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The High Field Side Reflectometry (HFSR) is one of the ITER diagnostics which provides information about plasma state by measuring the electron density profile. Five various designs have been developed and studied since 2014 as a result of interaction of Peter the Great St. Petersburg Polytechnic University, NRC “Kurchatov Institute”, and Fusion Centre. Multiphysics engineering finite-element analysis has been performed for each of the HFSR designs. The study deals with thermal loads which were discovered to be the most significant from structural integrity point of view. On the basis of the obtained results the decisions on introducing structural changes into individual parts of the diagnostics were made by the designers of the Kurchatov Institute. Key features of the proposed mathematical model and results of numerical simulations, which demonstrate a stress state of the diagnostics, are given. The process of multistage diagnostic design development based on the finite element analysis results is presented.

The research is devoted to the application of parametric optimization methods to the design of the waveguide line of ITER High Field Side Reflectometry (HFSR) diagnostic system. This waveguide line in the Interspace and Port Cell zones, on the one hand, must provide enough flexibility of the structure to withstand interface mechanical loads, and on the other hand, must ensure low signal losses. In order to satisfy these two contradicting requirements, an optimization problem was formulated and solved. Two different parametric optimization algorithms were used in conjunction with MATLAB and ANSYS software to find a reasonable design of the HFSR waveguide. MATLAB was used to construct the bend path with the minimal signal losses, and ANSYS was used to calculate the mechanical response of the constructed waveguide to the applied loads.

The High Field Side Reflectometry (HFS-R) is diagnostic equipment subjected to the conditions that are severe even for ITER and make development of the appropriate design challenging: magnetic field over 9T, highly non-uniform temperature field with temperatures up to 700 °C, long in-vessel waveguides. A special global electromagnetic finite element model for performing poloidal field variation, toroidal field variation and halo current analyses for HFS-R was developed. The most dangerous for the equipment loads were identified by performing electromagnetic, thermal and stress analyses. Results indicate that current design of the HFS-R in-vessel elements withstands the possible electromagnetic loads, but thermal stresses above the allowable indicate the necessity of further design modifications and elaboration of computational approach for thermal analysis.

At TEXTOR, an O-mode heterodyne reflectometer system is installed and operated for the measurement of plasma density fluctuations and turbulence investigations. With two antenna arrays in the equatorial and top positions having two and three horn antennae, respectively, poloidal correlations are investigated under different plasma scenarios. From the amplitude, cross-phase and coherency spectrum, differences in the ohmic and auxiliary heated discharges are investigated. Furthermore the dynamic behaviour of the turbulence is studied in the SOC-IOC transition and in the precursor phase of a disruption. For the latter an increased integrated power spectral density was observed at the X-point of the mode compared with the O-point. Stationary m = 2 mode activity is observed for the first time at TEXTOR by reflectometry. The fluctuation level is calculated for different conditions and rises significantly increasing heating power which is consistent with the L-mode confinement degradation. Correlation measurements yield the measured phase delays which are used to calculate the poloidal phase velocity perpendicular to the magnetic field. In ohmic plasmas the turbulence rotates like a 'rigid body' with constant angular velocity inside the q = 2 surface. The rigid body rotation is broken up during tangential neutral beam injection. From the deduced poloidal wavenumber of the turbulence, most likely ion temperature gradient modes are the driving mechanism of the turbulence.

Advanced tokamak regimes obtained in ASDEX Upgrade, DIII-D, FT-U, JET, JT-60U, TCV and Tore Supra experiments are assessed both in terms of their fusion performance and capability for ultimately reaching steady-state using data from the international internal transport barrier database. These advanced modes of tokamak operation are characterized by an improved core confinement and a modified current profile compared to the relaxed Ohmically driven one. The present results obtained in these experiments are studied in view of their prospect for achieving either long pulses ('hybrid' scenario with inductive and non-inductive current drive) or ultimately steady-state purely non-inductive current drive operation in next step devices such as ITER. A new operational diagram for advanced tokamak operation is proposed where the figure of merit characterizing the fusion performances and confinement, H × βN / q 295, is drawn versus the fraction of the plasma current driven by the bootstrap effect. In this diagram, present day advanced tokamak regimes have now reached an operational domain that is required in the non-inductive ITER current drive operation with typically 50% of the plasma current driven by the bootstrap effect (Green et al 2003 Plasma Phys. Control. Fusion 45 587). In addition, the existence domain of the advanced mode regimes is also mapped in terms of dimensionless plasmas physics quantities such as normalized Larmor radius, normalized collisionality, Mach number and ratio of ion to electron temperature. The gap between present day and future advanced tokamak experiments is quantitatively assessed in terms of these dimensionless parameters.

