Search

The paper is devoted to an energy confinement study at the Globus-M spherical tokamak (ST). Experiments were performed in single null divertor configuration with elongation as high as 1.8–1.9 for variable plasma current and fixed toroidal magnetic field. The confinement time (τ E) dependence on density for ohmic-heated (OH) deuterium plasma is presented. It was found that τ E rises linearly with plasma current in H-mode with pure ohmic heating. Pronounced electron and ion heating was achieved in discharges with neutral beam injection at a moderate density level. The dependence of τ E on absorbed power was weak.

The first results of an investigation of geodesic acoustic modes (GAMs) in the Globus-M spherical tokamak are presented. The experiments were performed using Doppler reflectometry. The conditions of the appearance of GAMs, the nature of their development, and their radial localization have been studied. The obtained data are compared to experimental results obtained in tokamaks with large aspect ratios.

This paper describes the challenges of Thomson scattering implementation for core plasma on DEMO and evaluates the capability to measure extremely high electron temperature range 0.5-40keV. A number of solutions to be developed for ITER diagnostics are suggested in consideration of their realization for DEMO. New approaches suggested for DEMO may also be of interest to ITER and currently operating magnetic confinement devices.

The targeted plasma parameters of the compact spherical tokamak (ST) Globus-M have basically been achieved. The reasons that prevent further extension of the operating space are discussed. The operational limits of Globus-M together with an understanding of the limiting reasons form the basis for defining the design requirements for the next-step, Globus-M2. The recent experimental and theoretical results achieved with Globus-M are discussed, the operational problems and the research programme are summarized and finally, the targeted Globus-M2 parameters are presented. The magnetic field and plasma current in Globus-M2 will be increased to 1?T and 0.5?MA, respectively. The plasma dimensions will remain unchanged. With auxiliary heating at a high average plasma density, the temperatures will be in the keV range and the collisionality parameter with ?*???1 will define the operational conditions. Noninductive current drive will be a major element of the programme. The engineering design issues of Globus-M2 tokamak are discussed and the technical tokamak parameters are confirmed by thermal load and stress analysis simulations. The experimental results obtained on Globus-M2 and the limits of its performance should clarify the feasibility of an ST-based super compact neutron source.

The edge plasma of five Globus-M discharges was modeled by the B2SOLPS5.2 code. Plasma current varied in 114–198 kA range, and all discharges were in the H-mode. The modeled scrape-off layer (SOL) width appeared to be inversely proportional to the plasma current. Such a relation is observed in many other tokamaks. Heat flux to the outer divertor target and radiation power was examined for these discharges. Radiation was found to be responsible for 40% of the energy loss. Energy flux to the low outer divertor target was about 1/3 of the energy loss in the single-null low X-point discharges.

Fast degradation of in-vessel optics is one of the most serious problems for all optical diagnostics in ITER. To provide the resistance to mechanical and thermal stresses along with a high stability of optical characteristics under deposition-dominated conditions we suggest using high-reflective metallic (Ag or Al) film mirrors coated on silicon substrate and protected with thin oxide film in the divertor Thomson Scattering (TS) diagnostics. The mirrors coated with Al2O3 and ZrO2 films were tested under irradiation by deuterium ions. The experimental results on the oxide films sputtering are discussed in the context of their applicability for the first mirror protection in ITER.

One of the main challenges of the implementation of divertor Thomson scattering system on ITER is weak laser scattering signal to be detected against intense background plasma radiation. The paper review briefly the line and continuum radiation data from present magnetic fusion devices in the spectral range of interest to TS diagnostics. The results will form the basis of design and development of the TS diagnostics for the ITER divertor.

The structure and surface morphology of aluminum and silver mirrors covered with protective dielectric oxide layer were studied by means of TEM and SEM. The presence of needle-like pores throughout the thickness of the ZrO(2) film and bubble-like pores in Al(2)O(3) was observed. The test for resistivity to deuterium ion bombardment shows that the exposition to a fluence of similar to 2 x 10(20) ions/cm(2) with the ion energy of 40-50 eV results in appearance of blisters on the surface of mirrors covered wit h Al(2)O(3). For the mirrors protected with ZrO(2) no noticeable changes in surface morphology and reflectivity were found even after order of magnitude higher ion fluence. The effect of different porous structures on blistering phenomena is discussed. (C) 2010 Elsevier B.V. All rights reserved.

