This chapter describes the importance of burning plasma research, explains why continued participation as an International Thermonuclear Experimental Reactor (ITER) partner is important to U.S. fusion energy research, and describes how ITER participation will inform the design of a compact fusion pilot plant as a new element of the U.S. magnetic fusion program. While decades of international collaborative research have been spent establishing confidence in the science and technology basis for reliable pulsed operation in ITER, a compact fusion pilot plant requires extending the burning plasma regime beyond ITER and achieving uninterrupted operation having simultaneously high normalized confinement, high fusion power density, and a long-life divertor and first-wall. Because the physics basis of the compact fusion pilot plant is less mature than the physics basis for ITER, hands-on experience and study of a burning plasma experiment, like ITER, will be essential along the path forward to the U.S. compact pilot plant.
This chapter begins with a brief summary of the committee’s Interim Report assessment of the importance of burning plasma research to the development of fusion energy and of the ongoing activities of the United States that support burning plasma science and the ITER project is included. Next, research activities that will maximize the benefits from U.S. participation in ITER and advance the science and technology needed to reduce risks in the design and operation of a compact fusion pilot plant are presented. Chapter 3 concludes with three findings and with two recommendations for how the U.S. Department of Energy’s (DOE’s) Office of Fusion Energy Sciences (FES) should conduct both near-term and long-term ITER research. Chapter 3 also includes two findings describing research needs if the U.S.
withdraws from the ITER project and states the committee’s recommendation that the United States should not withdraw. However, if the United States does withdraw from the ITER partnership, the U.S. DOE/FES should still initiate a plan leading toward the construction of a compact fusion pilot plant. In the scenario without ITER, an alternate means to study a burning plasma and to engage the international community would be required.
The importance of a burning plasma experiment as a required step in the development of practical fusion energy has been endorsed by all previous strategic planning documents prepared for the U.S. fusion research program.1 As ITER partners, the fusion research programs of the United States and other nations have focused on preparations for the study of burning plasmas using the ITER experiment for more than a decade.
Access to a burning plasma in ITER will open a new frontier in science. For the first time, scientists will observe the complex processes of a plasma that is primarily heated by the energetic particles created by fusion reactions within the plasma itself. Even beyond the observation of a self-heated fusion plasma, the study, control, and manipulation of a burning plasma in ITER will give scientists their first opportunity to demonstrate many technical capabilities needed by any energy-producing magnetic fusion device. ITER will extend the frontier of burning plasma research to include opportunities to learn how to sustain a fusion plasma for durations much longer than previously achieved and how to maximize the fusion power produced from a burning plasma. The frontier of burning plasma research also leads to progress in the applied and engineering sciences needed to design reliable structures that surround the plasma and provides the first opportunity to gather materials data needed to design systems that convert fusion energy into useful heat and power. In all of these ways, the science learned from study of burning plasma in ITER will extend beyond ITER and contribute to the engineering science of large fusion energy facilities and contribute to the applied science of predicting, controlling, and sustaining a magnetically confined burning plasma.
The committee’s Interim Report reaffirmed the importance of burning plasma research to the development of fusion energy, as well as to plasma science and other disciplines. Furthermore, because ITER is the only existing project to create a burning plasma at the scale of a power plant and also a major component of the U.S. fusion energy program, the scientific and technical goals of ITER give clear illustration of the importance of burning plasma research. ITER will demonstrate the scientific and technical feasibility of fusion energy for peaceful purposes. The ITER target is to sustain high fusion power gain (Q ~ 10) for more than 5 minutes and to provide the capabilities to study advanced operating scenarios and explore a wide
range of fusion confinement parameters. Long plasma durations exceeding 15 minutes are planned that will be useful for limited nuclear testing of tritium blanket modules. For shorter durations, ITER will test advanced scenarios with elevated plasma pressure and higher fusion power output that may allow investigation of ignited or near ignited burning plasmas. These ITER research targets are ambitious, but the understanding of burning plasma science has advanced significantly and the international effort to prepare for ITER operation has further increased confidence in the burning plasma performance that can be achieved in ITER. As explained in Chapter 2, this is due to the discovery of new ideas to control and sustain a burning plasma, the substantial improvements in the ability to predict plasma confinement and fusion energy performance, and the demonstration of burning plasma scenarios that are expected to simultaneously satisfy the requirements for stability, confinement, fuel purity, and compatibility with plasma-facing components.
The importance of burning plasma research to the development of fusion energy is described below in two parts: (1) understanding and controlling a burning plasma and (2) advancing fusion technology and engineering science.
Burning plasma experiments in ITER will address critical research areas and answer key scientific questions, including:
- A burning plasma experiment will represent the first time that a confined fusion plasma is dominated by fusion-born alpha particles. Energetic alpha particles from fusion reactions heat and sustain the burning plasma, but they are also predicted to drive plasma instabilities. While the onset of energetic particle instabilities is understood, detailed identification of nonlinear mechanisms is just beginning. Beyond understanding energetic particle dynamics, important research opportunities include methods to control energetic particles instability for helpful purposes such as favorably modifying the current profile or to govern the nonlinear dynamics to control fusion burn.
- A burning plasma experiment advances understanding of plasma transport properties from the core to the boundary. Although there has been considerable progress made in predicting plasma transport, the validity of these predictions needs to be tested in future burning plasma experiments. In particular, confinement studies using a burning plasma will help determine confinement scaling in desirable fusion reactor conditions characterized by high plasma beta, steady state, and compatible divertors. In addition, transport of plasma impurities eroded from material surfaces, as well as helium produced by fusion reactions, are not yet understood. ITER experiments
will give scientists a first look at heat and particle transport at the scale of a fusion power plant.
- A burning plasma experiment enables critical tests to control plasma transients. Due to the large energy of a burning plasma, transient events, which cause rapid energy loss from the plasma, may damage first-wall materials. Transients include plasma current disruptions and edge localized modes (ELMs). The United States has led the world in the development of techniques for understanding, as well as predicting, avoiding, and/or controlling plasma transients. These techniques are critical for ITER and other burning plasma devices, and ITER will provide a critical test of disruption avoidance and mitigation systems at reactor scale and power density.
- A burning plasma experiment advances divertor science necessary for a fusion power source. Unless controlled, the power escaping from a burning plasma will lead to inner wall damage. Control of escaping heat and particles is made by carefully shaping the magnetic field so that plasma flows along the plasma boundary and into a divertor, where the plasma heat and particle flux can be nearly extinguished by interaction with recycling neutrals. The U.S. research program has significantly advanced understanding of burning plasma boundary physics, including improved understanding of the narrow “scrape-off layer” connecting the confined plasma to the divertor2 and the successful testing of innovative divertor concepts.3 However, further developments for a divertor with long lifetime remains a major fusion research challenge, and ITER will enable detailed tests of divertor physics, including detachment, retention, erosion, and re-deposition in reactor-like conditions.
