• Energy Research

Future tokamak reactors of conventional design will require high levels of exhaust power dissipation (more than 90% of the input power) if power densities at the divertor targets are to remain compatible with active cooling. Impurity seeded H-mode discharges in JET-ITER-like Wall (ILW) have reached a max- imum radiative fraction ( F rad ) of ∼75%. Divertor Langmuir probe (LP) measurements in these discharges indicate, however, that less than ∼3% of the thermal plasma power reaches the targets, suggesting a missing channel for power loss. This paper presents experimental evidence from limiter LP for enhanced cross-field particle fluxes on the main chamber walls at high F rad . In H-mode nitrogen-seeded discharges with F rad increasing from ∼30% to up to ∼75%, the main chamber wall particle fluence rises by a factor ∼3 while the divertor plasma fluence drops by one order of magnitude. Contribution of main chamber wall particle losses to detachment, as suggested by EDGE2D-EIRENE modeling, is not sufficient to explain the magnitude of the observed divertor fluence reduction. An intermediate detached case obtained at F rad ∼60% with neon seeding is also presented. Heat loads were measured using the main chamber wall thermocouples. Comparison between thermocouple and bolometry measurements shows that the frac- tion of the input power transported to the main chamber wall remains below ∼5%, whatever the divertor detachment state is. Main chamber sputtering of beryllium by deuterium is reduced in detached condi- tions only on the low field side. If the fraction of power exhaust dissipated to the main chamber wall by cross-field transport in future reactors is similar to the JET-ILW levels, wall plasma power loading should not be an issue. However, other contributions such as charge exchange may be a problem. EURATOM 633053

In order to estimate the fuel loss in ITER and further future fusion devices, the deuterium permeation through different wall and structural materials are studied. In order to determine the effective permeability, gas-driven deuterium permeation measurements are performed on Cu and ITER grade CuCrZr. The obtained permeabilites for Cu and ITER grade CuCrZr are very similar and in agreement to literature values for Cu. For a better estimation for fusion reactor components, combined material samples are studied. Cu layers were applied on steel substrates by magnetron sputter deposition. With these studies, the influence of interfaces and microstructure on the hydrogen permeation is investigated. Our study reveals that in the case of Cu layered steel substrates the influence of the interface on the permeation flux is minor compared to the influence of the microstructure on the permeability. The Cu layer permeability is around one order of magnitude smaller than the Cu bulk permeability in the temperature range between 300∘C and 550∘C.

Tritium (T) distributions on tungsten (W)-coated plasma-facing tiles used in the third ITER-like wall campaign (2015–2016) of the Joint European Torus (JET) were examined by means of an imaging plate technique and β-ray induced x-ray spectrometry, and they were compared with the distributions after the second (2013–2014) campaign. Strong enrichment of T in beryllium (Be) deposition layers was observed after the second campaign. In contrast, T distributions after the third campaign was more uniform though Be deposition layers were visually recognized. The one of the possible explanations is enhanced desorption of T from Be deposition layers due to higher tile temperatures caused by higher energy input in the third campaign.

AbstractTo investigate the synergistic effects of fuel retention in tungsten, experiments were performed in the linear plasma device PSI-2 where the transient heat loads were applied by a high energy laser during the steady-state plasma operation. The impact on the deuterium (D) retention of simultaneous and sequential exposures to laser and plasma has been investigated. A significant increase of D retention, more than a factor of 12, has been observed during the simultaneous transient heat loads and plasma exposure. Successive exposures to transient heat loads and deuterium plasma also show the increased content of D-atoms by more than a factor of 3.6 in comparison to the pure plasma loading. In both cases the increase is most likely due to enhanced hydrogen clustering by the thermal shock exposures, due to an increased mobility of D atoms along the shock-induced cracks as well as to increased diffusion of D atoms into the W material caused by strong temperature gradients during the laser pulse exposure.Comparison of the NRA and TDS measurements shows that only 34% of the retained deuterium in the tungsten specimen is located inside the near-surface layer (d<4µm) after simultaneous as well as sequential exposures of W to heat load.Enhanced blister formation has been observed under combined loading conditions at power densities close to the threshold for damaging. It is shown that blisters are not mainly responsible for the pronounced increase of the D retention.

Tungsten migration in the open divertor region in the Large Helical Device is analyzed for validating the three-dimensional plasma-wall interaction simulation code ERO2.0. The ERO2.0 simulation reproduced the measurement of localized tungsten migration from a tungsten-coated divertor plate installed in the inboard side of the torus. The simulation also explained the measurement of the high tungsten areal density in the private side on a carbon divertor plate, next to the tungsten-coated divertor plate, by the tungsten prompt redeposition in plasma discharges for a low magnetic field strength in a counterclockwise toroidal direction. However, the simulation disagreed with the measurement of low tungsten areal density on the plasma-wetted areas on the carbon divertor plates, which indicated that the actual erosion rate of the redeposited tungsten should be much higher than that used in the ERO2.0 code.

Physical and chemical assisted physical sputtering were characterised by the Be I and Be II line and BeD band emission in the observation chord measuring the sightline integrated emission in front of the inner beryllium limiter at the torus midplane. The 3D local transport and plasma-surface interaction Monte- Carlo modelling (ERO code [18] ) is a key for the interpretation of the observations in the vicinity of the shaped solid Be limiter. The plasma parameter variation (density scan) in limiter regime has provided a useful material for the simulation benchmark. The improved background plasma parameters input, the new analytical expression for particle tracking in the sheath region and implementation of the BeD re- lease into ERO has helped to clarify some deviations between modelling and experiments encountered in the previous studies [4,5] . Reproducing the observations provides additional confidence in our ‘ERO-min’ fit for the physical sputtering yields for the plasma-wetted areas based on simulated data. EURATOM 633053