Recent Publications

Stellarators Linking Axisymmetric Mirrors Part 1: Coil Design, MHD Equilibrium, and Physics Metrics

Rahul Gaur, Djin Patch, Robert Babin, Florian Hindenlang, Xu Chu, Tony Qian

We present a computational design study of a stellarator-mirror hybrid concept, Stellarators Linking Axisymmetric Mirrors (SLAMs), in which an optimized poloidally omnigenous (OP or QI) stellarator serves as the rotational-transform-generating element that links long, axisymmetric (circular cross-section) magnetic mirrors. Starting from an OP precursor configuration obtained from the DESC omnigenity database, we construct hybrid coil sets for field period nFP = 2 SLAMs by separating the OP modular coils along the device midplane and inserting axisymmetric planar mirror coils into the resulting gap. Field-line tracing shows that nested vacuum flux surfaces survive this extrusion across a range of mirror-section lengths, with the cross-section qualitatively preserved at the point where the mirror is inserted. We use the field-line tracing data to map the shape of a surface and use the GVEC fixed-boundary solver to calculate an ideal magnetohydrodynamic (MHD) equilibrium and evaluate physics metrics such as the neoclassical transport proxy (eps_eff), the vacuum magnetic well (W), and the gyrokinetic heat fluxes (Q), providing a first-principles analysis of such concepts using modern stellarator analysis tools.

Feasibility of a Flexible, Hybrid Tokamak-Stellarator Experiment using an Axisymmetric Dipole Coil Array

Jacob Halpern, Mohammed Haque, Elizabeth Paul, Carlos Paz-Soldan, Rithik Banerjee, Talia Angles, Frederick Sheehan, Ian Stewart

We demonstrate the design of a flexible, university-scale hybrid tokamak-stellarator experiment based on an axisymmetric array of planar HTS dipole coils. Because the coil array has few geometric degrees of freedom, we use single-stage optimization of the coil currents initialized from two-stage solutions to obtain mutually consistent equilibria and coil sets within realistic engineering limits. We find that the field error and coil current thresholds set minimum and maximum coil-plasma distances that confine the boundary to a roughly fixed axisymmetric envelope, within which rotational transform, volume, coil current, and quasi-symmetry (QS) error trade off against one another. Tighter current limits delocalize the non-axisymmetric shaping and raise QS error at fixed transform. From this single coil array we obtain a broad range of equilibria-quasi-axisymmetric vacuum stellarators with $ι$ up to 0.2, finite-$β$ hybrids with realistic profiles reaching on-axis $ι$ $\approx$ 1 and vacuum transform relevant for MHD stabilization, and strongly shaped tokamaks with elongation $κ$ = 1.7 and triangularity $δ$ = $\pm$0.6, all at peak pointwise coil forces well below the HTS tolerance. We show the same array can additionally correct toroidal field (TF) coil ripple, reducing the number of TF coils required compared to the equivalent tokamak. These results establish the design as a promising platform for hybrid tokamak-stellarator research.

Mechanical Analysis of Helium Inlet and Outlet Structures in the Central Solenoid of the Next-Generation Fully Superconducting Tokamak

Yi Yu, Houxiang Han, Yongsheng Wu, Huan Jin

Neutronics Analysis of Shielding Performance for VNS Inboard Toroidal Field Coil with Various Material Options

Jin Hun Park, Pavel Pereslavtsev, Roman Afanasenko, Dieter Leichtle

Designing a buildable optimised stellarator to confine electron–positron plasmas

Pedro Francisco Gil, Jason Smoniewski, Paul Huslage, Rogerio Jorge, Timo Thun, Elisa Buglione-Ceresa, Tristan Schuler, Stefan Fingl, Grégoire-Hubert Ducas, Eve Virginia Stenson

In this paper, the design of the plasma equilibrium and superconducting coils for the electrons and positrons in an optimised stellarator electrons and positrons in an optimized stellarator (EPOS), experiment is presented. With newly developed stellarator optimisation tools, including single-stage and stochastic optimisation, as well as high-temperature superconductor strain, this work demonstrates that it is possible to achieve key metrics for the buildability and confinement properties of the device. In particular, satisfactory quality of quasisymmetry and stellarator robustness is designed, and engineering requirements are met for eight different candidates. A feasibility study is presented that optimises multiple candidates for different plasma major radii and coil currents, as well as the best EPOS candidate to date, named C4_R19.

