Plasma Operation Research Articles

Understanding the role of boron in plant adaptation to soil salinity.

Soil salinity is a major environmental constraint affecting the sustainability and profitability of agricultural production systems. Salinity stress tolerance has been present in wild crop relatives but then lost, or significantly weakened, during their domestication. Given the genetic and physiological complexity of salinity tolerance traits, agronomical solutions may be a suitable alternative to crop breeding for improved salinity stress tolerance. One of them is optimizing fertilization practices to assist plants in dealing with elevated salt levels in the soil. In this review, we analyse the causal relationship between the availability of boron (an essential metalloid micronutrient) and plant's adaptive responses to salinity stress at the whole-plant, cellular, and molecular levels, and a possibility of using boron for salt stress mitigation. The topics covered include the impact of salinity and the role of boron in cell wall remodelling, plasma membrane integrity, hormonal signalling, and operation of various membrane transporters mediating plant ionic and water homeostasis. Of specific interest is the role of boron in the regulation of H+-ATPase activity whose operation is essential for the control of a broad range of voltage-gated ion channels. The complex relationship between boron availability and expression patterns and the operation of aquaporins is also discussed.

First results of non-evaporable getter pump studies for the application in the EAST tokamak

Enhanced particle exhaust is pivotal in optimizing fuel recycling and bolstering plasma performance within fusion reactors. Non-Evaporable Getter (NEG) pumps, renowned for their robust high-temperature tolerance, high pumping speeds and large adsorption capabilities, emerge as highly promising candidates for augmenting the particle exhaust efficiency of tokamaks. Rigorous evaluations have been undertaken to ascertain their pumping efficacy within a vacuum testing setup and the EAST tokamak environment. Experimental data have consistently demonstrated that NEG pumps maintain stable operation across a broad temperature range, from ambient conditions to 200 °C, with a notable direct correlation observed between the pumping speed and the ambient temperature. Their performance surpasses that of conventional cryopumps, especially at D2 gas pressures exceeding 0.1 Pa. Additionally, the application of nitrogen passivation has been validated as an effective method to modulate the pumping speed and to mitigate the risk of hydrogen embrittlement and oxygen and carbon contaminations. Despite exposure to the extreme operational demands, including 30 h of plasma exposure, and lithium processing, atmospheric conditions, and water leaks, the NEG pumps preserved approximately 70 % of their initial pumping capacity. These results underscore the compatibility of NEG pumps with the challenging environments of plasma operations, positioning them as a viable and promising solution for particle exhaust in the next-generation fusion reactors.

Investigation of fast ion effects on core turbulence in FIRE mode plasmas

Further investigation of fast ion effects on turbulence and transport in the fast ion regulated enhancement (FIRE) mode discharge (Han et al 2022 Nature 609 269–275) was performed in this work as a continuation of a previous study (Kim et al 2023 Nucl. Fusion 63 124001) that showed that the dominant turbulence suppression mechanism by fast ions is the dilution effect in the FIRE mode discharge. The current study includes (i) the impact of the fast ion relevant mode observed in the simulation of thermal energy flux, (ii) dilution effects by fast ions compared to dilution effects by other species, and (iii) fast ion effects on electron-scale turbulence. First, nonlinear gyrokinetic simulation results show that turbulence is significantly suppressed even without the fast ion relevant mode, indicating that the impact of this mode on thermal transport is not significant in this discharge. Second, further analysis on the dilution effects shows the three following results: Turbulence is not completely suppressed by the reduced main ion density fraction effect due to impurities; the reduction in energy flux can be limited by a certain impurity mode that is destabilized by a high impurity density gradient from adjusting the main ion density gradient; electrons can contribute to turbulence suppression through the main ion density gradient change, although this effect is less significant compared to other species. Third, we observe that two fast ion effects can influence the linear growth rate of the electron-scale turbulence mode. The growth rate decreases by an increase in β∗(≡(−8π/B2)dp/dr) and increases by dilution effects, suggesting that fast ion effects on electron-scale turbulence can differ depending on the operation scenario, such as the fast ion fraction. The comprehensive analysis performed in this study can enhance our understanding of fast ion physics, required for burning plasma operation in the future.

