturing of enhanced tungsten heat exchangers. With these emerging techniques, important new opportunities are emerging to develop the resilient materials and components for a fusion energy system.
A fusion breeding blanket—that is, a nuclear system that creates tritium via interaction of the fusion-produced 14-MeV neutrons with lithium—is a key fusion nuclear technology needed for the development of fusion energy. Fusion reactors should operate with more tritium produced and recovered than is burned. The vast majority of the fuel injected in a fusion chamber will not be burned in a single pass. Unburned deuterium-tritium fuel will be continuously transported to the plasma edge, where it will be exhausted, stripped of impurities, and then reinjected into the plasma. A burning plasma experiment provides the opportunity to test and evaluate the performance of prototypical blanket modules and demonstrate technologies for tritium extraction from blankets and for fuel processing systems that can be operated efficiently at large scale.167,168,169
Some recent successful activities in the United States include construction and operation of the MaPLE (Magnetohydrodynamic PbLi Experiment) facility170
international blanket research partnerships (e.g., Korean National Fusion Research Institute-University of California, Los Angeles (UCLA)-INL collaboration framework and UCLA-EUROfusion collaboration), and innovative ideas in the tritium fuel cycle.171,172 In particular, significant progress has been made in experiments and massively parallel simulations to understand magneto-hydrodynamic flows of liquid metal, self-cooled, dual-coolant, and helium-cooled lead-lithium blanket concepts.
It is well recognized that the development of a fusion breeding blanket is an outstanding challenge for fusion because scientific gaps exist related to controlling tritium permeation and minimizing tritium inventory.
Although the 2004 Burning Plasma Assessment Committee report noted the central role of the ITER Test Blanket Module (TBM) in the U.S. fusion research program, the United States is not a partner in the ITER TBM Project, nor will the decision to incorporate TBMs on ITER be made until much later in the ITER construction schedule. The signing of all of the TBM arrangements occurred in 2015; however, TBM preliminary designs are not expected until 2023. The final
TBM designs are not expected until 2025, and the first installation of the TBMs will occur after 2030. Because the decision to install TBMs on ITER will not be made until after additional legal arrangements are signed dealing with all TBM phases through decommissioning,173 the careful consideration for the best approach to advance fusion tritium blanket technology could include becoming a supporting partner with a TBM activity along with other approaches.
The 2018 FESAC report recognizes this challenge, and identifies several opportunities to develop technologies with potential to address existing gaps, such as novel blanket technologies for tritium breeding that allow for higher thermal to electrical efficiencies and improved tritium breeding ratios,174,175 advanced tritium extraction technologies,176 and new fuel recycling technologies that allow for minimization of tritium inventories.177
A burning plasma experiment offers the opportunity to begin development of the technologies needed for a fusion reactor, including important safety-related technologies. Many components and systems needed for fusion’s safety objectives are unique, such as source diagnostics and cleaning technologies, state-of-the-art safety analyses tools, technologies for the remote handling of large activated components, technologies for the control of routine tritium releases, and innovative approaches for the control of tritiated and mixed waste streams.178 A burning plasma experiment will be an integrated demonstration of the safety, reliability, and effectiveness of these technologies.179 In the United States, recent progress has been remarkable in the areas of safety code development and understanding of tritium behavior in fusion systems.
INL’s Safety and Tritium Applied Research (STAR) facility is a worldwide unique facility capable of handling tritium and radioactive materials, as well as other controlled chemical elements such as beryllium and lead.180,181 This facility provides opportunities for researchers to investigate the synergistic effects of hydrogen isotopes in neutron irradiated fusion materials at burning plasma operating conditions. Experiments at the STAR Facility include the Tritium Plasma Experiment (Figure 2.14), the Tritium Gas Absorption Permeation Experiment, the Neutron Irradiated Material Ion Implantation Experiment, and the Experimental Chamber for Evaluation of Exploding Dust. Recent progress has allowed researchers to expand the understanding of tritium permeation and retention behavior in tungsten at prototypical fusion conditions. Not only have the number and types of traps been measured, but also how these crucial safety parameters vary with neutron irradiation history and temperature. In addition, excellent progress has been made in the area of safety code development. The MELCOR for fusion and Tritium Migration Analysis Program codes182 developed at INL, have been used for safety analyses in
licensing ITER and design studies for many national and international future fusion reactors. These computer codes are in use at more than 30 fusion institutions worldwide and have recently been merged allowing for tritium tracking within a fusion facility during normal operation and accident conditions.
Integrated systems studies guide research and identify programs that can reduce cost and lower risk to the development of fusion power. Integrated systems studies combine burning plasma science, materials science, fusion nuclear science, and systems engineering to evaluate safety, environmental and maintainability issues, and technical requirements to progress toward fusion energy. Recent progress in systems engineering studies in the United States and by our international partners
has had an enormous impact on the direction of research to address fusion economics, public perception, and regulatory framework.
Systems engineering combines plasma physics and engineering constraints into a self-consistent integrated design for large-scale fusion facilities. Systems engineering studies have been carried out for various types of fusion reactors, including the advanced tokamak,183 high-field tokamak,184 and the spherical tokamak and stellarator.185 The United States has made significant progress in the area of fusion nuclear systems study,186 leading to the definition of requirements for a Fusion Nuclear Science Facility187 for integrated testing of fusion components. Figure 2.15 illustrates several examples of integrated systems studies, ranging from a pilot plant that would generate 73 MW of net electricity to a large commercial demonstration facility generating up to 300 MW of net electricity, operating without interruption,
and designed with sufficient cooling for 1.5 GW of fusion power production. These systems engineering and integrated fusion systems studies are the primary means to combine the knowledge obtained from the study of burning plasma with the interdisciplinary sciences needed to define the most optimal path toward fusion energy demonstration.
Since the United States joined the ITER partnership, experiments using research facilities in the United States and in other nations have been highly productive. New ideas to control and sustain a burning plasma have been discovered, and theoretical and computational models developed in the United States have substantially improved the ability to control plasma stability, predict plasma confinement, and enhance fusion energy performance. Methods to control and mitigate transients and scenarios that will guide operation of ITER have been successfully tested. The understanding of burning plasma science has advanced significantly.
Finding: The U.S. fusion energy science program as part of the international research effort has made leading advances in burning plasma science and technology that have substantially improved our confidence that a burning plasma experiment such as ITER will succeed in achieving its scientific mission.
Although the primary focus of the world’s fusion research program is the preparation for ITER experiments, progress has also resulted in the research aimed beyond ITER to address remaining science and technology challenges and demonstrate innovative solutions that lead to a reduced size, lower cost, full-scale power source. While this research is much less developed than the science and technology required for a burning plasma experiment, opportunities exist to increase the readiness to benefit from ITER experiments and progress toward the demonstration of economical fusion power. New technologies, such as high-field superconducting magnets, may reduce the size of next-step fusion demonstration experiments. Advances in understanding and predictive models also suggest that opportunities exist to pursue more compact follow-on experiments.
