S. Marsen

Fuel retention studies with the ITER-Like Wall in JET

JET underwent a transformation from a full carbon-dominated tokamak to a fully metallic device with beryllium in the main chamber and a tungsten divertor. This material combination is foreseen for the activated phase of ITER. The ITER-Like Wall (ILW) experiment at JET shall demonstrate the plasma compatibility with metallic walls and the reduction in fuel retention. We report on a set of experiments (Ipxa0=xa02.0xa0MA, Btxa0=xa02.0–2.4xa0T, δxa0=xa00.2–0.4) in different confinement and plasma conditions with global gas balance analysis demonstrating a strong reduction in the long-term retention rate by more than a factor of 10 with respect to carbon-wall reference discharges. All experiments are executed in a series of identical plasma discharges in order to achieve maximum plasma duration until the analysis limit of the active gas handling system is reached. The composition analysis shows high purity of the recovered gas, typically 99%xa0D. For typical L-mode discharges (Pauxxa0=xa00.5xa0MW), type III (Pauxxa0=xa05.0xa0MW) and type-I ELMy H-mode plasmas (Pauxxa0=xa012.0xa0MW) a drop of the deuterium retention rate normalized to the operational time in divertor configuration is measured from 1.27xa0×xa01021, 1.37xa0×xa01021 and 1.97xa0×xa01021xa0Dxa0s−1 down to 4.8xa0×xa01019, 7.2xa0×xa01019 and 16xa0×xa01019xa0Dxa0s−1, respectively. The dynamic retention increases in the limiter phase in comparison with carbon-fibre composite, but also the outgassing after the discharge has risen in the same manner and overcompensates this transient retention. Overall an upper limit of the long-term retention rate of 1.5xa0×xa01020xa0Dxa0s−1 is obtained with the ILW. The observed reduction by one order of magnitude confirms the expected predictions concerning the plasma-facing material change in ITER and is in line with identification of fuel co-deposition with Be as the main mechanism for the residual long-term retention. The reduction widens the operational space without active cleaning in the DT phase in comparison with a full carbon device.

Impact of nitrogen seeding on confinement and power load control of a high-triangularity JET ELMy H-mode plasma with a metal wall

This paper reports the impact on confinement and power load of the high-shape 2.5xa0MA ELMy H-mode scenario at JET of a change from all carbon plasma-facing components to an all metal wall. In preparation to this change, systematic studies of power load reduction and impact on confinement as a result of fuelling in combination with nitrogen seeding were carried out in JET-C and are compared with their counterpart in JET with a metallic wall. An unexpected and significant change is reported on the decrease in the pedestal confinement but is partially recovered with the injection of nitrogen.

Impact of carbon and tungsten as divertor materials on the scrape-off layer conditions in JET

The impact of carbon and beryllium/tungsten as plasma-facing components on plasma radiation, divertor power and particle fluxes, and plasma and neutral conditions in the divertors has been assessed in JET both experimentally and by edge fluid code simulations for plasmas in low-confinement mode. In high-recycling conditions the studies show a 30% reduction in total radiation in the scrape-off (SOL) layer when replacing carbon (JET-C) with beryllium in the main chamber and tungsten in the divertor (JET-ILW). Correspondingly, at the low-field side (LFS) divertor plate a two-fold increase in power conducted to the plate and a two-fold increase in electron temperature at the strike point were measured. In low-recycling conditions the SOL was found to be nearly identical for both materials configurations. Saturation and rollover of the ion currents to both low- and high-field side (HFS) plates was measured to occur at 30% higher upstream densities and radiated power fraction in JET-ILW. Past saturation, it was possible to reduce the ion currents to the LFS targets by a factor of 2 and to continue operating in stable, detached conditions in JET-ILW; in JET-C the reduction was limited to 50%. These observations are in qualitative agreement with predictions from the fluid edge code package EDGE2D/EIRENE, for which a 30% reduction of the total radiated power is also yielded when switching from C to Be/W. For matching upstream parameters the magnitude of predicted radiation is, however, 50% to 100% lower than measured, independent of the materials configuration. Inclusion of deuterium molecules and molecular ions, and temperature and density dependent rates in EIRENE reproduced the experimentally observed rollover of the ion current to the LFS plate, via reducing the electron temperature at the plate.

First operation with the JET International Thermonuclear Experimental Reactor-like wall

To consolidate International Thermonuclear Experimental Reactor (ITER) design choices and prepare for its operation, Joint European Torus (JET) has implemented ITERs plasma facing materials, namely, Be for the main wall and W in the divertor. In addition, protection systems, diagnostics, and the vertical stability control were upgraded and the heating capability of the neutral beams was increased to over 30u2009MW. First results confirm the expected benefits and the limitations of all metal plasma facing components (PFCs) but also yield understanding of operational issues directly relating to ITER. H-retention is lower by at least a factor of 10 in all operational scenarios compared to that with C PFCs. The lower C content (≈ factor 10) has led to much lower radiation during the plasma burn-through phase eliminating breakdown failures. Similarly, the intrinsic radiation observed during disruptions is very low, leading to high power loads and to a slow current quench. Massive gas injection using a D2/Ar mixtu...

Recent improvements of the JET lithium beam diagnostic

A 60 kV neutral lithium diagnostic beam probes the edge plasma of JET for the measurement of electron density profiles. This paper describes recent enhancements of the diagnostic setup, new procedures for calibration and protection measures for the lithium ion gun during massive gas puffs for disruption mitigation. New light splitting optics allow in parallel beam emission measurements with a new double entrance slit CCD spectrometer (spectrally resolved) and a new interference filter avalanche photodiode camera (fast density and fluctuation studies).

A near infra-red video system as a protective diagnostic for electron cyclotron resonance heating operation in the Wendelstein 7-X stellarator

The Wendelstein 7-X stellarator is a large nuclear fusion device based at Max-Planck-Institut für Plasmaphysik in Greifswald in Germany. The main plasma heating system for steady state operation in W7-X is electron cyclotron resonance heating (ECRH). During operation, part of plama facing components will be directly heated by the non-absorbed power of 1 MW rf beams of ECRH. In order to avoid damages of such components made of graphite tiles during the first operational phase, a near infra-red video system has been developed as a protective diagnostic for safe and secure ECRH operation. Both the mechanical design housing the camera and the optical system are very flexible and respect the requirements of steady state operation. The full system including data acquisition and control system has been successfully tested in the vacuum vessel, including on-line visualization and data storage of the four cameras equipping the ECRH equatorial launchers of W7-X.