The effects of poloidal asymmetries and heated minority species are shown to be necessary to accurately describe heavy impurity transport in present experiments in JET and ASDEX Upgrade. Plasma ...rotation, or any small background electrostatic field in the plasma, such as that generated by anisotropic external heating can generate strong poloidal density variation of heavy impurities. These asymmetries have recently been added to numerical tools describing both neoclassical and turbulent transport and can increase neoclassical tungsten transport by an order of magnitude. Modelling predictions of the steady-state two-dimensional tungsten impurity distribution are compared with tomography from soft x-ray diagnostics. The modelling identifies neoclassical transport enhanced by poloidal asymmetries as the dominant mechanism responsible for tungsten accumulation in the central core of the plasma. Depending on the bulk plasma profiles, turbulent diffusion and neoclassical temperature screening can prevent accumulation. Externally heated minority species can significantly enhance temperature screening in ICRH plasmas.
The behaviour of tungsten in the core of hybrid scenario plasmas in JET with the ITER-like wall is analysed and modelled with a combination of neoclassical and gyrokinetic codes. In these discharges, ...good confinement conditions can be maintained only for the first 2-3 s of the high power phase. Later W accumulation is regularly observed, often accompanied by the onset of magneto-hydrodynamical activity, in particular neoclassical tearing modes (NTMs), both of which have detrimental effects on the global energy confinement. The dynamics of the accumulation process is examined, taking into consideration the concurrent evolution of the background plasma profiles, and the possible onset of NTMs. Two time slices of a representative discharge, before and during the accumulation process, are analysed with two independent methods, in order to reconstruct the W density distribution over the poloidal cross-section. The same time slices are modelled, computing both neoclassical and turbulent transport components and consistently including the impact of centrifugal effects, which can be significant in these plasmas, and strongly enhance W neoclassical transport. The modelling closely reproduces the observations and identifies inward neoclassical convection due to the density peaking of the bulk plasma in the central region as the main cause of the accumulation. The change in W neoclassical convection is directly produced by the transient behaviour of the main plasma density profile, which is hollow in the central region in the initial part of the high power phase of the discharge, but which develops a significant density peaking very close to the magnetic axis in the later phase. The analysis of a large set of discharges provides clear indications that this effect is generic in this scenario. The unfavourable impact of the onset of NTMs on the W behaviour, observed in several discharges, is suggested to be a consequence of a detrimental combination of the effects of neoclassical transport and of the appearance of an island.
Disruption-generated runaway electron (RE) beams represent a severe threat for tokamak plasma-facing components in high current devices like ITER, thus motivating the search of mitigation techniques. ...The application of 3D fields might aid this purpose and recently was investigated also in the ASDEX Upgrade experiment by using the internal active saddle coils (termed B-coils). Resonant magnetic perturbations (RMPs) with dominant toroidal mode number n = 1 have been applied by the B-coils, in a RE specific scenario, before and during disruptions, which are deliberately created via massive gas injection. The application of RMPs affects the electron temperature profile and seemingly changes the dynamics of the disruption; this results in a significantly reduced current and lifetime of the generated RE beam. A similar effect is observed also in the hard-x-ray (HXR) spectrum, associated to RE emission, characterized by a partial decrease of the energy content below 1 MeV when RMPs are applied. The strength of the observed effects strongly depends on the upper-to-lower B-coil phasing, i.e. on the poloidal spectrum of the applied RMPs, which has been reconstructed including the plasma response by the code MARS-F. A crude vacuum approximation fails in the interpretation of the experimental findings: despite the relatively low β ( < 0.5 % ) of these discharges, a modest amplification (factor of 2 ) of the edge kink response occurs, which has to be considered to explain the observed suppression effects.
The solution of the problem of heat exhaust has been pointed out as one of the main challenge towards the realization of magnetic confinement fusion. In the last years, two concepts have been ...proposed in alternative to the conventional divertor solution adopted for ITER: modification of the magnetic topology in the divertor region and liquid metal as plasma facing component. The role of the Divertor Tokamak Test facility (DTT) in the power exhaust implementation strategy is discussed. The evolution of the project, since the original proposal in 2015 to the present design, is shown. The DTT facility is well integrated in the European strategy and the final decision on the divertor configuration will be made, within 2022-23, on the basis of the indication of the Power Exhaust Group constituted by the EUROfusion Consortium. Finally, the main milestones and the timeline of the project are illustrated.
The ITER full size plasma source device design Sonato, P.; Agostinetti, P.; Anaclerio, G. ...
