Monday, July 25
Charles Greenfield (General Atomics): Welcome
Dr. Greenfield will welcome participants to San Diego and IIS2022. The talk will introduce the school, discuss some of the logistics, and introduce the DIII-D National Tokamak Facility that some of you will visit on Monday evening's tour.
Dmitri Orlov (UCSD): Introduction to UCSD
The lecture will introduce the IIS-2022 participants to the University of California San Diego and will include an overview of the nuclear fusion research activities at the UC San Diego, including the research done at the UCSD laboratories, the DIII-D national facility, US and international facilities.
Chris Holcomb (Lawrence Livermore National Laboratory): DIII-D Scenario Development and Control Research
To give a flavor of the types of scenarios and controls topics that will be discussed throughout this summer school, this opening lecture will give a high-level overview of plasma operating scenarios and controls research being conducted on the DIII-D tokamak. DIII-D is a mid-sized (R=1.67 m, a=0.67 m, B=2.1 T) experimental tokamak facility with relatively high levels of heating and current drive, flexible plasma shaping, and extensive diagnostics. A mature and adaptable plasma control system (PCS) is used for everything from routine vertical stability control to sophisticated acquisition of target plasma profiles and active instability avoidance. The talk will discuss a range of scenarios including (1) inductive ITER Baseline, which satisfies ITER’s fusion goals using relatively conservative normalized parameters, and (2) fully non-inductive High-qmin, which pushes normalized stability and confinement to the upper limits to achieve attractive conditions for steady-state DEMOs or fusion pilot plants. The actuators, diagnostics, and key related controls research required to obtain high performance, avoid disruptions, and protect the device also will be covered.
Work supported by US DOE under DE-FC02-04ER54698.
Peter de Vries (ITER Organization): Status of the ITER Project and its Scenario Development and Control
In southern France, the EU, China, India, Japan, Korea, the Russian Federation and the United States, are collaborating to build the world's largest tokamak, ITER, a magnetic fusion device that has been designed to prove the feasibility of fusion as a large-scale and carbon-free source of energy. It will be the first magnetic fusion device to produce net energy and thus to operate with burning plasmas, testing the physics regimes necessary for the commercial production of fusion-based electricity. This presentation will first show recent progress at the ITER construction site and in assembly of the device. It will then provide an outline the ITER Research Plan (IRP), showing how the tokamak will be brought into operation and how the operation range will be systematically expanded in stages, towards the ITER project goals. Finally, the design and development of the ITER Plasma Control System (PCS) will be discussed, focussing in particular on requirements and on how the various plasma operational scenarios forming the basis of the IRP are integrated into the PCS design.
Hartmut Zohm (Max-Planck-Institut für Plasmaphysik, DEMO Central Team, EUROfusion PMU): Introduction to Tokamak Operation Scenarios and Development Considerations
The term tokamak operation scenario is used to describe a plasma state characterized by the radial profiles of plasma pressure and toroidal current density. Depending on these distributions, the resulting operational point will have different properties such as confinement quality, MHD stability or achievable discharge length. Hence, the operation scenario can be optimized in different directions, e.g. to maximise fusion power or pulse length for a given tokamak hardware. Since the profiles are interlinked by the plasma physics of transport and stability, the optimization strategies are non-trivial and benefit greatly from detailed knowledge of the underlying physics.
In the talk, I will start with a discussion of the underlying physics that determines the current and pressure profiles. I will then review different tokamak scenarios and point out their benefits and drawbacks for application in future fusion reactors. I will also discuss recent progress in integrated modelling of tokamak scenarios, including the development of a tool that models the full tokamak discharge including the feedback control system (so-called ‘tokamak flight simulator’).
