This appendix briefly describes major research facilities in the United States and other nations. These facilities include the DIII-D facility (San Diego, California); National Spherical Tokamak Experiment Upgrade (NSTX-U) facility (Princeton Plasma Physics Laboratory [PPPL]); smaller U.S. confinement research facilities; and the larger devices in Europe and Asia. All fusion research experiments with superconducting magnets are located outside the United States: Experimental Advanced Superconducting Tokamak (EAST, located in China); Korea Superconducting Tokamak Advanced Research (KSTAR, located in Republic of Korea); WEST (formerly Tore Supra, located in France; Figure G.1); Large Helical Device (LHD, located in Japan); and Wendelstein 7-X (located in Germany; see Figure G.2.) The JT-60SA superconducting experiment is under construction in Japan and operation is expected to begin shortly after 2020. Recently, Italy announced its intention to design and construct the superconducting Divertor Tokamak Test (DTT) facility.
DIII-D1,2 is a medium-size tokamak at conventional aspect ratio (R/a ≥ 2.5) that is the largest magnetic fusion research experiment in the United States (Figure G.3). It is a multi-institutional user facility whose primary research goals are to
- Provide solutions to physics and operational issues critical to the success of the International Thermonuclear Experimental Reactor (ITER);
- Develop the physics basis for steady-state tokamak operation required for efficient power production;
- Contribute substantially to the technical basis for a fusion nuclear science facility; and
- Advance the fundamental understanding and predictive capability of fusion science.
The DIII-D project commenced in 1986, and its technical capabilities have continually evolved so that DIII-D is presently a flexible device that can study confinement, stability, and divertor physics with a variety of heating and current drive
techniques. This, in turn, allows for the development of the high-performance, advanced tokamak concept, which requires targeted simultaneous control of multiple plasma profiles both in the plasma core and at the edge. Near-term research on DIII-D addresses the development of plasma scenarios scalable to the high fusion gain ITER target. Longer-term research focuses on developing techniques
to produce stable, high-performance, steady-state (i.e., noninductive) operation for ITER and beyond.
DIII-D has major and minor radii of 1.67 and 0.67 m, respectively, with a nominal aspect ratio of 2.5. It has a maximum operating capacity of 2.2 T toroidal magnetic field and 3 MA plasma current, although it generally operates at lower currents, ≤2 MA. Eighteen field-shaping coils operated by a plasma control system provide great flexibility in plasma shape, discharge evolution, and divertor configuration. Divertor cryopumps control the plasma density. DIII-D presently has 26 MW of external heating capability, split between 20 MW of neutral beam (NB) heating and 6 MW of electron cyclotron (EC) heating and current drive. The neutral beams are configured on- and off-axis, and in the co- and counter-current direction, to provide a range of torque and neutral beam driven noninductive current profiles. Another key feature of DIII-D is the set of internal and external coils that can provide a wide spectral range of applied three-dimensional (3D) magnetic perturbations for edge-localized mode (ELM) suppression and other edge profile control studies. Shattered pellet injection and argon pellet systems are employed for disruption and runaway electron mitigation, a lithium and boron “dropper” is used for wall-conditioning, and a laser blow-off instrument is available for impurity transport studies. DIII-D has an outstanding, comprehensive set of core, edge, and divertor diagnostics. A close relation between theory and experiment enables the data to be readily used to validate first-principles physics simulations for the development of high-confidence predictive tools. A few examples from the diverse, multifaceted research program are given below.
In ITER, unmitigated ELMs will rapidly erode first-wall materials. DIII-D was the first tokamak to use 3D magnetic perturbations to suppress ELMs. Design of the ITER 3D coils relied heavily on DIII-D results. In support of developing operational scenarios for ITER, many recent experiments explore stability and confinement in plasmas with low injected torque, and with significant electron heating, as is expected in ITER. In particular, validation of transport models against ITER baseline discharges in DIII-D have revealed the importance of certain drift-wave modes that cause particle transport into the center of the plasma. This model,3 when used to simulate ITER scenarios, predicts a peaking of the density profile, which is a necessary condition for achieving the Q = 10 ITER target.4 Electron cyclotron heating was shown to be effective in expelling impurities from the core of these plasmas.5
Demonstration of noninductive scenarios with high confinement is another major program element. Experiments in the “hybrid” operational regime have achieved a normalized pressure of βN = 3.7 with a confinement enhancement factor of 1.6 and zero loop voltage, indicating that a significant portion of the plasma current was self-driven.6
Divertor and plasma-material solutions are key for a successful fusion reactor. A recent upgrade to the upper divertor module allowed DIII-D to study the physics of the Small-Angle Slot (SAS) configuration. This configuration can cause the divertor to radiatively “detach” from the hot upstream plasma at lower densities, giving lower temperatures across the divertor region, and presents a potential divertor solution to mitigate high heat fluxes. In materials studies, tungsten inserts installed during a metal-rings campaign provided measurements of tungsten erosion, migration, and redeposition.
