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This article presents a new method for estimating the electron temperature of the Protosphera's screw pinch. The temperature radial profile is obtained by a self-consistent modeling of a 1D MHD equilibrium along with a 0D power balance of the plasma column, given measurements and estimates of the axial pinch plasma current, of the plasma rotational frequency and, at the equatorial plane, of the electron density radial profile, of the edge poloidal magnetic field, of the edge electron temperature and of the neutrals pressure in the vacuum vessel. The plasma is considered in equilibrium with its neutral phase and in constant rotation. A MATLAB code has been developed with the aim of estimating the MHD radial equilibrium profiles, the thermodynamic plasma state and the neutrals profile. The numerical estimates are compared with available experimental data showing a good agreement., Comment: 4 pages, 6 figures, 1 table, research presented to the "6th ICFDT"

The tokamak à configuration variable (TCV) continues to leverage its unique shaping capabilities, flexible heating systems and modern control system to address critical issues in preparation for ITER and a fusion power plant. For the 2019–20 campaign its configurational flexibility has been enhanced with the installation of removable divertor gas baffles, its diagnostic capabilities with an extensive set of upgrades and its heating systems with new dual frequency gyrotrons. The gas baffles reduce coupling between the divertor and the main chamber and allow for detailed investigations on the role of fuelling in general and, together with upgraded boundary diagnostics, test divertor and edge models in particular. The increased heating capabilities broaden the operational regime to include T e/T i ∼ 1 and have stimulated refocussing studies from L-mode to H-mode across a range of research topics. ITER baseline parameters were reached in type-I ELMy H-modes and alternative regimes with ‘small’ (or no) ELMs explored. Most prominently, negative triangularity was investigated in detail and confirmed as an attractive scenario with H-mode level core confinement but an L-mode edge. Emphasis was also placed on control, where an increased number of observers, actuators and control solutions became available and are now integrated into a generic control framework as will be needed in future devices. The quantity and quality of results of the 2019–20 TCV campaign are a testament to its successful integration within the European research effort alongside a vibrant domestic programme and international collaborations.

Since the 2018 IAEA FEC Conference, FTU operations have been devoted to several experiments covering a large range of topics, from the investigation of the behaviour of a liquid tin limiter to the runaway electrons mitigation and control and to the stabilization of tearing modes by electron cyclotron heating and by pellet injection. Other experiments have involved the spectroscopy of heavy metal ions, the electron density peaking in helium doped plasmas, the electron cyclotron assisted start-up and the electron temperature measurements in high temperature plasmas. The effectiveness of the laser induced breakdown spectroscopy system has been demonstrated and the new capabilities of the runaway electron imaging spectrometry system for in-flight runaways studies have been explored. Finally, a high resolution saddle coil array for MHD analysis and UV and SXR diamond detectors have been successfully tested on different plasma scenarios.

Pellet injection represents, to date, the most promising option for core fuelling of the EU-DEMO tokamak. Simulations with the HPI2 pellet ablation/deposition code indicate, however, that sufficiently deep fuel deposition requires injection from the High Field Side (HFS) at velocities ≳1 km/s. Two complementary inboard injection schemes are being explored: one makes use of guide tubes with curvature radii ≥6 m in the attempt of preserving pellet integrity at speeds of ˜1 km/s, the other is investigating the feasibility of injecting high-speed (˜3 km/s) pellets along “direct line of sight” (DLS) trajectories, from either the HFS or a vertical port. Options using quasi-vertical DLS paths routed across the upper vertical port have been explored first, as they can be more easily integrated, Unfortunately, the radial position of the available vertical access (≳9 m from the machine axis) turns out to be unfavorable; further simulations with the HPI2 code predict indeed that vertical injection may be effective only if pellets trajectories are well inboard the magnetic axis. High-speed injection through oblique inboard “DLS” paths, not interfering with the Central Solenoid (CS), are instead predicted to yield good performance, provided that the injection location is ≲2.5 m from the equatorial mid-plane. The angular spread of high-speed free-flight pellets, recently measured using an existing facility, turns out to be enclosed within ˜ 0.7°. This scatter cone may require significant cut off volume of the Breeding Blanket (BB). Moreover, DLS in-vessel conical penetrations may increase the neutron flux outside of the bio-shield, and also result in a significant heat load in the cryogenic pellet source. These issues are being investigated, to identify suitable shielding strategies; preliminary results are reported. The suitability of straight guide tubes to reduce the scatter cone, and hence the corresponding open cross section on BB penetration and the neutron streaming, will be explored as a further step. © 2019 Elsevier B.V.

Results from the last FTU campaigns on RE mitigation strategies for quiescent and post-disruption RE beams are presented. We provide experimental evidence that for some RE quiescent scenarios D-2 solid pellets achieve complete RE suppression capability, mainly due to the induced burst MHD activity expelling RE seed, whereas in other cases we report clear indications of avalanche multiplication of RE. Results on the assimilation of solid deuterium pellets on RE quiescent scenarios are provided. Quantitative indications of dissipative effects of anomalous Doppler instabilities (ADI) and MHD activity, in terms of critical electric field increase, is introduced and supported by experimental evidence. Multiple analysis are provided to show the significant energy conversion/dissipation of large ADI on post-disruption RE beams suggesting new strategies for RE energy suppression. We also demonstrate experimentally that modulated ECRH could be used for ADI pacing.