Finding: In addition to burning plasma studies, further research in burning plasma science, fusion nuclear science, and fusion materials science is needed to reduce the cost and fully enable the fusion power system.
Finding: New technologies for fusion, advances in understanding and predictive modeling, the improved confidence in the science and operation of ITER, and the engineering systems studies conducted both within the United States
and by our international partners demonstrate a readiness to undertake the research leading to a cost-effective next step toward the realization of commercial fusion energy.
2. NRC, 2004, Burning Plasma.
3. R. Hawryluk, 2009, Principal physics developments evaluated in the ITER design review, Nuclear Fusion 49:065012.
4. M. Shimada, D.J. Campbell, V. Mukhovatov, M. Fujiwara, N. Kirneva, K. Lackner, M. Nagami, et al. 2007, Chapter 1: Overview and summary, Nuclear Fusion 47:S1, http://dx.doi.org/10.1088/0029-5515/47/6/S01.
5. A.C.C. Sipps, G. Giruzzi, S. Ide, C. Kessel, T.C. Luce, J.A. Snipes, J.K. Stober, and the Integrated Operation Scenario Topical Group of the ITPA, 2015, Progress in preparing scenarios for operation of the International Thermonuclear Experimental Reactor, Physics of Plasmas 22:021804.
6. ITER Organization, 2018, ITER Research Plan within the Staged Approach, ITR-18-003, September 17.
7. T. Goerler and F. Jenko, 2010, Scale separation between electron and ion thermal transport, Physical Review Letters 100:185002.
8. S. Maeyama, Y. Idomura, T.-H. Watanabe, M. Nakata, M. Yagi, N. Miyato, A. Ishizawa, and M. Nunami, 2015, Cross-scale interactions between electron and ion scale turbulence in a tokamak plasma, Physical Review Letters 114:255002.
9. N.T. Howard, C. Holland, A. White, M. Greenwald, and J. Candy, 2015, Multi-scale gyrokinetic simulation of tokamak plasmas: Enhanced heat loss due to cross-scale coupling of plasma turbulence, Nuclear Fusion 56:014004.
10. L. Bardóczi, T.A. Carter, R.J. La Haye, T.L. Rhodes, and G.R. McKee, 2017, Impact of neoclassical tearing mode–turbulence multi-scale interaction in global confinement degradation and magnetic island stability, Physics of Plasmas 24:122503.
11. A. Bañón Navarro, L. Bardóczi, T.A .Carter, F. Jenko, and T.L. Rhodes, 2017, Effect of magnetic islands on profiles, flows, turbulence and transport in nonlinear gyrokinetic simulations, Plasma Physics and Controlled Fusion 59:034004.
12. M. Kotschenreuther, D.R. Hatch, S. Mahajan, P. Valanju, L. Zheng, and X. Liu, 2017, Pedestal transport in H-mode plasmas for fusion gain, Nuclear Fusion 57:064001.
13. O. Meneghini, P.B. Snyder, S.P. Smith, J. Candy, G.M. Staebler, E.A. Belli, L.L. Lao, J.M. Park, D.L. Green, W. Elwasif, B.A. Grierson, and C. Holland, 2016, Integrated fusion simulation with self-consistent core-pedestal coupling, Physics of Plasmas 23:042507.
14. W.M. Solomon, B. Snyder, K.H. Burrell, M.E. Fenstermacher, A.M. Garofalo, B.A. Grierson, A. Loarte, G.R. McKee, R. Nazikian, and T.H. Osborne, 2014, Access to a new plasma edge state with high density and pressures using the quiescent H-mode, Physical Review Letters 113:135001.
15. P.B. Snyder, W.M. Solomon, K.H. Burrell, A.M. Garofalo, B.A. Grierson, R.J. Groebner, A.W. Leonard, R. Nazikian, T.H. Osborne, E.A. Belli, J. Candy, and H.R. Wilson, 2015, Super H-mode: Theoretical prediction and initial observations of a new high-performance regime for tokamak operation, Nuclear Fusion 55:83026.
16. J.W. Hughes, P.B. Snyder, M.L. Reinke, B. LaBombard, S. Mordijck, S. Scott, E. Tolman, et al., 2018, Access to pedestal pressure relevant to burning plasmas on the high magnetic field tokamak Alcator C-Mod, Nuclear Fusion 58:112003.
17. P.B. Snyder, J.W. Hughes, T.H. Osborne, C. Paz-Soldan, W. Solomon, D. Eldon, T. Evans, et al., 2018, “High Fusion Performance in Super H-Mode Experiments on Alcator C-Mod and DIII-D,” presented at the 2018 IAEA Fusion Energy Conference, October 22-27.
18. P.H. Diamond, S.-I. Itoh, K. Itoh, and T. S. Hahm, 2005, Zonal flows in plasma—A review, Plasma Physics and Controlled Fusion 47:R35.
19. W. Guttenfelder, J. Candy, S.M. Kaye, W.M. Nevins, R.E. Bell, G.W. Hammett, B.P. LeBlanc, and H. Yuh, 2012, Scaling of linear microtearing stability for a high collisionality National Spherical Torus Experiment discharge, Physics of Plasmas 19:022506; W. Guttenfelder, J. Candy, S.M. Kaye, W. M. Nevins, E. Wang, J. Zhang, R. E. Bell, et al., 2012, Simulation of microtearing turbulence in national spherical torus experiment, Physics of Plasmas 19:056119.
20. W.X. Wang, T.S. Hahm, S. Ethier, L.E. Zakharov, and P.H. Diamond, 2011, Trapped electron mode turbulence driven intrinsic rotation in tokamak plasmas, Physical Review Letters 106:085001.
21. C. Holland, C.C. Petty, L. Schmitz, K.H. Burrell, G.R. McKee, T.L. Rhodes, and J. Candy, 2012, Progress in GYRO validation studies of DIII-D H-mode plasmas, Nuclear Fusion 52:114007.
22. P.B. Snyder, R.J. Groebner, J.W. Hughes, T.H. Osborn, M. Beurskens, A.W. Leonard, H.R. Wilson, and X.Q. Xu, 2011, A first-principles predictive model of the pedestal height and width: Development, testing and ITER optimization with the EPED model, Nuclear Fusion 51:103016.
23. D.R. Hatch, D. Told, F. Jenko, H. Doerk, M.G. Dunne, E. Wolfrum, E. Viezzer, the ASDEX Upgrade Team, and M.J. Pueschel, 2015, Gyrokinetic study of ASDEX upgrade inter-ELM pedestal profile evolution, Nuclear Fusion 55:063028; D.R. Hatch, M. Kotschenreuther, S. Mahajan, P. Valanju, F. Jenko, D. Told, T. Görler, and S. Saarelma, 2016, Microtearing turbulence limiting the JET-ILW pedestal, Nuclear Fusion 56:104003; D.R. Hatch, M. Kotschenreuther, S. Mahajan, P. Valanju, and X. Liu, 2017, A gyrokinetic perspective on the JET-ILW pedestal, Nuclear Fusion 57:036020.
