Plasma Control System Research Articles

Preemptive RMP-driven ELM crash suppression automated by a real-time machine-learning classifier in KSTAR

Suppression or mitigation of edge-localized mode (ELM) crashes is necessary for ITER. The strategy to suppress all the ELM crashes by the resonant magnetic perturbation (RMP) should be applied as soon as the first low-to-high confinement (L–H) transition occurs. A control algorithm based on real-time machine learning (ML) enables such an approach: it classifies the H-mode transition and the ELMy phase in real-time and automatically applies the preemptive RMP. This paper reports the algorithm design, which is now implemented in the KSTAR plasma-control system, and the corresponding experimental demonstration of typical high-δ KSTAR H-mode plasmas. As a result, all initial ELM crashes are suppressed with an acceptable safety factor at the edge (q 95) and with RMP field adjustment. Moreover, the ML-driven ELM crash suppression discharges remain stable without further degradation due to the regularization of the plasma pedestal.

Deep learning based surrogate models for first-principles global simulations of fusion plasmas

The accurate identification and control of plasma instabilities is important for successful fusion experiments. First-principle simulations that can provide physics-based instability information such as the mode structure are generally not fast enough for real-time applications. In this work, a workflow has been presented to develop deep-learning based surrogate models for the first-principle simulations using the gyrokinetic toroidal code (GTC). The trained surrogate models of GTC (SGTC) can be used as physics-based fast instability simulators that run on the order of milliseconds, which fits the requirement of the real-time plasma control system. We demonstrate the feasibility of this workflow by first creating a big database from GTC systematic linear global electromagnetic simulations of the current-driven kink instabilities in DIII-D plasmas, and then developing SGTC linear internal kink instability simulators through supervised training. SGTC linear internal kink simulators demonstrate predictive capabilities for the mode instability properties including the growth rate and mode structure.

The Engineering Commissioning of the Magnets System On HL-2M Tokamak

HL-2M tokamak is one of largest magnetic confinement fusion devices designed and constructed by Southwestern Institute of Physics (SWIP), China National Nuclear Corporation (CNNC), the first plasma was obtained on December, 4 <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">th</sup> , 2020. The magnets system which plays an important role for the plasma confinement consists of 20 toroidal field (TF) coils and 18 poloidal field (PF) coils. The magnets system commissioning was carried out by two phases: self-commissioning and integrated commissioning. The first phase: self-commissioning including joint resistance commissioning, turn to turn insulation commissioning, ground insulation commissioning, deionized water sealing commissioning was accomplished during the installation and test of the magnets system. The second phase: the integrated commissioning with power supply system, ground system, measurement and data acquisition system, plasma control system, interlock and safety system was accomplished based on the results of the first stage. The total commissioning of magnets system in HL-2M tokamak for the first plasma have been accomplished successfully.

Overview of C-2W: high temperature, steady-state beam-driven field-reversed configuration plasmas

TAE Technologies, Inc. (TAE) is pursuing an alternative approach to magnetically confined fusion, which relies on field-reversed configuration (FRC) plasmas composed of mostly energetic and well-confined particles by means of a state-of-the-art tunable energy neutral-beam (NB) injector system. TAE’s current experimental device, C-2W (also called ‘Norman’), is the world’s largest compact-toroid device and has made significant progress in FRC performance, producing record breaking, high temperature (electron temperature, T e > 500 eV; total electron and ion temperature, T tot > 3 keV) advanced beam-driven FRC plasmas, dominated by injected fast particles and sustained in steady-state for up to 30 ms, which is limited by NB pulse duration. C-2W produces significantly better FRC performance than the preceding C-2U experiment, in part due to Google’s machine-learning framework for experimental optimization, which has contributed to the discovery of a new operational regime where novel settings for the formation section and the confinement region yield consistently reproducible, hot, and stable plasmas. An active plasma control system has been developed and utilized in C-2W to produce consistent FRC performance as well as for reliable machine operations using magnets, electrodes, gas injection, and tunable NBs. The active control system has demonstrated stabilization of FRC axial instability. Overall FRC performance is well correlated with NBs and edge-biasing system, where higher total plasma energy is obtained by increasing both NB injection power and applied-voltage on biasing electrodes. C-2W divertors have demonstrated a good electron heat confinement on open-field-lines using strong magnetic mirror fields as well as expanding the magnetic field in the divertors (expansion ratio > 30); the energy lost per electron ion pair, η e ∼ 6–8, is achieved, which is close to the ideal theoretical minimum.