Michael Walker (General Atomics): The Science of Control
The primary objectives of control are somewhat different from those of much of fusion plasma physics. Magnetic fusion physics has historically focused on understanding the physics of plasmas in magnetic confinement devices, while fusion plasma control seeks to capitalize on the understanding already gained to cause the system (fusion device plus plasma) to behave in certain desirable ways. For example, early uses of plasma control in fusion devices had simple goals such as extending the survival of discharges by minimizing plasma-wall interaction or by regulating density. Present applications are primarily aimed at achieving conditions with better potential fusion performance or conditions under which fusion plasmas can be more easily studied. The demanding performance requirements and significant constraints expected on control of future fusion reactors suggest that plasma control is a critical enabling technology for progress toward commercial fusion power. A greater understanding of control techniques for fusion plasmas and a more widespread use of these techniques in existing devices is required in order to develop the solutions needed.
The science of control is old  - much older than either the science of plasma physics or its application to magnetic fusion. The maturity of control theory and thousands of proven applications in dozens of disciplines imply that a wealth of techniques already exist for developing control solutions for fusion plasmas. This talk will provide an introduction to the language, concepts, and tools of this broad discipline. Some of the concepts introduced have counterparts in approaches that will be familiar to many physicists, but use a somewhat different language to describe those concepts.
Our focus in this presentation is to provide attendees with a basic understanding of the tools used to analyze systems and to develop controls for those systems, as well as some intuition for how these tools work and under what conditions to apply them. We first describe several ways in which systems to be controlled and the algorithms used to control them may be represented. We then describe the basic concepts behind feedback and feedforward control and introduce the metrics by which controlled systems are evaluated. Finally, we describe several frequently used control techniques and discuss their advantages, disadvantages, and areas of intended application.
This material is based upon work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences, using the DIII-D National Fusion Facility, a DOE Office of Science user facility, under Award(s) DE-FC02-04ER54698. Disclaimer: This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.
 Stuart Bennett, A Brief History of Automatic Control, IEEE Control Systems Magazine, June 1996, 17-25
Tuesday, July 26
Wolfgang Treutterer (Max Planck Institute for Plasma Physics): Control Simulation
Simulation of complex systems nowadays is an indispensible tool to estimate their behaviour under nominal as well as extraordinary conditions. The lecture will deal with the question, on how simulation can support the design of control strategies and secure the operation of feedback controlled systems on the example of the ITER plasma control. Looking at typical use cases, two characteristic types of simulations will be elaborated: high-fidelity simulation for accurate prediction but limited to details of the full system, and for control simulation for a quick, holistic preview, albeit with reduced accuracy.
Finally a short overview on the ITER Plasma Control System Simulation Platform PCSSP will be presented, illustrating typical features and tasks of a control simulator.
Gianmaria De Tommasi (University of Naples): Magnetic Equilibrium and Instability Control
The objective of the magnetic equilibrium and instability control system is to track the desired plasma current, shape and position, by reacting to the foreseen disturbances, and by compensating for uncertainties on the reference model used to design the scenario waveforms.
In this lecture the main advantages of feedback control will be first recalled. Then the following control problems, which are tackled by the magnetic equilibrium and instability control system, are introduced:
Plasma Current Control
Plasma Shape Control
Vertical Stabilization of elongated plasmas
Finally, an overview of a modular and flexible architecture for magnetic equilibrium and instability control is presented, together with a possible implementation of the various control algorithms.
The attendees are supposed to be acquainted with the following basic tools for the analysis and design of control systems: state space representation of linear systems, transfer functions, frequency response, proportional-integral-derivative control. To this aim, the interested students are referred to:
K. J. Åström and R. M. Murray, “Feedback Systems – An Introduction for Scientists and Engineers”, available on line
David Eldon (General Atomics): First Wall Heat Load Control, ELM and Divertor, Detachment Control
Control systems are implemented to mitigate intense heat flux expected in future fusion devices. Without intervention, heat and particle fluxes reaching divertor target plates tend to concentrate in narrow (~cm in R) regions and thus the peak heat load will likely be well above the material’s tolerable limit. Adding extrinsic impurities to the plasma promotes line radiation and other dissipation processes that spread the plasma’s heat exhaust across a greater wall area. With strong enough dissipation, the zone of primary interaction between the plasma and neutrals from the surface can detach from the divertor target plate, shielding the plate from most of the direct heat load from the plasma. In wall-limited plasmas, impurity line radiation is useful for spreading heat loads across wider areas. While this is an excellent way to protect the wall and divertor from melting or sputtering, the extrinsic impurities are also a potent means of reducing core plasma confinement quality, diluting fusion fuel, or even prompting a disruption. It is the job of the control system to moderate the flow of impurity gas to achieve divertor/wall protection without harmful excess. It is not guaranteed that every plasma scenario is compatible with both detachment and good core performance at the same time. The detachment control system and core scenario must also be compatible with an Edge Localized Mode (ELM) removal solution, since the intermittent heat flux from an ELM in a reactor would potentially exceed the material’s tolerable limit. Further complicating the problem, ability to diagnose and affect plasma conditions in future devices will be limited as many popular diagnostics are unlikely to be feasible in a fusion power reactor, and key actuators will be subject to constraints as well, such as delays due to longer gas lines.
