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The effect of a high-frequency discharge on the change in the topography of the surface of KU-1 optical quartz and transmission of visible light (400–1000 nm) is studied. The working gases of the discharge are D2 and a D2/N2 mixture, in which the fraction of N2 is 25 mol %. The addition of nitrogen increases the rate of sputtering from 60 to 300 nm/h without changing the stoichiometry of the surface layers. After the exposure in plasma, the root-mean-square roughness of the surface decreases from 1.3 to 0.6 nm. The transparence of quartz remains unchanged. The analysis of the roughness and calculation of the diffuse light scattering with a wavelength of 400 nm after passing through quartz are executed using the power spectral density functions.

Migration of plasma erosion products in plasma facilities is studied experimentally and numerically within the framework of modeling transport of plasma-facing materials in the diagnostic ducts of fusion devices. Material transport simulation is discussed for two cases of low and high background neutral gas pressures. Monte Carlo software KITe was used to simulate transport at a neutral gas background pressure 0.1–0.5 Pa—typical during steady-state tokamak operation and during pressure pulses caused by edge localized modes (ELMs). The simulation approach was implemented to describe experiments at the MAGNUM-PSI facility. Fluid dynamic code FLUENT is used to simulate transport during pressure surges as high as 1000 Pa, which can occur in the case of severe disruptions in tokamak plasma discharges, such as vertical displacement events (VDE) or accidental events. The hydrodynamic approach was verified in simulation of target sputtering in the QSPA plasma gun facility.

This article reports the results of the assessment performed on the ITER Divertor Thomson Scattering (DTS). It is designed to provide an instrument capable of measuring the electron temperature and the related density profiles in the outer divertor plasma, relying on Thomson scattering method. The DTS main components are front rack, back rack, neutron shielding and diagnostic mirrors. Current DTS design has been modified compared to the one described in the previous article dedicated to this object. The most significant changes are: • Upper neutron shielding has been removed. It is now not connected directly with the DTS system. It helped to reduce total weight of the construction; • Dielectric insertions were added in order to reduce influence of the ponderomotive forces on construction strength; • DTS supports were redesigned. Connections of the DTS with radial rails on which the construction is mounted on the sides were explored. Corresponding contact problem was solved and the structure integrity in this area was assessed. Mirrors made of SiC material are the DTS diagnostic elements. They are installed into front and back diagnostic racks. Other elements of the DTS are made of 316L(N)-IG steel. The contact pads on the radial rails are made of 660 steel. The whole Divertor Thomson Scattering system has to withstand all the loads (thermal, mechanical, electromagnetic and seismic) acting on the in-vessel region in which it is located without compromising its integrity. Its design layout is being modified and optimized according to the outcome of the structural assessment. Every step of the design update brings the construction closer to satisfying all of the criterions of structural integrity. This paper describes some important steps that really help moving forward to the DTS implementation.

We developed high energy Q-switched 1064 nm and 946 nm Nd:YAG lasers for ITER Divertor Thomson Scattering (DTS). The 1064 nm Nd:YAG laser is designed to be used as the main DTS probing laser and features output energy of 2 J at 3 ns pulse duration. MOPA configuration with phase conjugation mirror provides divergence of 1.3 times the diffraction limit and high stability of the output beam parameters with repetition rate varied from 1 to 50 Hz. The 946 nm Nd:YAG laser designed for calibration by an auxiliary Thomson spectrum at the neighboring wavelength. The 946 nm laser consists of MO based on cavity dumping, high gain regenerative preamplifier and zero diffraction length ring cavity regenerative power amplifier with image-rotation providing 1 J × 50 Hz at 3 ns pulse duration with divergence of 1.5 the diffraction limit. As far as we know this is the highest value of the output energy obtained at 946 nm with nanosecond pulse duration until now. Both lasers use unified pump modules providing high maintainability of the system. The paper describes laser design approaches, state of the art and outlook of further developments.

