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This paper describes an important part of ITER Research Plan, on plasma–wall interaction (PWI). In order to maximize the flexibility of the machine during the initial operation (H and D phases), CFC will be used for the targets. Tungsten will be used for the other plasma-facing components of the divertor. In order to minimize the tritium retention, tungsten will fully cover the divertor targets before the DT phase. Extrapolation of heat loads on plasma-facing components (PFCs) during disruption and ELMs to ITER parameters indicates serious consequences of these phenomena. Therefore schemes for prediction and mitigation or avoidance of these phenomena need to be developed during construction and demonstrated in the early phase of ITER operation. T-retention and dust have important impacts on safety. Therefore the methods of measurement and removal of tritium and dust must be developed during construction and demonstrated in the early phase of ITER operation.

The chapter summarizes the physics issues of the demonstration toroidal fusion power plant (Demo) that can be addressed by ITER operation. These include burning plasma specific issues, i.e. energetic particle behaviour and plasma self-heating effects, and a broader class of power-plant scale physics issues that cannot be fully resolved in present experiments. A critical issue for Demo is whether MHD and energetic particle modes driven by fast particles will become unstable and affect plasma performance. Self-heating effects are expected to be especially important for control of steady-state plasmas with internal transport barriers (ITBs) and high bootstrap current fractions. Experimental data from ITER will improve strongly the physics basis of projections to Demo of major plasma parameters such as the energy confinement time, beta and density limits, edge pedestal temperature and density, and thermal loads on in-vessel components caused by ELMs and disruptions. ITER will also serve as a test bed for fusion technology studies, such as power plant plasma diagnostics, heating and current drive systems, plasma facing components, divertor and blanket modules. Finally, ITER is expected to provide benefits for the understanding of burning plasma behaviour in other magnetic confinement schemes.

The understanding and predictive capability of transport physics and plasma confinement is reviewed from the perspective of achieving reactor-scale burning plasmas in the ITER tokamak, for both core and edge plasma regions. Very considerable progress has been made in understanding, controlling and predicting tokamak transport across a wide variety of plasma conditions and regimes since the publication of the ITER Physics Basis (IPB) document (1999 Nucl. Fusion 39 2137-2664). Major areas of progress considered here follow. (1) Substantial improvement in the physics content, capability and reliability of transport simulation and modelling codes, leading to much increased theory/experiment interaction as these codes are increasingly used to interpret and predict experiment. (2) Remarkable progress has been made in developing and understanding regimes of improved core confinement. Internal transport barriers and other forms of reduced core transport are now routinely obtained in all the leading tokamak devices worldwide. (3) The importance of controlling the H-mode edge pedestal is now generally recognized. Substantial progress has been made in extending high confinement H-mode operation to the Greenwald density, the demonstration of Type I ELM mitigation and control techniques and systematic explanation of Type I ELM stability. Theory-based predictive capability has also shown progress by integrating the plasma and neutral transport with MHD stability. (4) Transport projections to ITER are now made using three complementary approaches: empirical or global scaling, theory-based transport modelling and dimensionless parameter scaling (previously, empirical scaling was the dominant approach). For the ITER base case or the reference scenario of conventional ELMy H-mode operation, all three techniques predict that ITER will have sufficient confinement to meet its design target of Q = 10 operation, within similar uncertainties.

The 'Progress in the ITER Physics Basis' (PIPB) document is an update of the 'ITER Physics Basis' (IPB), which was published in 1999 [1]. The IPB provided methodologies for projecting the performance of burning plasmas, developed largely through coordinated experimental, modelling and theoretical activities carried out on today's large tokamaks (ITER Physics R&D). In the IPB, projections for ITER (1998 Design) were also presented. The IPB also pointed out some outstanding issues. These issues have been addressed by the Participant Teams of ITER (the European Union, Japan, Russia and the USA), for which International Tokamak Physics Activities (ITPA) provided a forum of scientists, focusing on open issues pointed out in the IPB. The new methodologies of projection and control are applied to ITER, which was redesigned under revised technical objectives. These analyses suggest that the achievement of Q > 10 in the inductive operation is feasible. Further, improved confinement and beta observed with low shear (= high βp = 'hybrid') operation scenarios, if achieved in ITER, could provide attractive scenarios with high Q (> 10), long pulse (>1000 s) operation with beta

