Table of contents

The ITER Physics Basis presents and evaluates the physics rules and methodologies for plasma performance projections, which provide the basis for the design of a tokamak burning plasma device whose goal is to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes. This Chapter summarizes the physics basis for burning plasma projections, which is developed in detail by the ITER Physics Expert Groups in subsequent chapters. To set context, the design guidelines and requirements established in the report of ITER Special Working Group 1 are presented, as are the specifics of the tokamak design developed in the Final Design Report of the ITER Engineering Design Activities, which exemplifies burning tokamak plasma experiments. The behaviour of a tokamak plasma is determined by the interaction of many diverse physics processes, all of which bear on projections for both a burning plasma experiment and an eventual tokamak reactor. Key processes summarized here are energy and particle confinement and the H-mode power threshold; MHD stability, including pressure and density limits, neoclassical islands, error fields, disruptions, sawteeth, and ELMs; power and particle exhaust, involving divertor power dispersal, helium exhaust, fuelling and density control, H-mode edge transition region, erosion of plasma facing components, tritium retention; energetic particle physics; auxiliary power physics; and the physics of plasma diagnostics. Summaries of projection methodologies, together with estimates of their attendant uncertainties, are presented in each of these areas. Since each physics element has its own scaling properties, an integrated experimental demonstration of the balance between the combined processes which obtains in a reactor plasma is inaccessible to contemporary experimental facilities: it requires a reactor scale device. It is argued, moreover, that a burning plasma experiment can be sufficiently flexible to permit operation in a steady state mode, with non-inductive plasma current drive, as well as in a pulsed mode where current is inductively driven. Overall, the ITER Physics Basis can support a range of candidate designs for a tokamak burning plasma facility. For each design, there will remain a significant uncertainty in the projected performance, but the projection methodologies outlined here do suffice to specify the major parameters of such a facility and form the basis for assuring that its phased operation will return sufficient information to design a prototype commercial fusion power reactor, thus fulfilling the goal of the ITER project.

Physics knowledge in plasma confinement and transport relevant to design of a reactor-scale tokamak is reviewed and methodologies for projecting confinement properties to ITER are provided. Theoretical approaches to describing a turbulent plasma transport in a tokamak are outlined and phenomenology of major energy confinement regimes observed in tokamaks, including those with edge and internal transport barriers, is described. The chapter is focused on the energy confinement in the high confinement regime (H-mode) with the edge localized MHD modes, the basic operational regime of ITER. Three approaches are being pursued: (i) derivation of empirical global scaling laws; (ii) non-dimensionally similar studies; and (iii) one dimensional transport modelling codes, with the first approach recommended as the most robust at the present time. Special attention is paid to analysis of uncertainties in confinement predictions. Empirical scaling relations for projecting the L-mode to H-mode power threshold based on regression analysis of an extensive database are discussed. Particle and toroidal momentum confinement and their relation to energy confinement are reviewed.

The present physics understandings of magnetohydrodynamic (MHD) stability of tokamak plasmas, the threshold conditions for onset of MHD instability, and the resulting operational limits on attainable plasma pressure (beta limit) and density (density limit), and the consequences of plasma disruption and disruption related effects are reviewed and assessed in the context of their application to a future DT burning reactor prototype tokamak experiment such as ITER. The principal considerations covered within the MHD stability and beta limit assessments are (i) magnetostatic equilibrium, ideal MHD stability and the resulting ideal MHD beta limit; (ii) sawtooth oscillations and the coupling of sawtooth activity to other types of MHD instability; (iii) neoclassical island resistive tearing modes and the corresponding limits on beta and energy confinement; (iv) wall stabilization of ideal MHD instabilities and resistive wall instabilities; (v) mode locking effects of non-axisymmetric error fields; (vi) edge localized MHD instabilities (ELMs, etc.); and (vii) MHD instabilities and beta/pressure gradient limits in plasmas with actively modified current and magnetic shear profiles. The principal considerations covered within the density limit assessments are (i) empirical density limits; (ii) edge power balance/radiative density limits in ohmic and L-mode plasmas; and (iii) edge parameter related density limits in H-mode plasmas. The principal considerations covered in the disruption assessments are (i) disruption causes, frequency and MHD instability onset; (ii) disruption thermal and current quench characteristics; (iii) vertical instabilities (VDEs), both before and after disruption, and plasma and in-vessel halo currents; (iv) after disruption runaway electron formation, confinement and loss; (v) fast plasma shutdown (rapid externally initiated dissipation of plasma thermal and magnetic energies); (vi) means for disruption avoidance and disruption effect mitigation; and (vii) `integrated' modelling of disruptions and fast shutdown and of the ensuing effects. In each instance, the presentation within a given topical area progresses from a summary of present experimental and theoretical understanding to how this understanding projects or extrapolates to an ITER class reactor regime tokamak. Examples of extrapolations to the specific ITER design concept developed during the course of the ITER EDA are given, and assessments of the degree of adequacy of present understanding are also provided. In areas where present understanding is identified to be less than fully adequate, areas in which continuing or new research is needed are identified.

