Piezoelectric motors designed for operation in the ITER tokamak-reactor must be tested for stability under severe radiation conditions. Properties of lead zirconate-titanate that is the most common type of piezoelectric materials were analyzed from the point of view on influence of radiation. It is shown that, at the expected in ITER level of radiation, this piezoceramics has a good potential for resistance to radiation induced amorphization and depolarization.

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.

As a part of ITER diagnostics design and development, ultrasonic and stick-slip piezoelectric motors and several types of piezoelectric ceramics were tested under severe neutron and gamma ray irradiation (-10(19) per cm(2) at E > 0.1 MeV) in a pool-type fission reactor. The fission neutron and gamma ray spectra were corrected to comply with conditions expected under ITER divertor cassettes. Taking into account possible neutron-induced amorphization of crystal structure, four types of piezoelectric ceramics were pre-selected. To restore the electronic structure of piezoelectric ceramic damaged by gamma rays, recovery polarization (i.e. heating while applying reverse bias) was suggested. The motors and piezoceramic samples were assessed for performance before and after irradiation as well as after recovery polarization. Recovery polarization was performed at 330 degrees C as it is expected close to the divertor cassettes during baking. Two most promising types of piezoelectric ceramics survived multiple cycles of depolarization/polarization at both 330 degrees C and 200 degrees C as appropriate for baking ITER divertor and vacuum vessel, respectively. All the experiments and procedures are described in detail.

Piezoelectric motors designed for operation in ITER tokamak-reactor must be tested for stability under severe radiation conditions. Properties of lead zirconate-titanate - the most common type of piezoelectric materials - were analysed from the point of view on influence of radiation. It is shown that at the expected in ITER level of radiation, this piezoceramic has a good potential for resistance to radiation amorphization and depolarization.

Studies of contaminants obtained by spraying ITER-grade Be on a quasi-stationary plasma gun facility QSPA-Be are presented. Contaminating films, consisting mainly of Be and O in approximately equal proportions, were deposited on substrates of quartz, sapphire, single crystalline silicon (SC-Si) and NaCl crystal. Characterization of the deposits was performed using SEM, XPS, EBS, AFM, SE, TEM&SAED and micro-interferometry showing polycrystalline BeO films similar as found in JET-ILW. Films on SC-Si and NaCl were used to characterize their composition and morphology. The cleaning rates in the 81.36 MHz RF discharges in He or D2 at 2 Pa were measured on the SC-Si target. The measured etching rate of the deposited films was several times higher than the rate calculated from the theoretical value of beryllium oxide sputtering yield. Test cleaning of these contaminants was carried out in a capacitively coupled RF discharge (CCRF) from the surface of sapphire and quartz plates, which were considered as a mock-ups of the protective window of the first mirror unit (FMU) designed for ITER divertor Thomson scattering diagnostic system (DTS).

Combined Thomson scattering (TS) and laser-induced fluorescence (LIF) diagnostics are being developed. The Thomson scattering and laser-induced fluorescence are laser diagnostics, with joint both probing and c-ollecting optical systems, which are the most complex and expensive parts of the diagnostic systems of large tokamaks, can be combined. Thomson scattering by free electrons is the use-proven diagnostic method for measuring profiles of important parameters of the plasma electron component (electron tempera-ture T e and electron density n e ), which requires a minimum of model assumptions. Almost all existing tokamaks are equipped with one or more TS systems, and by now, considerable experience has been accumulated in practical methods for implementation of these systems. The use of laser-induced fluorescence for measuring parameters of ion and neutral plasma components is less common, also because it requires knowledge of electron parameters to calculate populations of excited levels. The joint diagnostics of the Thomson scattering and laser-induced fluorescence in divertor plasma will be used to simultaneously measure the plasma parameters necessary for fundamental understanding physics of plasma detachment from divertor plates. These parameters are: the local parameters of plasma electrons ( T e , n e ), ion temperature ( T i ) measured from the data on emission of helium ions (HeII), as well as densities of helium atoms ( n a (He)) and hydrogen isotopes ( n a (H,D,T)). The measured parameters make it possible to calculate the following characteristics: (i) the ionization and recombination rates (using the data on T e , n e , T i , and n a (H, D, T)); (ii) the friction force of the plasma flow due to collisions with neutral particles (using the data on T i , n i (assuming n i = n e ), and n a (H, D, T)); and (iii) the pressure of the oncoming plasma flow (using the data on T e , n e , T i , and n i ). The article discusses advantages of combining laser diagnostics and ways of further development of the joint diagnostcs, based on the experience of creating similar diagnostics for domestic tokamaks and the similar diagnostics developed for ITER.