The ITER plasma performance is assessed on the basis of the newly-proposed confinement scalings, using 1.5D simulations with the ASTRA code. In the simulations, the transport coefficients are adjusted to satisfy HH ≡ τE/τE(scaling) = 1 for each scaling. It is shown that for plasma current I = 15 MA and plasma minor radius a = 2 m for each scaling there exists an operational window within the range of moderate densities n/nG = 0.65–0.85 where nG is the Greenwald density nG (1020 m−3) = I/π a2. The fusion power Pfus = 250–600 MW can be obtained with neutral beam injection of 33 MW and ion cyclotron central heating of 0–20 MW. The operational windows predicted by the three scalings are close to each other. The fusion multiplication factor Q can achieve values in the range of 10–20 even with a more conservative assumption on helium pumping that corresponds to the ratio of the effective helium confinement time to the energy confinement time . The impact of enhanced particle confinement and increased threshold power for L to H mode transition on the ITER operational space is studied.

The requirements for pellet injection parameters for plasma fuelling are assessed for ITER scenarios with enhanced particle confinement. The assessment is based on the integrated transport simulations including models of pedestal transport, reduction of helium transport and boundary conditions compatible with SOL/divertor simulations. The requirements for pellet injection for the inductive H-mode scenario (HH98y,2 = 1) are reconsidered taking account of a possible reduction of the particle loss obtained in some experiments at low collisionalities. The assessment of fuelling requirements is carried out for the hybrid and steady state (SS) scenarios with enhanced confinement with HH98y,2 > 1. A robustness of plasma performance to the variation of particle transport is demonstrated. A new type of SS scenario is considered with neutral beam current drive and electron cyclotron current drive instead of lower hybrid current drive (LHCD) to extend the range of stable operation and to avoid the reduction of the edge LHCD efficiency caused by pellet injection.

Recent progress in experiments, modeling and theory has increased confidence that ITER can achieve its goal of high Q (>10) in inductive operation. Further, experimental results in the ‘improved hybrid’ regime suggest a possibility of high Q (>10) operation in a long pulse (>1000 s) with benign ELMs. However, considerable uncertainty still exists in the prediction of several key aspects; for example, tritium retention, disruption, impurity control, ELMs, SOL transport and dust. ITER requires flexibility of operation scenarios at least in the early operation phase to accommodate uncertainties in prediction, to explore wide operational spaces and to incorporate newly developed control schemes. Because of these uncertainties conservative assumptions are adopted in performance predictions, and step-wise implementation of reactor-relevant plasma-facing material, such as tungsten, are planned.

Tokamak discharges with improved energy confinement properties arising from internal transport barriers (ITBs) have certain attractive features, such as a large bootstrap current fraction, which suggest a potential route to the steady-state mode of operation desirable for fusion power plants. This paper first reviews the present state of theoretical and experimental knowledge regarding the formation and characteristics of ITBs in tokamaks. Specifically, the current status of theoretical modelling of ITBs is presented; then, an international ITB database based on experimental information extracted from some nine tokamaks is described and used to draw some general conclusions concerning the necessary conditions for ITBs to appear, comparing these with the theoretical models. The experimental situation regarding the steady-state, or at least quasi-steady-state, operation of tokamaks is reviewed and finally the issues and prospects for achieving such operational modes in ITER are discussed. More detailed information on the characteristics of ITBs in some 13 tokamaks (as well as helical devices) appears in the appendix.

Recent performance analysis has improved confidence in achieving Q (= fusion power/auxiliary heating power)≥ 10 in inductive operation in ITER. Performance analysis based on empirical scalings shows the feasibility of achieving Q ≥ 10 in inductive operation, particularly with improved modelling of helium exhaust. Analysis has also indicated the possibility that ITER can potentially demonstrate Q ~ 50, enabling studies of self-heated plasmas. Theory-based core modelling indicates the need for a high pedestal temperature (3.2–5.3 keV) to achieve Q ≥ 10, which is in the range of projections with presently available pedestal scalings. Pellet injection from the high-field side would be useful in enhancing Q and reducing edge localized mode (ELM) heat load in high plasma current operation. If the ELM heat load is not acceptable, it could be made tolerable by further tilting the target plate. Steady state operation scenarios at Q = 5 have been developed with modest requirements on confinement improvement and beta (HH98(y,2) ≥ 1.3 and βN ≥ 2.6). Stabilization of the resistive wall modes (RWMs), required in such regimes, is feasible with the present saddle coils and power supplies with double-wall structures taken into account. Recent analysis shows a potential of high power steady state operation with a fusion power of 0.7 GW at Q ~ 8. Achievement of the required βN ~ 3.6 by RWM stabilization is a possibility. Further analysis is also needed on reduction of the divertor target heat load.