Tritium retention inside the vacuum vessel is a potentially serious constraint in the operation of large-scale fusion machines like ITER. An in situ diagnostics for first wall H/D/T retention by laser induced desorption spectroscopy (LIDS) is proposed for use between plasma discharges. The technique is based on local baking of the first wall by laser irradiation and subsequent analysis of the in-vessel gas by optical emission spectroscopy of plasma radiation. The local heating implementation, kinetics of H/D/T thermal extraction and the accuracy of optical emission spectroscopy measurements are analysed. To resolve the H/D/T lines spectroscopically, their thermal broadening should be minimized to prevent overlapping of the line shapes. A comparative performance analysis of several types of plasma sources with relatively cold ions is made including the following types of discharges: Penning, RF multipactor, laser torch and ECR. All these radiation sources require rather low power and could be used for remote in situ measurements of relative densities of the thermally extracted hydrogen isotopes.

Thomson scattering (TS) is a proven diagnostic technique that will be implemented in ITER in three independent systems. The Edge TS will measure electron temperature Te and electron density ne profiles at high resolution in the region with r/a>0.8 (with a the minor radius). The Core TS will cover the region r/a

The study is stimulated by the expected noticeable deviation of the electron velocity distribution function (eVDF) from a Maxwellian under condition of a strong auxiliary heating of electron plasmas in tokamak-reactors. The key principles of accuracy estimation of the Thomson scattering diagnostic of non-Maxwellian plasmas in tokamak-reactors are presented. The algorithm extends the conventional approach to the assessment of non-Maxwellian plasmas measurements errors for a broad class of deviations of the eVDF from a Maxwellian. The algorithm is based on solving the inverse problem many times to determine main parameters of the eVDF with allowance for all possible sources of error and statistical variation of the input parameters of the problem. The method is applied to a preliminary analysis of the advantages of the formerly suggested use of various wavelengths of probing laser radiation in the Thomson diagnostics of non-Maxwellian plasma on the example of the core plasma Thomson scattering diagnostic system which is under design for ITER tokamak. The results obtained confirm the relevance of the diversification of the probing laser radiation wavelength.

The thesis describes the Charge-eXchange Recombination Spectroscopy (CXRS) diagnostics setup and its application to the ion temperature determination in the NBI-heated discharges in the Globus-M spherical tokamak. Measurements of the spectral line shape of the radiation emitted by the hydrogen-like carbon ion CVI (5290.5 A) are presented. Comparison of CXRS and Neutral Particle Analyser (NPA) data demonstrates good agreement.

Multi-diagnostic approach developed for the GAM research in the spherical tokamak Globus M is described. Doppler backscattering (DBS) method as the tool for the GAM study, together with the diagnostics of plasma density and magnetic field GAM oscillations, were simultaneously used in experiments. The version of the DBS diagnostics with two cut-offs positioned at different poloidal angles of the minor cross-section was employed in Globus-M. For the GAM plasma density oscillation study, the Dα emission was observed at different angles to restore the spatial mode structure of the GAM plasma density oscillations. At the same time, the array of Mirnov coils was used for the GAM-like magnetic oscillation study, and that made it possible to restore the magnetic field perturbation spatial structure. The coherent and cross-bicoherence analyzes were employed to identify the interaction between the GAM velocity oscillation and plasma turbulent fluctuations. A shorter version of this contribution is due to be published in PoS at: 1st EPS conference on Plasma Diagnostics

First experiments on non-inductive current drive (CD) in the spherical tokamak using lower hybrid (LH) waves at 2.45 GHz are described. The basic characteristics of the new experimental approach utilizing poloidal LH wave slowing-down scheme obtained by modeling are presented at the beginning of the paper. The experimental results for a poloidal slowing-down scheme are described. The CD efficiency comparable or even higher than theoretically predicted is demonstrated in the experiment using a 10 waveguide grill with 120° phase shift between neighboring waveguides.