Laser-induced fluorescence (LIF) on ITER will be used for local measurement of helium density nHe and ion temperature Ti in divertor. The LIF diagnostic is combined with the divertor Thomson scattering (DTS) via common laser injection and signal collection optics. Collisional-radiative model developed for helium ions is used to estimate laser parameters required for robust measurement of laser-induced signals against plasma background radiation. The estimations demonstrate feasibility of measuring nHeII with laser parameters provided by available optical parametric oscillator pumped by Nd:YAG. New approach for measuring Ti is proposed and the expected errors are estimated using numerical simulations.

Hot ion mode was recently achieved at the Globus-M2 spherical tokamak when the toroidal magnetic field was as high as 0.9 T and plasma current 0.35 MA. The injection of two neutral beams with a total power of 0.7 MW and an energy of 30 keV into a plasma with major radius 0.36 m and minor radius 0.22 m (aspect ratio 1.6) made it possible to heat the plasma ions up to 4 keV, while electron density in the plasma center was as high as 8 × 1019 m−3. The achievement of high temperatures became possible due to the good thermal insulation of the plasma.

Leucosapphire (c-LS) structure and transmission in visible after surface treatment in 90%H2-10%N2 RF discharge are studied. According to AFM, the number of scratches of the mechanically polished surface decreased significantly after removal of about a 300 nm layer (exposure time of 12 h) under unchanged rms of roughness. According to TEM, a two-layer structure formed in the near-surface region consists of an outer 10 nm amorphous layer followed by a crystalline layer of 40-50 nm with a high defect density. The c-LS transmission in the angle of 400-1000 nm either slightly increased or remained unchanged. The demonstrated transmission stability during exposure in 90%H2-10%N2 RF discharge allows us to consider the plasma sputtering as a promising technique for cleaning contaminated windows protecting first mirror of divertor Thomson scattering being developed for ITER divertor Keywords: Leucosapphire, RF discharge, hydrogen, nitrogen, AFM, TEM, surface layer structure, visible transmission.

The work presents the results of the energy confinement study carried out on the compact spherical tokamak Globus-M2 with a toroidal magnetic field as high as 0.8 T. A reproducible and stable discharge was obtained with the average plasma density (5–10) × 1019 m−3. Despite the increase in the magnetic field, the neutral beam injection (NBI) led to clear and reproducible transition to the H-mode accompanied by a decrease in the turbulence level at the plasma edge. NBI allowed effectively heat the plasma: electron and ion temperatures in the plasma core exceeded 1 keV. Compared to the previous experiments carried out with a toroidal magnetic field as high as 0.4 T plasma total stored energy was increased by a factor of 4. The main reason or this phenomenon is a strong dependence of the energy confinement time on the toroidal magnetic field in the spherical tokamak. It was experimentally confirmed that such kind of dependence is valid for ST with magnetic field up to 0.8 T. It has also been shown that the enhancement of the energy confinement in Globus-M2 with collisionality decrease is associated with an improvement of both electron and ion thermal insulation.

ITER (International Thermonuclear Experimental Reactor) is a nuclear fusion research and engineering project. It is supposed to be the first fusion device to test the integrated technologies, materials and operational regimes necessary for the commercial production of fusion-based electricity. Front (explored in this paper) and Back diagnostic racks are used as supporting structures of the so-called ITER Divertor Thomson Scattering (DTS) Diagnostic System that is created to measure plasma characteristics. They will be attached to the lower port walls (ITER ports are used to locate diagnostic equipment and its supporting structures) through the so-called Divertor Rails. The developed design of the Front (Diagnostic) Rack locking (e.g. Rack attachment to the rails mechanism) was analyzed in this research to be consistent with thermal stresses, electromagnetic and seismic loads taking into account some specific assembly and maintenance loads including preloading under installation. The analysis of the rack locking design's resistance to the mentioned loads should prove correspondence of the suggested common rack design with the structure integrity criterions. Construction strength under the most severe loading scenario was estimated in this research and appropriate conclusions were made. It was found out that the stress criterion was not fully satisfied, so a new rack locking design concept was proposed. Its main details are also described at the end of this article.