Advanced tokamak regimes obtained in ASDEX Upgrade, DIII-D, FT-U, JET, JT-60U, TCV and Tore Supra experiments are assessed both in terms of their fusion performance and capability for ultimately reaching steady-state using data from the international internal transport barrier database. These advanced modes of tokamak operation are characterized by an improved core confinement and a modified current profile compared to the relaxed Ohmically driven one. The present results obtained in these experiments are studied in view of their prospect for achieving either long pulses ('hybrid' scenario with inductive and non-inductive current drive) or ultimately steady-state purely non-inductive current drive operation in next step devices such as ITER. A new operational diagram for advanced tokamak operation is proposed where the figure of merit characterizing the fusion performances and confinement, H × βN / q 295, is drawn versus the fraction of the plasma current driven by the bootstrap effect. In this diagram, present day advanced tokamak regimes have now reached an operational domain that is required in the non-inductive ITER current drive operation with typically 50% of the plasma current driven by the bootstrap effect (Green et al 2003 Plasma Phys. Control. Fusion 45 587). In addition, the existence domain of the advanced mode regimes is also mapped in terms of dimensionless plasmas physics quantities such as normalized Larmor radius, normalized collisionality, Mach number and ratio of ion to electron temperature. The gap between present day and future advanced tokamak experiments is quantitatively assessed in terms of these dimensionless parameters.

ITER will be the first magnetic confinement device with burning DT plasma and fusion power of about 0.5 GW. Parameters of ITER plasma have been predicted using methodologies summarized in the ITER Physics Basis (1999 Nucl. Fusion 39 2175). During the past few years, new results have been obtained that substantiate confidence in achieving Q ≥ 10 in ITER with inductive H-mode operation. These include achievement of a good H-mode confinement near the Greenwald density at high triangularity of the plasma cross section; improvements in theory-based confinement projections for the core plasma, even though further studies are needed for understanding the transport near the plasma edge; improvement in helium ash removal due to the elastic collisions of He atoms with D/T ions in the divertor predicted by modelling; demonstration of feedback control of neoclassical tearing modes and resultant improvement in the achievable β-values; better understanding of edge localized mode (ELM) physics and development of ELM mitigation techniques; and demonstration of mitigation of plasma disruptions. ITER will have a flexibility to operate also in steady-state and intermediate (hybrid) regimes. The 'advanced tokamak' regimes with weak or negative central magnetic shear and internal transport barriers are considered as potential scenarios for steady-state operation. The paper concentrates on inductively driven plasma performance and discusses requirements for steady-state operation in ITER.

The values of Q = (fusion power)/(auxiliary heating power) predicted for ITER by three different methods are compared. The first method utilizes an empirical confinement-time scaling and prescribed radial profiles of transport coefficients; the second approach extrapolates from specially designed ITER similarity experiments, and the third approach is based on partly theory-based transport models. The energy confinement time given by the ITERH-98(y, 2) scaling for an inductive scenario with a plasma current of 15 MA and a plasma density 15% below the Greenwald density is 3.7 s with one estimated technical standard deviation of ±14%. This translates, in the first approach, for levels of helium removal, and impurity concentration, that, albeit rather stringent, are expected to be attainable, into an interval for Q of [6–15] at the auxiliary heating power, Paux = 40 MW, and [6–30] at the minimum heating power satisfying a good confinement ELMy H-mode. All theoretical transport-model calculations have been performed for the plasma core only, whereas the pedestal temperatures were taken as estimated from empirical scalings. Predictions of similarity experiments from JET and of theory-based transport models that we have considered—Weiland, MMM, and IFS/PPPL—overlap with the prediction using the empirical confinement-time scaling within its estimated margin of uncertainty.

