While construction and operation of a burning plasma experiment is a critical next step to the development of magnetic fusion energy, further research is needed to improve and fully enable the fusion power system. The interconnected science and technology for the hardware that surrounds the plasma, converts fusion power into useful heat, and breeds and recovers tritium will need to be developed for a commercial fusion power source. Additionally, research and innovations are needed to reduce the size and cost of the fusion power system and to attract industries and utilities to pursue fusion energy-based electricity production for the United States.
This chapter describes the research needed to advance magnetic fusion toward an economical energy source beyond what will be conducted with the International Thermonuclear Experimental Reactor (ITER). It is motivated by recent science and technology achievements that make feasible a research strategy that may shorten the time and reduce the cost required to develop commercial fusion energy. Instead of following ITER experiments with a large, and potentially costly demonstration fusion power plant (DEMO),1 research opportunities are described for a reduced cost pathway to fusion development. Industry can now produce commercial quantities of high-temperature superconducting tapes that have the potential to be used in very high field fusion magnets that will reduce the size needed to magnetically confine a burning plasma. Advanced burning plasma operating scenarios have been investigated that may allow uninterrupted plasma confinement at high fusion power density and with very low recirculating power requirements. Emerging new technologies have been identified, like additive and advanced manufacturing and novel tritium processing technology, that may reduce the cost and improve
the reliability of fusion nuclear components and systems. Taken together, these advances create a new approach to the development of commercial fusion energy, called the “compact fusion pilot plant.” The goal of a compact fusion pilot plant is to use advances in fusion science and technology to address mission elements of previous pathways to a fusion DEMO but in a facility having a small size and the lowest possible capital cost.
This pathway to a compact fusion pilot plant merges fusion science and technology research with burning plasma studies conducted in parallel with ITER. This pathway converges in the time-frame after burning plasma demonstration in ITER (near 2040) and enables the construction of a compact pilot plant to begin near this time. When ITER operation establishes key burning plasma science and when the accompanying research program has simultaneously advanced burning plasma science, materials science, fusion nuclear science, and engineering science, the design of the compact fusion pilot plant can be finalized and construction commence. As identified in the committee’s interim report,2 without the research accompanying ITER aimed to improve and fully enable the fusion system, the United States risks being overtaken as our partners advance the science and technology required to deliver fusion energy.
Advancing magnetic fusion energy by developing a compact fusion pilot plant involves risks. A compact fusion pilot plant requires developing operation scenarios for sustaining high-power density burning plasma with the plasma exhaust capability required for compact fusion. The engineering design and fabrication of large bore, high-field superconducting magnets for fusion needs to be established. Long-lifetime materials will need to be developed and qualified for use in the compact fusion pilot plant. Tritium science and fusion breeding blanket development needs to be developed sufficiently for integrated nonnuclear testing of prototypes that can serve as the basis for blanket components that will ultimately be installed in the compact pilot plant. Finally, to be successful, a detailed system engineering effort is needed to guide a “pre-pilot-plant” research program toward construction of a low capital-cost fusion pilot plant through cost-effective research and development (R&D). The only way to retire these risks is to carry out the needed R&D.
This chapter is organized into four parts. First, previous pathways to commercial fusion are summarized, and the mission needs for commercial fusion power are listed. Second, the compact fusion pathway is presented including the research needed to establish and sustain burning plasma conditions at high power density, with low recirculating power, and using low-cost research facilities. Third, fusion nuclear science and technology is discussed followed by description of a “pre-pilot-plant” research program that will increase the technical readiness of the superconducting magnets, fusion nuclear materials, fusion nuclear components, and enabling technologies that are needed to design and fabricate a compact fusion
pilot plant. Possible facilities and partnerships that further accelerate magnetic fusion energy are also described.
Finally, the committee presents its findings and recommendations for a national program of accompanying research and technology leading to the construction of a compact pilot-plant which produces electricity from fusion at the lowest possible capital cost. A detailed finding itemizing the technical and scientific support motivating a new national research program leading to the construction of a compact pilot plant is followed by four recommendations to resolve five critical research needs, initiate planning for the construction of new research facilities, and the adoption of a two-phase approach to its plans for the compact pilot plant so that scientific and technical risks can be addressed cost-effectively.
The long-term objective of previous pathways to commercial fusion energy is the design and operation of a DEMO. A DEMO would produce electricity, operate routinely, and eliminate all technical barriers to the commercialization of fusion power. Previous designs for a fusion DEMO are for large, high-power facilities that build upon the ITER design with technologies needed to produce net electricity from fusion for long periods of time.
Two approaches have been proposed to reduce technical risks of a DEMO facility and to satisfy fusion nuclear licensing requirements. One approach builds an intermediate facility, called a fusion nuclear science facility (FNSF), prior to DEMO. An FNSF would establish the materials and component database in the real fusion in-service environment before proceeding to a larger DEMO facility. The other approach would design and build a large DEMO facility with a “slow start” where necessary fusion nuclear components would be installed over time in a staged approach.
Figure 4.1 illustrates the timelines for various pathways to DEMO and to a commercial fusion power plant developed for Korea, Europe, Japan, and China and previously considered for the United States. All pathways have the same goal of operating a DEMO near 2050. The United States has not adopted an official strategy toward a DEMO, but the U.S. Department of Energy’s (DOE’s) Fusion Energy Sciences Advisory Committee (FESAC) has recommended research leading to the construction of an FNSF prior to DEMO.3,4 The FNSF would be the first part of a two-step approach to a fusion power plant that would commence in parallel with the study of burning plasma in ITER. The second step is a DEMO device. Variations of the FNSF differ in size and capability, but the strategic argument for the FNSF is to understand the behaviors of materials and fusion components prior to the pursuit of electricity production in the larger DEMO device. Several options for an FNSF have been considered in the United States, depending how closely the
FNSF approaches the characteristics of a power-producing fusion energy device and that vary the shape of the magnets that confine the toroidal burning plasma.5
The pathways considered by Korea and China proceed directly from ITER to DEMO, but the Korean and Chinese DEMO facilities would be designed and operated in two phases. Higher-power and longer-duration fusion power would occur in the second phase after the first phase established burning plasma operating scenarios and some fusion nuclear technologies. For Japan and Europe, construction of DEMO would begin near the end of ITER operation in order to start DEMO operation at the earliest possible date.6 The European Union fusion roadmap
Horizon 20207 requires the European Union DEMO to be based on mature technologies and use reliable regimes of operation extrapolated from the ITER facility. All of the pathways listed in Figure 4.1 include a DEMO device built at a size and power level larger than ITER.
The various international pathways to commercial fusion energy address equivalent technical R&D needs. In the United States, these research needs were most recently described in the 2007 report of the DOE FESAC committee, Priorities, Gaps and Opportunities: Towards A Long-Range Strategic Plan for Magnetic Fusion Energy,8 and the 2009 report of the DOE Office of Science Research Needs Workshop for Magnetic Fusion Sciences.9 Scientific and technical questions were organized into three broad themes defined in terms of the knowledge required prior to DEMO. These were as follows:
- Creating predictable high-performance steady-state plasmas sufficient to create and sustain a burning plasma meeting all of the conditions required for practical production of fusion energy.
- Taming the plasma-material interface sufficiently to design and build robust material components that interface the hot plasma in the presence of energetic fusion neutrons.
- Harnessing fusion power sufficiently to design and build reliable systems that convert fusion energy to useful forms of energy and breed a self-sufficient supply of tritium fuel.
These reports recommended an integrated program of research to address all of the scientific challenges of fusion energy including fusion engineering, materials science, and plasma physics.
As presented in Chapters 2 and 3, the worldwide focus on the construction and preparation for burning plasma experiments on ITER has resulted in significant progress in the first theme to create predictable high-performance steady-state plasmas sufficient to create and sustain a burning plasma. Important progress has also been made toward the second theme of plasma-materials interaction needed to design and predict performance of the ITER divertor. Progress in the third research theme, the technical readiness to harness fusion power, is least well developed. In a future nuclear fusion power facility, virtually every major component will require novel materials compatible with fusion’s energetic neutron production. Additionally, fusion nuclear components will need to safely and efficiently fuel, exhaust, breed, confine, extract, and separate unprecedented quantities of tritium.10
The development path and technical missions to progress beyond ITER toward a commercial fusion power plant were recently summarized based on a detailed system engineering study for an FNSF. In addition to the study of burning plasma and control, the development of fusion technology is categorized into three steps.
Fusion-relevant neutron exposure of materials and nonnuclear testing of fusion components are tests that can be prepared in the short term. The purpose of the second step, the FNSF, is to produce environments similar to a fusion power plant for evaluating heat removal and the impact of energetic fusion neutrons. Integration of those systems related to electricity production, including high efficiency heating, current-drive, thermal management, breeding, and gas processing, are deferred to the last step, the large DEMO device.
