Search

The focus of effort in ITER EDA since 1998 has been on the development of a new design to meet revised technical objectives and a cost reduction target of about 50% of the previously accepted cost estimate. Drawing on the design solutions already developed, and using the latest physics results and outputs from technology R&D projects, the Joint Central Team and Home Teams, working together, have been able to progress towards a new design which will allow the exploration of a range of burning plasma conditions, with a capacity to progress towards possible modes of steady state operation. The new ITER design, whilst having reduced technical objectives from those of its predecessor, will nonetheless meet the programmatic objective of providing an integrated demonstration of the scientific and technological feasibility of fusion energy. The main features of the current design and of its projected performance are introduced and the outlook for construction and operation is summarized.

The magnet design of the new ITER-FEAT machine comprises 18 toroidal field (TF) coils, a central solenoid (CS), 6 poloidal field coils and correction coils. A key driver of this new design is the requirement to generate and control plasmas with a relatively high elongation (κ95 = 1.7) and a relatively high triangularity (δ95 = 0.35). This has led to a design where the CS is vertically segmented and self-standing and the TF coils are wedged along their inboard legs. Another important design driver is the requirement to achieve a high operational reliability of the magnets, and this has resulted in several unconventional designs, and in particular the use of conductors supported in radial plates for the winding pack of the TF coils. A key mechanical issue is the cyclic loading of the TF coil cases due to the out-of-plane loads which result from the interaction of the TF coil current and the poloidal field. These loads are resisted by a combination of shear keys and `pre-compression' rings able to provide a centripetal preload at assembly. The fatigue life of the CS conductor jacket is another issue, as it determines the CS performance in terms of the flux generation. Two jacket materials and designs are under study. Since 1993, the ITER magnet R&D programme has been focused on the manufacture and testing of a CS and a TF model coil. During its testing, the CS model coil has successfully achieved all its performance targets in DC and AC operations. The manufacture of the TF model coil is complete. The manufacture of segments of the full scale TF coil case is another important and successful part of this programme and is near completion. New R&D effort is now being initiated to cover specific aspects of the ITER-FEAT design.

ITER is planned to be the first fusion experimental reactor in the world operating for research in physics and engineering. The first ten years of operation will be devoted primarily to physics issues at low neutron fluence and the following ten years of operation to engineering testing at higher fluence. ITER can accommodate various plasma configurations and plasma operation modes, such as inductive high Q modes, long pulse hybrid modes and non-inductive steady state modes, with large ranges of plasma current, density, beta and fusion power, and with various heating and current drive methods. This flexibility will provide an advantage for coping with uncertainties in the physics database, in studying burning plasmas, in introducing advanced features and in optimizing the plasma performance for the different programme objectives. Remote sites will be able to participate in the ITER experiment. This concept will provide an advantage not only in operating ITER for 24 hours a day but also in involving the worldwide fusion community and in promoting scientific competition among the ITER Parties.

A configuration in which the vacuum vessel (VV) fits tightly to the plasma aids the passive plasma vertical stability, and ferromagnetic material in the VV reduces the toroidal field ripple. The blanket modules are supported directly by the VV. A full scale VV sector model has provided critical information related to fabrication technology and for testing the magnitude of welding distortions and achievable tolerances. This R&D validated the fundamental feasibility of the double wall VV design. The blanket module configuration consists of a shield body to which a separate first wall is mounted. The separate first wall has a facet geometry consisting of multiple flat panels, where 3-D machining will not be required. A configuration with deep slits minimizes the induced eddy currents and loads. The feasibility and robustness of solid hot isostatic pressing joining were demonstrated in the R&D by manufacturing and testing several small and medium scale mock-ups and finally two prototypes. Remote handling tests and assembly tests of a blanket module have demonstrated the basic feasibility of its installation and removal.

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.

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.

The remote maintenance strategy and the associated component design of the International Thermonuclear Experimental Reactor (ITER) have reached a high degree of completeness, especially with respect to those components that are expected to require frequent or occasional remote maintenance. Large scale test stands, to demonstrate the feasibility in principle of the remote maintenance procedures and to develop the required equipment and tools, were operational at the end of the Engineering Design Activities phase. The initial results are highly encouraging: major remote equipment deployment and component replacement operations have been successfully demonstrated.

The magnet system design for the International Thermonuclear Experimental Reactor (ITER) has reached a high degree of integration to meet performance and operation requirements, including reliability and maintainability, in a cost effective manner. The article identifies the requirements of long inductive burn time, large number of tokamak pulses, operational flexibility for the poloidal field system, magnet reliability and the cost constraints as the main design drivers. Key features of the magnet system which stem from these design drivers are described, together with interfaces and integration aspects of certain auxiliary systems.

The present status of the ITER project is summarised in terms of progress made on its scientific, engineering and safety/environmental characteristics and of its position in world-wide fusion development. The selection of materials and joining technologies for ITER is discussed, with emphasis on materials choices and test results for in-vessel components. Progress of major R&D projects to validate the technologies is presented. The future prospects for ITER are considered.

Present status and understanding of the principal plasma-performance determining physics issues that affect the physics design and operational capabilities of the International Thermonuclear Experimental Reactor (ITER) [ITER EDA Agreement and Protocol 2 (International Atomic Energy Agency, Vienna, 1994)] are presented. Emphasis is placed on the five major physics-basis issues—energy confinement, beta limit, density limit, impurity dilution and radiation loss, and the feasibility of obtaining partial-detached divertor operation—that directly affect projections of ITER fusion power and burn duration performance. A summary of these projections is presented and the effect of uncertainties in the physics-basis issues is examined. ITER capabilities for experimental flexibility and plasma-performance optimization are also described, and how these capabilities may enter into the ITER physics program plan is discussed.

