Plasma Facing Elements Research Articles

Comparison of plasma start-up with high Z and low Z first wall in WEST

Abstract The choice of first wall material is of paramount importance for the plasma start-up conditions in ITER and future fusion power plants. In this context, the present work focuses on the correlations between first wall impurity sources and total radiated power during plasma start-up in the tungsten (W) Environment in Steady-state Tokamak (WEST). The objective is to highlight experimental indications for a preferable combination of start-up plasma scenario and first wall materials. Until 2019, WEST featured a full high Z first wall with all limiters exposing only W surfaces to the plasma. To study the impact of a low Z first wall in WEST, boron nitride (BN) tiles were installed in the central part of the inner and outer limiters in 2020. Although visible spectroscopy and bolometry measurements show respectively a strong weakening of the WI line intensity on the limiters and a reduction of radiated power after the changeover, a degradation occurs with the accumulation of plasma exposure. In addition, the different plasma facing elements of the main chamber do not influence equally the radiated power during start-up. In both high Z and low Z environments, a clear non-linear dependence is found between the start-up radiated power and the outer limiter W impurity source. Since W seems to be the main cause for core radiation, correlation between outer limiter W sources and other impurity sources are investigated. Finally, analysis of the legacy of B powder drops on a number of start-up plasmas suggests that it is less effective at reducing radiated power when the first wall is covered with W.

Options for Implementation of IR Thermography Diagnostics in a Tokamak with Reactor Technologies TRT

AbstractOptions 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.

First study of lithium boron composite as plasma facing material

Lithium-Boron Composite Material (LBCM) is presented as a plasma-facing material with a lot of promise from the perspective of liquid metal first wall. LBCM consists of ∼ 20 w.% boron and ∼ 80 w.% lithium, arranged in the form of a Li5B4 matrix filled with metal lithium, similar to capillary porous systems (CPS) already used as plasma-facing elements. The characteristic matrix cell size is approximately 5 µm. LBCM was subjected to electron flux and helium plasma in a linear PR-2 device. LBCM maintained physical integrity up to T > 900 °C. Boron didn’t evaporate from LBCM under electron irradiation, but was sputtered during helium bombardment. CPS-like properties, wherein Li from the whole of the LBCM sample was transported to the area of the highest thermal load were observed.

Comparative Analysis of Lithium First Wall Concepts for Tokamak with Reactor Technologies

When developing the stationary fusion reactor, an unresolved issue is the design of its intra-chamber plasma-facing elements. It has now become obvious that among the materials conventionally used for intra-chamber elements, there are no solid structural materials that would meet the requirements for the long-term operation under the effect of the flux of fusion neutrons (14 MeV) with a density of ~1014 cm–2 s–1 and the heat flux with a power density of 10–20 MW/m2. An alternative solution to this problem is the use of liquid metals as a plasma-facing materials, and, first of all, the use of lithium, which has a low atomic number (low charge number Z). Other easily-melting metals are also considered, which have higher Z number, but lower saturation vapor pressure than lithium. This will make it possible to create the long-lived, heavy-to-damage and self-renewing surface of the intra-chamber elements, which will not contaminate the plasma. The main ideas of the alternative concept of the intra-chamber elements can be formulated based on the comprehensive analysis of the problems and requirements arising during the development of intra-chamber elements of the stationary reactor, for example, the DEMO-type reactor. The article presents the analysis of the possible design of the lithium-coated intra-chamber elements and discusses the main ideas of the lithium first wall concept for the tokamak with reactor technologies.

Properties of graphite to be used as a material for the T-15MD tokamak plasma-facing elements

The results of the measurement of physical properties, sputtering yield during irradiation by hydrogen ions and a study of surface morphology modification of graphite to be used for the plasma-facing materials of the T-15MD tokamak are presented in this work.

Using the Scattering Spectroscopy of keV-Energy Protons to Analyze the Deposition of Lithium on Tungsten

Controlling the thickness of thin layers of lithium when they are deposited on plasma-facing elements is an important task in the field of plasma-surface interaction in fusion devices. It is shown the keV‑energy proton scattering spectroscopy can be used for this purpose. The sensitivity of technique to chemical reactions on a surface is demonstrated, and estimates of the charge state of hydrogen ions reflected from a lithium surface are given.

Modern Diagnostics for Investigation of Lithium Element Behavior in Tokamaks

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.