The diagnostic performances of photodetectors based on Chemical Vapor Deposition (CVD) single crystal diamonds that were installed on one of the equatorial ports of the FTU tokamak during the last six months of operation of the machine are reported. Examples of plasma fast events have been collected in several different plasma conditions, confirming the fast response capabilities of diamond detectors and the high S/N ratio. During the Runaway Control and Mitigation experiments, for example, the so-called Anomalous Doppler Instabilities were observed as sharp peaks of Ly-α emission caused by fast electrons hitting the wall, as a result of RE beam instabilities, perfectly correlated in time with other magnetic diagnostics and fast EC polychromator signals. Fast oscillations in the plasma emission are also observed in the presence of tearing modes, while other interesting observations relate to pellet ablation processes. The CVD diamond detectors were installed on FTU in view of their possible use for replacement of the Si photodiodes currently adopted for Soft X-Ray tomography (SXT) on future devices with harsher radiation environments. In fact, it was realized that the relatively flat response of the UV diamond sensor opened the possibility of using these detectors as bolometers. The first encouraging results have prompted launching an R&D program for the development of full-fledged diamond bolometers, which will be especially well suited for the coverage of the divertor and edge regions in high performance devices, in the energy range from 10 eV to about 20 keV.

Pellet injection represents to date the most realistic candidate technology for core fueling of a demonstration fusion power reactor tokamak fusion reactor. Modeling of both pellet penetration and fuel deposition profiles, for different injection locations, indicates that effective core fuelling can be achieved launching pellets from the inboard high field side at speeds not less than ~1 km/s. Inboard pellet fueling is commonly achieved in present tokamaks, using curved guide tubes; however, this technology might be hampered at velocities ≥1 km/s. An innovative approach, aimed at identifying suitable inboard “direct line” paths, to inject high-speed pellets (in the 3 to 4 km/s range), has recently been proposed as a potential complementary solution. The fuel deposition profiles achievable by this approach have been explored using the HPI2 simulation code. The results presented here show that there are possible geometrical schemes providing good fueling performance. The problem of neutron flux in a direct line-of-sight injection path is being investigated, though preliminary analyses indicate that, perhaps, this is not a serious problem. The identification and integration of straight injection paths suitably tilted may be a rather difficult task due to the many constraints and to interference with existing structures. The suitability of straight guide tubes to reduce the scatter cone of high-speed pellets is, therefore, of main interest. A preliminary investigation, aimed at addressing these technological issues, has recently been started. A possible implementation plan, using an existing Italian National Agency for New Technologies, Energy and Sustainable Economic Development-Oak Ridge National Laboratory facility is shortly outlined.

The injection of cryogenic pellets from the low field side (LFS) has long been in use for core fuelling of fusion devices, but injection from the high field side (HFS) has proved to provide a more effective core particle deposition, despite the severe limitations imposed to the pellet speed (≤300 m/s) by inboard accessibility. In the future, an alternative approach may be that of injecting high-speed pellets from the HFS, through suitable “free-flight” paths, eliminating curved transfer systems. Furthermore, the expected length of the plasma discharges will require steady-state repetitive systems. ORNL and ENEA have been collaborating on high-speed injectors since 1990; they successfully realized a high-speed repeating pellet injector (2.55 km/s at 1 Hz). Since then, good progress has been achieved on both fronts of steady-state extruders, and operation and reliability of two-stage guns. A comprehensive R&D program is therefore proposed to investigate how far speed limits and repetition rates of combined two-stage guns and steady-state extruders technologies can be extended. Simulations results are presented showing pellet penetration for several injection locations on a tokamak under construction such as JT60-SA, on the basis of one set of design plasma parameters.

The research program of the TCV tokamak ranges from conventional to advanced-tokamak scenarios and alternative divertor configurations, to exploratory plasmas driven by theoretical insight, exploiting the device's unique shaping capabilities. Disruption avoidance by real-time locked mode prevention or unlocking with electron-cyclotron resonance heating (ECRH) was thoroughly documented, using magnetic and radiation triggers. Runaway generation with high-Z noble-gas injection and runaway dissipation by subsequent Ne or Ar injection were studied for model validation. The new 1 MW neutral beam injector has expanded the parameter range, now encompassing ELMy H-modes in an ITER-like shape and nearly non-inductive H-mode discharges sustained by electron cyclotron and neutral beam current drive. In the H-mode, the pedestal pressure increases modestly with nitrogen seeding while fueling moves the density pedestal outwards, but the plasma stored energy is largely uncorrelated to either seeding or fueling. High fueling at high triangularity is key to accessing the attractive small edge-localized mode (type-II) regime. Turbulence is reduced in the core at negative triangularity, consistent with increased confinement and in accord with global gyrokinetic simulations. The geodesic acoustic mode, possibly coupled with avalanche events, has been linked with particle flow to the wall in diverted plasmas. Detachment, scrape-off layer transport, and turbulence were studied in L- and H-modes in both standard and alternative configurations (snowflake, super-X, and beyond). The detachment process is caused by power 'starvation' reducing the ionization source, with volume recombination playing only a minor role. Partial detachment in the H-mode is obtained with impurity seeding and has shown little dependence on flux expansion in standard single-null geometry. In the attached L-mode phase, increasing the outer connection length reduces the in-out heat-flow asymmetry. A doublet plasma, featuring an internal X-point, was achieved successfully, and a transport barrier was observed in the mantle just outside the internal separatrix. In the near future variable-configuration baffles and possibly divertor pumping will be introduced to investigate the effect of divertor closure on exhaust and performance, and 3.5 MW ECRH and 1 MW neutral beam injection heating will be added., This work was supported in part by the Swiss National Science Foundation.