Fusion engineering and design,
06/2009, Letnik:
84, Številka:
2
Journal Article, Conference Proceeding
Recenzirano
In the framework of the strategy for the development and the procurement of the NB systems for ITER, it has been decided to build in Padova a test facility, including two experimental devices: a full ...size plasma source with low voltage extraction and a full size NB injector at full beam power (1
MV). These two different devices will separately address the main scientific and technological issues of the 17
MW NB injector for ITER. In particular the full size plasma source of negative ions will address the ITER performance requirements in terms of current density and uniformity, limitation of the electron/ion ratio and stationary operation at full current with high reliability and constant performances for the whole operating time up to 1
h. The required negative ion current density to be extracted from the plasma source ranges from 290
A/m
2 in D
2 (D
−) and 350
A/m
2 in H
2 (H
−) and these values should be obtained at the lowest admissible neutral pressure in the plasma source volume, nominally at 0.3
Pa. The electron to ion ratio should be limited to less than 1 and the admissible ion inhomogeneity extracted from the grids should be better than 10% on the whole plasma cross-section having a surface exposed to the extraction grid of the order of 1
m
2.
The main design choices will be presented in the paper as well as an overview of the design of the main components and systems.
Runaway electron generation and control Esposito, B; Boncagni, L; Buratti, P ...
Plasma physics and controlled fusion,
01/2017, Letnik:
59, Številka:
1
Journal Article
Recenzirano
Odprti dostop
We present an overview of FTU experiments on runaway electron (RE) generation and control carried out through a comprehensive set of real-time (RT) diagnostics/control systems and newly installed RE ...diagnostics. An RE imaging spectrometer system detects visible and infrared synchrotron radiation. A Cherenkov probe measures RE escaping the plasma. A gamma camera provides hard x-ray radial profiles from RE bremsstrahlung interactions in the plasma. Experiments on the onset and suppression of RE show that the threshold electric field for RE generation is larger than that expected according to a purely collisional theory, but consistent with an increase due to synchrotron radiation losses. This might imply a lower density to be targeted with massive gas injection for RE suppression in ITER. Experiments on active control of disruption-generated RE have been performed through feedback on poloidal coils by implementing an RT boundary-reconstruction algorithm evaluated on magnetic moments. The results indicate that the slow plasma current ramp-down and the simultaneous reduction of the reference plasma external radius are beneficial in dissipating the RE beam energy and population, leading to reduced RE interactions with plasma facing components. RE active control is therefore suggested as a possible alternative or complementary technique to massive gas injection.
Ion cyclotron resonance frequency (ICRF) heating has been an essential component in the development of high power H-mode scenarios in the Jet European Torus ITER-like wall (JET-ILW). The ICRF ...performance was improved by enhancing the antenna-plasma coupling with dedicated main chamber gas injection, including the preliminary minimization of RF-induced plasma-wall interactions, while the RF heating scenarios where optimized for core impurity screening in terms of the ion cyclotron resonance position and the minority hydrogen concentration. The impact of ICRF heating on core impurity content in a variety of 2.5 MA JET-ILW H-mode plasmas will be presented, and the steps that were taken for optimizing ICRF heating in these experiments will be reviewed.
•Design review of the Italian Divertor Tokamak Test facility: engineering aspects.•Rationale for design choices and main differences from the original proposal.•Main goal of DTT: explore and qualify ...alternative power exhaust solutions for DEMO.•Edge conditions similar to DEMO in terms of dimensionless parameters.•Test of alternative divertor solutions in integrated scenarios.
This paper presents the engineering aspects of the design review of the Italian DTT (Divertor Tokamak Test facility), illustrating the rationale for the design choices and focusing on the main differences with respect to the original proposal.
Disruption-generated runaway electron (RE) beams represent a potentially severe threat for tokamak plasma-facing components. Application of properly designed 3D fields can act as a mitigation ...mechanism, as recently investigated in ASDEX Upgrade (AUG) and COMPASS experiments and in the tokamak discharges of RFX-mod. In all of these devices, the dynamics of the disruption are affected by the application of magnetic perturbations (MPs), and the resulting RE beam current and lifetime are significantly reduced. These experiments show, in particular, that the strength of the observed effects strongly depends on the poloidal spectrum of the applied MPs, which has been reconstructed including the plasma response. This paper reports the main findings on RE mitigation from the previously mentioned three devices, highlighting the common physics behind them and their interpretation by using the guiding center code ORBIT.
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