24. C.S. Chang, S. Ku, G.R. Tynan, R. Hager, R.M. Churchill, I. Cziegler, M. Greenwald, A.E. Hubbard, and J.W. Hughes, 2017, Fast low-to-high confinement mode bifurcation dynamics in a tokamak edge plasma gyrokinetic simulation, Physical Review Letters 118:175001.
25. J.K. Park, Y.M. Jeon, J.E. Menard, W.H. Ko, S.G. Lee, Y.S. Bae, and M. Joung, 2013, Rotational resonance of nonaxisymmetric magnetic braking in the KSTAR tokamak, Physical Review Letters 111:095002.
26. Z.R. Wang, M.J. Lanctot, Y.Q. Liu, J-K. Park, and J.E. Menard, 2015, Three-dimensional drift kinetic response of high-β plasmas in the DIII-D tokamak, Physical Review Letters 114:145005.
27. N.M. Ferraro, S.C. Jardin, L.L. Lao, M.S. Shephard, and F. Zhang, 2016, Multi-region approach to free-boundary three-dimensional tokamak equilibria and resistive wall instabilities, Physics of Plasmas 23:056114.
28. V.A Izzo, P.B. Parks, N.W. Eidietis, D. Shiraki, E.M. Hollmann, N. Commaux, R.S. Granetz, et al., 2015, The role of MHD in 3D aspects of massive gas injection, Nuclear Fusion 55:073032.
29. S. Ku, R. Hager, C.S. Chang, J.M. Kwon, and S.E. Parker, 2016, A new hybrid-Lagrangian numerical scheme for gyrokinetic simulation of tokamak edge plasma, Journal of Computational Physics 315:467.
30. C.S. Chang, S. Ku, A. Loarte, V. Parail, F. Köchl, M. Romanelli, and R. Maingi, 2017, Gyrokinetic projection of the divertor heat-flux width from present tokamaks to ITER, Nuclear Fusion 57:116023.
31. E.J. Strait, 2015, Magnetic control of magnetohydrodynamic instabilities in tokamaks, Physics of Plasmas 22:021803.
32. P.C. de Vries, G. Pautasso, E. Nardon, P. Cahyna, S. Gerasimov, J. Havlicek, T.C. Hender, et al., 2016, Scaling of the MHD perturbation amplitude required to trigger a disruption and predictions for ITER, Nuclear Fusion 56:026007, doi:10.1088/0029-5515/56/2/026007.
33. F. Turco, T.C. Luce, W. Solomon, G. Jackson, G.A. Navratil, and J.M. Hanson, 2018, The causes of the disruptive tearing instabilities of the ITER baseline scenario in DIII-D, Nuclear Fusion 58:106043.
34. M.J. Lanctot, J.A. Snipes, H. Reimerdes, C. Paz-Soldan, N. Logan, J.M. Hanson, R.J. Buttery, et al., 2016, A path to stable low-torque plasma operation in ITER with test blanket modules, Nuclear Fusion 57:036004.
35. R.J. La Haye, R. Prater, R.J. Buttery, N. Hayashi, A. Isayama, M.E. Maraschek, L. Urso and H. Zohm, 2006, Cross-machine benchmarking for ITER of neoclassical tearing mode stabilization by electron cyclotron current drive, Nuclear Fusion 46:451.
36. E. Kolemen, A.S. Welander, R.J. La Haye, N.W. Eidietis, D.A. Humphreys, J. Lohr, V. Noraky, B.G. Penaflor, R. Prater, and F. Turco, 2014, State-of-the-art neoclassical tearing mode control in DIII-D using real-time steerable electron cyclotron current drive launchers, Nuclear Fusion 54:073020
37. E.S. Strait, J.M. Bialek, I.N. Bogatu, M.S. Chance, M.S. Chu, D.H. Edgell, A.M. Garofalo, et al., 2004, Resistive wall mode stabilization with internal feedback coils in DIII-D, Physics of Plasmas 11:2505.
38. S.A. Sabbagh, J.-W. Ahn, J. Allain, R. Andre, A. Balbaky, R. Bastasz, D. Battaglia, et al., 2013, Overview of physics results from the conclusive operation of the National Spherical Torus Experiment, Nuclear Fusion 53:104007.
39. See, for example, H. Zohm, F. Träuble, W. Biel, E. Fable, R. Kemp, H. Lux, M. Siccinio, and R. Wenninger, 2017, A stepladder approach to a tokamak fusion power plant, Nuclear Fusion 57:086002, for discussion of this formula.
40. S. Coda, O. Sauter, M.A. Henderson, and T.P. Goodman, 2008, “Full Bootstrap Discharge Sustainment in Steady State in the TCV Tokamak,” in Proceedings of the 22nd IAEA Fusion Energy Conference, EX/2-3.
41. S.M. Kaye, G.L. Jahns, A.W. Morris, S. Sesnic, K. Bol, M.S. Chance, P. Couture, et al., 1988, Characteristics of low-q disruptions in PBX, Nuclear Fusion 28:1963.
42. S.A. Sabbagh, J.M. Bialek, R.E. Bell, A.H. Glasser, B.P. LeBlanc, J.E. Menard, and F. Paoletti, 2004, The resistive wall mode and feedback control physics design in NSTX, Nuclear Fusion 44:560.
43. Ya.I. Kolesnichenko, A. Könies, V.V. Lutsenko, and Yu.V. Yakovenko, 2011, Affinity and difference between energetic-ion-driven instabilities in 2D and 3D toroidal systems, Plasma Physics and Controlled Fusion 53:024007.
44. K. Toi, K. Ogawa, M. Isobe, M. Osakabe, D.A. Spong, and Y. Todo, 2011, Energetic-ion-driven global instabilities in stellarator/helical plasmas and comparison with tokamak plasmas, Plasma Physics and Controlled Fusion 53:024008.
45. K. Ghantous, N.N. Gorelenkov, H.L. Berk, W.W. Heidbrink, and M.A. Van Zeeland, 2012, 1.5D quasilinear model and its application on beams interacting with Alfvén eigenmodes in DIII-D, Physics of Plasmas 19:092511.
46. M. Podestà, M. Gorelenkova, and R.B. White, 2014, A reduced fast ion transport model for the tokamak transport code TRANSP, Plasma Physics and Controlled Fusion 56:055003.
47. Y. Todo, M.A. Van Zeeland, and W.W. Heidbrink, 2016, Fast ion profile stiffness due to the resonance overlap of multiple Alfvén eigenmodes, Nuclear Fusion 56:112008.
48. E. Belova, N.N. Gorelenkov, N.A. Crocker, J.B. Lestz, E.D. Fredrickson, S. Tang, and K. Tritz, 2017, Nonlinear simulations of beam-driven compressional Alfvén eigenmodes in NSTX, Physics of Plasmas 24:042505.