Within the EUROfusion MST1 work package, a series of experiments has been conducted on AUG and TCV devices to disentangle the role of plasma fueling and plasma shape for the onset of small ELM regimes. On both devices, small ELM regimes with high confinement are achieved if and only if two conditions are fulfilled at the same time. Firstly, the plasma density at the separatrix must be large enough (), leading to a pressure profile flattening at the separatrix, which stabilizes type-I ELMs. Secondly, the magnetic configuration has to be close to a double null (DN), leading to a reduction of the magnetic shear in the extreme vicinity of the separatrix. As a consequence, its stabilizing effect on ballooning modes is weakened. Peer Reviewed Article escrit per 365 autors/autores: Labit, B.; Eich, T.; Harrer, G. F.; Wolfrum, E.; Bernert, M.; Dunne, M. G.; Frassinetti, L.; Hennequin, P.; Maurizio, R.; Merle, A.; Meyer, H.; Saarelma, S.; Sheikh, U.; Adamek, J.; Agostini, M.; Aguiam, D.; Akers, R.; Albanese, R.; Albert, C.; Alessi, E.; Ambrosino, R.; Andr be, Y.; Angioni, C.; Apruzzese, G.; Aradi, M.; Arnichand, H.; Auriemma, F.; Avdeeva, G.; Ayllon-Guerola, J. M.; Bagnato, F.; Bandaru, V. K.; Barnes, M.; Barrera-Orte, L.; Bettini, P.; Bilato, R.; Biletskyi, O.; Bilkova, P.; Bin, W.; Blanchard, P.; Blanken, T.; Bobkov, V.; Bock, A.; Boeyaert, D.; Bogar, K.; Bogar, O.; Bohm, P.; Bolzonella, T.; Bombarda, F.; Boncagni, L.; Bouquey, F.; Bowman, C.; Brezinsek, S.; Brida, D.; Brunetti, D.; Bucalossi, J.; Buchanan, J.; Buermans, J.; Bufferand, H.; Buller, S.; Buratti, P.; Burckhart, A.; Calabr, G.; Calacci, L.; Camenen, Y.; Cannas, B.; Cano Megías, P.; Carnevale, D.; Carpanese, F.; Carr, M.; Carralero, D.; Carraro, L.; Casolari, A.; Cathey, A.; Causa, F.; Cavedon, M.; Cecconello, M.; Ceccuzzi, S.; Cerovsky, J.; Chapman, S.; Chmielewski, P.; Choi, D.; Cianfarani, C.; Ciraolo, G.; Coda, S.; Coelho, R.; Colas, L.; Colette, D.; Cordaro, L.; Cordella, F.; Costea, S.; Coster, D.; Cruz Zabala, D. 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P.; Gorini, G.; Gospodarczyk, M.; Granucci, G.; Grekov, D. 1; Grenfell, G.; Griener, M.; Groth, M.; Grover, O.; Gruca, M.; Gude, A.; Guimarais, L.; Gyergyek, T.; Hacek, P.; Hakola, A.; Ham, C.; Happel, T.; Harrison, J.; Havranek, A.; Hawke, J.; Henderson, S.; Hesslow, L.; Hitzler, F.; Hnat, B.; Hobirk, J.; Hoelzl, M.; Hogeweij, D.; Hopf, C.; Hoppe, M.; Horacek, J.; Hron, M.; Huang, Z.; Iantchenko, A.; Iglesias, D.; Igochine, V.; Innocente, P.; Ionita-Schrittwieser, C.; Isliker, H.; Ivanova-Stanik, I.; Jacobsen, A.; Jakubowski, M.; Janky, F.; Jardin, A.; Jaulmes, F.; Jensen, T.; Jonsson, T.; Kallenbach, A.; Kappatou, A.; Karpushov, A.; Kasilov, S.; Kazakov, Y.; Kazantzidis, P. 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The implications of the adoption of a tokamak's toroidal field coil characterized by differential tilting in the azimuthal direction are investigated. From an engineering point of view, the major advantage introduced by such coils is a drastic reduction of some components of the electromagnetic forces in certain areas. As a beneficial side-effect, they generate poloidal field, in addition to toroidal field. The former advantage allows for a partial relaxation of the reinforcing structural material required in the machine design, while the poloidal flux generated during the current rise, when used in conjunction with the conventional central solenoid, would allow for discharges of longer duration. This paper presents results obtained by applying the tilting optimization procedure to circular and D-shaped coils, and characterized by geometrical and physical parameters proper of high-field compact tokamaks, since the issue of electromagnetic force reduction is most relevant in these devices.

The four-barrel, two-stage gun Ignitor Pellet Injector (IPI) was developed in collaboration between ENEA and ORNL. The prototype injector is presently located at Oak Ridge (TN, USA), and is normally operated locally through a control and data acquisition system developed in LabVIEW. More recently, a remote-control system has been set up, based on RealVNC®, which allows to operate the IPI from a control room in Italy. Tools for data transfer and storage into ENEA ICT area have also been provided. A Staging, Storage and Sharing system, named E3S, developed using OwnCloud as architectural component, is used for file synchronization and sharing of the data acquired by the diagnostic systems. It provides a homogeneous platform able to store and share heterogeneous data produced by many data acquisition systems in large nuclear fusion experiments. This paper reports about the implementation of the IPI remote control, and presents the application of E3S to this specific case, allowing easy storage and sharing of experimental data onto a wide-area distributed file-system, as well as remote data access via web-services based on MDS+ tool, integrated with MySQL metadata. A performance analysis of the architectural components is also introduced.

