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A 24MW Electron Cyclotron (EC) system operating at 170 GHz and 3600 s pulse length is to be installed on ITER. The EC plant shall deliver 20MW of this power to the plasma for Heating and Current Drive (H&CD) applications. The EC system is designed for plasma initiation, central heating, current drive, current profile tailoring, and Magneto-hydrodynamic control (in particular, sawteeth and Neo-classical Tearing Mode) in the flat-top phase of the plasma. A preliminary design review was performed in 2012, which identified a need for extended application of the EC system to the plasma ramp-up, flattop, and ramp down phases of ITER plasma pulse. The various functionalities are prioritized based on those applications, which can be uniquely addressed with the EC system in contrast to other H&CD systems. An initial attempt has been developed at prioritizing the allocated H&CD applications for the three scenarios envisioned: ELMy H-mode (15 MA), Hybrid (~12 MA), and Advanced (~9 MA) scenarios. This leads to the finalization of the design requirements for the EC sub-systems. [ABSTRACT FROM AUTHOR]

Understanding the origin of rotation in ion cyclotron resonance frequency (ICRF) heated plasmas is important for predictions for burning plasmas sustained by alpha particles, being characterized by a large population of fast ions and no external momentum input. The angular velocity of the plasma column has been measured in JET H-mode plasmas with pure ICRF heating both for the standard low toroidal magnetic ripple configuration, of about ∼0.08% and, for increased ripple values up to 1.5% (Nave et al 2010 Phys. Rev. Lett. 105 105005). These new JET rotation data were compared with the multi-machine scaling of Rice et al (2007 Nucl. Fusion 47 1618) for the Alfv'en-Mach number and with the scaling for the velocity change from L-mode into H-mode. The JET data do not fit well any of these scalings that were derived for plasmas that are co-rotating with respect to the plasma current. With the standard low ripple configuration, JET plasmas with large ICRF heating power and normalized beta, βN 1.3, have a very small co-current rotation, with Alfv'en-Mach numbers significantly below those given by the rotation scaling of Rice et al (2007 Nucl. Fusion 47 1618). In some cases the plasmas are actually counter-rotating. No significant difference between the H-mode and L-mode rotation is observed. Typically the H-mode velocities near the edge are lower than those in L-modes. With ripple values larger than the standard JET value, between 1% and 1.5%, H-mode plasmas were obtained where both the edge and the core counter-rotated. [ABSTRACT FROM AUTHOR]

Electron cyclotron (EC) heating and current drive (HCD), at 170 GHz, 20MW, is one of the heating systems foreseen to assist and sustain the development of various ITER scenarios since the early phase of ITER. It is usually assumed that EC is efficient only at magnetic fields operating around full field (5.3 T) and half field (2.65 T), and most of the analyses presented so far have focused on the ECRH&CD performances at flat-top. Here, the EC capabilities are investigated for different plasma parameters corresponding to different phases of the ITER plasma discharge, from current ramp-up up to burn, and for a wide range of magnetic fields, focusing in particular on the EC potential for heating and for L- to H-mode assist. It is found that the EC system can contribute to a wide range of heating scenarios during the ramp-up of the magnetic field, significantly increasing the applicable range as a function of magnetic field as compared with traditional views. [ABSTRACT FROM AUTHOR]

Abstract: Experiments at JET with Be as plasma-facing material demonstrate that although Be PFCs are easily damaged, the consequences for plasma operation are mild. L-mode plasmas up to a current of 7MA and heating power ∼30MW were obtained on heavily damaged Be limiter with ∼12MJ plasma energy, albeit with a very high Be content. The behaviour of H-mode plasmas in contact with a molten Be divertor target was studied. Despite the heavy damage to the target (∼3mm molten grooves) no plasma performance deterioration was observed for medium density ELMy H-modes. Type I ELMs, in these conditions, lead to ∼40μm melt layer formation but to an average melt layer loss of only ∼4%. Based on these results, conclusions for ITER limiter and divertor operation are extracted. [Copyright &y& Elsevier]

This paper presents the experimental characterization of pedestal parameters, edge localized mode (ELM) energy, and particle losses from the main plasma and the corresponding ELM energy fluxes on plasma facing components for a series of dedicated experiments in the Joint European Torus (JET). From these experiments, it is demonstrated that the simple hypothesis relating the peeling-ballooning linear instability to ELM energy losses is not valid. Contrary to previous observations at lower triangularities, small energy losses at low collisionality have been obtained in regimes at high plasma triangularity and q95∼4.5, indicating that the edge plasma magnetohydrodynamic stability is linked with the transport mechanisms that lead to the loss of energy by conduction during type I ELMs. Measurements of the ELM energy fluxes on the divertor target show that their time scale is linked to the ion transport along the field and the formation of a high energy sheath, in agreement with kinetic modeling of ELMs. Higher density ELMs, of a convective nature, lead to overall much longer time scales for the ELM energy flux, with more than 80% of the ELM energy flux arriving after the surface divertor temperature has reached its maximum value. On the contrary, for low density ELMs, of a conductive nature, up to 40% of the energy flux arrives at the divertor target before the surface divertor temperature has reached its maximum value. These large and more conductive ELMs may lead to up to ∼50% of the ELM energy reaching the main wall plasma facing components instead of the divertor target. The extrapolation to the International Thermonuclear Experimental Reactor of the obtained results is described and the main uncertainties discussed. [ABSTRACT FROM AUTHOR]