For all of these development pathways, the long-term objective is the design and operation of a DEMO. The pathways are distinguished by the number and size of research facilities needed to achieve the integrated technical efficiency to produce electricity from fusion energy.
New technologies, such as high-temperature superconducting magnets and advanced materials, and new achievements in burning plasma science now make smaller, less-costly research devices possible and fusion R&D more affordable and attractive. This has two strategic implications:
- In place of a single-step approach to a large DEMO, the opportunity exists today to start the interconnected science and technology research leading to construction of a compact pilot plant and, ultimately, the production of electricity with a device with significantly lower cost.
- In place of two-step approach with a FNSF having a mission limited to fusion component development followed by a second DEMO facility, recent science and technology advances suggest a compact fusion pilot plant might be built at a cost comparable to previous FNSF designs while being ultimately capable of demonstrating the overall systems efficiency required to produce electricity.
Relative to previous pathways to commercial fusion energy, a compact fusion pathway targets smaller device size, lower capital cost, and shorter development steps. A research approach that minimizes the capital cost of major research facilities is a less costly pathway to the demonstration of fusion electricity. A research approach that includes the production of electricity as an objective motivates, from the beginning, efforts to optimize overall systems efficiency as an essential part of the evaluation of the compact fusion pilot plant.
This compact fusion pilot plant would be a pre-commercial research facility with a burning plasma at its core and surrounded by a blanket to capture fusion heat and neutrons. In addition to the production of fusion electricity, the pilot plant would ultimately be capable of uninterrupted operation for weeks and produce
tritium, the heavy isotope of hydrogen in fusion fuel, from lithium-containing blankets that surround the plasma. As a pilot plant, its purpose will be learning, but the knowledge obtained would be sufficient for the design of the first commercial fusion power systems.
The pilot plant pathway to commercial power has been examined favorably by several studies for many years.11Figure 4.2 compares one DEMO design to an example of a pilot plant designed with superconducting magnets, B = 6 T, small size, R = 4 m, and relatively low fusion power, 500 MW.12 By comparison, the K-DEMO (R = 6.8 m, B = 7.4 T) design would be much larger and costlier but produce 2,000 MW of fusion power.13 Besides decreasing the cost of fusion research facilities, a smaller, compact fusion pilot plant addresses market trends from U.S. electrical utilities seeking lower capital-cost capacity additions, shorter construction times, and more flexible siting options that result from smaller power-plant footprint.14
Another advantage of a compact approach is that it reduces the financial risk in combing two fusion research missions in a single device, provided that sufficient progress is made to retire technology risks. The initial pilot plant operation would demonstrate net-electric equivalent performance in a compact fusion system, focusing on integrated core/edge performance, assessing plasma material interactions, demonstrating tritium pumping, limited breeding, safe handling, and extraction. This initial phase would not include long-pulse fusion power production and would not demonstrate self-sufficient tritium production. The second phase of the pilot plant would seek near continuous operation, allow for materials/component testing with neutron fluences approaching power-plant levels, and provide integrated blanket testing. Upon success of this second phase, the compact fusion pilot plant studies would have reduced both the economic and technical risks for fusion energy-based electricity production and will motivate further involvement from industries and utilities in the United States.
The scientific and technical opportunities for developing a compact fusion pilot plant are described below. The important relationship between compact size and high magnetic field is discussed along with the engineering challenges associated with high-temperature superconducting magnets and the plasma science and materials science challenges associated with continuous operation and high-power plasma exhaust. Fusion nuclear science and technology, opportunities for new research facilities, and opportunities for expanded international collaboration are also discussed further.
The relationship between high magnetic field and compact fusion confinement has been appreciated for decades.15,16 Fusion power density is proportional to the square of the plasma pressure, 0.08 P2 (MW m-3), where P is the plasma pressure in atmospheres. Because the maximum achievable plasma pressure increases in proportion to the magnetic pressure, fusion power density increases rapidly with increasing magnetic field, in proportion to B4. Furthermore, the confinement time in a magnetized plasma scales with the number of gyro-radii across the plasma. At fusion temperatures, this number scales with the product of the plasma size and the magnetic field, R × B. Thus, if the magnetic field doubles, the fusion power density can increase as much as 16-fold while maintaining the same fusion power gain at half the size.
The magnet technology with the highest possible magnetic field generally determines the size of magnetic fusion devices. A smaller compact fusion power source becomes possible with the availability of higher field magnets provided the compatible plasma components can be developed to control, sustain, and utilize the high-power density burning plasma.
Early compact magnetic fusion experiments were built with copper magnets, where high magnetic field strength could be reached for short pulses using novel engineering to deal with very large magnetic forces. The world’s record for volume-averaged plasma pressure was set in September 2016 in the Alcator C-Mod device at the Massachusetts Institute of Technology (MIT), which operated at high magnetic field and physics-based optimization. (See Figure 2.3 in Chapter 2.) Two of the three burning plasma experiments discussed in the 2004 report of the the Burning Plasma Assessment Committee, Burning Plasma: Bringing a Star to Earth,17 were designed with copper coils in order to reach burning plasma conditions during short pulses in devices that would be smaller, but less capable, than ITER.
Copper coils require large currents and consume large power when operating at high magnetic field. For this reason, continuous high-field operation for fusion is not possible with copper coils. All power-producing magnetic fusion concepts considered today require superconducting magnets to reduce recirculating power and significantly improve the efficiency and economics of electricity production.18 At the time ITER was designed, the highest-field superconductor available was niobium-tin (Nb3Sn). To reach burning plasma conditions, the ITER superconducting magnets are the largest ever built, with a total magnetic stored energy of 51 GJ, a nominal mechanical stress of 600 MPa, and a magnetic field strength of B = 5.3 T within the plasma.19 Today’s opportunity for compact magnet fusion energy results from the potential for high-field superconducting magnets. New high-temperature superconductors may make possible fusion magnets that can achieve fusion gain and power equivalent to ITER but at a significantly lower size and cost.
Figure 4.3 illustrates the sizes for several proposed superconducting next-step burning plasma devices at A ~ 3, including ITER. Various DEMO designs have magnetic fields comparable to ITER and achieve higher power at larger size. The large EU-DEMO device (R = 9 m) requires a system of superconducting magnets having a magnet energy about four times larger than ITER. In contrast, recent FNSF conceptual designs20 (which are not designed to achieve the same power gain as ITER) and the proposed ARC (affordable, robust, compact) device21 are much smaller fusion devices that operate with higher magnetic field strength but smaller stored magnetic field energy than in ITER.
Two engineering challenges will need to be addressed to increase the magnetic field in a compact fusion device. The first challenge is the development of superconductors with higher critical current density and higher critical magnetic field strengths. With the new commercial availability of rare-earth barium copper oxide superconducting materials deposited on steel tapes, the prospects for meeting this challenge appear good. The second challenge is the mechanical design of the high field tokamak. At high magnetic fields, the magnetic force becomes extremely large. Within larger devices, like ITER, the central force of each toroidal
field magnet is supported by wedges between adjacent coils and includes a large ohmic heating solenoid. The CFETR,22 EU-DEMO,23 K-DEMO,24 and J-DEMO25 designs listed in Figure 4.3 are examples. By comparison, with a compact fusion device, innovative mechanical configurations are needed to support the large forces on the superconducting magnets. The magnetic field strength used for conceptual design studies of compact fusion devices are limited by the high stresses within the materials needed to support the magnets.26 However, DOE’s 2018 FESAC report Transformative Enabling Capabilities for Efficient Advance Toward Fusion Energy27 concluded a “consensus within the magnet community that existing high strength stainless steel and superalloy materials are adequate for projected fusion requirements” (p. 28). Additionally, smaller and higher-field designs incorporate multiple load paths, called bucked and wedged, to support the magnet forces.28
In a fusion device built using superconductors, the field from the toroidal magnets does not change in time. However, the poloidal field magnets that are used to start up the burning plasma and control its position and shape will need to change in time. The development of high-field superconductors for the poloidal field coils
require the solution of challenging technical issues related to time-varying magnet currents. These engineering issues will influence the magnet configuration, aspect ratio, and location of poloidal field magnets within the device.
While new high-temperature superconductors require engineering developments to reach the high magnetic field strength for compact fusion, they may provide additional benefits as compared to low-critical temperature superconductors. For example, demountable toroidal field magnets may allow simplified maintenance, as proposed in the ARC compact fusion device.29 Because the new superconductor is deposited in thin layers on flexible steel tapes, magnet winding and manufacturing may be superior than the wind-and-react method required for Nb3Sn. Low aspect ratio compact design has been proposed both in the United States30 and by Japanese scientists.31,32 In each case, size reduction resulting from high-field compact tokamak reactors is significant when compared with conventional reactor designs. Further mechanical design is required to assure adequate structural integrity and compatibility with high fusion power density, a compact tritium breeding blanket, effective neutron and gamma shielding, and noninductive operation.
