Steady-state Tokamak Research Articles

Diagnostics and control for the steady state and pulsed tokamak DEMO

The present paper is devoted to a first assessment of the DEMO diagnostics systems and controls in the context of pulsed and steady state reactor design under study in Europe. In particular, the main arguments treated are: (i) The quantities to be measured in DEMO and the requirements for the measurements; (ii) the present capability of the diagnostic and control technology, determining the most urgent gaps, and (iii) the program and strategy of the research and development (R&D) needed to fill the gaps. Burn control, magnetohydrodynamic stability, and basic machine protection require improvements to the ITER technology, and moderated efforts in R&D can be dedicated to infrared diagnostics (reflectometry, electron cyclotron emission, polarimetry) and neutron diagnostics. Metallic Hall sensors appear to be a promising candidate for magnetic measurements in the high neutron fluence and long/steady state discharges of DEMO.

The WEST project: validation program for WEST tungsten coated plasma facing components

The W—for tungsten—Environment in Steady-state Tokamak (WEST) project is based on an upgrade of the Tore Supra tokamak from a carbon limiter to an X-point divertor device. A new set of actively cooled tungsten coated plasma facing components will cover a part of the vessel to provide a fully metallic environment. This paper deals with the validation program performed for tungsten coatings (≥15 μm) on a CuCrZr substrate. The first step was dedicated to the qualification under high heat flux tests of the coating on small inertially cooled samples. To study the thermal behavior and the non-uniformity, the second step was dedicated to the validation of the coating on large inertially cooled samples with geometry and shape (540 × 120 mm) representative of the WEST coated components. The last step was dedicated to the optimization of the coating and to the high heat flux tests up to 10.5 MW m−2 on relevant coated actively cooled prototypes. Non-uniformity and thickness of the coating (15 and 30 μm) correspond to specifications. As no delamination was observed, coatings of 15 and 30 μm were qualified with regard to their application on WEST coated components. In order to decrease the risk of coating delamination under thermal loading, it was decided to cover the upper divertor and baffle targets with the thinnest coating option of 15 μm.

Experimental test of the system of vertical and longitudinal lithium limiters on T-11M tokamak as a prototype of plasma facing components of a steady-state fusion neutron source

A new functional model of the prototype of closed Li circuit for protection of the chamber wall was tested in T-11M tokamak by simultaneous use of the vertical Li limiter as an emitter of Li and a new longitudinal Li limiter as its collector. Such technological scheme can be suggested for the steady-state fusion neutron source on the tokamak basis. During plasma shots the cryogenic target of T-11M collected Li flow emitted by the vertical capillary Li limiter almost completely (up to 80%). These Li and hydrogen isotopes were captured and extracted outside the tokamak vacuum chamber without venting of the vessel which is a key requirement for the use of Li in the steady-state tokamak reactor.

Use of virtual reality for optimizing the life cycle of a fusion component

Efficient development of a complex system such as a fusion component needs a stringent integration of standard and new constraints. For example, compared to the previous fusion experimental devices, remote handling (RH) and safety requirements are in ITER key parameters which must be integrated since the earliest design. For optimizing such integration studies, CEA, IRFM decided in 2010 to implement the use of virtual reality (VR) tools during the life cycle (from design to operation) of a fusion component. This paper describes a first feedback of such use for fusion engineering purposes.After a short overview of the CEA, IRFM VR platform capabilities, three main uses will be described: design review, simulation of remote handling and hands-on operations, with man in the loop.The Design review mode was intensively used within the framework of a fruitful collaboration with ITER design Integration Team. This mode, fully compatible with CAD software, enables scale one data visualization with stereoscopic rendering. It improves the efficiency in detecting inconsistencies inside models and machine sub-system design optimization needs. Several accessibility cases of major Safety Important Components (SIC-1) were studied giving important requirements to the design at an early stage.CEA, IRFM, in close collaboration with expertise of CEA, LIST for VR simulation software, applies VR technologies for designing RH maintenance scenario for ITER Test Blanket System (TBS) and Ion cyclotron Resonance Heating (ICRH) Port Plugs. RH compatibility studies using VR pointed out major design drivers while helping to propose credible solution. VR platform is intensively used in the design of WEST (Tungsten (W) Environment Steady-state Tokamak) components and assembly studies, providing important information about the feasibility of assembly processes, optimization of physical mock-ups and ergonomic posture and gestures of operator.Finally, new perspectives, as the integration of safety constraints (dose calculation) will be described, demonstrating the powerful of VR tools at different stages of the component lifecycle.