Control system design includes control policy/algorithm design, selection (in flexible devices like DIII-D) or at least awareness (in single-point designs) of the base scenario or operating point, selection of sensors and formulation of control parameters, and selection of actuators, in this case by choosing which gas species to inject into the plasma. Each of these facets will be reviewed, followed by a look at challenges and potential solutions for future devices compared to the current state of the art.
David Humphreys (General Atomics): Disruption Prevention and Avoidance
An attractive power plant candidate must provide power more than 80% of the time in a given operational year, typically implying that the frequency of key component failures resulting in unplanned loss of plant availability must be reduced to below 0.001/year . Present fusion devices typically have little motivation to operate with such high reliability, and allow relatively frequent instability-driven plasma-terminating events known as disruptions. The vision of an operational fusion reactor therefore includes a level of reliable control performance and confidence well beyond that of presently operating devices. Maximizing use of the limited number of discharges planned for ITER also implies a major advance in control reliability. Fortunately, the mature field of control theory offers methods that routinely provide such levels of performance in many fields from aerospace to process control.
This lecture discusses the control design methods and issues involved in providing high reliability control for tokamaks, including control to minimize the frequency of disruptions and minimize the damaging effects of faults in general. Providing this level of control requires high-reliability model-based design of algorithms, a systematic approach to verifying and validating performance, and a detailed mapping of operational control space to ensure this performance across a wide range of real-world conditions. Our consideration of reliable control will take us to the next steps in design beyond nominal stability assurance and satisfaction of performance requirements under ideal conditions. These next steps include designing for robustness to noise and disturbances, as well as for tolerance to variability in plasma, auxiliary system, and diagnostic responses. We introduce the concept of a Control Operating Space (COS) to describe the degree and robustness of closed loop control used in different physics regimes in terms of control performance metrics, and consider the design options and control consequences of various choices. We can apply the COS to designing a layered control hierarchy, in which different control layers provide simple regulation and tracking of system and plasma states, or stabilization of particular instabilities, while another layer explicitly regulates the degree of controllability or stability of the plasma to maintain a specified level of distance from dangerous boundaries. Both the layered architecture approach to control authority and this “boundary proximity control” can be powerful tools for reducing disruptivity to commercially-effective levels.
Although the goal of high reliability control is to reduce the frequency of control failure and incidence of catastrophic faults to an acceptably low level, a properly designed control system must also include a comprehensive and provable response to faults, should they occur. The fault response design must be comprehensive in order to sufficiently account for likely fault modes and assess their probabilities, and provable in order to reliably calculate the probability of various response scenarios being effective. We discuss methods of detecting and responding to faults so as to recover and continue operation if possible, or execute algorithms and scenarios for soft rapid shutdown, or in extreme (and extremely unlikely) events, to execute a hard rapid shutdown along with application of damage mitigation tools. We draw analogies between fault responses in high performance aircraft and tokamaks, and argue that the same methods that yield less than 1 fatality per billion airline passenger miles  can also produce a highly reliable tokamak power reactor.