Preliminary design of real-time plasma control system for CFETR

The China Fusion Engineering Test Reactor (CFETR), aiming at demonstrating fusion energy production in self-sufficient burning in different advanced operation scenarios, has entered the stage of engineering design and R&D (research and development). The real-time plasma control system (PCS), as a critical system controlling the evolution of plasma parameters that are necessary to operate CFETR, is designed synchronously. The CFETR PCS is composed of system architecture, integrated control algorithms, and two assistant tools PCS-SDP (Software Development Platform) and PCS-VP (Verification Platform) for algorithm development and verification. Taking the steady-state operation and machine safety into account, the requirement and conceptual design of the system architecture of CFETR PCS is introduced in this paper. PCS-SDP, a visual algorithm development platform, has completed the preliminary design and a prototype has been built to validate the design and technical feasibility with the control algorithm development for EAST. PCS-VP is under preliminary design taking ITER PCSSP (Plasma Simulation Platform) design as reference. These two tool platforms will strongly support and accelerate the development of CFETR PCS.

ITER plasma control system final design and preparation for first plasma

The ITER plasma control system (PCS) has successfully completed its final design for the first plasma (FP) and engineering operations (EOs) phase and plant system commissioning has begun as ITER prepares for this first operation phase. Commissioning of the essential plant systems will continue as each plant system is completed and made ready for operation. Tokamak assembly has begun with the base and lower cylinder of the cryostat and the lower most poloidal field (PF) coil installed in the tokamak pit. The first vacuum vessel (VV) sector and accompanying two toroidal field (TF) coils are being prepared for transfer to the pit. Once the tokamak is assembled, the cryostat top lid is closed and pump down begins, this will start approximately one year of integrated commissioning (IC) to prepare all of the relevant plant systems for FP operation. After a scheduled one month of plasma operation with the goal of achieving a plasma current > 100 kA for at least 100 ms, there will be about six months of EO to complete commissioning of the superconducting central solenoid, PF, and TF magnet systems to full current, without plasma, to complete this initial ITER operation phase. The PCS final design for FP will be described as well as the IC sequence of the main plant systems required for this operations phase. The plans for the FP operation campaign will be described, including specific challenges present on ITER due to large VV eddy currents, issues associated with electron cyclotron heating (ECH) assist, neutral pressure and impurities. This will be followed by the EO phase to commission full current operation of the superconducting magnets, possibly including plasma operation at full TF of 5.3 T to have improved conditions both for Ohmic plasma initiation and ECH absorption.

Designand implementation of plasma electron density measurements based on FPGA with all-phase FFT for tokamak devices.

Plasma electron density is one of the most fundamental parameters when studying tokamak plasma physics, which is widely used in plasma control systems and plasma physics analyses. A hydrogen cyanide laser interferometer is generally applied to measure the plasma electron density in many tokamak devices. Therein, the plasma electron density is calculated by measuring the phase difference between the reference signal and the detector signal. This work provides a new way to realize real-time measurements of the plasma electron density with a phase comparator and processing system based on a field-programmable gate array chip. The system integrates a signal processing module, an all-phase fast Fourier transform (ap-FFT) module realized via matrix operations and phase comparisons, and a network communication module all in one board. This work concludes that the ap-FFT is robust and accurate for phase calculations compared with a windowing FFT. A data-reuse method and a phase shift method are proposed to improve the time resolution and phase range. The phase error is less than 0.1° and the time resolution is 0.025ms, which is better than hardware methods and traditional software methods. This system is highly flexible with reduced designcosts to meet the requirements of a tokamak, which can provide a valuable reference for other tokamak and phase difference comparisons.