Transport and redeposition of plasma-facing materials in future fusion devices may have a critical influence on performance of in-vessel components. Evaluation of the impurity migration in diagnostic ports is important to predict degradation of in-vessel optics during the main plasma discharges as well as for development of in-situ cleaning maintenance. The Monte-Carlo code KITe was developed for modeling transport of neutral particles in weakly ionized plasma. The code employs advanced neutral particle collision model. The first KITe simulation and experimental validation of sputtered atom transport demonstrate performance of the proposed approach.

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.

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.

In plasma devices, in which plasma–wall interaction and material migration are significant, the diagnostic-window transmission may decrease due to the contamination of plasma-cleaning components by sputtered materials. We discuss the cleaning of KU-1 fused silica, which simulates the diagnostic window, from aluminum films in RF plasma generated in pure H2(D2) and Ne and in H2(D2)–0.23Ne mixtures. Aluminum is used as a chemical analogue of Be, which is the main material of the International Thermonuclear Experimental Reactor (ITER) first wall. The morphology of the plasma-treated surface is investigated by atomic force microscopy, the chemical composition is analyzed by X-ray photoelectron spectroscopy, and the transmission spectra in the range 400−1000 nm shows that plasma cleaning is accompanied by the slight reduction of quartz to suboxides and a simultaneous decrease in roughness Rq from 1.3 up to 1.0 nm. After plasma treatment of the quartz surface, a decrease in light transmission by 1.5–2% in the wavelength range 400–750 nm is observed. Further sputtering of the purified quartz surface with the removal of a layer with a thickness of more than 300 nm is accompanied by gradual smoothing of the surface and a decrease in Rq to 1 nm, but with the retention of reduced light transmission. All investigated gases, hydrogen isotopes, neon and D2(H2)–Ne mixtures, are suitable for removing Al films from the quartz surface at a RF power of several W/cm2 and temperatures of 20–100°C.

The effect of a cleaning discharge plasma on the KU-1 optical quartz surface topography and transmission in visible and near IR ranges (400-1000 nm) was studied. For the study, we used crossed DC and RF discharges in D2 and D2/N2 mixture, where the N2 fraction was ~25 mol %. The addition of nitrogen increased sputtering rate from 250 to 1200 nm / h while maintaining stoichiometry of the surface layers. The rms surface roughness decreased after the plasma exposure from 1.3 to 0.6 nm, while the quartz transmittance has not changed. The surface roughness analysis and calculation of diffusion scattered light at 400 nm transmitted through the quartz were performed using the power spectral density function of AFM.

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.

Laser-induced fluorescence (LIF) on ITER will be used for local measurement of helium density n(He) and ion temperature T-i 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 n(HeII) with laser parameters provided by available optical parametric oscillator pumped by Nd:YAG. New approach for measuring T-i is proposed and the expected errors are estimated using numerical simulations.

Divertor Thomson scattering (DTS) and laser-induced fluorescence (LIF) are both laser aided diagnostics well suited to combination with common probing and collecting optics that are the most sophisticated and expensive part of any ITER optical diagnostic system. The combination of DTS and LIF are used for simultaneous measurement of local electron (Te, ne), ion (Ti, nHeII) and atom (nHeI, nH(D,T)) parameters and provide basic information on rates of electron and ion processes to allow basic understanding of the physics of divertor plasma detachment. The measured parameters permit the calculation of rates of ionization and recombination using Te, ne, Ti, ni, nHeI and nH(D,T); emission intensity—Te, ne, ni, nHeI and nH(D,T); frictional force of the plasma flow due to collisions with neutrals—Ti, ni, T0, nHeI and nH(D,T) and pressure of the incoming plasma flow—Te, ne, Ti and ni. The paper discusses the benefits of DTS and LIF integration, suggests new approaches to the estimation of DTS capability, LIF implementation and possibilities for further diagnostic development.

Scientific and technological program of ITER requires measurement of plasma parameters using variety of diagnostics, ~ 20 of which are the optical ones [1]. The plasma emitted light and/or laser beams will travel by in-vessel optical labyrinths between torus and analyzing equipment. The closest to plasma optical elements, so-called first mirrors (FMs) and windows (FWs), will operate in harsh environmental conditions such as high temperature, particle fluxes and deposition of eroded first wall materials (Be, W etc.) [3]. During periods of Mirror Conditioning States (MCS), the optical performance of FMs and FWs has to be restored by an appropriate cleaning system. The system based on the radio frequency capacitively coupled (RFCC) discharge with a mirror or window used as a RF electrode is currently considered as the main approach [4]. However, most research of the plasma cleaning were performed using Be proxy materials (Al, Au etc.) due to its high toxicity. The study of sputtering Be deposits are of high importance to prove the plasma cleaning efficiency. The report presents results of pilot experiments on deposition and in-situ sputtering of Be deposits in the QSPA-Be facility located in Bochvar institute. QSPA-Be facility was designed to simulate plasma heat loads related to ITER ELMs / disruptions [5]. The pure Be target evaporation by one pulse of QSPA-Be provided deposition of ~7 nm contaminant layer with ~70% of beryllium. The beryllium films were characterized using scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy (EDX). The in-situ plasma cleaning was realized by exposing in 81 MHz helium discharge at pressure of 2 Pa. The cleaning process was controlled observing beryllium emission line in plasma column spectrum. The FW mock-up characterization was performed after 2 hours of the plasma exposure. Cyclic removal of Be-depositions in QSPA facility is planned to be done in 2020.

