Steady-state Tokamak Research Articles

Qualification, manufacturing and assembly of the WEST divertor structure and coils

In order to fully validate “ITER-like” actively water cooled tungsten plasma facing units, addressing the issues of long plasma discharges, an axisymmetric divertor structure has been studied and manufactured for the implementation in the WEST (W-Tungsten Environment in Steady state Tokamak) tokamak platform.This assembly, called divertor structure and coils (4m diameter, 20t), is composed of two stainless steel casings containing an actively water cooled (up to 180°C, 4MPa) copper winding pack designed for a conductor current in the range of 12.5kA (up to 1000s). It must sustain harsh environmental conditions in terms of ultra-high vacuum, high temperatures and electrodynamic loads. One major difficulty is the assembly by induction brazing of individual bended conductor sectors inside the vacuum vessel and the consecutive sealing of the casings by TIG welding.Therefor development activities have been carried out on a scale one dummy coil, such as brazing, assembly, thermal cycling and electrical insulation tests (5kV ground voltage). Whereas the brazing assembly technics and the conductor installation were validated without major difficulties, different technical solutions for the electrical insulation had to be tested. The chosen solution is a resin epoxy impregnated fiber glass fabric layered around the conductors followed by a polymerization procedure.In parallel the manufacturing of divertor structure components started in the second half of 2013 with a total delivery at the end of 2015.The paper will illustrate the technical developments which have been performed in order to fully validate the design. It concerns mainly the dummy coil and the complex conductor installation procedure assisted by virtual reality tools. The manufacturing methods proposed by industry in order to fulfil the technical requirements will be also addressed. Finally the processes and associated tools used in order to implement this large component inside the WEST vacuum vessel will be detailed.

Mechanical analysis of WEST divertor support plate

The Tore Supra tokamak is being transformed in an x-point divertor fusion device in the frame of the WEST (W-for tungsten-Environment in Steady-state Tokamak) project, launched in support to the TER tungsten divertor strategy. The WEST project aims to test W monoblock Plasma Facing Units (PFU) under long plasma discharge (up to 1000s), with thermal loads of the same magnitude as those expected for ITER. Therefore the divertor is a key component of the WEST project, and so is its support structure, which has to handle strong mechanical loads. The WEST upper and lower divertor are made of 12 30° sectors, each one composed of 38PFU that can be made of tungsten, CuCrZr or graphite. A generic 316L stainless steel 30° conic support plate is used to hold the 38PFU together, regardless of their material. The PFUs are fixed on the support plate thanks to 152 Xm19 stainless steel fixing elements (4 per PFU), and in each of this fixing element an Aluminium-Nickel-Bronze alloy (Al-Ni-Br) pin is engaged in a slotted hole, in order to allow thermal expansion in the length direction of the PFU. The support plate is fixed on the divertor coil casing thanks to 10M10 screws. Mechanicals loads which act on the PFUs are transmitted to the support plate through the fixing elements. These loads are due to Vertical Displacement Event (VDE), disruptions and thermal expansion of the PFU. First the different load cases, PFU configurations and scenario are presented. Then an ANSYS plastic mechanical simulation is performed in order to validate the number of cycles of the support plate for each scenario: 30,000 cycles in steady-state and 3000 cycles in VDE. Finally reactions forces from the previous ANSYS simulation are used in order to calculate the stress in the M10 screws.

A new simpler way to obtain high fusion power gain in tandem mirrors

From the earliest days of fusion research, Richard F. Post and other advocates of magnetic mirror confinement recognized that mirrors favor high ion temperatures where nuclear reaction rates begin to peak for all fusion fuels. In this paper we review why high ion temperatures are favored, using Post’s axisymmetric Kinetically Stabilized Tandem Mirror as the example; and we offer a new idea that appears to greatly improve reactor prospects at high ion temperatures. The idea is, first, to take advantage of recent advances in superconducting magnet technology to minimize the size and cost of End Plugs; and secondly, to utilize parallel advances in gyrotrons that would enable intense electron cyclotron heating (ECH) in these high field End Plugs. The yin-yang magnets and thermal barriers that complicated earlier tandem mirror designs are not required. We find that, concerning end losses, intense ECH in symmetric End Plugs could increase the fusion power gain Q, for both DT and Catalyzed DD fuel cycles, to levels competitive with steady-state tokamaks burning DT fuel. Radial losses remain an issue that will ultimately determine reactor viability.