49. W.W. Heidbrink, C.S. Collins, M. Podestà, G.J. Kramer, D.C. Pace, C.C. Petty, L. Stagner, M.A. Van Zeeland, R.B. White, and Y.B. Zhu, 2017, Fast-ion transport by Alfvén eigenmodes above a critical gradient threshold, Physics of Plasmas 24:056109.
50. E.D. Fredrickson, E.V. Belova, N.N. Gorelenkov, M. Podestà, R.E. Bell, N.A. Crocker, A. Diallo, B.P. LeBlanc, and the NSTX-U Team, 2018, Global Alfvén eigenmode scaling and suppression: Experiment and theory, Nuclear Fusion 58:082022.
51. M.A. Van Zeeland, W.W. Heidbrink, R. Nazikian, M.E. Austin, C.Z. Cheng, M.S. Chu, N.N. Gorelenkov, et al., 2009, Measurements, modelling and electron cyclotron heating modification of Alfvén eigenmode activity in DIII-D, Nuclear Fusion 49:065003.
52. P.C. de Vries, M.F. Johnson, B. Alper, P. Buratti, T.C. Hender, H.R. Koslowski, V. Riccardo, and JET-EFDA Contributors, 2011, Survey of disruption causes at JET, Nuclear Fusion 51:053018.
53. S. Gerhardt, D.S. Darrow, R.E. Bell, B.P. LeBlanc, J.E. Menard, D. Mueller, A.L. Roquemore, S.A. Sabbagh, and H. Yuh, 2013, Detection of disruptions in the high-β spherical torus NSTX, Nuclear Fusion 53:063021.
54. J.W. Berkery, S.A. Sabbagh, R.E. Bell, S.P. Gerhardt, and B.P. LeBlanc, 2017, A reduced resistive wall mode kinetic stability model for disruption forecasting, Physics of Plasmas 24:056103.
55. P.C. de Vries, G. Pautasso, D. Humphreys, M. Lehnen, S. Maruyama, J.A. Snipes, A. Vergara, and L. Zabeo, 2016, Requirements for triggering the ITER Disruption Mitigation System, Fusion Science and Technology 69:471, doi: 10.13182/FST15-176.
56. G. Pautasso, C. Tichmann, S. Egorov, T. Zehetbauer, O. Gruber, M. Maraschek, K.-F. Mast, et al., 2002, On-line prediction and mitigation of disruptions in ASDEX Upgrade, Nuclear Fusion 42:100.
57. J. Kates-Harbeck, A. Svyatkovskiy, and W. Tang, 2018, “Accelerating Progress Towards Controlled Fusion Power via Deep Learning at the Largest Scale,” submitted to Science.
58. C. Rau and R. Granetz, to be published in Fusion Science and Technology (2018).
59. H. Reimerdes, M.S. Chu, A.M. Garofalo, G.L. Jackson, R.J. La Haye, G.A. Navratil, M. Okabayashi, J.T. Scoville, and E.J. Strait, 2004, Measurement of the resistive-wall-mode stability in a rotating plasma using active MHD spectroscopy, Physical Review Letters 93:135002.
60. J.W. Berkery, S.A. Sabbagh, A. Balbaky, R.E. Bell, R. Betti, A. Diallo, S.P. Gerhardt, B.P. LeBlanc, J. Manickam, J.E. Menard, and M. Podestà, 2014, Measured improvement of global magnetohydrodynamic mode stability at high-beta, and in reduced collisionality spherical torus plasmas, Physics of Plasmas 21:056112.
61. E.M. Hollman, 2006, “DIII-D Studies of Massive Gas Injection Fast Shutdowns for Disruption Mitigation,” 33rd European Physical Society Conference on Plasma Physics, ECA 301 5.136, Curran Associates, Inc., http://www.proceedings.com/15959.html.
62. M. Bakhtiari, G. Olynyk, R. Granetz, D.G. Whyte, M.L. Reinke, K. Zhurovich, and V. Izzo, 2011, Using mixed gases for massive gas injection disruption mitigation on Alcator C-Mod, Nuclear Fusion 51:063007.
63. M. Lehnen, A. Alonso, G. Arnoux, N. Baumgarten, S.A. Bozhenkov, S. Brezinsek, M. Brix, et al., 2011, Disruption mitigation by massive gas injection in JET, Nuclear Fusion 51:123010.
64. N. Commaux, D. Shiraki, L.R. Baylor, E.M. Hollmann, N.W. Eidietis, C.J. Lasnier, R.A. Moyer, T.C. Jernigan, S.J. Meitner, S.K. Combs, and C.R. Foust, 2016, First demonstration of rapid shutdown using neon shattered pellet injection for thermal quench mitigation on DIII-D, Nuclear Fusion 56:046007.
65. D. Shiraki, N. Commaux, L.R. Baylor, C.M. Cooper, N.W. Eidietis, E.M. Hollmann, C. Paz-Soldan, S.K. Combs, and S.J. Meitner, 2018, Dissipation of post-disruption runaway electron plateaus by shattered pellet injection in DIII-D, Nuclear Fusion 58:056006.
66. N. Commaux, L.R. Baylor, S.K. Combs, N.W. Eidietis, T.E. Evans, C.R. Foust, E.M. Hollmann, et al., 2011, Novel rapid shutdown strategies for runaway electron suppression in DIII-D, Nuclear Fusion 51:103001.
67. Z. Guo, C.J. McDevitt, and X.-Z. Tang, 2018, Control of runaway electron energy using externally injected whistler waves, Physics of Plasmas 25:032504.
68. T. Evans, R.A. Moyer, K.H. Burrell, M.E. Fenstermacher, I. Joseph, A.W. Leonard, T.H. Osborne, et al., 2006, Edge stability and transport control with resonant magnetic perturbations in collisionless tokamak plasmas, Nature Physics 21:1.
69. Y. Liang, H.R. Koslowski, P.R. Thomas, E. Nardon, B. Alper, P. Andrew, Y. Andrew, et al., 2007, Active control of Type-I edge-localized modes with n = 1 perturbation fields in the JET tokamak, Physical Review Letters 98:265004.
70. A. Kirk, E. Nardon, R. Akers, M. Bécoulet, G. De Temmerman, B. Dudson, B. Hnat, Y.Q. Liu, R. Martin, P. Tamain, D. Taylor, and the MAST Team, 2010, Resonant magnetic perturbation experiments on MAST using external and internal coils for ELM control, Nuclear Fusion 50:034008.
71. J.M. Canik, 2010, On demand triggering of edge localized instabilities using external nonaxisymmetric magnetic perturbations in toroidal plasmas, Physical Review Letters 104:045001.
72. P.T. Lang, A. Alonso, B. Alper, E. Belonohy, A. Boboc, S. Devaux, T. Eich, et al., 2011, ELM pacing investigations at JET with the new pellet launcher, Nuclear Fusion 51:033010.