The application of the laser-aided diagnostic using the nanosecond laser for direct measurement of hydrogen concentration on the walls of the Globus-M2 tokamak is under discussion. The D release after the laser irradiation with the fluence in the range of 0.1 - 4 J/cm2 was measured in laboratory experiments at the Large MEPhI mass-monochromator facility for W-D, Ti-D films, and for the W tile exposed in the Globus-M tokamak. The desorption flux was high enough to be detected during similar measurements inside the Globus-M2 tokamak. TMAP 7 modeling of D release from W-D films was done for the relatively low laser fluence (

ITER is a nuclear fusion research and engineering project. It is supposed to be the first fusion device designed for testing the integrated technologies, materials, and physical aspects necessary for development of the commercially available fusion-power plant. One of the important components of the project is optical diagnostic systems with collecting mirrors. These mirrors have to provide stability of optical systems under severe loads of different types that could possibly arise in the tokamak. The collecting mirrors of several ITER diagnostics have a large scale and should be installed into diagnostic ports. Thermal stress analysis of the mirror updated design is aimed to obtain deformation and rotation values of the mirrors’ reflecting surfaces in order to conduct ray tracing analysis and to edit mirrors alignment to provide correct functioning of the optical systems. The maximum temperature values of the Divertor Thomson Scattering collecting mirrors were estimated for the normal operation mode. The FE model of the second mirror takes into account all the force boundary conditions, basic kinematic boundary conditions and constraints. Boundary conditions taken for the simulation were applied on surfaces contacting with the diagnostic rack. Thermally stressed state was calculated and corresponding displacement and rotation distributions were obtained.

Numerous plasma-near mirrors of optical diagnostics in ITER require protection from erosion and deposition caused by impinging particles. This is achieved by 63 Diagnostic Shutters, which shall operate in vacuum under high thermal and neutron fluxes over 20 years without maintenance, ruling out standard engineering solutions. In fact, these conditions are unprecedented even on fusion devices. Hence, qualification R&D efforts are extensive, because if a shutter fails, so does the respective diagnostic. Shutter design tasks are widespread among Domestic Agencies and their suppliers, because every shutter is part of the diagnostic it shall protect when not in use. However, as these highly resembling systems have obvious synergy potential, a coordination strategy for all ITER shutters was implemented at IO. An extensive shutter experience collection including failure reports from 14 fusion devices was performed. These are summarized in the present work. For the first time, the state-of-the-art of shutters is thereby defined and assessed as to its applicability to ITER. Furthermore, design-driving environmental effects due to the specific operating conditions are recalled and evaluated. The findings of both assessments are put into context with the current designs of all ITER shutters. In a next step, these are reviewed with emphasis on synergies between different shutter systems. Finally, recommendations on necessary prototyping and generic components are given.

The ITER Divertor Thomson Scattering system is designed to provide an instrument capable of measuring the profiles of electron temperature and relative profiles of electron density in the outer divertor plasma. Main components of construction are front rack, back rack, neutron shielding and diagnostic mirrors. The whole Divertor Thomson Scattering system has been analyzed to be consistent with thermal, mechanical, electromagnetic and seismic loads which can affect construction. All the obtained results, such as displacement and stress maps, temperature distribution and values of forces and torques in construction during plasma disruption events, can lead to some significant changes in construction design in order to improve the system.