Systems analysis of reactor-scale stellarators with optimised plasma–coil spacing

Jedrzej Walkowiak, Felix Warmer, Sehyun Kwak, Andrea Pavone

This study evaluates the feasibility of a stellarator power plant using a new Stable Quasi-Isodynamic Designs (SQuID) configuration within the PROCESS systems code. Our results indicate that for a stellarator constructed with high-temperature superconducting (HTS) technology, a neutron wall load is a primary restriction for size reduction. With an optimistic assumption of 1.5 MW m−2 neutron wall load limit, a major radius of 16.5 m is achievable for 1GW of plant net electric power. We achieve the reduction in size by changing the coil radius in relation to major radius. It is shown that right preselection of the minimal plasma-coil distance at the stage of coil set optimization allows to find optimal solution in machine optimization. Analysis of the proposed coil set suggests that both HTS and low-temperature superconducting (LTS) technologies are viable magnet options. The PROCESS cost model shows that HTS becomes economically advantageous when its material cost is less than five times that of LTS, assuming comparable manufacturing complexity and risk.

Strain analysis on ITER central solenoid conductors and the effect of short twist pitch cabling

Tomone Suwa, Takaaki Isono, Keiya Takebayashi, Yasuhiro Uno, Tsutomu Kawasaki, Masaru Kawabe, Tsutomu Hemmi

In developing the ITER central solenoid (CS) conductors, performance degradation due to electromagnetic loading cycles was identified as a critical issue. This issue was addressed by developing conductors having a short twist pitch, resulting in stable Tcs under electromagnetic loading cycles. While this degradation issue has been resolved, the contribution of the strain state of the Nb₃Sn strands within the conductor to its performance remains unclear, and the mechanism by which the short twist pitch mitigates Tcs degradation has not yet been clarified. Since the electromagnetic loading in DEMO will be larger than that in ITER, it is important to mitigate the strain in the conductor under electromagnetic loading in order to develop conductor whose performance can withstand the electromagnetic loading. To support the development of future conductors for DEMO, strain analysis based on the results of SULTAN test, CS insert test with internal strain measurement by neutron diffraction to investigate both the mechanism by which the short twist pitch mitigates performance degradation and the strain state under actual operating conditions. By integrating these results, it was clarified that the mitigation of Tcs degradation achieved by the short twist pitch cabling originates from strain suppression, especially through the reduction of bending strain. As a result, the short twist pitch design is expected to maintain its performance due to the reduction of the bending strain, and relaxation of axial compressive thermal strain is induced by hoop strain during the operation.

WEST Operation - Reliability and availability of a long pulse fusion tokamak

Valérie Lamaison, Cyril Brun, Elodie Corbel, Annika Ekedahl, Laurent Gargiulo, Sebastien Hacquin, Michael Houry, Lionel Meunier, Philippe Moreau, Lionel Toulouse

Since 2016, the WEST tokamak has demonstrated its capability to perform long plasma discharges approaching 1000 seconds in a fully metallic environment. It operates with a permanent magnetic field of up to 3.65T generated by 18 superconducting toroidal field coils cooled with helium at 1.8K produced by a cryogenic system. Since 2021, all plasma-facing components (PFCs), including the tungsten ITER-grade divertor, are actively cooled by pressurized water, making WEST representative of future superconducting fusion devices. Between 2022 and 2024, experimental campaigns achieved significant improvements in performance. The number of long-duration discharges (>100s) increased threefold thanks to the non-inductive current drive from the Lower Hybrid Current Drive (LHCD) system, culminating in a world record plasma duration of 22 minutes with 2.6 GJ injected energy. Total plasma time exceeded five hours per year, with over 70% of successful pulses. These results are enabled thanks to the availability of the WEST machine and all subsystems, higher than 70%. The study of downtimes recorded during the last three years experimental campaigns, shows four main elements/systems impacting WEST operation: the water and air leaks in vacuum vessel, the poloidal field system, the CODAC (Control, Data Access and Communication) system and the cryogenic system. Key lessons for future fusion devices to achieve plasma long pulses include a high availability of the machine based on a targeted maintenance plan to maximize reliability of all sub-systems and an effective responsiveness in incident diagnosis and repair, particularly in water leak detection on actively cooled Plasma-Facing Components (PFC).

One-dimensional analytic model for the evaluation of the pre-load for the DTT hybrid CS system

F. Petreri, L. D’Alia, L. Muzzi, F. Giorgetti, M. Bombardieri, G. Ruta, A. Anemona, D. Indrigo, G.M. Polli, G. Ramogida

BLUEMIRA: a modular, open-source framework for designing tokamak fusion reactors

Matti Coleman, James Cook, Fabrizio Franza, Ivan Maione, Simon Mcintosh, Hudson Baker, Alexander Ian Blair, Shail Desai, Oliver Funk, Georgina Graham, et al.

