Advances and challenges in KSTAR plasma control toward long-pulse, high-performance experiments

Self Cite

High-performance long-pulse plasma operation is essential for producing economically viable fusion energy in tokamak devices. To achieve such discharges in KSTAR, firstly, the rapid increase in the temperature of plasma-facing components was mitigated. The temperature increase of the poloidal limiter, especially, was associated with beam-driven fast ion orbit loss and the discrepancy of the equilibrium reconstructed with heated magnetic probes of signal drift. The fast ions lost to the poloidal limiter were reduced by optimizing the plasma shape and the composition of neutral beam injection (NBI). This nonlinear signal drift was successfully reduced by a new thermal shielding protector on the magnetic probes. Secondly, a lower loop voltage approach was implemented to reduce a poloidal flux consumption rate. A plasma current of 400 kA and a line-averaged electron density of ∼2.0 × 1019 m−3 were chosen by considering the L–H power threshold, fast ion orbit loss, and beam shine-through power loss for low loop voltage in KSTAR. In addition, the application of electron cyclotron heating also helped maintain the plasma with low loop voltage (∼25 mV) by enhancing the NBI-driven current and achieving a high poloidal beta (β P) state. KSTAR has achieved a long pulse (∼90 s) operation with the high performance of β P ⩽ 2.7, thermal energy confinement enhancement factor (H98y2) ∼ 1.1, and fraction of non-inductive current (f NI) ⩽ 0.96. Still, gradual degradation of the plasma performance has been observed over time in the discharges. In one of the long-pulse discharges, β P reduced by ∼18% over the time of ∼8τ R (current relaxation time, τ R ∼ 5 s) and ∼1067τ E,th (thermal energy confinement time, τ E,th ∼ 45 ms). The degradation may be closely associated with weak, yet growing, and persistent toroidal Alfvén eigenmodes and their effect on fast ion confinement.

Section: Experimental Results Of High-performance Long-pulse Dischargesmentioning

confidence: 99%

“…KSTAR interlock systems [6], operated to protect the PFCs, continuously monitor the temperature of PFCs using thermocouples at a rate of 1 Hz. If the temperature of PFCs exceeds the designated threshold for more than 3 s after initial detection, the interlock system terminates the discharge process.…”

Section: Mitigation Of the Temperature Increase In Pfcsmentioning

confidence: 99%

Development of high-performance long-pulse discharge in KSTAR

Kim,

Jeon,

Han

et al. 2023

Self Cite

“…With a minor radius of a = 57 cm, figure 5 achieves ∆Z max /a ∼ 6%−7% on SPARC, implying that safe operation can be expected. Similar metrics (∆Z max /a = 5%) have been adopted by ITER and demo designs [8,46,47]. We note, however, that the maximum excursion of a controlled trajectory would lead likely to a limited plasma on SPARC, especially if radial motion is included in the disturbance.…”

Section: Maximum Controllable Displacementmentioning

confidence: 86%

Implications of vertical stability control on the SPARC tokamak

Nelson,

Garnier,

Battaglia

et al. 2024

To achieve its performance goals, SPARC plans to operate in equilibrium configurations with a strong elongation of κ areal ∼ 1.75 , which in turn will destabilize the n = 0 vertical instability. However, SPARC also features a relatively thick conducting wall that is designed to withstand disruption forces, leading to lower vertical instability growth rates than usually encountered. In this work, we use the TokSyS framework to survey families of accessible shapes near the SPARC baseline configuration, finding maximum growth rates in the range of γ ≲ 100 s−1. The addition of steel vertical stability plates has only a modest ( ∼ 25 % ) effect on reducing the vertical growth rate and almost no effect on the plasma controllability when the full vertical stability system is taken into account, providing flexibility in the plate conductivity in the SPARC design. Analysis of the maximum controllable displacement on SPARC is used to inform the power supply voltage and current limit requirements needed to control an initial vertical displacement of 5% of the minor radius. From the expected spectra of plasma disturbances and diagnostic noise, requirements for filter latency and vertical stability coil heating tolerances are also obtained. Small modifications to the outboard limiter location are suggested to allow for an unmitigated vertical disturbance as large as 5% of the minor radius without allowing the plasma to become limited. Further, investigations with the 3D COMSOL code reveal that strategic inclusion of insulating structures within the VSC supports are needed to maintain sufficient magnetic response. The workflows presented here help to establish a model for the integrated predictive design for future devices by coupling engineering decisions with physics needs.

“…Nevertheless, KSTAR deploys a rather conservative protection of the critical superconducting and other systems. For example, it aims at preventing quenching of its poloidal field coils [36], mitigating damage of the first wall by plasma heating modules etc. These protection schemes are included in the drivers of the USD DECAF event and it can be seen in figure 3(b) that prior to KSTAR 2021 this event was mainly deployed at plasma current I p ⩽ 0.1 MA.…”

Section: Disruption Recognition Via I P Quench Detectionmentioning

confidence: 99%

DECAF cross-device characterization of tokamak disruptions indicated by abnormalities in plasma vertical position and current

Zamkovska,

Sabbagh,

Tobin

et al. 2024

Self Cite

Abnormal (deviating from target) variations in the plasma vertical position Z and current Ip (such as vertical displacements, transient Ip 'spikes' and quenches) constitute common elements of a disruption, a phenomenon that is to be mitigated, or ultimately avoided in future reactor-relevant tokamaks. While those abnormalities are generally recognized cross-shot and cross-device, details in terms of appearance (or not) and order of those abnormalities in disruption event chains is bound with the plasma state at the time of the chain initiation. Detection of those abnormalities is thus indicative not only of the onset of the plasma collapse itself, but also of the disruption driving cause that is promoted at the particular plasma state. Here, occurrence of disruptions, explored via detection of a Ip quench, and analysis of disruption event chains constituted by Ip and Z abnormalities, is reported for in total 7 full years of operation of 3 devices (KSTAR, MAST-U and NSTX-U) using the DECAFTM code expanded tools and capabilities. It is shown that the disruption occurrence depends not only on details of the plasma state, but also on (device-dependent) technical elements of the shot exit scenario. A year-to-year change in main disruption causes and a reduction of the disruptivity rate, bound with device and operation upgrades, are reported. Particular trigger instances of disruption event chains (and the full chains, when applicable) are shown to occupy different parts of the operation space diagrams, in accordance with prior expectation. Plasma elongation is identified as an important factor influencing details of the chains and its role will be further explored.