- A burning plasma experiment tests integrated scenarios that simultaneously test the requirements for stability, confinement, fuel purity, and compatibility with plasma-facing components needed for a fusion energy source. Plasma operation and control scenarios have been developed and tested in preparation for ITER experiments, and integrated models using the latest advances in high-performance computing now routinely interpret experimental measurements and make progress in predicting the results of burning plasma experiments. Additionally, the U.S. research program has led the world in the development of quiescent plasma scenarios not subject to damaging transient events, which can achieve the same plasma performance at reduced plasma current and so minimize the risk of disruption damage. A burning plasma experiment can also test other advanced scenarios, like the so-called “super H-mode,” which represents an attractive area of innovation aimed to reduce the size of a fusion device with improved confinement. An ITER research program focused strongly on advanced scenarios and physics model validation will enable an advance in understanding needed to develop an attractive, compact fusion pilot plant.
As an ITER partner, the United States receives full benefit from the technology developed for ITER. Because U.S. industry is fabricating major systems for ITER, burning plasma research is also building industrial capacity in fusion nuclear science, superconducting magnet engineering, nuclear instrumentation, and plasma heating and control systems. Construction and operation of the ITER facility addresses important research in fusion nuclear science and engineering science, including:
- Fusion fuel processing, blanket design, and tritium breeding. The release of fusion energy results from the fusion reactions of tritium and deuterium ions heated to 100 million K. While deuterium is abundant, tritium will be produced from lithium within a fusion breeding blanket, which is a key fusion nuclear technology needed for the development of fusion energy. Tritium for ITER will be obtained from the tritium removal facility of the Canadian-owned Ontario Power Generation utility using tritium produced within Canadian Deuterium Uranium (CANDU) nuclear power stations. The vast majority of the fuel injected in a fusion chamber will not be burned in a single pass. ITER will provide the first opportunity for large-scale fuel processing of deuterium-tritium, which will be transported continuously from the plasma edge, then exhausted from the vacuum chamber, stripped of impurities, and reinjected into the plasma. ITER will also provide the first opportunity to test and evaluate the performance of prototypical blanket modules and demonstrate technologies for tritium extraction from blankets in ITER Test Blanket Modules.4,5
- Fusion safety, remote handling, and waste management. As a burning plasma experiment, ITER will provide the first 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.6 ITER will be the first integrated demonstration of the safety, reliability, and effectiveness of these technologies.7
- Fusion materials science. As a burning plasma experiment, ITER will aid in the development of high-heat-flux components and will evaluate the performance of the components in a fusion environment at the scale of a power plant. 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.8,9,10 Additionally, the behavior and-integrity of materials irradiated by fusion neutrons are of great importance to the long-term viability of fusion energy.11 The high flux of energetic neutrons to the vessel and structural materials poses a serious materials problem that will require substantial testing, some of which will be done on ITER.12
- Plasma heating and current drive systems for fusion. Plasma heating by electromagnetic waves and neutral particle beams is needed to heat the plasma to a burning state,13 sustain plasma current,14 modify temperature and current profiles, and control plasma instabilities.15 Fusion reactor research continues to push the frontiers of high-power mm-wave and radio-frequency technology, and these technologies will be extensively tested at the reactor scale on ITER.16
- High-field magnet technology for fusion. 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. The superconducting magnets being constructed for ITER will be the largest ever made and are designed to operate with the highest practical magnetic field strength for large toroidal field coils made of niobium-tin (Nb3Sn) superconductors and supported by steel.17 Experience operating ITER’s superconducting magnets will advance physics understanding of the role of magnetic field in a burning plasma and inform the development of large-scale superconducting magnet technology.
As the world’s first burning plasma experiment at the scale of power plant, ITER will provide scientists the first opportunity to access the frontier of burning plasma research. The complex processes within a burning plasma that couple plasma confinement, energetic particles created by fusion reactions, plasma stability, and fusion materials and technologies will be investigated. Because ITER is built at the scale of a power plant, ITER offers the first opportunity to begin development of nearly all of the technologies needed for a fusion reactor. In addition, as the first licensed fusion research facility, within the licensing code requirements used by the French nuclear industry, the safe operation of ITER will demonstrate necessary safety and operational procedures of a fusion power system.
Burning plasma research in support of ITER and in preparation for ITER experiments is a primary focus of the fusion research programs of the United States and other nations. Preparation for ITER experiments is a central element of today’s burning plasma research activities. Participation in the ITER project also provides formal mechanisms for U.S. scientists to take leading roles in the international effort to develop fusion energy and to benefit significantly from international collaboration. Because the United States is a key contributor to ITER construction, participation in ITER has resulted in significant advances in U.S. domestic industrial capabilities and capacities that would not have happened without ITER participation.
The important role of ITER in the U.S. fusion research program is described below. First, the U.S. role in ITER construction is discussed and the resulting benefits to the United States in industrial capabilities. Next, advances in fusion technology resulting from U.S. participation in ITER are presented. Finally, research in preparation for ITER experiments provides a focus for ongoing burning plasma research that has already resulted in substantial scientific progress in understanding, predicting, and controlling a burning plasma.
In December 2017, the ITER project passed the construction milestone of having completed 50 percent of the tasks required for first plasma operation. ITER is a large and ambitious facility. The fusion containment vessel, superconducting magnets, and cryostat will weigh 23,000 metric tons. The largest superconducting magnets are the poloidal field coils with a 24 m diameter. Each of 18 superconducting toroidal field coils is 17 m tall, weighs 360 metric tons, and will be installed to a precision of less than 0.1 mm in the radial plates which support the large magnet forces.
Since the 2013 ITER Management Assessment Report,18 the new director, Bernard Bigot, has played a key role in enacting rigorous project discipline in a nuclear project culture, enabling milestones to be met consistently. Project decision processes and accountability have been substantially improved through improved integration of the ITER Central Team and the domestic agencies. Component design is now finalized, and a new optimized resource-loaded schedule has been developed to minimize the time to first plasma. Since 2016, all scheduled project milestones have been achieved on schedule. The recently revised ITER Research Plan within the Staged Approach19 has a first plasma in 2025, first scientific experiments in 2028, and a first burning plasma experiment with deuterium and tritium fuel in 2035. The ITER Organization now has a team of over 800 people, of which
5.5 percent come from the United States, with many thousands working on the construction site and in the globally distributed supply chain.
The external structure surrounding the tokamak is now largely complete, as is the assembly hall standing 60 m tall and with a crane lifting capacity of 1,500 metric tons. The ITER cryo-plant will be the largest single platform cryo-cooling facility in the world. It will distribute liquid helium to various machine components, including the superconducting magnets, thermal shield, and divertor cryopumps. The last of 18 skids supporting the helium compressors was installed atop its massive 4-meter-high concrete pad in November 2017. The 400 kV switchyard to provide power from the grid has been successfully installed and commissioned. Many ancillary heating and diagnostic buildings are now erected and awaiting fit out.
A 12,000 m2 facility has been built to wind the largest poloidal field coils since they are too large to be transported to the site. The first two poloidal field coils are now being wound and all the superconducting strand is now onsite. The 30 m × 30 m cryostat is being assembled on site and undergoing final welding. General Atomics has completed the first module of the 1,000-metric-ton central solenoid, which will produce ITER’s highest magnetic field of 13 T. Besides initiating plasma current within ITER, the currents in the six modules of the central solenoid will be independently controlled to shape and position the plasma.