Experimental advanced superconducting tokamak

Experimental advanced superconducting tokamak (EAST) is the first fully superconducting tokamak device in the world, designed and built in our own country. As a ‘medium-size superconducting tokamak, EAST has successfully bridged the gap between the short-pulse medium-size tokamak and the long-pulse large-size tokamak. EAST is equipped with the advanced international thermonuclear experimental reactor (ITER) based technology, and can realize 1000-s-long pulse discharges. In the next 10 years, EAST will be one of the few tokamaks in the world used as an experimental platform, which could be used for studying plasma physics and engineering technology under conditions of high-performance long-pulse discharges. During the past few years, EAST has achieved 400-s-long pulses in ‘low confinement mode plasma of divertor configuration, and repeatable 30-s-long pulses in ‘high confinement mode, setting a new world record. EAST will be used for solving key physical and engineering issues related to the high-performance steady state operation at the ITER and future fusion reactors. The studies on EAST will provide an important basis for the design and operation of future Chinese fusion reactors, and will provide the technological and physical basis for a steady-state, high-efficiency and safe tokamak fusion reactor.

Application of virtual reality tools for assembly of WEST components: Comparison between simulations and physical mockups

The WEST project (Tungsten (W) Environment in Steady state Tokamak) is an upgrade of the existing fusion machine, Tore Supra. The goal is to equip the tokamak with a fully cooled tungsten divertor and to transform the machine in a test platform open to all ITER partners.The main assembly challenge of this project consists of an implementation of two magnet systems, called divertors, with an accuracy of 1mm. Indeed, each divertor has about 4m as diameter and has a heavy weight of 10 tons; also it introduces piece by piece in the original vessel through tight ports then assembled inside. To ensure a perfect fitting between these new components and a very constrained environment, it is necessary to use the latest CAD technologies available.Beyond conventional CAD tools, the virtual reality (VR) room of the institute provides several useful tools. Thanks to the 185″ stereoscopic 3D screen and a force feedback arm linked to clash detection software developed by the CEA LIST, a new way to carry out design and assembly studies was performed. In order to improve VR results, metrology data (3D scan) enhance simulations. Therefore, it becomes possible to be aware of the real size of a component and future difficulties in assembling it.At last, performance of such simulations is evaluated and compared to physical mockup in order to bring enhancement to the VR tools, before to be compared to the real operations on Tore Supra. The aim is to build a design tool that helps the designer since early stage of the design of complex systems, taking into consideration integration, assembly and maintenance aspects while reducing costs and schedule of a project.

Manufacturing monitoring and mock-ups validation of the WEST divertor structure and coils

In order to fully validate “ITER-like” actively water cooled tungsten plasma facing units, the implementation of an axisymmetric divertor structure in the Tokamak Tore-Supra has been studied. With this major upgrade, the so-called WEST (Tungsten Environment in Steady state Tokamak), Tore-Supra will be able to address the issues of long plasma discharges using a tungsten divertor based on monoblock targets.The divertor structure and coils assembly are made up of two stainless steel casings containing a copper winding pack cooled by a pressurized hot water circuit (up to 180°C, 4MPa) in which a total divertor current of up to 16×13kA is circulating in steady state. The conductor is electrically insulated and wedged inside the casing in order to be mechanically protected.The divertor which is designed to perform steady state plasma operation (up to 1000s), must sustain harsh environmental conditions in terms of ultra light vacuum conditions, electromagnetical loads and electrical insulation (5kV ground voltage) under high temperature (180°C).Therefore, a feasibility study of such a complex structure has been performed. It implied activities on a scale one dummy coil, such as installation, assembly issues and representative tests (electric, thermal and hydraulic).The manufacturing of the divertor structure, which is a large assembly of 4-m diameter representing a total weight of around 20tonnes, started in the second half of 2013 and is expected to be delivered by the end of 2014.The paper will illustrate the technical developments and tests performed during 2013 and beginning of 2014 in order to fully validate the design concept before the industrial phase. The manufacturing methods proposed by the contractor in order to fulfil the technical requirements will be also addressed.

