B. Sieglin

Scaling of the tokamak near the scrape-off layer H-mode power width and implications for ITER

A multi-machine database for the H-mode scrape-off layer power fall-off length, λq in JET, DIII-D, ASDEX Upgrade, C-Mod, NSTX and MAST has been assembled under the auspices of the International Tokamak Physics Activity. Regression inside the database finds that the most important scaling parameter is the poloidal magnetic field (or equivalently the plasma current), with λq decreasing linearly with increasing Bpol. For the conventional aspect ratio tokamaks, the regression finds , yielding λq,ITER 1 mm for the baseline inductive H-mode burning plasma scenario at Ip = 15 MA. The experimental divertor target heat flux profile data, from which λq is derived, also yield a divertor power spreading factor (S) which, together with λq, allows an integral power decay length on the target to be estimated. There are no differences in the λq scaling obtained from all-metal or carbon dominated machines and the inclusion of spherical tokamaks has no significant influence on the regression parameters. Comparison of the measured λq with the values expected from a recently published heuristic drift based model shows satisfactory agreement for all tokamaks.

Advances in the physics basis for the European DEMO design

In the European fusion roadmap, ITER is followed by a demonstration fusion power reactor (DEMO), for which a conceptual design is under development. This paper reports the first results of a coherent effort to develop the relevant physics knowledge for that (DEMO Physics Basis), carried out by European experts. The program currently includes investigations in the areas of scenario modeling, transport, MHD, heating & current drive, fast particles, plasma wall interaction and disruptions.

Impact and mitigation of disruptions with the ITER-like wall in JET

Disruptions are a critical issue for ITER because of the high thermal and magnetic energies that are released on short timescales, which results in extreme forces and heat loads. The choice of material of the plasma-facing components (PFCs) can have significant impact on the loads that arise during a disruption. With the ITER-like wall (ILW) in JET made of beryllium in the main chamber and tungsten in the divertor, the main finding is a low fraction of radiation. This has dropped significantly with the ILW from 50?100% of the total energy being dissipated during disruptions in CFC wall plasmas, to less than 50% on average and down to just 10% for vertical displacement events (VDEs). All other changes in disruption properties and loads are consequences of this low radiation: long current quenches (CQs), high vessel forces caused by halo currents and toroidal current asymmetries as well as severe heat loads. Temperatures close to the melting limit have been locally observed on upper first wall structures during deliberate VDE and even at plasma currents as low as 1.5 MA and thermal energy of about 1.5?MJ only. A high radiation fraction can be regained by massive injection of a mixture of 10% Ar with 90% D2. This accelerates the CQ thus reducing the halo current and sideways impulse. The temperature of PFCs stays below 400??C. MGI is now a mandatory tool to mitigate disruptions in closed-loop operation for currents at and above 2.5?MA in JET.

High-accuracy characterization of the edge radial electric field at ASDEX Upgrade

The installation of a new poloidal charge exchange recombination spectroscopy (CXRS) diagnostic at ASDEX Upgrade (AUG) has enabled the determination of the radial electric field, Er, using the radial force balance of impurity ions. Er has been derived from charge exchange (CX) spectra measured on different impurity species, such as He2+, B5+, C6+ and Ne10+. The resulting Er profiles are found to be identical within the uncertainties regardless of the impurity species used, thus, demonstrating the validity of the diagnostic technique. The Er profile has been compared to the main ion pressure gradient term, which is found to be the dominant contribution at the plasma edge, thus, supporting that the Er well is created by the main ion species. The Er profile has been measured in different confinement regimes including L-, I- and H-mode. The depth of the Er well and the magnitude of the Er shear are correlated with the ion pressure at the pedestal top. The temporal evolution of the measured CX profiles and the resulting Er have been studied during an edge-localized mode (ELM) cycle. At the ELM crash, the Er minimum is less deep resulting in a reduction of the E???B shear. Within 2?ms after the ELM crash, the edge kinetic profiles have nearly recovered and the Er well is observed to recover simultaneously. In high density type-I ELM mitigated H-mode plasmas, obtained via externally applied magnetic perturbations (MPs) with toroidal mode number n?=?2, no clear effect on Er due to the MPs has been observed.

Divertor studies in nitrogen induced completely detached H-modes in full tungsten ASDEX Upgrade

The first stable completely detached H-mode plasma in the full tungsten ASDEX Upgrade has been achieved. Complete detachment of both targets is induced by nitrogen seeding into the divertor. Two new phases are added to the detachment classification described in Potzel et al (2014 Nucl. Fusion 54 013001): first, the line integrated density increases by about 15% with partial detachment of the outer divertor. Second, complete detachment of both targets is correlated to the appearance of intense, strongly localized, stable radiation at the X-point. Radiated power fractions, frad, increase from about 50% to 85% with nitrogen seeding. X-point radiation is accompanied by a loss of pedestal top plasma pressure of about 60%. However, the core pressure at ?pol?<?0.7 changes only by about 10%. H98?=?0.8?1.0 is observed during detached operation. With nitrogen seeding the edge-localized mode (ELM) frequency increases from the 100?Hz range to a broadband distribution at 1?2?kHz with a large reduction in ELM size.

