Operating a full tungsten actively cooled tokamak: overview of WEST first phase of operation

WEST is an MA class superconducting, actively cooled, full tungsten (W) tokamak, designed to operate in long pulses up to 1000 s. In support of ITER operation and DEMO conceptual activities, key missions of WEST are: (i) qualification of high heat flux plasma-facing components in integrating both technological and physics aspects in relevant heat and particle exhaust conditions, particularly for the tungsten monoblocks foreseen in ITER divertor; (ii) integrated steady-state operation at high confinement, with a focus on power exhaust issues. During the phase 1 of operation (2017–2020), a set of actively cooled ITER-grade plasma facing unit prototypes was integrated into the inertially cooled W coated startup lower divertor. Up to 8.8 MW of RF power has been coupled to the plasma and divertor heat flux of up to 6 MW m−2 were reached. Long pulse operation was started, using the upper actively cooled divertor, with a discharge of about 1 min achieved. This paper gives an overview of the results achieved in phase 1. Perspectives for phase 2, operating with the full capability of the device with the complete ITER-grade actively cooled lower divertor, are also described

Velocity-space sensitivity of the time-of-flight neutron spectrometer at JET

The velocity-space sensitivities of fast-ion diagnostics are often described by so-called weight functions. Recently, we formulated weight functions showing the velocity-space sensitivity of the often dominant beam-target part of neutron energy spectra. These weight functions for neutron emission spectrometry (NES) are independent of the particular NES diagnostic. Here we apply these NES weight functions to the time-of-flight spectrometer TOFOR at JET. By taking the instrumental response function of TOFOR into account, we calculate time-of-flight NES weight functions that enable us to directly determine the velocity-space sensitivity of a given part of a measured time-of-flight spectrum from TOFOR

Relationship of edge localized mode burst times with divertor flux loop signal phase in JET

A phase relationship is identified between sequential edge localized modes (ELMs) occurrence times in a set of H-mode tokamak plasmas to the voltage measured in full flux azimuthal loops in the divertor region. We focus on plasmas in the Joint European Torus where a steady H-mode is sustained over several seconds, during which ELMs are observed in the Be II emission at the divertor. The ELMs analysed arise from intrinsic ELMing, in that there is no deliberate intent to control the ELMing process by external means. We use ELM timings derived from the Be II signal to perform direct time domain analysis of the full flux loop VLD2 and VLD3 signals, which provide a high cadence global measurement proportional to the voltage induced by changes in poloidal magnetic flux. Specifically, we examine how the time interval between pairs of successive ELMs is linked to the time-evolving phase of the full flux loop signals. Each ELM produces a clear early pulse in the full flux loop signals, whose peak time is used to condition our analysis. The arrival time of the following ELM, relative to this pulse, is found to fall into one of two categories: (i) prompt ELMs, which are directly paced by the initial response seen in the flux loop signals; and (ii) all other ELMs, which occur after the initial response of the full flux loop signals has decayed in amplitude. The times at which ELMs in category (ii) occur, relative to the first ELM of the pair, are clustered at times when the instantaneous phase of the full flux loop signal is close to its value at the time of the first ELM

Current measurements in the scrape-off layer of Tore Supra with the new CIEL limiter

The current flowing between the new toroidal pump limiter of Tore Supra and the inner wall components was investigated. The response of the current to a change of the wall clearance as well as to a variation of the poloidal and toroidal magnetic field was studied. A simple estimate of the thermoelectric currents resulting from the radial temperature gradient is presented. We find, that the dependance of the measured current on the wall clearance can be reproduced by prescribing a deviation of the plasma potential from the sheath potential. (C) 2003 Elsevier Science B.V. All rights reserved