73. D.K. Mansfield, A.L. Roquemore, T. Carroll, Z. Sun, J.S. Hu, L. Zhang, Y.F. Liang, et al., 2013, First observations of ELM triggering by injected lithium granules in EAST, Nuclear Fusion 53:113023.
74. P.T. Lang, A. Burckhart, M. Bernert, L. Casali, R. Fischer, O. Kardaun, G. Kocsis, et al., 2014, ELM pacing and high-density operation using pellet injection in the ASDEX Upgrade all-metal-wall tokamak, Nuclear Fusion 54:083009 (2014).
75. A. Bortolon, R. Maingi, D.K. Mansfield, A. Nagy, A.L. Roquemore, L.R. Baylor, N. Commaux, et al., 2016, High frequency pacing of edge localized modes by injection of lithium granules in DIII-D H-mode discharges, Nuclear Fusion 56:056008.
76. D.K. Mansfield, A.L. Roquemore, H. Schneider, J. Timberlake, H. Kugel, M. G. Bell, and the NSTX Research Team, 2010, A simple apparatus for the injection of lithium aerosol into the scrape-off layer of fusion research devices, Fusion Engineering and Design 85:890.
77. R. Maingi, J.S. Hu, Z. Sun, K. Tritz, G.Z. Zuo, W. Xu, M. Huang, et al., 2018, ELM elimination with Li powder injection in EAST discharges using the tungsten upper divertor, Nuclear Fusion 58:024003.
78. X. Chen, K.H. Burrell, T.H. Osborne, W.M. Solomon, K. Barada, A.M. Garofalo, R.J. Groebner, et al., 2017, Stationary QH-mode plasmas with high and wide pedestal at low rotation on DIII-D, Nuclear Fusion 57:022007.
79. D.G. Whyte, A.E. Hubbard, J.W. Hughes, B. Lipschultz, J.E. Rice, E.S. Marmar, M. Greenwald, et al., 2010, I-mode: An H-mode energy confinement regime with L-mode particle transport in Alcator C-Mod, Nuclear Fusion 50:105005.
80. NRC, 2004, Burning Plasma.
81. C.P.C. Wong, M. Abdou, Y. Katoh, R.J. Kurtz, A. Lumsdaine, E. Marriott, B. Merrill, et al., 2013, Progress on DCLL blanket concept, Fusion Science and Technology 64:623.
82. M. Shimada, C.N. Taylor, R.J. Pawelko, L.C. Cadwallader, and B.J. Merrill, 2017, Tritium plasma experiment upgrade and improvement of surface diagnostic capabilities at STAR facility for enhancing tritium and nuclear PMI sciences, Fusion Science and Technology 71:310.
83. B.J. Merrill, L.C. Cadwallader, M. Shimada, P.W. Humrickhouse, C.N. Taylor, D.A. Stewart, and R.J. Pawelko, 2018, Recent accomplishments of the fusion safety program at the Idaho National Laboratory, Fusion Engineering and Design, 136:1106.
84. L. Waganer, C. Kessel, S. Malang, E. Marriott, S. Reyes, and A. Davis, 2018, The examination of the FNSF maintenance approach, Fusion Engineering and Design 135:394.
85. B.D. Wirth, K.D. Hammond, S.I. Krasheninnikov, and D. Maroudas, 2015, Challenges and opportunities of modeling plasma’s surface interactions in tungsten using high-performance computing, Journal of Nuclear Materials 463:30.
86. S.A. Gourlay, 2016, The U.S. Magnet Development Program Plan, https://science.energy.gov/~/media/hep/pdf/Reports/MagnetDevelopmentProgramPlan.pdf.
87. C. Kessel, 2015, The Fusion Nuclear Science Facility, the critical step in the pathway to fusion energy, Fusion Science and Technology 68:2.
88. P.C. Duckworth, L.R. Baylor, S.J. Meitner, S.K. Combs, D.A. Rasmussen, M. Hechler, and T. Edgemon, 2012, Development and demonstration of a supercritical helium-cooled cryogenic viscous compressor prototype for the ITER vacuum system, Advanced Cryogenic Engineering 57A-5B:1234.
89. A.N. Perevezentsev, L.A. Bernstein, L.A. Rivkis, I.G. Prykina, V.V. Aleksandrov, I.A. Ionessian, M.I. Belyakov, and I.B. Kuprianov, 2017, Study of outgassing and removal of tritium from metallic construction materials of ITER vacuum vessel components, Fusion Science and Technology 72:1.
90. J.E. Klein, A.S. Poore, and D.W. Babineau, 2015, Development of fusion fuel cycles: Large deviations from US defense program systems, Fusion Engineering and Design 96:113.
91. M.S. Lyttle, L.R. Baylor, R.E. Battle, S.J. Meitner, D.A. Rasmussen, and J.M. Shoulders, 2017, Tritium challenges and plans for ITER pellet fueling and disruption mitigation systems, Fusion Science and Technology 71:251.
92. P. Libeyre, C. Cormany, N. Dolgetta, E. Gaxiola, C. Jong, C. Lyraud, N. Mitchell, et al., 2016, Starting manufacture of the ITER central solenoid, IEEE Transactions on Applied Superconductivity 26:4203305.
93. R. Maingi, 2016, Chapter 3 in Magnetic Fusion Energy: From Experiments to Power Plants (G.H. Neilson, ed.), Woodhead Series in Energy, No. 99, Elsevier.
94. R.P. Wenninger, M. Bernert, T. Eich, E. Fable, G. Federici, A. Kallenbach, A. Loarte, et al., 2014, DEMO divertor limitations during and in between ELMs, Nuclear Fusion 54:114003.
95. H. Zohm, C. Angioni, E. Fable, G. Federici, G. Gantenbein, T. Hartmann, K. Lackner, et al., 2013, On the physics guidelines for a tokamak DEMO, Nuclear Fusion 53:073019.
96. P.C. Stangeby and A.W. Leonard, 2011, Obtaining reactor-relevant divertor conditions in tokamaks, Nuclear Fusion 51:063001.
97. V.A. Soukhanovskii, 2017, A review of radiative detachment studies in tokamak advanced magnetic divertor configurations, Plasma Physics and Controlled Fusion 59:064005.
98. A.W. Leonard, 2018, Plasma detachment in divertor tokamaks, Plasma Physics and Controlled Fusion 60:044001.
99. M. Kotschenreuther, P.M. Valanju, S.M. Mahajan, and J.C. Wiley, 2007, On heat loading, novel divertors, and fusion reactors, Physics of Plasmas 14:072502.
100. V.A. Soukhanovskii, S.L. Allen, M.E. Fenstermacher, C.J. Lasnier, M.A. Makowski, A.G. McLean, W.H. Meyer, et al., 2018, Developing physics basis for the snowflake divertor in the DIII-D tokamak, Nuclear Fusion 58:036018.
101. H. Reimerdes, B.P. Duval, J.R. Harrison, B. Labit, B. Lipschultz, T. Lunt, C. Theiler, et al., 2017, TCV experiments towards the development of a plasma exhaust solution, Nuclear Fusion 57:126007.
