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At present, the Kazakhstan Tokamak for Material testing (KTM) project is in progress. The optimization of innovative design elements of the divertor is carried out at the KTM, in addition to studying materials applied for fusion technologies. The design of the lithium divertor module based on lithium capillary porous systems has been developed and tested successfully in real tokamak conditions. In this design, the problem of removing high-density heat flows is solved by using liquid metal coolant based on the Na–K eutectic alloy. The requirements for the improved safety and the divertor design compatibility with other in-vessel water-cooled elements of the tokamak and temperature limitation of the receiving lithium surface to the level

This paper presents the results of the experiments on hydrogen isotopes interaction with lithium capillary-porous systems (CPS) under reactor irradiation, when sample’s temperatures were from 180 to 550 °C. A special ampoule device with a tubular sample of lithium CPS was developed for experiments. The IVG.1 M reactor (Kurchatov) was used as a source of radiation. The experiments were carried out by the method of dynamic sorption: a constant flow of deuterium was swept through the tube sample with constant degassing and fixation of gas composition in the experimental ampoule device. The paper presents initial data on temperature dependencies of tritium release from lithium CPS under reactor irradiation. The initial analysis showed some features of tritium release: insignificant tritium release in solid lithium (samples temperature below 180 °C), decrease of tritium release under temperature of cell’s center ˜ 250℃, stop of tritium release under temperature of cell’s center ˜ 300℃; then tritium release increased with increase of temperature. The conclusion was made that tritium capture by lithium has a significant influence on the processes of tritium release from lithium CPS under reactor irradiation. The work was performed within the framework of the ISTC #K-2204 project.

Capillary-porous systems (CPS) with liquid lithium are considered as a prospective alternative to traditional structural materials (W, Be, CFC) during the production of in-vessel elements contacting plasma of stationary thermonuclear reactors. The behavior of CPS with liquid lithium under the impact of pulsed deuterium plasma flows is considered. The basic processes determining the high stability of CPS against damage are revealed and the critical parameters defining the CPS resistance are specified. The behavior of the tungsten-lithium CPS after its interaction with atmospheric gases is studied in experiments on a plasma facility at the Ioffe Physical-Technical Institute. The irradiated samples of the CPS were studied by optical and scanning electronic microscopy and local X-ray spectral and X-ray diffraction analyses. The surface temperature was measured using a bicolored pyrometer during deuterium plasma irradiation. The outcome of the plasma effect on the heavily oxidized CPS surface substantially differs from that for a pure surface: there is damage to the CPS structure. The specific power of the plasma flow was 22–41 GW/m2, and the number of pulses per irradiated target was 100.

Prototypes of lithium emitters and collectors designed on the basis of capillary-porous systems (CPS) are being tried-out at the T-11M tokamak within the concept of continuous lithium circulation, previously proposed for stationary fusion neutron sources (FNSs). One of the goals of the T-11M tokamak research program are to implement the closure of the lithium circulation circuit and to develop the technology of recuperation (extraction and removal from the tokamak chamber) of hydrogen isotopes by using special lithium CPS-based collectors in order to prevent hydrogen isotope accumulation on the inner surface of the discharge chamber. In order to determine the optimal temperature regime for the future FNS operation, the effect of the temperature of the collector surface on its ability to capture lithium and hydrogen isotopes was studied. It is ascertained that the efficiency of capturing lithium and hydrogen isotopes by metal (12Cr18Ni10Ti) collectors in the operating regimes of the T-11M tokamak depends on the temperature of their collecting surfaces; namely, it remains almost constant in the temperature range from–196 to 50°С and then gradually decreases as the temperature rises to 300–400°С. In this case, the amount of the collected lithium is reduced by no more than sixfold, whereas the amount of collected hydrogen isotopes decreases by more than two orders of magnitude. Thus, the wall of a tokamak reactor chamber, which will be heated to 400°С but still coated with residual lithium (or its chemical compounds), will serve as a “mirror” for the incident hydrogen isotopes. It is found that the liquid lithium surface of the CPS-based collector can efficiently capture hydrogen isotopes falling onto it. As the lithium CPS-based collector is heated from 100 to 240°С, the amount of captured deuterium is reduced only twofold. This means that, in a steady-state mode, if, e.g., an MHD pump is used to transport liquid lithium from the collector back to the emitter, then, together with lithium, the captured hydrogen isotopes can also be transported into the recuperation zone.

