In order to support the operation of ITER and the planned experimental programme an extensive set of plasma and first wall measurements will be required. The number and type of required measurements will be similar to those made on the present-day large tokamaks while the specification of the measurements-time and spatial resolutions, etc-will in some cases be more stringent. Many of the measurements will be used in the real time control of the plasma driving a requirement for very high reliability in the systems (diagnostics) that provide the measurements.The implementation of diagnostic systems on ITER is a substantial challenge. Because of the harsh environment (high levels of neutron and gamma fluxes, neutron heating, particle bombardment) diagnostic system selection and design has to cope with a range of phenomena not previously encountered in diagnostic design. Extensive design and R&D is needed to prepare the systems. In some cases the environmental difficulties are so severe that new diagnostic techniques are required.The starting point in the development of diagnostics for ITER is to define the measurement requirements and develop their justification. It is necessary to include all the plasma parameters needed to support the basic and advanced operation (including active control) of the device, machine protection and also those needed to support the physics programme. Once the requirements are defined, the appropriate (combination of) diagnostic techniques can be selected and their implementation onto the tokamak can be developed. The selected list of diagnostics is an important guideline for identifying dedicated research and development needs in the area of ITER diagnostics.This paper gives a comprehensive overview of recent progress in the field of ITER diagnostics with emphasis on the implementation issues. After a discussion of the measurement requirements for plasma parameters in ITER and their justifications, recent progress in the field of diagnostics to measure a selected set of plasma parameters is presented. The integration of the various diagnostic systems onto the ITER tokamak is described. Generic research and development in the field of irradiation effects on materials and environmental effects on first mirrors are briefly presented. The paper ends with an assessment of the measurement capability for ITER and a forward of what will be gained from operation of the various diagnostic systems on ITER in preparation for the machines that will follow ITER. Performance assessment relative to requirements Design meets requirements S339 A.J.H. Donné et alPhysics Basis [7] and remains essentially the same. However, for ITER, the specific limits have changed. 2.1.2.Measurements needed for plasma control and evaluation. The measurements needed for plasma control and evaluation are naturally directly linked to the experimental programme, and particularly to the operating phase (i.e. H, D or D/T) and the operating scenario (H-mode, hybrid, etc). Since there is expected to be a phased introduction of po...

After many years of fusion research, the conditions needed for a D–T fusion reactor have been approached on the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. For the first time the unique phenomena present in a D–T plasma are now being studied in a laboratory plasma. The first magnetic fusion experiments to study plasmas using nearly equal concentrations of deuterium and tritium have been carried out on TFTR. At present the maximum fusion power of 10.7 MW, using 39.5 MW of neutral-beam heating, in a supershot discharge and 6.7 MW in a high-βp discharge following a current rampdown. The fusion power density in a core of the plasma is ≊2.8 MW m−3, exceeding that expected in the International Thermonuclear Experimental Reactor (ITER) [Plasma Physics and Controlled Nuclear Fusion Research (International Atomic Energy Agency, Vienna, 1991), Vol. 3, p. 239] at 1500 MW total fusion power. The energy confinement time, τE, is observed to increase in D–T, relative to D plasmas, by 20% and the ni(0) Ti(0) τE product by 55%. The improvement in thermal confinement is caused primarily by a decrease in ion heat conductivity in both supershot and limiter-H-mode discharges. Extensive lithium pellet injection increased the confinement time to 0.27 s and enabled higher current operation in both supershot and high-βp discharges. Ion cyclotron range of frequencies (ICRF) heating of a D–T plasma, using the second harmonic of tritium, has been demonstrated. First measurements of the confined alpha particles have been performed and found to be in good agreement with TRANSP [Nucl. Fusion 34, 1247 (1994)] simulations. Initial measurements of the alpha ash profile have been compared with simulations using particle transport coefficients from He gas puffing experiments. The loss of alpha particles to a detector at the bottom of the vessel is well described by the first-orbit loss mechanism. No loss due to alpha-particle-driven instabilities has yet been observed. D–T experiments on TFTR will continue to explore the assumptions of the ITER design and to examine some of the physics issues associated with an advanced tokamak reactor.

Alpha particle loss was measured during the TFTR DT experiments with a scintillator detector located at the vessel bottom in the ion Del B drift direction. The DT alpha particle loss to this detector was consistent with the calculated first orbit loss over the whole range of plasma current I=0.6-2.7 MA. In particular, the alpha particle loss rate per DT neutron at a given plasma current did not increase significantly with fusion power up to 10.7 MW, indicating the absence of any new 'collective' alpha particle loss processes in these experiments