Laser-induced fluorescence diagnostics on ITER will be used for local measurements of helium density (nHeI) and ion temperature (Ti) in the divertor area. The diagnostics is combined with divertor Thomson scattering via common laser injection and signal collection optics. Multichannel filter polychromators with design similar to those developed for Thomson scattering will be used to perform spectral analysis of the fluorescent signals. Both the fluorescent signals and the background radiation are estimated using the collisional-radiative models developed for He I and He II. The estimations demonstrate the feasibility of measuring nHeI with the temporal resolution 20 ms and accuracy ΔnHeI

The ITER divertor Thomson scattering (TS) system has been significantly upgraded in the last few years in terms of equipment and experimental setup design. In this paper, we are focusing on lasers, digital filter polychromator and piezo actuators, which could be also useful for other diagnostics. Capabilities of the probing and calibration lasers are specified and compared against ITER requirements. Key features of the digital polychromator design for TS applications are presented and obtained signal-to-noise ratios are discussed. The main reason to use piezo actuators for in-vessel remotely controlled devices is high radiation and magnetic resistance of piezo elements. Piezo actuators of inertial and ultrasonic types were produced and their promising performance for fusion devices was demonstrated.

This report summarizes the status of several R&D tasks devoted to characterization of the basic behavior and definition of some features of the RF discharge mirror cleaning systems for ITER spectroscopy diagnostics. First results of mirror cleaning system engineering development and its implementation on ITER are described. Key requirements and specifications for such mirror cleaning systems for ITER conditions are presented.

This note is devoted to a filter polychromator combining built-in optical, data acquisition and processing subsystems designed for electron temperature and density measurements in fusion plasma by Thomson scattering (TS). Special attention is paid to assessing the accuracy of TS signals measured in the polychromator spectral channels. The potential for estimating TS error caused by background emission is demonstrated using high-pass signal. The recent polychromator tests have shown APD gain stabilization better than ± 0.5% over the ambient temperature range from 17.5 °C to 35 °C

The LabBot Framework project is intended to implement control of experiment and data acquisition without writing special platform codes, while achieving industrial-grade results. The Framework is intended for fast and advanced development of controlling and data acquisition software for laboratory-scale experimental setups; more sophisticated software for middle-scale laboratory equipment; industrial-grade software for self-made commercial instrumentation developed by small- and middle-scale companies. The paper presents review of the Framework top-level architecture and its implementation addressing development of ITER Divertor Thomson Scattering (DTS) diagnostic. The first examples of the Framework implementation include the set-ups for vacuum heating testbench; RF cleaning of diagnostic mirrors as well as DTS equipment, which must be controlled by the ITER CODAC system.

The paper describes the digital filter polychromator developed for Thomson scattering (TS) diagnostics. The device having ultra-low noise, high-speed detection system and up to 7 spectral channels is equipped with digitizer and single board computer for system control and data processing. The data acquisition system is capable to detect very short laser pulses of few ns width with low magnitude. This feature opens up new possibilities for the TS diagnostics design in modern fusion devices, which includes: operation with high background illumination, measurement of very low electron temperature and density at the plasma edge and scrape-off layer (SOL), separation of the TS signal and the stray light coming with a delay of several ns. The analysis of the TS signal waveform also gives a possibility to distinguish the TS and Mie scattering from the dust particles that is important for measurements in SOL. Owing to compactness and ultra-low energy consumption both optical and data processing units interfaced with optical Gigabit Ethernet were combined in a stand-alone device. Such all-in-one design with optical fiber input and optical digital output provides enhanced resistance to electromagnetic disturbances and simplify operation. A set of such polychromators can be mounted in a standard 19” rack and can be integrated with outer CODAC (Control, Data Access and Communication) system. Testing of the polychromators at the Globus-M2 tokamak has shown excellent performance of the developed serial devices.

The support structure for divertor Thomson scattering equipment – the front diagnostic rack, which actually plays plugging role of the divertor port, should be designed to sustain the severe ITER conditions. Meeting the requirements of multifield analyses (which often contradict each other) results in an iterative design process. A number of design variants based on engineering analyses results were developed in 2011–2012. We study here the close to the final design of the diagnostic rack for consistency to electromagnetic, thermal and seismic loads. The specific ITER environment imposes a restricted list of materials and requires a careful design of optical elements to accommodate their thermal expansion. Special attention is focused on the mirror deformed shape under operating loading conditions and its effect on optical system performance, which is vital for all optical systems with mirrors specially designed for the ITER.

