The consequences of tungsten (W) melting on divertor lifetime and plasma operation are high priority issues for ITER. Sustained and controlled W-melting experiment has been achieved for the first ...time in WEST on a poloidal sharp leading edge of an actively cooled ITER-like plasma facing unit (PFU). A series of dedicated high power steady state plasma discharges were performed to reach the melting point of tungsten. The leading edge was exposed to a parallel heat flux of about 100 MW.m-2 for up to 5 s providing a melt phase of about 2 s without noticeable impact of melting on plasma operation (radiated power and tungsten impurity content remained stable at constant input power) and no melt ejection were observed. The surface temperature of the MB was monitored by a high spatial resolution (0.1 mm/pixel) infrared camera viewing the melt zone from the top of the machine. The melting discharge was repeated three times resulting in about 6 s accumulated melting duration leading to material displacement from three similar pools. Cumulated on the overall sustained melting periods, this leads to excavation depth of about 230 m followed by a re-solidified tungsten bump of 200 m in the JxB direction.
Abstract The consequences of tungsten (W) melting on divertor lifetime and plasma operation are high priority issues for ITER. Sustained and controlled W-melting experiment has been achieved for the ...first time in WEST on a poloidal sharp leading edge of an actively cooled ITER-like plasma facing unit (PFU). A series of dedicated high power steady state plasma discharges were performed to reach the melting point of tungsten. The leading edge was exposed to a parallel heat flux of about 100 MW.m −2 for up to 5 s providing a melt phase of about 2 s without noticeable impact of melting on plasma operation (radiated power and tungsten impurity content remained stable at constant input power) and no melt ejection were observed. The surface temperature of the MB was monitored by a high spatial resolution (0.1 mm/pixel) infrared camera viewing the melt zone from the top of the machine. The melting discharge was repeated three times resulting in about 6 s accumulated melting duration leading to material displacement from three similar pools. Cumulated on the overall sustained melting periods, this leads to excavation depth of about 230 μ m followed by a re-solidified tungsten bump of 200 μ m in the JxB direction.
A selection of achievements and first physics results are presented of the European Integrated Tokamak Modelling Task Force (EFDA ITM-TF) simulation framework, which aims to provide a standardized ...platform and an integrated modelling suite of validated numerical codes for the simulation and prediction of a complete plasma discharge of an arbitrary tokamak. The framework developed by the ITM-TF, based on a generic data structure including both simulated and experimental data, allows for the development of sophisticated integrated simulations (workflows) for physics application. The equilibrium reconstruction and linear magnetohydrodynamic (MHD) stability simulation chain was applied, in particular, to the analysis of the edge MHD stability of ASDEX Upgrade type-I ELMy H-mode discharges and ITER hybrid scenario, demonstrating the stabilizing effect of an increased Shafranov shift on edge modes. Interpretive simulations of a JET hybrid discharge were performed with two electromagnetic turbulence codes within ITM infrastructure showing the signature of trapped-electron assisted ITG turbulence. A successful benchmark among five EC beam/ray-tracing codes was performed in the ITM framework for an ITER inductive scenario for different launching conditions from the equatorial and upper launcher, showing good agreement of the computed absorbed power and driven current. Selected achievements and scientific workflow applications targeting key modelling topics and physics problems are also presented, showing the current status of the ITM-TF modelling suite.
Electricity production through the control of nuclear fusion reactions is a promising option to meet the increasing demand for clean energy sources. The magnetically confined approach based on the ...tokamak is presently investigated through the international ITER project. ITER is the largest fusion development project aimed at demonstrating the capability of generating 500MW of fusion power with an input of 50MW (Q=10) for long durations (> 5 min) by controlling fusion reactions. The project requires a multidisciplinary development approach from materials to cooling system and, of course, control. The high temperature plasmas generated in a tokamak require continuous control to guarantee the achievement and sustainment of the required performance to demonstrate a high fusion energy gain. This requires an integrated set of controls with multiple inputs and multiple outputs, sharing the available actuators supported by an effective system to avoid, where possible, conditions that could ultimately, compromise the integrity of the device. This role is in the scope of the Plasma Control System (PCS). A lot of experience has been already gained from present fusion devices, but the scale of ITER and future reactors requires further developments in the field. The control in ITER is similar to other devices, but more complexity is required by the new physics to be explored in ITER scenarios, the robustness against larger loads (e.g. high neutron flux rate) and the systematic application of multiple controls sharing resources or by narrow operational spaces/proximity to instability boundaries associated with high Q operation. Moreover, the design of the PCS requires a model-based approach, considering that the machine for which it is designed is under construction. On the other hand, while the control schemes will be validated through simulations ahead of operation, the PCS will only be fully tested and commissioned through ITER operation itself. In this paper, an overview of a PCS and its design is presented followed by recent development targeting the ITER PCS for First Plasma and future operations. Constraints, limitations and challenges will be illustrated and the adopted solutions presented.