The United States has committed to contributing 9.09 percent of ITER’s construction costs. Participation in ITER in this fashion has resulted in significant advances in U.S. domestic industrial capabilities and capacities, with the vast majority (approximately 80 percent) of U.S. ITER construction funding remaining within the U.S. supply chain.20 For example,
- The United States has proven its capability for fabricating superconductor in bulk, producing over 4 miles of cable-in-conduit superconductor for the toroidal field magnets;
- The United States is fabricating a first-of-a-kind 13 m tall, 13 T central solenoid electromagnet, which is unique worldwide and has required the development of bespoke fabrication and testing infrastructure;
- U.S. industry is developing microwave and radio-frequency transmission lines to provide unprecedented power transfer for heating in ITER;
- High-throughput cryogenic pellet fueling systems and tritium processing systems have been developed by U.S. national laboratories; and
- Instrumentation for the fusion nuclear environments has been developed.
In addition, the United States has been tasked with the research, design, and fabrication of the AC power system (delivered), the tokamak cooling water system, the vacuum pumping systems and the tokamak exhaust processing system. The United States has also been a key contributor toward the approval of ITER’s license
to start construction, by providing a “pedigreed” version of the fusion-modified safety code MELCOR, developed and maintained by the Fusion Safety Program at Idaho National Laboratory, which has been used extensively for the safety analyses presented to the French Nuclear Regulator (Autorité de Sureté Nucléaire) as part of the Construction Authorization Request.
According to the DOE project execution plan for ITER,21 the United States has “made considerable progress in completing its assigned hardware design, research and development (R&D), and fabrication work.” Final design of about two-thirds of U.S. hardware is complete, and 2 of 13 in-kind hardware systems have been delivered. A total of $942 million has been obligated by the U.S. ITER project with contracts spread across U.S. industry, universities, and national laboratories, across 44 states.22 Approximately 50 U.S. personnel are working as members of the staff of the ITER International Organization. More than 100 full-time equivalents (FTEs) are working in the U.S. ITER Project Office, with most on the central solenoid construction and tokamak cooling water system. It is anticipated that with full funding, approximately 150 FTEs will be working in the U.S. ITER Project Office. The technical leadership and contributions made by the U.S. fusion science team are and will continue to be important to the eventual success of the ITER design, operation, diagnostics, and analyses. In addition, the U.S. financial commitment is highly leveraged by the sharing of costs and technology with its international partners. The performance of the United States in its ITER obligations has been very favorably assessed by the U.S. Government Accountability Office23 and DOE assessments and quality assurance audits conducted in 2015.
The U.S. contribution to ITER construction is a prime driver for the U.S. fusion technology program, and the United States is contributing a number of key systems that are both important for ITER and for any future fusion power system. U.S. ITER participation has the effect of not only providing leadership in the field of fusion technology, but also building capability in U.S. industry and capacity to provide components for future fusion experiments and facilities. The United States has several critical items to deliver for the completion of ITER construction and it should remain committed to delivering these key systems.
The International Tokamak Physics Activity (ITPA) operates under the auspices of ITER, and the ITPA provides an international framework for coordinated fusion research useful for all fusion programs and for broad progress toward fusion energy. The United States continues to make significant contributions to the ITPA, which coordinates the international tokamak physics research and development activities and provides the physics basis for the ITER project. Presently, the United States chairs or co-chairs three of the seven ITPA Topical Working groups. The United
States also actively participates in multiple-facility, joint tokamak experimental exercises. For example, joint experiments coordinated among MAST, ASDEX Upgrade, and DIII-D have recently evaluated the use of resonant magnetic field perturbations and pellet injection to suppress ELMs24 leading to the introduction of ELM control systems to be tested in ITER.
Ongoing advances in understanding burning plasma physics enable improvements in the prediction and optimization of ITER and in future burning plasma experiments. Progress has resulted by combining physics-based dimensionless analysis with the development and validation of advanced physics models using high-performance scientific computing. While this approach increases confidence in ITER performance, the burning plasma regime is an extrapolation of fundamental underlying processes. The normalized size of a burning plasma experiment is much larger than in existing devices, and this may alter the confinement and stability properties. Heating and current sources, including the fusion alpha particles that will heat a burning plasma, will involve new science. Finally, the boundary of the burning plasma, where atomic physics and interaction with material surfaces become important, may lead to new phenomena.
While addressing all these issues self-consistently awaits a burning plasma experiment, there are numerous opportunities to establish the physics basis for optimal fusion performance. These include: (1) exploring plasma confinement physics expected in ITER, especially core plasma energy transport, plasma edge pedestal physics, and the transition between low-confinement (L-mode) and high-confinement (H-mode), (2) exploring the strongly coupled physics of a burning plasma regime when fusion alpha particle heating interacts with plasma transport and stability; (3) optimizing methods to control ITER plasmas to reach performance targets, maintain stability, and avoid or mitigate transients and disruptions; and (4) further development and validation of plasma-material-interaction theory and simulation, leveraging advanced computational techniques and next generation exascale computers.
Integrated understanding of the plasma core-edge integration with the materials science of the divertor and first-wall is an ongoing research area that will benefit burning plasma experiments in ITER and contribute to U.S. efforts toward an attractive compact fusion pilot plant. Because the core and edge plasma are guided by different physics, bringing the two together in high fusion performance, reactor-relevant scenarios represent a grand challenge of burning plasma science. Doing so will enable validation of theory and simulation in a reactor relevant physics regime, while qualifying exhaust scenarios at relevant heat loads. Core-edge integration requires high plasma pressure in order to simultaneously achieve the high particle density needed for attractive exhaust solutions, while maintaining reactor-relevant low plasma collision rates. Experimental investigation of core-edge integration will also need to address the conflicting needs for a cold, dense diver-
tor to avoid material erosion and, simultaneously, a hot, high power density core for high fusion performance. ITER will provide a critical opportunity to explore a high-power density core at reactor-like dimensionless parameters, combined with a tungsten divertor.
As U.S. fusion researchers approach and access frontier studies of the burning plasma state, additional research topics provide opportunities for progress. These include the following:
- Disruption mitigation research, including development of prediction and avoidance algorithms using passive and active control as well as mitigation (i.e., shattered pellet injection, SPI);
- ELM control through pellet injection, applied external magnetic perturbations, natural ELM-free regimes;
- Controlling heat exhaust through detached divertor and innovative divertor configurations;
- Development of ITER-relevant steady-state, noninductive high-performance scenarios with acceptable divertor power loading;
- Development of predictive computational tools within integrated simulations enabling extrapolation to ITER regimes, including models for noninductive current drive, core confinement, pedestal physics, core-edge coupling, energetic particle-induced instabilities, plasma transient control, and plasma surface interactions;
- Development of integrated frameworks for plasma control algorithm development;
- Understanding material properties under the combined load condition and neutron loading present in ITER with verification of models against irradiated samples from test reactors; and
- Understanding the redistribution and loss of energetic particles and how Alfvénic instabilities can be mitigated in burning plasmas.