Finding: Although additional research, including magnet engineering research, is needed to demonstrate the viability of the compact pathway to fusion power, the combination of new high-field superconducting magnet technology with advanced burning plasma science is a significant opportunity to decrease the size and cost of a magnetic fusion power system.
Power handling is one of the crucial challenges for magnetic confinement fusion, and the compact pathway to economical fusion path may either heighten or help to mitigate this challenge.
The escaping plasma exhaust that contacts the divertor will need to be configured to avoid unacceptable thermal damage or erosion of plasma-facing components due to sputtering. Compared with present-day devices, the escaping plasma heat flux will increase in commercial fusion power devices and require design improvements to properly cool the divertor and first-wall and allow continuous uninterrupted operation. Additionally, plasma relaxation phenomena, called edge localized modes (ELMs) driven by instabilities localized to the plasma edge region,33 should be controlled or prevented in order to avoid first-wall material damage.
Variations of the divertor approach taken for ITER may also prove applicable for the compact fusion pilot plant. The divertor is constructed from water-cooled blocks with plasma-facing tungsten armor. The very high heat flux from the plasma is prevented from direct contact with the divertor armor by volumetric radiation and recombination, called a “detachment.” Uncertainties exist how divertor detachment
can be achieved in next-step fusion devices, and the results from ITER experiments will provide crucial measurements with which to test predictive models. Nevertheless, the power flow of the escaping plasma exhaust is observed to be narrow, with a width inversely proportional to the poloidal magnetic field, Bp ≈ B/Aq, where A is the plasma aspect ratio and q is the plasma safety factor (inversely proportional to the helical twist of the magnetic field lines). A fusion device with the same peak power flux to the divertor as expected in ITER will require that the thermal power escaping the plasma surface scale in proportion to RqA/B and likely also require seeding with impurity atoms (like argon, neon, or other radiating atoms) that scale in proportion to qA/B.34 Self-consistent models for detached divertors indicate the impurity fraction required for detachment scales in proportion to the ratio of the escaping fusion power to the poloidal magnetic field. For this reason, a compact pilot plant operating at lower power but higher magnetic field may be preferred to a higher-power fusion system because the compact design allows lower impurity concentration within the detached divertor region.
As the magnetic field increases, the fusion device becomes more compact and produces less total power at equivalent gain. Figure 4.4 illustrates these trends, which show how the limits imposed by the divertor heat flux and fusion power change as the magnetic field, B, and the device size, R, change while keeping the
aspect ratio at A ~ 3. Because the exhaustible heat is proportional to the plasma surface area while the device cost is proportional to the magnetic energy within the plasma volume, a compact fusion approach can be a cost-effective approach for magnetic confinement fusion from the power handling point of view as well as by reducing total capital investment for a power plant.
Another challenge is controlling the transient heat flux due to ELMs. ELMs are edge relaxation phenomena believed to be driven by the peeling/ballooning mode.35 Several methods have been demonstrated to suppress ELMs, including the application of 3D magnetic perturbations in DIII-D36 and steady operation without any transient ELM events. These include the quiescent “QH-mode”37,38 associated with edge harmonic oscillation to enhance edge transport, “I-mode”39 associated with the weakly coherent mode, and improved energy confinement with reduced particle confinement. Eliminating or reducing the transient heat flux due to ELMs significantly lengthens the lifetime of the divertor armor. Because of the higher magnetic field of a compact fusion reactor, operation can occur at reduced plasma current and higher q, which has been found to assist in ELM-free operating modes.
Finally, ongoing research, discussed in Chapter 2, is now evaluating promising advanced divertor configurations such as the Snowflake divertor,40 Super-X-divertor,41 and small-angle slot (SAS) divertor.42 A recent low aspect ratio high-temperature superconductor (HTS) FNSF/Pilot plant design by Menard43 successfully showed that the long-leg and Super-X-divertor can be implemented for the outboard divertor leg in a compact fusion system. The SAS divertor, first tested in DIII-D, is a compact divertor geometry in which stable detachment has been obtained experimentally, showing that a slot with a V-shaped corner is very promising for a compact fusion power source.
Finding: While methods to remove heat from the divertor and reduce material erosion due to plasma sputtering remain active research areas, current understanding of divertor scaling indicates that the compact fusion pathway at higher field and lower total power may benefit power handling solutions for fusion energy.
A commercial fusion power source will need to produce electricity continuously for several months at a time. A critical goal for a compact fusion pilot plant is to demonstrate uninterrupted operation and to establish the basic science and technology needed for commercial fusion power.
Figure 4.5 shows diagrams of fusion plasma performance and plasma pulse duration. The fusion energy gain is roughly proportional to the fusion “triple
product,” n(0) × τE × Ti(0) equal to the product of the central plasma density, the characteristic time for energy loss, and the central temperature of the fusion fuel. The fusion gain and triple product generally increase quickly with the size and magnetic field strength of the fusion containment device, n(0) × τE × Ti(0) ∝ (R B)3 at fusion temperatures and densities. The highest fusion performance has so far been achieved using large copper magnets (e.g., the JT-60U, JET [Joint European Torus], and TFTR [Tokamak Fusion Test Reactor] experiments). In these devices, long-pulse operation requires reduced toroidal magnetic field, resulting in reduced plasma performance. Studies of long duration plasma confinement have become possible with superconducting magnets such as the Tore Supra,44 EAST, and K STAR experiments. The longest plasma duration was achieved in TRIAM-1M,45 but TRIAM-1M could not simultaneously achieve high fusion performance due to its small size, R = 0.84 m. Two superconducting stellarators, the W-7X (R = 5.5 m, B = 2.4 T) and LHD device (R = 3.9 m, B = 3.0 T), have, respectively, two and three times higher energy superconducting magnet systems than in the EAST tokamak, and they are sufficiently large to allow study of fusion performance at pulse lengths comparable to the superconducting tokamaks. As shown in Figure 4.4, the technical achievement of high fusion gain and long plasma duration will be tested for more than 1 minute in the JT-60SA device,46 expected to begin plasma experiments in 2020 in Japan, and for several minutes in ITER and additional facilities, like FNSF, have been proposed to extend fusion pulse lengths from days to weeks.
To achieve efficient steady-state operation, most of the plasma current will need to be self-generated instead of driven by injection of particles or electromagnetic waves. Indeed, the plasma current in high-gain fusion experiments have been driven by magnetic induction, and the plasma current in the highest performance experiments in ITER also results from induction.
Self-generated current is called “bootstrap current,” and the fraction of plasma current sustained by the bootstrap effect scales in proportion to the plasma poloidal beta, βp, where the plasma poloidal beta measures the ratio of the plasma energy to the magnetic energy of the plasma current. The poloidal beta is proportional to the normalized plasma beta and the plasma safety factor, βp ≈ 0.03 q A βN. The high poloidal beta regime occurs when q βN ~ 25 (when the plasma aspect ratio is near 3) and the self-generated bootstrap current fraction, fBS, reaches 100 percent. This mode of operation is called the advanced tokamak operation.47 JT-60U48 sustained high bootstrap current fraction (~75 percent) discharges for 7.4 s; TCV achieved 100 percent bootstrap current; and DIII-D49 achieved a fully noninductive high bootstrap current fraction (~ 60 percent) operation around βN = 3.5, which is twice the normalized pressure in the ITER reference scenario. These high poloidal beta regimes also have improved energy confinement, making high poloidal beta regimes promising modes to operate a commercial fusion power device and a fusion DEMO. Higher magnetic field in the compact fusion pathway allows higher
fusion power density while also operating with high poloidal beta and lower plasma current. As shown in Figure 4.4(a), a compact fusion pilot plant (shown as the star symbol and labeled “HT-SC Compact”) operating at high magnetic field and within an advanced fusion confinement regime would produce significant fusion power while operating at lower plasma current and full bootstrap current. Similar levels of fusion power were described by the HTS-ST pilot plant design developed by Menard and co-authors.50
As was described in the Chapter 3 section on “Extending ITER Performance,” just as recent advances in theory and simulation provide opportunities to significantly extend ITER performance, these advances also improve the prospects for a compact fusion pilot plant. Experiments at U.S. research facilities have achieved record plasma pressure, demonstrated advanced operating scenarios, like the Super H-Mode, and tested methods to improve divertor power handling. Simulations of integrated core-pedestal performance have already been used to optimize steady-state scenarios on DIII-D and make initial predictions for ITER. These advances optimizing fusion performance for ITER combine with advances in magnet technology to motivate the reduced size and capital cost of the compact fusion pilot plant.
Finding: While significant progress is needed to achieve uninterrupted operation of a high-performance fusion confinement device, the higher magnetic field in the compact fusion pathway, when combined with advanced operating scenarios, appears to allow operation at high fusion power density, high poloidal beta, and high bootstrap current fraction more easily than other pathways to commercial fusion power.
