Tokamak Components Research Articles

Study of temporal pulse shape effects on W using simulations and laser heating

Transient heat pulses with triangular, square, and ELM-like temporal shapes are investigated in order to further understand how transient plasma instabilities will affect plasma facing components in tokamaks. A solution to the 1D heat equation for triangular pulses allows the peak surface temperature to be written analytically for arbitrary rise times. The solution as well as ANSYS simulations reveal that a positive ramp (maximum rise time) triangular pulse has a higher peak surface temperature by a factor of compared to that from a negative ramp (rise time = 0) pulse shape with equal energy density, peak power, and pulse width. Translating the results to ITER, an ELM or disruption pulse with the shortest rise time is the most benign compared to other pulse shapes with the same peak heat flux and same energy density.

High heat flux testing of divertor plasma facing materials and components using the HHF test facility at IPR

The High Heat Flux Test Facility (HHFTF) was designed and established recently at Institute for Plasma Research (IPR) in India for testing heat removal capability and operational life time of plasma facing materials and components of the ITER-like tokamak. The HHFTF is equipped with various diagnostics such as IR cameras and IR-pyrometers for surface temperature measurements, coolant water calorimetry for absorbed power measurements and thermocouples for bulk temperature measurements. The HHFTF is capable of simulating steady state heat load of several MW m−2 as well as short transient heat loads of MJ m−2. This paper presents the current status of the HHFTF at IPR and high heat flux tests performed on the curved tungsten monoblock type of test mock-ups as well as transient heat flux tests carried out on pure tungsten materials using the HHFTF. Curved tungsten monoblock type of test mock-ups were fabricated using hot radial pressing (HRP) technique. Two curved tungsten monoblock type test mock-ups successfully sustained absorbed heat flux up to 14 MW m−2 with thermal cycles of 30 s ON and 30 s OFF duration. Transient high heat flux tests or thermal shock tests were carried out on pure tungsten hot-rolled plate material (Make:PLANSEE) with incident power density of 0.49 GW m−2 for 20 milliseconds ON and 1000 milliseconds OFF time. A total of 6000 thermal shock cycles were completed on pure tungsten material. Experimental results were compared with mathematical simulations carried out using COMSOL Multiphysics for transient high heat flux tests.

Progress of JT-60SA Project: EU-JA joint efforts for assembly and fabrication of superconducting tokamak facilities and its research planning

Aiming at supporting the early realization of fusion energy, the JT-60SA Project has shown steady progress for several years toward the first plasma in 2019 under the dual frameworks: the Satellite Tokamak Programme of the Broader Approach Agreement between EU and Japan, and the Japanese national programme. JT-60SA is a superconducting tokamak designed to operate in break-even equivalent conditions for a long pulse duration (typically 100s) with a maximum plasma current of 5.5MA. A variety of plasma control capabilities enable JT-60SA to contribute directly to the ITER project and also to DEMO by addressing key engineering and physics issues for advanced plasma operation. Design and fabrication of JT-60SA components, shared by the EU and Japan, started in 2007. Assembly in the torus hall started in January 2013, and welding work of the vacuum vessel sectors (seven 40° sectors and two 30° sectors) is currently ongoing on the cryostat base. Other components such as TF coils, PF coils, power supplies, cryogenic system, cryostat vessel, thermal shields and so on were or are being delivered to the Naka site for installation, assembly and commissioning. This paper gives technical progress on fabrication, installation and assembly of tokamak components and ancillary systems, as well as progress of the JT-60SA Research Plan being developed jointly by European and Japanese fusion communities.

Proof-of-concept experiment for on-line laser induced breakdown spectroscopy analysis of impurity layer deposited on optical window and other plasma facing components of Aditya tokamak.

In the present manuscript, we demonstrate the design of an experimental setup for on-line laser induced breakdown spectroscopy (LIBS) analysis of impurity layers deposited on specimens of interest for fusion technology, namely, plasma-facing components (PFCs) of a tokamak. For investigation of impurities deposited on PFCs, LIBS spectra of a tokamak wall material like a stainless steel sample (SS304) have been recorded through contaminated and cleaned optical windows. To address the problem of identification of dust and gases present inside the tokamak, we have shown the capability of the apparatus to record LIBS spectra of gases. A new approach known as "back collection method" to record LIBS spectra of impurities deposited on the inner surface of optical window is presented.