A capillary porous liquid tin limiter (TLL) was exposed as a plasma-facing component (PFC) in the Frascati Tokamak Upgrade. The TLL was progressively inserted deeper into the scrape-off-layer, very close to the last closed magnetic surface (

A promising field of application of the X-ray radiation generated by intense laser-matter interaction is the contact-microscopy of biological samples. In order to optimize the yield on the desired spectral window, it is fundamental to have an accurate characterization of this emitted radiation. To this purpose, an experimental campaign is underway with the three-nanosecond phosphate Nd:glass laser at the ABC laboratory in ENEA. The plasma was generated by irradiating solid targets with energy up to 30 J, and intensity on target up to 1014 Wcm-2. A transmission grating provided low-resolution spectra of a wide spectral range (2-50 Å). This covered also the so-called "water window" region, namely the spectral region between Oxygen and Carbon K-absorption edges (23 to 44 Å) where the absorption of carbon is ten times that of oxygen, and for this reason is of high interest for the contact-microscopy application. The X-ray yield in this spectral region was also monitored by a PIN diode filtered with 0.5 μm vanadium foil and coupled with a grazing-incidence copper mirror. A spherically-bent mica crystal was used to obtain high resolution spectra of the region going from 5.1 to 5.8 Å. This narrow spectral range was analyzed for determining the temperature of the produced plasma through the identification of different lines and of their intensity.

We will discuss the transmission spectra of plasmas generated by spectroscopic lamps filled with different gas compounds. The spectral response is measured with a standard THz-TDS spectrometer with 2 THz bandwidth. © 2018 IEEE.

Introduction FTU is conducting an extensive program aimed at controlling and mitigating purposely generated runaway electron (RE) beams in natural and induced disruptions [1, 2]. During the latest experimental campaign different mitigation techniques have been tested, which involved injection of Laser Blow Off (LBO) metal impurities and of multiple Deuterium pellets (1÷2×1020 atoms). A large number of discharges has been analysed correlating various plasma signals to REs behaviour. The injections have been performed both during the plasma current flat top, with seed REs embedded in a hot plasma, and also after current quench (CQ), with current mainly carried by REs, in order to extrapolate information for ITER predictions. The main RE measurements used in this work were the following diagnostic systems: BF3 chambers, only sensitive to neutrons; NEU213 organic liquid scintillator, sensitive both to neutrons and hard X-rays (HXR); Gamma camera to measure HXR produced by the REs through bremsstrahlung in the plasma; Cherenkov probes to detect REs escaping the plasma [2,3].

Core fuelling of DEMO tokamak fusion reactor is under investigation within the EUROfusion Work Package “Tritium, Fuelling and Vacuum”. An extensive analysis of fuelling requirements and technologies, suggests that pellet injection still represents, to date, the most realistic option. Modelling of both pellet penetration and fuel deposition profiles for different injection locations, assuming a specific plasma reference scenario and the ITER reference pellet mass (6 × 1021 atoms), indicates that: 1) Low Field Side (LFS) injection is inadequate; 2) Vertical injection may be effective only provided that pellets are injected at ∼ 10 km/s from a radial position ≤∼8 m; 3) effective core fuelling can be achieved launching pellets from the High Field Side (HFS) at ∼1 km/s. HFS injection was therefore selected as the reference scheme, though scenarios featuring less steep density and temperature gradients at the plasma edge could induce to reconsider vertical injection at speeds in the range of 4–5 km/s. To deliver intact pellets at 1 km/s from the HFS, the use of guide tubes with a bend radius ≥6 m is envisaged. The results of above simulations rely on the hypothesis that pellets are delivered at the plasma edge with the desired mass and speed. However, mass erosion and fracturing of pellets inside the guide tube (severely limiting the transfer speed), as well as pressure build up and speed losses at relevant injection rates, might hamper the use of curved guide tubes. An additional innovative approach, aimed at identifying inboard straight “free flight” injection paths, to inject pellets from the HFS at significantly higher speeds, is proposed and discussed as a backup solution. Outboard high-speed injection is still being considered, instead, for JT-60SA.

A safety analysis study has been applied to the Ignitor machine. Deterministic evaluations of radioactive environmental releases have been performed. The preliminary radiological impact analysis for the normal operation and the main accidental sequences of Ignitor, in case of its localization in the TRINITI site in Russia are presented. The site of TRINITI, hosting since decades nuclear installations, is well characterized, both from the meteorological and population aspects: many data have been collected over the years. The Ignitor machine, both during routine functioning and accidental sequences, presents a negligible radiological impact. Ignitor does not need any concrete primary containment, like those used in fission nuclear power plants. There is no need of people evacuation or emergency countermeasures even in presence of the worst accident.