Integrated Real-Time Supervisory Management for Off-Normal-Event Handling and Feedback Control of Tokamak Plasmas

For long-pulse tokamaks, one of the main challenges in control strategy is to simultaneously reach multiple control objectives and to robustly handle in real-time (RT) unexpected events (off-normal-events -- ONEs) with a limited set of actuators. We have developed in our previous work a generic architecture of the plasma control system (PCS) including a supervisor and an actuator manager to deal with these issues. We present in this paper recent developments of real-time decision-making by the supervisor to switch between different control scenarios (normal, backup, shutdown, disruption mitigation, etc.) during the discharge, based on off-normal-event states. We first standardize the evaluation of ONEs and thereby simplify significantly the supervisor decision logic, as well as facilitate the modifications and extensions of ONE states in the future. The whole PCS has been implemented on the TCV tokamak, applied to disruption avoidance with density limit experiments, demonstrating the excellent capabilities of the new RT integrated strategy.

Development of synthetic diagnostics for ITER First Plasma operation

ITER First Plasma (FP) operation aims to produce a plasma with current higher than 100 kA for a duration longer than 100 ms. However, the low values of plasma density and temperature associated with the initiation phase make it difficult to diagnose the plasma accurately. It is therefore essential to develop models for the available diagnostics for ITER FP operation in order to determine the necessary measurement ranges for plasma initiation and use them as inputs for controller development and assessment within the plasma control system (PCS). The use of accurate diagnostic models also helps to optimally prepare for, and analyse, ITER FP operation. A model developed for the Hα main chamber visible spectroscopy for FP shows that it is possible to measure Hα emission soon after breakdown for plasma temperatures higher than 3 eV and average plasma densities higher than 1017 m−3. The diagnostic model for the interferometer shows that for the proposed ITER FP scenarios, accurate measurements are possible for average plasma densities higher than 5 × 1017 m−3. The results from modelling of the hard x-ray monitor demonstrate that it can provide signals useful for runaway electron detection and can measure current levels typically down to 3 kA. Thus, these measurements can thus be used effectively by the PCS for runaway electron (RE) detection during FP operation. Improved diagnosis of the plasma state deducing, for example, parameters such as the plasma size and position, is possible by combining the measurements from the different FP diagnostics (density interferometer polarimeter and Hα main chamber visible spectroscopy), which will be available for ITER FP operation.

Real-time prediction of high-density EAST disruptions using random forest

A real-time disruption predictor using random forest was developed for high-density disruptions and used in the plasma control system (PCS) of the EAST tokamak for the first time. The disruption predictor via random forest (DPRF) ran in piggyback mode and was actively exploited in dedicated experiments during the 2019–2020 experimental campaign to test its real-time predictive capabilities in oncoming high-density disruptions. During dedicated experiments, the mitigation system was triggered by a preset alarm provided by DPRF and neon gas was injected into the plasma to successfully mitigate disruption damage. DPRF’s average computing time of ∼250 μs is also an extremely relevant result, considering that the algorithm provides not only the probability of an impending disruption, i.e. the disruptivity, but also the so-called feature contributions, i.e. explainability estimates to interpret in real time the drivers of the disruptivity. DPRF was trained with a dataset of disruptions in which the electron density reached at least 80% of the Greenwald density limit, using the zero-dimensional signal routinely available to the EAST PCS. Through offline analysis, an optimal warning threshold on the DPRF disruptivity signal was found, which allows for a successful alarm rate of 92% and a false alarm rate of 9.9%. By analyzing the false alarm causes, we find that a fraction (∼15%) of the misclassifications are due to sudden transitions of plasma confinement from H- to L-mode, which often occur during high-density discharges in EAST. By analyzing DPRF feature contributions, it emerges that the loop voltage signal is that main cause of such false alarms: plasma signals more apt to characterize the confinement back-transition should be included to avoid false alarms.