Research aimed at developing a fusion-based power plant has, to date, focused mainly on the plasma physics and confinement itself, including the plasma, the divertor and first wall, as well as the magnets and heating systems. These are all necessary features of a power plant, and significant progress has culminated with the construction of ITER. But, these aspects alone are not sufficient for a fusion energy pilot plant. The attractiveness of a fusion system, in terms of economics as well as safety and environmental considerations, is mainly determined by the materials and design of systems that will extract the fusion power in order to convert it to electricity and sustainably generate, or breed, tritium. At present, these systems for the divertor and first wall and integrated blanket are at a very low technology readiness level, and significant fusion nuclear science and technology research is needed to provide the technological foundation required for the design and construction of a compact fusion pilot plant.
As noted earlier, a number of transformative enabling capabilities illustrate how rapidly developing technology advances may speed progress and lower the cost of fusion energy development. These transformative capabilities include advanced materials, high-temperature and/or high field magnets, and tritium processing, all of which offer the potential to significantly increase the technical readiness level to enable construction and mitigate risks toward the initial operation of the compact, advanced fusion energy test facility. In particular, the 2018 Transformative Enabling Capabilities report noted that “magnet systems are the ultimate enabling technology for magnetic confinement fusion devices. Advances in the development of superconductors that operate at higher temperature and higher field, referred to as HTS, present a potentially transformative opportunity to significantly enhance the performance and feasibility of a large variety of magnetic confinement devices.”51
In addition to advancement in superconducting magnet technology, the Transformative Enabling Capabilities report described recent advances in “novel synthesis, manufacturing and materials design are providing the most promising transformation enabling technologies in PMI [plasma-material interaction] and nuclear fusion materials to enable fusion energy for the future.”52 The report also identified a number of potentially transformative developments in tritium extraction and processing that show promise for a commercial fusion reactor, but will require further R&D, in addition to demonstration in a compact fusion energy test facility.
In the following, opportunities to advance magnetic fusion energy through engineering science, materials science, and fusion nuclear technology are described. Taking advantage of these research opportunities would increase the technical readiness needed for the design and construction of a compact fusion pilot plant.
Magnetic fusion energy requires access to the highest possible magnetic fields that can be maintained with superconducting magnets. In general, the highest magnetic fields achievable in practical large-bore superconducting magnets have been limited by the properties of the superconducting materials themselves. The two main well-established options are low-critical-temperature superconductors. Niobium-titanium is used in Tore-Supra, EAST, K STAR, and the two superconducting stellarators. The higher-field Nb3Sn superconductor is used for ITER. Since the mid-2000s, a new class of HTS materials was successfully used for large-scale applications: rare-earth barium copper oxides (REBCO) tapes and Bi-2212 round strands. Figure 4.6 shows a plot of the critical current density versus magnetic field for low-temperature superconducting and HTS magnets. REBCO has performance anisotropy parallel and perpendicular to the magnetic field but has high critical current density at high field. BSCCO (bismuth strontium calcium copper oxide) superconductors, such as Bi 2212 and Bi 2223, have lower current density at
high field, but each has its own merit compared to REBCO. Today, REBCO tapes are commercially available from at least seven manufacturers (AMSC, Fujikura, Shanghai Superconductor, SuNAM, SuperOx, SuperPower, and SWCC Showa).53
REBCO superconductors offer the potential to carry sufficient current density for magnet applications at fields up to 100 T.54 REBCO has been successfully used to reach fields over 40 T in solenoids55 and has demonstrated engineering current densities exceeding 10 A/mm2.56 This is an order-of-magnitude higher current density compared to conventional low-temperature superconductor fusion magnets, and has generated considerable interest from the private sector. Indeed, the recent MIT-Commonwealth Fusion Systems collaboration has announced it is moving forward with plans to design, construct, and test a large-bore magnet using HTS that is central to the SPARC design of a compact fusion device with a peak field of 23 T.57
At present, HTS magnets have not been tested in the configuration or at the scale needed for fusion experiments. Key challenges include magnet quench detection and protection, conductor stress/strain management, and characterization of radiation resistance. There is consensus within the magnet community that existing high-strength stainless steel and super-alloy materials are adequate for projected fusion requirements. Although REBCO is extremely stable in operation, quench detection is a significant issue due to very slow propagation of the normal zone. The status and future directions of high magnetic field science, including the potential for fusion energy applications, were assessed in 2013.58 Important goals of the Magnet Development Program of the DOE Office of High Energy Physics are to investigate fundamental aspects of magnet design that lead to substantial performance improvements and cost reduction.59 Progress in the use of HTS magnets for fusion energy applications have been reported; however, additional engineering research is needed to gain full-size operating experience with fusion magnets, including magnet quench detection and protection, demountable coil development and testing, conductor stress/strain management, and characterization of radiation resistance.
Finding: While additional R&D is needed to establish the technical basis for large high-field HTS magnets, the growing industrial capability to produce HTS conductor, opportunities to partner with industry and other DOE program offices, and the rapid progress in HTS magnets may enable significant reductions in the size of magnetic fusion devices and support the compact lower-cost pathway to fusion development.
Overcoming the performance challenges of materials and structures in the fusion energy environment is a daunting challenge but yet is critically important to realize the promise of fusion as a practical energy source. Development of materials, components, and structures for any complex engineered system generally occurs in a series of steps, proceeding from relatively simple single-variable experiments to very complex, fully integrated multiple-variable tests. Commercially available additive manufacturing tools exist today, and the Transformative Enabling Capabilities report noted the rapid advancement of capabilities including mixed material printing, multi-scale features, and large component manufacture.60
The first step on the development path begins with screening experiments that are performed under carefully controlled laboratory conditions. These experiments establish the basic mechanical and physical properties, chemical compatibility, and fabrication and joining technologies of candidate materials. If satisfactory results are obtained, more complex experiments are performed in order to identify materials
that perform well in partially integrated or multiple-variable tests. The final stage of material development involves fully integrated experiments with prototypical test sections that are carried out in an environment combining the appropriate nuclear, chemical, thermomechanical and magnetic fields necessary to establish lifetime and performance limits. The ultimate goal is to develop an extensive material database that includes chemical and mechanical properties, and physics-based models to describe the material performance within a fusion power system.
Fortunately, as noted previously in this chapter, significant advances in materials synthesis, manufacturing, and materials design provide promising transformative capability to develop the integrated divertor, first wall, and blanket materials to survive the fusion environment. Advanced manufacturing offers the potential to locally tailor the material microstructure within a single component by varying the manufacturing process parameters, and to create complex structures that were simply not possible with conventional methods. Further, advanced manufacturing will lead to complex lattice or composite structures for lightweight yet strong components that could optimize cooling channels to increase heat removal capability. While these tailored materials and complex structures require testing in the neutron and high heat flux environments to characterize their properties for optimizing the material microstructure and component design, advanced manufacturing offers substantial opportunities to develop complex hierarchical composites and self-healing materials to enable breakthroughs in emergent fusion materials.
Complex composites and complex solid phase alloys also have promise to demonstrate radiation tolerance, or even radiation resistance that could be transformative for fusion divertor, first wall, and blanket materials. This use of complex composite geometries may significantly improve the performance of refractory materials for fusion energy applications, and recent developments in continuous fiber and laminate composites demonstrate promising thermal fatigue and thermomechanical response in laboratory experiments.61 As well, advanced manufacturing techniques could lead to the development of hybrid material systems that encompass the self-healing and renewable characteristics of liquid metals with a solid-state matrix that could provide a beneficial plasma-facing component with optimal heat removal. Furthermore, significant developments in SiC/SiC composites have been demonstrated in the ceramic gas turbine industry, and have transformative potential for nuclear fusion pilot plant components. The Transformative Enabling Capabilities report also called attention to novel tritium extraction technologies proposed for liquid metal breeding blankets and for plasma-facing components. Tritium science, extraction technologies, and fuel processing are critical challenges for fusion energy systems, and significant challenges will need to be overcome including the need to develop effective tritium permeation barriers to prevent release of sizable quantifies of tritium.62
Finding: While advanced manufacturing and complex material component design have transformative potential, research is required to move beyond the early stage of developing these alloys and composites. This includes radiation effects, chemical compatibility and corrosion, unknown response to plasma material interactions and tritium permeation, and component performance and degradation in the complex neutron, plasma material, and thermal-mechanical loading conditions. Studies will need to proceed from relatively simple single-variable experiments to very complex, fully integrated, multiple-variable tests.
Continued development of the technologies used to heat, control, and measure burning plasma will be needed as the level of fusion power increases and the duration time of the plasma approaches steady-state conditions. Additionally, engineering strategies need to be developed for subsystem and component reliability and efficient remote maintenance of fusion nuclear components.
