METIS is a numerical code aiming at fast full tokamak plasma analyses and predictions. It combines 0D scaling-law normalised heat and particle transport with 1D current diffusion modelling and 2D ...equilibria. It contains several heat, particle and impurities transport models, as well as heat, particle, current and momentum sources, which allow faster than real time scenario simulations. This paper gives a first comprehensive description of the METIS suite: overall structure of the code, main available models, details on the simulation workflow and numerical implementation. Some examples of applications to the analysis of experimental discharges and the predictions of ITER scenarios are also given.
The WEST project recently launched at Cadarache consists in transforming Tore Supra in an X-point divertor configuration while extending its long pulse capability, in order to test the ITER divertor ...technology. The implementation of a full tungsten actively cooled divertor with plasma facing unit representative of ITER divertor targets will allow addressing risks both in terms of industrial-scale manufacturing and operation of such components. Relevant plasma scenarios are foreseen for extensive testing under high heat load in the 10–20MW/m2 range and ITER-like fluences (1000s pulses). Plasma facing unit monitoring and development of protection strategies will be key elements of the WEST program.
WEST is scheduled to enter into operation in 2016, and will provide a key facility to prepare and be prepared for ITER.
In this paper, new results of plasma ι-profile and β control on TCV, using total plasma current Ip, and ECCD (Electron Cyclotron heating and Current Drive) heating source have been discussed. The ...control model is governed by the resistive diffusion equation coupled with the thermal transport equation, written in PCH (Port-Controlled Hamiltonian) formulation. The IDA-PBC (Interconnection and Damping Assignment – Passivity based Control) controller is developed and tested on simulation as well as on TCV real plant. Two test scenarios are considered: ι control only, and ι and β control. The spatial distributions of ECCD profiles are pre-defined and only input powers are used for control design. Thus, a stationary control is defined in order to consider all non-linearity and actuator constraint, and a linear feedback IDA-PBC will ensure the convergence speed and the robustness of the closed-loop system. The obtained results are encouraging towards using routinely such plasma advanced control algorithm in a near future.
The integrated commissioning of the tungsten (W) environment in steady-state tokamak (WEST) was started with the superconducting magnet cool-down and completed one year after by the injection of 2.5 ...MW of additional power. It consisted of several steps: local system commissioning, integrated commissioning without plasma, and first plasma operation. This article reviews these WEST commissioning phases and highlights the experiences and the lessons learned. The magnet cooling down operation was initiated in late September 2016. The vacuum vessel was pumped out three weeks before the first plasma. The impregnation and curing of the two in-vessel divertor coils was performed at 180 °C during two days contributing to the baking of the vacuum vessel. In the meanwhile, rehearsal sessions for testing the CODAC systems were organized every month. The toroidal, poloidal, and divertor field coils were energized and their performance was tested successfully allowing moving to the first plasma operation phase. After a few attempts, plasma breakdown was achieved confirming that all systems were properly running and synchronized. Due to several micro air leaks on fiber optic feedthroughs and issues with the design of the in-vessel stabilizing plates, plasma current ramp-up could not be achieved in a first step. Once these issues have been fixed, confined plasmas could be obtained. Diverted plasmas were then achieved and the plasma current was increased up to 800 kA for more than 10 s. Up to 2.5 MW of additional power was eventually successfully coupled to the plasma.
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.
The successful performance of a model predictive profile controller is demonstrated in simulations and experiments on the TCV tokamak, employing a profile controller test environment. Stable ...high-performance tokamak operation in hybrid and advanced plasma scenarios requires control over the safety factor profile (q-profile) and kinetic plasma parameters such as the plasma beta. This demands to establish reliable profile control routines in presently operational tokamaks. We present a model predictive profile controller that controls the q-profile and plasma beta using power requests to two clusters of gyrotrons and the plasma current request. The performance of the controller is analyzed in both simulation and TCV L-mode discharges where successful tracking of the estimated inverse q-profile as well as plasma beta is demonstrated under uncertain plasma conditions and the presence of disturbances. The controller exploits the knowledge of the time-varying actuator limits in the actuator input calculation itself such that fast transitions between targets are achieved without overshoot. A software environment is employed to prepare and test this and three other profile controllers in parallel in simulations and experiments on TCV. This set of tools includes the rapid plasma transport simulator RAPTOR and various algorithms to reconstruct the plasma equilibrium and plasma profiles by merging the available measurements with model-based predictions. In this work the estimated q-profile is merely based on RAPTOR model predictions due to the absence of internal current density measurements in TCV. These results encourage to further exploit model predictive profile control in experiments on TCV and other (future) tokamaks.