There are several protecting techniques managing with contamination on optical surfaces of in-vessel diagnostic components in ITER. Analysis of impurity transport in narrow and curved gaps gave us the idea that it can’t be explained by convection flows. The proposed protecting construction, situated between plasma and irradiating laser mirror launcher, was analysed for effectiveness. The protecting ability of this construction is based on the principles of hydrodynamic, in particular on bevelled entrance, which provides redirection of polluting gas flow away from the optical components due to angling optical and geometrical channel axes. Several different numerical simulations were studied. The design, setting objectives as well as equations and parameters are under discussion. Results of 2D and 3D numerical simulations are provided.

Erosion and re-deposition of plasma-facing components in tokamaks require the development of in-vessel optics protection and cleaning techniques. Front-end diagnostic optics in large-scale fusion m...

The divertor port racks used as a support structure of the divertor Thomson scattering equipment has been carefully analyzed to be consistent with electromagnetic and seismic loads. It follows from the foregoing simulations that namely these analyses demonstrate critical challenges associated with the structure design. Based on the results of the reference structure [2] a modified design of the diagnostic racks is proposed and updated simulation results are given. The results signify a significant improvement over the previous reference layout and the design will be continued towards finalization.

The operation of ITER requires high-quality estimates of the plasma electron density over multiple regions in the plasma for plasma evaluation, plasma control and machine protection purposes. Although the density regimes of ITER are not very different from those of existing tokamaks (1018–1021 m−3), the severe conditions of the fusion plasma environment present particular challenges to implementing these density diagnostics. In this paper we present an overview of the array of ITER electron density diagnostics designed to measure over the entire ITER domain: plasma core, pedestal, edge, scrape-off layer and divertor. It will focus on the challenges faced in making these measurements, and the technical solutions of the current designs.

First mirrors of optical diagnostics in ITER are subject to charge exchange fluxes of Be, W, and potentially other elements. This may degrade the optical performance significantly via erosion or deposition. In order to restore reflectivity, cleaning by applying radio frequency (RF) power to the mirror itself and thus creating a discharge in front of the mirror will be used. The plasma generated in front of the mirror surface sputters off deposition, restoring its reflectivity. Although the functionality of such a mirror cleaning technique is proven in laboratory experiments, the technical implementation in ITER revealed obstacles which needs to be overcome: Since the discharge as an RF load in general is not very well matched to the power generator and transmission line, power reflections will occur leading to a thermal load of the cable. Its implementation for ITER requires additional R&D. This includes the design of mirrors as RF electrodes, as well as feeders and matching networks inside the vacuum vessel. Mitigation solutions will be evaluated and discussed. Furthermore, technical obstacles (i.e., cooling water pipes for the mirrors) need to be solved. Since cooling water lines are usually on ground potential at the feed through of the vacuum vessel, a solution to decouple the ground potential from the mirror would be a major simplification. Such a solution will be presented.

The metal mirrors for deflecting laser radiation in the divertor Thomson scattering system of ITER are found to be inapplicable due to high density of laser radiation on a mirror. This is caused by the short distance from the laser focus to the mirror surface. We report on investigations of the possibility to use dielectric mirrors and protective sapphire plates in a rotary unit. We study the laser damage thresholds of optical elements, which are supposed to be used in the laser input channel. These optical elements are the dielectric mirrors deposited on different substrates (quartz, sapphire, and single-crystal molybdenum) and the protecting sapphire plate. A number of sapphire samples were irradiated by a neutron flux up to 1019 n/cm2 (E > 100 keV) and annealed. Laser damage thresholds of the elements were measured at room temperature and at the ITER operating temperature of 150°C.A YAG:Nd laser operating with 10-Hz repetition rate was used in the experiments. The laser pulse parameters ...