The ability to exhaust the plasma power loss from a large tokamak onto material walls surrounding the plasma has been perceived to be a large obstacle to the successful production of a fusion power reactor in the past. There have been tremendous strides in understanding the physics relevant to this power exhaust over the past five years. This improvement in understanding has arisen because of both improved diagnostics of the plasma outside the last closed flux surface, and because of improved two dimensional computer models of this plasma. This understanding has led to innovative plasma solutions that reduce the power load to the divertor region of ITER to levels that are acceptable for a successful engineering design of the divertors. These plasma solutions have been realized in the devices that are active today. Analysis using the improved plasma models also indicates that particle control, both of fuel and impurity particles, is adequate for successful operation of ITER. This paper presents the current status of both the experimental and theoretical understanding of the plasma, neutral and atomic physics relevant to the plasma at the edge of fusion devices. Since understanding of the subject of this paper is progressing rapidly, we should emphasize that this paper was written in the spring of 1998 and, as such, presents the status of the subject at that time.

Physics knowledge (theory and experiment) in energetic particles relevant to design of a reactor scale tokamak is reviewed, and projections for ITER are provided in this Chapter of the ITER Physics Basis. The review includes single particle effects such as classical alpha particle heating and toroidal field ripple loss, as well as collective instabilities that might be generated in ITER plasmas by energetic alpha particles. The overall conclusion is that fusion alpha particles are expected to provide an efficient plasma heating for ignition and sustained burn in the next step device. The major concern is localized heat loads on the plasma facing components produced by alpha particle loss, which might affect their lifetime in a tokamak reactor.

Heating and current drive systems must fulfil several roles in ITER operating scenarios: heating through the H-mode transition and to ignition; plasma burn control; current drive and current profile control in steady state scenarios; and control of MHD instabilities. They must also perform ancillary functions, such as assisting plasma start-up and wall conditioning. It is recognized that no one system can satisfy all of these requirements with the degree of flexibility that ITER will require. Four heating and current drive systems are therefore under consideration for ITER: electron cyclotron waves at a principal frequency of 170GHz; fast waves operating in the range 40-70MHz (ion cyclotron waves); lower hybrid waves at 5GHz; and neutral beam injection using negative ion beam technology for operation at 1MeV energy. It is likely that several of these systems will be employed in parallel. The systems have been chosen on the basis of the maturity of physics understanding and operating experience in current experiments and on the feasibility of applying the relevant technology to ITER. Here, the fundamental physics describing the interaction of these heating systems with the plasma is reviewed, the relevant experimental results in the exploitation of the heating and current drive capabilities of each system are discussed, key aspects of their application to ITER are outlined, and the major technological developments required in each area are summarized.

The physics issues of the measurements of the plasma properties necessary to provide both the control and science data for achieving the goals of the ITER device are discussed. The assessment of the requirements for these measurements is first discussed, together with priorities that relate to the experimental program. Subsequently, some of the proposed measurement techniques, the plasma diagnostics, are described with particular emphasis on their implementation on ITER and their capability to meet the requirements. A judgement on the present status of the diagnostic program on ITER is provided with some indication of the research and development program necessary to demonstrate viability of techniques or their implementation.