Millimeter-wave imaging of magnetic fusion plasmas: technology innovations advancing physics understanding

Electron cyclotron emission (ECE) imaging is a passive radiometric technique that measures electron temperature fluctuations; and microwave imaging reflectometry (MIR) is an active radar imaging technique that measures electron density fluctuations. Microwave imaging diagnostic instruments employing these techniques have made important contributions to fusion science and have been adopted at major fusion facilities worldwide including DIII-D, EAST, ASDEX Upgrade, HL-2A, KSTAR, LHD, and J-TEXT. In this paper, we describe the development status of three major technological advancements: custom mm-wave integrated circuits (ICs), digital beamforming (DBF), and synthetic diagnostic modeling (SDM). These have the potential to greatly advance microwave fusion plasma imaging, enabling compact and low-noise transceiver systems with real-time, fast tracking ability to address critical fusion physics issues, including ELM suppression and disruptions in the ITER baseline scenario, naturally ELM-free states such as QH-mode, and energetic particle confinement (i.e. Alfvén eigenmode stability) in high-performance regimes that include steady-state and advanced tokamak scenarios. Furthermore, these systems are fully compatible with today’s most challenging non-inductive heating and current drive systems and capable of operating in harsh environments, making them the ideal approach for diagnosing long-pulse and steady-state tokamaks.

Overview of the different processes of tungsten coating implemented into WEST tokamak

The Tore Supra tokamak is being transformed in an x-point divertor fusion device within the frame of the WEST (W-for tungsten-Environment in Steady-state Tokamak) project, launched in support to the ITER tungsten divertor strategy. The WEST project aims at testing W monoblock Plasma Facing Units (PFU) under long plasma discharge, with thermal loads of the same magnitude as those expected for ITER. The others Plasma Facing Components (PFC) will also be modified and coated with W to transform Tore Supra into a fully metallic environment. Different coating techniques have been selected, taking into account the specifications of the various PFC: heat loads, complex geometries (length up to 1m) and different substrates (CuCrZr for actively cooled PFC, graphite and CFC for other components). This paper gives an overview on the different processes used and the associated validation program and concludes on the adequacy of the W coating with the WEST experimental program requirements.

10th Asia Plasma and Fusion Association Conference

Welcome to APFA-2015It gives me immense pleasure to note that 10th Asia Plasma Fusion Association Conference is being held at Institute for Plasma Research (IPR), Bhat, Gandhinagar 382 428, INDIA during 14-17 December, 2015. This is the second time this conference is being held in India, the first time being 2006. Main aim of the conference is to bring together researchers and students working in the field of fusion plasma science and technology, especially, in the South East Asia region, every alternate year under one umbrella to discuss technology advancement made during this period.Fusion activities have increased in all the member countries during this period. Smaller countries of this region have also started experimental programmes in plasma physics. Therefore, APFA executive council may look into more participation of these countries and more opportunities for interactions on important fusion science and technology related areas. More Industry participation is also welcome in the meetings. This is being helpful to prepare human resource for the future need in fusion field. Interactions in these meetings are generating interesting regional collaborations.As India is a full partner in the ITER programme and has now developed and commissioned successfully its own Superconducting Steady state Tokamak (SST-1), research in fusion science and technology needs to increase substantially. APFA is playing an important role in advancing this effort through this meeting.Few awards for younger colleagues have been instituted which should create a very healthy competition among the participants. We request members to instate more such awards.IPR colleagues are holding the conference this year to bring the plasma researchers together once again.I wish APFA, all the best in this endeavour.Dhiraj BoraConference Chair