Диагностика томсоновского рассеяния (ТР) является ключевой диагностикой при проведении экспериментальных исследований на токамаках. Изучение динамики профилей температуры и концентрации электронов, измеренных методом ТР, позволяет получать сведения о режимах удержания и термоизоляции высокотемпературной плазмы, что является одной из важнейших проблем УТС. На токамаке Глобус-М выполнена модернизация диагностики томсоновского рассеяния. Число пространственных каналов увеличено до 10, что позволило измерить профили температуры и концентрации электронов от внутренней до внешней границы плазмы. Благодаря применению новой системы сбора света и регистрации рассеянного излучения увеличена точность измерений. В статье обсуждаются особенности калибровки диагностики и обработки сигналов рассеяния, приводятся первые результаты эксперимента. К настоящему времени модернизированная система ТР задействована в программе исследований режимов с инжекцией высокоэнергетичной плазменной струи, нейтральной инжекцией и режимов с внутренними транспортными барьерами.

Analysis of stability of the peeling-ballooning modes in the edge plasma of the Globus-M tokamak is presented. Studies were performed using the ideal 3D MHD code written in the BOUT++ framework. The edge localized modes which are frequently observed in the Globus-M plasma could be described using MHD theory. It was found that the parameters of the Globus-M edge plasma are in the range of stability for the peeling-ballooning modes. The instability modes with a toroidal number of n = 12 and typical distances between filaments of approximately ~12 cm were found to be the most unstable. This statement agrees with the Doppler backscattering measurements. It was also considered whether the plasma in the Globus-M2 tokamak will be stable against the PB instability.

25th Fusion Energy Conference (FEC), Govt Russian Federat, St Petersburg, RUSSIA, OCT 13-18, 2014; International audience; In this paper we present the fusion related activities of the Plasma Physics Division at the Ioffe Institute. The first experiments on lower hybrid current drive (LHCD) in a spherical tokamak performed at the Globus-M tokamak (R = 0.36 m, a = 0.24 m, B-t = 0.4 T, I-p = 200 kA) with a novel poloidally oriented grill resulted in an RF driven current of up to 30 kA at (100kW, 2.5 GHz), exceeding the modelling predictions. At the FT-2 tokamak (R = 0.56 m, a = 0.08 m, B-t = 3T, I-p = 30 kA) experiments with a traditional toroidally oriented grill revealed no strong dependence of the LHCD density limit on the H/D ratio in spite of LH resonance densities differing by a factor of 3. Microwave Doppler reflectometry (DR) at the Globus-M, and DR and heavy ion beam probe measurements at the tokamak TUMAN-3M (R = 0.53 m, a = 0.24 m, B-t = 1.0T, I-p = 190 kA) demonstrated geodesic acoustic mode (GAM) suppression at the L to H transition. Observations at FT-2 using Doppler Enhanced Scattering showed that the GAM amplitude is anti-correlated both spatially and temporally to the drift turbulence level and electron thermal diffusivity. For the first time turbulence amplitude modulation at the GAM frequency was found both experimentally and in global gyrokinetic modelling. A model of the L-H transition is proposed based on this effect. The loss mechanisms of energetic ions' (EI) were investigated in the neutral beam injection (NBI) experiments on Globus-M and TUMAN-3M. Empirical scaling of the 2.45 MeV DD neutron rate for the two devices shows a strong dependence on toroidal field B-t(1.29) and plasma current I-p(1.34) justifying the B-t and I-p increase by a factor of 2.5 for the proposed upgrade of Globus-M. Bursts of similar to 1 MHz Alfvenic type oscillations correlating with sawtooth crashes were observed in ohmic TUMAN-3M discharges. The possibility of low threshold parametric excitation of Bernstein and upper hybrid waves trapped in drift-wave eddies resulting in anomalous absorption in electron cyclotron resonance heating (ECRH) experiments in toroidal plasmas was identified theoretically. A novel method of radial correlation Doppler reflectometry is shown to be capable of measuring the turbulence wave-number spectrum in realistic 2D geometry. On the progress in design and fabrication of three diagnostics for ITER developed in the Ioffe institute is reported: neutral particle analysis, divertor Thomson scattering and gamma spectroscopy.