The experience gained in developing, creating, and testing prototypes of stationary in-vessel devices based on capillary-porous systems with liquid lithium for the T-11M, T-10, KTM, and FTU tokamaks is presented. A promising scheme is proposed for designing liquid-metal-based in-vessel components with heat removal for thermonuclear reactors of the DEMO type and thermonuclear neutron sources.

The introduction of lithium as a material for production of in-vessel plasma-facing elements of the tokamak necessitates the development of the corresponding diagnostic instruments. A series of diagnostic devices have been developed and fabricated for tokamaks T-10 and T-11 M which make it possible to investigate the processes of lithium transport in the tokamak plasma scrape-off layer (SOL), real-time dynamics of lithium deposition at various temperatures of a collecting surface by means of the microbalance technique, adsorption/desorption process of hydrogen isotopes on the lithium surface, and influence of an electric field on lithium trapping. Scanning of plasma parameters is provided by Langmuir probes. Such devices can be used to extract the lithium deposited on the inner walls of the tokamak vacuum chamber without opening it. For these purposes, a lock chamber and bellows-free vacuum input allowing movement and rotation is provided. It is planned to perform the study of the tokamak plasma interaction with in-vessel lithium-based elements by means of infrared (IR) thermometry. It is planned to try out this technique on the Т-11М tokamak using a special device based on a IR camera.

The feasibility of getting “D” shaped plasma configurations in FTU, with a possible X point close to the first wall, has been investigated with the aim of achieving an H-mode regime in this machine. This regime could allow both evaluating the thermal effects on the liquid lithium limiter due to the possible Edge Localized Modes and studying the L-H transition properties in low recycling conditions due to the presence of lithium.. An alternative design for the magnetic system in FTU has been also proposed, to realize an X-point inside the plasma chamber, close to the Liquid Lithium Limiter.Preliminary experiments with elongated configurations and limited ECRH additional heating power did not allowed approaching the L-H transition but they were used to develop a proper elongation control. This controller allowed guaranteeing the vertical stability in elongated configurations despite the reduced power available for the horizontal field coils in FTU. The elongation was stably keep over 1.2, while the lithium limiter was very close to the last close flux surface. Neither limiter damages nor plasma pollution were observed in these configurations.A possible alternative connection of the poloidal field coils in FTU is here proposed, with the aim of achieving a true X-point configuration with a magnetic single null well inside the plasma chamber and strike points on the lithium limiter. A preliminary assessment of this design allowed estimating the required power supply upgrade and showed its compatibility with the existing mechanical structure and cooling system, at least for plasmas with current up to 300 kA and flat-top duration up to 4s. Keywords: FTU, Liquid lithium limiter, L-H transition, X-point, Plasma elongation

This paper describes the results on neutron irradiation of liquid lithium saturated with deuterium at the IVG.1 M research reactor. The neutron flux at the reactor core center at 2 MW was 5 10−13 cm−2 s−1. The efficiency coefficients of helium and tritium release from lithium saturated with deuterium were calculated. The tritium interaction with lithium atoms (formation and dissociation of lithium tritide) has an effect on tritium release. An increment of sample’s temperature results in tritium release acceleration due to rising of the dissociation rate of lithium tritide. At temperatures below 573 K the efficiency coefficient of tritium release is well described by the expression K = 0.015 exp(−14/RT), and above 623 K − K = 109 exp(-144/RT). The T2 molecules contribution into the overall tritium release becomes apparent at temperatures higher than 673 K and increases with the temperature rise.

The paper describes the analysis of liquid lithium interaction with materials based on carbon, the manufacture technology of capillary-porous system (CPS) matrix on basis of CNT-reinforced carboxylic fabric. Preliminary study of carboxylic fabric wettability with liquid lithium is presented. The development of technology includes: microstructural studies of carboxylic fabric before its CNT-reinforcing; validation of CNT-reinforcing technology; mode validation of CVD-method for CNT synthesize; study of synthesized carbon structures. Preliminary studies of carboxylic fabric wettability with liquid lithium consist in carrying out of experiments at temperatures 673, 773 and 873 К in vacuum during long time. The scheme of experimental device for manufacturing of lithium CPS and matrix filling procedure with liquid lithium are presented. The concept of lithium limiter with CPS on basis of CNT-reinforced carboxylic fabric is proposed.