As efforts across the world to deliver fusion power enter their respective conceptual design stages, many begin to encounter the difficulties of generating complete designs, evaluating them, performing trade-off studies, and comparing different alternatives. Many organisations successfully use fusion reactor systems codes with low-fidelity formulations of the critical physics and engineering aspects to inform the initial design stage and steer them towards a region of the design space they consider attractive, but struggle to explore the design space in more depth. This work is an overview of the open-source tokamak reactor design framework, BLUEMIRA, built from the BLUEPRINT [1, 2] and MIRA [3, 4] codes, which goes beyond the “classical” systems code paradigm and enables the user to parameterise their reactor designs at higher fidelity. Reactor design workflows ranging from 0-D “radial build” design, equilibrium design, to 3-D CAD and automated neutronics analyses can be run within minutes. We present the functionality implemented in the BLUEMIRA code and demonstrate its application to the conceptual design of a conventional aspect ratio fusion reactor, performing an indicative design study over aspect ratio and number of toroidal field coils.

Permanent-magnet discretization error in stellarators scales linearly with magnetization tolerance

Xiaowei Meng, Ketai He, Guoxuan Ming, Hao Dong, Yangwei Du, Chaofang Dong

Stellarator permanent-magnet concepts discretize a continuously optimized magnetization distribution into a finite number of physical blocks, yet an analytical framework for predicting how the resulting normal-field error scales with the block partitioning parameter has been lacking. Starting from the Biot-Savart integral of the equivalent magnetization currents, this letter derives the power-law scaling of the normalized root-mean-square error (NMSE, defined as the root-mean-square normal-field error on the plasma surface divided by the peak-to-peak normal field of the toroidal-field coils) with the magnetization similarity threshold phi_max (the upper bound on the magnetization-direction angle between any two points within one block). The analysis shows that assigning the volume-weighted average as the equivalent magnetization vector of each block causes the zeroth-order moment of the block error to vanish exactly. When the block linear size l_k is much smaller than the magnet-to-plasma distance r (Regime 1), NMSE ~ C_1 phi_max^2; when l_k ~ r (Regime 2), NMSE ~ C_2 phi_max, where C_1 and C_2 depend only on the configuration geometry. Finite-element simulations of the NCSX quasi-axisymmetric stellarator yield a power-law exponent alpha = 0.94 from a fit to four data points over phi_max = 30-75 degrees, in quantitative agreement with the Regime 2 prediction (alpha = 1). The derived scaling provides a quantitative basis for selecting block partitioning parameters in stellarator and similar tailored-field permanent-magnet systems.

Thermal-hydraulic analysis of the HTS part of the hybrid Central Solenoid designed for the next generation experimental fusion device

Monika Lewandowska, Aleksandra Dembkowska, Rafał Ortwein, Gianluca De Marzi

Alpha particle losses due to ripple and field errors in a SPARC like tokamak

Hugo Emilio Ferrari, Agustín Mairotta, Cesar F Clauser, Ricardo Farengo, Justo Andres Gonzalez Litardo, Alexandra LeViness, Steve Scott

In this work we study the alpha particle losses due to ripple and coil misalignments in the SPARC tokamak. To calculate the effects of ripple and coil misalignment, we developed a code named B3D. The ripple field calculated by this code is combined with 2D equilibrium fields and predicted plasma profiles and used in the FOCUS code to calculate the alpha particle losses. We benchmarked our results against previous studies with the ASCOT and SPIRAL codes, obtaining excellent agreement. When the effect of coil misalignment is included, the alpha particle losses increase and localized regions with significantly larger values of the power density through the last closed flux surface appear. This appears to be due to the low order modes induced in the magnetic field by the misalignment.

The Helios cryostat, maintenance schemes, and cryogenic production system

J. Wasserman, W. Harris, N. de Boer, K. Brodek, J. Jeon, S. Pasmann, C.P.S. Swanson, S. Walsh, B. Berzin, D.A. Gates

A structural beam-based approach for preliminary structural sizing of toroidal coil system: A supplementary module to MADE algorithm

Lorenzo Giannini, Laura Pittaluga, Giuseppe Ruta

Compact Experimental Negative TriAngUlarity Reactor (CENTAUR): A design study for a compact, affordable breakeven tokamak

The CENTAUR Collaboration, Samuel W. Freiberger, Evan Bursch, Javier Chiriboga, Hiro J. Farre-Kaga, Eliot Felske, Sophia Guizzo, John Labbate, Shreyas Seethalla, Frederick Sheehan, et al.

This work presents the compact experimental negative triangularity reactor (CENTAUR), a low overnight cost, high-field tokamak, breakeven reactor design, achieving a predicted total fusion power of 40MW and scientific energy gain of 1.3. Ballooning stability calculations confirm that the device's pedestal is within the first stability regime, which is consistent with the expected ELM-free operation associated with negative triangularity (NT) plasmas. The geometry of the NT divertor allows for high fraction of radiated power (13.5$\%$) between the separatrix and plasma facing components. Heat transport modeling based on simulations of the edge region show heat loads into plasma facing components well below material limits. The magnet system employs rare-earth barium copper oxide (REBCO) high-temperature superconductors in 18 toroidal field coils, an hourglass-shaped central solenoid, and six poloidal field coils to support high-field ($B_0=10.9$ T) plasma confinement, shaping, and current drive. Neutronics analysis shows that a 12 cm $B_4C$ shield keeps superconducting magnet heating below the 33~K quench limit during 10 s, 40 MW DT pulses. With this shielding, the modeled fluence indicates HTS components can survive more than ten times the 3000-pulse design lifetime. Iteration of economic analysis in tandem with the technical design process allows CENTAUR to achieve its overnight cost goal of $\$$2B determined using a custom costing model that predicts a total overnight cost of $1.6$B$\pm0.2$B.