Design and operation results of KSTAR ECH system

The electron cyclotron heating (ECH) system at the Korea Superconducting Tokamak Advanced Research (KSTAR) is a crucial heating devices, which is employed in various research endeavors, including plasma startup, wall cleaning, neoclassical tearing mode control, long pulse operation, and hybrid scenarios development from the first plasma operation. For facilitating advanced scenarios development and magnetohydrodynamics research in KSTAR, it was deemed necessary to install a total of 6 MW ECH systems, consisting of four 105/140 GHz and two 170 GHz systems. To address this, a new ECH room was developed and the existing ECH system was totally rearranged. This study describes the design of the novel KSTAR ECH system and its operation results, and discusses the problems that occur during the ECH system operation. Each ECH system to deliver 1 MW for 300 s was designed to enable stable, safe, and efficient operation in the new ECH room and four 105/140 GHz dual frequency ECH were injected to KSTAR successfully. The polarization angle was experimentally confirmed by using an electron cyclotron wall conditioning, and the beam angle change by ECH antenna was verified by a newly developed IR measurement. However, mixed patterns with higher-order modes were observed in the output beam from matching optics unit of some gyrotrons, which affect the operation efficiency and the output power. In future works, we will analyze the causes for low mode purity and improve the operation efficiency.

SMART materials for DEMO: Towards industrial production

Self-passivating Metal Alloys with Reduced Thermo-oxidation (SMART) are under development for a fusion power plant. SMART exhibits similar sputtering resistance as that of pure tungsten during regular plasma operation. Under accident conditions, SMART demonstrates the suppression of oxidation – a remarkable advantage over pure tungsten. The viability of SMART technology has been shown on a laboratory scale.Presently, the industrial scale-up of SMART technology is underway. This activity comprises the industrially supplied feedstock materials for SMART, mechanical alloying of powder at industry and sintering of SMART materials using industrial equipment. Industrial mechanical alloying can be accomplished within 21,5 h of milling, providing at least four kilograms of fully alloyed powder. Rectangular industrial SMART samples feature linear dimensions of 10 cm, a thickness of 0.5 cm, a weight slightly below 1 kg and a relative density exceeding 97 %.The transition to industrial feedstock suppliers allowed SMART to outperform pure tungsten in powder procurement costs needed for plasma-facing components. The status of industrial SMART technology is given along with an outlook on future activities.

Trigger transceiver and timing control system for ADITYA-U tokamak

The trigger transceiver and timing control system described in this study is developed to generate multiple triggers with varying delays to synchronize various subsystems with plasma discharges in the ADITYA Upgrade (ADITYA-U) tokamak (Ghosh et al., 2016) [1]. The Toroidal Field Power Supply (TFPS) is a fundamental component for each plasma discharge in the ADITYA-U, which generates a master trigger pulse for the particular discharge. The rise time of Toroidal Field Power Supply is 2 s and the loop voltage which is required to initiate the plasma discharge begins at 2.4 s after the master trigger. So the loop voltage is time-stamped at t = 0 and the master trigger from TFPS is time-stamped at t = −2.4 s and considered as a pre-trigger. Apart from this master trigger at t = −2.4 s, two more pre-trigger pulses, one at ∼ −40 ms and another at ∼ −4 ms, prior to the initiation of loop voltage trigger are generated by the power supply. These pre-trigger pulses generated by TFPS constitute the backbone of the entire discharge timing sequence, and they operate/control/initiate numerous ADITYA-U subsystems such as the gas puffing system, data acquisition system, radio frequency pre-ionization and heating system, plasma position control, and so on. So a suitable trigger generator system was needed to be developed in which three pre-trigger pulses (time stamped @ −2.4 s, −40 ms, and −4 ms prior to the loop voltage trigger) arriving from the TFPS may be selected and altered based on the experimental needs. To cater these requirements in ADITYA-U, a trigger transceiver and timing control system has been developed. The system has been successfully operated over ∼6000 discharges of ADITYA-U and functioned satisfactorily in all discharges in daily experiments with no failures. This study will discuss the detailed design elements and implementation of the Trigger Transceiver and Timing Control System for ADITYA-U plasma operations.

Effect of Gas Composition on Temperature and CO2 Conversion in a Gliding Arc Plasmatron reactor: Insights for Post-Plasma Catalysis from Experiments and Computation.