The role and application of ion beam analysis for studies of plasma-facing components in controlled fusion devices

First wall materials in controlled fusion devices undergo serious modification by several physical and chemical processes arising from plasma–wall interactions. Detailed information is required for the assessment of material lifetime and accumulation of hydrogen isotopes in wall materials. The intention of this work is to give a concise overview of key issues in the characterization of plasma-facing materials and components in tokamaks, especially in JET with an ITER-Like Wall. IBA techniques play a particularly prominent role here because of their isotope selectivity in the low-Z range (1–10), high sensitivity and combination of several methods in a single run. The role of 3He-based NRA, RBS (standard and micro-size beam) and HIERDA in fuel retention and material migration studies is presented. The use of tracer techniques with rare isotopes (e.g. 15N) or marker layers on wall diagnostic components is described. Special instrumentation, development of equipment to enhance research capabilities and issues in handling of contaminated materials are addressed.

The ITER project construction status

The pace of the ITER project in St Paul-lez-Durance, France is accelerating rapidly into its peak construction phase. With the completion of the B2 slab in August 2014, which will support about 400 000 metric tons of the tokamak complex structures and components, the construction is advancing on a daily basis. Magnet, vacuum vessel, cryostat, thermal shield, first wall and divertor structures are under construction or in prototype phase in the ITER member states of China, Europe, India, Japan, Korea, Russia, and the United States. Each of these member states has its own domestic agency (DA) to manage their procurements of components for ITER. Plant systems engineering is being transformed to fully integrate the tokamak and its auxiliary systems in preparation for the assembly and operations phase. CODAC, diagnostics, and the three main heating and current drive systems are also progressing, including the construction of the neutral beam test facility building in Padua, Italy. The conceptual design of the Chinese test blanket module system for ITER has been completed and those of the EU are well under way. Significant progress has been made addressing several outstanding physics issues including disruption load characterization, prediction, avoidance, and mitigation, first wall and divertor shaping, edge pedestal and SOL plasma stability, fuelling and plasma behaviour during confinement transients and W impurity transport. Further development of the ITER Research Plan has included a definition of the required plant configuration for 1st plasma and subsequent phases of ITER operation as well as the major plasma commissioning activities and the needs of the accompanying R&D program to ITER construction by the ITER parties.

Analytical Analysis of the Electromagnetic Response for the Vacuum Vessel of EAST

During plasma disruptions and vertical displacement events (VDEs), time-varying eddy currents are induced on Vacuum Vessel (VV) and Plasma Facing Components (PFCs) of Tokamak. In this paper, we calculate eddy currents induced forces on VV and inner limiter (one of the PFC) during plasma disruption in Experimental Advanced Superconducting Tokamak (EAST). Various plasma transients—VDEs and major disruptions (MDs) are considered. And the study includes assessments of integral values (net vertical, hoop forces), time evolution, peaks of the force distribution. It is shown that the distribution pattern of the eddy currents for different scenarios differ greatly, therefore the resulting EMFs and torques cause different mechanical response.

Integration of remote refurbishment performed on ITER components

Internal components of the ITER Tokamak are replaced and transferred to the Hot Cell by remote handling equipment. These components include port plugs, cryopumps, divertor cassettes, blanket modules, etc. They are brought to the refurbishment area of the ITER Hot Cell Building for cleaning and maintenance, using remote handling techniques. The ITER refurbishment area will be unique in the world, when considering combination of size, quantity of complex component to refurbish in presence of radiation, activated dust and tritium. The refurbishment process to integrate covers a number of workstations to perform specific remote operations fully covered by a mast on crane system. This paper describes the integration of the Refurbishment Area, explaining the functions, the methodology followed, some illustrations of trade-off and safety improvements.

Initial DEMO tokamak design configuration studies

To prepare the DEMO conceptual design phase a number of physics and engineering assessments were carried out in recent years in the frame of EFDA concluding in an initial design configuration of a DEMO tokamak. This paper gives an insight into the identified engineering requirements and constraints and describes their impact on the selection of the technologies and design principles of the main tokamak components. The EU DEMO program aims at making best use of the technologies developed for ITER (e.g., magnets, vessel, cryostat, and to some degree also the divertor). However, other systems in particular the breeding blanket require design solutions and advanced technologies that will only partially be tested in ITER. The main differences from ITER include the requirement to breed, to extract, to process and to recycle the tritium needed for plasma operation, the two orders of magnitude larger lifetime neutron fluence, the consequent radiation dose levels, which limit remote maintenance options, and the requirement to use low-activation steel for in-vessel components that also must operate at high temperature for efficient energy conversion.

Disruptions in ITER and strategies for their control and mitigation

The thermal and electromagnetic loads related to disruptions in ITER are substantial and require careful design of tokamak components to ensure they reach the projected lifetime and to ensure that safety relevant components fulfil their function for the worst foreseen scenarios. The disruption load specifications are the basis for the design process of components like the full-W divertor, the blanket modules and the vacuum vessel and will set the boundary conditions for ITER operations. This paper will give a brief overview on the disruption loads and mitigation strategies for ITER and will discuss the physics basis which is continuously refined through the current disruption R&D programs.