High Field Side Lower Hybrid Current Drive Simulations for Off- axis Current Drive in DIII-D

Efficient off-axis current drive scalable to reactors is a key enabling technology for developing economical, steady state tokamak. Previous studies have focussed on high field side (HFS) launch of lower hybrid current drive (LHCD) in double null configurations in reactor grade plasmas and found improved wave penetration and high current drive efficiency with driven current profile peaked near a normalized radius, ρ, of 0.6-0.8, consistent with advanced tokamak scenarios. Further, HFS launch potentially mitigates plasma material interaction and coupling issues. For this work, we sought credible HFS LHCD scenario for DIII-D advanced tokamak discharges through utilizing advanced ray tracing and Fokker Planck simulation tools (GENRAY+CQL3D) constrained by experimental considerations. For a model and existing discharge, HFS LHCD scenarios with excellent wave penetration and current drive were identified. The LHCD is peaked off axis, ρ∼0.6-0.8, with FWHM Δρ=0.2 and driven current up to 0.37 MA/MW coupled. For HFS near mid plane launch, wave penetration is excellent and have access to single pass absorption scenarios for variety of plasmas for n|| =2.6-3.4. These DIII-D discharge simulations indicate that HFS LHCD has potential to demonstrate efficient off axis current drive and current profile control in DIII-D existing and model discharge.

Simulation results for PLATO: a prototype hybrid X-ray photon counting detector with a low energy threshold for fusion plasma diagnostics

PLATO is a prototype hybrid X-ray photon counting detector that has been designed to meet the specifications for plasma diagnostics for the WEST tokamak platform (Tungsten (W) Environment in Steady-state Tokamak) in southern France, with potential perspectives for ITER. PLATO represents a customized solution that fulfills high sensitivity, low dispersion and high photon counting rate. The PLATO prototype matrix is composed of 16 × 18 pixels with a 70 μm pixel pitch. New techniques have been used in analog sensitive blocks to minimize noise coupling through supply rails and substrate, and to suppress threshold dispersion across the matrix. The PLATO ASIC is designed in CMOS 0.13 μm technology and was submitted for a fabrication run in June 2016. The chip is designed to be bump-bonded to a silicon sensor. This paper presents pixel architecture as well as simulation results while highlighting novel solutions.

Results of high heat flux qualification tests of W monoblock components for WEST

One goal of the WEST project (W Environment in Steady-state Tokamak) is the manufacturing, quality assessment and operation of ITER-like actively water-cooled divertor plasma facing components made of tungsten. Six W monoblock plasma facing units (PFUs) from different suppliers have been successfully evaluated in the high heat flux test facility GLADIS at IPP. Each PFU is equipped with 35 W monoblocks of an ITER-like geometry. However, the W blocks are made of different tungsten grades and the suppliers applied different bonding techniques between tungsten and the inserted Cu-alloy cooling tubes. The intention of the HHF test campaign was to assess the manufacturing quality of the PFUs on the basis of a statistical analysis of the surface temperature evolution of the individual W monoblocks during thermal loading with 100 cycles at 10 MW m−2. These tests confirm the non-destructive examinations performed by the manufacturer and CEA prior to the installation of the WEST platform, and no defects of the components were detected.

Modelling of the soft X-ray tungsten spectra expected to be registered by GEM detection system for WEST

Abstract In the future International Thermonuclear Experimental Reactor (ITER), the interaction between the plasma and the tungsten chosen as the plasma-facing wall material imposes that the hot central plasma loses energy by X-ray emission from tungsten ions. On the other hand, the registered X-ray spectra provide alternative diagnostics of the plasma itself. Highly ionized tungsten emits extremely complex X-ray spectra that can be understood only after exhaustive theoretical studies. The detailed analyses will be useful for proper interpretation of soft X-ray plasma radiation expected to be registered on ITER-like machines, that is, Tungsten (W) Environment in Steady-state Tokamak (WEST). The simulations of the soft X-ray spectra structures for tungsten ions have been performed using the flexible atomic code (FAC) package within the framework of collisional-radiative (CR) model approach for electron temperatures and densities relevant to WEST tokamak.