Plasma-based CO2 conversion has attracted increasing interest. However, to understand the impact of plasma operation on post-plasma processes, we studied the effect of adding N2, N2/CH4 and N2/CH4/H2O to a CO2 gliding arc plasmatron (GAP) to obtain valuable insights into their impact on exhaust stream composition and temperature, which will serve as feed gas and heat for post-plasma catalysis (PPC). Adding N2 improves the CO2 conversion from 4 % to 13 %, and CH4 addition further promotes it to 44 %, and even to 61 % at lower gas flow rate (6 L/min), allowing a higher yield of CO and hydrogen for PPC. The addition of H2O, however, reduces the CO2 conversion from 55 % to 22 %, but it also lowers the energy cost, from 5.8 to 3 kJ/L. Regarding the temperature at 4.9 cm post-plasma, N2 addition increases the temperature, while the CO2/CH4 ratio has no significant effect on temperature. We also calculated the temperature distribution with computational fluid dynamics simulations. The obtained temperature profiles (both experimental and calculated) show a decreasing trend with distance to the exhaust and provide insights in where to position a PPC bed.

Study on hot isostatic pressing conditions of ARAA for fabrication of the breeding blanket first wall

The helium-cooled concept with solid breeding materials is one of the candidates for demonstration fusion power reactor (DEMO) breeding blanket. The first wall (FW) is a core component of the breeding blanket. The FW should keep the structural integrity during the normal plasma operation which is subject to high surface heat flux, neutron wall load, and high coolant pressure simultaneously. Various fabrication methods have been studied for the development of the FW fabrication technologies. This study focuses on FW fabrication technologies with hot isostatic pressing (HIP) bonding which is considered as the most promising method for achieving such complex shape like FW having internal cooling channels. We investigated the HIP bonding conditions using Advance Reduced Activation Alloy, ARAA, which is a reduced activation ferritic-martensitic steel under development in Korea. Temperature, holding time and pressure were considered as the parameters to optimize the HIP bonding conditions. The microstructures around the HIP joints changed with respect to the HIP bonding temperature. The amount of oxide particles around the HIP joints varied depending on the degassing conditions. The tensile strength of the HIP joint was not greatly affected by the presence of the HIP bonding line and degassing conditions. On the other hand, it has been confirmed that the Charpy absorbed energy of the HIP joint was significantly affected by degassing conditions. An increase in oxide particles around the HIP joints occurred when degassing temperature and vacuum level were low, and as the quantity of oxide particles increased, the Charpy absorbed energy of the HIP joint decreased substantially.

Non-thermal plasma for decontamination of bacteria trapped in particulate matter filters: plasma source characteristics and antibacterial potential

The aims of this study encompass the characterization of process parameters and the antimicrobial potential during operation of a novel non-thermal plasma (NTP) source in a duct system containing a particulate matter (PM) filter thus mimicking the interior of an air purifier. Simulating conditions of a long-term operation scenario, in which bacterial aerosols in indoor environments accumulate on PM filters, the filter surfaces were artificially inoculated with Escherichia coli (E. coli) and exposed to an air stream enriched with reactive species. Electrical power consumption, key plasma parameters, volume flow and air flow velocity, reactive gas species concentrations as well as inactivation rates of E. coli were assessed. The NTP operated at a gas temperature close to ambient air temperature and featured a mean electron energy of 9.4 eV and an electron density of 1∙1019 m−3. Ozone was found to be the dominating reactive gas species with concentrations of approx. 10 ppm in close vicinity to the PM filters. An inactivation rate of 99.96 % could be observed after exposure of the PM filters to the gas stream for 15 min. This inactivation efficiency appears very competitive in combating realistic bacterial aerosol concentrations in indoor environments.

Pop-up Langmuir probe diagnostic in the water cooled divertor of Wendelstein 7-X.

The design, development, and successful implementation of pop-up Langmuir probes installed in the water-cooled divertor of W7-X are described. The probes are controlled by drive coils (actuators) installed behind the divertor plates. These drive coils make use of the magnetic field in W7-X to move the probe tips into and out of the plasma. The drive coils were installed in the vacuum vessel after extensively testing the durability of the coils and analyzing the criteria for safe operation. The probe design is carefully tailored for each of the 36 probe tips in order to be suitable for the different magnetic field configurations used in W7-X and ensure that the probes do not present leading edges to the magnetic flux tubes. An electronic bridge circuit is used for measurement to compensate for the effects of signal propagation time on the long cable lengths used. The diagnostic is integrated with the segment control of W7-X for automated operation and control of the diagnostic. The evaluation of the results from the plasma operation is presented after accounting for appropriate sheath expansion for negative bias voltage on the probes.