Scrape-off layer ion temperature measurements at the divertor target during type III ELMs in MAST measured by RFEA

Edge-localised modes (ELMs) can carry significant fractions of their energy as far as main chamber plasma-facing components in divertor tokamaks. Since in future devices (e.g. ITER, DEMO) these energies could cause issues for material lifetime and impurity production, the energy and temperature of ions in ELMs needs to be investigated. In MAST, novel divertor measurements of Ti during ELMs have been made using the divertor retarding field energy analyser (RFEA) probe. These measurements have shown instantaneous ion energy distributions corresponding to an effective Ti at 5cm from the strike point at the target that can be as high as 60eV and that this decreases with time after the ELM start. This is consistent with the hottest, fastest ions arriving at the target first by parallel transport, followed by the lower end of the ion energy distribution. This analysis will form a basis for future data analysis of fast swept measurements of ion distributions in ELMs.

Long‐Term Hydrogen Outgassing from Plasma Facing Components

AbstractThe time evolution of hydrogen outgassing fluxes from plasma facing components (PFCs) in tokamaks at room temperature is investigated using a diffusion‐reaction model with a single trap in the bulk of PFCs material. The release of hydrogen is surface‐limited or diffusion‐limited depending on properties of material and surface processes. In both surface and diffusion‐limited regimes, the long‐term decrease of the hydrogen outgassing flux follows a time power‐law ∼t–κ, for which κ depends on the concentration of hydrogen in the material and on surface desorption processes. In surface‐limited regime, reaction‐diffusion models of hydrogen outgassing from wall material(with hydrogen diffusion in the material bulk) are shown to be equivalent to models of outgassing from wall surfaces. Experimental results showing power‐law decays of hydrogen outgassing fluxes from PFCs during long off‐plasma events (κ = 0.7 for graphite and κ = 0.8 for ITER‐like JET configuration) can be then reproduced using a diffusion‐reaction model in surface‐limited regime. (© 2014 WILEY‐VCH Verlag GmbH & Co. KGaA, Weinheim)

Plasma tomographic reconstruction from tangentially viewing camera with background subtraction

Light reflections are one of the main and often underestimated issues of plasma emissivity reconstruction in visible light spectral range. Metallic and other specular components of tokamak generate systematic errors in the optical measurements that could lead to wrong interpretation of data. Our analysis is performed at data from the tokamak COMPASS. It is a D-shaped tokamak with specular metallic vessel and possibility of the H-mode plasma. Data from fast visible light camera were used for tomographic reconstruction with background reflections subtraction to study plasma boundary. In this article, we show that despite highly specular tokamak wall, it is possible to obtain a realistic reconstruction. The developed algorithm shows robust results despite of systematic errors in the optical measurements and calibration. The motivation is to obtain an independent estimate of the plasma boundary shape.

Present Status and Achievements of Broader Approach Activities

Five years have passed since the Broader Approach (BA) activities launched in 2007 under the framework of collaboration between Japan and EURATOM. In the International Fusion Materials Irradiation Facility (IFMIF) Engineering Validation and Engineering Design Activities (EVEDA) project, commissioning of the prototype lithium test loop was completed in March 2011, and a lithium flow of 25 mm thick at 15 m/s was obtained in July 2012. Components of the prototype accelerator are being fabricated. Injector will be delivered in Rokkasho in March 2013 followed by the validation test of the accelerator. In the International Fusion Energy Research Center (IFERC) project, five R&Ds mainly on DEMO blanket are being carried out at a newly constructed DEMO R&D facility. A high performance supercomputer with a LINPAC performance of 1.27 Pflops is now operational for the fusion simulation. In the Satellite Tokamak Program (STP) project, JT-60 torus disassembly is almost completed, and manufacture of tokamak components such as PF coils, vacuum vessel, cryostat are being fabricated on schedule. JT-60SA construction will start in Dec.2012. Present status and recent achievements of these BA activities are summarized.

Thermal fatigue damage of Cu–Cr–Zr alloys

The primary aim of this investigation is to examine thermal fatigue damage (TFD) in Cu–Cr–Zr alloys used in High Heat Flux components of Tokamak and its subsystems. Thermal fatigue experiments have been carried out between 290°C and 30°C, which is analogous to the condition of service application on two Cu–Cr–Zr alloys having different aging treatments. The extents of TFD have been examined by standard measurements of electrical conductivity, lattice strain, residual stress and dynamic elastic modulus, supplemented by characterizations of microstructure and determination of hardness and tensile properties. The results lead to infer that the relative amounts of damage are different in the two alloys which are further dependent on their aging conditions; the reasons for the observed difference have been explained. The operative mechanisms of TFD are revealed to be as formation and subsequent coalescence of microvoids, and/or initiation and growth of microcracks.

