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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.

The solution of the problem of heat exhaust has been pointed out as one of the main challenge towards the realization of magnetic confinement fusion. In the last years, two concepts have been proposed in alternative to the conventional divertor solution adopted for ITER: modification of the magnetic topology in the divertor region and liquid metal as plasma facing component. The role of the Divertor Tokamak Test facility (DTT) in the power exhaust implementation strategy is discussed. The evolution of the project, since the original proposal in 2015 to the present design, is shown. The DTT facility is well integrated in the European strategy and the final decision on the divertor configuration will be made, within 2022-23, on the basis of the indication of the Power Exhaust Group constituted by the EUROfusion Consortium. Finally, the main milestones and the timeline of the project are illustrated. © 2019 Elsevier B.V.

The DTT proposal not only considers the fundamental technical and scientific aspects, but also includes a careful financial and managerial analysis. It is organized taking into account that only a careful management of all human and financial resources can guarantee the success of the initiative. Regarding the financial aspects, the project analyzes the costs of both engineering design and construction phases, reporting a breakdown of the expenses and their time evolution. The study also considers the expected financial revenues in terms of consistency, funding channels, and time evolution. The organization and government layout (organs, functions, relationships) has been designed pursuing the best balance between autonomy and accountability, transparency and cooperation. The paper also discusses the capability of the ENEA Frascati site to host the device and, in addition, examines the socio-economic impact expected by the long, expansive presence of a high number of scientists and technicians in the area.

One of the main objectives of the DTT project is to test many divertor designs and configurations, so that the concept of the machine could change from the initial single null (SN) configuration to other configurations such as the SnowFlake Divertor (SFD). Furthermore the design of Vacuum Vessel, ports and In-Vessel Components should take into account the application and testing of a Liquid Metal Divertor. For this reason the divertor design has been developed having in mind the possibility of easily replacing the divertor itself by remote handling.

Materials such as lithium, gallium and tin have been proposed as candidates for liquid metals plasma facing components. The properties of these materials introduce novel methods to cope with the high power loads on the divertor region of nuclear fusion reactors. Although several results have been reported using lithium less experiments were done using gallium or tin. With the purpose of comprehending the impact of the utilisation of tin on a tokamak's plasma performance it was deemed relevant to perform a series of experiments with particular emphasis on plasma contamination. In this work several diagnostics, such as Hα, bolometry, visible spectroscopy and electrostatic probes are used to characterise the interaction of plasmas with tin samples and impact this material had on the tokamak discharge performance. Liquid tin samples were exposed in the ISTTOK's edge and subject to deuterium plasmas. The sample's surface temperature was monitored by pyrometry from which plasma induced heat load was inferred. Heat loads in excess of 300 kWm−2 was regularly observed which are in agreement with Langmuir probe analysis. Hα, bolometry and plasma resistivity were used as proxies for plasma performance and these were found to remain unchanged throughout the sample exposure. The radial impurity content across the plasma column of ISTTOK is characterised by spectroscopic measurements of the Sn-II line 579.88 nm. The line intensity was found to exponentially decrease towards the plasma center with a decay length of ≈2.3 mm. These results suggest that, under specific experimental conditions, liquid tin had a negligible impact in the plasma performance of ISTTOK discharges.

Peaked density profiles are observed in FTU discharges when the recycling condition of the chamber is influenced by the action of the liquid lithium limiter (LLL) [1] . Turbulence analysis of lithium doped FTU plasmas [2] , [3] has shown that the presence of the light impurity modifies the phase between fluctuating fields responsible for transport and consequently leads to an inward deuterium pinch and outward impurity flux. Analogous peaked discharges were produced by Ne-gas puffing in different L mode plasma scenarios that have been recently obtained on FTU with following plasma parameters: I = 360 kA, B = 5–6.5 T, ne0 = 0.2–1 × 1020 m−3, Te0 = 1–4 keV, as well as in similar experiments on other machines [4] . In fact the Ne seeded plasmas show an increase of the peaking factor around 30% [5] . UV spectroscopy measurements confirm that the electron-density peaking arises from a convective flow and cannot be attributed to the contribution of the injected Ne alone. The Ne doped discharges analysis together with lithium conditioned ones is useful to extend the interpretative framework of the particle transport. In this work a comparison of the diffusion coefficient and of the pinch velocity of the two cases is conducted. By using a two-colors scanning interferometer providing very high spatial and time resolution, it is indeed possible to estimate the D and U coefficients of a simple model for the particle flux [6] .

The DTT facility is designed to study, within the European Fusion Road Map, the completely integrated (physics-technology and bulk-edge) power exhaust problems. The scientific program is characterized by the tight balance between clear guidelines and the “flexibility” to tackle all the present scientific questions as well as the ideas possibly arising along the years of its operative life; in particular, it will be ready to answer to the necessities arising from both ITER operation and/or DEMO design. The paper discusses the guideline of DTT operations program to be carried out during its operative life, planned in about thirty years. © 2017 Elsevier B.V.

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.