Note: Conference of the Japan Society of Plasma Science and Nuclear Fusion Research, Mito, Japan, November 2003 (invited paper) Reference CRPP-CONF-2003-038 Record created on 2008-05-13, modified on 2017-05-12

This paper presents the current status of the Electron Cyclotron Emission (ECE) diagnostic on the Reduced Technical Objectives/Reduced Cost International Thermonuclear Experimental Reactor (RTO/RC ITER). It discusses the implications of the new machine design on the measurement requirements, the ability of the diagnostic technique to meet these, and the changes in the implementation imposed by the new layout. Finally, it outlines the physics studies, design and R&D work required prior to the detailed design and construction of the diagnostic. Key results are: (i) that the localisation of the measurement is similar to that in ITER–FDR (40–100 mm in X-mode, 60–200 mm in O-mode for the reference scenario), so that the relative spatial resolution degrades in this, smaller, machine, and (ii) the expected effect of transport barriers on the temperature profile in the high temperature region will be poorly resolved, because the effect of the temperature gradient on the outboard side is to degrade the resolution to (∼250 mm in X-mode, ∼350 mm in O-mode). Nevertheless ECE will be able to make a unique and useful contribution to the RTO/RC ITER measurement set.

The JET experimental campaign has focused on studies in support of the ITER physics basis. An overview of the results obtained is given for the reference ELMy H mode and advanced scenarios, which in JET are based on internal transport barriers. JET studies for ELMy H mode have been instrumental in the definition of ITER FEAT. Positive elongation and current scaling in the ITER scaling law have been confirmed, but the observed density scaling fits a two term (core and edge) model better. Significant progress in neoclassical tearing mode limits has been made showing that ITER operation with q(95) around 3.3 seems to be optimized. Effective helium pumping and divertor enrichment is found to be well within ITER requirements. Target asymmetries and hydrogen isotope retention are well simulated by modelling codes taking into account drift flows in the scrape-off plasmas. Striking improvements in fuelling effectiveness have been made with the new high field pellet launch facility. Good progress has been made on scenarios for achieving good confinement at high densities, both with radiation improved modes and with high field side pellets. Significant development of advanced scenarios, in view of their application to ITER, has been achieved. Progress towards integrated advanced scenarios is well developed with edge pressure control (impurity radiation). An access domain has been explored showing, in particular, that the power threshold increases with magnetic field but can be significantly reduced when lower hybrid current drive is used to produce target plasmas with negative shear. The role of ion pressure peaking on MHD has been well documented. Lack of sufficient additional heating power and interaction with the septum at high beta prevents assessment of the beta limits (steady plasmas achieved with beta (N) up to 2.6). Plasmas with a non-inductive current (I(NI)/I(p) = 60%), well aligned with the plasma current, high beta and good confinement have also been obtained.

Correlation of density turbulence suppression and reduced plasma transport is observed in the internal transport barrier (ITB) region of JET tokamak discharges with optimized magnetic shear. The suppression occurs in two stages. First, low frequency turbulence and ion transport are reduced across the plasma core by a toroidal velocity shear generated by intense auxiliary heating. Then with the ITB formation, high frequency turbulence and electron transport are reduced locally within the steep pressure gradient region of the ITB.

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.

An integrated system of correlation reflectometers have been installed on the Joint European Torus tokamak. The diagnostic consists of three X-mode, heterodyne, dual-antenna (bistatic), correlation reflectometers. Two of the reflectometers are dual channel systems at fixed frequencies of 75 GHz (edge measurements) and 105 GHz (core measurements) with a selectable poloidal (25 mm) or toroidal (40 mm) antenna separation. The third reflectometer is a two frequency system for radial measurements with one channel fixed at 92 GHz and the other swept between 92 and 96 GHz. All three reflectometers operate simultaneously through the same antenna cluster. The diagnostic can provide unique information on the spatial and temporal characteristics of plasma fluctuations. The capabilities of the system are illustrated with a selection of results.

The unique roles played by ICRH in the preparation, formation and sustainment of internal transport barriers (ITBs) in high fusion performance JET optimized shear experiments using the Mark II poloidal divertor are discussed. Together with LHCD, low power ICRH is applied during the early ramp-up phase of the plasma current, `freezing in' a hollow or flat current density profile with q(0)>1. In combination with up to ∼20 MW of NBI, the ICRH power is stepped up to ∼6 MW during the main low confinement (L mode) heating phase. An ITB forms promptly after the power step, revealed by a region of reduced central energy transport and peaked profiles, with the ion thermal diffusivity falling to values close to the standard neoclassical level near the centre of both DD and DT plasmas. At the critical time of ITB formation, the plasma contains an energetic ICRF supported hydrogen minority ion population, contributing ∼50% to the total plasma pressure and heating mainly electrons. As both the NBI population and the thermal ion pressure develop, a substantial part of the ICRF power is damped resonantly on core ions (ω = 2 ωcD = 3ωcT), contributing to the ion heating. In NBI step-down experiments, high performance has been sustained by maintaining central ICRH; analysis shows the efficiency of central ICRH ion heating to be comparable to that of NBI. The highest DD fusion neutron rates (RNT = 5.6 × 1016 s-1) yet achieved in JET plasmas have been produced by combining a low magnetic shear core with a high confinement (H mode) edge. In DT, a fusion triple product niTiτE = (1.2 ± 0.2) × 1021 m-3 keV s was achieved with 7.2 MW of fusion power obtained in the L mode and with up to 8.2 MW of fusion power in the H mode phase.