Automated signal classification in the C-2W fusion experiment.

In TAE Technologies' current experimental fusion device, C-2W (also called "Norman"), record breaking, advanced beam-driven field-reversed configuration plasmas are produced and sustained in steady state utilizing variable-energy neutral beams, expander divertors, end-bias electrodes, and an active plasma control system. With a rapid shot-pace and an extensive number of data channels, the amount of data generated necessitates automated signal processing. To this end, a machine learning algorithm consisting of a multi-layered neural network as well as other data processing software has been designed for signal feature identification, allowing for accurate and fast signal classification, anomalous condition detection, and providing for signal pre-processing. With a small set of training data, the neural network can be "bootstrapped" to provide a robust classification system while minimizing human oversight. An example using data from the theta pinch plasma formation pulsed power system is presented. With an overall accuracy of ∼97%-having classified more than 5 × 106 pulsed power signals-the classification scheme is more than sufficient to fine-tune machine set points. However, this technique can be used for near-real-time preprocessing of any plasma physics signal and has wide ranging application in fusion experiments for the varied data produced by plasma diagnostics.

Wire calorimeter for direct neutral beam power measurements on C-2W.

The C-2W experiment produces advanced beam-driven field reversed configuration (FRC) plasmas, which are sustained in steady state utilizing variable energy neutral beams (15-40 keV, total power up to 21MW), advanced divertors, end bias electrodes, and an active plasma control system. Since heating, current drive, and refueling from neutral beam injection are essential to FRC sustainment, it is crucial to have accurate measurements of the beam power being injected into the plasma. A new tungsten wire calorimeter has been designed, built, calibrated on a test stand, and implemented to make the first direct measurements of the time-average injected beam power into C-2W. This paper outlines the design of the diagnostic, along with calibration methods and experimental results.

Custom application of PCS software development platform on EAST

The plasma control system software development platform (PCS-SDP) will provide an effective, convenient, and visual development environment for PCS software developers. The detailed design of PCS-SDP for the Chinese fusion engineering testing reactor has been completed. A custom application of PCS-SDP is implemented on EAST to verify its effectiveness and reliability, and validate the platform architecture and technical proposal. PCS-SDP customizes components corresponding to the algorithm control interface. Developers can directly drag and drop the components to design data structure, waveform information, and other parameters of the algorithm required for communication with the GUI interface. Moreover, the algorithm logic master file, including the basic modules and functions, is predefined in the platform. Developers only need to add the algorithm logic code to the corresponding module and then save the project. The new algorithm can be compiled, linked with other files, and finally successfully integrated into PCS. Therefore, PCS algorithm development becomes more intuitive and efficient.

Conceptual design of the COMPASS-U control systems

The COMPASS-U tokamak is an advanced and more complicated successor of the COMPASS tokamak. Almost all main technological parts will be different. Both the higher performance of COMPASS-U and the increased number of parts and systems lead to a higher risk of failure. Therefore, new control systems must be developed regarding handling dangerous situations and integrating new systems. The COMPASS-U also opens the opportunity to make significant changes in the current systems based on the experience gained from COMPASS operation and to upgrade the systems to nowadays technologies.This article focuses on the conceptual design of the COMPASS-U control systems. It shows all CODAC (Control, Data Access, and Communication) systems their purpose, functionality, operation principle, interaction, and main used technologies.The COMPASS-U will use EPICS (Experimental Physics and Industrial Control System) for Slow control and supervision system. The real-time plasma control system will utilize MARTe2 (Multi-threaded Application Real-Time executor). The base control loop period of the Real-time plasma control system will be 50 μs as of COMPASS.