The DIII-D program emphasizes scientific understanding to develop a predictive capability that improves fusion performance. A model was developed that explains the observed height and width of the pressure “pedestal” at the plasma edge when the tokamak operates in the “H-mode” confinement regime.7,8 Further analysis suggested that, by judicious choice of the plasma shape and discharge evolution, access to a higher pressure “super H-mode” was possible. Subsequent experiments accessed this higher performance regime, and produced plasmas with equivalent QDT of up to 0.6.
Modifications of DIII-D are currently under way. A major goal is demonstration of a steady-state condition with high confinement and pressure. To drive more off-axis current, one neutral beamline is being reoriented to inject off-axis. New methods to use plasma waves to drive current off-axis are also being prepared, including installation of a high-power helicon antenna.
Organizationally, DIII-D is managed by a private company, General Atomics (GA). Multiple national laboratory and university personnel, as well as GA employees, constitute the scientific staff. Generally, GA employees operate the major systems,
with several major subsystems the responsibility of national laboratory teams; diagnostic systems are the responsibility of university, national laboratory, and GA personnel. Experiments are selected after a “Research Opportunities Forum” that is open to all; review by a “Research Council” with experienced team members from GA, laboratories, and universities; and final allocations by GA management. Experiments are conducted by multi-institutional teams that often include international visitors. The research program is influential. As a measure of impact, consider the papers selected for oral presentations at the most recent International Atomic Energy Agency (IAEA) Fusion Energy Conference (the highest visibility conference in the field). Of 42 experimental magnetic fusion papers, 15 utilized DIII-D data, the most of any facility in the world.
The National Spherical Torus Experiment Upgrade (NSTX-U; Figure G.4)9 is one of 17 tokamaks designed to operate in the low aspect ratio regime. It is a high-powered, medium-size device that is one of the two largest and most capable low aspect ratio tokamaks in the world, the other being the Mega Ampere Spherical Tokamak Upgrade (MAST-U)10 in the UK. The mission of NSTX-U is to
- Advance the spherical tokamak (ST)11 as a candidate for a fusion nuclear science facility (FNSF);
- Develop solutions for the plasma-material interface, including the snowflake divertor and lithium/liquid metal plasma facing components (PFCs);
- Advance toroidal confinement physics’ predictive capability for ITER and beyond; and
- Develop the ST for fusion energy production—for example, as an ST pilot plant.
The ST concept, in which R/a ≤ 2, offers a potential development path for a more compact and lower-cost energy production system or materials testing facility through optimization of the fusion triple product nTτ, where n is density, T is temperature, and τE is energy confinement time. In particular, improvements in energy confinement are inherent in the ST due to the stabilizing properties of its high toroidicity, high plasma flow velocities, and high flow shear. STs also naturally achieve high-β due to operation at lower toroidal magnetic field, and the spherical nature of the plasma configuration leads to high natural elongation. Also because of the low toroidal field, the fast ion population that results from neutral beam injection in STs resides in a parameter space expected for α-heated plasmas at both conventional and low aspect ratio. These unique physics regimes, along with the compact nature of the ST, which leads to stringent requirements for developing
power handling and noninductive current drive capabilities, offers great leverage in testing tokamak physics models for improved predictive capability.
Many of the ST physics challenges were explored in NSTX,12,13 the predecessor device to NSTX-U. NSTX had an aspect ratio of R/a = 0.85/0.68~1.25; operated with plasma currents and toroidal magnetic fields of up to 1.5 MA and 0.55 T, respectively; had pulse lengths of up to 1.5 s; and operated in either D+ of He++. NSTX was equipped with a three-source neutral beam capable of injecting 6 MW of D0 power at 90 keV, and up to 6 MW of high harmonic fast wave (HHFW) RF power for heating and current drive. Co-axial helicity injection (CHI) was used for noninductive plasma start-up. Close-fitting passive conductors, coupled with application of active control algorithms using applied 3D magnetic fields as actuators, were used to stabilize magnetohydrodynamics (MHD) instabilities and maintain high-performance operations.