Work supported by the Department of Energy under DE-FC02-04ER54698
 N.P. Taylor, “A Model of Availability of a Fusion Power Plant,” ISFNT-5, Rome, 1999
 A.T. Wells, C.C. Rodriguez, Commercial Aviation Safety, McGraw-Hill (2004)
Wednesday, July 27
Egemen Kolemen (Princeton University): Model Based Control
Physics-based models of complicated phenomena occurring in a fusion reactor many times not available. However, we have many years of experimental data available from various fusion devices. I’ll talk about the progress we made on control of plasmas using experimental data, in combination with simulations.
Eugenio Schuster (Lehigh University): Core Kinetic and Magnetic Control in Tokamak Reactors
As the first burning-plasma experiment, the need for regulating the burn condition is a unique characteristic of ITER. Regulating the amount of fusion power produced by ITER and future fusion reactors will require precise control over the plasma density and temperature. Therefore, the control of the core-plasma kinetic state, usually referred to as burn control, arises as one of the most fundamental problems in fusion reactors, and will be critical to the success of burning-plasma devices like ITER. Due to the nonlinear coupled dynamics of the plasma, feedback control of the burn condition will be necessary to avoid undesirable transient performance and to respond to changes in plasma confinement, impurity content, or operation conditions, which could significantly alter the plasma burn. The goals of the active burn controller are threefold: i- access to and exit from the burning plasma mode; ii- regulation around a desired burning equilibrium point by rejecting perturbations in temperature and density even when the equilibrium point is unstable; iii- drive the plasma system from one operating point to another during burning plasma operation (e.g., different Q or fusion power). Moreover, for nuclear fusion to become an economical energy source, tokamak reactors must be capable of not only operating in a burning plasma mode characterized by a large value of Q, the ratio of fusion power to auxiliary power, but also doing this stably for extended periods of time. This will demand simultaneous shaping of the plasma internal profiles, which has been demonstrated to be a key condition for the realization and sustainment of advanced modes of operation characterized by MHD stability, improved confinement, and steady-state performance.
This lecture will provide an overview of the state of the art of both burn control and profile control in the plasma core, their mutual coupling, and their relationship with the plasma scenario, magnetic equilibrium, MHD stability, and divertor heat/particle flux load. It is important to realize that there is a strong coupling between the plasma dynamics in the core (inside the last closed magnetic surface) and the plasma dynamics in the scrape-off layer (SOL). Therefore, any practical solution to the problems of burn control and wall/divertor heat/particle-load control demands an integrated approach. This implies that plasma-performance goals in the core may need to be relaxed, either by re-shaping the plasma profiles or eventually changing the burn condition, to be able to guarantee secure divertor operation. Different modeling and model-based control approaches will be discussed. Traditional actuators for burn control will be considered, including auxiliary power regulation to prevent quenching of the fusion burn, impurity injection to increase radiation losses and stop thermal excursions, and fueling (gas puffing + pellet injection) to regulate the density. Moreover, both isotopic fuel tailoring and confinement modification via in-vessel coils will also be considered as actuators. All these actuation mechanisms will impact the dynamics of not only the density and temperature average magnitudes but also their spatial profiles. In addition to heating and particle sources for density, temperature, and pressure profile control in a burning plasma, non-inductive current drives and torque sources will be considered for current profile control and rotation profile control, respectively. Finally, mechanisms to tackle both actuator and diagnostic constraints in the control-design process will be discussed.
George Sips (General Atomics): Actuators and Sensors for Tokamak Control
With the construction of ITER progressing towards completion, the worlds tokamak experiments provide a supporting and preparation role for ITER while continuing research on exploring more advanced operation scenarios compatible with a future steady state power reactor and to avoid plasma instabilities completely. To achieve its scientific objectives, the ITER device will need to implement solutions to several challenging control problems, including avoidance of operating limits before more draconian hardwired protection systems activate. Several strategies are used for using sensor data for control with actuators, from simple single-input-single output (SISO) systems to multiple-input-multiple output (MIMO) systems with actuator (priority) sharing. Processing of sensor data can be performed by the real-time (RT) data system or alternatively be outsourced on computing nodes that provide reduced inputs to the RT control system. Sensors discussed are the reconstruction of magnetic data into equilibrium information in real time (EFIT), IR data for wall protection and divertor performance, real time kinetic (profile) data, bolometry and plasma impurities, instrumentation of key components and MHD instabilities. Actuators vary from external coils for shape and divertor control, internal coils for ELM and disruption control, heating and current drive systems, gas injection system and pellet injectors.