Multi-chord IR-visible two-color interferometer on KSTAR.

Major parts of an IR-visible two-color interferometer (TCI) on KSTAR have been upgraded for the multi-chord operation: (1) a diode-pumped-solid-state (DPSS) laser (660nm) replacing the former HeNe laser (633nm), (2) vacuum-compatible vibration isolator with titanium retro-reflectors, and (3) full digital phase comparator for multi-chord real-time density signals. The commercial compact DPSS laser suits the multiple chord configuration with its strong beam power (500 mW) and long coherent length (>100 m). Ti retro-reflectors are mounted on vacuum-compatible vibration isolators. The isolators are essential for the visible beams to avoid any fringe skips due to their short wavelength, considering the speed of the mechanical vibration (up to hundreds of μm). Field-programmable-gate-array (FPGA) modules count the entire fringes fast enough with a signal output rate up to 1.25MHz, solving the fringe skip issues. The FPGA module enables the full digital processing of the phase comparator with a CORDIC algorithm after the sampling rate of 160 MS/s for the 40MHz intermediate frequency of each beam. The full digital signals are transferred to the main plasma control system in real-time. Stable single-input-single-output operation of the KSTAR density control was demonstrated with the TCI. The real-time density profile control is also promising in the near future, with multiple actuators such as pellets and gas puffings.

Avoidance of vertical displacement events in DIII-D using a neural network growth rate estimator

Robust disruption avoidance techniques are critical for the development of reliable fusion reactor devices. A viable reactor will require non-disruptive, long pulse operation where simply shutting down a discharge is undesirable. To achieve such performance, the plasma must be controlled to continuously avoid hazardous regimes instead of asynchronously aborting. A recent experiment on DIII-D demonstrated for the first time real-time control of proximity to a disruptive instability boundary. In particular, the vertical growth rate, an eigenvalue that characterizes the degree of instability of the plasma's vertical position, was regulated so as not to exceed DIII-D's vertical controllability limit. The open-loop growth rate was estimated in real time on the DIII-D plasma control system using a neural network model trained with tens of thousands of DIII-D shots. The model was trained to replicate the results of RZRIG [1], a rigid displacement code for calculating the growth rate. Once trained, producing an estimate using the neural network is multiple orders of magnitude faster than RZRIG, thereby making the calculation suitable for real-time execution. The control system regulated the estimated growth rate by adjusting plasma elongation and distance to the inner wall of the vessel, and this regulation was shown to reliably avoid vertical displacement event disruptions (i.e. uncontrolled vertical oscillations) of the plasma. This work presents these experimental results, including the dynamic performance and the effectiveness of the control technique. Details are presented on the training of the neural network model, including concerns such as hyperparameter tuning and uncertainty quantification. Additionally, the methodology for embedding the neural network into the control system is discussed.

Non-Resonant n = 1 Helical Core Induced by m/n = 2/1 Neoclassical Tearing Mode in JT-60U

In JT-60U, simultaneous excitation of n = 1 helical cores (HCs) and m/n = 2/1 Tearing Modes (TMs) was observed [T. Bando et al., Plasma Phys. Control. Fusion 61 115014 (2019)]. In this paper, we have investigated the excitation mechanism of n = 1 HCs with m/n = 2/1 TMs based on the experimental observations and a simple quasi-linear MHD model. In the previous study, it was reported that a "coupling" on the phase of the MHD mode is observed between n = 1 HCs and m/n = 2/1 TMs. In this study, it is found that the coupling is observed with the mode frequency from several Hz to 6 kHz. This indicates that the resistive wall and the plasma control system do not induce the coupling because the both time scales are different from the mode frequency. In addition, n = 1 HCs appear to be the non-resonant mode from the two observations: n = 1 HCs do not rotate with the plasma around the q = 1 surface in the core and the coupling is also observed even when qmin > 1. It is also observed that the electron fluctuation due to an n = 1 HC in the core region disappears with the stabilization of an m/n = 2/1 neoclassical tearing mode by electron cyclotron current drive, implying that n = 1 HCs are driven by m/n = 2/1 TMs. This perspective, n = 1 HCs are driven by m/n = 2/1 TMs, is supported by the observation that the saturated amplitude of the m/n = 1/1 component of the radial displacement in the core is smaller than that of the m/n = 2/1 component. Finally, we revisit a quasi-linear MHD model where the m/n = 1/1 HC is induced directly by the sideband of the current for the m/n = 2/1 TM, which allows to excite the non-resonant m/n = 1/1 mode. The model also describes the characteristic of the coupling, fm/n=1/1(HC) = 2fm/n=2/1(TM).