NSTX made significant progress in achieving a goal of high-β, long-pulse performance, including achieving βT values up to 35 percent, with βN up to 6.5 m-T-MA−1 and βN/li, a metric for maximizing bootstrap current, to 14. Its accomplishments include the following:14
- Pulse lengths up to 1 s, with 60 to 65 percent of the current being driven noninductively by both bootstrap and neutral beam current drive H-mode operation with τE/τ98y,2 values up to 1.5 and τE/τ89p values over 2.
- Confinement trends showing that performance improved with decreasing collisionality in a nearly linearly inverse fashion.
- Impurity transport rates near predicted neoclassical values in turbulent L-mode plasmas.
- Identification and development of approaches to control neoclassical tearing modes and resistive wall modes.
- Observation of different classes of fast ion-induced MHD, with modes in the conventional Alfvén eigenmode (AE) range of frequencies (tens of kHz), but also with frequencies of 0.5-1 MHz, near the ion cyclotron frequency.
- Significant electron heating (Te0 > 6 keV) and indications of current drive with HHFW.
- Noninductive start-up currents of up to 400 kA using CHI.
- Utilization of advanced divertor configurations (e.g., snowflake) with partial detachment to mitigate divertor heat flux.
- Use of lithium wall coatings to improve plasma performance and mitigate ELMs.
Coupled with these experimental achievements was the development of the theoretical underpinnings necessary for understanding the results. For instance, first-principles gyrokinetic simulations identified the microtearing mode, which is electromagnetic in nature and exists at high-β, as the microinstability responsible for most of the energy loss from the plasma, which was through the electron channel.15 This mode becomes more stable as collisionality is reduced, consistent with the strong increase of global confinement time with decreasing collisionality. Theory development related to the fast ion-driven AE modes led to a deeper understanding of how these instabilities affect both the fast ion and thermal populations.16 This understanding led to the development of models of fast ion transport that have been applied successfully at low and conventional aspect ratio. Furthermore, development of the theory of kinetic stabilization of resistive wall modes in NSTX17 was found to also explain stability trends at conventional aspect ratio.
NSTX-U will continue to explore physics issues critical to both low aspect ratio, but with enhanced capabilities. The toroidal magnetic field will be increased from 0.55 to 1 T, the plasma current from 1.5 to 2 MA, and the pulse length from 1 to 5 sec. A second, more tangentially injecting neutral beam was added, doubling the total available power up to 12 MW under normal operating conditions. These additions make NSTX-U the most powerful ST in the world, with the highest toroidal field and highest accessible pressure and β. This will allow NSTX-U to achieve up to 10 times higher fusion triple produce (nTτ) and 4 times higher divertor heat fluxes, reaching levels expected in ITER.
The increased current, field, and power will enable NSTX-U to operate at higher temperature and up to 5 times lower collisionality than in NSTX. Opera-
tion at reduced collisionality is critical to resolving how confinement varies with this parameter. If the favorable confinement trend with collisionality continues at these lower values, in contrast to the weaker dependence of confinement on collisionality at conventional aspect ratio, this would certainly be critical information for optimizing the ultimate design of a tokamak reactor, and would present the low aspect ratio, high-β regime as a potentially attractive one for a compact, more attractive reactor.
NSTX-U is an excellent testbed for simulating α-particle physics applicable to burning plasmas and ITER. Neutral beam-heated NSTX-U plasmas will operate in the largest fast ion dynamic range of parameter space of any ST or conventional aspect ratio tokamaks, and in the regime expected for α-heated plasmas at both low and higher aspect ratio. Experiments on NSTX-U have already shown the flexibility of the more tangential neutral beam in being able to phase space engineer the fast ion distribution in pitch angle and deposition profile in order to control the fast-ion instabilities.18
NSTX-U will be the leader in assessing whether high-performance STs can be sustained without a transformer, a critical research component since the compact nature of an ST-based pilot plant, for instance, will preclude a substantial OH transformer. The flexibility of the more tangential neutral beam will allow for additional noninductive current with profiles that can be controlled actively. Beam torque will induce rotation, and both active and passive stabilization of global MHD modes through the passive conducting plate and applied 3D magnetic fields, along with production of favorable current profiles, will allow sustainment of high performance. Additional noninductive current will be produced by the plasma through the bootstrap effect, which can be optimized through the high βN/li that will be attained, and that could be twice as high as that produced on NSTX. Predictive simulations indicate that 100 percent noninductive operation at 1 MA is possible.