Development of a compact fusion pilot plant with higher magnetic field strength will require development of a new generation of higher frequency sources for radio waves and millimeter waves and also technology research to extend the capabilities and efficiency of higher-power launching apparatus and transmission systems. Similar to the shorter duration needs for ITER, a compact fusion power plant will require continuous injection of high-power electromagnetic waves for plasma heating, plasma profile control, pre-ionization/startup, and plasma current drive.
Diagnostics have two important roles in preparation for a compact pilot plant. First, new measurement techniques are needed to provide the data to validate the physical models and simulation codes used to extrapolate to future devices. As noted in the 2007 report Plasma Science: Advancing Knowledge in the National Interest,63 “quite simply, we cannot understand what we cannot measure.” Hence, diagnostic development is a key building block of the predictive understanding that will enable a compact fusion device, and these diagnostics will have to function reliably in the harsh neutron radiation environment near the burning plasma. New diagnostics are needed to replace those techniques incompatible with continuous production of fusion power. Measurements are needed to address the new issues associated with burning plasma experiments, such as detection of the alpha-particle population in ITER and erosion of material surfaces in the pilot plant.
In addition to measurement instruments, the plasma and fusion energy system will be actively controlled. Sensors will provide input to algorithms that control actuators in real-time to maintain the plasma in the desired state. Advances in machine learning and mathematical control theory may enable effective control of fusion plasmas despite imperfect knowledge of the plasma state.
Finding: While continued R&D is needed to adapt enabling technologies for use in the compact fusion pathway, ongoing advances in heating, diagnostics, and control under way in support of ITER provide confidence that these technologies can be developed for higher-power longer-pulse fusion devices.
The integrated first wall and breeding blanket of a fusion reactor will need to operate at high temperature to ensure efficient conversion of fusion power into electricity in addition to generating tritium in the blanket. The tritium generated in the blanket, as well as the unburned tritium fuel from the plasma exhaust, will need to be efficiently extracted and processed for re-introduction to the plasma. Although the performance, safety, and economics of a fusion system depend on successful power extraction and tritium breeding, these systems have a low technical readiness and significant uncertainty regarding performance and operating limits requiring technology advancement to extrapolate handling increasing tritium concentration and temperature, as illustrated in Figure 4.7.
Testing in prototypic environments is incredibly challenging. Because of the unique operating environment surrounding a burning plasma, the development of power handling and breeding blanket systems involves a complex set of interactions among numerous disciplines including materials, thermo-mechanics, thermofluids, magnetohydrodynamics, and corrosion chemistry. The complex, extreme fusion environment needed to test these components is not currently available. In addition to the science within each discipline, it is likely that synergistic effects among disciplines will be discovered when components are tested in a fusion nuclear environment.
The breeding and recovery of tritium as it is processed raises a number of safety concerns to protect workers, the public, and the environment. Tritium is highly mobile, and can readily permeate through metallic components, especially at elevated temperatures. Tritium will have to be accurately tracked to assure safety and nonproliferation. The grand challenges of tritium require improved scientific understanding of many interconnected phenomena including permeation, radiolytic chemistry, surface science and kinetics, liquid metal magnetohydrodynamics, and mass transfer. Systems and processes should be developed that can efficiently and safely continuously process tritium at flow rates and quantities beyond current practice.
The United States has developed a potentially attractive family of blanket concepts, in which a dual cooled, lead-lithium eutectic alloy serves as both breeder and coolant. In this concept, the reduced activation ferritic steel integrated first wall and blanket structure have separate gas cooling and thermal- and electrical-insulating inserts based on silicon carbide composites that control the structural material
temperatures at critical interfaces. However, this concept, along with the other solid and liquid blanket concepts that have been proposed, remain relatively immature due to a lack of research and testing capability to assess performance limits. A significant enhancement of research activities is needed to validate blanket concepts and the science and technologies of the tritium fuel cycle prior to constructing a compact fusion pilot plant.
The 2018 Transformative Enabling Capabilities report64 highlighted a number of technologies that show tremendous potential for fusion power development. These include advances in tritium fuel production in breeding blankets involving either the dual-cooled liquid lithium concept or cellular ceramic blankets. Research into tritium fuel production would also address unresolved sintering problems in proposed ceramic pebble based fusion blankets. In terms of tritium extraction, the Transformative Enabling Capabilities report identified recent advances in electrolytic membrane extraction and permeable membrane extraction methods with the potential to efficiently process and extract tritium from liquid metal blankets. The Transformative Enabling Capabilities report also discussed the potential for a super-permeable metal foil pump that would effectively decouple the plasma and
tritium plant operation and thereby reduce the size and inventory of the tritium plant substantially. While the Transformative Enabling Capabilities report noted these potentially transformative tritium fuel cycle technologies, it also noted the relatively low scientific and technical maturity of the fusion blanket system and the systems to manage the full tritium fuel cycle.
Finding: Technical concepts needed to harness fusion energy are ready for design and testing. These concepts, along with innovations and promising new methods to separate and process tritium, will be essential to the development of a compact, lower cost fusion reactor.
Conceptual design studies for various compact fusion pilot plants have concluded that in addition to a burning plasma experiment research is needed to understand the interconnected science and technology for the high-field superconducting magnets and for the fusion components that surround the plasma, convert fusion power into useful heat, and breed and recover tritium. Important optimization is needed to configure the high-field HTS magnets and determine the burning plasma scenarios for high-power density uninterrupted operation at very low recirculating power. The most cost-effective configuration or shape for a compact fusion pilot plant has yet to be determined. Configuration options include divertor geometry, plasma aspect ratio, geometric parameters like elongation and triangularity, and others that may improve the operation of the fusion power system. This “pre-pilot-plant” research program needs to be guided by an interdisciplinary systems engineering effort to reduce the size and cost of the fusion development pathway and to attract industrial engagement in the development of fusion energy-based electricity for the United States. By the time construction of the compact fusion pilot plant takes place, industry should be prepared to deliver fully functioning industrially produced components fabricated using materials that have been appropriately qualified for use in the fusion pilot plant environment.
This section describes strategic elements within a pre-pilot-plant research program. They are intended to provide guidance for a program extending beyond ITER’s burning plasma experiments and leading to the construction of a compact fusion pilot plant having the lowest possible capital cost. Many technical details, research schedules, and issues pertaining to the proper balance between international and national research are beyond the scope of the committee’s study. However, the scientific and technical requirements for the pre-pilot-plant research program are well documented through previous studies of a FNSF or a fusion pilot plant.
Figure 4.8 illustrates the science and technology research elements of a “pre-FNSF” program that would still be necessary for the compact pilot plant.
The pre-pilot-plant science and technology research program will need a coordinated engineering approach to guide a variety of laboratory scale experimental facilities, computational modeling and analysis, and collaborations with the international effort. New nonnuclear program elements will be needed for testing and evaluation of HTS fusion magnets, and for establishing critical design data for simultaneous thermo-mechanical loading in a high-temperature environment. A facility will be needed for evaluating the performance of divertor and plasma-facing components at high temperature with representative particle and heat fluxes and compatible with high-performance and high-pressure fusion confinement. The mission requirements for this facility integrates the science and technology from the high-performance core to the divertor and expands upon the mission requirements for the Divertor Test Tokamak proposed in both the 2015 FESAC report
on Plasma Materials Interactions and for construction in Italy.65 The experimental demonstration of integrated uninterrupted high-power density magnetic confinement regimes is needed in order to validate the configuration and operation space for a compact fusion pilot plant. A new national research facility is probably needed to provide this. If so, the construction of such a new facility would occur after the most critical research missions of the national DIII-D and NSTX-U facilities are completed.
Fusion nuclear testing is essential for materials development and qualification for all aspects of fusion research. Fusion nuclear tests can include both fission and fusion relevant neutron exposure of individual material samples. A larger volumetric fusion neutron source can provide experimental data and initial nuclear testing of integrated first wall and breeding blanket systems as well as critical data on the effect of radiation damage and transmutation effects from a 14-MeV peaked neutron spectrum. While the strategies adopted in previous FNSF and DEMO pathways required qualification of all materials near the fusion core to the neutron fluence they will experience, a staged-approach to the operation of the compact fusion pilot plant may have simpler materials qualification requirements in a first-stage because of lower neutron fluences. This two-stage approach reduces the cost and accelerates fusion demonstration in the compact fusion pathway.
The key elements of the pre-pilot-plant research strategy, in addition to the burning plasma science and technology that will be learned from ITER operation, are the following:
- Systems engineering for a compact fusion pilot plant,
- Advanced materials modeling for fusion technology,
- Testing of large-bore, high-field HTS magnets for magnetic fusion,
- Developing long-lifetime materials for fusion,
- Advancing tritium science and blanket technologies,
- A fusion neutron irradiation facility for prototypical materials qualification,
- Demonstrating sustained high-power-density fusion plasmas with optimized plasma exhaust configuration for compact fusion, and
- Continued development of fusion-enabling technologies needed to heat, measure, and control the burning plasma and to safely maintain the components within the compact fusion pilot plant.
