The new magnetic diagnostics in the WEST tokamak

Abstract

The 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.

Renewal of the interfero-polarimeter diagnostic for WEST

“…The diagnostic has routinely been running during the 2017-2018 WEST restart campaigns and helped to improve the plasma performances up to a 10 s duration and a 3 × 10 19 m-2 central LID. With the help of the last closed surface (LCS) that is given in real time by the magnetics diagnostic [9] -figure 16a shows an example for pulse 53259 -, the distance of each chord in the plasma can be estimated to calculate the averaged LID. Moreover, the LID integration of the 8 almost horizontal channels along the vertical line defined by the centre of the plasma enables to calculate the volume average density by dividing by the surface, assuming that the LID is null on the LCS.…”

Effective Area Measurements of Magnetic Pick-Up Coil Sensors for RFX-mod2

“…The nominal value of A s defined by the design, is inevitably subjected to variations, mainly due to the manufacturing process and subsequent thermal cycles during operation. Even if the associated errors due to this processes can be contained in a reasonable range (say within 5%), in the case of fusion relevant experiments each single measurement requires a nominal error of around 1% [15]. Obviously this goal can be achieved only by an individual calibration for each sensor.…”

An overview of the numerical methods for tokamak plasma equilibrium computation implemented in the NICE code

“…This mode of NICE also known as VacTH is routinely used at WEST for plasma shape identication and control during a shot [27]. Here is provided a typical example of plasma boundary reconstruction for a WEST-like equilibrium.…”

“…It makes intensive use of Newton method and SQP method to solve non linear problems. NICE is used routinely for the WEST tokamak operation [27]. NICE is adapted to the IMAS (ITER Modelling and Analysis Suite [21,31]) format which makes it usable on many dierent tokamaks [8].…”