The status and plans for sensors and actuators for control is given, with examples from DIII-D and JET, while projecting the requirements for sensors and actuators in the harsh ITER environment.
Federico Felici (Swiss Plasma Center): Integration of plasma real-time control, state monitoring and event detection for reliable and routine operation of tokamaks
Routine operation of tokamaks in future reactors and power plants will require high-reliability control systems that address multiple tokamak control challenges simultaneously. In addition to basic magnetic and kinetic control, which are routinely performed on present-day tokamaks, advanced control functions such as MHD instability control, radiation control, detachment control, fusion burn control and profile control to be carried out, all simultaneously. Moreover, the tokamak state and various events will need to be monitored and categorized in real-time to assess the proximity to disruptions and determine and execute the appropriate response.
These requirements pose new challenges which are currently at the frontier of plasma control research. For example, interaction between different control tasks due to their effect on the plasma introduces the need for a trade-off between multivariable controller design for several quantities at once and separate design of independent control loops. Multiple concurrent controllers may request to use the same actuator (e.g. a heating or fueling system), requiring real-time prioritization of control tasks and automated allocation of actuators to a given controller. Various real-time diagnostics provide partial information about the plasma state and events, requiring integration of diagnostic information in real-time into a generic plasma state description to be used by the supervision and control algorithms.
Further challenges in this domain concern the implementation of advanced control functions in tokamak control system software and hardware. Novel architectural paradigms have emerged to enable handling complex simultaneous control tasks, including event handling, in a tractable manner by enforcing strict separation between various components and responsibilities. From the software engineering point of view, this increasing complexity in plasma control system algorithms calls for ever increasing integration of model-based design, verification and continuous validation of control software across the lifetime of a tokamak experiment.
This lecture will discuss the state-of-the-art of control integration using examples from various tokamaks world-wide and present the open challenges for future research.
Indranil Bandyopadhyay (ITER India): Scenario Development Constraints (operation limits, control constraints)
This lecture will cover various operational limits in tokamaks – density limits, beta limits, vertical stability limits etc and the constraints on various actuators in scenario development in some detail. The Hugill diagram, pushing the boundaries of the Hugill diagram and especially the Greenwald density limit will be discussed. The control constraints on various control actuators - both for magnetic and kinetic control, especially in the ITER scenario development, e.g., flux and force limits on the central solenoid and poloidal field coils will be presented. Some of the ITER scenario development and control simulation results using mainly 2D MHD and transport solver codes will be presented.
Thursday, July 29
Francesca Poli (Princeton Plasma Physics Laboratory): Integrated Operation Scenarios
In this lecture Integrated Operation Scenarios are described from the point of view of modeling for experimental planning. Integrated modeling for operation is the place where the physicist meets the engineer, where plasma discharges are designed – or better ‘engineered’ – to be safe and to satisfy at the same time target performance, hardware and system requirements and control constraints. The elements of a fusion plasma simulator will be described, with focus on the physics fidelity required for specific applications, with example from specific applications, for ITER plasmas and for present-day experiments.