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An Ignitor-like tokamak that is compact, high field, and high density device, could make full use of the its intense neutron flux, without reaching ignition as a source of neutrons for materials testing in support of a fission-fusion hybrid device. The main features of this High Field Neutron Source Facility, which would have about 50% more plasma volume than Ignitor, are illustrated and the R&D required in order to achieve relevant dpa quantities in test materials are discussed. Several full-power months of operation are sufficient to obtain relevant radiation damage values in terms of dpa, and a scoping study of shielding the magnetic insulators to reduce radiation damage has been performed

Nuclear fusion with polarized fuel looks especially attractive in the perspective of increasing the reaction rates or providing a preferential direction to the emission of the fusion products. Considering the difficulties encountered in reaching the necessary plasma conditions, however, the problem of producing and maintaining a polarized plasma may appear as an additional, and unwarranted, complication. Nevertheless, it can be observed that since most of the fusion reactions take place in the very central part of the plasma column in a tokamak, it may not be impossible to sustain long enough the fuel polarization in that region, where both particle and energy confinement are higher, and instabilities are lower. The possibility of utilizing standard auxiliary heating systems at the ion cyclotron range of frequency to interact with the nuclear magnetic moments of deuterium or tritium is considered, with some preliminary results from a radiofrequency propagation code showing the polarization states of two waves, resonating respectively with the D and T precession frequency.

The 0.1-30 THz frequency range routinely studied with time-domain spectroscopy (TDS), has an unquestionable diagnostic potential for tokamak plasmas, thus THz-TDS applications appear to be the natural evolution of FIR and submillimeter wave diagnostics. By illuminating the plasma with a THz beam, phase, intensity and polarization of reflected and transmitted wave can be simultaneously detected, potentially measuring several plasma parameters with a single diagnostic. © 2016 IEEE.

The foreseen future expansion of nuclear power would involve a solution to burn the long half-life transuranics (TRU) in the spent nuclear fuel discharged from LWRs. Moreover, high-energy neutrons can be very useful for many processes. The study addresses the development of a tokamak neutron source for a hybrid fission-fusion demonstration experimental device, using Ignitor-based technologies. Ignitor is a proposed compact high magnetic field tokamak, aimed at reaching ignition in DT plasmas. A revision of its operating parameters, in order to act a as a suitable neutron source in a hybrid device, is discussed, and a new operating scenario is proposed. Features of the tokamak neutron source are illustrated and preliminary results are discussed. A two phase project is described, in order to test an Ignitor-based neutron source and then implement an experimental device for demonstration of technical feasibility of TRU transmutation

Fusion creates more neutrons per energy released than fission or spallation, therefore DT fusion facilities have the potential to become the most intense sources of neutrons for material testing. An Ignitor-like device, that is a compact, high field, high density machine could be envisaged for this purpose making full use of the intense neutron flux that it can generate, without reaching ignition. The main features of this High Field Neutron Source Facility, which would have about 50% more volume than Ignitor, are illustrated and the R&D required in order to achieve relevant dpa quantities in the tested materials are discussed, in particular the adoption of superconducting magnet coils. Radiation damage evaluations have been performed by means of the ACAB code for some fusion-relevant materials, like pure iron, ASI316L, EUROFER, SiC/SiC, Mo, Graphite, V-15Cr-5Ti. Values ranging from 1.6 x10-26 to 2.4 x10-25 dpa per source neutron have been obtained. Some full-power months of operation are sufficient to obtain relevant radiation damage values in terms of dpa: the setup of a duty cycle for the device in order to obtain such operation times is the next required step to proceed with the evaluation

IGNITOR is the only experiment today designed with the ambitious goal of achieving ignition. During the discharge, the First-Wall Limiter (FWL) will receive about 20 MW [B. Coppi, A. Airoldi, F. Bombarda, et al., Nucl. Fusion 41 (200 1) 1253-1257] power, split between conduction/convection and radiation channels. Previous estimates suggested a peak heat flux on the FWL close to 1 MW/m(2). This value should be re-evaluated, accounting for the latest developments in the design of the IGNITOR first wall and changes in the operational scenarios, in order to assess the risk of damages to the FWL deriving from the combination of thermal and mechanical stresses. For this purpose, the development of the ASPOEL code, implementing a 2D model of the edge plasma and including a detailed representation of the FWL, was started at Politecnico di Torino. Here we present the main features of the new code, and illustrate its potential via a preliminary application to IGNITOR. (c) 2007 Elsevier B.V. All rights reserved.

The Frascati Tokamak Upgrade (FTU) is a high field and high density device which allows one to study confinement and transport at plasma parameters generally not accessible to other tokamaks and is relevant to next generation experiments (ITER). In this paper we study the effect of different density, temperature and q profiles on confinement and transport in various types of plasmas including ohmic, externally heated and pellet-fuelled discharges. Peaking of the electron density profile in pellet-fuelled discharges has been found to enhance the energy confinement time on FTU to values as high as 120 ms at densities of the order of those expected in ITER. Experiments designed to further improve confinement by injecting pellets at higher densities show the existence of a second threshold for saturation of confinement when the local electron heat conductivity corresponds to the same order of the local ion neoclassical conductivity in the region of maximum temperature gradient. Reflectometry measurements show turbulence suppression in pellet-injected discharges (as predicted by microstability analysis) and give insights to the change in turbulence in electron internal transport barrier plasmas. Finally, particle transport has been studied in experiments with full LH current drive.

Radiation damage evaluations have been performed with the ACAB code for fusion-relevant materials in an Ignitor-like compact fusion device that could be used as a neutron source for materials testing. Values ranging from 1.6 × 10 −26 to 2.4 × 10 −25 dpa per source neutron have been obtained, which translates into 16–250 dpa/y at full operating power and demonstrates the potential of this neutron-rich device for fusion materials testing. It will be shown that only a few full-power months of operation with a feasible operating duty cycle are sufficient to obtain relevant radiation damage values in terms of dpa. An estimate of the radiation damage on selected machine components will be presented, and solutions to solve the problem of radiation damage to the insulator of the toroidal field insulator will be discussed.