Fusion technology advances have been driven by ITER research needs and by next-step goals to fully enable the fusion energy system. ITER provides important experience for the critical development of fusion technology and engineering needed in a fusion power device. Examples include: operation of superconducting magnets, tritium handling systems, and tungsten divertor performance. Key contributions from the U.S. fusion technology program involve the fusion fuel cycle, fusion materials, fusion materials modeling,25 fusion plasma power handling, superconducting magnets, and liquid metals. These contributions have resulted from joint international projects in support of ITER and from tasks directed by
U.S. researchers. Examples include vacuum and gas species management,26,27 tritium fusion fuel cycle development,28 pellet injection for fueling and disruption mitigation,29 and the manufacture of the ITER central solenoid.30 The capabilities of the U.S. pellet injection technology will be used in future fusion experiments at the JET (Joint European Torus) device (see Figure 3.1).
Many of the program advancements in fusion technology and engineering science in the United States are coordinated with the Virtual Laboratory for Technology (VLT). The VLT functions as a “virtual” laboratory with many collaborating institutions within the United States, including eight universities, nine national laboratories, and one private company.31 The VLT facilitates fusion technology and engineering science in the United States by (1) developing the enabling technology for existing and next-step experimental devices, (2) exploring and understanding key materials and technology feasibility issues for attractive fusion power sources, and (3) conducting advanced design studies that provide integrated solutions for next-step and future fusion devices and call attention to research opportunities in the field.32 As was described in Chapter 2, the United States has also advanced the science of high-power plasma-material interactions using linear plasma simulators.
Research preparation for ITER’s scientific mission has focused on science and technology needed to extend ITER operation for long pulse durations and to provide safe, reliable operation of the machine for long duration pulses. Research results in support of ITER’s scientific mission also support the scientific missions of fusion energy facilities that will follow ITER. Three active research areas are: controlling the high heat flux on the divertor, preventing or minimizing transient edge instabilities that could damage the divertor armor, and preventing or mitigating an uncontrolled loss of plasma confinement, called a plasma current disruption, that could damage the first wall or the vessel structure. Significant progress in all three of these areas were described in Chapter 2.
The large heat flux to the divertor is determined, in part, by the width of the channel in which hot plasma escapes the confined plasma region and is directed to the divertor region. The ITER design can accommodate a peak heat flux of 10 MW/m2. This heat flux is achieved by operating the divertor in a “detached” or “partially detached” state where the majority of the escaping plasma heat flux is radiated by recycling neutral atoms in the diverter region. Detached divertors have been achieved on a number of devices worldwide, and the science of the escaping plasma heat flux continues to be an active area of research.33,34
The baseline operating scenario for ITER is high-confinement (H-mode) with a strong edge transport barrier and high pedestal pressure. The steep pressure gradient at the H-mode pedestal can destabilize ELMs. However, experiments on a number of devices worldwide have demonstrated that it is possible to suppress or mitigate the effects of these ELMs by either applying non-axisymmetric magnetic fields locally to the plasma edge, or injecting small pellets at high frequency. These control schemes are now included in the ITER design.
Finally, ITER will have the largest plasma current ever produced in a magnetic confinement device, and an unmitigated plasma current disruption has the potential to significantly damage first-wall components and divertor armor. To mitigate the effects of disruptions, ITER is planning to use massive injection of gases and injection of shattered pellets to provide injection of a substantially larger volume of particles than is in the fuel, which will radiate the stored energy across all of the material surfaces, thereby mitigating any risk of damage. These techniques to mitigate disruptions have been demonstrated, notably in DIII-D, and will soon be tested in JET using the shattered pellet injector (SPI) anticipated for ITER (see the section “Mitigation of Transients and Abnormal Events” in Chapter 2). Following a recent disruption mitigation workshop,35 the United States is actively participating in an ITER Disruption Task Force and exploring even more effective techniques for disruption mitigation.
Recent advances in validated theory and simulation provide opportunities to significantly extend ITER performance, including higher fusion power gain, longer plasma duration, demonstration of advanced operating scenarios, and improvements in divertor power handling. Simulations of integrated core-pedestal performance have already been used to optimize steady-state scenarios on DIII-D36 and make initial predictions for ITER.37 New high-performance regimes such as Super H-Mode have been predicted first, and later experimentally confirmed. These optimized regimes resulted in the highest plasma pedestal pressure (pped ~ 80 kPa) ever achieved in a magnetic fusion device using the Alcator C-Mod tokamak,38 and the transient demonstration of the highest value of peak fusion gain (Qequiv ~ 0.5) ever achieved on a medium-scale (R < 2 m) tokamak by operating with advanced scenarios on DIII-D.39 These achievements are summarized in Figure 3.2.40
Existing medium-scale and large-scale fusion experiments have achieved performance levels consistent with ITER’s goals and have briefly achieved normalized
performance levels that would enable ITER to exceed its fusion performance goals. These might allow high fusion power gain (Q ~ 10) at lower plasma currents and fusion ignition, or near ignition, which corresponds to achieving both high plasma confinement and high fusion power. Achieving and sustaining these levels of performance in ITER would represent a significant development toward smaller, less costly, compact fusion power systems. Figure 3.2 highlights these achievements using two metrics of fusion performance: the plasma pedestal pressure and a new metric motivated by recent advances in simulation <p>W/PhIaB. This is a metric of fusion performance based on the product of volume averaged pressure (<p>) and stored energy (W) divided by the product of heating power (Ph), plasma current (I), minor radius (a) and magnetic field (B), and is plotted in units of kPa MJ / MW MA m T. The normalized plasma pressure is βN; the average plasma pressure is given by <p> (Pa) = 4000 βN (B I/a); and the plasma energy confinement time is W/Ph. In this way, the fusion performance parameter, <p>W/PhIaB, measures both high plasma pressure and high plasma confinement relative to the provided plasma current, field, and minor radius. High performance has also been demonstrated relative to empirical H-mode confinement scalings.41
Existing devices such as DIII-D and National Spherical Torus Experiment Upgrade (NSTX-U) provide opportunities to address key science that may extend the fusion performance in ITER. These include continuous operation while avoiding transients, maximizing normalized core performance at high bootstrap fraction, and developing exhaust solutions. Collaborations with the European JET experiment on the planned deuterium-tritium (D-T) campaign should prove very valuable for validating physics models in high performance core plasmas consistent with an ITER-like beryllium wall and tungsten divertor. Collaborations with EAST and KSTAR, and in the future JT-60SA, enable model testing in long pulse mode. JT-60SA in particular provides a compelling opportunity to test model predictions made in advance of an experiment, as will be necessary in ITER. U.S.-developed models such as EPED and TGLF have been applied in initial predictive studies of JT-60SA, and these models, coupled to others, can be applied in comprehensive predictions of performance and performance optimization in JT-60SA, which can then be tested as the machine enters high-power operation.