Development of 28 GHz Gyrotron for Cooperative ECH Study

At the Plasma Research Center at University of Tsukuba, development of megawatt gyrotrons is being performed as a collaborative electron cyclotron heating (ECH) study with some research organizations. A 28 GHz 1 MW 1 s gyrotron has been developed to upgrade the GAMMA 10/PDX ECH systems. To improve the oscillation efficiency in high current regions, the magnetron injection gun (MIG) of the 28 GHz gyrotron has been modified. Output power of 1.25 MW has been achieved with this gyrotron. For the first step of the collaborative research between Tsukuba University and Kyushu University, the Tsukuba 28 GHz gyrotron was adapted to the Q-shu University Experiment with Steady-State Spherical Tokamak (QUEST) ECH system, and the plasma heating and current drive effect were demonstrated. We obtained successful results, including an electron cyclotron–driven plasma current of 66 kA in the QUEST plasma experiment. For the next step of the collaborative research, the design targets of a 2 MW 3 s and 0.4 MW continuous wave have been achieved in a design study of a new 28 GHz gyrotron.

JT-60SA superconducting magnet system

The most distinctive feature of the superconducting magnet system for JT-60SA is the optimized coil structure in terms of the space utilization as well as the highly accurate coil manufacturing, thus meeting the requirements for the steady-state tokamak research: a conceptually new outer inter-coil structure separated from the casing is introduced to the toroidal field coils to realize their slender shape, allowing large-bore diagnostic ports for detailed plasma measurements. A method to minimize the manufacturing error of the equilibrium-field coils has been established, aiming at the precise plasma shape/position control. A compact butt-joint has been successfully developed for the Central Solenoid, which allows an optimized utilization of the limited space for the Central Solenoid to extend the duration of the plasma pulse.

Development of DEMO-FNS tokamak for fusion and hybrid technologies

The history of fusion–fission hybrid systems based on a tokamak device as an extremely efficient DT-fusion neutron source has passed through several periods of ample research activity in the world since the very beginning of fusion research in the 1950s. Recently, a new roadmap of the hybrid program has been proposed with the goal to build a pilot hybrid plant (PHP) in Russia by 2030. Development of the DEMO-FNS tokamak for fusion and hybrid technologies, which is planned to be built by 2023, is the key milestone on the path to the PHP. This facility is in the phase of conceptual design aimed at providing feasibility studies for a full set of steady state tokamak technologies at a fusion energy gain factor Q ∼ 1, fusion power of ∼40 MW and opportunities for testing a wide range of hybrid technologies with the emphasis on continuous nuclide processing in molten salts. This paper describes the project motivations, its current status and the key issues of the design.

High internal inductance for steady-state operation in ITER and a reactor

Increased confinement and ideal stability limits at relatively high values of the internal inductance () have enabled an attractive scenario for steady-state tokamak operation to be demonstrated in DIII-D. Normalized plasma pressure in the range appropriate for a reactor has been achieved in high elongation and triangularity double-null divertor discharges with at , near the ideal kink stability limit calculated without the effect of a stabilizing vacuum vessel wall, with the ideal-wall limit still higher at . Confinement is above the H-mode level with . At , the current is overdriven, with bootstrap current fraction , noninductive current fraction and negative surface voltage. For ITER (which has a single-null divertor shape), operation at is a promising option with and the remaining current driven externally near the axis where the electron cyclotron current drive efficiency is high. This scenario has been tested in the ITER shape in DIII-D at , so far reaching and at with performance appropriate for the ITER Q=5 mission, . Modeling studies explored how increased current drive power for DIII-D could be applied to maintain a stationary, fully noninductive high discharge. Stable solutions in the double-null shape are found without the vacuum vessel wall at , and , and at with the vacuum vessel wall.

Major upgrade of the articulated inspection arm control system to fulfill daily operation requirements

An articulated inspection arm (AIA) has been developed by CEA for visual inspection between pulses inside the Tore Supra tokamak vacuum vessel without breaking temperature and vacuum conditions. The eight meters length robot is composed of a shuttle and six articulated segments with a video camera at its end. A demonstration prototype has been achieved in 2008 at Tore Supra (Gargiulo, 2007; Houry, 2008; Perrot, 2003).A project to upgrade the AIA into a fully operational robot has been undertaken by IRFM and ASIPP in an Associated Laboratory. It will be in operation first in the EAST machine and afterwards in Tore Supra in its WEST (W/Tungsten Environment in Steady-state Tokamak) configuration where it is of paramount importance to survey possible degradation of W component surface.The control system of the robot has been extensively upgraded. The effort has been focused on three areas: (1) improvement of the arm position accuracy, (2) increase of the operational robustness, (3) use of a powerful graphical user interface including simulation of trajectories and robot deployment capabilities in a 3D viewer environment.The aim of this paper is to detail the architecture of the AIA control system.