High-frequency fluctuation and EHO-like mode in the H-mode pedestal on the EAST tokamak

In the pedestal region of the Experimental Advanced Superconducting Tokamak (EAST) during high confinement mode plasma operations with radio-frequency heating, two distinct fluctuations are observed: high-frequency fluctuations (HFFs) and edge harmonic oscillation-like (EHO-like) modes. The HFFs are characterized by intermittent fluctuations with a broadband frequency range of 1−3 MHz and a poloidal wave number ( kθ ) greater than 0.9 cm−1 . On the other hand, the EHO-like mode exhibits characteristics similar to magnetohydrodynamics (MHD)-like modes with n = 1−5 and lower poloidal wave numbers ( kθ⩽0.12 cm−1 ). During the pedestal establishing phase following the L–H transition, a significant concurrent presence of HFF and EHO-like modes in high-density pedestal regions has been noted. In this phase, the EHO-like mode not only modulates the amplitude of the HFF but also engages in nonlinear interactions. The occurrence of EHO-like mode and HFF is associated with particle transport toward the divertor, though it is notably less than that caused by edge coherent modes. During the inter-edge localized mode (ELM) period, a significant decrease in the Dα baseline is observed whenever the low frequency fluctuation (LFF) weakens and the HFF grows, prior to each large ELM. One possible explanation is that the rapid increase of E×B shear stabilizes the LFF and destabilizes the HFF, which lowers the pedestal transport and enables the further growth of the pedestal until the onset of the ELM.

Study on damage behavior of the outer horizontal target in the EAST lower divertor after plasma operations

The EAST lower divertor features a partially implemented flat-type structure, and visible damage has been observed, particularly at the outer horizontal target location after plasma operations. This study conducts a detailed analysis of the damage to the outer horizontal target. The findings reveal instances of plastic deformation, cracking, and even tungsten flake detachment at specific locations on the outer horizontal target. Furthermore, different degrees of recrystallization are identified in the tungsten plates after plasma operations, with a more pronounced effect closer to the surface of the tungsten plates. These observations offer pertinent insights for ongoing research into the microstructural enhancement design of tungsten materials and the enhancement of crack thresholds at the W/Cu interface.

Micro-particle injection experiments in ADITYA-U tokamak using an inductively driven pellet injector

A first-of-its-kind, inductively driven micro-particle (Pellet) accelerator and injector have been developed and operated successfully in ADITYA-U circular plasma operations, which may ably address the critical need for a suitable disruption control mechanism in ITER and future tokamak. The device combines the principles of electromagnetic induction, pulse power technology, impact, and fracture dynamics. It is designed to operate in a variety of environments, including atmospheric pressure and ultra-high vacuum. It can also accommodate a wide range of pellet quantities, sizes, and materials and can adjust the pellets’ velocities over a coarse and fine range. The device has a modular design such that the maximum velocity can be increased by increasing the number of modules. A cluster of lithium titanate/carbonate (Li2TiO3/Li2CO3) impurity particles with variable particle sizes, weighing ∼50–200 mg are injected with velocities of the order of ∼200 m s−1 during the current plateau in ADITYA-U tokamak. This leads to a complete collapse of the plasma current within ∼5–6 ms of triggering the injector. The current quench time is dependent on the amount of impurity injected as well as the compound, with Li2TiO3 injection causing a faster current quench than Li2CO3 injection, as more power is radiated in the case of Li2TiO3. The increase in radiation due to the macro-particle injection starts in the plasma core, while the soft x-ray emission indicates that the entire plasma core collapses at once.