Reasons for a decrease in recovery of lanthanides by sulfuric acid leaching from separate lots of phosphosemihydrate produced by Voskresenskie Mineral'nye Udobreniya joint-stock company were studied.

Factors were examined that determine passing of lanthanides into solution during nitric acid leaching of phosphohemihydrate obtained in the course of production of wet-process phosphoric acid from Khibiny apatite concentrate by the hemihydrate technique. The data obtained were compared to the results of sulfuric acid leaching.

High density plasmas (n0 ≈ 8 × 1020m-3) featuring steady improved core confinement have been obtained in FTU up to the maximum nominal toroidal field (8 T) by deep multiple pellet injection. These plasmas also feature high purity efficient electron-ion coupling and peaked density profiles sustained for several confinement times. Neutron yields in excess of 1 × 1013 n/s are measured, consistent with the reduction of the ion transport to neoclassical levels.

Small tokamaks may significantly contribute to the better understanding of phenomena in a wide range of fields such as plasma confiement and energy transport; plasma stability in different magnetic configurations; plasma turbulence and its impact on local and global plasma parameters; processes at the plasma edge and plasma‐wall interaction; scenarios of additional heating and non‐inductive current drive; new methods of plasma profile and parameter control; development of novel plasma diagnostics; benchmarking of new numerical codes and so on. Furthermore, due to the compactness, flexibility, low operation costs and high skill of their personnel small tokamaks are very convenient to develop and test new materials and technologies. Small tokamaks are suitable and important for broad international cooperation, providing the necessary environment and manpower to conduct dedicated joint research programmes. In addition, the experimental work on small tokamaks is very appropriate for the education of students,...

An overview of the Frascati tokamak upgrade (FTU) results during the period 2000-2002 is presented. Long duration internal transport barriers (ITBs) have been obtained on FTU with combined injection of lower hybrid (LH) and electron cyclotron (EC) waves in 5 T/0.5 MA discharges. The ITB phase lasts about 10 energy confinement times and is characterized by an energy confinement time up to 1.6 times the ITER97 L-mode scaling. Temperatures up to I I keV are measured at 0.9 x 10(20) m(-3) central density. ITB studies using IBW injection have also been continued up to 8 T/0.8 MA. The LH system has operated at full power allowing us to complete the current drive (CD) studies at ITER-relevant densities. At these density values the electrons and ions are coupled and an increase in the ion temperature is clearly observed. A preliminary sign of enhanced CD efficiency has been obtained in the combined injection of EC and LH waves at magnetic field values lower than the resonant field for thermal EC absorption. Pellet optimization studies have been performed in order to test the conditions under which a quasi-steady-state confinement improvement can be obtained and impurity accumulation can be avoided. The confinement time in Ohmic discharges is generally in agreement with the ITER97 L-mode scaling. Transient confinement improvement is observed for a duration less than one energy confinement time. Radiation improved mode studies have been started thanks to the recently inserted boronization system, which has allowed reduction of the radiated power. Confinement improvement with neon injection has been observed in 6 T/0.9 MA discharges. Transport studies on profile stiffness and magnetohydrodynamic (MHD) studies of fast reconnection and snakes will also be presented.

Our understanding of the physics of internal transport barriers (ITBs) is being advanced by analysis and comparisons of experimental data from many different tokamaks worldwide. An international database consisting of scalar and two-dimensional profile data for ITB plasmas is being developed to determine the requirements for the formation and sustainment of ITBs and to perform tests of theory-based transport models in an effort to improve the predictive capability of the models. Analysis using the database indicates that: (a) the power required to form ITBs decreases with increased negative magnetic shear of the target plasma, and: (b) the E x B flow shear rate is close to the linear growth rate of the ion temperature gradient (ITG) modes at the time of barrier formation when compared for several fusion devices. Tests of several transport models (JETTO, Weiland model) using the two-dimensional profile data indicate that there is only limited agreement between the model predictions and the experimental results for the range of plasma conditions examined for the different devices (DIII-D, JET, JT-60U). Gyrokinetic stability analysis (using the GKS code) of the ITB discharges from these devices indicates that the ITG/TEM growth rates decrease with increased negative magnetic shear and that the E x B shear rate is comparable to the linear growth rates at the location of the ITB.