Improvements in high-performance computing hardware and algorithms, including the advent of exascale computing, also provide an opportunity to extend the fusion performance of ITER. During the next few years, advanced scientific simulations will be able to incorporate higher resolution time and space scales and explore the complex couplings between electrons, ions, and global collective plasma physics. Advances in analytic theory, including development of new formalisms capable of efficiently treating the full range of scales associated with both magnetohydrodynamics and gyrokinetics, will lead to improved study of the boundary plasma where equilibrium and turbulence scales overlap. A comprehensive theory
and simulation program including high fidelity multi-scale simulations, reduced models incorporating insight from those simulations, and very fast neural net interpolations of more complex models will enable theoretical understanding to be involved in all aspects from experimental planning to control, design and optimization of the fusion concept. Integrated simulation, moving toward whole device modeling by connecting physics models from the core to the pedestal, boundary, and material interface, provides a timely opportunity for comprehensive planning of burning plasma experiments on ITER.
Continued physics model development and validation using the results from U.S. facilities, DIII-D, NSTX-U, and the international experiments will further increase confidence in ITER predictions. Additionally, experiments using U.S. facilities can develop advanced scenarios, closely guided by validated models, to enable high fusion performance in both inductive and steady-state scenarios. In particular, advanced scenarios, such as the “Super H-mode,” are expected to be achievable on ITER42 and lead to important studies of enhanced fusion performance in ITER and improved confidence in the design of the compact pilot plant.
The United States has also contributed to the development of steady-state scenarios, where the plasma current is largely self-driven, through the “bootstrap” effect, and does not rely on inductively driven current. A strong focus on high performance steady-state scenarios on ITER will also advance understanding of the high-performance steady-state plasma, where the heating source (primarily fusion produced alpha particles) and current drive (primarily pressure-driven bootstrap current) are both strongly coupled to confinement and energetic particle physics.
The move from existing fusion experiments to experiments at the scale of a power plant, like ITER, brings important changes to the underlying plasma and atomic physics. The ratio of the ion gyroradius to the device size, called ρ*, will be much smaller than on existing devices. This is true because the reactor’s minor radius, a, is larger, the magnetic field is relatively stronger, and the gyroradius decreases with magnetic field. Numerous physical processes in both the core and edge plasma are expected to have important dependencies on ρ*, and in some cases these dependencies are still not well understood.
In addition, a fusion device at the scale of a power plant is influenced by the atomic processes associated with the fueling and penetration of neutral particles at the plasma edge. In existing magnetic confinement experiments, the neutrals penetrate a significant distance into the pedestal region of the confined plasma, and the pedestal region is directly fueled by recycling neutral atoms, which are ionized upon penetrating into the plasma. This neutral penetration depth relative to the width of the pedestal is denoted by λ*, which scales roughly in proportion to the inverse
product of plasma density and plasma size. However, the large scale and relatively high density of a device like ITER prevent neutral penetration into the pedestal region. The study and understanding of plasma confinement processes under conditions of small ρ*, short neutral penetration depth λ*, and high plasma pressure and fusion power density is a scientific frontier of burning plasma research.
Pedestal Physics at the Power Plant Scale
High confinement regimes are necessary for high fusion gain in ITER and in compact fusion pilot plants. These regimes are characterized by stabilization of plasma turbulence and radial shear of the plasma flow near the pedestal. The increased plasma confinement associated with the high-confinement operating mode, called the “H-mode,” occurs when the plasma shear flow is self-generated near the edge pedestal region. How the turbulence suppression and plasma confinement properties scale with decreasing ρ* is an important question to answer at the scale of a fusion power plant. Because the plasma flow shear stabilization takes place across a region which scales with the turbulent eddy size, there is no significant ρ* scaling of the pedestal width in front-propagation paradigm. To date, observations on existing devices in carefully controlled dimensionless experiments have found no significant ρ* scaling of the pedestal width.43 However, it remains to be learned whether ρ* physics may enter at very small ρ* values which may impact the pedestal pressure in ITER and other future fusion power devices. Ongoing model tests on existing devices,44,45 particularly at higher field, can shed light on this and further extend comparisons such as that shown in Figure 2.2(b). Detailed testing of physics models in the early stages of ITER operation should improve understanding before entering the burning plasma phase of ITER. The role of direct fueling by neutrals in determining the pedestal density and density profile also plays an important role. An ongoing series of high-density experiments on DIII-D is currently exploring this physics, and experiments on ITER will provide critical data.
Predicting the L-H Transition
The physics associated with the initial formation of the edge transport barrier (known as the “low-to-high” or L-H transition) remains poorly understood. Empirical scaling of the heating power required for the transition suggests high power will be needed in an ITER-scale device because the needed L-H power is observed to increase strongly with magnetic field, density, and plasma surface area at an aspect ratio of A ~ 3. Being able to predict the L-H power threshold precisely is important for development of high-performance scenarios on ITER. Both the empirical scaling and some proposed physics models, such as those that include the role of ion orbit loss,46 suggest ρ* scaling. In addition, changes in fueling and
recycling, such as divertor leg length, have been observed to affect the L-H transition, suggesting a role for neutral penetration physics. Results from NSTX suggest a possible collisionality dependence. Three-dimensional (3D) fields, such as those ITER plans to use to control ELMs, also are known to impact the L-H transition. Ongoing studies on existing devices such as DIII-D and NSTX-U can further explore the role of geometry, fueling and 3D fields on the L-H transition. The early operation phase of ITER, including operation at reduced field to reduce the L-H threshold power, will provide valuable data to further develop physics understanding, and burning plasma experiments on ITER will address L-H transition physics at high power density and reactor-like physics parameters.
Understanding the Plasma Power Exhaust Width
As heat leaves the confined plasma it is quickly lost to the divertor across a narrow layer known as the scrape-off-layer width, λq. As shown in Figure 3.3, observations of the scrape-off-layer width shows that it scales inversely with the poloidal magnetic field, or λq ∝ q/AB, where q is the plasma safety factor (inversely proportional to the twist rate of the magnetic field), A is the plasma aspect ratio, and B is the magnetic field strength. At high q or low magnetic field, λq is wide and spreads the escaping plasma heat across a wider surface within the divertor. In contrast, at low q or higher magnetic field, λq becomes narrow and the peak power dissipation in the divertor region increases. Empirical scaling of λq suggests it narrows with increasing poloidal magnetic field, and may therefore be very small, as small as 1 mm in ITER.47 Physics models based on drift scaling also predict narrow λq for ITER.48,49 However, physics models incorporating micro-instabilities suggest that as ρ* gets smaller, gradients across the scrape-off-layer will become high, strongly driving turbulence which transports heat radially, broadening λq.50,51 These models predict much larger values of λq ~ 5 to 8 mm for ITER. A similar broadening is predicted for high plasma current (2 MA) NSTX-U plasmas, and validation of these calculations on both NSTX-U and ITER can lead to a full understanding of this important physics at reactor scale.
Core Heat, Particle, Momentum and Impurity Transport
In the plasma core, the competition between turbulent and neoclassical transport is expected to be strongly affected by ρ*. In particular, transport of high Z impurities such as tungsten are often dominated by the neoclassical pinch in existing devices, but turbulent transport is predicted to play a much more important role in ITER.52,53 ITER is also expected to operate much closer to the critical gradient for micro-instabilities which has important effects requiring large scale, long duration simulations to explore.54 Low ρ* is also predicted to weaken coupling
between equilibrium and turbulent eddy scales, making transport more local. The physics of momentum transport, and the mechanism for generation of so-called intrinsic rotation, are relatively poorly understood, and ITER results at low ρ* and low injected torque will clarify important physics. In addition, ITER will operate with a variety of hydrogen isotopes, including protium, deuterium, and tritium, and so will provide important data on the isotope effect on transport.