While there is significant overlap between the two major ST devices, NSTX-U will focus on core physics, and in a complementary fashion MAST-U will focus on boundary physics. MAST-U is equipped with a significant number of poloidal field coils that will allow for much more flexible, long-legged divertor configurations than those that can be produced in NSTX-U. However, NSTX-U can contribute and, in some instances, lead in power exhaust studies. NSTX-U will be using solid lithium coatings to protect PFCs from high heat fluxes, to improve confinement and to suppress ELMs, as was done in NSTX. Solid lithium injectors on both the top and bottom of the vessel will serve to double the lithium deposition over that in NSTX. Long-term plans include the development of liquid metal divertors as a possible transformative wall solution.
NSTX-U operated for 10 weeks in 2016 and had a productive scientific campaign. However, by the end of that period, it was discovered that one of the poloidal field coils failed, necessitating NSTX-U to shut down for an extended recovery
outage. The NSTX-U recovery is ongoing, with numerous design improvements, including modification of the vacuum chamber, in order to support flexible operations and increase reliability to achieve key mission goals. New requirements for the divertor heat fluxes have been defined, based on recent scrape off layer heat flux width models. New halo currents loads have been determined based on data from NSTX, NSTX-U, MAST, and conventional aspect ratio devices. New error field analysis has been conducted, with the goal of both optimizing the global MHD stability and minimizing PFC heat flux asymmetries for scenarios with large poloidal flux expansion. New designs of graphite PFCs utilize castellations to reduce the mechanical stresses, allowing tiles to reach surface temperature limits, ~1600°C. Improved divertor coil designs simplify fabrication and facilitate turn-to-turn testing. The NSTX-U recovery project is on track to enhance reliability and safety and provide the highest performance ST device as a fusion research user facility. NSTX-U is expected to resume operations during CY2020.
SMALLER CONFINEMENT RESEARCH FACILITIES WITHIN THE UNITED STATES
Pegasus is an ultra-low aspect ratio tokamak at the University of Wisconsin that operates with R ~ 0.35 m, R/a ~ 1.13-1.3 BT = 0.17 T and elongation ~2. Its mission is to explore very high-β confinement and stability, and to develop noninductive discharge start-up techniques. Pegasus has achieved βT values near 100 percent, and it has also achieved H-mode plasmas, with threshold powers for accessing the H-mode well above (~15x) that predicted for the Pegasus operating parameters.19,20 Localized DC helicity injection utilizing plasma guns has produced induction-free plasmas with plasma currents up to 100 kA, with the plasma current scaling with injected edge current in accordance with the Taylor relaxation mechanism.21
The Lithium Tokamak Experiment Upgrade (LTX-β) is also a low aspect ratio tokamak, situated at PPPL, with R = 40 cm, R/a ~ 1.55, BT ≤ 0.17 T, and Ip ≤ 100 kA. It is the follow-on device to LTX. The purpose of LTX-β is to develop the approach to using liquid lithium walls, and to study their effect on plasma performance. LTX used lithium coatings on a high-Z wall, and it exhibited flat electron temperature profiles and enhanced confinement without having the lithium dilute the core plasma or radiate power.22 LTX-β will extend the capabilities of LTX with 700 kW of neutral beam heating and fueling, 100 kW of electron cyclotron heating/electron Bernstein waves (ECH/EBW) for electron heating, higher BT and Ip, longer pulse length, and upgraded diagnostics.
The Madison Symmetric Torus (MST)23 at the University of Wisconsin is a reversed-field pinch (RFP) physics experiment, which relies on a transient burst of current to create the plasma and the confining magnetic fields. In the RFP, the toroidal magnetic field is weaker than the poloidal magnetic field, and it actually reverses
direction in the plasma near the edge. The mission of MST, presently a formal user facility, is to study fusion and astrophysical implications of reconnection,24 turbulence,25 and dynamo formation. A 1 MW neutral beam injector will be used to heat the plasma and enable studies of fast particles and their role in the reconnection process. A wide range of diagnostics is available for characterizing the plasma.