The ignition experiment Ignitor will produce high performance plasma regimes with neutron fluxes at the first wall comparable to those expected in power generating reactors (1015n/cm2/s at the first wall). An appreciable degradation of the inorganic insulators surrounding the conductors is expected and the measurement of basic plasma parameters, such as plasma current and position, by electromagnetic diagnostics can become problematic. Thus an R&D program, aimed at the development of effective and affordable devices for electromagnetic diagnostics with higher damage threshold, is being carried out. Prototype diagnostic coils of different shapes have been manufactured, in collaboration with SALENTEC and Università di Lecce. At the same time, an alternative plasma position control method is being explored, based on the diffraction and detection of the soft X-ray radiation emitted, mostly by Molybdenum impurities, near the top or bottom of the plasma column, where the distance of the plasma last closed magnetic surface from the wall is only few millimeters. The underlying principles and general layout of the new system is essentially an adaptation of a 2-dimensional curved crystal spectrometer. © 2008 American Institute of Physics.

The physics of sawtooth oscillations in plasmas produced by the Alcator C-Mod machine is discussed by identifying the regimes that characterize these oscillations, which occur in the central value of the electron temperature, as measured by X ray and/or ECE diagnostics. Sawteeth are generally present in ohmic as well as in RF heated plasmas. It is shown by numerical analysis that these plasmas are stable to ideal internal, pressure driven modes. Hence, the modes responsible for the crash phase of the sawteeth have to be resistive in nature. This is confirmed by numerical computations using a resistive code. The role of energetic ions, produced by ion cyclotron heating, is also discussed.

Impurity toroidal rotation has been observed in the centre of Alcator C-Mod ion cyclotron range of frequencies (ICRF) heated plasmas, from the Doppler shifts of argon X ray lines. Rotation velocities greater than 1.2 × 107 cm/s (ω = 200 krad/s) in the co-current direction have been observed in H mode discharges that had no direct momentum input. There is a correlation between the increase in the central impurity rotation velocity and the increase in the plasma stored energy (confinement enhancement), induced by ICRF heating, although other factors may be at play. The toroidal rotation velocity is highest near the magnetic axis, and decreases with increasing minor radius. A radial electric field of 300 V/cm at r/a = 0.3 has been inferred from the force balance equation. The direction of the rotation changes when the plasma current direction is reversed, remaining co-current. Impurity toroidal rotation in ICRF heated plasmas is in the direction opposite to the rotation in ohmic L mode plasmas; co-current rotation has also been observed during purely ohmic H modes. When the ICRF heating is turned off, the toroidal rotation decays with a characteristic time of order 50 ms, similar to the energy confinement time, and much shorter than the calculated neoclassical momentum damping time.

A series of experiments, examining the confinement properties of ion cyclotron range of frequencies (ICRF) heated H mode plasmas, has been carried out on the Alcator C-Mod tokamak. Alcator C-Mod is a compact tokamak that operates at high particle, power and current densities at toroidal fields up to 8 T. Under these conditions the plasma is essentially thermal with very little contribution to the stored energy from energetic ions (typically no more than 5%) and with Ti~Te. Most of the data were taken with the machine in a single null `closed' divertor configuration with the plasma facing components clad in molybdenum tiles. The data include those taken both before and after the first wall surfaces were coated with boron, with emphasis on the latter. H modes obtained from plasmas run on boronized walls typically had a lower impurity content and radiated power and attained a higher stored energy than those run on bare molybdenum. Confinement enhancement, the energy confinement time normalized to L mode scaling, for discharges with boronized walls, ranged from 1.6 to 2.4. The unique operating regime of the Alcator C-Mod device provided a means for extending the tests of global scaling laws to parameter ranges not previously accessible. For example, the Alcator C-Mod edge localized mode (ELM)-free data were found to be 1.1 to 1.6 times the ITERH93 scaling and the ELMy data almost 2.0 to 2.8 times the ITERH92 ELMy scaling law, suggesting that the size scaling in both scalings may be too strong. While both ELM-free and ELMy discharges were produced, the ELM characteristics were not easily compared with observations on other devices. No large, low frequency ELMs were seen despite the very high edge pressure and temperature gradients that were attained. For all of our H mode discharges, a clear linear relationship between the edge temperature pedestal and the temperature gradient in the core plasma was observed; the discharges with the `best' transport barriers also showing the greatest improvement in core c

Highly localized direct electron heating [full width at half-maximum (FWHM)≅0.2a] via mode converted ion Bernstein waves has been observed in the Alcator C-Mod Tokamak [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)]. Electron heating at or near the plasma center (r/a⩾0.3) has been observed in H(3He) discharges at B0=(6.0–6.5) T and ne(0)≅1.8×1020 m−3. [Here, the minority ion species is indicated parenthetically.] Off-axis heating (r/a⩾0.5) has also been observed in D(3He) plasmas at B0=7.9 T. The concentration of 3He in these experiments was in the range of n3He/ne≅(0.2–0.3) and the locations of the mode conversion layer and electron heating peak could be controlled by changing the 3He concentration or toroidal magnetic field (B0). The electron heating profiles were deduced using a rf modulation technique. Detailed comparisons with one-dimensional and toroidal full-wave models in the ion cyclotron range of frequencies have been carried out. One-dimensional full-wave code predictions were found to ...