Tokamaks are observed empirically to encounter a density limit that scales roughly with the ratio of the plasma current to the square of the minor radius.55
The physics mechanism for this density limit is not well understood though a number of theories have been proposed, such as the radiative island theory.56 On existing devices, the plasma density limit is also associated with high collisionality in the plasma edge. However, at the high pressure expected in ITER, it should be possible to encounter the density limit at low collisionality, providing insight on the role of plasma collisionality in density limits. This physics is very important for ITER and reactor performance, because state-of-the art performance projections predict that fusion performance increases with density even to densities above the empirical limit observed in present devices.
Alpha Particle Transport
Fusion plasmas contain energetic ions created by fusion reactions in the plasma. In present devices, energetic ions can be produced by injecting beams of high-energy neutrals into the plasma that, after ionization, subsequently heat it; alternatively, radio frequency waves accelerate an energetic ion population. In a fusion power device, fusion reactions between deuterium and tritium produce energetic alpha particles (also known as the nuclei of helium gas) at 3.5 MeV. The alpha particles are trapped within the plasma and heat the plasma and sustain its temperature. Besides heating the plasma, the energetic ions may drive instabilities that degrade their confinement. At their worst, alpha-particle instabilities may cause beams of energetic ions to be lost to the first wall and erode wall materials. Figure 3.4 illustrates the orbit of an energetic ion in the DIII-D device that resonates with a 115 kHz Alfvén eigenmode. Even if the ions remain in the plasma, their radial redistribution can degrade plasma performance. Near-term research concentrates on acquiring the ability to predict which instabilities will be unstable, the nonlinear coupling among multiple simultaneous modes, what their consequences will be, and developing methods to mitigate adverse consequences. Reduced models and high-fidelity physics simulations using high-power computing play a key role. Development of simpler models that are less expensive computationally enable efficient prediction and model validation. Experiments on both DIII-D and NSTX-U will help to validate these simulations. Experiments on ITER are a vital step in developing confidence in these projections at reactor scale. One difference between ITER and present-day devices is that the energetic-particle orbit is a smaller fraction of the machine radius. This is predicted to change the spectrum of unstable waves, which may alter their saturation mechanism. Using neutral beams, these predictions can be tested in the nonnuclear phase.
A major goal of ITER is to study the behavior of alpha particles, for the first time, in plasmas with dominant fusion-product heating. The plasma will be a highly coupled, nonlinear system, as changes in alpha confinement will alter the heating profile which may, in turn, alter the production of alpha particles. High
power deuterium-tritium experiments in ITER constitute a crucial experimental test of our ability to predict alpha-particle behavior in a power plant.
A burning plasma is a highly coupled, nonlinear system. The energy and momentum that alpha particles provide to the plasma will affect the plasma current, transport, and stability in a manner that will alter the density and temperature of the burning fuel, which, in turn, changes the rate of fusion reactions. Achieving and controlling the burning plasma state involves understanding these internal
nonlinear couplings and the self-organized plasma configuration resulting from self-heating. A burning plasma is fundamentally different from plasmas that have been created in all research facilities to date, and ITER will provide the first opportunities to study, sustain, and control a burning plasma.
In preparation for ITER experiments, many techniques to control the coupled plasma state can be explored beforehand. Couplings between the current profile, transport properties, and macroscopic stability are already operative in existing experiments. In general, plasma control hinges on three elements: sensors that measure the plasma state, actuators that alter the plasma state, and algorithms that direct the actuator response to the sensor input. In the case of plasma control, the goal is often to suppress an instability or, if that is not possible, to mitigate its impact. Instabilities that can damage the first wall of the plasma chamber include plasma current disruptions, ELMs, neoclassical tearing modes, and Alfvén eigenmodes. Exploration of these instabilities benefit from the development of new sensors and instrumentation that can be operated in the radiation environment of a burning plasma, testing of control actuators, such as localized microwave heating, and the development of new control algorithms and real-time control systems.
Previous sections of this chapter describe the importance of burning plasma research, explain why continued participation as an ITER partner is important to U.S. fusion energy research, and describe how ITER participation will inform the design of a compact fusion pilot plant as a new element of the U.S. magnetic fusion program. The benefits of continued U.S. participation in ITER are compelling. Planning and preparation for ITER experiments are the major focus of the U.S. fusion research program. The development of national expertise in burning plasma science requires the participation of experts and will not result from mere study of the research achievements of other nations. ITER construction is more than half complete, and the first plasma experiments are expected to begin in less than 10 years. ITER is the only existing experiment with a mission to explore burning plasma physics at the power plant scale. ITER is also an ambitious research project that integrates multiple advanced technologies and combines the scientific and engineering expertise, industrial capacity, and financial resources of several nations. As an ITER partner, the United States receives full benefit from the technology that will establish the feasibility of fusion while providing only a fraction of the financial resources.
Even though ITER is recognized as “the best candidate today to demonstrate sustained burning plasma,” the committee was tasked to provide long-term guidance for the scenario in which the United States is not a partner in ITER. Because
any strategy to develop magnetic fusion energy requires study of a burning plasma, a decision by the United States to withdraw from the ITER project would require a new approach to study a burning plasma and a new focus to the U.S. fusion research program. Currently, there is no mature burning plasma experiment as an alternative to ITER. The design, construction, and licensing of such an alternative to ITER would require significant development by the U.S. program. Because participation in ITER and the ITPA aids international cooperation and collaboration in fusion energy science, withdrawal from the ITER project will also require a new approach to avoid isolation from the international fusion energy research effort.
Irrespective of whether the United States remains an ITER partner, the committee recommends that the United States should start a national program of accompanying research and technology leading to the construction of a compact pilot plant at the lowest possible capital cost and the production of electricity from fusion. In this way, the committee’s long-term strategic guidance is generic and applies to both scenarios.
All previous strategic plans reviewed by the committee call for construction and operation of a burning plasma experiment and the demonstration of scientific and technical feasibility prior to construction of a facility capable of electricity production. This committee concurs with this assessment. A burning plasma experiment is a critical next step toward the realization of fusion energy, and the science and technology gained from a burning plasma experiment, like ITER, will answer key questions needed to design a compact pilot plant. With access to a burning plasma experiment, scientists will have the means to answer fundamental questions pertaining to energetic alpha particles created by fusion reactions, plasma transport processes in fusion reactor conditions, methods to control of plasma transients, divertor science, and the integrated scenarios that simultaneously test the requirements for stability, confinement, fuel purity, and compatibility with plasma-facing components needed for a fusion energy source.
If the United States wishes to maintain scientific and technical leadership in fusion energy development and undertake a program toward a compact pilot plant, national expertise in burning plasma science needs to be developed through hands-on operational participation and scientific study by U.S. fusion scientists.