102. G. Calabrò, B.J. Xiao, S.L. Chen, Y.M. Duan, Y. Guo, J. G. Li, L. Liu, et al., 2015, EAST alternative magnetic configurations: Modelling and first experiments, Nuclear Fusion 55:083005.
103. H.Y. Guo, C.F. Sang, P.C. Stangeby, L.L. Lao, T.S. Taylor, and D.M. Thomas, 2017, Small angle slot divertor concept for long pulse advanced tokamaks, Nuclear Fusion 57:044001.
104. H.Y. Guo, T. Abrams, B. Covele, D.N. Hill, A. Leonard, P.C. Stangeby, and D. Thomas, 2018, “DIII-D as a Key User Facility in a National Divertor and Materials Science Program,” white paper submitted to the committee.
105. P.M. Valanju, M. Kotschenreuther, S.M. Mahajan, and J. Canik, 2009, Super-X divertors and high power density fusion devices, Physics of Plasmas 16:056110.
106. G. Fishpool, J. Canik, G. Cunningham, J. Harrison, I. Katramados, A. Kirk, M. Kovari, H. Meyer, R. Scannell, and the MAST-Upgrade Team, 2013, MAST-upgrade divertor facility and assessing performance of long-legged divertors, Journal of Nuclear Materials 438:S356.
107. B. LaBombard, E. Marmar, J. Irby, J.L. Terry, R. Vieira, G. Wallace, D.G. Whyte, et al., 2015, ADX: A high field, high power density, advanced divertor and RF tokamak, Nuclear Fusion 55:053020.
108. T.W. Morgan, P. Rindt, G.G. van Eden, V. Kvon, M.A. Jaworksi, and N.J. Lopes Cardozo, 2017, Liquid metals as a divertor plasma-facing material explored using the Pilot-PSI and Magnum-PSI linear devices, Plasma Physics and Controlled Fusion 60:014025.
109. H.W. Kugel, M.G. Bell, J.-W. Ahn, J.P. Allain, R. Bell, J. Boedo, C. Bush, et al., 2008, The effect of lithium surface coatings on plasma performance in the National Spherical Torus Experiment, Physics of Plasmas 15:056118.
110. D.P. Boyle, R. Majeski, J.C. Schmitt, C. Hansen, R. Kaita, S. Kubota, M. Lucia, and T.D. Rognlien, 2017, Observation of flat electron temperature profiles in the lithium tokamak experiment, Physical Review Letters 119:015001.
111. R. Maingi, T.H. Osborne, B.P. LeBlanc, R.E. Bell, J. Manickam, P.B. Snyder, J.E. Menard, et al., 2009, Edge-localized-mode suppression through density-profile modification with lithium-wall coatings in the National Spherical Torus Experiment, Physical Review Letters 103:075001.
112. J.S. Hu, G.Z. Zuo, J. Ren, Q.X. Yang, Z.X. Chen, H. Xu, L.E. Zakharov, et al., 2016, First results of the use of a continuously flowing lithium limiter in high performance discharges in the EAST device, Nuclear Fusion 56:046011.
113. G. Ramogida, G. Calabro, F. Crisanti, M.L. Apicella, G. Artaserse, W. Bin, L. Boncagni, et al., 2017, D-shaped configurations in FTU for testing liquid lithium limiter: Preliminary studies and experiments, Nuclear Materials and Energy 12:1082.
114. G.Z. Zuo, J.S. Hu, R. Maingi, J. Ren, Z. Sun, Q.X. Yang, Z.X. Chen, et al., 2017, Mitigation of plasma–material interactions via passive Li efflux from the surface of a flowing liquid lithium limiter in EAST, Nuclear Fusion 57:046017.
115. M. Kikuchi and M. Azumi, 2015, Frontiers in Fusion Research II—Introduction to Modern Tokamak Physics, Springer.
116. M.A. Jaworski, 2018, “First-Wall, Plasma-Material Interaction, Liquid Metals and Strategic Elements for Advancing Liquid Metal Science and Technology,” presentation to the committee, April.
117. B. LaBombard, P.C. Stangeby, R. Majeski, and J.P. Allain, 2018, “Elements of a U.S. R&D Plan to Solve Plasma-Material Interaction Challenges for Magnetic Fusion Energy,” white paper submitted to the committee.
118. M. Kikuchi and M. Azumi, 2012, Steady-state tokamak research: Core physics, Review of Modern Physics 84:1807.
120. C. Gormezano, A.C.C. Sips, T.C. Luce, S. Ide, A. Becoulet, X. Litaudon, A. Isayama, et al., 2007, Progress in the ITER Physics Basis, Chapter 6: Steady state operation, Nuclear Fusion 47:S285.
121. R. Prater, C.P. Moeller, R.I. Pinsker, M. Porkolab, O. Meneghini, and V.L. Vdovin, 2014, Application of very high harmonic fast waves for off-axis current drive in the DIII-D and FNSF-AT tokamaks, Nuclear Fusion 54:083024.
122. A. Ekedahl, L. Delpech, M. Goniche, D. Guilhem, J. Hillairet, M. Preynas, P.K. Sharma, et al., 2011, Long pulse operation with the ITER-relevant LHCD antenna in Tore Supra, AIP Conference Proceedings 1406:399.
123. S.J. Wang, J. Kim, J.H. Jeong, H.J. Kim, M. Joung, Y.S. Bae, and J.G. Kwak, 2015, Recent experimental results of KSTAR RF heating and current drive, AIP Conference Proceedings 1689:030014.
124. X. Gong, B. Wan, J. Li, J. Quan, E. Li, F. Liu, Y. Zhao, et al., 2017, Realization of minute-long steady-state H-mode discharges on EAST, Plasma Science and Technology 19:032001.
125. T. Mutoh, R. Kumazawa, T. Seki, K. Saito, T. Watari, Y. Torii, N. Takeuchi, et al., 2004, Long-pulse operation and high-energy particle confinement study in ICRF heating of LHD, Fusion Science and Technology 46:175.
126. A. Boozer, 1998, What is a stellarator? Physics of Plasmas 5:1647.
127. J.-K. Park, Y.M. Jeon, Y. In, J.-W. Ahn, R. Nazikian, G. Park, J. Kim, et al., 2018, 3D field phase-space control in tokamak plasmas, Nature Physics 14:1223.
128. A.H. Boozer, 2018, Enhanced control, Nature Physics 14:1157.
129. A.H. Boozer, 1983, Transport and isomorphic equilibria, The Physics of Fluids 26:496.
130. H.E. Mynick, N. Pomphrey, and P. Xanthopoulos, 2010, Optimizing stellarators for turbulent transport, Physical Review Letters 105:095004.
131. P. Xanthopoulos, H.E. Mynick, P. Helander, Y. Turkin, G.G. Plunk, F. Jenko, T. Görler, D. Told, T. Bird, and J.H.E. Proll, 2014, Controlling turbulence in present and future stellarators, Physical Review Letters 113:155001.
