K.A. Podushnikova

First mirrors in ITER: material choice and deposition prevention/cleaning techniques

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 ITER divertor Thomson scattering system: engineering and advanced hardware solutions

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.

Investigation of beam– and wave–plasma interactions in spherical tokamak Globus-M

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.

Perspectives of Use of Diagnostic Mirrors with Transparent Protection Layer in Burning Plasma Experiments

We evaluate using of metal mirrors over‐coated with transparent protection layer for the in‐vessel diagnostic systems in reactor‐grade fusion devices. Ideally, these should satisfy the contradictory demands of high reflectivity and small rate degradation when being bombarded by CX atoms. The serious threat to the performance of diagnostic mirrors is surface contamination with carbon‐based material eroded from carbon tiles. Via coupling the protective layer to a bulk mirror we can mitigate the deposit infiuence on the reflectance spectra. The regards are given to survivability in plasma environment of protected coated metallic mirrors.

Thomson scattering diagnostics for ITER divertor

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.

Research on mirror cleaning in inductively and capacitively driven radio-frequency discharges

Bench tests are used to compare cleaning performance of inductively and capacitively driven radio-frequency (RF) discharges as a potential tool for in-situ maintenance of in-vessel diagnostic mirrors in fusion devices. The effect of erosion of hydrogenated carbon coating is studied in different processing conditions. Stainless steel (SS) mirrors have been exposed to CH4‒Ar and H2‒Ar plasmas in an RF discharge at a pressure of 10−2 Torr with an input power of 0.5 kW at 13.6 MHz. The samples, which exhibit a slow rate of chemical erosion, become essentially erosive in both inductively and capacitively driven RF discharges. The cleaning ability of a capacitively driven RF discharge is studied in dedicated experiments with SS samples retrieved from the tokamaks T-10 and Globus-M after long-term exposure to the working and wall conditioning discharges.

Near-infrared plasma spectroscopy in support of divertor Thomson scattering diagnostics development for 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.

Experimental results from the TUMAN 3 tokamak

The ‘‘TUMAN‐3’’ Tokamak programme concentrates on issues of improved confinement. In 1989 the transition from an ordinary Ohmic regime into an improved confinement mode was achieved. The signatures of the H‐mode in auxiliary heated tokamaks have been observed in this regime. The crucial role of the boundary radial electric field was found in the experiments with internal bias probe. Other techniques were demonstrated to disturb the boundary plasma which led to H‐mode triggering: short increase of working gas puffing, minor radius magnetic compression and pellet injection. The role scaling of the energy confinement time in the Ohmic H‐mode was obtained, which differs dramatically from the scaling for the ordinary Ohmic regime. There were found a strong dependence of τE on plasma current and a weak dependence on density. The maximum value of τE was 10 times longer than in the ordinary Ohmic region. The τE scaling for the Ohmic H‐mode is consistent with the scaling proposed for devices with powerful auxiliar...

Thomson scattering diagnostic systems in the ITER divertor

A diagnostic array has been developed for studying the operating modes of the divertor in the ITER tokamak-reactor using the Thomson scattering technique. The aim of this study is to measure the spatial profiles of the electron temperature and density. The structure of the diagnostic setup was selected on the basis of a classical diagnostic geometry and the high-resolution LIDAR system, which provide access to different regions of the divertor plasma. A severe radiation environment, limited access to the plasma in the ITER divertor, and a high-dust environment (the divertor plate erosion material) in the divertor volume pose many problems for performing diagnostics under unique conditions having no analogs in the tokamaks that are now in operation. Different methods for protecting optical surfaces from plasma-enriched deposition are proposed and analyzed. The efficiency of these methods has been demonstrated in bench tests. The concept of laser and detector systems and diffraction polychromators capable of operating at different electron temperatures with a lower limit of 1 eV, has been justified and approved.

Plasma processes accompanying fast current decrease in the Tuman-3 tokamak

The authors report an investigation of the influence of radial current distribution on the Tuman-3 tokamak plasma parameters under the conditions of ohmic heating with a twofold current decrease for a period 4-6 times shorter than the magnetic field diffusion time-scale. A slight change in the parameters was observed in the current ramp-down phase. The characteristic relaxation time of the plasma parameters after the current drop was considerably larger than the energy confinement time and comparable with the field diffusion time-scale. The influence of the current ramp-down on the boundary plasma is also studied. Narrowing of the current density profile is accompanied by a 1.5-2-fold rise of the energy confinement time compared with its steady-state value.