In the framework of the liquid lithium limiter experiment in Frascati a new auxiliary system was developed in order to provide a better control of the energy fluid vector. The cooled liquid lithium system (CLL) was installed for the first time at the end of 2013, it uses overheated water to heat the lithium and to extract, at the same time, the heat from the metal surface when it gets wet by the plasma. A first version of the system, developed and presented in previous papers, has been modified to optimize the heat flux measurement on the liquid lithium surface. The changes include a new power supply logic for the heating system, new sensors and new read-out electronics compatible with the implementation of a real time control system. The prototype was updated with the aim of achieving a low cost and versatile control system.

Deuterium release from Li-D films co-deposited on a Mo substrate at room temperature in magnetron discharge was investigated by means of thermal desorption spectroscopy. The deuterium concentration in the films was estimated to be D/Li = (14 ± 4)%. TDS from Li-D films just after co-deposition had a sharp peak at 670–710 K. Exposure of deposited Li-D films in the air at room temperature led to deuterium release. Comparison of release in air, water vapor, nitrogen, and oxygen demonstrated that water plays a major role in deuterium release in the air at low temperatures.

This paper describes the main parameters of tritium generation and release from lithium capillary-porous system (CPS) under neutron irradiation at the IVG.1 M research reactor. The experiments were carried out using the method of mass-spectrometric registration of released gases and using a specially constructed ampoule device. Irradiation was carried out at different reactor thermal powers (1, 2 and 6 MW) and sample temperatures from 473 to 773 K. In the experiments a very small tritium release was detected likely due to its high solubility in liquid lithium. It can be caused by formation of lithium tritide during tritium diffusion to the lithium surface.

Power handling experiments with a special liquid metal divertor module based on the capillary porous system technology were performed in the tokamak COMPASS. The performance of two metals (Li and LiSn alloy) were tested for the first time in a divertor under ELMy H-mode conditions. No damage of the capillary mesh and a good exhaust capability were observed for both metals in two separate experiments with up to 12 MW/m(2) of deposited perpendicular, inter-ELM steady-state heat flux and with ELMs of relative energy similar to 3% and a local peak energy fluence at the module similar to 15 kJ.m(-2). No droplets were directly ejected from the mesh top surface and for the LiSn experiment, no contamination of the core and SOL plasmas by Sn was observed. The elemental depth profile analysis of 14 stainless-steel samples located around the vacuum vessel for each experiment provides information about the migration of evaporated/redeposited liquid elements.

Lithium capillary-porous system (CPS) as a mat between molybdenum grids fixed in a molybdenum module was tested in PLM plasma device for 3 hours in a steady state helium discharge with plasma parameters relevant to scrape-off-layer and divertor plasma of large tokamak. The plasma load on the CPS was of 1 MW / m2 led to surface heating up to 700 °C and vaporization of lithium to plasma. Lithium ions was registered by optical emission spectra from plasma. The redeposition of vaporized lithium was detected on the PLM in-vessel components. Inspections of CPS modules after plasma irradiation did not reveal any damage of molybdenum grids.

The PLM plasma device was constructed for high heat flux tests of fusion plasma-facing materials and in-vessel components of a fusion reactor. The ITER-grade tungsten samples were irradiated with steady-state plasma in PLM. The combined tests of ITER-grade tungsten samples with an e-beam load of 40 MW/m2 and stationary steady-state plasma load of ∼1 MW/m2 led to erosion, cracking, and nanostructured “fuzz” structure growth on the material surface. The capillary porous system of liquid tin was tested with steady-state plasma in PLM during ∼200 min demonstrating sustainability under high heat plasma load. Lithium materials deposited in the T-10 tokamak during experiments with lithium capillary-porous system have been irradiated with stationary steady-state plasma in PLM to test the evolution of the deposits under long-term plasma load.