Recent experiments at the Joint European Torus [Rebut et al., Fusion Eng. Des. 22, 7 (1993)] aim to improve confinement quality in high-confinement-mode (H-mode) plasmas at high densities. Energy confinement time as predicted by the International Thermonuclear Experimental Reactor ITER-H98(y,2) scaling at densities near or in excess of 85% of the Greenwald density limit scaling has been obtained by (i) strong plasma shaping (triangularity 0.35<δ<0.5), or (ii) impurity seeding, or (iii) high-field side pellet injection. Slow peaking of central density without confinement degradation is observed. Loss of sawteeth and core impurity accumulation is prevented by central ion cyclotron resonance heating. In high triangularity and impurity seeded plasmas, reduction of average power loss associated with type I edge localized modes (ELMs) is found which is attributed to the occurrence of additional losses in between ELMs. Broad band magnetic fluctuations are seen which are reminiscent of regimes with small ELMs in other tokamaks. Plasma configurations have been varied to find best combinations of edge pedestal parameters and small ELM losses. © 2002 American Institute of Physics. [ABSTRACT FROM AUTHOR]

Abstract Disruption mitigation by shattered Ne pellet injection (SPI) has been simulated for an H-mode ITER D-T discharge of 280 MJ stored energy using the 3D TOKES code. Heating of the first wall from the radiation flash during the thermal quench of the disruption initiated by the injected pellets has been assessed. The dynamics of Ne plasma expansion in the core with radiation from this plasma have been simulated for the series of pellet sizes from 5.4 × 1024 to 8.6 × 1022 Ne atoms. The threshold pellet size for irradiation of the core energy has been found. The results of SPI and massive gas injection simulations are compared. [ABSTRACT FROM AUTHOR]

The SOLPS-ITER package represents a renewed coupling of the up-to-date parallelized EIRENE Monte-Carlo code and the B2.5 edge plasma solver. Each part of the fluid-kinetic code system has been dismantled and analysed. For SOLPS-ITER, full backwards compatibility with previous SOLPS4.3 ITER reference simulations is critical. We present the results of the ongoing comprehensive benchmark activity and discuss how SOLPS-ITER can be tuned to reproduce ITER simulations performed with SOLPS4.3. In particular in SOLPS-ITER a quantitative “metric” to measure the level of code convergence has been implemented, with corresponding code diagnostics carried over from SOLPS4.3, allowing code run times to be accurately compared. Good agreement is found between the two codes in the tests so far. Remaining discrepancies in the solution in front of the divertor targets are likely related to different formulations of the equations and the different discretization schemes used by the codes. [ABSTRACT FROM AUTHOR]

Highlights: [•] Front steering mirror design. [•] Plasma facing blanket shield module/first wall panel design. [•] Fixed frequency torus CVD diamond window serving as first tritium barrier. [•] Prototypes and tests of the above key components in the Launcher Handling and Testing Facility. [Copyright &y& Elsevier]

Abstract: The structural components of the ITER ECH upper port plug are presented which integrates the mm-wave beam lines of the front steering launcher for the control of plasma instabilities. The design of the main structure considers two alternatives. Firstly, a double wall configuration for the port plug frame responds to homogeneous baking requirements at 240°C. Secondly, an alternative frame is presented with a single wall in sections that do not require active cooling. The blanket shield module (BSM) is designed as a double-wall housing with a first wall panel (FWP) and individual shield blocks respecting the open space for the mm-wave beams. The cut-out sections to the design of the FWP and the neighbouring lower regular blanket module are defined for the beams of the actual front steering model. The thermo-mechanical stresses of the FWP and the BSM housing caused by nuclear heating and plasma radiation were analysed for the more severe case of the earlier remote steering (“3/8 RS”) reference model. As a result of the calculations for stationary loads and for transient loads during a typical plasma burn of 400s, no critical stresses have to be expected in the BSM housing. [Copyright &y& Elsevier]

Abstract: The paper reviews recent theoretical and experimental results focussing on the identification of the key factors controlling ELM energy and particle losses both in natural ELMs and in the presence of external controlling mechanisms. Present experiment and theory pointed out the benefit of the high plasma shaping, high q95 and high pedestal density in reducing the ELM affected area and conductive energy losses in Type I ELMs. Small benign ELMs regimes in present machines (EDA, HRS, Type II, Grassy, QH, Type III in impurity seeded discharges at high δ ) and their relevance for ITER are reviewed. Recent studies of active control of ELMs using stochastic boundaries, small pellets and edge current generation are presented. [Copyright &y& Elsevier]

A new divertor (MKII-HP) has been designed to be implemented in JET as part of a possible enhancement programme of the JET facility (JET EP). The aim is to handle up to 40 MW of injected power for 10 s with plasma triangularities up to 0.5 while keeping enough flexibility for other scenarios. The divertor is shaped to optimise the wetting fraction without exposing sharp edges or metallic parts and the general design allows for high halo currents. [Copyright &y& Elsevier]