Power and Sizes of Tokamak Fusion Neutron Sources with NBI-Enhanced Reaction Rate

Steady state tokamak with deuterium–tritium plasma is considered as a basis for fusion neutron source for a hybrid fusion–fission reactor. Prototypes of such a system can be developed on the basis of the present day tokamaks as the plasma power gain factor Q ~ 1 is required for hybrid applications. Significant population of fast ions can be supported by a powerful neutral beam injection heating in regimes with Q ~ 1. The reaction rate for fast ions greatly exceeds the rate for thermal Maxwellian ions. The possible ranges of parameters are discussed for medium size tokamaks with minor plasma radius a = 0.5–1 m. Power and sizes of the neutron source are determined by the value of the injection energy. Power gain Q ≈ 1 can be achieved with injection energy of deuterium about 130 keV and tritium energy about 200 keV. Neutron power of 30–40 MW can be realized with a ≈ 1 m, and about of few megawatts with a ≈ 0.5 m.

Compact ECEI system with in-vessel reflective optics for WEST.

An electron cyclotron emission imaging (ECEI) diagnostic system for WEST (W Environment for Steady state Tokamak) is under development to study the MHD instabilities affected by tungsten impurities. The system will provide 2-D Te fluctuation images (width × height = ∼18 cm × ∼ 34 cm at low field side and ∼13 cm × ∼ 39 cm at high field side) from a poloidal cross section with high spatial (≤1.7 cm) and temporal (≤2 μs) resolutions. While the key concept and electronic structure are similar to that of prior ECEI systems on other tokamak devices such as KSTAR, DIII-D, or ASDEX-U, part of the imaging optics have to be placed inside the vacuum vessel in order to resolve issues on limited installation space and longer beam path to the detector position. The in-vessel optics consisting of two large curvature-radius mirrors are expected to withstand the extreme heating on long-pulse operation scenario (∼1000 s). The out-vessel optical housing is constructed as compact as possible to remove easily from the installation site in case of necessity. Commissioning of the system is scheduled on the second experimental WEST campaign end of 2017.

Tungsten coatings for application in WEST project

The WEST (Tungsten Environment in Steady-state Tokamak) project is aimed to implement and test the ITER W divertor technology in relevant plasma operation conditions. In addition to the ITER-like W monoblock divertor target, the WEST project involves W coated CuCrZr target and W coated Fine Grain Graphite (FGG) and CFC-N11 (Carbon Fiber Composite) tiles.Combined Magnetron Sputtering and Ion Implantation (CMSII) technology was developed and applied in the framework of ITER-like wall (ILW) project at JET for W coating of CFC Dunlop 780 DMS tiles Ruset et al. (2007) [1]. The coating thickness was in the range of 10–25 µm depending on the position of the tiles in the wall. The same technology was also used for W coating of more than 1300 FGG tiles for ASDEX Upgrade tokamak.In the present paper W coatings of 12.8 µm and 20 µm deposited on PFC (Plasma Facing Components) materials used in WEST project will be analyzed. As substrate materials CFC N11, FGG samples were used. These substrate materials have different properties compared to materials used for ILW project. GDOES (Glow Discharge Optical Emission Spectrometry) and XRD (X-Ray Diffraction) measurements were used for assessment of chemical composition and structure of the coatings.The behavior of the coatings to the High Heat Flux tests (HHFT) has been investigated by using an electron beam test facility. The testing program consisted of ∼300 pulses with duration of 420 s each. The surface temperature during the tests was 1200 °C whereas the power density was about 3.5 MW/m2.