Enhanced confinement modes up to a toroidal field of BT=8 T have been studied with up to 3.5 MW of radiofrequency (rf) heating power in the ion cyclotron range of frequencies (ICRF) at 80 MHz. H-mode is observed when the edge temperature exceeds a threshold value. The high confinement mode (H-mode) with higher confinement enhancement factors (H) and longer duration became possible after boronization by reducing the radiated power from the main plasma. A quasi-steady state with high confinement (H=2.0), high normalized beta (βN=1.5), low radiated power fraction (Pradmain/Ploss=0.3), and low effective charge (Zeff=1.5) has been obtained in Enhanced Dα H-mode. This type of H-mode has enhanced levels of continuous Dα emission and very little or no edge localized mode (ELM) activity, and reduced core particle confinement time relative to ELM-free H-mode. The pellet enhanced performance (PEP) mode is obtained by combining core fueling with pellet injection and core heating. A highly peaked pressure profile with a central value of 8 atmospheres was observed. The steep pressure gradient drives off-axis bootstrap current, resulting in a shear reversed safety factor (q) profile. Suppression of sawteeth appears to be important in maintaining the highly peaked pressure profile. Lithium pellets were found to be more effective than deuterium pellets in raising q0.

Impurity toroidal rotation has been observed in the centre of ohmic plasmas in the Alcator C-Mod tokamak from the Doppler shifts of argon and molybdenum X-ray lines. The rotation is highest (~6 × 106 cm/s) in the early portion of the discharges, when the loop voltage is highest and the electron density is lowest, and then typically settles to values ≤ 2 × 106 cm/s during the steady state period. The impurity rotation is in the same direction as the electron toroidal drift, opposite to the plasma current, and reverses direction when the plasma current direction is reversed. Molybdenum and argon ions rotate with the same velocity. These observations are in qualitative agreement with neoclassical theory

X-ray spectra from Alcator C-Mod plasmas have been collected using a high wavelength resolution, five-spectrometer array during a wide range of operating conditions, providing a large variety of diagnostic information. Each independently scannable von Hamos-type spectrometer has a wavelength range of 2.6–4.1 A and the complete Rydberg series of He- and H-like argon has been observed. Spectra of Δn = 1 ground state transitions and satellites taken along various chords have been simulated using the results from a collisional radiative model and the MIST transport code. Line ratios are used to infer the electron temperature profile and good agreement is found with electron cyclotron emission profiles. Line intensities have been used to obtain absolute argon densities and argon recycling coefficients, and up-down asymmetric density profiles are sometimes observed. The widths of the strongest lines have been used to deduce ion temperature profiles, and wavelength shifts have been used to determine toroidal rotation velocities. Transitions from around n = 9 to the ground state are populated by charge exchange in the outer regions of the plasma and these line intensities have been used to measure neutral hydrogen density profiles. Spectra from helium-like scandium have been obtained during injection experiments and time histories and line intensities have been used to determine impurity transport coefficients. Δn ≥2 ground state transitions in molybdenum charge states around neon-like; have been observed and used to measure the absolute molybdenum density and to test atomic structure calculations.

A five chord, high resolution X-ray spectrometer array has been used to measure vertical brightness profiles of helium-like argon emission from Alcator C-Mod plasmas, out to the last closed flux surface. During standard Alcator C-Mod operation, with the X point and the ion B* Del |B| drift downward, the helium-like argon brightness is a factor of ~8 larger at the top of the plasma than at the bottom, near the plasma edge. In these edge regions, where the electron temperature is low, the upper levels of observed transitions are populated by radiative recombination of hydrogen-like argon. This up-down brightness asymmetry can be explained by a factor of ~8 enhancement of hydrogen-like argon at the top of the machine at r/a=0.9. This implies a vertical impurity drift from inside the plasma, since the hydrogen-like argon is born near the plasma centre. With the ion B* Del |B| drift upward, the enhancement switched to the bottom of the machine, but did not change when the X point was moved up. This asymmetry agrees qualitatively with the predictions of neoclassical parallel impurity transport

Transport physics in the divertor and scrape-off layer of Alcator C-Mod is investigated for a wide range of plasma conditions. Parallel (∥) transport topics include: low recycling, high-recycling, and detached regimes, thermoelectric currents, asymmetric heat fluxes driven by thermoelectric currents, and reversed divertor flows. Perpendicular (⊥) transport topics include: expected and measured scalings of ⊥ gradients with local conditions, estimated χ⊥ profiles and scalings, divertor neutral retention effects, and L-mode/H-mode effects. Key results are: (i) classical ∥ transport is obeyed with ion-neutral momentum coupling effects, (ii) ⊥ heat transport is proportional to local gradients, (iii) χ⊥ αTe−0.6 n−0.6 L−0.7 in L-mode, insensitive to toroidal field, (iv) χ⊥ is dependent on divertor neutral retention, (v) H-mode transport barrier effects partially extend inside the SOL, (vi) inside/outside divertor asymmetries may be caused by a thermoelectric instability, and (vii) reversed ∥ flows depend on divertor asymmetries and their implicit ionization source imbalances.

The behavior of argon and nitrogen, injected into single null diverted Ohmic discharges, has been compared. The screening of the divertor and scrape-off layer plasma is studied by comparing the number of impurity atoms in the core with the number injected by a calibrated gas puff. The effect of the poloidal position of injection on screening has been investigated. The core nitrogen density is dependent of an poloidal position of injection while the argon density is not. The spatial distribution of low charge states in the divertor has also been studied using a multichord visible spectrometer. Initial results of nitrogen modeling with the Monte Carlo code DIVIMP are presented.