132. D.A. Spong, S.P. Hirshman, L.A. Berry, J.F. Lyon, R.H. Fowler, D.J. Strickler, M.J. Cole, et al., 2001, Physics issues of compact drift optimized stellarators, Nuclear Fusion 41:711.
133. H.-S. Bosch, R.C. Wolf, T. Andreeva, J. Baldzuhn, D. Birus, T. Bluhm, T. Bräuer, et al., 2013, Technical challenges in the construction of the steady-state stellarator Wendelstein 7-X, Nuclear Fusion 53:126001.
134. T.S. Pedersen, A. Dinklage, Y. Turkin, R. Wolf, S. Bozhenkov, J. Geiger, G. Fuchert, et al., 2017, Key results from the first plasma operation phase and outlook for future performance in Wendelstein 7-X, Physics of Plasmas 24:055503.
135. D.A. Spong and I. Holod, 2015, “Analysis of Energetic Particle Driven Alfvén Instabilities in 3D Toroidal Systems Using a Global Gyrokinetic Model,” presented at14th IAEA Technical Meeting on Energetic Particles in Magnetic Confinement Systems.
136. E.A. Lazarus, M.C. Zarnstorff, S.R. Hudson, L.-P. Ku, D.C. McCune, D.R. Mikkelsen, D.A. Monticello, N. Pomphrey, and A.H. Reiman, 2004, Simulation of a discharge for the NCSX stellarator, Fusion Science and Technology 46:209.
137. S. Sakakibara, H. Yamada, K.Y. Watanabe, Y. Narushima, K. Toi, S. Ohdachi, M. Takechi, et al., 2001, MHD characteristics in the high beta regime of the Large Helical Device, Nuclear Fusion 41:1177.
138. S. Sakakabira, K.Y. Watanabe, Y. Takemura, M. Okamoto, S. Ohdachi, Y. Suzuki, Y. Narushima, et al., 2015, Characteristics of MHD instabilities limiting the beta value in LHD, Nuclear Fusion 55:083020.
139. G.J. Hartwell, S.F. Knowlton, J.D. Hanson, D.A. Ennis, and D.A. Maurer, 2017, Design, construction, and operation of the compact toroidal hybrid, Fusion Science and Technology 72:76.
140. T. Morisaki, S. Masuzaki, A. Komori, N. Ohyabu, M. Kobayashi, J. Miyazawa1, M. Shoji, et al., 2006, Review of divertor studies in LHD, Plasma Science and Technology 8:14.
141. A. Bader, A.H. Boozer, C.C. Hegna, S.A. Lazerson, and J.C. Schmitt, 2017, HSX as an example of a resilient non-resonant divertor, Physics of Plasmas 24:032506.
142. H. Renner, D. Sharma, J. Kißlinger, J. Boscary, H. Grote, and R. Schneider, 2004, Physical aspects and design of the Wendelstein 7-X divertor, Fusion Science and Technology 46:318.
143. D.T. Anderson, A. Bader, R.J. Fonck, C.C. Hegna, J.S. Sarff, O. Schmitz, and J.N. Talmadge, 2018, “A Mid-Scale Quasihelically Symmetric Experiment Would Significantly Accelerate Fusion Development Through the Stellarator Line,” white paper submitted to the committee.
144. D.A. Gates, D. Anderson, S. Anderson, M. Zarnstorff, D.A. Spong, H. Weitzner, G.H. Neilson, et al., 2018, Stellarator research opportunities: A report of the National Stellarator Coordinating Committee, Journal of Fusion Energy 37:51.
145. A. Bhattacharjee, P. Bonoli, A. Boozer, J. Callen, J. Candy, C.-S. Chang, A. Friedman, et al., 2018, “Accelerating Fusion Through Integrated Whole Device Modeling,” white paper submitted to the committee.
146. F. Ebrahimi, et al., 2018, “Importance of Theory, Computation and Predictive Modeling in the U.S. Magnetic Fusion Energy Strategic Plan,” white paper submitted to the committee.
147. U.S. Department of Energy, Report of the Workshop on Integrated Simulations for Magnetic Fusion Energy Sciences, June 2-4, 2015, Office of Fusion Energy Sciences, https://science.energy.gov/~/media/fes/pdf/workshop-reports/2016/ISFusionWorkshopReport_11-12-2015.pdf.
148. P. Bonoli and L.C. McInnes, 2015, Report of the Workshop on Integrated Simulations for Magnetic Fusion Energy Sciences, https://science.energy.gov/fes/community-resources/workshop-reports.
149. U.S. Department of Energy (DOE), 2018, Report on Transformative Enabling Capabilities for Efficient Advance Toward Fusion Energy, Fusion Energy Sciences Advisory Committee, Washington, DC.
150. C. Rea and R.S. Granetz, 2018, Exploratory machine learning studies for disruption prediction using large databases on DIII-D, Fusion Science and Technology 74(1-2):89-100, doi: 10.1080/15361055.2017.1407206.
151. J. Minervini, Y. Zhai, X. Wang, and R.C. Duckworth, 2018, “Developing HTS Magnets for Fusion Applications,” white paper submitted to the committee.
152. N. Mitchell and A. Devred, 2017, The ITER magnet system: Configuration and construction status, Fusion Engineering and Design 123:17.
153. M. Takayasu, L. Chiesa, P.D. Noyes, and J.V. Minervini, 2017, Investigation of HTS twisted stacked-tape cable (TSTC) conductor for high-field, high-current fusion magnets, IEEE Transactions on Applied Superconductivity 27:1.
154. W.H. Fietz, C. Barth, S. Drotziger, W. Goldacker, R. Heller, S.I. Schlachter, and K.-P. Weiss, 2013, Prospects of high temperature superconductors for fusion magnets and power applications, Fusion Engineering and Design 88:440.
155. S.J. Zinkle and L.L. Snead, 2015, Designing radiation resistance in materials for fusion energy, Annual Review of Materials Research 44:241.
156. A.R. Raffray, R. Nygren, D.G. Whyte, S. Abdel-Khalik, R. Doerner, F. Escourbiac, T. Evans, et al., 2010, High heat flux components—Readiness to proceed from near term fusion systems to power plants, Fusion Engineering and Design 85:93.
157. D. Stork, P. Agostini, J.L. Boutard, D. Buckthorpe, E. Diegele, S.L. Dudarev, C. English, et al., 2014, Developing structural, high-heat flux and plasma facing materials for a near-term DEMO fusion power plant: The EU assessment, Journal of Nuclear Materials 455:277.
158. C.M. Parish, K.A. Unocic, L. Tan, S.J. Zinkle, S. Kondo, L.L. Snead, D.T. Hoelzer, and Y. Katoha, 2017, Helium sequestration at nanoparticle-matrix interfaces in helium plus heavy ion irradiated nanostructured ferritic alloys, Journal of Nuclear Materials 483:21.