Development of plasma facing material (PFM) based on the Capillary-Porous System (CPS) with lithium and activity on realization of lithium application strategy are addressed to meet the challenges under the creation of steady-state tokamak fusion reactor and fusion neutron source. Presented overview of experimental study of lithium CPS in plasma devices demonstrates the progress in protection of tokamak plasma facing components (PFC) from damage, stabilization and self-renewal of liquid lithium surface, elimination of plasma pollution and lithium accumulation in tokamak chamber. The possibility of PFC protection from the high power load related to cooling of the tokamak boundary plasma by radiation of non-fully stripped lithium ions supported by experimental results. This approach demonstrated in scheme of closed loops of Li circulation in the tokamak vacuum chamber and realized in a series of design of tokamak in-vessel elements.

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.

At present time the project of Kazakh material science tokamak KTM is under implementation. In addition to solving problems in the study of materials for fusion technologies on tokamak is working out innovative design elements of the divertor. The design of the lithium divertor module based on lithium capillary-porous systems has been developed and successfully tested under real tokamak conditions. In this design, the solution to the problem of removal of heat flows of high density was implemented through the use of a liquid metal coolant based on the eutectic alloy Na-K. Requirements to improve the safety and compatibility of the design of the divertor with other in-vessel elements of the tokamak, cooled by water, and limiting the temperature of the lithium receiving surface at 2 led to the creation of a new design solution for the experimental module of the divertor and the use of a new coolant – gas dispersed water flow (gas-water spray). The article describes and justifies the design solutions of the new version of the module, the parameters of the coolant and the cooling system scheme. Experimental results on determination of heat transfer coefficient for coolant based on gas-water spray are considered

Plasma tests of a tin capillary-porous system (CPS) were carried out on the PLM plasma device. This CPS is made similar to the lithium CPS used in experiments in tokamaks T-10, T-11M. The CPS system was immersed in the PLM plasma device and tested for 3 hours in a stationary helium discharge with plasma load of 1 MW / m2. The tin CPS has not damaged after such plasma load suggesting it as plasma facing component in fusion reactors and plasma engines.

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. © 2013 Elsevier B.V.

The T-11M lithium program is focused to find solutions for technological issues of a steady-state tokamak. A recent lithium activity in T-11M was focused on three directions: investigation of different Li-limiters, investigations of long-term effects of lithium behavior in hydrogen plasmas and development of a new kind of CPS limiters. The new so-called R limiter has been used as Li collector for investigation of lithium fluxes near the plasma boundary and chamber walls. It was shown, that the Li circulation near the limiter exceeds up to four times the primary lithium flux from limiters. 90% of primary flux is collected by lateral sides of limiter and only 10% arrives at the chamber wall. The deuterium glow discharge was used to test the long-term lithium degradation under the deuterium bombardment. It was shown that such degradation is small. That means that the Li-limiter can be used as Li-emitter in steady-state tokamak operation. A new vertical lithium limiter was successfully tested in T-11M.

Liquid metal infused capillary porous structures (CPSs) are considered as a potential divertor solution for DEMO due to their potential power handling capability and resilience to long term damage. In this work the power handling and performance of such Sn-based CPS systems is assessed both experimentally and via modelling. A Sn-CPS target was exposed to heat fluxes of up to 18.1 MW m−2 in He plasma in the Pilot-PSI linear device. Post-mortem the target showed no damage to nor any surface exposure of the underlying W-CPS felt. The small pore size (∼40µm) employed resulted in no droplet formation from the target in agreement with calculated Rayleigh-Taylor and Kelvin-Helmoholtz instability thresholds. The temperature response of the Sn-target was used to determine the thermal conductivity of the mixed Sn-CPS material using COMSOL modelling. These values were then used via further finite element analysis to extrapolate to DEMO relevant monoblock designs and estimate the maximum power handling achievable based on estimated temperature windows for all component elements of the design. For an optimized design a heat-load of up to 20 MW m−2 may be received while the use of CPS also offers other potential design advantages such as the removal of interlayer requirements. Keywords: Liquid-metal, DEMO, CPS, Tin, Power-exhaust, Plasma-wall interaction