Options for a compact fusion pilot plant are at a preconceptual design stage, and iteration of the engineering design is needed to evaluate configuration options for reaching the optimum design goals. Systems engineering involves the integration of multiple science and technology realms including burning plasma science,
configuration optimization and development of HTS fusion magnets, a compact blanket, licensing procedures, tritium processing technology, and enabling technologies like efficient plasma heating and current-drive systems. The design of the pilot plant should include both optimization of the burning plasma configuration and also considerations for efficient and low-cost balance-of-plant (BoP) systems needed for operating the plant. Experience from ITER operation will be important input to the design optimizations regarding plasma initiation, divertor science, burning plasma control, and other supporting fusion technologies. Examples include experience with ITER’s tritium-handling system, cryogenic systems, and measurement and control systems.
A systematic study is needed to explore configuration options for the compact fusion pilot plant. It should identify research needs and facility needs that can be addressed in the pre-pilot-plant program. Operational scenarios will need to be developed and supported by validated modeling incorporating state-ofthe-art advances in coupled core-edge plasma simulations. Metrics for evaluating design options should include the capability for uninterrupted operations, very low recirculating power, demonstration of fusion electricity, a staged approach to tritium self-sufficiency, and a justified estimate for the construction cost. At the end of the concept design study, focused engineering design activities should commence with significant involvement of industry. Industrial experience could be developed by industrial participation in testing of component prototypes prior to the decision to construct a compact fusion pilot plant.
The United States has made significant advances in multi-scale modeling of plasma materials interactions and high-energy neutron induced degradation of structural materials.66,67,68 These multi-scale models attack the complex materials degradation issues from both a “bottom-up” atomistic-based approach and a “top-down” continuum perspective, and they focus on the hierarchical integration of kinetic processes for species reactions and diffusion to model microstructure evolution over experimental timescales. The simultaneous use of both an atomistic and continuum approach has furthered the development of scale-bridging or multi-scale integration, and has led to fundamental insight into helium–hydrogen synergies controlling tungsten PMI as well as the long-term microstructural evolution due to radiation damage in structural materials. However, it is important to note that these emerging modeling capabilities are in the early stages of development, and continued research activities are required to further develop this capability. Central to the development of advanced modeling is to closely coordinate the modeling activities with experimental studies to provide both model validation and guidance for future modeling activities, as well as to design experiments to resolve
specific scientific challenges such as simulating changes in multi-component surfaces and materials that are formed by erosion and re-deposition processes of PFC surfaces and neutron transmutation of structural materials.
The compact fusion pathway requires development of large-bore, high-field HTS magnets. Currently HTS magnet R&D is focusing on commercial production improvements, characterization of conductor performance, and the scale-up and integration of the magnet assemblies and components, such as the cables needed to fabricate full-size fusion-class magnets.
On March 9, 2018, MIT and a newly formed private company, Commonwealth Fusion Systems, announced the start of a staged research effort for fusion experiments and fusion power systems based on advances in high-temperature superconductors.69 Other efforts to develop large HTS magnets include the Magnet Development Program sponsored by the DOE Office of High Energy Physics, the National Institute for Fusion Science at Toki, Japan, the National High Magnetic Field Laboratory (Tallahassee, Florida), which announced the world’s largest magnetic field generated with superconducting solenoid in December 2017, and several European efforts including CERN, CEA (FR), KIT (DE), University of Geneva (CH), University of Twente (NL) and Bruker HTS (DE).
Because of the importance of HTS magnet development for fusion, an opportunity exists to initiate a large-coil HTS test facility modeled after the Large Coil Test Facility (LCTF) that was hosted by the Oak Ridge National Laboratory (ORNL) in the 1980s. Figure 4.9 shows a photograph of the installation of one of six low-temperature superconducting magnets and a schematic showing the cross-section of the LCTF. The international magnet test facility evolved from the Large Coil Project (LCP) imitated through contracts awarded to industrial teams led by GDC (General Dynamics Convair Division), General Electric Company, and Westinghouse Electric Cooperation. Later, the International Energy Agency Fusion Power Coordinating Committee sponsored an international agreement with partners of EURATOM, Switzerland, and Japan. Each of the six coils that were tested had different features but met equivalent magnet requirements. The GDC coil used Nb3Sn, while the other five coils used NbTi as the superconductor. A U.S.-sponsored HTS test facility would likely attract participation from industries from several nations. Such international cooperation, along with the opportunity to partner with private industry within the United States, would be quite useful for accelerating optimizing HTS coil design as well as identifying the most cost-effective manufacturing method.
In addition to a large coil HTS test facility, several important smaller scale components, including a new, very near-term test facility, are needed if the United
States is to play a leading role in superconducting magnet development for fusion. The purpose of such an additional facility is to allow testing of full-size, HTS-jacketed cable samples at high field over a range of temperatures and currents. This HTS sample test facility has been strongly urged by U.S. superconducting magnet researchers. It would be much less costly than a full coil test facility, and research use may include non-fusion applications of HTS magnets including those for high-energy physics. This relatively small facility has the potential to substantially speed up magnet development since new designs and concepts can be rapidly tested under realistic conditions with small sample sizes. As of now, testing of such cable samples takes place either in National Institute for Fusion Science (Japan) or SULTAN (Switzerland), neither of which has the high field capabilities needed for HTS. Furthermore, the United States has little control over the prioritization of usage time at these facilities of other nations. The other equally important components of a U.S. HTS program include the development of magnet quench detection and protection systems, demountable coil development and testing, conductor stress/strain management, characterization of radiation resistance, and continued R&D of advanced HTS materials.
One of the keys to understanding and controlling plasma material interactions in the divertor and plasma-facing components is to collect data on the evolution of material surfaces during and following long-term plasma exposure. The U.S. program has contributed significantly to this research through PMI experimental studies based on the PISCES facility.70 Additional research is needed to explore very long duration exposure under conditions of high heat and particle flux with a representative geometry. The opportunity exists for a cost-effective linear plasma material interaction test stand that would utilize a high-intensity radio frequency (RF) plasma source for experimental PMI studies. The new linear test stand should be able to operate uninterrupted for many days and expose targets at glancing angles with an applied magnetic field in order to test models of the plasma-material sheath, the acceleration of ionized impurities, emission of secondary electrons, while maintaining ion and neutral fluxes directly relevant to a fusion divertor.71 The knowledge gained from PMI test stands would need to be coordinated with plasma divertor research and inform the selection of divertor materials needed to sustain a high-power density fusion plasma with an optimized plasma exhaust.
A commercial source of fusion electricity requires a closed tritium fuel cycle. In order to fully realize commercial fusion power, the science and technology of tritium need to be developed. This includes all tritium processes and the methods needed for safe operation and benign environmental impact. Virtually all of the technologies related to the tritium fuel cycle are at low technical readiness, with uncertain parameters that describe tritium migration through materials and across interfaces, its retention in bulk solids and liquids, and its retention and behavior in plasma-facing materials. Building technical readiness for fusion power requires a program of materials testing and component performance when exposed to fusion neutrons.72,73
In a fusion power system, the breeding blanket is a critical component that consists of a set of modules covering the interior of the fusion vacuum vessel, capable of supporting a high heat load and an intense neutron flux.74 The breeding blanket will need to (1) assure self-sufficiency of the fusion reactor with regard to tritium, (2) maximize the net efficiency of the power plant, (3) act as a radiation barrier, and (4) act as structural barrier to limit dispersion of the tritium and potential activation products suspended in the coolant.75
ITER provides an opportunity to answer questions regarding tritium processing at large scale within the fueling/exhaust tritium loop. The ITER Test Blanket Module (TBM) program is the first opportunity to study whether tritium can be
generated in a blanket and whether heat can be extracted for power production.76 Although the amount of bred tritium to be handled will be relatively low, due to a small testing area and low plasma duty cycle, the tritium transport processes involved in the TBM program are prototypical of the fusion blanket tritium fuel cycle, including deuterium and tritium neutral ion flux implantation and the resulting transport and permeation under prototypical tokamak plasma-facing surface and operating conditions.
The complexity involved in understanding the behavior of breeding blanket concepts in the fusion environment has led to a rather detailed planning for TBM testing program in ITER.77 During the nonnuclear phase of ITER operation, the TBM testing objectives are to (1) test the electromagnetic response, (2) verify the test blanket system operation in the ITER operating environment, (3) demonstrate the cooling capability and TBM resistance to ITER disruptions, (4) obtain data required for the nuclear licensing process, and (5) confirm that the TBMs do not jeopardize the quality of plasma confinement. During the nuclear phase with operation with deuterium and deuterium-tritium plasma, the main ITER TBM objectives are to (1) test the thermal-neutronic behavior, neutronic-tritium/thermo-mechanic response, and integral performance of the modules, (2) validate the predictions with modeling codes and nuclear data, (3) assess the TBMs thermomechanical behavior, (4) demonstrate the tritium management capability, and (5) demonstrate TBM performance for an extended period of time.
