1-D Kinetic Control Simulations of ITER with Maximization of Ion Cyclotron Power Absorption

V. Graber, E. Schuster, C.C. Klepper, E. Lerche, T.M. Biewer, J.D. Lore

34th Symposium on Fusion Technology (SOFT)

Aix-en-Provence, France, September 21-25, 2026

Future reactor-grade tokamaks that produce fusion power, such as ITER, will require real-time control of the plasma temperature and density. To enable the development and assessment of kinetic control strategies for ITER, the control-friendly, MATLAB©-based COTSIM© (Control-Oriented Transport SIMulator) code has been updated with improved one-dimensional (1-D) transport models [1] that can simulate the entire plasma discharge, including the entry to and exit from burn. In this work, a zero-dimensional (0-D) burn control strategy specifically design for ITER [2] is evaluated in the 1-D COTSIM© environment. This nonlinear burn control strategy is designed to maximize the plasma’s absorption of Ion Cyclotron Resonance Heating (ICRH) power through the injection of helium-3 gas. A key component of this control strategy is the Diagnostic Residual Gas Analyzer (DRGA), which is a specific diagnostic tool planned for ITER operations. Unlike present-day experimental tokamaks, much of the first wall and divertor in future fusion reactors will be covered in blanket modules to maximize tritium breeding and energy conversion. Because the blanket modules will occupy most of this valuable space, diagnostic systems will have much less access to the interior of the vessel for sampling the plasma. This presents a significant challenge for feedback control algorithms that depend on real-time measurement data. The DRGA is designed to circumnavigate this limitation by accessing only the sub-divertor region. More specifically, the DRGA is capable of providing real-time measurements of particle concentrations, including hydrogen isotopes (the deuterium and tritium fuel) and helium isotopes (the helium-4 fusion product and the injected helium-3 particles), in the sub-divertor region. Tokamak experiments have shown that particle concentrations in the sub-divertor region correlate well with their concentrations in the core-plasma region after a time delay [3], meaning that the DRGA can monitor the composition of the main plasma. With time-delayed measurements of the particle concentrations, the feasibility of the proposed DRGA-enabled control strategy is evaluated in a 1-D simulation environment. By maximizing the plasma’s ICRH power absorption through external helium-3 gas injection, the control strategy is shown to substantially increase in the fusion power output.

[1] V. Graber and E. Schuster 2025 Fusion Engineering and Design 221 115362
[2] C. C. Klepper, et al 2025 Nuclear Fusion 65 086015
[3] C. C. Klepper, et al 2019 Nuclear Fusion 60 016021

*Supported by the US DOE under DE-SC0010661.