The Helically Symmetric Experiment (HSX),26 also at the University of Wisconsin, is a quasi-helically symmetric (QHS) stellarator with R = 1.2 m, a = 0.15 m, and BT up to 1.25 T. It has up to 200 kW of EC heating, which can heat the electrons up to 2 to 2.5 keV in the core. By nature of its QHS design, neoclassical electron thermal transport was reduced.27 Furthermore, HSX exhibited reduced damping of plasma flow,28 important for ultimately reducing turbulence-driven transport, reduced bootstrap and Pfirsch-Schlüter currents for maintaining plasma stability,29 and good particle confinement of trapped high-energy electrons.30 HSX also serves as a flexible divertor test platform, able to produce either an island or a nonresonant divertor.
The Hybrid Illinois Device for Research and Applications (HIDRA) at the University of Illinois is a classical stellarator with R = 0.72 m and a = 0.19 m, with magnetic fields up to 0.5 T. The main focus of HIDRA is to study plasma-material interactions, including liquid lithium science and technology.31
The Compact Toroidal Hybrid (CTH)32 device at Auburn University is designed to study how MHD stability in a stellarator depends on 3D shaping of the plasma. It has R = 0.75 m, a = 0.29 m, and BT = 0.7 T, and it has independently controlled magnet coils that can produce magnetic configurations over a large range of vacuum transforms, as well as having additional coils to control plasma shape as well as horizontal and vertical position. There is an ohmic system that produces plasma current, and when operated in this mode, disruptions due to vertical displacement events, density limits, and low-q have been observed.33
The mission of the High Beta Tokamak-Extended Pulse (HBT-EP)34 device at Columbia University is to utilize an adjustable close-fitting conducting wall for passive stabilization,35 and applied external magnetic perturbations36 for active control of MHD modes to study and extend the β-limit. It has R = 0.92 m, a = 0.15 m, and BT = 0.35 T.
The Helimak37 is an R = 1 m, BT = 0.1 T toroidal device that is used to study plasma turbulence at high collisionality.38 Because its magnetic field lines have low pitch, its geometry approximates that of an infinite cylinder. Flow shear is externally applied and can be controlled. The plasma is colder, with Te ~ 10 eV and densities of only 1017 m3.
INTERNATIONAL FUSION RESEARCH FACILITIES
The current U.S. fusion research strategy has an increasing focus on U.S. participation in newer international long-pulse experiments with superconduct-
ing magnets including EAST (China; Figure G.5),39 KSTAR (Republic of Korea; Figure G.6),40 and Wendelstein 7-X (Germany).41 EAST began operation in 2006 and KSTAR began in 2009. The Wendelstein 7-X stellarator began operation in December 2015, requiring €350 million for the stellarator device42 and additional amounts for personnel and materials during construction. The HL-2M tokamak is under construction at the Southwestern Institute of Physics43 as an upgrade to the existing HL-2A44 device. HL-2M will have higher plasma heating power and magnetic field strength to explore higher-pressure, fusion-relevant plasma. The JT-60SA tokamak in Japan (Figure G.7) is under construction as a Japan-Europe project and is expected to begin operation in 2020.45 Non-U.S. proposals for new facilities include the superconducting DTT facility46 that would be built by the Italian National Agency for New Technologies, Energy, and Sustainable Economic Development’s fusion laboratory in Frascati, Italy, and the China Fusion Engineering Test Reactor (CFETR) under consideration as a new fusion facility to demonstrate self-sufficient tritium breeding. While researchers in the U.S. fusion community welcome these international opportunities, presentations to the committee47 and during the first fusion community workshop48 did not foresee how international cooperation by itself will allow U.S. fusion researchers to maintain a world leadership position without new facility starts within the United States.
The United States has made and continues to make important contributions to the world’s largest currently operating fusion device, Joint European Torus (JET).
This includes involvement in testing important auxiliary systems relevant to ITER (e.g., the ITER-like Shattered Pellet Injector49), plasma diagnostics (e.g., Faraday cups), and experimental operating scenarios (e.g., involvement in developing deuterium-tritium scenarios50). Additionally, simulation codes (e.g., TRANSP51) developed by U.S. scientists have been adopted by international partners and are now routinely used for scenario modeling within the JET program and across
EUROfusion ITER-related activities. Since 2016, 9 of the 33 articles appearing in the IAEA journal Nuclear Fusion and reporting results from the JET device involved co-authors from the United States.
For intermediate-size tokamaks (ASDEX Upgrade, Germany; TCV, Switzerland; MAST Upgrade, United Kingdom), many bilateral collaborations exist between the United States and EU partners. Prominent recent examples of U.S. contributions include temporarily moving diagnostic devices from U.S. facilities to EU machines and joint experiments on multiple machines to develop understanding and robust demonstration of control schemes and new plasma scenarios. Since 2016, about 10 percent of the articles appearing in Nuclear Fusion describing research with these medium-size tokamaks involved co-authors from the United States.