Recent research at three small tokamaks with different parameters located at the Ioffe Institute—the spherical tokamak Globus-M, the large aspect ratio tokamak FT-2 and the compact tokamak TUMAN-3M—are reviewed. This overview covers energy confinement (Globus-M and FT-2), L–H transition (TUMAN-3M and FT-2), Alfvén waves (Globus-M and TUMAN-3M), ion cyclotron emission (TUMAN-3M), major plasma discharge disruption (Globus-M) and scrape-off layer (Globus-M) studies. A full-f global gyrokinetic modeling benchmark using synthetic diagnostics in FT-2 is described. Anomalous absorption and emission in electron cyclotron resonance heating experiments due to the parametric excitation of localized upper hybrid waves are analyzed theoretically. Progress in the development of the neutral particle analysis, gamma-ray spectrometry and divertor Thomson scattering combined with laser-induced fluorescence diagnostics for ITER is discussed. The status of the new Globus-M2 spherical tokamak is reported.

Physical and engineering aspects of RF plasma cleaning developed for actively cooled first mirror serving as grounded DC-coupled electrode are under consideration. Feasibility of the water-cooling lines implemented as the coaxial notch filter for ITER in-vessel diagnostic mirrors was demonstrated for Capacitively Coupled Radio Frequency (CCRF) discharge in neon at 1-10 Pa. The impinging ions energy as a function of absorbed RF power was measured for both the DC-coupled and -decoupled schemes. Sputtering yield variation over rectangular surface of stainless steel mirror was studied experimentally. The measured sputtering yield profiles were found to be in qualitative agreement with the RF plasma numerical simulations. It was demonstrated that Al/Al2O3 used as Be/BeO proxy can be removed from the water cooled mirror surface by 100 eV neon ions with the use of the notch filter. The 14 h plasma exposure didn't affect the mirror reflectivity.

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.

Mirrors will be used as first plasma-viewing elements in optical and laser-based diagnostics in ITER. Deterioration of the mirror performance due to e.g. sputtering of the mirror surface by plasma particles or deposition of impurities will hamper the entire performance of the affected diagnostic and thus affect ITER operation. The Specialists Working Group on First Mirrors (FM SWG) in the Topical Group on Diagnostics of the International Tokamak Physics Activity (ITPA) plays an important role in finding solutions for diagnostic first mirrors. Sound progress in research and development of diagnostic mirrors in ITER was achieved since the last overview in 2009. Single crystal (SC) rhodium (Rh) mirrors became available. SC rhodium and molybdenum (Mo) mirrors survived in conditions corresponding to similar to 200 cleaning cycles with a negligible degradation of reflectivity. These results are important for a mirror cleaning system which is presently under development. The cleaning system is based on sputtering of contaminants by plasma. Repetitive cleaning was tested on several mirror materials. Experiments comprised contamination/cleaning cycles. The reflectivity SC Mo and Rh mirrors has changed insignificantly after 80 cycles. First in situ cleaning using radiofrequency (RF) plasma was conducted in EAST tokamak with a mock-up plate of ITER edge Thomson Scattering (ETS) with five inserted mirrors. Contaminants from the mirrors were removed. Physics of cleaning discharge was studied both experimentally and by modeling. Mirror contamination can also be mitigated by protecting diagnostic ducts. A deposition mitigation (DeMi) duct system was exposed in KSTAR. The real-time measurement of deposition in the diagnostic duct was pioneered during this experiment. Results evidenced the dominating effect of the wall conditioning and baking on contamination inside the duct. A baffled cassette with mirrors was exposed at the main wall of JET for 23,6 plasma hours. No significant degradation of reflectivity was measured on mirrors located in the ducts. Predictive modeling was further advanced. A model for the particle transport, deposition and erosion at the port-plug was used in selecting an optical layout of several ITER diagnostics. These achievements contributed to the focusing of the first mirror research thus accelerating the diagnostic development. Modeling requires more efforts. Remaining crucial issues will be in a focus of the future work of the FM SWG.