The integrated diagnostic suite of the C-2W experimental field-reversed configuration device and its applications.

In the current experimental device of TAE Technologies, C-2W (also called "Norman"), record breaking advanced beam-driven field-reversed configuration (FRC) plasmas are produced and sustained in steady state utilizing variable energy neutral beams (15-40 keV, total power up to 20MW), advanced divertors, bias electrodes, and an active plasma control system. This fully operational experiment is coupled with a fully operational suite of advanced diagnostic systems. The suite consists of 60+ individual systems spanning 20 categories, including magnetic sensors, Thomson scattering, interferometry/polarimetry, spectroscopy, fast imaging, bolometry, reflectometry, charged and neutral particle analysis, fusion product detection, and electric probes. Recently, measurements of main ion temperatures via a diagnostic neutral beam, axial profiles of energy flux from an array of bolometers, and divertor and edge plasma parameters via an extensive set of electric probes, interferometers, and spectrometers have all been made available. All the diagnostics work together to provide a complete picture of the FRC, fast-ion inventory, and edge plasma details enabling tomographic reconstruction of plasma parameter profiles and real-time plasma control.

Keras2c: A library for converting Keras neural networks to real-time compatible C

With the growth of machine learning models and neural networks in measurement and control systems comes the need to deploy these models in a way that is compatible with existing systems. Existing options for deploying neural networks either introduce very high latency, require expensive and time consuming work to integrate into existing code bases, or only support a very limited subset of model types. We have therefore developed a new method called Keras2c, which is a simple library for converting Keras/TensorFlow neural network models into real-time compatible C code. It supports a wide range of Keras layers and model types including multidimensional convolutions, recurrent layers, multi-input/output models, and shared layers. Keras2c re-implements the core components of Keras/TensorFlow required for predictive forward passes through neural networks in pure C, relying only on standard library functions considered safe for real-time use. The core functionality consists of ∼1500 lines of code, making it lightweight and easy to integrate into existing codebases. Keras2c has been successfully tested in experiments and is currently in use on the plasma control system at the DIII-D National Fusion Facility at General Atomics in San Diego.

WEST operation with real time feed back control based on wall component temperature toward machine protection in a steady state tungsten environment

A real time Wall Monitoring System (WMS) is used on the WEST tokamak during the C4 experimental campaign. The WMS uses the wall surface temperatures from 6 fields of view of the Infrared viewing system. It extracts the raw digital data from selected areas, converts it to temperatures using the calibration and write it on the shared memory network being used by the Plasma Control System (PCS). The PCS feeds back to actuators, namely the injected power from 5 antennae's of the lower hybrid and ion cyclotron resonance radiofrequency (RF) heating systems. WMS activates feed back control 63 times during C4, which is 14 % of the plasma discharges. It activates mainly as the result of a direct RF loss to the upper divertor pipes. The feedback control maintains the wall temperature within the operation envelope during 97 % of the occurrences, while enabling plasma discharge continuation. The false positive rate establishes at 0.2 %. WMS significantly facilitated the operation path to high power operation during C4, by managing the technical risks to critical wall components.