The sputtering of a number of materials due to an intense polyenergetic flux of hydrogen particles has been investigated. The irradiation of pure tungsten, copper, aluminium, titanium, aluminium-lithium alloys, stainless steel and tungsten-copper composition has been carried out at particle flux densities of 1017-1018 cm~2 s~' and at fluences of 1020-1022 cm~2. Furthermore, W-Cu composition has been subjected to the effect of high-current plasma pulses for simulating the disruption heat loads in a thermonuclear reactor.
We have examined conditions and parameters of irradiation of materials perspective for use in the mainstream nuclear fusion facilities for two devices of the Dense Plasma Focus (DPF) type (PF-6 and PF-1000) with a number of diagnostics in comparison with conditions expected in the first-wall materials in Iter and NIF. It is found that a so-called “damage factor” helps in modelling of the fusion reactor conditions. Optical microscopy, SEM, Atomic Emission Spectroscopy, images in secondary electrons and in characteristic X-ray luminescence of different elements, and X-ray elemental analysis, present results for a number of materials including low-activated ferritic and austenitic stainless steels, β-alloy of Ti, as well as the double-forged W (candidate material for divertor in Iter). With an increase of the power flux density of hot plasma and fast ion streams irradiating the surface, its morphology changes from a weak wave-like structure of the surface to the strongly developed one for the same material. It was melted with the appearance of the fracturing pattern – first along the borders of grains and then with the intergranular net of microcracks. At the highest values of power flux densities multiple blisters appeared. Besides, in this last case cracks develop because of microstresses at the solidification of melt. Presence of deuterium within the surface nanolayers of irradiated ferritic steel is explained by capture of deuterons in lattice defects of the types of impurity atoms, pores and oxycarbonitride particles presented in the material.
Both the plasma poloidal curvature and the curvature of the probing wave-front play an important role for the spectral resolution of microwave reflectometry. The propagation of electromagnetic waves in a cylindrical plasma with a refractive index monotonically decreasing towards its center leads to elongation of the wave-front (i.e. an increase of the radius of curvature near the axis of radiation). Such a phase-front alteration can result in a major modification of the reflectometry response regarding on high kθ fluctuations. At the same time, the curvature of the cutoff surface in cylindrical plasma is always less than the curvature of the cutoff density for an O-mode wave. For specific geometry of the TEXTOR tokamak we perform numerical two-dimensional full-wave reflectometry synthesizing. The response from poloidaly rotating coherent density perturbation was analyzed on the variety of the plasma cutoff curvature and density scale length, which affects the incident wave-front curvature. The simulations are shown that the increase of the incident wave front curvature extends the spectral resolution of conventional reflectometer with gain antennas. On the other hand, the effect from the cutoff curvature is found to be much weaker as compare to the wave-front effect.
Activity on projects of ITER and DEMO reactors has shown that solution of problems of divertor target plates and other plasma facing elements (PFEs) based on the solid plasma facing materials cause serious difficulties. Problems of PFE degradation, tritium accumulation and plasma pollution can be overcome by the use of liquid lithium–metal with low Z. Application of lithium will allow to create a self-renewal and MHD stable liquid metal surface of the in-vessel devices possessing practically unlimited service life; to reduce power flux due to intensive re-irradiation on lithium atoms in plasma periphery that will essentially facilitate a problem of heat removal from PFE; to reduce Zeff of plasma to minimally possible level close to 1; to exclude tritium accumulation, that is provided with absence of dust products and an opportunity of the active control of the tritium contents in liquid lithium. Realization of these advantages is based on use of so-called lithium capillary-porous system (CPS) – new material in which liquid lithium fill a solid matrix from porous material. The progress in development of lithium technology and also activity in lithium experiments in the tokamaks TFTR, T-11M, T-10, FTU, NSTX, HT-7 and stellarator TJ II permits of solving the problems in development of steady-state operating lithium divertor module project for Kazakhstan tokamak KTM. At present the lithium divertor module for KTM tokamak is under development in the framework of ISTC project # K-1561. Initial heating up to 200 °C and lithium surface temperature stabilization during plasma interaction in the range of 350–550 °C will be provided by external system for thermal stabilization due to circulation of the Na–K heat transfer media. Lithium filled tungsten felt is offered as the base plasma facing material of divertor. Development, creation and experimental research of lithium divertor model for KTM will allow to solve existing problems and to fulfill the basic approaches to designing of lithium divertor and in-vessel elements of new fusion reactor generation, to investigate plasma physics aspects of lithium influence, to develop technology of work with lithium in tokamak conditions. Results of this project addresses to the progress in the field of fusion neutrons source and fusion energy source creation.
The behaviour and erosion of tungsten, copper and W-Cu composition under irradiation by high intensive hydrogen plasma have been investigated. The erosion coefficients of these materials have been determined. The importance of copper redepositions in the mechanism of sputtering and erosion of W-Cu composition has been emphasised.
Experiments with lithium plasma facing components (PFCs) show promising results for the operation of hot plasma facilities and the general improvement of plasma parameters. The design and development of new tokamak plasma facing material (PFM) based on lithium capillary porous systems (CPS) are described in this paper.
The recent progress in the development of limiters with different kinds of CPS is relevant for protection of tokamak PFCs from damage under normal operation, ELMs and disruptions. New PFM eliminates the lithium flux into plasma, its pollution and lithium accumulation.
Here we present an overview of the design and the experimental tests of the liquid lithium limiters. These limiters are based on CPS with hard matrix from stainless steel mesh, molybdenum and tungsten. Different types of limiter have been taken into account: the horizontal and vertical rail type limiters with passive and active cooling for investigation the possibility to provide the closed lithium circulation in tokamak chamber; the ring CPS-based limiter for investigation of lithium behavior in limiter scrape-off layer (SOL).
Here we also present the preliminary results of the application of the cryogenic techniques for lithium removal from the chamber wall after operation in hot plasma.
The projects of ITER and DEMO reactors showed that there are serious difficulties with solving the issues of plasma facing elements (PFE) based on the solid materials. Problems of PFE can be overcome by the use of liquid lithium. Application of lithium will allow to create a self-renewal and MHD stable liquid metal surface of the in-vessel devices possessing practically unlimited service life. Realization of these advantages is based on use of so-called lithium capillary-porous system (CPS) – new material, in which liquid lithium fills a solid matrix from porous material. The progress in development of lithium technology and also lithium experiments in the tokamaks TFTR, T-11M, T-10, FTU, NSTX, LTX, HT-7 and stellarator TJ II is a good basis for development of the project of steady-state operating lithium divertor module for Kazakhstan tokamak. At present the lithium divertor module for KTM tokamak is development and manufacturing. The paper describes main design features of the module of lithium divertor (MLD). The first step of the hydraulic tests of MLD with fully assembled external thermo-stabilization system, which was connected to in-vessel lithium unit, were performed using ethanol as a model heat transfer media. Test results of MLD have shown that operating parameters of designed and manufactured system for thermo-stabilization are sufficient for proper operation; basic hydraulic characteristics of the system are close to expected values.