Almost all optical diagnostic systems in ITER will require the implementation of mirror recovery and protection systems. Plasma cleaning is considered to be the most promising technique for the removal of metal deposits from optical surfaces. The engineering and physical aspects of RF discharge application for continuous or periodic plasma treatment are discussed with a focus on implementation under ITER conditions. The ion flux parameters obtained in capacitively coupled (CC) RF discharge were measured in the mock-up of a plasma cleaning system. The uniformity of sputtering in CC RF discharge with and without a magnetic field was studied experimentally for the cylindrical discharge reactor geometry and compared with numerical simulations. The sharp increase in the sputtering rate resulting from the non-uniform radial distribution of the ion flux was observed near the electrode edges. The longitudinal magnetic field improves sputtering uniformity. It was demonstrated that Al/Al2O3 deposits can be removed in the Ne and D2 plasma of CC RF discharge but long-term exposition results in the degradation of the polycrystalline molybdenum mirror surface. The efficiency of Al sputtering in the atmosphere containing O2 and N2 fractions was studied in the D2/O2 and D2/N2 plasma of glow discharge. The addition of 2% of oxygen or nitrogen increases the sputtering yield by 3–4 times as compared with that in a nominally pure D2 discharge. The impact of metal deposits on the performance of diagnostic mirrors is discussed. It was shown that an ultrathin metallic film with a thickness as low as a few nm may cause a significant degradation of diagnostic mirrors with a transparent coating.

The electron component is the main channel for anomalous power loss and the main indicator of transient processes in the tokamak plasma. The electron temperature and density profiles mainly determine the operational mode of the machine. This imposes demanding requirements on the precision and on the spatial and temporal resolution of the Thomson scattering (TS) measurements. Measurements of such high electron temperature with good accuracy in a large fusion device such as ITER using TS encounter a number of physical problems. The 40 keV TS spectrum has a significant blue shift. Due to the transmission functions of the fibres and to their darkening that can occur under a strong neutron irradiation, the operational wavelength range is bounded on the blue side. For example, high temperature measurements become impossible with the 1064 nm probing wavelength since the TS signal within the boundaries of the operational window weakly depends on Te. The second problem is connected with the TS calibration. The TS system for a large fusion machine like ITER will have a set of optical components inaccessible for maintenance, and their spectral characteristics may change with time. Since the present concept of the TS system for ITER relies on the classical approach to measuring the shape of the scattered spectra using wide spectral channels, the diagnostic will be very sensitive to the changes in the optical transmission. The third complication is connected with the deviation of the electron velocity distribution function from a Maxwellian that can happen under a strong ECRH/ECCD, and it may additionally hamper the measurements. This paper analyses the advantages of a 'multi-laser approach' implementation for the current design of the core TS system. Such an approach assumes simultaneous plasma probing with different wavelengths that allows the measurement accuracy to be improved significantly and to perform the spectral calibration of the TS system. Comparative analysis of the conservative and advanced approaches is given.

The behaviour of the fast particle population during 18 keV hydrogen and 26 keV deuterium neutral beam injection in deuterium plasmas is investigated. Experiments reveal large fast ion losses. The experimental results are confirmed using different types of modelling: simulation using the NUBEAM module, solution of the Boltzmann kinetic equation and solution of the 3D fast ion tracking algorithm. The dynamics of the energetic particle redistribution and losses during sawtooth oscillation and toroidal Alfven eigenmodes are studied. A method to decrease fast ion losses under the current conditions (0.4 T, 0.2 MA) is shown. The influence of the plasma parameters on the energetic ion confinement rate is investigated. Modelling for the Globus-M2 conditions (1 T, 0.5 MA) is performed.

The results of geodesic acoustic mode (GAM) studies in the spherical torus Globus-M via Doppler reflectometry are presented. The intermittent character of the GAM evolution is similar to the limit-cycle oscillation behavior of zonal flows. The evident correlation between the GAM rotational velocity and both Dα emission and plasma density oscillations is exhibited and discussed. The obtained experimental results are compared with those from tokamaks with large aspect ratios.

