The possibility of using the method of dual-wavelength digital holographic interferometry to assess the wear of protective elements of the Globus-M2 spherical tokamak after working plasma discharges is demonstrated. At this stage of the work, the protective elements were removed from the tokamak discharge chamber and used as samples in the holographic setup. A diagram of a holographic interferometer for recording primary holographic images is presented, in which control of the radiation wavelength recording and monitoring systems is carried out through a hardware and software complex in real time. The results of measurements of the shape of tokamak elements are presented. It is shown that when the difference in wavelengths changes, the sensitivity of the measurement method changes, and in the proposed configuration of the optical scheme it is possible to determine the minimum value of the shape change at a level of 10–30 μm. At the same time, the error in determining the phase difference, by which the surface profile is assessed, in the digital method can reach about 2π/40.

A diagnostic system for Thomson scattering of the central, edge and divertor plasma regions of a tokamak with reactor technologies is discussed. The rationale and choice of technical solutions are given, the composition of the Thomson scattering diagnostic complex is discussed, as well as an estimate of the accuracy of measuring the electron temperature and plasma density in the central edge and divertor regions of the TRT tokamak. Particular attention is paid to ensuring the functionality of the proposed diagnostics in the reactor mode of the tokamak operation and the results of testing diagnostic equipment in experiments on the Globus-M2 tokamak.

Thomson scattering of the core edge and divertor plasma regions of a tokamak with reactor technologies is discussed. The rationale and choice of technical solutions are given, the composition of the Thomson scattering diagnostic complex is discussed, as well as an estimate of the accuracy of measuring both electron temperature and density. Particular attention is paid to ensuring the functionality of the proposed diagnostics in the reactor mode of the tokamak operation and the results of testing diagnostic equipment in the experiments on Globus-M2 tokamak.

Mock-up of the system for remote monitoring of the accumulation of hydrogen isotopes in the walls of the Globus-M2 tokamak was assembled and tested. The measurements were performed using the LIA-QMS laser diagnostics (laser-induced ablation with registration using quadrupole mass-spectrometry). The data were obtained on the content of hydrogen isotopes in deposits appearing after exposing tungsten collectors to the loads in the volume of the Globus M tokamak. After testing the diagnostics under laboratory conditions, it was mounted at the Globus-M2 facility. In-situ measurements of the content of hydrogen isotopes in the graphite tiles of the tokamak divertor were performed. The possibility of combining the L-IA‑QMS diagnostics with the LIBS (laser-induced emission spectroscopy) diagnostics has been confirmed, in order to obtain information on the composition of the ablated material. In addition, the LIBS method was used for obtaining the deuterium/protium isotopic ratio during measurements in the Globus-M2 facility.

Options for implementing the IR thermography diagnostic system in a TRT facility are considered. Two variants of the optical scheme for measuring the temperature of the first wall and divertor targets are proposed: a wide-angle system combined with two divertor channels and a four-channel viewing system. The optical resolution of both systems and the levels of the collected signal are numerically simulated. On the basis of the calculations performed, conclusions are drawn about the compliance of the systems with the requirements for measuring the temperature of TRT plasma-facing elements. The sources of the temperature measurement error are considered, and the error caused by the reflection of radiation from structural plasma-facing elements by the surface under study is estimated. Calibration issues for IR thermography diagnostics are also discussed.

A conceptual design for diagnosing erosion of the first wall and divertor plates of a tokamak with reactor technologies TRT is proposed. The principles of constructing a diagnostic complex based on the following systems are developed: laser radar, dual-wavelength digital holographic interferometry and active laser IR thermography. An optical scheme is developed for combining the optical paths to input laser radiation and collect scattered light from diagnostic systems. To view the maximum area of the first wall, a scheme for optical scanning of the surface of the first wall and divertor is proposed. Based on optical simulation, the spatial distribution of the power density and phase of interferometry laser radiation in the illuminated region of the first wall is constructed, and the dimensions of the light fields and power density for IR thermography and laser radar diagnostics are determined. An image formation scheme is proposed and the spatial resolution is determined for interferometry and IR thermography methods. The light scattering function on models of the ITER divertor cladding is studied experimentally. The energy of the collected signal is calculated on the basis on the experimental data for all three diagnostic methods and the requirements for the diagnostic equipment are formulated.