Sun-Hee Kim (ITER Organization): ITER Pre-Fusion Power Operation phase scenarios
Pre-Fusion Power Operations (PFPO) in ITER, following the First Plasma milestone of achieving plasma breakdown in hydrogen or helium, will be important for commissioning the tokamak systems and for validating various techniques necessary for establishing operations in deuterium-tritium. The PFPO phase consists of two periods, PFPO-I and PFPO-II, within the Staged Approach of the ITER project, and enables physics and technology research activities associated with machine assembly and integrated commissioning activities in hydrogen and helium plasmas. In the PFPO-I phase, extensive system commissioning activities with plasma will be performed and diverted plasma operation will be developed up to 7.5MA/2.65T along with the establishment of plasma control, diagnostics, electron cyclotron (EC) heating and current drive (H&CD), and disruption mitigation capabilities. Early access to high confinement mode (H-mode) through operation at 1.8T with EC heating up to 30MW is an option that is being considered. In the PFPO-II phase, heating/diagnostic neutral beams and ion cyclotron H&CD will be commissioned to their full power, while advancing the capabilities of plasma control, edge localized mode (ELM) and divertor heat load controls, fuelling and disruption mitigation. As a key milestone, high power L-mode operation will be developed up to 15MA/5.3T to demonstrate the full technical capability of the device. Various H-mode scenarios at fields above beyond 1.8T by utilizing various mixes of H&CD systems (up to 83MW) will establish the physics and operational basis required for the transition to Fusion Power Operation phase. This lecture presents various plasma operation scenarios developed for the ITER PFPO phases and explains their operational goals, physics assumptions and constraints applicable.
Francesca Turco (Columbia University): ITER Baseline Scenario (Q=10 ELMy H-mode)
The ITER Baseline Scenario (IBS) is the main scenario foreseen for reaching the ITER Q=10 mission, which requires the production of 500 MW of fusion power in stationary conditions for ~400 s, with a power multiplication factor of Q=Pfus/Paux~10. This mission is arguably one of the crucial achievements that ITER needs to demonstrate, as early as possible in its DT phase. Therefore assessing the options to maximise the stability, performance and safety of this scenario is one of the central topics of research in fusion aimed at producing a viable source of energy. In this lecture we will go over the main characteristics of this scenario and its goals and physics+hardware constraints, as well as the available experiments designed in present tokamaks to demonstrate the feasibility and assess the issues for this kind of plasmas. We will also present projections of the IBS demonstration discharges in a range of different core and edge conditions, to connect the present operational space and achievable results to the ITER parameter space and requirements. The issue of MHD stability of these high current plasmas will be discussed, where a disruptive 2/1 tearing mode limits the duration and the performance of the scenario, as well as the passive and active methods to avoid it or suppress it, with the trade-offs for each route. Confinement and heating efficiency considerations based on experimental results will be presented, showing the trade-offs between core and edge heating, the impact of external torque, direct electron heating and collisionality level.
Joëlle Mailloux (UKAEA): Preparation and Execution of JET Deuterium-Tritium Campaign
In 2021, high fusion power deuterium-tritium experiments (DTE2) were performed in JET for the first time since the 1997 D-T campaigns in TFTR and JET (DTE1). The preparation for DTE2 took place over many years, starting with the conception of the JET ITER-like wall (ILW: W divertor and Be main chamber) and included several enhancements. Dedicated experimental campaigns to expand the operational space in JET-ILW and to prepare the plasmas required to exploit JET’s currently unique tritium handling capabilities were performed. DTE2 delivered record fusion energies and demonstrated the compatibility of sustained high performance D-T plasmas with the ILW, while also highlighting operational challenges from changing from deuterium to D-T plasmas, as ITER will need to do. Experiments designed to address specific physics questions of interest to ITER provided several notable results. Compared to their deuterium counterparts, D-T plasmas require lower input power to reach H-mode and show better particle confinement. The plasma pedestal pressure increases with the ratio of T to D, with improved measurements allowing better understanding of its contribution to global confinement. The total W source is higher for T ions than D, but the concentration in the high power D-T plasmas remains tolerable, at least in part due to strong core heating and screening. Efficient core heating and impurity control was observed with the Ion Cyclotron Radio Frequency schemes considered for ITER D-T operations: e.g. 3He minority heating, and a novel D-9Be-T three-ion scheme with the Be naturally present in the ILW. Unambiguous observations of alpha particles and of alpha-driven instabilities were obtained. Plasmas with neon seeding for reducing divertor heat loads, as required in ITER, were performed for the first time in D-T. DTE2 provides a unique dataset for testing predictive capabilities and better prepare ITER fusion power operation.