For the scenario with the United States remaining an ITER partner, research toward the second goal of compact, attractive fusion power generation will build upon the ITER experience and focus on high power density plasmas, and the integration of core and edge physics in the regime required for a compact fusion pilot plant. Using research results from the DIII-D, NSTX-U, and Alcator C-Mod programs, from advances in fusion confinement theory and simulation, a follow-on experiment as an intermediate step toward a high-pressure, compact pilot plant need not be a fusion nuclear facility, resulting in significant savings in facility cost, research access, instrumentation, and project schedule.
However, if the United States were to withdraw from ITER, it would need to design and construct a larger and more ambitious research facility with a capability to explore burning plasma science with deuterium-tritium operation. The direct study of high-gain burning plasma physics and access to research opportunities necessary to evaluate long plasma duration and burning plasma control methods are central long-term goals of the U.S. program. As an alternative to ITER, the addition of an expanded fusion nuclear program for the high-power density burning plasma facility would very be expensive for the United States to undertake without international support, and it would delay progress in the field.57 Such an expanded fusion research program, however, would be critical for directly addressing the physics of a strongly coupled burning plasma, and addressing the key challenges discussed above.
A decision by the United States to withdraw from the ITER partnership would make international collaboration more difficult. Nevertheless, the United States would need to explore other avenues for collaboration and international cost-sharing, such as the engagement of the United States in the physics design for the China Fusion Energy Test Reactor. Such international collaborations, particularly in the event of a U.S. withdrawal from ITER, would benefit the United States, provided a vibrant national fusion program can provide value to offer to other collaborating nations.
In summary, in both scenarios, whether the United States remains an ITER partner or not, the committee recommends the United States should start a national research and technology program leading to the construction of a compact pilot plant at the lowest possible capital cost. However, without ITER participation and since the long-term objective would still be the compact pilot plant, the primary initial focus of a U.S. program would be a high-power density research tokamak with expanded capabilities allowing study and operational experience with a burning plasma. Systems engineering and conceptual design studies for this machine, building on results from new experiments on DIII-D and NSTX-U, and state-ofthe-art theory and simulation, should begin as quickly as possible in response to a U.S. decision to withdraw from ITER in order to provide the necessary means to study burning plasma science and technology as part of a new focus to the U.S. fusion research program and maintain progress toward the long-term development of commercial fusion power.
Major advances in both experimental and theoretical fusion science provide a strong foundation for rapid progress toward fusion development. Progress in theory and computation of fusion plasmas, coupled with well-diagnosed flexible U.S. experiments, have increased confidence in predictions of burning plasma
performance, and clarified requirements for an attractive fusion energy source. Realization of a burning plasma is essential to developing a complete understanding of the strongly coupled system and advancing the technology needed to make fusion energy attractive.
ITER is a burning plasma experiment and the critical next step in the development of fusion energy. Methods to control plasma stability, plasma interactions with first wall materials, plasma confinement, and fusion power output will be tested. Theoretical predictions of energetic particles produced by fusion reactions and methods to sustain a burning plasma will be explored and validated. Equally important are gains in fusion engineering science and industrial capability that result from ITER fabrication and operation.
Finding: The scientific and technical benefits from the study and operation of ITER are compelling and critical to the development fusion energy for the United States.
ITER is a large and ambitious project that integrates multiple advanced technologies and combines the scientific and engineering expertise, industrial capacity, and financial resources of several nations. As an ITER partner, the United States receives full benefit from the technology that will establish the feasibility of fusion while providing only a fraction of the financial resources.
Finding: ITER plays a central role in U.S. burning plasma research activities and is currently the only existing project to create a burning plasma at the scale of a power plant. Because the ITER partnership is the central focus in the large international effort to develop fusion energy, the United States significantly benefits from participation in the ITER partnership.
Recommendation: Because the scientific and technical benefits from ITER are compelling and because ITER is the only existing project to create a burning plasma at the scale of a power plant, the Committee recommends that the U.S. government fulfill its commitment to construct and operate ITER as the primary experiment in the burning plasma component of its long-term strategic plan for fusion energy.
Recommendation: A near-term focus of the U.S. DOE Office of Fusion Energy Sciences research program should maximize the scientific and technical benefits from its partnership in a burning plasma experiment.
Finding: Advances in understanding toroidal magnetic confinement, plasma control, and integrated solutions to whole-plasma optimization point to
improvements beyond the ITER baseline and show how careful design and simulation can be used to lower the cost and accelerate fusion energy development.
Recommendation: In the longer term, the U.S. DOE Office of Fusion Energy Sciences research program should encourage the development and testing of burning plasma scenarios on ITER that contribute to reliable operation of a compact fusion pilot plant.
Finding: If the United States withdraws from the ITER project, the national research effort would be significantly disrupted, U.S. researchers would be isolated from the international effort, and any benefit from sharing the cost in critical burning plasma studies and fusion demonstration would be eliminated.
Finding: Without ITER, the United States would need to design, license, and construct an alternative means to gain experience creating and controlling an energy-producing burning plasma. The scale of research facilities within the United States would be more costly. The achievement of electricity production from fusion in the United States would be delayed.
Recommendation: Nevertheless, if the United States decides to withdraw from the ITER project, the U.S. DOE Office of Fusion Energy Sciences should initiate a plan to continue research that will lead toward the construction of a compact fusion pilot plant. This should include the construction of an alternative means to study the burning plasma regime and an alternate method to engage in the international effort in the pursuit of its long-term objective for fusion demonstration.
1. See previous studies listed in Appendix D and, for example, National Research Council (NRC), 2004, Burning Plasma: Bringing a Star to Earth, The National Academies Press, Washington, DC, https://doi.org/10.17226/10816.
2. R.J. Goldston, 2012, Heuristic drift-based model of the power scrape-off width in low-gas-puff H-mode tokamaks, Nuclear Fusion 52:013009.
3. S.I. Krasheninnikov, A.S. Kukushkin, W. Lee, A.A. Phsenov, R.D. Smirnov, A.I. Smolyakov, A.A. Stepanenko, and Y. Zhang, 2017, Edge and divertor plasma: Detachment, stability, and plasma-wall interactions, Nuclear Fusion 57:102010.
4. 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.
5. L.M. Giancarli, M. Abdou, D.J. Campbell, V.A. Chuyanov, M.Y. Ahn, M. Enoeda, C. Pan, et al., 2012, Overview of the ITER TBM Program, Fusion Engineering and Design 87:395.
6. J.-Ph. Girard, P. Garin, N. Taylor, J. Uzan-Elbez, L. Rodríguez-Rodrigo, and W. Gulden, 2007, ITER, safety and licensing, Fusion Engineering and Design 82:506.
7. 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.
8. 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.
9. S.J. Zinkle and L.L. Snead, 2014, Designing radiation resistance in materials for fusion energy, Annual Review of Materials Research 44:241.
10. NRC, 2004, Burning Plasma.
11. S.J. Zinkle and L.L. Snead, 2014, Designing radiation resistance in materials for fusion energy, Annual Review of Materials Research 44:241.
12. NRC, 2004, Burning Plasma.
13. T. Omori, M.A. Henderson, F. Albajar, S. Alberti, U. Baruah, T.S. Bigelow, B. Beckett, et al., 2011, Overview of the ITER EC H&CD system and its capabilities, Fusion Engineering and Design 86:951.