This paper reviews the status of the design of the divertor and first-wall/shield, the main in-vessel components for ITER. Under nominal ignited conditions, 300 MW of alpha power will be produced and must be removed from the divertor and first-wall. Additional power from auxiliary sources up to the level of 100 MW must also be removed in the case of driven burns. In the ignited case, about 100 MW will be radiated to the first wall as bremsstrahlung. Allowing the remaining power to be conducted to the divertor target plates would result in excessive heat fluxes. The power handling strategy is to radiate an additional 100–150 MW in the SOL and the divertor channel via a combination of radiation from hydrogen, and intrinsic and seeded impurities. Vertical targets have been adopted for the baseline divertor configuration. This geometry promotes partial detachment, as found in present experiments and in the results of modelling runs for ITER conditions, and power densities on the target plates can be ≤ 5 MW / m 2 . Such regimes promote relatively high pressure (> 1 Pa ) in the divertor and even with a low helium enrichment factor of 0.2, the required pumping speed to pump helium is ≤ 50 m 3 / s . An important physics question is the quality of core confinement in these attractive divertor regimes. In addition to power and particle handling issues, the effects of disruptions play a major role in the design and performance of in-vessel components. Both centered disruptions and VDE's produce stresses in the first-wall/shield modules, backplate and the divertor wings and cassettes that are near or even somewhat in excess of allowables for normal operation. Also plasma-wall contact from disruptions, including at the divertor target, together with material properties are major factors determining component lifetime. Considering the potential for impurity contamination and minimizing tritium inventory as well as thermomechanical performance, the present material selection calls for carbon divertor targets near the strike point, tungsten on the rest of the target and on the baffle where the charge-exchange flux could be high, and beryllium elsewhere. All three materials and relevant joining techniques are being developed in the R&D program and the final selection for the first assembly will be made at the end of the EDA.

Neutron diagnostics will play a prominent role in the control and evaluation of thermonuclear plasmas in ignition device to test engineering concepts (ITER). As in present D-T experiments, measurements of neutron yield and of fusion power and power density are essential. In addition, the spectral width of the 14.1-MeV t(d,n)α neutron emission should be a reliable indicator of ion temperature in an ignited plasma. More detailed measurements of the neutron spectrum may allow determination of the densities of tritium, deuterium, and confined alpha particles. Although the central fusion power density in ITER will be comparable to the maximum values obtainable in TFTR and JET, neutron flux on the first wall will be ten times higher, and the neutron yield per discharge will be about five orders of magnitude greater than previously experienced. The thermal and radiation shielding necessary to protect the ITER superconducting coils from the intense flux at the first wall will restrict diagnostic access for neutron cameras and spectrometers, complicate the design of material activation systems, and limit the applicability of conventional calibration techniques for neutron source strength monitors. These considerations, together with unprecedented reliability requirements and the need for full remote handling of many components, pose demanding challenges for the design of the ITER neutron diagnostic systems.

Reflectometry will be used on ITER to measure the density profile in the main plasma and divertor regions, and to measure the plasma position and shape in order to provide a standby reference for the magnetic diagnostics in long pulse discharges. The high temperatures of the ITER core and the resultant significant relativistic downshift of the second-harmonic electron cyclotron absorption imply that both low-field side O-mode and high-field side lower cut-off (X−l mode) systems are required to access the full plasma profile. A low-field side upper cut-off (X−u mode) system will also be required for measurements of the scrape-off layer. For measurements of the plasma position and shape, an O-mode system is optimum due to the large range of magnetic field along the plasma periphery and the wide range of possible plasma configurations achievable on ITER. A robust real-time calibration technique of the whole transmission line is required. It is likely that an accurate estimate of the position of the plasma will require the simultaneous use of signals from the profile reflectometer. For the divertor, profiles with peak densities in the range 1019–1022/m3 are to be measured with a target resolution of 3 mm. The large density range will necessitate the use of more than one system. Installing these reflectometers on ITER incurs additional difficulties such as the routing of the millimetre wave radiation around the complicated first wall and divertor structures and design of antennas able to operate through the first wall and blanket.

The Engineering Design Activities (EDA) for an International Thermonuclear Experimental Reactor (Iter) were established on 21 July 1992 by an agreement between the European Atomic Energy Community, the Government of Japan, the Government of the Russian Federation and the Government of the USA. The overall programmatic objective is to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes. The engineering objectives are to establish the design of Iter in such a way that controlled ignition and extended burn of a fusion plasma can be demonstrated, essential reactor technologies can be demonstrated in an integrated system, and integrated tests of high heat flux and nuclear components can be performed. This article outlines the fusion process and the tokamak concept used for Iter and summarises the potential advantages of fusion energy. The objectives of the EDA are presented together with the organisational structure to translate them into a design. Finally the Iter parameters and design status are reported; and a brief overview of time planning and cost is given.

The design being developed for the International Thermonuclear Experimental Reactor (ITER) in the Engineering Design Activities (EDA) phase has to assure the achievement of controlled ignition and extended burn, and must provide a demonstration of technologies that are essential to a reactor. These requirements are being addressed by the application of present physical understanding of tokamaks, and by emphasis on simplicity in design and in maintenance concepts. The approach places severe demands on materials used in the design, and the concept developed so far requires earlier qualification of materials to more stringent specifications than was expected previously in the fusion program.

The ITER design, jointly conducted by Euratom, Japan, USSR and USA under the auspices of the IAEA, is progressing with an optimized tokamak machine, having a major radius of 6.0 m, magnetic field on axis of 4.85 T, and a nominal plasma current of 22 MA. This paper describes major parameters of the machine as well as the design of the superconducting magnets, first wall, divertor and blanket/shield which are particularly relevant to fusion reactor materials research and development.