Design and development of ITER high-frequency magnetic sensor

High-frequency (HF) inductive magnetic sensors are the primary ITER diagnostic set for Toroidal Alfvén Eigenmodes (TAE) detection, while they also supplement low-frequency MHD and plasma equilibrium measurements. These sensors will be installed on the inner surface of ITER vacuum vessel, operated in a harsh environment with considerable neutron/nuclear radiation and high thermal load. Essential components of the HF sensor system, including inductive coil, electron cyclotron heating (ECH) shield, electrical cabling and termination load, have been designed to meet ITER measurement requirements. System performance (e.g. frequency response, thermal conduction) has been assessed. A prototyping campaign was initiated to demonstrate the manufacturability of the designed components. Prototypes have been produced according to the specifications. A series of lab tests have been performed to examine assembly issues and validate electrical and thermo-mechanical aspects of the design. In-situ microwave radiation test has been conducted in the MISTRAL test facility at IPP-Greifswald to experimentally examine the microwave shielding efficiency and structural integrity of the ECH shield. Low-power microwave attenuation measurement and scanning electron microscopic inspection were conducted to probe and examine the quality of the metal coating on the ECH shield

Prototyping and testing of the Continuous External Rogowski ITER magnetic sensor

The measurement of the plasma current in ITER plays an outstanding role as it is part of the machine protection and is a safety-relevant measurement: it will be used in relation with regulatory limits to show that the operation remains within the safe envelope defined in the ITER license. The Continuous External Rogowski (CER) is an inductive sensor designed for current measurements and located in the casing of 3 Toroidal Field Coils (TFCs). After the completion of the design of the CER, 4 prototypes of the sensor were manufactured and R&D activities were performed under a Grant with the European Domestic Agency (F4E-GRT-012). The work was carried out between 2010 and 2011 by the ITERMAG consortium comprising 3 laboratories: CRPP (Switzerland) as leader, CEA (France) and RFX (Italy). The R&D campaign on CER prototypes consisted in the measurement of about 100 parameters to characterize the CER in terms of electrical, thermal, mechanical and also of vacuum compatibility. From these results, electromagnetic modeling of the CER response was performed. It is demonstrated that the CER fulfills ITER requirements. However, the vacuum compatibility of the prototype has to be improved and solutions to cope with this issue are proposed. (C) 2013 Elsevier B.V. All rights reserved

Design and development of ITER high-frequency magnetic sensor

High-frequency (HF) inductive magnetic sensors are the primary ITER diagnostic set for Toroidal Alfvén Eigenmodes (TAE) detection, while they also supplement low-frequency MHD and plasma equilibrium measurements. These sensors will be installed on the inner surface of ITER vacuum vessel, operated in a harsh environment with considerable neutron/nuclear radiation and high thermal load. Essential components of the HF sensor system, including inductive coil, electron cyclotron heating (ECH) shield, electrical cabling and termination load, have been designed to meet ITER measurement requirements. System performance (e.g. frequency response, thermal conduction) has been assessed. A prototyping campaign was initiated to demonstrate the manufacturability of the designed components. Prototypes have been produced according to the specifications. A series of lab tests have been performed to examine assembly issues and validate electrical and thermo-mechanical aspects of the design. In-situ microwave radiation test has been conducted in the MISTRAL test facility at IPP-Greifswald to experimentally examine the microwave shielding efficiency and structural integrity of the ECH shield. Low-power microwave attenuation measurement and scanning electron microscopic inspection were conducted to probe and examine the quality of the metal coating on the ECH shield

The new magnetic diagnostics in the WEST tokamak

International audienceThe WEST tokamak consists of a major upgrade of the superconducting medium size tokamak Tore Supra aiming at testing ITER divertor components. Such modification has required rebuilding a full set of magnetic diagnostics. The project was started in 2013 and completed in 2016. The diagnostic consists of a set of 469 sensors (421 pick-up coils, 36 flux loops, and 12 Rogowski coils) installed in the WEST vacuum vessel. New analog integrators have been developed in order to obtain the magnetic field and flux from the raw signal of the sensors. During the startup phase of WEST, plasma currents of the order of a few kilo amperes were measured despite much larger current of the order of hundreds of kilo amperes flowing in nearby conducting structures. The diagnostic is now fully operational and exhibits a noise level of about 0.5 mT on the magnetic field, and 2.0 mWb on flux loops allowing identifying the plasma boundary with an accuracy of a few millimeters on a 2 ms time cycle