Evaluation of the impact of plasma operation scenarios on the fusion reactor blanket design using an integrated numerical plasma and neutronics analysis suite

A very limited operation scenario has been used for fusion neutronics for the design of tritium breeding blankets. However, many advanced operation scenarios are under development and the impact of choosing a plasma operation scenario for fusion neutronics, which requires integration of neutronics analysis with plasma analysis, needs to be evaluated. An automated process from plasma analysis to neutronics analysis to create a viable fusion plasma neutron source was considered using a few different codes. In this work, we develop an integrated suite combining plasma modeling codes and neutronics codes. McCARD, a Monte-Carlo neutron-photon transport simulation code, is integrated into TRIASSIC, a tokamak reactor integrated automation suite for simulation and calculation, as the module for neutronics analysis of various plasma operation scenarios. McCARD integration allows TRIASSIC to perform neutronics analysis along with analytical and predictive plasma simulations. To evaluate this integrated suite, KSTAR experimental data were used to calculate the neutron flux at the first wall via TRIASSIC with the McCARD module in three different cases. First, the effect of plasma shape parameters, elongation and triangularity, on neutron flux is analyzed. Second, plasma scenario with the changes to the shape and the performance of plasma over time is studied to evaluate the time varying simulation. Third, two different plasma scenarios are compared to analyze the impact of plasma performance on neutron flux.

Nuclear analyses for the integration of ITER equatorial Port 2

The present work is devoted to nuclear analyses in support of the ITER diagnostic Equatorial Port 2 (EP#2) integration. ITER EP #2 is a diagnostic port based on the long-modular Diagnostic Shielding Module (DSM) housing the following systems: Disruption Mitigation System (DMS) in DSM#1 and #3 and X-Ray Crystal Spectroscopy Core (XRCS-Core) in DSM#2. Ensuring adequate radiation shielding is a major challenge since the diagnostic systems require several apertures from the Vacuum Vessel (VV) through the Port Interspace (PI) and up to the Port Cell (PC). In the present study, a three-dimensional MCNP model of EP#2 has been developed, starting from the latest design available from Preliminary Design Review stage (PDR), and successively integrated into the reference 40° ITER C-Model. Comprehensive nuclear analyses have been carried out employing the D1SUNED v3.1.4 code based on the MCNP Monte Carlo transport code. Relevant nuclear quantities during and at the end of plasma operations have been evaluated: i.e., neutron and gamma fluxes and energy spectra along the port from the Diagnostic First Wall (DFW) up to the Bio-Shield Plug (BP), nuclear heating, neutron damage, helium and tritium production, and shutdown dose rate. This analysis allowed the identification of potentially critical areas, and therefore, the implementation of additional shielding options aimed at reducing the neutron streaming and mitigating of the radiation field in the PI region. In this work, the results of the analyses are presented and discussed. Some solutions to mitigate nuclear loads and to improve the shielding in PI area are proposed and their impact has been assessed. Finally, some recommendations for the optimization of the design of EP#2 are provided as well.

An extension of NASCA for SDDR analysis of mobile activated components with global variance reduction

During the plasma operation period of fusion reactor facilities, neutrons generated by DT reaction can penetrate deeply into structures and induce strong activation. Highly activated components have to be transferred from the on-load position to the hot cell for the purpose of maintenance. This movement produces a high-dose field, originating from the mobile radiation source, which could affect the functionality of important devices or pose risks to the workers. Therefore, it is imperative to compute the shutdown dose rate (SDDR) maps while transporting activated components in the facility and develop an optimized solution for this calculation process. In this study, we established a new extension of the NASCA method combined with the OTF GVR method to compute SDDR maps. NASCA-OTF is a method developed based on the shutdown dose simulation code NASCA to ensure automatic modification of the geometry structure on the fly. In the meantime, it uses OTF method for speeding up the Calculation. Thus, this modified method produces convergent SDDR maps corresponding to different positions of activated components within a single Monte Carlo (MC) simulation. The NASCA-OTF was verified by comparison with R2S method through benchmark analysis. The method was also used for SDDR analysis of the mobile blanket sector for the China Fusion Engineering Test Reactor (CFETR), and the results indicated that the NASCA-OTF is feasible for computing the SDDR resulting from the movement of activated components.