First scenario development with the JET new ITER-like wall

In the recent JET experimental campaigns with the new ITER-like wall (JET-ILW), major progress has been achieved in the characterization and operation of the H-mode regime in metallic environments: (i) plasma breakdown has been achieved at the first attempt and X-point L-mode operation recovered in a few days of operation; (ii) stationary and stable type-I ELMy H-modes with beta(N) similar to 1.4 have been achieved in low and high triangularity ITER-like shape plasmas and are showing that their operational domain at H = 1 is significantly reduced with the JET-ILW mainly because of the need to inject a large amount of gas (above 10(22) Ds(-1)) to control core radiation; (iii) in contrast, the hybrid H-mode scenario has reached an H factor of 1.2-1.3 at beta(N) of 3 for 2-3 s; and, (iv) in comparison to carbon equivalent discharges, total radiation is similar but the edge radiation is lower and Z(eff) of the order of 1.3-1.4. Strong core radiation peaking is observed in H-mode discharges at a low gas fuelling rate (i. e. below 0.5 x 10(22) Ds(-1)) and low ELM frequency (typically less than 10 Hz), even when the tungsten influx from the diverter is constant. High-Z impurity transport from the plasma edge to the core appears to be the dominant factor to explain these observations. This paper reviews the major physics and operational achievements and challenges that an ITER-like wall configuration has to face to produce stable plasma scenarios with maximized performance.

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 30 MW. 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...

Power load studies in JET and ASDEX-Upgrade with full-W divertors

For the design and operation of large fusion devices, a detailed understanding of the power exhaust processes is necessary. This paper will give an overview of the current research on divertor power load mechanisms. The results shown are obtained in JET with the ITER-like wall (ILW)and ASDEX-Upgrade with tungsten coated plasma-facing components (PFCs). The challenges of infrared thermography on an ITER-like bulk tungsten divertor are presented. For the steady-state heat load, the power fall-off length ?q in JET-ILW is compared to an empirical scaling found in JET and the ASDEX-Upgrade with carbon PFCs. A first attempt to scale the divertor broadening S in the ASDEX-Upgrade with tungsten PFCs is shown. The edge localized mode (ELM) duration tELM in JET-C and JET-ILW is compared. For similar pedestal conditions (Te,ped and ne,ped), similar ELM durations are found in JET-C and JET-ILW. For higher ne,ped at the same pedestal pressure pe,ped, longer ELM durations are found in JET-ILW. The pedestal pressure pe,ped is found to be a good qualifier for the ELM energy fluency in both JET-C and JET-ILW. Improved diagnostic capabilities reveal ELM substructures on the divertor target occurring a few milliseconds before the ELM crash.

A protection system for the JET ITER-like wall based on imaging diagnostics

The new JET ITER-like wall (made of beryllium and tungsten) is more fragile than the former carbon fiber composite wall and requires active protection to prevent excessive heat loads on the plasma facing components (PFC). Analog CCD cameras operating in the near infrared wavelength are used to measure surface temperature of the PFCs. Region of interest (ROI) analysis is performed in real time and the maximum temperature measured in each ROI is sent to the vessel thermal map. The protection of the ITER-like wall system started in October 2011 and has already successfully led to a safe landing of the plasma when hot spots were observed on the Be main chamber PFCs. Divertor protection is more of a challenge due to dust deposits that often generate false hot spots. In this contribution we describe the camera, data capture and real time processing systems. We discuss the calibration strategy for the temperature measurements with cross validation with thermal IR cameras and bi-color pyrometers. Most importantly, we demonstrate that a protection system based on CCD cameras can work and show examples of hot spot detections that stop the plasma pulse. The limits of such a design and the associated constraints on the operations are also presented.

Mitigation of type-I ELMs with n?=?2 fields on JET with ITER-like wall

Recently, strong mitigation of Type-I Edge Localized Modes (ELMs) has been observed with application of the n = 2 field in high collisionality (nu^*_e=2.0) H-mode plasma on JET tokamak with ITER-like wall. In this experiment, the EFCC power supply system has been enhanced with a coil current up to 88kAt (twice than before). With an n = 2 field, the large type-I ELMs with frequency of ~ 45 Hz was replaced by the high frequency (few hundreds Hz) small ELMs. No density pump-out was observed during an application of the n = 2 field. The influence of the n = 2 field on the core and the pedestal electron pressure profiles is within the error bar and it can be neglected. During the normal type-I ELM H-mode phase, the maximal surface temperature (Tmax) on the outer divertor plate was overall increasing and associated with large periodical variation due to the type-I ELMs. However, during an application of the n = 2 field, Tmax was saturated and has only small variation in few degrees due to the small mitigated ELMs. Splitting of the outer strike point has been observed during the strong mitigation of the type-I ELMs.