Friday, July 29
Yong-Su Na (Seoul National University): Alternative Tokamak Operation Scenarios
This lecture deals with alternative operation scenario development in tokamaks by dealing with their status and future prospect. Requirements of fusion reactors are addressed based on figure of merits for fusion performance and non-inductive current drive. The limitations of conventional H-mode scenarios are discussed with respect to these figure of merits. Several alternative operation scenarios to overcome them, so-called advanced scenarios are introduced. Firstly, the alternative scenarios developed so far are categorised according to the shape of the q-profile into strongly-reversed shear, weakly-reversed shear, and low or flat shear scenarios. Secondly, MHD stability and beta limits of each scenario are discussed with its pressure profile and current density profile. Thirdly, confinement characteristics of each scenario is discussed with respect to the existence of internal transport barrier (ITB), the edge pedestal property, etc. This lecture mainly covers various ITB scenarios including high βp scenario [1,2], high βN scenario , quiescent double barrier scenario , and FIRE mode , advanced inductive or hybrid scenario [6,7], high li mode , and some enhanced pedestal modes. Finally, prospect of tokamak operation scenario development is to be discussed.
 Y. Koide et al, Phys. Review Letters 72 3662 (1994)
 A. M. Garofalo et al, Nucl. Fusion 55 123025 (2015)
 Y. Kamada et al, Nucl. Fusion 34 1605 (1994)
 E. J. Doyle et al, Plasma Phys. Controlled Fusion 43 A95 (2001)
 H.S. Han et al, Nature, accepted (2022)
 T. C. Luce et al, Nucl. Fusion 54 013015 (2014)
 Yong-Su Na et al, Nucl. Fusion 60 086006 (2020)
 L. L. Lao et al, Phys. Review Letters 70 3435 (1993)
Sang-Hee Hahn (Korea Institute of Fusion Energy): Long-Pulse Tokamak Operations
Extension of plasma pulse length is essential for economically meaningful fusion reactor. The task is, however, not equivalent to the mere physical extrapolations of known short pulses; At least not for the tokamak-type devices, which creates a column of plasma current (Ip) by fast poloidal flux swing. The long-pulse design, a sustainment of plasma current (Ip) as long as possible under the given hardware constraints, would have two different aspects from that of general short-pulse discharges; The first aspect is that only a few scenarios with high fraction of non-inductive current drive (NICD) can sustain the pulse over hundreds or thousands of energy confinement time. Since the available poloidal flux consumption is limited by the ohmic coil capacity of the device, any inductive scenarios would have their own limitation on the pulse length. Toward the steady-state tokamak discharge, defined as one whose duration is determined by the daily rhythm of the electric supply demand rather than by any physics or engineering limits [Lister 2000], candidate scenarios with zero- or low loop voltage are ought to be developed.
The second aspect is that the long pulse discharges interacts with the devices, changing the known engineering constraints during the discharge; Changes include 1) vacuum vessel chamber conditions, 2) auxiliary heating devices, 3) diagnostics and even 4) control magnets. Especially, devices with adaptations of superconducting PF/TF magnets (including ITER) have machine-specific constraints related to the allowed coupling AC losses. The plasma-facing components (PFCs) and their interactions with the plasma are the main reason of the chamber condition changes; Changes on recycling, increase of carbon wall retention, and expected but unpredictable events such as erosion, melting or sputtering of particles due to the PFC temperature increase. Diagnostics, many of them located nearby the PFCs, are also affected by the temperature increase, which would result in the quality change on the accuracy. The reliability of auxiliary heating devices (usually beams or RF) is essential to sustain the plasma current, as noted before, and a safe termination of the discharge against possible shutdown of heating/CD is also desired.
In this lecture, through a case study of the long-pulse discharge experiences obtained over a decade in a mid-size long pulse machine, KSTAR [ Lee 2000, Oh 2018], we will discuss relevant controls for realization of long pulse tokamak operation and emphasize a few important points on designing such discharges from a pedagogical view.
[Lister 2000] Lister et al., Nucl. Fusion 40(6), p1167-1181 (2000).
[Lee 2000] Lee et al., Nucl. Fusion 40(3Y), p575-582 (2000).
[Oh 2018] Oh et al., J. of the Korean Physics Society, 73(6), p712-735 (2018).