Observation of fast electron redistribution during saturated kink mode in high βp H-mode discharge with central heating in EAST tokamak

In the EAST tokamak, we have developed an internal transport barrier (ITB) high-confinement mode (H-mode) scenario characterized by dominant electron heating and centrally peaked electron temperature profiles, facilitated primarily through the combustion of lower hybrid current drive and electron cyclotron radio heating (ECRH). Hard x-ray diagnostics reveal a marked increase in the population of fast electrons with energy from 30 keV to 80 keV, concurrent with augmented ECRH power during H-mode plasma operations. This surge in fast electron population precedes the formation of the electron temperature ITB (Te-ITB). Within the Te-ITB H-mode discharge, a mild and long-lived m/n = 1/1 mode (where m and n denote the toroidal and poloidal mode numbers, respectively) emerges proximal to the ITB region. This mode precipitates a redistribution of fast electrons, contributing to an increase in the safety factor near the magnetic axis and thereby promoting the stability of the Te-ITB. Furthermore, we explore the influence of fast electrons on plasma pressure and examine the effects of the profile of fast electrons on the central Te. Strategies to maintain the m/n = 1/1 mode at a moderate amplitude are also discussed, highlighting their significance in the sustained management of Te-ITB.

In Situ Plasma Studies Using a Direct Current Microplasma in a Scanning Electron Microscope

AbstractMicroplasmas can be used for a wide range of technological applications and to improve the understanding of fundamental physics. Scanning electron microscopy, on the other hand, provides insights into the sample morphology and chemistry of materials from the mm‐ down to the nm‐scale. Combining both would provide direct insight into plasma‐sample interactions in real‐time and at high spatial resolution. Up till now, very few attempts in this direction have been made, and significant challenges remain. This work presents a stable direct current glow discharge microplasma setup built inside a scanning electron microscope. The experimental setup is capable of real‐time in situ imaging of the sample evolution during plasma operation and it demonstrates localized sputtering and sample oxidation. Further, the experimental parameters such as varying gas mixtures, electrode polarity, and field strength are explored and experimental V–I curves under various conditions are provided. These results demonstrate the capabilities of this setup in potential investigations of plasma physics, plasma‐surface interactions, and materials science and its practical applications. The presented setup shows the potential to have several technological applications, for example, to locally modify the sample surface (e.g., local oxidation and ion implantation for nanotechnology applications) on the µm‐scale.

Effects of size and static temperature exposure on the shear strength of tungsten bonded CFC blocks for divertor in JT-60SA

The construction of a satellite tokamak, JT-60SA, was completed in 2020. JT-60SA is expected to contribute the research of ITER and developments for DEMO. JT-60SA employs carbon and carbon fiber strengthened carbon (CFC) composites divertors. The divertors will be set on an actively cooled cassette and be remotely handled because of human access restriction into the vacuum vessel due to the radioactivation caused by repeated deuterium plasma operations. An important mission of JT-60SA is high-β steady state plasma research, and during the mission, the use of carbon divertors will be continued. The possibility of metal wall experiments after the complete of high-β steady state plasma research is under discussion. Then the use of remote handled divertor cassette will be a condition, therefore, high performance and appropriate weight divertors are convenient for the operation. One idea is to use clad plates of tungsten and CFC. In previous researches revealed that a sinter bonding method using an insert material of SiC powders + additives was able to join W and CFC. But the research accidentally found out the insert material without explore of the other materials and the size of specimens was small. In present research, some candidates for the insert material were investigated and also the uniformity of strength on a W/CFC block of the expected size were evaluated.

Avoiding fusion plasma tearing instability with deep reinforcement learning

For stable and efficient fusion energy production using a tokamak reactor, it is essential to maintain a high-pressure hydrogenic plasma without plasma disruption. Therefore, it is necessary to actively control the tokamak based on the observed plasma state, to manoeuvre high-pressure plasma while avoiding tearing instability, the leading cause of disruptions. This presents an obstacle-avoidance problem for which artificial intelligence based on reinforcement learning has recently shown remarkable performance1–4. However, the obstacle here, the tearing instability, is difficult to forecast and is highly prone to terminating plasma operations, especially in the ITER baseline scenario. Previously, we developed a multimodal dynamic model that estimates the likelihood of future tearing instability based on signals from multiple diagnostics and actuators5. Here we harness this dynamic model as a training environment for reinforcement-learning artificial intelligence, facilitating automated instability prevention. We demonstrate artificial intelligence control to lower the possibility of disruptive tearing instabilities in DIII-D6, the largest magnetic fusion facility in the United States. The controller maintained the tearing likelihood under a given threshold, even under relatively unfavourable conditions of low safety factor and low torque. In particular, it allowed the plasma to actively track the stable path within the time-varying operational space while maintaining H-mode performance, which was challenging with traditional preprogrammed control. This controller paves the path to developing stable high-performance operational scenarios for future use in ITER.