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...

A Thomson scattering system is being developed for Joint European Torus with 15 mm spatial resolution and a foreseen accuracy for temperature better than 15% at a density of 1019 m−3. This resolution is required at the internal transport barrier and edge pedestal and it can not be fully achieved with the present light detection and ranging systems. The laser for this system is Nd:YAG, 5 Joule, 20 Hz. Scattering volumes from R=2.9 m to R=3.9 m are imaged onto 1 mm diameter fibers, with F/25 collection aperture. Two fibers are used per scattering volume. Using optical delay lines, three scattering volumes are combined in each of the 21 filter polychromators. The signals are recorded with transient digitizers, which allow the combined time delayed signals to be resolved. Knowledge of the time delay between signals allows the use of correlation techniques in determining signal levels. The ac output of the amplifier is used, which tolerates a higher level of background signal without affecting dynamic range. The noise resulting from plasma light is determined directly.

Results are presented from a series of dedicated experiments carried out on JET in tritium, DT, deuterium and hydrogen plasmas to determine the dependence of the H mode power threshold on the plasma isotopic mass. The Pthr ∝ Aeff-1 scaling is established over the whole isotopic range. This result makes it possible for a fusion reactor with a 50:50 DT mixture to access the H mode regime with about 20% less power than that needed in a DD mixture. Results on the first systematic measurements of the power necessary for the transition of the plasma to the type I ELM regime, which occurs after the transition to H mode, are also in agreement with the Aeff-1 scaling. For a subset of discharges, measurements of Te and Ti at the top of the profile pedestal have been obtained, indicating a weak influence of the isotopic mass on the critical edge temperature thought to be necessary for the H mode transition.

The scaling of the energy confinement in H-mode plasmas with different hydrogenic isotopes (H, D, D-T and T) is investigated in JET. For ELM-free H-modes the thermal energy confinement time τ th is found to decrease weakly with the isotope mass (τ th ~ M-0.25 ± 0.22) whilst in ELMy H-modes the energy confinement time shows practically no mass dependence (τ th ~ M 0.03 ± 0.1). Detailed local transport analysis of the ELMy H-mode plasmas reveals that the confinement in the edge region increases strongly with the isotope mass whereas the confinement in the core region decreases with mass (τ thcore ∝ M-0.16) in approximate agreement with theoretical models of the gyro-Bohm type (τ gB ~ M-0.2).

A power-balance model, with radiation losses from impurities and neutrals, gives a unified description of the density limit (DL) of the stellarator, the L-mode tokamak, and the reversed field pinch (RFP). The model predicts a Sudo-like scaling for the stellarator, a Greenwald-like scaling, , for the RFP and the ohmic tokamak, a mixed scaling, , for the additionally heated L-mode tokamak. In a previous paper (Zanca et al 2017 Nucl. Fusion 57 056010) the model was compared with ohmic tokamak, RFP and stellarator experiments. Here, we address the issue of the DL dependence on heating power in the L-mode tokamak. Experimental data from high-density disrupted L-mode discharges performed at JET, as well as in other machines, are taken as a term of comparison. The model fits the observed maximum densities better than the pure Greenwald limit.