Progress of fusion technology: summary of FIP, FNS, MPT, and SEE from the 27th IAEA Fusion Energy Conference

Two hundred and twenty-five papers relevant to fusion technology and engineering were presented at the 27th IAEA Fusion Energy Conference. The ITER project has reached 57.4% of the total construction work scope through to first plasma. Construction of ITER is indeed progressing at full speed, with some ITER components having already been manufactured, such as the toroidal field coil structures and the gyrotrons of the electron cyclotron heating and current drive system, as a result of decades of technological development. Construction of JT-60SA and WEST is also progressing on schedule to support ITER in various engineering and physics aspects. The EU and Japan are advancing with their pre-conceptual designs for DEMO, and the experience and lessons learned from ITER are already being applied to both designs. China reported on the status of the China Fusion Engineering Test Reactor engineering design. One of the technical issues evident from these DEMO designs is the interfacing of the plasma-facing, in-vessel components with plasmas. Advanced plasma-facing elements having liquid metals and an advanced in situ boron coating technology were applied to existing fusion reactors, the results of which were reported at this conference. It was also reported that commissioning of the injector for the International Fusion Materials Irradiation Facility engineering validation and engineering design activities (IFMIF/EVEDA) project has been completed and testing of the Radio Frequency Quadrupole accelerator has begun. Europe and Japan presented their plans to construct neutron irradiation test facilities: the DEMO-Oriented Neutron Source (DONES) and Advanced Fusion Neutron Source (A-FNS), respectively. A new strategy based on a probability-based design method was proposed to mitigate the difficulty of defining the allowable limit for irradiation-induced degradation with limited empirical evidence.

Analysis of mixed-material layers deposited on the toroidal array probes during the FY 2012 LHD plasma campaign

Processes occurring on the plasma-facing walls in the Large Helical Device (LHD) and in-vessel material migration were investigated using a technique of material deposition probes. Ten Si plates were located on the outer side of the first-wall surface in each 36° toroidal section (Nos. 1–10). Mixed-material layers deposited on the probes during the fiscal year 2012 plasma campaign contained carbon, boron, hydrogen, and metals from which the plasma-facing elements were made. Metallic impurities and hydrogen content in deposited layers were examined by Rutherford backscattering spectrometry (RBS) and elastic recoil detection analysis (ERDA), respectively. The cross-sectional observations of the deposited mixed-material layers were performed with the help of a scanning electron microscope equipped with an energy dispersive X-ray (EDX) spectrometer. The results of the combined RBS/ERDA analysis and TEM-EDX examination allow tracing the processes occurring on plasma-facing components during the plasma campaign.

PROJECT ON CONSTRUCTION OF KAZAKHSTAN MATERIAL TESTING TOKAMAK KTM

Article presents a short overview of the modern power engineering condition and its impact on global ecology. The main advantages of thermonuclear power engineering are highlighted as a potential solution of power crisis in the future. KTM facility overview is presented, meant for solution the issue on large-scale study of materials, planned to be used for intrachamber plasma facing elements production for future thermonuclear reactors. The problem of the basic technological systems development, introduction and commissioning is exposed on the example of power supply pulse system.

Experience in the Development of Liquid Metal Plasma Facing Elements Based on Capillary-Porous Structures for a Steady-State Tokamak (Survey)

Realization of steady-state operation of a fusion reactor requires the development of essentially new designs and materials for plasma facing elements (PFE). The most promising solution in this field is the concept of capillary-porous systems (CPS) with liquid metal (first of all, lithium) that provide PFE surface self-renewal, closed circulation of their corrosion products, and plasma performance improvement and promote achievement of almost stationary modes of plasma burning. Furthermore, the problem of power exhaust with high specific density (20–30 МW/m2) and maintaining a reasonable temperature level on the PFE surface can be solved by the introduction of a special heat removal system. The survey covers the experience in the development, manufacture, and experimental study of liquid metal CPS based on the tests of PFE element models with thermal stabilization systems in steady-state conditions for the T-11М, Т-10, FTU, and КТМ tokamaks and different aspects of liquid metal application.

Comparative analysis of the possibility of applying low-melting metals with the capillary-porous system in tokamak conditions

The use of capillary-porous systems (CPSs) with liquid Li, Ga, and Sn is considered as an alternative for solving the problem of creating plasma-facing elements (PFEs) of the fusion neutron source (FNS) and the DEMO-type reactor. The main advantages of CPSs with liquid metal compared with hard materials are their stability with respect to the degradation of properties in tokamak conditions and capability of surface self-restoration. The evaluation of applicability of liquid metals is performed on the basis of the analysis of their physical and chemical properties, the interaction with the tokamak plasma, and constructive and process features of in-vessel elements with CPSs implementing the application of these metals in a tokamak. It is shown that the upper limit of the PFE working temperature for all low-melting metals under consideration lies in the range of 550–600°С. The decisive factor for PFEs with Li is the limitation on the admissible atomic flux into plasma, while for those with Ga and Sn it is the corrosion resistance of construction materials. The upper limit of thermal loads in the steady-state operating mode for the considered promising PFE design with the use of Li, Ga, and Sn is close to 18–20 MW/m2. It is seen from the analysis that the use of metals with a low equilibrium vapor pressure of (Ga, Sn) gives no gain in extension of the region of admissible working temperatures of PFEs. However, with respect to the totality of properties, the possibility of implementing the self-restoration and stabilization effect of the liquid surface, the influence on the plasma discharge parameters, and the ability to protect the PFE surface in conditions of plasma perturbations and disruption, lithium is the most attractive liquid metal to create CPS-based PFEs for the tokamak.