Scaling of H-mode edge pedestal pressure for a Type-I ELM regime based on a simple analytical formula (Hatae T et al 2001 Nucl. Fusion 41 285) can be significantly improved by introducing shaping factors into the scaling, which is motivated by a recent development of the MHD model for the critical pressure gradient (Snyder P B et al 2002 Phys. Plasmas 5 2037, Lao L L et al 2002 ITPA H-mode Edge Pedestal Physics Topical Group Meeting (San Diego)). The physics basis of these shaping factors is an enhancement of the critical pressure gradient determined by a high toroidal mode number, n, ideal ballooning mode through the effect of the magnetic well. On an increase in the magnetic well, peeling and ballooning modes decouple and the critical gradient is determined by the intermediate n mode, which enhances the critical gradient substantially. The improved scaling can reproduce the pedestal pressure reasonably well for ASDEX-U, JET and JT-60U Type-I ELMy data archived in the International Pedestal Database version 3.

ITER is planned to be the first fusion experiment operating under reactor-relevant conditions. The following requirements are given: ITER should (1) achieve extended burn in inductively driven plasma at Q ≥ 10, whilst not precluding the possibility of controlled ignition, (2) aim at demonstrating steady-state operation using non-inductive current drive at Q ≥ 5, (3) demonstrate availability and integration of essential fusion technologies, and (4) test components for a future reactor including tritium breeding module concepts, with the 14 MeV neutron power load on the first wall ≥ 0.5 MW m-2 and fluence ≥ 0.3 MW am-2. The ITER tokamak is designed not only to satisfy these requirements but also to have flexibility of plasma operations. This flexibility will provide a wide range of opportunities to develop inductive and non-inductive reactor core plasmas and to study burning plasma physics. Possible operational modes studying physics and technical issues of burning plasmas are overviewed.

The plasma performance in two design options of the reduced-technical objectives/reduced cost (RTO/RC) ITER, i.e. IAM (intermediate aspect ratio machine) and LAM (low aspect ratio machine) is analysed. It is shown that Q = Pfus/Paux~10 can be obtained in both options at inductively driven ELMy H-mode operation. The operation domain in LAM is found to be marginally larger than that in IAM. The non-inductive operation with Q≈5 will be possible in both machines, provided a large amount of power with a high current drive efficiency is applied, or substantial improvement of the energy confinement time relative to the ELMy H-mode (HH = 1.2-1.4) is obtained. The required values of HH and βN are marginally smaller in IAM. The IAM-like machine, ITER-FEAT (fusion energy advanced tokamak), proposed for a detailed engineering design is discussed in brief.

Features incorporated in the design of the International Thermonuclear Experimental Reactor (ITER) tokamak and its ancillary and plasma diagnostic systems that facilitate operation and control of ignited and/or high?Q DT plasmas are presented. Control methods based upon straightforward extrapolation of techniques employed in the present generation of tokamaks are found to be adequate and effective for ITER plasma control with fusion powers of up to 1.5?GW and burn durations of ? 1000?s. Examples of simulations of key plasma control functions, including plasma magnetic configuration control and fusion burn (power) control, are given. The prospects for the creation and control of steady state plasmas sustained by non-inductive current drive and bootstrap current are also discussed.

Plasma control requirements for the International thermonuclear Experimental Reactor (ITER) are identified, and an overview of proposed ITER plasma control concepts is presented. ITER will operate with a burning deuterium-tritium plasma to produce 1.5 GW of fusion power for durations of 1000 s or more. Key plasma control requirements to achieve these objectives encompass (a) plasma scenario and sequencing: plasma initiation, current rampup, divertor formation, auxiliary heating, ignition and burn, deignition (fusion power shutdown), and current rampdown and termination; (b) plasma magnetics control: plasma current and shape (R 0 , a, κ, δ) versus time, plus control of critical plasma-to-first-wall clearance gaps, including ion-cyclotron coupling gap and divertor magnetic configuration, during the diverted heating/ignition/burn/deignition phase of the plasma scenario; (c) plasma kinetics and divertor control: core plasma density and/or fusion power, core impurity content and/or radiated power fraction; core profile control (auxiliary heating and/or current drive), and divertor control (pumping, in-divertor gas and/or impurity injection and magnetic configuration optimization for divertor performance); and (d) fast plasma shutdown: fusion power and current shutdown by means of impurity injection. Physics and hardware concepts are presented as to how these plasma control junctions will be implemented. Diagnostic measurements needed for plasma control are summarized. The relationship of plasma control to machine protection and public safety is also addressed.

