Recent Publications

Yesterday

On the feasibility of model-based feedback control of vertical instability growth rate using out-vessel coils in ARC-like scenarios

yesterday

Arunav Kumar, Cesar Clauser, Theodore Golfinopoulos, Jon C. Hillesheim

In this work, we propose a model-based feedback controller that regulates the vertical instability growth rate ($γ_{gr}$) of a high-elongation, double-null tokamak directly, using only out-vessel poloidal field (PF) coils. High elongation raises the achievable plasma current and fusion performance but makes the plasma vertically unstable, and in a fusion power plant the in-vessel coils that present devices rely on for stabilization may be absent, leaving only distant out-vessel circuits. The controller couples a machine learning surrogate of non-rigid, profile agnostic vertical instability metric to a constrained quadratic program: the surrogate supplies real-time $γ_{gr}$ estimates and, via automatic differentiation, the actuator sensitivities, while the program allocates coil voltages to track a target growth rate, maintain double-null divertor balance, and respect electromechanical limits. We tested this method on the ARC~V3A power plant design configuration across 24 closed-loop simulations spanning equilibrium variations, actuator degradations, and transient disturbances. We achieved full or marginal success in 83\% of these cases (full in 50\%, marginal in a further 33\%) and lose control in the remaining 17\%; the failures map the boundary of out-vessel controllability (occurring at the highest growth rates) and under actuator limits. The controller does not regulate boundary shape explicitly: separatrix geometry follows indirectly from growth rate and flux balance control and would require a separate shape control layer for sustained scenario evolution.

Neutronic Analyses for the Design of the Chamber of Xcimer’s Athena Fusion Pilot Plant

yesterday

Francisco Ogando, Robert Earley, Wayne Meier, Rene Raffray, Patrick Sauvan, Jaime Marian, Susana Reyes, Conner Galloway

Xcimer Energy Corporation, Universidad Nacional de Educación a Distancia (UNED), Lawrence Livermore National Laboratory, University of California, University of California Los Angeles

Aug 22

Aug 21

Zero-dimensional multi-physics-constrained parameter design and optimization for advanced quasi-isodynamic stellarators

4 days ago

Ziyuan Sun, Zixuan Guo, Xianglin Hao, Longjun Qin, Qian Liu, Xiang Teng

A zero-dimensional (0D) multi-physics-constrained framework for parameter design and optimization of Stable Quasi-Isodynamic Designs (SQuIDs) is presented. Single- and multi-objective optimizations for three staged devices are carried out using an in-house stellarator 0D systems code: YF-1 for discharge demonstration, YF-2 for scientific break even, and YF-3 for a commercial demonstration plant. Pareto searches map the main design trade-offs across the three generations. The equal weight optima for YF-2 and YF-3 both lie in the electron-root favorable regime of the adopted root proxy: YF-2 recovers $Q_{phys} \sim 1$, while YF-3 reaches an ignited point at reactor scale. Future work will couple engineering feasibility and economic assessment modules for integrated plant evaluation.

Aug 20

Aug 17

The status and design challenges of the heating and current drive systems for DTT

Aug 17, 2026

Gustavo Granucci, Silvio Ceccuzzi, Afra Romano, Andrea Murari, Gian Luca Ravera, Piero Agostinetti, Sofia Bertolami, Falk Braunmüller, Alessandro Bruschi, Daniele Busi, et al.

Consorzio RFX, Consiglio Nazionale delle Ricerche, ENEA Agenzia Nazionale per Le Nuove Tecnologie l'Energia e lo Sviluppo Economico Sostenibile, EniProgetti SpA, Ecole Polytechnique Federale de Lausanne

This paper reports the main design keys and the challenging issues of the Heating and Current Drive System (HCD) of the Divert Tokamak Test (DTT) facility that is under construction at ENEA site in Frascati with the aim to perform studies on the power exhaust in a flexible and easily modifiable environment. The selected HCD systems for DTT are those with the most consolidated technology and expected to be relevant for the future reactor. The status of each system is reported, both in terms of design and procurement, which are well advanced for the system required in the first phase of the DTT exploitation: electron and ion cyclotron resonant heating. The third system is neutral beam injector, based on negative ion acceleration, which will be installed in DTT in a second phase, after the first five years of operation. The full heating power will be reached with the doubling of the radiofrequency power to reach the 45 MW at plasma needed to test the divertor with a power density at reactor level.

Aug 14

Overview of the European Breeding Blanket programme for Helias 5-B: Dual Coolant Lead-Lithium design, integration and tools

Aug 14, 2026

Iole Palermo, Javier Alguacil, Gaetano Bongiovi, Juan Pablo Catalan, Ilenia Catanzaro, Iván Fernández-Berceruelo, Salvatore Giambrone, Guillermo Gómez Fonfría, Jose Ángel Noguerón, Vicente Manuel Queral Mas, et al.