Investigations on the heat flux and impurity for the HL-2M divertor

The controllability of the heat load and impurity in the divertor is very important, which could be one of the critical problems to be solved in order to ensure the success for a steady state tokamak. HL-2M has the advantage of the poloidal field (PF) coils placed inside the demountable toroidal field (TF) coils and close to the main plasma. As a result, it is possible to make highly accurate configuration control of the advanced divertor for HL-2M. The divertor target geometry of HL-2M has been designed to be compatible with different divertor configurations to study the divertor physics and support the high performance plasma operations. In this paper, the heat loads and impurities with different divertor configurations, including the standard X-point divertor, the snowflake-minus divertor and two tripod divertor configurations for HL-2M, are investigated by numerical simulations with the SOLPS5.0 code under the current design of the HL-2M divertor geometry. The plasmas with different conditions, such as the low discharge parameters with = 0.5 MA at the first stage of HL-2M and the high parameters with = 2.0 MA during the normal operations, are simulated. The heat load profiles and the impurity distributions are obtained, and the control of the peak heat load and the effect of impurity on the core plasma are discussed. The compatibility of different divertor configurations for HL-2M is also evaluated. It is seen that the excellent compatibility of different divertor configurations with the current divertor geometry has been verified. The results show that the snowflake-minus divertor and the tripod divertor with present good performance in terms of the heat load profiles and the impurity distributions under different conditions, which may not have a big effect on the core plasma. In addition, it is possible to optimize the distance between the two X-points, , to achieve a better performance in terms of the parameters of discharges.

RETRACTED: A new perspective on hydrogen plasma equilibrium calculation along with the MHD potential energy

In this contribution we have presented two techniques for determination of plasma equilibrium position in IR-T1 tokamak: Relaxation and optical techniques. An analysis method of tokamak plasma equilibrium by a relaxation method with specified magnetic axis was presented. The degrees of freedom due to designated positions of the magnetic axis are possible by using poloidal field coil currents. Stable steady state tokamak plasma equilibria are calculated along with the Magnetohydrodynamic (MHD) potential energy. The plasma generates a plasma current which partially or fully cancels the magnetic field from the poloidal field coils. For low-temperature plasmas, the plasma current distribution is centrally peaked; for high-temperature plasmas, the plasma current has a hole. A centrally peaked current distribution in a low-temperature plasma is evolved into a current distribution with a hole by increasing the plasma pressure by Ohmic heating, RF heating, or by neutral beam injection heating. In the second technique, an image processing technique is used for the output signal of CCD camera. Then, plasma emission intensity profile and plasma position were obtained. Results were compared and discussed.

Thermomechanical simulation of WEST actively cooled upper divertor

The Tore Supra tokamak is being transformed in an x-point divertor fusion device in the frame of the WEST (W-for tungsten-Environment in Steady-state Tokamak) project, launched in support to the ITER tungsten divertor strategy. The WEST project aims to test ITER-like W monoblock Plasma Facing Units (PFU). This ITER-like divertor will be tested under long plasma discharge up to 1000s, with high heat flux density up to 20MW/m2. This paper presents the results of ANSYS thermal-structural simulations of the WEST upper divertor. The upper divertor is made of twelve 30° sectors, each one composed of 38PFU. The PFUs are actively cooled CuCrZr heat sinks and the incidence surface is coated with a thin tungsten layer. The fixing system is made of pins engaged in slotted holes. Besides, the fixing system of the sector assembly is the same as WEST lower divertor, so one upper divertor sector can be used indifferently in upper or Lower position during transitional operation phases in WEST. The total surface of the upper divertor is 8m2, and it has to be able to extract up to 4MW in steady-state, with peak heat flux values up to 8MW/m2. The fixing system was designed to handle structural loads such as forces and torques resulting from halo and eddy current, respectively, especially during disruptions and Vertical Displacement Event (VDE). The torque resulting from eddy current is first calculated thanks to an internal CEA ANSYS APDL routine. Then the ANSYS structural and thermal-structural simulations of the PFU are presented, and its design is validated thanks to A-level RCC-MRx criteria. Finally, the most conservative load case is determined in order to validate the design of the pins and the support structure.

Development of visible spectroscopy diagnostics for W sources assessment in WEST.