A new type of magnetohydrodynamic (MHD) mode, provisionally termed the wash board (WB) mode, has been observed during H-mode plasmas in JET. It occurs in all types of H-mode discharges, but is not seen during L-mode even at high values of . The WB mode appears to be linked with saturation in the plasma confinement and central plasma temperatures. These modes have high m and n numbers and are localized in the outer part of the plasma, typically from the q = 2 surface to the plasma edge. They rotate with the electron diamagnetic frequency and have a strong ballooning character. There is a good correlation between increasing plasma pressure and the growth of both the spectral extent and amplitude of the WB modes. Changes in the electron temperature profile also correlate well with changes in the amplitude of these modes. They are, therefore, regarded as a possible candidate to play a role in the confinement degradation of H-mode plasmas. Although these modes appear in all types of H-modes in JET, most of the analysis has been carried out during the hot-ion H-mode period of the 100% T discharges, since these modes are practically the only MHD activity left and the small-scale density fluctuations can, therefore, be analysed.

Steady state high performance with improved core confinement and sustainable plasma edge conditions has been approached on JET in a double barrier (DB) mode. The DB mode combines an internal transport barrier of the optimized shear regime with an edge transport barrier of an ELMy H mode regime. Improved confinement with an H factor HITER-89 ≈ 2 has been maintained for four energy confinement times. Ion and electron temperature profiles remain peaked in the DB mode, while the density profile is broad and similar in shape to the conventional ELMy H mode profile. The energy confinement improves across the whole plasma cross-section, and the ion heat conductivity falls to the neoclassical level in the core. Particle transport studies show that impurity accumulation can be avoided in the DB mode. In DT discharges the DB mode has attained a fusion gain of Q ≈ 0.4, producing 6.8 MW of fusion power, compared with Q ≈ 0.2 in the conventional sawtoothing ELMy H mode. The ELMs are more benign, with an amplitude an order of magnitude smaller in the DB mode. The DB mode has a high potential to improve performance in reactor relevant conditions.

High fusion performance has been achieved in optimized shear discharges in JET with the early appearance of internal transport barriers (ITBs) in the high power phase in both DD and DT plasmas. An ITB has been produced in a large variety of conditions, including LHCD only, ICRH only, NBI only and NBI + ICRH with LHCD preheat. An ITB may coexist with the edge barrier. The high pressure gradient produced by ICRH combined with high power NBI heating starting in a low density plasma with q(0) ≈ 1.5-2 is the main prerequisite for an efficient ITB in high performance discharges. Strong coupling between q profiles, MHD activity, energy confinement and fusion yield has been observed. The transition from L to H mode can be triggered by MHD events. The transition is accompanied by significant momentary changes in the density fluctuation spectrum in the peripheral plasma. The ITB may persist for some time in both ELM-free and ELMy H mode.

Recent experiments in D-T plasmas on the JET and TFTR tokamaks have evaluated a wide range of ITER relevant ion cyclotron heating scenarios. Absorption of fast waves at the second-harmonic tritium resonance has provided bulk ion heating in TFTR supershots and electron heating in JET H-mode discharges. In JET, deuterium minority heating has generated 1.7 MW of fusion power with 6 MW of radio frequency power giving a record steady-state Q-value of 0.22. Strong bulk ion heating has been achieved with He-3 minority heating with central ion temperatures up to 13 keV being produced in H-modes with a density of 3.6 x 10(19) m(-3) Hydrogen, deuterium and He-3 minority heating methods have produced plasmas with normalized confinement times greater than or equal to that required by ITER for ignition. These H-modes are characterized by small-amplitude, high-frequency ELMs, each of which transports less than 1.5% of the plasma energy content to the limiters. The heavy minority scheme of tritium in a deuterium plasma has been demonstrated both as a heating scheme and a generator of suprathermal neutrons. On TFTR mode conversion to an ion Bernstein wave has achieved central bulk ion heating in supershots with target ion temperatures greater than 20 keV.

An analysis of the complete JET D-D threshold database, from 1990 to 1997 inclusive is presented. A JET scaling is derived far the threshold power through the separatrix, estimated by considering also radiation losses from the bulk plasma. Such scalings are applied to obtain a first qualitative estimate of the amount of power above threshold that is needed to obtain steady-state high confinement (H89 > 1.8). The L-H and H-L transitions are also analysed in terms of local edge parameters. The first results on isotope scaling of the H-mode power threshold in D-T and T-T plasmas are also presented, confirming the A(-1) scaling.

This paper summarises the study which has been made of the physics of ECE for ITER. Two ECE simulation codes have been used to obtain quantitative information about the spatial localisation and accessibility limits of T-e, measurements. The calculations include the effects of spatial averaging due to the antenna pattern. Results are presented for three different scenarios: the core of full performance plasmas, the edge region and the startup plasma. It is shown that for thermal plasmas it will generally be possible to make core T-e measurements with spatial resolution at or close to the ITER target (similar to 0.1 m) using the second harmonic extraordinary mode, although access to the plasma centre will be limited at high T-e. Access is better for the first harmonic ordinary mode, but the resolution is poor. Edge T-e measurements using ECE will have a spatial resolution >50 mm, much worse than the ITER target (5 mm), and it will generally not be possible to measure low values of T-e (<1 keV) close to the separatrix. It is shown that it will be possible to make core T-e measurements during much of the discharge startup if there is a suitable sightline.

The measurement of electron density in the JET pumped divertor by conventional reflectometry techniques is difficult because of poor access to the divertor region,unfavourable plasma geometry and the small plasma size. This paper presents a new reflectometer diagnostic which can operate in the difficult conditions of the pumped divertor. It is a `comb' reflectometer in which a number of microwave signals with different frequencies are launched simultaneously into the plasma, and the peak electron density along the sightline is estimated by determining which signals are transmitted and which are reflected. The design and operating principles of the diagnostic are described, and preliminary results illustrating its performance are presented.