Another important U.S. contribution to fusion research in the EU has been the participation in the Wendelstein 7-X (W7-X) stellarator project. This includes the construction and operation of five large auxiliary coils52 (installed on the outside of the device to assist in precise setting of the magnetic fields at the plasma edge) and an X-ray spectrometer, as well as the development of fluctuation diagnostics and a pellet injector. This work is carried out at three U.S. national laboratories (Princeton Plasma Physics Laboratory, PPPL; Oak Ridge National Laboratory, ORNL; and Los Alamos National Laboratory, LANL) and three U.S. universities (Auburn University; University Wisconsin, Madison; and Massachusetts Institute of Technology, MIT), supporting W7-X with equipment that has been funded, designed, and produced in the United States and with related magnetic field and plasma diagnosis and modeling. Since 2016, more than half of the articles appearing in Nuclear Fusion describing research with the W7-X stellarator involved coauthors from the United States.
The United States is actively playing a significant role in developing new fusion programs in Asia. Major contributions have been made to the programs on new Asian devices since the 2004 National Research Council Report (NRC) report—notably, EAST (China), KSTAR (Republic of Korea), HL-2A (China), and J-TEXT (Japan)—and a strong relationship continues with smaller spherical tokamaks—QUEST (Kyushu University, Japan), VEST (Seoul National University, Republic of Korea), SUNIST (Tsinghua University, China). In particular, noninductive plasma start-up and ramp-up using CHI and electron cyclotron wave heating and current drive is the focus of a multidomestic institution collaboration with QUEST. A major focus of this international partnership has been in the use of long-pulse superconducting devices to develop steady-state plasma scenarios.53 As an example, collaborations on EAST have made advances in plasma control and wall conditioning techniques developed collaboratively with and initially demonstrated on DIII-D. Novel computer science hardware and software infrastructure has improved data movement, visualization, and communication and allows scientists in the United States to remotely conduct experiments using the EAST facility.54 In July 2017,
the Chinese researchers using EAST achieved a stable 101.2-second steady-state high-confinement plasma, setting a world record in long-pulse H-mode operation.55 Similarly, physicists at Princeton Plasma Physics Laboratory have connected remotely to run experiments on KSTAR.
Recent U.S.-Asia cooperation is also seen in the development of HL-2M under construction in China and in the physics design of the CFETR burning plasma facility under consideration in China, where the United States provides design expertise and simulation codes.56
INTERNATIONAL COLLABORATION IN U.S. RESEARCH
International collaboration with U.S. researchers in burning plasma science involves all parts of the program, including use of experimental facilities and involvement with theory, simulation, and modeling groups. As a metric of international involvement since 2016, of those articles appearing in the IAEA journal Nuclear Fusion describing research with U.S. medium-size tokamaks, one-fourth involved co-authors from Europe and one-fourth involved co-authors from Asia. Half of all articles appearing in Nuclear Fusion since 2016 reporting advancements in fusion simulation involved collaborating international co-authors. In the area of fusion technology and engineering science, the EUROfusion Work Package for Plasma Facing Components pays to use the PISCES-B facility at the University of California, San Diego, helping to identify first-wall materials for ITER and future fusion energy systems. Currently, no other linear plasma facility is capable of performing experiments with beryllium samples. One main goal of this collaboration is to study the interaction between deuterium or helium plasmas with beryllium and tungsten surfaces. Another example of a long-standing U.S.-Japan collaboration is the study of high dose irradiation effects in an experiment on the High Flux Isotope Reactor at Oak Ridge National Laboratory.
International participation from Asia (China and Korea in particular) in the U.S. program also has the goal of importing established U.S. scientific knowledge such as 3D physics, tokamak scenario development, diagnostic techniques, heating and current drive technology (ECH, Klystron for helicon CD, LHCD high field launch), and advanced plasma control systems including real-time control and tokamak design and construction (e.g., HL-2M design and construction based on knowledge gained from DIII-D). Joint experiments such as those performed on EAST and DIII-D, simulation and modeling codes such as BOUT++, and technology transfer on linear plasma sources for plasma-material interaction (PMI) study such as PISCES are also areas where the current focus of the collaborating Asian scientists is to absorb leading scientific expertise of the United States.
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