Modelling of pulsed and steady-state DEMO scenarios

Scenario modelling for the demonstration fusion reactor (DEMO) has been carried out using a variety of simulation codes. Two DEMO concepts have been analysed: a pulsed tokamak, characterized by rather conventional physics and technology assumptions (DEMO1) and a steady-state tokamak, with moderately advanced physics and technology assumptions (DEMO2). Sensitivity to impurity concentrations, radiation, and heat transport models has been investigated. For DEMO2, the impact of current driven non-inductively by neutral beams has been studied by full Monte Carlo simulations of the fast ion distribution. The results obtained are a part of a more extensive research and development (R&D) effort carried out in the EU in order to develop a viable option for a DEMO reactor, to be adopted after ITER for fusion energy research.

On the interpretation of high-resolution x-ray spectra from JET with an ITER-like wall

The upgraded x-ray crystal spectrometer KX1 on the Joint European Torus (JET) can now measure the M x-ray lines from tungsten with sufficiently high resolution to evaluate how much tungsten may sputter from the plasma-facing tungsten wall planned for the International Thermonuclear Experimental Reactor (ITER). However, a test run on JET found that the L x-ray lines of a molybdenum impurity, which happen to occur in the targeted wavelength region of 5.00–5.35 Å, must be taken into account together with the radiation from the M x-ray lines of tungsten to match the high-resolution spectra. Such detailed radiation modeling is expected to be needed for ASDEX Upgrade tokamak, and for other tokamaks such as ITER and tungsten (W) Environment in Steady-state Tokamak (WEST), which will implement tungsten plasma-facing components and a high-resolution spectrometer to keep track of with a similar x-ray diagnostics.

WEST Physics Basis

With WEST (Tungsten Environment in Steady State Tokamak) (Bucalossi et al 2014 Fusion Eng. Des. 89 907–12), the Tore Supra facility and team expertise (Dumont et al 2014 Plasma Phys. Control. Fusion 56 075020) is used to pave the way towards ITER divertor procurement and operation. It consists in implementing a divertor configuration and installing ITER-like actively cooled tungsten monoblocks in the Tore Supra tokamak, taking full benefit of its unique long-pulse capability. WEST is a user facility platform, open to all ITER partners. This paper describes the physics basis of WEST: the estimated heat flux on the divertor target, the planned heating schemes, the expected behaviour of the L–H threshold and of the pedestal and the potential W sources. A series of operating scenarios has been modelled, showing that ITER-relevant heat fluxes on the divertor can be achieved in WEST long pulse H-mode plasmas.

Mechanical design and thermo-hydraulic simulation of the infrared thermography diagnostic of the WEST tokamak

The WEST (Tungsten (W) Environment in Steady state Tokamak) project is a partial rebuild of the Tore Supra tokamak to make it an X-point metallic environment machine aimed at testing ITER technologies in relevant plasma environment. For the safe operation of the WEST tokamak, infra-red (IR) thermography is a crucial diagnostic as it is a sound and reliable way to detect hotspots or abnormal heating patterns on the plasma facing components (PFCs). Thus WEST will be fitted with middle/short-IR (1.5–2μm or 3–5μm) cameras in the upper port plugs to get a full view of the critical PFCs (in particular the new lower divertor) and radio-frequency (RF) heating antennas and one camera at the equatorial level to monitor the new upper divertor and the first wall.This paper describes the design of the up-to-date optical system along with the hydraulic analysis and the thermal and mechanical finite element analysis conducted to ensure adequate heat extraction capabilities. Boundary conditions and simulation results will be presented and discussed as well as technological solutions retained.

Demonstration of real-time control for poloidal beta in KSTAR

Sustaining the plasma in a stable and a high performance condition is one of the important control issues for future steady state tokamaks. In the 2014 KSTAR campaign, we have developed a real-time poloidal beta (βp) control technique and carried out preliminary experiments to identify its feasibility. In the control system, the βp is calculated in real time using the measured diamagnetic loop signal, and compared with the target value leading to the change of the neutral beam (NB) heating power using a feedback PID control algorithm. To match the requested power of NB which is operated with constant voltage, the on-time periods of the intervals were adjusted as the ratio of the required power to the maximum achievable one. This paper will present the overall procedures of the βp control, the βp estimation process and NB operation scheme implemented in the plasma control system (PCS), and the analysis on the preliminary experimental results.