At the ITER Organization, the ITER TBM program will provide an opportunity for testing tritium breeding blanket concepts that would result in tritium self-sufficiency, an extraction of high-grade heat and net electricity production in future fusion reactors. Although all of the TBM arrangements were signed in 2015, the ITER International Organization has not made a final determination on proceeding with the ITER TBM program, and the decision to install the TBM on ITER requires signing additional legal arrangements dealing with all TBM phases through decommissioning.
The United States is not currently a partner in the ITER TBM Program. Whether or not the United States becomes a supporting partner in a TBM activity would need to be determined after further consideration of the schedule and capabilities of the ITER TMB activities, as well as the goals of the new national blanket technology program are defined. Should the United States decide not to seek supporting partnerships on one or more ITER TBMs (assuming they are approved by the ITER International Organization), then the United States will forego access to and experience with the resulting demonstrations and will need to develop this experience through the national blanket technology program. If an ITER TBM partnership collaboration were initiated, the United States, for example, could contribute critical property data such as recombination coefficients, tritium diffusivity in PbLi, MHD mixed convection on tritium transport, and tritium transport within permeation
barrier coatings, through small scale laboratory tritium experiments. Additionally, international collaboration on the various aspects of the tritium fuel cycle and the accompanying areas of fusion nuclear materials, plasma-facing materials, fusion nuclear science, and enabling technologies requires serious consideration. These objectives are critical steps toward developing working breeding blankets for future fusion concepts.
Other opportunities exist in blanket and tritium fuel cycle research that could be conducted in parallel with the ITER activities. Nonnuclear testing can be conducted to advance understanding of thermo-solid and thermo-fluid mechanics, tritium extraction and migration, and some aspects of a fuel cycle development facility. The thermal-fluid testing would incorporate surface and volumetric heating with high magnetic field strength and representative coolant flow rates. This nonnuclear fuel-cycle R&D investigates only deuterium/hydrogen isotopes to advance the required technologies for tritium extraction and processing. Because fusion neutron irradiation will change materials properties, such as barriers to tritium penetration, beneficial synergies should be explored, such as the possibility that enhanced trapping of tritium in solid material due to damage or even the nanostructured particles introduced may enhance the material’s radiation resistance. Access to a neutron irradiation facility is required for initial data on neutron exposure effects on welds, component lifetimes, and reliability.
The fusion nuclear engineering community has long advocated a dedicated fusion neutron irradiation facility for acquiring material irradiation test data in a simulated fusion environment for design, licensing, construction, and safe operation of fusion structural materials, and for benchmarking radiation responses of materials with computational material science. Existing neutron irradiation sources include accelerator-driven sources, like the SINQ Target Irradiation Program at the Swiss Paul Scherrer Institute (PSI), and fission-based sources like the High Flux Isotope Reactor at ORNL. Options for fusion neutron irradiation facilities with a 14 MeV fusion neutron spectrum do not now exist although proposals for such a source include a plasma-based volume neutron source or accelerator-based facilities with either gas, liquid, or solids targets.78 Based on joint Japanese and European research in the framework of the broader approach agreement, the International Fusion Materials Irradiation Facility (IFMIF)–DEMO-Oriented Neutron Source (DONES) is currently being developed for the irradiation of temperature controlled capsules containing test specimens filling a 54 cc volume. The capsule-averaged structural damage rate expected for IFMIF-DONES reaches 15 dpa/fpy.79
Opportunities should be explored to provide a larger-volume fusion irradiation environment that would study and increase the technical readiness level of
breeding blanket components and systems prior to the construction of the compact fusion pilot plant. A low-cost material testing facility has been proposed for near-term fusion materials research.80 Recognizing the need to conserve resources, fusion engineers have proposed either a low-cost linear mirror device for blanket testing or, when mirror research was abandoned in the United States, a driven tokamak.81,82 New results in the gas dynamic trap mirror device at Novosibirsk have revived interest in the mirror option.83
The licensing requirements for the first and second phases of the compact fusion pilot plant need to be determined early in its planning. If irradiation data are only required for structural material samples and welds, then full-size fusion components can be tested as part of the fusion pilot plant research program. If larger volume irradiations are required, then program planning would need to design and construct a cost-effective plasma-based neutron source. In particular, determining whether construction of a volumetric fusion neutron irradiation facility using a magnetic mirror device, or an alternate configuration, would more rapidly advance the technology readiness and lower the cost of the compact fusion pathway needs to be answered as part of the systems engineering studies.
The United States has made significant contributions to the development and understanding of high-performance, steady-state burning plasma operating scenarios that will be used in ITER to demonstrate fusion power gain for pulses of several minutes. However, the compact fusion pathway requires further advances in fusion performance to achieve high power density in a small-sized device with integrated core-edge plasma scenarios capable of uninterrupted operation. This approach is inspired by recent experiments, which were motivated by theoretical studies that demonstrated record performance on Alcator C-Mod (world record plasma pressure) and DIII-D (high equivalent fusion gain) through optimization of the pedestal performance. Separate experiments have demonstrated high-performance steady-state scenarios with no ELMs and the sustainment of high bootstrap fraction scenarios with internal transport barriers at near zero rotation. Further, the impact of reduced aspect ratio and improved boundary shaping on access to regimes of enhanced pedestal confinement and stability will be investigated utilizing NSTX-U operation at higher plasma current and toroidal field. Additionally, theoretical studies have indicated the distinct advantages of high-field side-launched RF in obtaining much higher current drive efficiencies, which if realized could lead to more efficient fusion systems.
The investigation of compact burning plasma conditions will require increased values of normalized size, R × B/T1/2, while maintaining achievable levels of plasma
collisionality and normalized plasma pressure, β ~ nT/B2. Currently the two U.S. national facilities (DIII-D and NSTX-U) are operating at relatively low toroidal fields (B = 2.2 T in DIII-D and B = 1 T in NSTX-U). A device built at the size of DIII-D but with a much larger magnetic field (about B ~ 4 T) and with larger heating power (Pheat ~ 50 MW) would be a potential facility. Opportunities exist to extend the capabilities of the U.S. national facilities. High-performance fully noninductive scenarios can be explored on NSTX-U84 and with planned upgrades of DIII-D85 and NSTX-U.86 A follow-on national facility, which combines the research efforts of the DIII-D and NSTX-U, could provide new advanced tokamak studies closer to burning-plasma conditions.87 This would involve a substantial upgrade, or a new facility.
A key issue in assessing the compact fusion pilot plant is confinement performance scaling at high βN as the normalized device size, R × B, increases. High βN reduces the size of the fusion device and increases the bootstrap current fraction, thereby reducing the current drive power and improving the overall efficiency of the pilot plant. While ITER and JT-60SA will provide some information about performance improvements with increasing βN, plasma edge pedestal models predict that the highest performance, and hence highest fusion power density, will be achieved with optimal shaping of the plasma, including aspect ratio, triangularity, and elongation. DIII-D and NSTX-U are well positioned to provide key information on the choice of the required parameters for a follow-on high-power density research facility needed to establish the science and technology basis for the compact pilot plant. This approach would also enable exploration of the physics benefit of high field operation, particularly the impact on improving plasma confinement, which is synergistic with the HTS research. In fact, practical experience could be gained with smaller HTS toroidal field magnets if the magnet R&D delivers such a capability on the necessary timeline. Considering the importance of long-pulse advanced tokamak operation to the compact fusion pathway, the availability of HTS magnets would be significant, but its use critically depends on the outcome of timely HTS magnet development.
Integrated simulation has long been part of fusion energy research, and recent years have seen a tremendous improvement in the ability to perform simulations with increasing physical fidelity. Theoretical and computational models developed in the United States have substantially improved the ability to predict plasma confinement, control plasma stability, and enhance fusion energy performance.88,89
In addition to integrated simulation making important contributions to developing burning plasma science with ITER, advancement along the compact fusion pathway will benefit from research on integrating more multiple physics, multi-
scale phenomena, and higher-fidelity algorithms into simulations as well as further expanding the simulation capabilities toward prediction and design. These advances in theoretical and computational models, which enable the development of whole device modeling with increasing physical fidelity, have been facilitated by the rapid growth in computational capabilities and the DOE Office of Science initiatives that have encouraged partnerships among fusion, applied mathematics, and computer science researchers.90
High-fidelity whole device simulations will improve physics understanding and further enable optimization of the high density, high performance plasma core coupled to a detached divertor. Continued research investments in whole device modeling should improve the fidelity of kinetic simulations, help develop model hierarchies for incorporating disruption and boundary physics, allow incorporating the interactions of fast particles with thermal plasma waves and instabilities, allow the strong coupling of core transport to sources and instabilities, and improve the coupling of the plasma surface interactions and boundary plasma physics to the pedestal and plasma core. Advances in whole device models, as well as individual physics models, are anticipated to enhance the capability of probabilistic whole device modeling to assess the likelihood of key physical transitions such as those leading to plasma disruptions, to optimize operational conditions for achieving specific fusion energy gains, and to optimize or even increase the divertor heat flux thresholds.