The present work concerns the development of a W sources assessment system in the framework of the tungsten-W environment in steady state tokamak project that aims at equipping the existing Tore Supra device with a tungsten divertor in order to test actively cooled tungsten Plasma Facing Components (PFCs) in view of preparing ITER operation. The goal is to assess W sources and D recycling with spectral, spatial, and temporal resolution adapted to the PFCs observed. The originality of the system is that all optical elements are installed in the vacuum vessel and compatible with steady state operation. Our system is optimized to measure radiance as low as 1016 Ph/(m2 s sr). A total of 240 optical fibers will be deployed to the detection systems such as the "Filterscope," developed by Oak Ridge National Laboratory (USA) and consisting of photomultiplier tubes and filters, or imaging spectrometers dedicated to Multiview analysis.

Status of the new WEST plasma control system

The WEST (W – for Tungsten – Environment in Steady state Tokamak) project is aiming at minimizing technology and operational risks of a full tungsten actively cooled divertor on ITER. It was started in 2013 and consists in transforming the Tore Supra tokamak into an X point divertor device, while taking advantage of its long discharge capability. To operate the next coming operations of WEST, new controllers are required. These developments are an opportunity to develop a new Plasma Control System (PCS) architecture featuring build-in real time handling of both plasma and plants events, thus addressing key ITER needs.The Tore Supra PCS will be refurbished including a new Pulse Schedule Editor (PSE). The main idea is to use a time segmented approach to describe the pulse schedule with a full integration of event handling both on PCS and PSE.Further to detailed requirement specifications and architecture design, two software tools were selected to define and execute a whole plasma discharge defined as a set of time segments. The PCS real-time framework (RTF) is based on an upgraded version of the AUG framework, called DCS (Discharge Control System). The PSE is the Xedit application used on WEGA and under further development for W7-X facility.This paper reports on the status of the new WEST PCS developments. The on-going developments to adapt DCS to the Tore Supra Control infrastructure networks (new real-time network, chronology system and pulse supervision) will be reported. The required preparations for the use of Xedit will be presented, mainly the appropriate formal description of the WEST control system and the implementation of the mapping between the Xedit experiment configuration and DCS configuration files.

Tomographic capabilities of the new GEM based SXR diagnostic of WEST

The tokamak WEST (Tungsten Environment in Steady-State Tokamak) will start operating by the end of 2016 as a test bed for the ITER divertor components in long pulse operation. In this context, radiative cooling of heavy impurities like tungsten (W) in the Soft X-ray (SXR) range [0.1 keV; 20 keV] is a critical issue for the plasma core performances. Thus reliable tools are required to monitor the local impurity density and avoid W accumulation. The WEST SXR diagnostic will be equipped with two new GEM (Gas Electron Multiplier) based poloidal cameras allowing to perform 2D tomographic reconstructions in tunable energy bands. In this paper tomographic capabilities of the Minimum Fisher Information (MFI) algorithm developed for Tore Supra and upgraded for WEST are investigated, in particular through a set of emissivity phantoms and the standard WEST scenario including reconstruction errors, influence of noise as well as computational time.

Progress toward steady-state tokamak operation exploiting the high bootstrap current fraction regime

Recent DIII-D experiments have increased the normalized fusion performance of the high bootstrap current fraction tokamak regime toward reactor-relevant steady state operation. The experiments, conducted by a joint team of researchers from the DIII-D and EAST tokamaks, developed a fully noninductive scenario that could be extended on EAST to a demonstration of long pulse steady-state tokamak operation. Improved understanding of scenario stability has led to the achievement of very high values of βp and βN, despite strong internal transport barriers. Good confinement has been achieved with reduced toroidal rotation. These high βp plasmas challenge the energy transport understanding, especially in the electron energy channel. A new turbulent transport model, named TGLF-SAT1, has been developed which improves the transport prediction. Experiments extending results to long pulse on EAST, based on the physics basis developed at DIII-D, have been conducted. More investigations will be carried out on EAST with more additional auxiliary power to come online in the near term.