Next-step Machines Research Articles

Validation of theory-based models for the control of plasma currents in W7-X divertor plasmas

A theory-based model for the control of plasma currents for steady-state operation in W7-X is proposed and intended for model-based plasma control. The conceptual outline implies the strength of physics-based models: it offer approaches applicable to future conditions of fusion devices or next-step machines. The application at extrapolated settings is related to the validity range of the theory model. Therefore, the predictive power of theory-based control models could be larger than for data-driven approaches and limitations can be predicted from the validity range for the prediction of bootstrap currents in W7-X. The model predicts the L/R response when density or heating power is changed. The model is based on neoclassical bootstrap current calculations and validated for different discharge conditions. While the model was found to be broadly applicable for conducted electron-cyclotron-heated discharges in W7-X, limits were found for cases when the polarization of the electron cyclotron heating was changed from X2 to O2-heating. The validity assessment attempts to quantify the potential of the derived model for model-based control in the operational space (density, heating power) of W7-X.

Innovative signal processing and data analysis methods on JET for control in the perspective of next-step devices

In the last few years, it has been realized that more sophisticated control schemes are necessary to push the boundaries of tokamak operation and the performance of reactor-like machines. In addition, JET needs to operate safely with the new metallic wall and such protection will be needed for ITER. These objectives have motivated the development, benchmark and validation of new signal processing and data analysis methods. Two new approaches for the determination of the magnetic topology in real time have been validated on an extensive database of JET discharges, including advanced tokamak scenarios. Robust methods of confinement regime identification and disruption prediction are a prerequisite for safe, general control schemes. New identifiers have been developed and their success rates exceed 99% in determining whether plasmas are in the L or H mode. A new disruption predictor is being developed and has already provided success rates higher than 90% in realistic real-time conditions. Moreover, the generalization capability of this new predictor has been confirmed by applying it to new experimental campaigns not used for the training. The success rate remains high even more than ten campaigns, or about four years, after the last one used for the training. The deployment of video cameras in real time requires the development of new image processing algorithms, which have already been implemented and validated successfully on JET for the real-time identification of hot spots with a time resolution of tens of milliseconds. A series of new feedback schemes has also been explicitly developed not much to control the plasma but to really improve the physics understanding of some phenomena. Particularly interesting are the simultaneous control of the safety factor and pressure profiles and the real-time tracking of toroidal Alfven eigenmode instabilities. These advanced feedback schemes for physics understanding often require more advanced signal processing techniques like adaptive filtering, which have already been implemented. The paper concludes by discussing the use of these real-time analysis and control developments in next-step machines such as ITER.

Ion beam analysis for fusion energy research

Proposed next-step devices for development of fusion energy present a major increase in the energy content and duration of plasmas far beyond those encountered in existing machines. This increases the importance of controlling interactions between the fusion plasma and first-wall materials. These interactions change the wall materials and strongly affect the core plasma conditions. Two critical processes are the erosion of materials by the plasma, and the redeposition of eroded material along with hydrogen isotopes from the plasma. These impact reactor design through the lifetime of plasma-facing components and the inventory of tritium retained inside the vessel. Ion beam analysis has been widely used to investigate these complex plasma-material interactions in most of the large fusion plasma experiments. The design and choice of plasma-facing materials for next-step machines rely on knowledge obtained from these studies. This paper reviews the use of ion beam analysis for fusion energy research, and shows how these studies have helped to guide the design and selection of materials for a next-step machine.

Technologies of Joining between ITER Reference Grade Beryllium and Copper Alloys by Diffusion Bonding Process

The design of plasma-facing components for ITER, as for any of the envisaged next-step machines, relies on the use of junctions for coupling the armour materials to the heat sink and cooling tubes....

Bragg spectroscopy of impurities during the JET preliminary tritium experiment

Two Bragg spectrometers shared a shielded beamline, and monitored all significant plasma impurities throughout the Joint European Torus (JET) preliminary tritium experiment. The JET Active Phase Double Crystal Monochromator and a Bragg rotor spectrometer together monitored K-shell lines of Be, C, O, and Cl, and L-shell lines from metals such as Ni and Zn, in the wavelength range 0.2–11 nm. Background radiation was recorded during and after discharges, to assess the signal-to-noise ratio, and to monitor any activation or tritium ingress. Bragg spectroscopy was shown to be suitable for all impurities in magnetically confined D–T plasmas, using a thin foil to isolate tritium, and a low f-number shielded beamline. Future improvements to the signal-to-noise ratio are discussed in the context of a next-step machine such as ITER.

Improvements of NET/ITER shielding to limit short-term radioactivity in magnet materials

The problem of neutron-induced radioactivity in magnet structural materials for fusion reactors is examined. Two near-term reactor designs (NET-DN and ITER) are considered. Long-term activation of these materials is low enough to permit their recycling after less than 25 years of intermediate decay. Short-term radioactivity, on the other hand, requires the remote handling of these components for relatively long periods after shutdown. To permit a nearly immediate direct handling, activation of magnet structural materials must be reduced. For this purpose, two possible actions are considered: materials selection and improvement of the shielding. All the candidate materials for structural use in magnet components have a comparable activation behaviour. As far as the improvement of the shielding is concerned, the most interesting problem turns out to be the modification of the inboard shield in ITER. With some slight modifications, the goal of the direct handling for magnet materials is eventually reached. Finally, the activation of magnet materials in a next-step machine like DEMO is briefly examined.

Remote radiative cooling in JT-60

Remote radiative cooling can significantly reduce the divertor plate heat flux, which may provide a solution to one of the major problems of present and next-step machines. Strong remote radiative cooling of up to 50% of the total input power of 20 MW was observed in JT-60. JT-60 divertor plasmas are dense and cold ( n ediv ≈ 2.4 × 10 20 m −3, T ediv ≈ 26 eV). Remote radiative cooling is a mixture of carbon and hydrogen radiation (+ CX). JT-60 divertor plasmas can be adequately modeled with a simple divertor model with radiative cooling.

Fuel cycle R&D for ITER — a summary of objectives and current status

Joint work on the design of the International Thermonuclear Experimental Reactor (ITER) has provided an opportunity for engineers and physicists from Japan, the EC, USA and Soviet Union to compare concepts and identify critical data needs which must be evolved over the next few years to support the design of the next-step machine for magnetic-confinement fusion. In the fuel cycle area, well-developed national programs exist, and the challenging task of the Fuel Cycle Design Unit in ITER is to promote a harmonisation of these tasks to minimise duplication and ensure gaps which would affect the development of satisfactory design concepts do not occur. The principle mechanisms for this involves: • - identification of a subset of R&D task where results are especially needed for design, • - establishment of task and subtask definitions and program milestones to facilitate task monitoring, • - reporting and information exchange as results become available. Results of this exercise to-date are described in this note.