The liquid Sn in a matrix of Capillary Porous System (CPS) has a high potential as plasma facing material in steady state operating fusion reactor owing to its physicochemical properties. However, up to now it has no experimental confirmation in tokamak conditions. First steady state operating limiter based on the CPS with liquid Sn installed on FTU tokamak and its experimental study is in progress. Several aspects of the design, structural materials and operation parameters of limiter based on tungsten CPS with liquid Sn are considered. Results of investigation of corrosion resistance of Mo and W in Sn and their wetting process are presented. The heat removal for limiter steady state operation is provided by evaporation of flowing gaswater spray. The effectiveness of such heat removal system is confirmed in modelling tests with power flux up to 5 MW/m2.

a b s t r a c t Stellarator concept is considered as a promising approach for power fusion reactor development because it basically free from disruptions and other extreme thermal load events. However, the potential problem of impurity accumulation in stellarator plasma should be taken into account. Very promising results in density control, plasma reproducibility and confinement characteristics have been obtained with appli- cation of "lithiation" technology. The next step in the improvement of TJ-II Heliac plasma performance is the development and creation of two poloidal liquid lithium limiters (LL). Experimental possibilities, design, structural materials and main parameters of LL based on capillary-pore structure (CPS) filled with liquid lithium are considered. Understanding in hydrogen isotope interaction with liquid lithium surface is an important aspect of lithium technology development for fusion reactor application. Therefore study of deuterium sorption/desorption process on a lithium surface of LL is stipulated. The development of lithium CPS based devices decreasing intensity of plasma-wall interaction on a central "groove" of TJ-II vacuum camera is proposed as the further step in plasma performance improvement owing to decrease in impurity flux from the wall.

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.

Capillary-pore systems (CPS) with liquid metals are considered as advanced plasma facing material for application in DEMO-type fusion reactor. The estimation of opportunity of liquid Li, Ga and Sn application is carried out on the basis of its physical, chemical and technological properties, and with respect to prospective design of the tokamak in-vessel elements and technology. It has been shown that Li now is the most attractive and most investigated liquid metal for fusion devices application with CPS. The temperature limit for normal operation is about 550 °C and determined by appropriate Li flux to plasma due to evaporation. Wide range of structural materials is appropriate for Li based in-vessel elements. Ga and Sn are very corrosive and embrittlement inducing metals. As a result the temperature limit of these application is determined by compatibility with structural materials of CPS and in-vessel element. Only W can be used with Ga and Sn up to 500 °C. Moreover these metals have lower thermal properties comparing to Li. Surface temperature analysis for possible in-vessel element design (1 mm thick of porous W based CPS) has shown the similar power flux limit ∼21 MW/m2 for Li, Ga and Sn application at normal operation. Taking into account the latent heat of vaporization and screening effect with re-radiation the CPS with Li has a priority at ELM and disruption conditions.

The following critical issues of liquid lithium used in tokamak conditions are considered: major physical properties of lithium, physico-chemical aspects of lithium interaction and compatibility with structural materials of fusion reactors. Lithium capillary-porous system (CPS) is considered as advanced plasma facing material for power fusion reactor and its main properties are presented. Review of plasma facing element (PFE) structures based on lithium CPS and tests results in T-11M, T-10 and FTU tokamaks are included. Brief review of projects of lithium limiter of FTU with active system for thermal stabilization and module of lithium divertor for KTM tokamak with liquid metal (Na–K) cooling system based on the lithium CPS use are presented.

A short review of the modern state of lithium use in experimental facilities and fusion devices is presented. This paper is intended to be used by the engineers and researchers involved in activiti...

A short analysis of lithium properties presented in this article was carried out in terms of lithium application in various systems of a fusion reactor. The article is intended for use by the engineers and researchers involved in activities on thermonuclear fusion.

A liquid lithium limiter (LLL) with a total area of 170 cm 2 and lithium amount of ≅80 g based on lithium filled capillary-pore system (CPS) as an advanced plasma facing material has been tested for the first time on the medium size high magnetic field tokamak FTU. The main technological aims of the experiment were to test: the compatibility of lithium CPS in real plasma tokamak conditions; the capacity of CPS to self-regenerate an exposed surface and to confine lithium during normal plasma operations and disruptions; the capability to withstand high heat loads without damage of limiter surface. Lithium CPS advantages, experimental base for the tokamak application, general description of LLL design and preparation method were described. No anomalous phenomenon like “lithium bloom” has occurred in the plasma and no damage of CPS surface and LLL structure has been observed after ∼60 discharges with power flux up to 5 MW/m 2 , as well as on plasma disruptions. Self-restoring of lithium surface has been confirmed. LLL application as a conditioning element with depositing a lithium film on the walls (litization) has resulted in radiation losses, plasma contamination and D 2 recycling reduction. Perspectives of lithium CPS application for further fusion reactors have been considered.