An overview of the preliminary design of the ITER plasma control system (PCS) is described here, which focusses on the needs for 1st plasma and early plasma operation in hydrogen/helium (H/He) up to ...a plasma current of 15 MA with moderate auxiliary heating power in low confinement mode (L-mode). Candidate control schemes for basic magnetic control, including divertor operation and kinetic control of the electron density with gas puffing and pellet injection, were developed. Commissioning of the auxiliary heating systems is included as well as support functions for stray field topology and real-time plasma boundary reconstruction. Initial exception handling schemes for faults of essential plant systems and for disruption protection were developed. The PCS architecture was also developed to be capable of handling basic control for early commissioning and the advanced control functions that will be needed for future high performance operation. A plasma control simulator is also being developed to test and validate control schemes. To handle the complexity of the ITER PCS, a systems engineering approach has been adopted with the development of a plasma control database to keep track of all control requirements.
•New calibration method for threshold ionization mass spectrometry (TIMS) in tokamak.•Uncertainty evaluation of TIMS in plasma discharge and post-discharge.•He and D gas balance in D-to-He plasma ...changeover in WEST tokamak.•Difference in outgassing time constants for He and D2 in a full-W tokamak.•Delayed outgassing for D2 depends on the plasma termination: disrupted or not?
Threshold ionization mass spectrometry (TIMS) is one of two methods envisioned in ITER to quantify the helium (He) fusion product in the exhaust pumping lines during plasma discharges. We present the first demonstration of another potential application of TIMS in a tokamak environment, namely, the analysis of deuterium (D) and He outgassing following a plasma discharge i.e. during the post-discharge. This method has been tested with sub-second temporal resolution in WEST during its first He plasma discharges in the so-called He changeover experimental campaign. The calibration method of TIMS using a D plasma discharge is presented while the uncertainties related to TIMS during rapid pressure variations, i.e. upon plasma breakdown and plasma termination, are discussed. The first results obtained with TIMS during consecutive D and He plasma discharges in the full tungsten (W) tokamak WEST are reported. It is found that the time evolutions for He and D outgassing in the post-discharge are markedly different. On one hand, He outgassing is instantaneous and decays within 60 s until the He signal gets below detection level. On the other hand, D outgassing can reach a maximum up to several tens of seconds after the termination of the plasma and this outgassing can last for about 10 min. These striking differences should be related to different retention and outgassing from WEST plasma facing components, presently constituted of actively-cooled ITER-like W units and inertially cooled W-coated graphite. Potential mechanisms at the origin of the different outgassing behavior for D and He in W plasma facing components are discussed in light of a systematic analysis of the He and D gas balance and a macroscopic rate equation modeling of the D outgassing from the divertor strike points.
In December 2016, the experimental Tokamak WEST (W -for Tungsten- Environment in Steady state Tokamak) has produced its first plasma using its new Plasma Control System (PCS) based on the AUG RT ...framework DCS (Discharge Control System) and adapted to the specific needs of WEST. Now, WEST project’s development phase ends and a first operational version is routinely used for experimental purpose.
The new WEST configuration of Tore Supra facility leads to control challenges and the need to handle events of a modern diverted and metallic machine (vertical stabilization, impurity control, …). To ...address them, a new Plasma Control System (PCS) has been built based on the ASDEX upgrade (AUG) real-time framework called DCS (Discharge Control System). This contribution summarizes the work done during the 5 years of the project from the definition of the PCS concepts to its operation during the WEST campaigns. The integration into the Tore Supra control infrastructure is detailed as well as the different real-time control processes. The efficiency and the versatility of the PCS are illustrated by several examples of plasma operation.
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