Wall conditioning of fusion devices involves removal of desorbable hydrogen isotopes and impurities from interior device surfaces to permit reliable plasma operation. Techniques used in present devices include baking, metal film gettering, deposition of thin films of low-Z material, pulse discharge cleaning, glow discharge cleaning, radio frequency discharge cleaning, and in situ limiter and divertor pumping. Although wall conditioning techniques have become increasingly sophisticated, a reactor scale facility will involve significant new challenges, including the development of techniques applicable in the presence of a magnetic field and of methods for efficient removal of tritium incorporated into co-deposited layers on plasma facing components and their support structures. The current status of various approaches is reviewed, and the implications for reactor scale devices are summarized. Creation and magnetic control of shaped and vertically unstable elongated plasmas have been mastered in many present tokamaks. The physics of equilibrium control for reactor scale plasmas will rely on the same principles, but will face additional challenges, exemplified by the ITER/FDR design. The absolute positioning of outermost flux surface and divertor strike points will have to be precise and reliable in view of the high heat fluxes at the separatrix. Long pulses will require minimal control actions, to reduce accumulation of AC losses in superconducting PF and TF coils. To this end, more complex feedback controllers are envisaged, and the experimental validation of the plasma equilibrium response models on which such controllers are designed is encouraging. Present simulation codes provide an adequate platform on which equilibrium response techniques can be validated. Burning plasmas require kinetic control in addition to traditional magnetic shape and position control. Kinetic control refers to measures controlling density, rotation and temperature in the plasma core as well as in plasma periphery and divertor. The planned diagnostics (Chapter 7) serve as sensors for kinetic control, while gas and pellet fuelling, auxiliary power and angular momentum input, impurity injection, and non-inductive current drive constitute the control actuators. For example, in an ignited plasma, core density controls fusion power output. Kinetic control algorithms vary according to the plasma state, e.g. H- or L-mode. Generally, present facilities have demonstrated the kinetic control methods required for a reactor scale device. Plasma initiation - breakdown, burnthrough and initial current ramp - in reactor scale tokamaks will not involve physics differing from that found in present day devices. For ITER, the induced electric field in the chamber will be ∼0.3V· m^-1 - comparable to that required by breakdown theory but somewhat smaller than in present devices. Thus, a start-up 3MW electron cyclotron heating system will be employed to assure burnthrough. Simulations show that plasma current ramp up and termination in a reactor scale device can follow procedures developed to avoid disruption in present devices. In particular, simulations remain in the stable area of the l_i-q plane. For design purposes, the resistive V·s consumed during initiation is found, by experiments, to follow the Ejima expression, 0.45μ₀ RI_p. Advanced tokamak control has two distinct goals. First, control of density, auxiliary power, and inductive current ramping to attain reverse shear q profiles and internal transport barriers, which persist until dissipated by magnetic flux diffusion. Such internal transport barriers can lead to transient ignition. Second, combined use poloidal field shape control with non-inductive current drive and NBI angular momentum injection to create and control steady state, high bootstrap fraction, reverse shear discharges. Active n = 1 magnetic feedback and/or driven rotation will be required to suppress resistive wall modes for steady state plasmas that must operate in the wall stabilized regime for reactor levels of β ⩾ 0.03.

A reactor scale tokamak plasma will exhibit three areas of physics phenomenology not accessible by contemporary experimental facilities. These are: (1) instabilities generated by energetic alpha particles; (2) self-heating phenomena; and (3) reactor scale physics, which includes integration of diverse physics phenomena, each with its own scaling properties. In each area, selected examples are presented that demonstrate the importance and uniqueness of physics results from reactor scale facilities for both inductive and steady state reactor options. It is concluded that the physics learned in such investigations will be original physics not attainable with contemporary facilities. In principle, a reactor scale facility could have a good measure of flexibility to optimize the tokamak approach to magnetic fusion energy.