Electron temperature and density spatial distribution dynamics on the Globus-M (R =0.36 m, a=0.24 m, B =0.4 T, I= 0.2 MA) spherical tokamak was investigated during the NBI heating. The key tool of this research was the upgraded Thomson scattering diagnostics with variable intervals between probing pulses (20 pulses in total). The scattering points were placed along the major plasma radius from the inner to the outer plasma border (10 points in total). Simulation of the electron transport in the L and H-mode, based on experimentally measured ne(R) Te(R) profiles was carried out using the code ASTRA for the OH and NBI dischargers.

This paper describes the challenges of Thomson Scattering implementation in the ITER divertor and evaluates the capability to satisfy project requirements related to the range of the measured electron temperature and density. A number of aspects of data interpretation are also discussed. Although this assessment and the proposed solutions are considered in terms of ITER compatibility, they may also be of some use in currently operating magnetic confinement devices.

A divertor Thomson scattering (TS) system being developed for ITER has incorporated proven solutions from currently available TS systems. On the other hand any ITER diagnostic has to operate in a hostile environment and very restricted access geometry. Therefore the operation in an environment of intensive stray light, plasma background radiation, the necessity meet the requirement using only a 20 mm gap between divertor cassettes for plasma diagnosis as well as to measure plasma temperatures as low as 1 eV severely constrain the divertor TS diagnostic design. The challenging solutions of this novel diagnostic system which has to ensure its steady performance and also the operability and maintenance are the focus of this report. One of the most demanding parts of the in-vessel diagnostic equipment development is the design assessment using different engineering analyses. The task definition and first results of thermal, e/m and seismic analyses are provided. The process of further improving of the design involves identification of susceptible areas and multiple iterations of the design, as needed. One of the key points for all Thomson scattering diagnostics are the laser capabilities. A high-performance and high-power laser system using a steady-state and high-repetitive mode Nd:YAG laser (2J, 50–100Hz, 3ns) has been developed. The reduced laser pulse duration matched with high-speed low-noise APD detector can be very important under high background light level. For diagnostics such as Thomson scattering and Raman spectroscopy, a high-degree of discrimination against stray light at the laser wavelength is required for successful detection of wavelength-shifted light from the laser-plasma interaction region. For this case of high stray light level, a triple grating polychromator characterized by high rejection and high transmission has been designed and developed. The novel polychromator design minimizes stray light while still maintaining a relatively high transmission.

We present here our recent results on the development and testing of the first mirrors for the divertor Thomson scattering diagnostics in ITER. The Thomson scattering system is based on several large-scale (tens of centimetres) mirrors that will be located in an area with extremely high (3–10%) concentration of contaminants (mainly hydrocarbons) and our main concern is to prevent deposition-induced loss of mirror reflectivity in the spectral range 1000–1064 nm. The suggested design of the mirrors—a high-reflective metal layer on a Si substrate with an oxide coating—combines highly stable optical characteristics under deposition-dominated conditions with excellent mechanical properties. For the mirror layer materials we consider Ag and Al allowing the possibility of sharing the Thomson scattering mirror collecting system with a laser-induced fluorescence system operating in the visible range. Neutron tests of the mirrors of this design are presented along with numerical simulation of radiation damage and transmutation of mirror materials. To provide active protection of the large-scale mirrors we use a number of deposition-mitigating techniques simultaneously. Two main techniques among them, plasma treatment and blowing-out, are considered in detail. The plasma conditions appropriate for mirror cleaning are determined from experiments using plasma-induced erosion/deposition in a CH4/H2 gas mixture. We also report data on the numerical simulation of plasma parameters of a capacitively-coupled discharge calculated using a commercial CFD-ACE code. A comparison of these data with the results for mirror testing under deuterium ion bombardment illustrates the possibility of using the capacitively-coupled discharge for in situ non-destructive deposition mitigation/cleaning.