Key objectives of the first ten years of ITER operation are the investigation of the physics of burning plasmas and the demonstration of long-pulse ignited plasma technologies. These include studies of plasma confinement and stability, divertor operation, disruption mitigation and control, noninductive current drive, and steady state operation under conditions when the plasma is heated predominantly by alpha particles. The ITER operational plan envisages two and a half years for commissioning and initial operation with hydrogen plasmas at up to 100 MW of auxiliary heating power when initial tests of divertor operation and evaluation of disruption effects will be made. In order to meet the operational and programmatic goals, it will be necessary to make a wide range of plasma measurements. In this article the preliminary operational plan and physics program are presented and the implications for plasma measurements are outlined.

H-mode operation is projected to be needed in ITER in order to achieve its objectives of sustained burn for 1000 s with 1.5 GW of fusion power. However, experiments have shown that a minimum level of input power is required to access and sustain the H-mode regime. This paper studies the impact on ITER operation of the presently available scalings of this minimum input or threshold power for both the direct L- to H- and reverse H- to L-mode transitions. We present scenarios for access into the H-mode, for maintaining the plasma in H-mode and for the fusion power shutdown with the reverse transition into L-mode. The commonly used empirical scaling for the L- to H-mode transition with a reduction of a factor of two for the reverse transition has severe consequences on ITER operation and performance and various methods to quantify this impact are presented. We conclude, however, that the ITER mission and physics objectives remain compatible with such a scaling.

Our understanding of the physics of internal transport barriers (ITBs) is being advanced by analysis and comparisons of experimental data from many different tokamaks worldwide. An international database consisting of scalar and two-dimensional profile data for ITB plasmas is being developed to determine the requirements for the formation and sustainment of ITBs and to perform tests of theory-based transport models in an effort to improve the predictive capability of the models. Analysis using the database indicates that: (a) the power required to form ITBs decreases with increased negative magnetic shear of the target plasma, and: (b) the E x B flow shear rate is close to the linear growth rate of the ion temperature gradient (ITG) modes at the time of barrier formation when compared for several fusion devices. Tests of several transport models (JETTO, Weiland model) using the two-dimensional profile data indicate that there is only limited agreement between the model predictions and the experimental results for the range of plasma conditions examined for the different devices (DIII-D, JET, JT-60U). Gyrokinetic stability analysis (using the GKS code) of the ITB discharges from these devices indicates that the ITG/TEM growth rates decrease with increased negative magnetic shear and that the E x B shear rate is comparable to the linear growth rates at the location of the ITB.

The physics of ITER (International Thermonuclear Experimental Reactor) is based on demonstrated tokamak physics and credible extrapolations of that physics. Assessment of the energy confinement and MHD stability requirements led to the choice of the major plasma parameters of 22 MA for the plasma current, a toroidal field of approximately 5 T, an aspect ratio of approximately 3, and an elongation of approximately 2. Among the major accomplishments of the physics group has been the development of a database and an empirical scaling for L-mode energy confinement and the facilitation of an H-mode database and scaling. The divertor heat loads have been estimated by using experimentally validated models. The thermal and mechanical loads due to off-normal events such as disruptions have been based on analysis of the data from other tokamaks. To achieve the required availability of 10%, the pulse length has been extended by the use of current drive using 75 MW 1.3 MeV neutral beam and 45 MW lower hybrid systems. A relatively complete set of plasma diagnostics is planned for ITER. Finally, a physics R and D program has been developed. >