A liquid lithium limiter (LLL) with capillary porous system has been tested for the first time on the high field medium size tokamak, FTU. Lithium acts as a first wall material in the liquid phase and as a conditioning technique by depositing a lithium film on the walls (lithization). Thermal loads exceeding 5 MW/m 2 have been so far applied to the LLL surface during plasma discharges: no anomalous Li influx, like ‘lithium bloom’, occurs and no surface damage is observed, even after plasma disruptions. Radiation losses, plasma contamination and working gas recycling are reduced after Li coating of the wall as for boronization but with better results. A large electron temperature increase (∼50%) in the scrape-off layer occurs that is well reproduced by the simulation of 2D code TECXY. The Greenwald density limit is easily reached and even exceeded in the explored plasma current ranges ( I p = 0.50–0.9 MA).

The paper is an overview of recent results of Li limiter testing in T-11M tokamak. The lithium limiter is based on the capillary-pore system (CPS) concept. The Li erosion process and deuterium (D2) and helium (He) sorption by Li first wall were investigated. The ability of capillary forces to confine the liquid Li in the CPS limiter during disruption was demonstrated. The idea of combined lithium limiter with thin (0.6 mm) CPS coating as a solution of the heat removal problem was realized. As a result the quasi steady-state tokamak regime with duration up to 0.3 s and clean (Zeff= 1) deuterium plasma has been achieved. The temporal evolution of the lithium surface temperature during discharge was measured by a IR radiometer and then was recalculated to the surface power load. For the estimation of the Li limiter erosion the Li neutral and ions spectral line emission were observed. The increase in lithium erosion as a result of limiter heating was discovered. The radial distribution of plasma column radiation measurements showed up to 90% of the total radiation losses in a relatively thin (5 cm) boundary layer and only 10% in a plasma centre during discharges with high Li influx. Oscillations of Li emission and saw-tooth-like oscillations of the limiter surface temperature have been detected in discharge regimes with highest Li limiter temperature (>600 °C). A version of Li CPS first wall of DEMO reactor and Li CPS limiter experiment in the International Thermonuclear Energy Reactor are suggested.

MHD pressure drop in flowing liquid metal for a tokamak with high magnetic fields is a key concern regarding the development of lithium self-cooled test module for ITER and lithium breeding blanket for DEMO-type projects. MHD losses on liquid metal pumping can be most efficiently reduced by applying of an electrically insulating coating to the inner surface of the channels. The developed method to evaluate the electrically insulating properties of a coating placed on the internal surface of a vanadium test channel is based on the pressure drop measurement in a liquid lithium forced circulation system with the test section in magnetic field. The tests were conducted on channels of V–4Ti–4Cr alloy with and without on insulating coating based on the aluminum nitride technique (AlN) [1] . The dimensions of the test section were 6 mm × 20 mm × 320 mm. The tests involved lithium flow velocities up to 3 m/s, temperatures up to 400 °C and a uniform transverse magnetic field up to 1.6 T. The measured hydraulic resistances have shown five times reduction for the coated wall channel in comparison with the conducting wall channel.

Recent experimental results on beryllium corrosion behavior in a V–4Ti–4Cr alloy, liquid lithium static system during testing for 200–500 h at temperatures from 600 to 800 °C are presented. The influence of test conditions (temperature, duration and lithium purity) and beryllium characteristics (microstructure, grain size and chemical composition) on weight loss of beryllium and penetration of lithium into beryllium are discussed. Results of compressive tests for beryllium specimens before and after corrosion testing are also introduced.