Design and optimization of the WEST ICRH antenna front face components based on thermal and hydraulic analysis

The WEST (Tungsten (W) Environment in Steady-state Tokamak) is an upgrade of Tore-Supra (TS) which aims it into an X-point magnetic configuration tokamak equipped with an actively cooled tungsten divertor. To be a platform of ITER technologies of high heat flux components testing, three sets of Ion Cyclotron Resonant Heating (ICRH) antennas have been designed to inject 9MW during 30s or 3MW during 1000s.The antenna design is based on a load resilient prototype successfully tested in Tore Supra in 2007. In order to allow continuous-wave (CW) operations, the mechanical design of the WEST ICRH antenna is emphasized on its cooling performances by designing fully active cooling structure. Two kinds of cooling water loops are used, with temperature and pressure of 70°C/30bar and 25°C/5.2bar, respectively. The hot water loop is used for the Faraday screen (FS) and the housing box (HB), while the cold water loop is used for the straps, the matching capacitors and the impedance transformer. To enhance the heat removal ability and control the pressure drop, the cooling channels in the FS and HB are drilled directly and parallel connected as much as possible.By performing the hydraulic–thermal analysis, the lack of cooling efficiency was found in the front face of lateral collector where 1MW/m2 is imposed and fluid dead zones were found in some of the bars. After optimization, the cooling performance of the cooling circuit increased significantly. With a mass flow rate of 2.5kg/s, the total pressure drop is 3.1bar, and the peak temperatures on the FS and HB are 500°C and 261°C, respectively. Besides, no cavitation is found in the total cooling circuit.

Plasma facing components integration studies for the WEST divertor

In the context of the Tokamak Tore-Supra evolution, the CEA aims at transforming it into a test bench for ITER actively cooled tungsten (ACW) plasma facing components (PFC). This project named WEST (Tungsten Environment in Steady state Tokamak) is especially focused on the divertor target. The modification of the machine, by adding two axisymmetric divertors will make feasible an H-mode with an X-point close to the lower divertor. This environment will allow exposing the divertor ACW components up to 20MW/m2 heat flux during long pulse. These specifications are well suited to test the ITER-like ACW target elements, respecting the ITER design.One challenge in such machine evolution is to integrate components in an existing vacuum vessel in order to obtain the best achievable performance. This paper deals with the design integration of ITER ACW target elements into the WEST environment considering magnetic, electric, thermal and mechanical loads. The feasibility of installation and maintenance has to be strongly considered as these PFC could be replaced several times. The ports size allows entering a 30° sector of pre-installed tungsten targets which will be plugged as quickly and easily as possible. The main feature of steady state operation is the active cooling, which leads to have many embedded cooling channels and bulky pipes on the PFC module including many connections and sealings between vacuum and water channels. The 30° sector design is now finalized regarding the ITER ACW elements specifications. No major modifications are expected.

Langmuir probes design for the actively cooled divertor baffle in WEST

The WEST project (W-Environment in Steady-State Tokamak) aims to transform the Tore Supra limiter configuration to an x-point divertor, providing a test bed for ITER-like plasma-facing components (actively cooled W monoblocs) under high heat flux, steady-state plasma irradiation. The lower divertor includes an actively cooled, W-coated CuCrZr baffle to provide neutral compression and improve particle exhaust. As part of the new diagnostic equipment of Tore Supra within WEST project, a set of Langmuir probes will find place on the baffle in order to provide plasma flux and electron temperature measurements for physics studies and real-time machine protection functions during steady-state discharges. On the baffle top surface, irradiation coming from the plasma, energetic ripple-ions losses, photons and energetic neutrals from charge exchange reactions produce power fluxes up to 3MW/m2, representing a challenge for the Langmuir probes operating conditions. In this paper Copper–Chrome–Zirconium (CuCrZr) cylindrical probe concept design is proposed. Finite element thermo-mechanical analysis (FEA) confirmed the consistency of this solution under the steady-state plasma condition in the worst case (highest thermal load).