Specific performance from imposes loads on other supporting subsystems and subsystems. This paper overviews system requirements dictated by the plasma's needs and impacts. Overall device parameters have been derived from the performance goals using plasma transport and stability models based on physics R and D results. Most transport issues relate to the extrapolation of the plasma behavior to ITER's large size (especially relative to the gyroradius). Plasma control systems must achieve shape and position flexibility sufficient for the performance of the power handling and auxiliary power systems. Induced currents in the conducting structures between the poloidal coils and the plasma both provide passive stabilization of plasma disturbances and delay plasma responses to external magnetic controls. Auxiliary power systems must heat the plasma to ignition and drive noninductive currents for steady state operation, and may have to drive plasma rotation if physics confirms the need for rotation indicated by recent theories of resistive wall stabilization. Diagnostics must support plasma control and performance assessment and contribute to physics understanding. The first wall must handle steady state power and energetic neutral bombardment, and will experience transient power and runaway electron loads. The divertor system must disperse the incoming power while confining the divertor's higher concentration neutral gas and pumping the helium. In-vessel structural design is driven by the forces related to plasma disruptions, which induce currents in the conducting structures and drive non-axisymmetric "halo currents" between the plasma and the conductors. Such plasma control needs and impacts from power and disruption handling are described.

The individual diagnostics which constitute the ITER Diagnostic System are outlined and the present state of development of the designs is summarised. The results of an assessment of the overall performance of the System are presented and the areas where the probable performance falls below the target specifications are identified. The design and RsD plans which are in place to address the shortcomings are outlined.

Measurement of plasma and key first wall parameters will have three main roles on ITER. Some of the measurements will be used in real time to prevent the on-set of conditions which could potentially damage the first wall and other in-vessel components (machine protection); others will be used in real-time feedback control loops to control the value of key parameters at values required for specific plasma performance (plasma control); while others will be used to evaluate the plasma performance and to provide information on key phenomena which may limit ITER performance (physics studies). The measurements of some parameters may contribute to all three roles although the requirements on the measurements (accuracies, resolutions etc.) may be different depending on the role.

The design of the major diagnostic systems for ITER has now progressed from generic approaches1 to the engineering and design integration phases. The design of relevant parts of the ITER machine, in particular the access points for diagnostics such as ports and blanket modules, has developed significantly2. Diagnostic equipment is still found in the three main port areas (equatorial, vertical and divertor) and in the blanket back plate, but no longer in the blanket modules. Diagnostic access at these sites is considered in terms of the access to the plasma through the first wall and the implementation of the signal transmission paths away from the Tokamak. The arrangements for the installation and maintenance of the diagnostic components are described.

Radiation effects on key components of diagnostic systems expected to be subjected to high neutron and gamma fluxes and fluences are being examined in irradiation tests to evaluate and establish an ITER-relevant database to support the design. A comprehensive irradiation database has been accumulated and permits conclusions to be drawn on the application of these components in ITER. The design studies on prototypical assemblies of diagnostic components are continuing based on the irradiation data bases, neutronics calculations for evaluating irradiation environment of diagnostic components and required specification of diagnostic systems. These studies will aid the recognition of detailed requirements of diagnostic systems leading to more specific irradiation tests on diagnostic components.

Publisher Summary The International Thermonuclear Experimental Reactor diagnostic system consists of about 40 individual measurement systems to control and evaluate the ignited plasmas. These measurement systems involve magnetics, neutronics, laser and optical systems, X-Ray to visible spectroscopy, microwaves, probes, pressure gauges and thermocouples. Components are located in several different locations: at main ports, in the vacuum vessel, inside and outside the cryostat, and in buildings. Each area poses its own set of design requirements on the equipment. Although the equipment in the vertical port varies, a standard hole is provided for the plugs of diagnostic equipment required to be installed in the penetrations. Diagnostic sensors, such as magnetic pick-up coils, bolometer cameras, retroreflectors, and antennas are installed in the blanket modules. The modules also require instrumentation, such as thermocouples and strain gauges. Several modules are modified with holes or enlarged edge gaps for diagnostic purposes, although as much use as possible is made of the 20mm gap which exists between modules for installation purposes.

The requirements for diagnostics on ITER are determined from a consideration of the role plasma measurements will play in the operation and evaluation of the tokamak. The proposed list of measurements and the detailed measurement specifications - parameter ranges, target measurement resolutions and accuracies - are presented. The principal practical problems that will have to be overcome in implementing the diagnostics are outlined and areas where particular difficulties arise are identified.

The magnetic diagnostic for ITER has to meet the classical requirements of the magnetic diagnostic on the present day large divertor tokamaks and has in addition the requirement to work during the long pulse in excess of 1000s and in a hostile radiation environment.