The scientific goal of the Ignitor experiment is to approach, for the first time, the ignition conditions of a magnetically confined D-T plasma. The IGNIR collaboration between Italy and Russia is centred on the construction of the core of the Ignitor machine in Italy and its installation and operation within the Triniti site (Troitsk). A parallel initiative has developed that integrates this programme, involving the study of plasmas in which high-energy populations are present, with ongoing research in high-energy astrophysics, with a theory effort involving the National Institute for High Mathematics, and with INFN and the University of Pisa for the development of relevant nuclear and optical diagnostics. The construction of the main components of the machine core has been fully funded by the Italian Government. Therefore, considerable attention has been devoted towards identifying the industrial groups having the facilities necessary to build these components. An important step for the Ignitor programme is the adoption of the superconducting MgB2 material for the largest poloidal field coils (P14) that is compatible with the He-gas cooling system designed for the entire machine. The progress made in the construction of these coils is described. An important advance has been made in the reconfiguration of the cooling channels of the toroidal magnet that can double the machine duty cycle. A facility has been constructed to test the most important components of the ICRH system at full scale, and the main results of the tests carried out are presented. The main physics issues that the Ignitor experiment is expected to face are analysed considering the most recent developments in both experimental observations and theory for weakly collisional plasma regimes. Of special interest is the I-regime that has been investigated in depth only recently and combines advanced confinement properties with a high degree of plasma purity. This is a promising alternative to the high-density L-regime that had been observed by the Alcator experiment and whose features motivated the Ignitor project. The provisions that are incorporated in the machine design, and in that of the plasma chamber in particular, in order to withstand or prevent the development of macroscopic instabilities with deleterious amplitudes are presented together with relevant analyses. © 2013 IAEA, Vienna.

Results of ICRF heating experiments in Alcator C-Mod during the November 1994 to June 1995 campaign are summarized. Efficient heating of high-density plasmas was demonstrated with high power densities (up to volume averaged, surface averaged). These experiments were carried out with RF powers up to 3.5 MW at 80 MHz, at magnetic fields up to 8 T and plasma currents up to 1.2 MA. For on-axis hydrogen minority heating at 5.3 T, near complete absorption is achieved. Energy confinement in L-mode plasmas was found to be consistent with the ITER89-P scaling. H-mode is routinely observed when the ion drift is directed toward the X-point. The H-mode power threshold was found to scale as low as , which is a factor of two lower than the scaling observed on other tokamaks. PEP modes with highly peaked density and ion temperature profiles and highly enhanced fusion reactivity were obtained with Li pellet injection followed by on-axis ICRF heating at both (H minority heating) and 8 T ( minority heating). In H - plasmas at T highly localized direct electron heating by the mode converted ion Bernstein wave was observed. Nearly complete absorption by electrons in a small volume resulted in an extremely high electron heating power density of with .

The divertor of the Alcator C‐Mod tokamak [Phys. Plasmas 1, 1511 (1994)] routinely radiates a large fraction of the power entering the scrape‐off layer. This dissipative divertor operation occurs whether the divertor is detached or not, and large volumetric radiative emissivities, up to 60 MW m−3 in ion cyclotron range of frequency (ICRF) heated discharges, have been measured using bolometer arrays. An analysis of both Ohmic and ICRF‐heated discharges has demonstrated some of the relative merits of detached divertor operation versus high‐recycling divertor operation. An advantage of detached divertor operation is that the power flux to the divertor plates is decreased even further than its already low value. Some disadvantages are that volumetric losses outside the separatrix in the divertor region are decreased, the neutral compression ratio is decreased, and the penetration efficiency of impurities increases.

Extremely broadband (100 GHz-30 THz) single cycle THz pulses are routinely generated with femtosecond laser for Time Domain Spectroscopy applications (TDS). The wide frequency range has an unquestionable diagnostic potential for Tokamak plasmas and not surprisingly THz TDS finds a natural field of application in this area, which is an evolution of the FIR and millimeter waves diagnostics, where ENEA Frascati holds historical expertise. By illuminating the plasma with a THz beam, phase, intensity and polarization of both reflected and transmitted beams can be detected, devising a single diagnostic instrument capable of measuring multiple plasma parameters. We will describe and discuss the laboratory work now in progress to realise a tailored THz-TDS spectrometer with design parameters optimised for the requirements of Tokamak plasmas and the tests of optical fibers and quasioptical couplers to optimise access to plasma. ENEA Frascati and the Photonics group of Physics Dept. of Oxford University are collaborating on this subject [1]. © 2016 IOP Publishing Ltd and Sissa Medialab srl.

A series of transport experiments has been carried out in Alcator‐C‐Mod. [Phys Plasmas 1, 1511 (1994)]. Data from both Ohmic and ICRF (ion cyclotron range of frequencies) heated plasmas can be fitted with an L‐mode (low mode) scaling law. The Ohmic τE’s show no scaling with density in any regime and can reach values of 2–3 times neo‐Alcator. Impurity confinement has been studied with the laser blow‐off technique with τI showing nearly linear scaling with plasma current. Ohmic and ICRF H modes are obtained over a wide range of discharge parameters, extending the range in the international database for nB, by almost a factor of 10. The power threshold for ELM‐free (edge localized mode) discharges is in rough agreement with the scaling P/S=0.044nB. Energy diffusivities of Ohmic and ICRF heated plasmas have been measured from local analysis of plasma profiles and power fluxes. The same analysis produces a value for plasma resistivity which lies between the Spitzer and neoclassical calculations. Analysis of pl...