► Real-time event handling requires extended functionalities of pulse schedule editors and plasma control systems ► A new pulse schedule editor, conceived for parameterization of systematic ...off-normal event handling, is described ► A global, generic approach on off-normal event handling is highlighted ► The functional architecture of an off-normal event handling oriented plasma control system is discussed ► The main objects of the pulse schedule editor are the segment-descriptor object and the scenario-descriptor object.
Coping with unexpected events is an important issue of nuclear fusion experiments. The future machines, characterized by very long plasma discharges and actively cooled metallic plasma-facing components, will require a systematic intervention in real time, in order to maximize the performance and protect the investment. The real-time management of events will require extending the functionalities of the current pulse schedule editors with the possibility of using reference waveforms provided with acceptability margins and setting up advanced mitigation strategies and event countermeasures. With this purpose, a new pulse schedule editor, based on a time-segment approach for the preparation of experimental scenarios, is being conceived on Tore Supra, together with a new plasma control system. This paper will report on their conceptual design and give account of the preliminary results of a feasibility study currently under way in order to prepare a possible implementation on Tore Supra.
The WEST platform consists in a major upgrade of Tore Supra towards a steady-state tungsten (W) diverted tokamak. In support of this, significant developments are performed on the measurement systems ...(diagnostics); the control, data access and communication (CODAC); the plasma control system (PCS), the monitoring and protection of the first wall and modelling to prepare the restart of the plasma. Thanks to collaboration agreements already in force, most of the developments and some hardware procurements are performed with the help of several international partners. This paper discusses the present status of developments regarding the measurements and control for the WEST project. In particular, the integration of about 50 diagnostics in ports is completed, and their in-vessel and ex-vessel installation is underway. The refurbishment of the CODAC network architecture has been completed. The development of the new acquisition units based on PXI and of the Plasma Control System (PCS) is ongoing and some units are already available. In parallel, to prepare the plasma restart, the development of plasma magnetic and kinetic controllers has been performed on simplified plant and actuator models and plasma models.
Fusion power is the most significant prospects in the long-term future of energy in the sense that it composes a potentially clean, cheap, and unlimited power source that would substitute the ...widespread traditional nonrenewable energies, reducing the geographical dependence on their sources as well as avoiding collateral environmental impacts. Although the nuclear fusion research started in the earlier part of 20th century and the fusion reactors have been developed since the 1950s, the fusion reaction processes achieved have not yet obtained net power, since the generated plasma requires more energy to achieve and remain in necessary particular pressure and temperature conditions than the produced profitable energy. For this purpose, the plasma has to be confined inside a vacuum vessel, as it is the case of the Tokamak reactor, which consists of a device that generates magnetic fields within a toroidal chamber, being one of the most promising solutions nowadays. However, the Tokamak reactors still have several issues such as the presence of plasma instabilities that provokes a decay of the fusion reaction and, consequently, a reduction in the pulse duration. In this sense, since long pulse reactions are the key to produce net power, the use of robust and fast controllers arises as a useful tool to deal with the unpredictability and the small time constant of the plasma behavior. In this context, this article focuses on the application of robust control laws to improve the controllability of the plasma current, a crucial parameter during the plasma heating and confinement processes. In particular, a variable structure control scheme based on sliding surfaces, namely, a sliding mode controller (SMC) is presented and applied to the plasma current control problem. In order to test the validity and goodness of the proposed controller, its behavior is compared to that of the traditional PID schemes applied in these systems, using the RZIp model for the Tokamak à Configuration Variable (TCV) reactor. The obtained results are very promising, leading to consider this controller as a strong candidate to enhance the performance of the PID-based controllers usually employed in this kind of systems.
This article deals with the identification of a space and time dependent material thermal diffusivity. Such parameter is involved in heat transfers described by partial differential equations. An ...iterative regularization method based on a conjugate gradient algorithm is implemented. Such approach is attractive in order to efficiently deal with measurement noises and model errors. Numerical results are illustrated according to several simulations.
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