Chapter 1: Overview and summary

Abstract

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.

Indirect measurement of poloidal rotation using inboard–outboard asymmetry of toroidal rotation and comparison with neoclassical predictions

“…For example, in ITER [40], an appropriate arrangement of a x-ray crystal based diagnostic [41] could provide a viable way to measure u t , u p and E r in the plasma core, where CX diagnostics will be limited by beam attenuation.…”

Experimental investigation of ion cyclotron range of frequencies heating scenarios for ITER's half-field hydrogen phase performed in JET

“…The non-active phase of ITER [1] will start with hydrogen plasmas at a reduced magnetic field. At the nominal halffield of B 0 = 2.65 T and with the auxiliary power currently foreseen to be available in this phase, 16.5 MW off-axis neutral beam injection (NBI), 15 MW off-axis electron cyclotron resonance heating (ECRH) and 10 MW on-axis ion cyclotron resonance heating (ICRH) or radio-frequency (RF) heating, the discharges are expected to be in L-mode, and typical central densities of n 0 ≈ 3 × 10 19 m −3 and central temperatures of approximately T i = 8 keV and T e = 10 keV have been estimated [2].…”

Fast-ion transport and neutral beam current drive in ASDEX upgrade

“…Well confined fast ions are, hence, essential to obtain good heating and current drive performances. In addition, detailed understanding of the fast-ion transport mechanisms is needed in view of future fusion devices, not only to guarantee good performance, but also to ensure the safety of the machine: fast helium ions will be produced in fusion reactions that could, if poorly confined, damage the first wall [2]. In toroidally axisymmetric devices such as tokamaks, the fast-ion transport induced by collisions and orbit effects, i.e.…”