The possibility of using liquid lithium as plasma facing component for the divertor target plates will be investigated on FTU by employing an innovative solution that exploits the surface tension forces in capillary channels to compensate j × B forces induced in lithium. For this structure, a high stability and resistance is foreseen as well as an intrinsic capability to self-regenerate the plasma facing surface. FTU represents the first very good opportunity to test this configuration, called capillary porous system (CPS), in an ITER relevant experiment at high plasma density (up to 3.2 × 10 20 m −3 ), high current (up to 1.6 MA) and high magnetic field (up to 8 T). As first step of its application on FTU, foreseen for the second half of 2005, CPS system will be employed for a “litization” experiment aimed to reduce plasma contamination and recycling.

At present the most promising principal solution of the divertor problem appears to be the use of liquid metals and primarily of lithium Capillary-Pore Systems (CPS) as of plasma facing materials. A solid CPS filled with liquid lithium will have a high resistance to surface and volume damage because of neutron radiation effects, melting, splashing and thermal stress-induced cracking in steady state and during plasma transitions to provide the normal operation of divertor target plates and first-wall protecting elements. These materials will not be the sources of impurities inducing an increase of Zeff and they will not be collected as dust in the divertor area and in ducts. Experiments with lithium CPS under simulating conditions of plasma disruption on a hydrogen plasma accelerator MK-200 [~(10 - 15) MJ/m2, ~50 μs] have been performed. The formation of a shielding layer of lithium plasma and the high stability of these systems have been shown. The new lithium limiter tests on an up-graded T-11M tokamak (plasma current up to 100 kA, pulse length ~0.3 s) have been performed. Sorption and desorption of plasma-forming gas, lithium emission into discharge, lithium erosion, deposited power of the limiter are investigated in these experiments. The first results of experiments are presented.

This paper is a review of the experimental investigations of Li-behavior as limiter material in real tokamak. All experiments were performed in tokamak T-11M with main parameters: plasma current, Ip=100 kA; duration of the discharge 0.1 s, toroidal magnetic field, B=1 T; major radius, R=0.7 m; minor radius, a=0.19–0.23 m; average electron density, ne∼(1–5)×1019 m−3 and electron temperature, Te(0)∼0.3–0.5 keV. Two moving limiters with similar geometry were installed for comparison in chamber—the conventional graphite-boron limiter and Li-limiter on the basis of capillary pore system (CPS). The Li-influx to plasma was measured by visible, UV and SX plasma emission. The main experimental results were: (a) no dramatic bursts of lithium injection at heat load close to the tokamak-reactor level, ∼10 MW/m2, were observed, (b) the total lithium erosion from limiter can be explained by deuterium plus lithium ions sputtering (self-sputtering), (c) high lithium radiation during disruptions prevent Li-limiter from high power load, and (d) the solid basis of CPS limiter had no damages after more than 200 shots with disruptions. The main effect of lithium use in T-11M was the rise of the first wall getter properties, i.e. recycling reduction for not only H2 and D2 but for He too. The temperature of hydrogen isotopes desorption was 350–400 °C and He desorption was 50–100 °C.

A liquid metal surface made with a capillary porous structure (CPS) (solid base) filled with lithium (liquid) has been offered for application in a magnetic confinement fusion reactor. The approach is investigated experimentally for divertor and first wall relevant conditions. The CPS ensured stability of the liquid surface under pulsed plasma impact in disruption simulation and tokamak experiments. Continuous operation of lithium capillary target was studied under electron beam load in stationary thermal conditions in the range 1–10 MW/m 2 of energy flux in steady state. Lithium evaporation was shown to dominate at temperatures higher than 400 °C and it removed up to 0.7 of incident power. Heat flux redistribution at the liquid lithium surface was analyzed. Lithium ionization, lithium plasma parameters near the liquid surface were evaluated. The importance and possibility of prompt lithium removal from the near surface layer in divertor conditions are emphasized.

In order to prevent the overheating of the liquid Li surface and the consequent Li evaporation for T > 500°C, an advanced version of the liquid lithium limiter has been realized and installed on FTU. This new system, named Cooled Lithium Limiter (CLL), has been optimized to demonstrate the lithium limiter capability to sustain thermal loads as high as 10 MW/m2 with up to 5 s of plasma pulse duration. The CLL operates with an actively cooled system with water circulation at the temperature of about 200°C, for heating lithium up to the melting point and for the heat removal during the plasma discharges. To characterize CLL during discharges, a fast infrared camera and the spectroscopic signals from Li and D atom emission have been used. The experiments analyzed so far and simulated by ANSYS code, point out that heat loads as high as 2 MW/m2 for 1.5 s have been withstood without problems. © 2014 Elsevier B.V.