Engineering design for the HL-2M tokamak components

The HL-2A tokamak will be modified into HL-2M. The Bt at the plasma center (major radius R=1.78m) is 2.2T, the minor radius is 0.65m. The plasma current IP of HL-2M will reach up to 2.5MA, the elongation and triangularity is more than 1.8 and more than 0.5, respectively. The vacuum vessel torus consists of 20 sectors with “D” shaped cross-section and double wall structure. 20 toroidal field coil bundles comprise 140 turns which are designed with demountable joints, the poloidal field coils system consists of 25 coils. The engineering design and calculation for field coil system, vacuum vessel, support structure, etc. are finished, many key issues for manufacture process have been discussed with industry and the fabrication of main components of HL-2M tokamak will be carried out in factories.

The effect of thermo-oxidation on plasma performance and in-vessel components in DIII-D

In April 2010, two thermo-oxidation experiments (‘O-bakes’) were performed in the DIII-D tokamak. Internal surfaces of the tokamak, as well as a number of specimens inserted into the torus, were exposed to a mixture of 20% O2/80% He at a nominal pressure of 9.5 Torr (1.27 kPa) at a temperature of 350–360 °C for a duration of 2 h. Three primary conclusions have been drawn from these experiments: (1) laboratory measurements on the release of deuterium from tokamak codeposits by oxidation have been duplicated in a tokamak environment, (2) no internal tokamak components or systems were adversely affected by the oxidation and (3) the recovery of plasma performance following oxidation was similar to that following regular torus openings.

Introducing artificial depth cues to improve task performance in ITER maintenance actions

Maintenance operations on ITER tokamak components will be largely performed by remote handling. In previous work it was shown that representative maintenance tasks could be performed significantly faster with direct visual feedback than with camera feedback. In post-test interviews, operators indicated that they regarded the lack of 3D perception as the primary factor hindering their performance.This paper discusses various techniques to improve depth perception in teleoperation, including stereo vision, head tracking, virtual camera views and depth gauges. The most promising techniques were tested. Performance metrics included time-to-complete, path analysis and operator work-load.In a first experiment, artificial depth gauges views were tested in a 1:1 scale hardware testbed with mechanical master-slave manipulators handled by experienced operators. Robust real-time image processing was achieved with marker-based objects. The simple depth gauge and graphical overlay did not significantly improve task performance. Operators commented on their view of the task being “obstructed” by the graphical overlay, and the depth gauge was judged not very informative.In a second experiment, real time tracking was combined with VR display including stereo and head tracking. While stereo was found to improve the task performance significantly over the 1 camera (mono) condition, head tracking unexpectedly did not.

Three-dimensional particle-in-cell simulations of gap crossings in castellated plasma-facing components in tokamaks

This paper presents the first three-dimensional (3D) particle-in-cell (PIC) simulations of castellated plasma-facing components (PFCs) in tokamaks. Special focus is given to crossings between poloidal and toroidal gaps where elevated heat loads are expected to occur. Moreover, the crossings may affect the plasma penetration into the gaps between tiles. Both of these problems are of high importance for ITER when estimating the lifetime of its PFCs. Localized heat loads can potentially lead to damage of the tiles, while the plasma penetration is related to fuel retention in the gaps due to redeposition of eroded wall material. This problem has previously been targeted by 2D PIC simulations using our in-house code SPICE2, where toroidal and poloidal gaps (PGs) had to be simulated separately. This paper presents the results of a full 3D3V code SPICE3, which allows us to simulate a more realistic geometry of the tiles including the gap crossings and includes the complete E × B drift, which could not be simulated in 2D. The results of self-consistent simulations show that the crossing acts as a transport channel for electrons, allowing them to enter the plasma shadowed region in PGs. As a consequence, the potential near the gap entrance is modified allowing more ions to flow deep inside the gap. The combination of the plasma flow and an E × B drift in the crossing directs ions onto one tile corner, which receives elevated heat load.

Assembly study for JT-60SA tokamak

The assembly scenarios and assembly tools of the major tokamak components for JT-60SA are studied in the following. (1) The assembly frame (with a dedicated 30-tonne crane), which is located around the JT-60SA tokamak, is adopted for effective assembly works in the torus hall and the temporary support of the components during assembly. (2) Metrology for precise positioning of the components is also studied by defining the metrology points on the components. (3) The sector segmentation for weld joints and positioning of the vacuum vessel (VV), the assembly scenario and tools for VV thermal shield (TS), the connection of the outer intercoil structure (OIS) and the installation of the final toroidal field coil (TFC) are studied, as typical examples of the assembly scenarios and tools for JT-60SA.