CIEMAT, Universita degli Studi di Palermo, UNED

As part of EUROfusion’s mission to bring stellarators to technological maturity, the Stellarator Power Plant Studies (SPPS) WPPRD began in 2021 to develop a HELIAS-class power plant. Building on DEMO tokamak experience, European teams are designing a Dual Coolant Lead-Lithium (DCLL) breeding blanket (BB) for HELIAS. This concept uses liquid PbLi as breeder/coolant and decoupled helium cooling for the first wall (FW). Two key adaptations address HELIAS’s complex geometry: a detached FW using Capillary Porous System (CPS) technology, and a quasi-toroidal segmentation (QTS) with PbLi flow aligned to magnetic field lines. QTS reduces magnetohydrodynamic (MHD) pressure drop by up to two orders of magnitude, potentially eliminating electrical insulation needs. Remote handling (RH) is rethought for 3D stellarators, where traditional vertical-port extraction is impractical. Alternatives include enlarged fixed coils, movable coils for temporary large ports, and detachable vessel periods. The detached FW strategy shifts maintenance from large BB segments to smaller, easily replaceable FW panels, extending BB lifetime. CPS designs with Li in tungsten matrices were analysed thermally, hydraulically, and neutronically, showing potential to lower displacement-per-atom (dpa) damage while maintaining tritium breeding. Be-based moderators behind the FW CPS matrix improved TBR while reducing back-structure damage. To accelerate design and analysis, HeliasGeom and SHANE tools were developed to rapidly generate realistic 3D parametric geometries for CAD, neutronic, and thermal-hydraulic coupling. Preliminary 3D studies addressed TBR optimisation, shielding, MHD in non-uniform fields (via GridapMHD), and multi-scale thermal–mechanical assessments. These innovations collectively advance stellarator blanket technology, integration, and maintainability toward viable power plant concepts.

Aug 13

Multiscale assessment of tritium behavior in preliminary fusion pilot plant design using surrogate models in TMAP8

Aug 13, 2026

Lin Yang, Pierre-Clément Simon, Emre Yildirim, Jose Manuel Trueba Cutillas, Matthew Robinson, Masashi Shimada

Idaho National Laboratory, Tokamak Energy Ltd

The complexity and significance of multiscale phenomena in fusion energy systems make advanced modeling necessary for designing, optimizing, and safely deploying fusion plants. Tritium accountancy is one of those challenges for deuterium–tritium fusion systems. Its availability is constrained by its short half-life (12.33 years) and limited natural abundance, which require fusion plants to breed tritium onsite. Therefore, accurate tritium accountancy is essential for effective resource management, safety, and economics in fusion plants. Through the U.S. Department of Energy milestone program, Tokamak Energy Ltd. is developing a fusion pilot plant design and evaluating tritium retention and loss in key components and their effect on the fuel cycle. To rapidly explore design trade-offs and quantify design decisions on tritium management, this study presents a multiscale analysis to investigate tritium diffusion, trapping, and recovery in key plasma-facing components. To enhance computational efficiency, we integrate surrogate models at the component-level within a fuel cycle model at the system-level, enabling rapid evaluation of tritium recycling dynamics and inventory under various operational scenarios. The goal of this study is twofold: (1) demonstrate the feasibility of utilizing surrogate models to increase the accuracy of fuel cycle modeling, and (2) rapidly evaluate the performance of fusion technologies to accelerate design iterations. This multiscale model provides the tritium transport and retention behavior and supports the plasma-facing components design optimization in normal and bake-out operations. The work is implemented using the Tritium Migration Analysis Program, Version 8 (TMAP8), an open-source application for tritium transport analysis in fusion systems.

Aug 12

Aug 11

Aug 10

Steady state, core, operational optimization of an ARC-like tokamak via plasma composition and shape

Aug 10, 2026

A. Saltzman, P. Rodriguez-Fernandez, A. Ho, G. Snoep, J. Han, J. Hall, M. S. Anastopoulos Tzanis, J. Hillesheim, A. J. Creely, P. Snyder, et al.

Impurity composition, plasma shape, and pedestal density all provide strong levers on fusion power. Here, we explore the ways in which their variation changes fusion power and seek to find the optimum of these parameters. The key impacts of these variables are through changes in the core turbulent transport, the density of the fuel species, the pedestal pressure, and the plasma volume. ITG stabilization due to increased amounts of impurities is observed. The dependence of all of these parameters on the pedestal pressure is especially complicated because of the separate impacts on the peeling and ballooning modes, which can each limit the pedestal. Optimization of this multidimensional operating space is enabled by the use of Bayesian optimization, resulting in an operating point similar to ARC V3A with ~30% more fusion power and a higher fusion power density. Increased shaping parameters, including elongation, triangularity, and squareness are all beneficial, as is high Zeff. When elongation is also allowed to vary, a ~65% increase in fusion power can be achieved. While not commonly considered, we find squareness is an important lever on fusion power. The plasma performance is limited by the Greenwald density limit constraint. This workflow developed here and demonstrated with the example of ARC V3A can readily be applied to other tokamak designs.

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