This work presents the results of a property study and analysis of V–Ti–Cr alloys with a total base component concentration of Ti + Cr in the range of 8–30% with consideration of the influence of impurity and possible transmutation elements. The effect of the base alloying elements (titanium, chromium) and impurity elements (iron, silicon, aluminum, scandium etc.) on the recrystallization temperature, hardness and mechanical properties of alloys at different temperatures has been estimated. The role of dynamic strain aging with respect to the increase of vanadium alloy strength and intermittence developed on a tensile stress–strain diagram at temperatures above 400 °C has been shown.

To date there is no adequate solution for high heat load plasma facing components of the next step fusion reactor among solid material options. A lithium-filled capillary porous systems (CPS) was proposed as a plasma facing material and experimental work on this subject is now in progress. Steady-state experiments with CPS-based target and lithium supply systems have shown successful operation at heat fluxes of 1–10 MW/m2 during several hours. Experimental data is obtained on lithium CPS stability at heat flux up to 25–50 MW/m2. The lithium CPS behaviour in contact with real tokamak plasma is considered for normal discharge condition at 10 MW/m2 and for plasma disruption at 15 MJ/m2. Erosion mechanism of lithium under tokamak plasma impact was analysed. Stability of lithium CPS in tokamak conditions was shown.

The ITER project development has shown that considerable difficulties are encountered when already known engineering solutions and materials are used for divertor and divertor plates for tokamaks of such a scale. We offer to use a Li capillary-pore system (CPS) as a plasma facing material for tokamak divertor. Evaporated Li serves as a gas target and redistributes thermal load. The heat flux from the plasma is transported to the first wall by Li radiation in the plasma periphery. This allows the divertor plate to reduce the heat flux. A solid CPS filled with liquid lithium has a high resistance to surface damage in the stationary mode and during plasma transitions (disruptions, ELMs, VDEs, runaways) to assure normal operation of the divertor target plates. These materials are not the sources of impurities giving rise to Zeff and they will not be collected as dust in the divertor area and in ducts. Experiments with lithium CPS in a steady-state mode (up to 25 MW m-2) and in plasma disruption simulation conditions (~5 MJ m-2, ~0.5 ms) have been performed. High stability of these systems have been shown. Li limiter tests on T-11M tokamak have revealed the lithium CPS compatibility with the edge plasma for energy loads of up to 10 MW m-2. In a stable discharge mode at lithium limiter temperature of 20-600°C, no Li abnormal erosion and injection to plasma have been detected. A high sorption of D+ and H+ ions on the vessel walls was the main substantial result of the replacement of a graphite limiter by lithium one. He and D sorption was terminated by wall heating up to 50-100°C and above 350°C, respectively. T-11 tests on helium discharge allowed to reduce limiter heat load by a factor of two due to lithium radiation. All the experimental results have shown considerable progress in the development of lithium divertor.

The existing technological approaches for the formation of nitride- and oxide-based self-healing electrical insulating coatings for vanadium alloy–lithium systems are considered. The results of the property study of coatings applied from liquid lithium containing Al, N, Si, B additions on various modes are considered. The formation conditions of AlN-based coatings with scale specific electrical resistivity (∼50 Ω m) on the V–4Ti–4Cr vanadium alloy are determined. The results of formation and stability research of coatings on the V–4Ti–4Cr vanadium alloy in convectional and forced circulating lithium with Al and N additions in the homogeneous and heterogeneous lithium systems are discussed.

The paper concerns two problems. First, the usage of a lithium material basing on capillary-pore systems as a material of rail limiter working in conditions of considerable thermal (∼10 MW/m2) and electromechanical loads caused by interaction with a tokamak plasma in a T-11M tokamak. The second problem is the influence of a modification of recycling for hydrogen, deuterium and helium due to lithium sputtering and deposition on the discharge conditions.