159. S.L. Zinkle, J.L. Boutard, D.T. Hoelzer, A. Kimura, R. Lindau, G.R. Odette, M. Rieth, L. Tan, and H. Tanigawa, 2017, Development of next generation tempered and ODS reduced activation ferritic/martensitic steels for fusion energy applications, Nuclear Fusion 57:092005.
160. L.L. Snead, T. Nozawa, M. Ferraris, Y. Katoh, R. Shinavski, and M. Sawan, 2011, Silicon carbide composites as fusion power reactor structural materials, Journal of Nuclear Materials 417:330.
161. M.J. Baldwin and R.P. Doerner, 2008, Helium induced nanoscopic morphology on tungsten under fusion relevant plasma conditions, Nuclear Fusion 4:035001.
162. L.M. Garrison, Y. Katoh, L.L. Snead, T.S. Byun, J. Reiser, and M. Rieth, 2016, Irradiation effects in tungsten-copper laminate composite, Journal of Nuclear Materials 481:134.
163. G. Tynan, M. Baldwin, R. Doerner, E. Hollmann, D. Nishijima, K. Umstadter, and J. Yu, 2010, Mixed material plasma-surface interactions in ITER: Recent results from the PISCES Group, in plasma interaction in controlled fusion devices, AIP Conference Proceedings 1237:78.
164. J. Rapp, T.M. Biewer, T.S. Bigelow, J.B.O. Caughman, R.C. Duckworth, R.J. Ellis, D.R. Giuliano, et al., 2016, The development of the material plasma exposure experiment, IEEE Transactions on Plasma Science 44:3456.
165. U.S. Department of Energy, 2018, Report on Transformative Enabling Capabilities for Efficient Advance Toward Fusion Energy, Fusion Energy Sciences Advisory Committee, Washington, DC, p. 35.
167. M.E. Sawan and M.A. Abdou, 2006, Physics and technology conditions for attaining tritium self-sufficiency for the DT fuel cycle, Fusion Engineering and Design 81:1131.
168. L.M. Giancarli, M. Abdou, D.J. Campbell, V.A. Chuyanov, M.Y. Ahn, M. Enoeda, C. Pane, et al., 2012, Overview of the ITER TBM Program, Fusion Engineering and Design 87:395.
169. G. Federici, W. Biel, M.R. Gilbert, R. Kemp, N. Taylor, and R. Wenninger, 2017, European DEMO design strategy and consequences for materials, Nuclear Fusion 57:092002.
170. S. Smolentsev, M. Abdou, C. Courtessole, G. Pulugundla, F.-C. Li, N. Morley, R. Munipalli, et al., 2017, Review of recent MHD activities for liquid metal blankets in the US, Magnetohydrodynamics 53:411.
171. B. Garcia-Diaz, L. Olson, H. Colon-Mercado, J. Teprovich, and D. Babineau, 2017, “Direct LiT Electrolysis in Molten Lithium,” presentation to the FESAC TEC Meeting at the Princeton Plasma Physics Lab, August.
172. P.J. Foster, R.S. Willms, W.K. Hollis, and D. Dogruel, 2016, “Measurement of Uranium Hydride Storage Bed Engineering Parameters,” 11th International Conference on Tritium Science and Technology 2016, American Nuclear Society, La Grange Park, Ill.
173. B.G. Hong, 2018, Overview of ITER TBM program objectives and management, International Journal of Energy Research 42:4.
174. C. Wong, P.C.M. Abdou, Y. Katoh, R.J. Kurtz, A. Lumsdaine, E. Marriott, B. Merrill, et al., 2013, Progress on DCLL blanket concept, Fusion Science and Technology 64:623.
175. F. Hernandez, P. Pereslavtsev, Q. Kang, P. Norajitra, B. Kiss, G. Nádasi, and O. Bitz, 2017, A new HCPB breeding blanket for the EU DEMO: Evolution, rationale and preliminary performances, Fusion Engineering and Design 124:882.
176. B. Garcinuno, D. Rapisarda, I. Fernández-Berceruelo, D. Jiménez-Rey, J. Sanz, C. Moreno, I. Palermo, and Á. Ibarra, 2017, Design and fabrication of a permeator against vacuum prototype for small scale testing at Lead-Lithium facility, Fusion Engineering and Design 124:871.
177. C. Day and T. Giegerich, 2014, Development of advanced exhaust pumping technology for a DT fusion power plant, IEEE Transactions on Plasma Science 42:1058.
178. J.-Ph. Girard, P. Garinb, N. Taylor, J. Uzan-Elbez, L. Rodríguez-Rodrigo, and W. Gulden, 2007, ITER, safety and licensing, Fusion Engineering and Design 82:506.
179. B. Bornschein, C. Day, D. Demange, and T. Pinna, 2013, Tritium management and safety issues in ITER and DEMO breeding blankets, Fusion Engineering and Design 88:466.
180. M. Shimada, Y. Oya, D.A. Buchenauer, and Y. Hatano, 2017, Hydrogen isotope retention and permeation in neutron-irradiated tungsten and tungsten alloys under PHENIX Collaboration, Fusion Science and Technology 72:652.
181. M. Shimada and R.J. Pawelko, 2018, Tritium permeability measurement in hydrogen-tritium system, Fusion Engineering and Design 129:134.
182. B.J. Merrill, P.W. Humrickhouse, and M. Shimada, 2016, Recent development and application of a new safety analysis code for fusion reactors, Engineering and Design 109-111:970.
183. V.S. Chan, R.D. Stambaugh, A.M. Garofalo, M.S. Chu, R.K. Fisher, C.M. Greenfield, D.A. Humphreys, et al., 2010, Physics basis of a fusion development facility utilizing the tokamak approach, Fusion Science and Technology 57:66.
184. D. Whyte, J. Minervini, B. LaBombard, E. Marmar, L. Bromberg, and M. Greenwald, 2016, Smaller and sooner: Exploiting high magnetic fields from new superconductors for a more attractive fusion energy development path, Journal of Fusion Energy 35:41.
185. J.E. Menard, L. Bromberg, T. Brown, T. Burgess, D. Dix, L. El-Guebaly, T. Gerrity, et al., 2011, Prospects for pilot plants based on the tokamak, spherical tokamak and stellarator, Nuclear Fusion 51:103014.
186. C.E. Kessel, M.S. Tillack, F. Najmabadi, F.M. Poli, K. Ghantous, N. Gorelenkov, X.R. Wang, et al., 2015, The ARIES advanced and conservative tokamak power plant study, Fusion Science and Technology 67:1, doi: 10.13182/FST14-794.
187. C.E. Kessel, J.P. Blanchard, A. Davis, L. El-Guebaly, N. Ghoniem, P.W. Humrickhouse, S. Malang, et al., 2015, The Fusion Nuclear Science Facility, the critical step in the pathway to fusion energy, Fusion Science and Technology 68(2):225-236, doi: 10.13182/FST14-953.