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.

ITER design has highlighted the fundamental need to monitor and deepen understanding of divertor operations. The designed Thomson scattering (TS) for the divertor leg and SOL near the X-point, has gained a lot from present State-of-the-Art in the diagnostics. However, hostile environment, very restricted access, and required spatial resolution, all impose severe constraints on the diagnostic design for ITER. Signal transmission from diagnostics introduces the further considerations of relative movements of the machine components. Despite the constraints it is feasible using of crossed probe and viewing beams to make measurements of the electron temperature and density profiles in outer divertor leg at the extent from the strike-point to X-point vicinity, and the LIDAR TS technique to make the electron temperature and density scan along the major radial line-of-sight at ∼0.5 m above the X-point. The interfaces for each of the subsystems are presented and discussed in the sections below.

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.

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.

Three diagnostic systems are being developed in Ioffe Institute for ITER. Those are Neutral Particle Analysis (NPA), Thomson Scattering in Divertor (TSD) and Gamma Spectroscopy (GS). The main objective of NPA in ITER is to measure D/T fuel ration in plasma on the basis of measurement of neutralized fluxes of D and T ions [1]. Fuel ratio is one of the key parameters needed by ITER control system to provide the optimal conditions in plasma and the most effective plasma burning. Another objective is to measure the distribution function of fast ions (including alpha particles) generated as a result of the additional heating and nuclear fusion reactions. Thomson Scattering in Divertor (TSD) [2] will be used to measure electron temperature and density in the scrape-off layer in outer leg of ITER divertor. The main task of TSD is to protect the machine from divertor overloading. Gamma Spectroscopy (GS) [3] is based on the measurement of spectral lines of MeV range gammas generated in nuclear reactions in plasma. 2-D gamma-ray emission measurements give valuable information on the confined alpha particles in DT plasma. They also provide important information on the location of MeV range runaway electron beams in ITER plasma. For all three cases the physical basis and instrumentation are presented. The simple NPA version for measurements of D/T ratio in DEMO is also briefly described.

Diagnostics which is currently installed or under active development for a newly commissioned spherical tokamak (ST) Globus-M is described. Among the commonly implemented plasma diagnostics the pri...

ITER divertor operation is dominated by the necessity to exhaust around 200MW power via the scrape-off layer. A large fraction of the input power must be irradiated by the impurities either intrinsic or seeded. It is important that the radiation source be well distributed over the entire divertor plasma. The plasma detachment at the divertor target should be precisely adjusted as to enable a partially attached operating, that is detached near the separatrix strike point and attached further out in the scrape-off layer. To provide information on key fenomena which may limit the divertor performance is the challenging task for diagnostics in ITER. The reliable T e , n e profile measurements in the divertor upstream (near X-point) and downstream (divertor bottom) regions address the highly promising Thomson scattering diagnostics. The high resolution time-of-flight LIDAR Thomson scattering for the X-point and the conventional Thomson scattering technique for the divertor leg fit the reference divertor configuration with minimal impact on ITER design.

A wide bandpass double-grating polychromator with high rejection and high transmission has been designed and manufactured for laser-aided plasma diagnostics. The special mount utilizes subtractive dispersion in the second stage of the double polychromator such that the larger dispersion of the second stage is reduced by that of the first stage. This affects the intensity of the stray light background at the laser wavelength. The background at the edge of the laser line was measured at 10−5 of the light incident on the input slit. At the short end of the 200 nm bandpass, the stray light relative intensity approached 10−7.

The ITER tokamak requires diagnostics that on the one hand have a high sensitivity, high spatial and temporal resolution and a high dynamic range, while on the other hand are robust enough to survive in a harsh environment.In recent years significant progress has been made in addressing critical challenges to the development of spectroscopic (but also other) diagnostics. This contribution presents an overview of recent achievements in 4 topical areas:• First mirror protection and cleaning• Nuclear confinement• Radiation mitigation strategy for optical and electronic components• Calibration strategies

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

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.