The compact fusion pathway is based on an advanced tokamak configuration operating at sufficiently high normalized beta and sustained continuously, without the need for significant current drive power, because of self-generated bootstrap current. The tokamak configuration has dominated the world’s fusion research program. Fusion performance achieved with the tokamak configuration is superior to other magnetic configurations; however, theory, simulation, and experiments with the stellarator configuration is strongly related to the tokamak and can contribute to the integrated science and technology needed to design the compact fusion pilot plant.
The stellarator concept was invented in the United States, and, in some configurations, the confinement field can be produced entirely by the external magnetic coils. Two large stellarator devices operate with superconducting magnets: the Large Helical Device (LHD)91 in Japan and the Wendelstein 7-X (W7-X)92 stellarator in Germany. A smaller, helically symmetric stellarator, built with copper magnets, is located at the University of Wisconsin in the United States. As shown in Figure 4.5, the plasma duration of the LHD stellarator has achieved pulse lengths comparable to what is expected in ITER, except at much lower fusion gain. Because most of the
rotational transform of a stellarator is generated from external coils, the stellarator magnetic configuration does not require current drive power. Additionally, because the magnetic configuration is controlled primarily with external magnets, a transient confinement loss due to plasma instability should avoid all plasma current disruptions and the generation of runaway electrons.93
W7-X has a high aspect ratio (A > 10) with a large major radius, R = 5.5 m. The W7-X stellarator is called a “heliac,” and HELIAS fusion reactor studies show this type of configuration leads to large fusion power systems.94,95 Currently, the United States is an active collaborator in the W7-X experiment, and stellarator research using W7-X contributes to understanding of all aspects of toroidal magnetic confinement, including energy and particle transport, energetic particle physics, and divertor science and technology. The two large stellarators with superconducting magnets (see Figure 4.5) have achieved operating conditions comparable to the superconducting tokamaks, EAST (China) and KSTAR (ROK), when operating for about 1-minute pulses. These superconducting fusion research facilities provide opportunities to develop the science and technology needed for continuous operation and to learn from the comparison between axisymmetric and three-dimensional (3D) magnetic geometry.
While the long-pulse capabilities of the superconducting LHD and W7-X stellarators contribute to the knowledge needed for the long-pulse operation of the compact pilot plant, stellarators with “hidden symmetry,” either quasi-axisymmetry in the toroidal direction or quasi-helical-symmetry in the helical direction, will also contribute to understanding the role of symmetry in high-performance magnetic confinement and stability through the development and validation of predictive models. As presented to the committee,96 the toroidal magnetic confinement with quasi-symmetry has not been well-tested in the laboratory, but opportunities exist for an optimized intermediate-scale research facility that may impact the national effort to design a compact fusion pilot plant. Such optimization is expected to be useful to generate flow in the quasi-symmetric direction and hence to enhance confinement by flow shear suppression of turbulence. In a quasi-symmetric stellarator, unlike the heliac configuration, plasma currents will change the magnetic structure as plasma pressure increases. Quasi-symmetric stellarators might help validate models to predict fusion performance and improved optimization of a compact fusion pilot plant. Because the only operating quasi-symmetric stellarator is located in the United States and, more importantly, because quasi-symmetric stellarators might lead to improved designs for a compact fusion confinement system, opportunities exist to explore this configuration and validate the physics of 3D magnetic fields and quasi-symmetry for toroidal magnetic confinement.
Fusion Enabling Technologies for Plasma Heating, Current Drive, Measurement and Control and Safe Maintenance of Core Components
The compact pathway to fusion requires higher magnetic fields and long, uninterrupted plasma operation. Continued development of fusion enabling technologies include: higher-frequency gyrotrons and transmission systems in the range of 250-336 GHz, higher power, more compact antenna/wave launching systems for a compact pilot plant, advancement in diagnostic and plasma control systems, and the implementation of remote maintenance of fusion components located near the burning plasma. Before the operational pulse times of fusion research facilities can significantly increase, R&D of the associated plasma heating, plasma current drive, and plasma control systems are required. Examples of such research are described earlier in this chapter, and Chapter 2 described ongoing research activities for robust current drive technologies, like injection of high-frequency “helicon” waves.
This chapter describes the interconnected science and technology research required within an expanded U.S. burning plasma research program in the near and mid-term leading to construction of a compact fusion pilot plant at the lowest possible capital cost.
Finding: Recent advances motivate a new national research program leading to the construction of a compact fusion pilot plant at the lowest possible capital cost that will accelerate the fusion development path. Significant progress has been made to predict and create the high-pressure plasma required for such a reactor. This progress combined with opportunities to develop technologies for fusion, such as high-temperature superconducting magnets and advanced materials, now make a compact device technically possible, affordable, and attractive for industrial participation. This finding is supported by the following:
- Although additional research, including magnet engineering research, is needed to demonstrate the viability of the compact pathway to fusion power, the combination of new high-field superconducting magnet technology with advanced burning plasma science is a significant opportunity to decrease the size and cost of a magnetic fusion power system.
- While methods to remove heat from the divertor and reduce material erosion due to plasma sputtering remain active research areas, current understanding of divertor scaling shows that the compact fusion pathway at higher field and lower total power may benefit power handling solutions for fusion energy.
- While significant progress needs to be demonstrated to achieve uninterrupted operation of a high-performance fusion confinement device, the higher magnetic field in the compact fusion pathway appears to allow operation at high fusion power density, high poloidal beta, and high bootstrap current fraction more easily than other pathways to commercial fusion power.
- While additional R&D is needed to establish the technical basis for large high-field HTS magnets, the growing industrial capability to produce HTS conductor, opportunities to partner with industry and other DOE program offices, and the rapid progress in HTS magnets promise significant reductions in the size of magnetic fusion devices and supports the compact pathway to fusion development.
- While advanced manufacturing and complex material component design have transformative potential, research is required to move beyond the early stage of developing these alloys and composites. This includes radiation effects, chemical compatibility and corrosion, unknown response to plasma material interactions and tritium permeation, and component performance and degradation in the complex neutron, plasma material, and thermal-mechanical loading conditions. Studies should proceed from relatively simple single-variable experiments to very complex, fully integrated, multiple variable tests.
- While continued R&D is needed to adapt enabling technologies for use in the compact fusion pathway, ongoing advances in heating, diagnostics, and control under way in support of ITER provide confidence that these technologies can be developed for higher-power longer-pulse fusion devices.
- Technical concepts needed to harness fusion energy are ready for design and testing. These concepts, along with innovations and promising new methods to separate and process tritium, will be essential to the development of a compact, lower cost fusion reactor.
Based on these findings, the committee offers the following recommendations:
Recommendation: Along with participation in international fusion research, including the ITER partnership, the U.S. DOE Office of Fusion Energy Sciences should start a national program of accompanying research and technology leading to the construction of a compact pilot plant, which produces electricity from fusion at the lowest possible capital cost.
Recommendation: In the near- and mid-terms, the U.S. DOE should resolve critical research needs for the construction of a compact fusion pilot plant:
- Understand the science, production, and control of burning plasma at the scale of a power plant through participation in ITER.
- Demonstrate the science and engineering needed to sustain a magnetically confined plasma having the high-confinement property and compatible plasma exhaust system that are needed for a compact fusion pilot plant.
- Advance high-critical-temperature superconductors and demonstrate the ability to achieve high magnetic fields using large, fusion-relevant coils.
- Expand significantly the U.S. research program in fusion nuclear technology, advanced materials, safety, and tritium and blanket technologies needed to fully enable fusion energy.
- Develop promising innovations in burning plasma science, such as optimized stellarator configurations and innovative approaches for a low-cost fusion irradiation facility, and fusion engineering science that reduce the cost and improve the fusion concept as a source of electricity.
Recommendation: In addition to study of a burning plasma, new research facilities should be built to increase the technical and scientific readiness of critical capabilities needed to construct a compact fusion pilot plant. This will require retiring one or more existing facilities as they complete their most important goals.
Recommendation: In recognition of the significant challenges that need to be addressed for the construction of a compact fusion pilot plant facility capable of electricity production, the U.S. DOE Office of Fusion Energy Sciences plan for a pilot plant should have a two-phase approach. The objectives of these two phases are:
- In the first phase, the pilot plant should be capable of demonstrating fusion electricity production for periods lasting minutes and establish the feasibility of electricity production in a compact fusion system including the assessment of plasma material interactions, tritium safety, pumping, recycling, breeding, and extraction.
- In the second phase, the pilot plant should be capable of uninterrupted operation for many days allowing fusion materials and component testing consistent with a commercial power plant, including full fuel cycle blanket testing.
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