The experimental and theoretical results obtained in the last two years on the interaction of neutral particle beams and high-frequency waves with a plasma using the spherical tokamak Globus-M are discussed. The experiments on the injection of low-energy proton beam of ~300 eV directed particle energy are performed with a plasma gun that produces a hydrogen plasma jet of density up to 3 × 1022 m−3 and a high velocity up to 250 km s−1. A moderate density rise (up to 30%) is achieved in the central plasma region without plasma disruption. Experiments on high-energy (up to 30 keV) neutral beam injection into the D-plasma are analysed. Modelling results on confinement of fast particles inside the plasma column that follows the neutral beam injection are discussed. The influence of the magnetic field on the fast particle losses is argued. A neutral beam injection regime with primary ion heating is obtained and discussed. The new regime with fast current ramp-up and early neutral beam injection shows electron temperature rise and formation of broad Te profiles until the q = 1 flux surface enters the plasma column. An energetic particle mode in the range of frequencies 5–30 kHz and toroidal Alfven eigenmodes in the range 50–300 kHz are recorded in that regime simultaneously with the Te rise. The energetic particle mode and toroidal Alfven eigenmodes behaviour are discussed. The toroidal Alfven eigenmode spectrum appears in Globus-M as a narrow band corresponding to n = 1. The first experimental results on plasma start-up and noninductive current drive generation are presented. The experiments are carried out with antennae providing mostly poloidal slowing down of waves with a frequency of 920 MHz, which is higher than a lower hybrid one existing under the experimental conditions. The high current drive efficiency is shown to be high (of about 0.25 A W−1), and its mechanism is proposed. Some near future plans of the experiments are also discussed.

The ITER design has highlighted the fundamental need to monitor the machine operation in more detail. The mission of the Thomson scattering diagnostics in the ITER divertor research/operation is discussed with due attention paid to challenges and capabilities of the existing diagnostic design.

Experiments and simulations to achieve high values of plasma parameters at the Globus-M spherical tokamak (ST) at moderate absolute auxiliary heating power (up to 0.8 MW) and high specific heating power (up to 2–3 MW m−3) are described. Important distinguishing features are the low edge safety factor range, which is unusual for STs, 2.7 < q < 5 and small plasma–outer wall space (3–5 cm). High ion heating efficiency with neutral beam injection (NBI) was demonstrated. Results of numerical simulation of fast ion trajectories are described and fast ion generation during the NBI and ion cyclotron resonance heating is discussed. Investigations on their confinement and slowing down are also presented. Reasons for achievement of high IC heating efficiency are outlined. Reliable H-mode regime achievement is described. Transport ASTRA modelling demonstrated that during NB heated H-mode ion heat diffusivity remains neoclassical and the particle diffusion coefficient inside transport barrier decreases significantly. Analysis was performed of divertor tile and special probe surfaces after irradiation by plasma during a large number of shots (3000–10 000 shots). Mixed layer composition is measured and deuterium retention in different tokamak first wall areas is estimated. Plasma jet injection experiments with upgraded plasma jet are described. Jet penetration to the plasma centre with immediate increase of density and temperature drop is proved and analogy with pellet injection is outlined.

The lifetime of front optical components unprotected from reactor grade plasmas may be very short due to intensive contamination with carbon and beryllium-based materials eroded by the plasma from beryllium walls and carbon tiles. Deposits result in a significant reduction and spectral alterations of optical transmission. In addition, even rather thin and transparent deposits can dramatically change the shape of reflectance spectra, especially for mirrors with rather low reflectivity, such as W or Mo. The distortion of data obtained with various optical diagnostics may affect the safe operation of ITER. Therefore, the development of optics-cleaning and deposition-mitigating techniques is a key factor in the construction and operation of optical diagnostics in ITER. The problem is of particular concern for optical elements positioned in the divertor region. The latest achievements in protection of in-vessel optics are presented using the example of deposition prevention/cleaning techniques for in-machine components of the Thomson scattering system in the divertor. Careful consideration of well-known and novel protection approaches shows that neither of them alone provides guaranteed survivability of the first in-vessel optics in the divertor. Only a set of complementary prevention/cleaning techniques, which include special materials for mirrors and inhibition additives for plasma, is able to manage the challenging task. The essential issue, which needs to be addressed in the immediate future, is an extensive development of techniques tested under experimental conditions (exposure time and contamination fluxes) similar to those expected in ITER.