The technique based on quadrupole mass spectrometry of the material ejected by laser-induced ablation - LIAQMS is proposed as a tool to measure H/D/T content in carbon co-deposits in the first wall and divertor regions of tokamak reactors. Tungsten tiles exposed in Globus-M2 tokamak were taken for validation experiments. Measurement accuracy was determined by comparison with the results obtained by laser-induced desorption - LID-QMS and conventional thermal desorption spectroscopy (TDS). Total amount of deuterium in hydrocarbon deposits was found to be 2.8 x 1017 D/cm2, when measured by - LIA-QMS, 3.0 x 1017 D/cm2 for LID-QMS and 2.9 x 1017 D/cm2 for TDS. The main uncertainty of D surface concentration are caused by inhomogeneity of the deposit thickness over the sample area or by presence of hydrocarbon species in the released gas.

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 laser-induced desorption (LID) and laser-induced ablation (LIA) methods are compared with each other regarding the possibility of measurements an absolute quantitative analysis of hydrogen isotopes content in first wall materials of fusion reactors. Deuterium containing tungsten films with a thickness of 300–400 nm on a silicon substrate were used as model samples. To implement the LID, the samples were irradiated with laser pulses with a duration of 200 microseconds and an energy density of 50–150 J/cm2, for LIA – 12 ns and 5–15 J/cm2. The registration of residual gases was carried out by quadrupole mass spectrometry. Computer simulation of laser pulse heating was performed for the LID process. The simulation results and experimental data showed that heating at an energy density of 100–150 J/cm2 is sufficient to degas tungsten films of the studied thickness. A comparison of the amount of desorbed deuterium in the LID (150 J/cm2) and LIA (15 J/cm2) modes shows that it is identical within the measurement error and is equal to 4.15±0.15·1014 cm-2.

In ITER, the first mirrors of optical diagnostics will be prone to deposition of the first wall materials which will result in the degradation of the mirror's optical parameters and the parameters of the diagnostic itself. To restore the mirror's reflectivity a cleaning system based on the gas discharge will be employed. The present paper discussed low-frequency power modulation of capacitively-coupled RF discharge utilized to decrease the mean RF power supplied through the in-vessel feeding line. The impact of power modulation on the incident ion flux, distribution of ion energy, and sputtering rates of the first wall materials were studied for the cases of DC-coupled and -decoupled mirrors with/without external magnetic field. The parameters of the plasma were measured depending on modulation frequency (1 Hz - 10 kHz) and duty cycle (10 - 90 %) for He plasma at a pressure of a few Pa. Based on the measurements of ion flux and ion energy the sputtered rates of Be and BeO were estimated. The estimations of the sputtering rate were followed by experiments on BeO sputtering in He plasma. Finally, the recommendations are given for selecting an appropriate RF power required to clean the mirror from Be/BeO deposits in ITER.

A project is proposed of the complex of diagnostics of the plasma-facing elements of the first wall and divertor of tokamak with reactor technologies (TRT). The main objectives of the complex are control of the integrity of plasma-facing elements, monitoring of the state of the first wall and the thermal flows onto the first wall and divertor targets as well as studies of the physical processes in the material of the first wall during its interaction with plasma. A review of expected effects of interaction of plasma with the plasma-facing elements is given; in the first place, with the tungsten facing of the divertor. Based on the analysis of the methods that were tested at existing tokamaks and methods that are being developed within the framework of the international ITER project, a complex of complementary diagnostics is proposed, which satisfies the TRT tasks. In the paper, a variant of implementation at TRT is discussed of frequency/amplitude modulated laser radar, two-wavelength digital holography, active (laser) and passive IR thermography, diagnostics of fuel build-up based on laser desorption and ablation, and the placement of witness samples. A variant is proposed of the placement of these diagnostics at TRT and their physical integration.

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.

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.

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.

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.