Making ICRF power compatible with a high-Z wall in ASDEX Upgrade

A comparison of the ASDEX Upgrade 3-strap ICRF antenna data with the linear electro-magnetic TOPICA calculations is presented. The comparison substantiates a reduction of the local electric field at the radially protruding plasma-facing elements of the antenna as a relevant approach for minimizing tungsten (W) sputtering in conditions when the slow wave is strongly evanescent. The measured reaction of the time-averaged RF current at the antenna limiters to the antenna feeding variations is less sensitive than predicted by the calculations. This is likely to have been caused by temporal and spatial fluctuations in the 3D plasma density distribution affected by local non-linear interactions. The 3-strap antenna with the W-coated limiters produces drastically less W sputtering compared to the W-coated 2-strap antennas. This is consistent with the non-linear asymptotic SSWICH-SW calculations for RF sheaths.

Selection of materials for tokamak plasma facing elements based on a liquid tin capillary pore system

Capillary-Pore Systems (CPS) filled by liquid metals are considered as an alternative solution of materials choice for plasma facing component of tokamak reactor. Tin is viewed as one of the candidates for CPS because it has lower corrosiveness than gallium and lower saturated vapour pressure compared to lithium. The corrosion resistance of Mo, Nb and W in pure liquid tin was investigated. The corrosion tests were carried out in the static isothermal conditions at a temperature up to 1050°C. As a result of the corrosion study, it was found that Mo does not corrode in liquid Sn, as opposed to Nb and is compatible with liquid tin in temperatures of up to approx. 1000°C. This allows considering Mo as an alloy base material of the in-vessel tokamak elements based on liquid tin capillary pore systems.

Application of Analytical Model of the Electric Potential Distribution for Calculation of Charged Particle Dynamics in a Near-Wall Layer and Sputtering of the Plasma Facing Surfaces

Simple analytical formulas are derived for calculation of the electric field potential distribution in the magnetic pre-layer and the Debye layer near the plasma facing surfaces. It is shown that the calculated potential profiles are in good agreement with the dependences of the potential distribution on the magnetic field inclination obtained by solving the magnetic hydrodynamic (MHD) equations and modeling using the PIC code SPICE2. Dependences of the angular distribution of ions incident on the surface of plasma facing elements on the magnetic field inclination are obtained. Results of calculations demonstrate that the surface areas, on which the magnetic field is incident at sliding angles, are critical from the viewpoint of the increase of sputtering.

Development of Lithium CPS Based Limiters for Realization of a Concept of Closed Lithium Circulation Loop in Tokamak

Cooling of tokamak boundary plasma owing to radiation of non-fully stripped lithium ions is considered as a promising way for protection of plasma facing elements (PFE) in tokamak. It may be effectively realized when the main part of lithium ions are involved in the closed circuit of migration between plasma and PFE surface. Such an approach may be implemented with the use of lithium device whose hot (500–600°C) area to be effected by plasma serves as a Li-emitter and the cold part (∼180°C) as a Li-collector in the shadow. Capillary-pore system (CPS) provides the returning of collected and condensed lithium to emitting zone by capillary forces. The main goals of the last T-11M lithium experiments were investigating Li ions transport in the tokamak scrape of layer (SOL) and their collecting by different kinds of limiters. The design of devices based on lithium CPS with different ratio of emitting/collecting area is the main subject of this paper.

Status of design and experimental activity on module of lithium divertor for KTM tokamak

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.

Module of lithium divertor for KTM tokamak

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.

Spectroscopic studies of homogeneous thin carbon erosion films on mirrors and flakes with a high deuterium content formed in tokamak T-10

Redeposited hydrocarbon films on plasma facing elements in tokamaks accumulate hydrogen isotopes. In the present study such films were made to redeposit on stainless steel mirror substrates as thin films and without any substrate as bare flakes with high deuterium content, under deuterium-plasma discharges inside T-10 tokamak vacuum chamber. These films were subjected to spectral characterizations through Fourier-transform infrared (FT-IR), electron paramagnetic resonance (EPR), and photoluminescence techniques. IR spectra showed the presence of two main deuterium states as observed by the CD 2,3 sp 3 stretching modes at 2100–2200 cm −1 and the CD 2 sp 3 bending modes at 600–1100 cm −1. Among these, CD 3 stretching mode at 2217 cm −1 may serve as a control for deuterium desorption during the cleanup process of the reactor. As a comparative measure, C60 films were also studied, the luminescence excitation spectrum of which showed similarity in peak positions with tokamak bare flakes pertained to sp 2 luminescence centers. The observed spectral differences are mainly due to more localized sp 2 states for C60 and sp 3 states for tokamak flakes. EPR spectra of the bare flakes showed the defective states with a high spin density, ∼10 19 cm −3 which serve as luminescence quenching centers, and provide a path for hydrogen isotopes adsorption.