14. R. Cesario, L. Amicucci, A. Cardinali, C. Castaldo, M. Marinucci, L. Panaccione, F. Santini, et al., 2010, Current drive at plasma densities required for thermonuclear reactors, Nature Communications 1:55.
15. O. Sauter, M.A. Henderson, G. Ramponi, H. Zohm, and C. Zucca, 2010, On the requirements to control neoclassical tearing modes in burning plasmas, Plasma Physics and Controlled Fusion 52:025002.
16. M. Thumm, 2014, Recent advances in the worldwide fusion gyrotron development, IEEE Transactions on Plasma Science 42:590.
18. Madia and Associates, 2013, Final Report of the 2013 ITER Management Assessment, Contract-ITER/CT/13/4300000830, October 18, https://www.documentcloud.org/documents/10319342013-iter-management-assessment.html.
19. ITER Organization, 2018, ITER Research Plan within the Staged Approach, ITR-18-003, September 17.
20. N. Sauthoff, 2017, “Perspectives from the US ITER Project,” presented to the committee on August 29.
21. U.S. Department of Energy, 2017, Project Execution Plan for U.S. ITER Subproject-1, DOE Project No. 14-SC-60, Office of Science, FES, Washington, DC, January.
22. See slides 56-59 of U.S. Department of Energy, 2017, Project Execution Plan.
23. U.S. Government Accountability Office, 2014, Fusion Energy: Actions Needed to Finalize Cost and Schedule Estimates for U.S. Contributions to an International Experimental Reactor, Report to Congress, GAO-14-499, June.
24. Y. Liu, A. Kirk, L. Li, Y. In, R. Nazikian, Y. Sun, W. Suttrop, et al., 2017, Comparative investigation of ELM control based on toroidal modelling of plasma response to RMP fields, Physics of Plasmas 24:056111.
25. 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.
26. R.C. Duckworth, L.R. Baylor, S.J. Meitner, S.K. Combs, D.A. Rasmussen, M. Hechler, T. Edgemon, et al., 2012, Development and demonstration of a supercritical helium-cooled cryogenic viscous compressor prototype for the ITER vacuum system, Advances in Cryogenic Engineering 57A-57B:1234-1242.
27. 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.
28. 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 96-97:113.
29. 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.
30. 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 Superconductors 26:4203305.
32. C.C. Baker, 2002, An overview of enabling technology research in the United States, Fusion Engineering and Design 61-62:37.
33. B. Chen, X.Q. Xu, T.Y. Xia, N.M. Li, M. Porkolab, E. Edlund, B. LaBombard, J. Terry, J.W. Hughes, M.Y. Ye, and Y.X. Wan, 2018, Progress towards modeling tokamak boundary plasma turbulence and understanding its role in setting divertor heat flux widths, Physics of Plasmas 25:055905.
34. C.S. Chang, S. Ku, A. Loarte, V. Parail, F. Köchl, M. Romanelli, R. Maingi, et al., 2017, Gyrokinetic projection of the divertor heat-flux width from present tokamaks to ITER, Nuclear Fusion 57:116023.
36. J.M. Park, J.R. Ferron, C.T. Holcomb, R.J. Buttery, W.M. Solomon, D.B. Batchelor, W. Elwasif, et al., 2018, Integrated modeling of high normalized beta steady state scenario on DIII-D, Physics of Plasmas 25:012506.
37. 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.
38. 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.
39. W.M. Solomon, P.B. Snyder, A. Bortolon, K.H. Burrell, A.M. Garofalo, B.A. Grierson, R.J. Groebner, et al., 2016, Exploration of the Super H-mode regime on DIII-D and potential advantages for burning plasma devices, Physics of Plasmas 23:056105.
40. 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,” 27th IAEA Fusion Energy Conference.
41. C.T. Holcomb, J.R. Ferron, T.C. Luce, T.W. Petrie, J.M. Park, F. Turco, M.A. Van Zeeland, et al., 2014, Steady state scenario development with elevated minimum safety factor on DIII-D, Nuclear Fusion 54:093009.
42. P.B. Snyder, W.M. Solomon, K.H. Burrell, A.M. Garofalo, B.A. Grierson, R.J. Groebner, A.W. Leonard, et al., 2015, Super H-Mode: Theoretical prediction and initial observations of a new high performance regime for tokamak operation, Nuclear Fusion 55:083026.
43. M.N.A. Beurskens, T.H. Osborne, P.A. Schneider, E. Wolfrum, L. Frassinetti, R. Groebner, P. Lomas, et al., 2011, H-Mode pedestal scaling in DIII-D, ASDEX Upgrade and JET, Physics of Plasmas 18:056120.
44. 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, https://doi.org/10.1088/1741-4326/aa6416.
46. S. Ku, C.S. Chang, R. Hager, R.M. Churchill, G.R. Tynan, I. Cziegler, M. Greenwald, et al., 2018, A fast low-to-high confinement mode bifurcation dynamics in the boundary-plasma gyrokinetic code XGC1, Physics of Plasmas 25:056107.
47. T. Eich, A.W. Leonard, R.A. Pitts, W. Fundamenski, R.J. Goldston, T.K. Gray, A. Herrmann, et al., 2013, Scaling of the tokamak near the scrape-off layer H-mode power width and implications for ITER, Nuclear Fusion 53:093031.
48. R.J. Goldston, 2012, Heuristic drift-based model of the power scrape-off width in low-gas-puff H-mode tokamaks, Nuclear Fusion 52:013009.
49. 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, https://doi.org/10.1088/1741-4326/aa6416.
50. C.S. Chang, S. Ku, A. Loarte, V. Parail, F. Köchl, M. Romanelli, R. Maingi, et al., 2017, Gyrokinetic projection of the divertor heat-flux width from present tokamaks to ITER, Nuclear Fusion 57:116023.
51. B. Chen, X.Q. Xu, T.Y. Xia, N.M. Li, M. Porkolab, E. Edlund, B. LaBombard, J. Terry, J.W. Hughes, M.Y. Ye, and Y.X. Wan, 2018, Progress towards modeling tokamak boundary plasma turbulence and understanding its role in setting divertor heat flux widths, Physics of Plasmas 25:055905.
52. E.A. Belli and J. Candy, 2018, Impact of centrifugal drifts on ion turbulent transport, Physics of Plasmas 25:032301.
53. F.J. Casson, C. Angioni, E.A. Belli, R. Bilato, P. Mantica, T. Odstrcil, T. Pütterich, et al., 2015, Theoretical description of heavy impurity transport and its application to the modeling of tungsten in JET and ASDEX upgrade, Plasma Physics and Controlled Fusion 57:014031.
54. G.M. Staebler, N.T. Howard, J. Candy, and C. Holland, 2017, A model of the saturation of coupled electron and ion scale gyrokinetic turbulence, Nuclear Fusion 57:066046.
55. M. Greenwald, 2002, Density limits in toroidal plasmas, Plasma Physics and Controlled Fusion 44:R27.
56. D.A. Gates and L. Delgado-Aparicio, 2012, Origin of tokamak density limit scalings, Physical Review Letters 108:165004.