WEST long-pulse achievements in support of next-step fusion devices

Remi Dumont, Theo Fonghetti, Patrick Maget, Pierre Manas, Jean-Francois Artaud, Tullio Barbui, Clarisse Bourdelle, Laurent Colas, Guido Ciraolo, Yann Corre, et al.

The WEST tokamak is equipped with a superconducting toroidal magnetic field system, a multi-megawatt radiofrequency auxiliary power system, and an actively cooled ITER-grade tungsten divertor. As such, it is well adapted to explore experimental aspects related to the long pulse operation of next-step devices. Supported by predict-first integrated modeling, bespoke scenario development has allowed zero-loop voltage pulses to be achieved. The resulting discharges, with plasma currents in the range I p ∼0.22-0.28 MA exclusively sustained by the Lower Hybrid Current Drive (LHCD) system as an auxiliary power source, have achieved durations in excess of 22 min and injected/extracted energies up to 2.61 GJ. Plasma performance is characterized by ranges of poloidal beta β p ∼1.6-2.0, normalized toroidal β N ∼0.6-0.9 and confinement factor H 96L ∼1.0-1.3. Mild MHD activity, identified as resulting from the interaction of 3/1 and 4/1 tearing modes, is occasionally present, depending on the LHCD antenna combination used. This article describes the predict-first approach that has been employed in the context of this long-pulse scenario development endeavor. The main achievements and the physics analyses performed are reviewed, including post-experiment integrated modeling aspects. Prospects for further long-pulse developments are drawn.

Magnetic equilibrium of the CN-H1 heliac: A comparative study of design,as-built, and historically built configurations

Zhengkun Gao, Axel Koenies, Jinjia Cao, Hengqian Liu, Guodong Yu, Dong Xiang, Caoxiang Zhu, Yong-Zhi Dai, Haipeng Wu, Michael Drevlak, et al.

The relocation of the Australian H-1 National Facility (H-1 NF) to China and its reconstruction into the CN-H1 device necessitated a rigorous reassessment of its magnetic configuration to address systematic deviations introduced dur ing the engineering process.This paper presents a comprehensive comparative study of the magnetic equilibrium across three distinct definitions of the de vice: the idealized H-1 Design, the historically built as-built H-1 NF, and the newly reconstructed CN-H1. High-precision laser tracker metrology was em ployed to generate a high-fidelity coil model for CN-H1, incorporating the re manufactured poloidal and helical windings. Calculations of the vacuum mag netic field reveal that, compared to the H-1 NF, the CN-H1 exhibits a systematic inward radial shift of the magnetic axis, a slight increase in the rotational trans form, while maintaining a comparable level of stellarator symmetry breaking. Three-dimensional MHD equilibria were reconstructed using the VMEC code and validated against field-line tracing. Comparative analysis across represen tative configurations demonstrates that the defining characteristics of the flexi ble heliac—including high rotational transform, shear tunability, and magnetic well depth—are preserved in CN-H1. Magnetic spectra in Boozer coordinates of CN-H1 are nearly indistinguishable from the H-1 NF historically built baseline, despite coil positioning errors in the millimeter range. These results validate the engineering fidelity of the CN-H1 reconstruction and establish a precise equilib rium basis for future finite-beta plasma experiments.

Overview of the helios design: A practical planar coil stellarator fusion power plant

C.P.S. Swanson, S.T.A. Kumar, D.W. Dudt, E.R. Flom, W.B. Kalb, T.G. Kruger, M.F. Martin, J.R. Olatunji, S. Pasmann, L.Z. Tang, et al.

Quasi-single-stage optimization for advanced stellarators

Guodong Yu, Yidong Xie, Hengqian Liu, Caoxiang Zhu

Advanced stellarator design requires a balance between plasma performance and the manufacturability of three-dimensional modular coils. In conventional two-stage optimization, the coils required to realize an optimized equilibrium can be limited by engineering feasibility. Here, we develop a quasi-single-stage (QSS) framework that incorporates coil feasibility directly into plasma-boundary optimization. QSS uses the maximum normalized normal-field error, evaluated rapidly from surface currents on a uniformly offset winding surface, as a coil-feasibility surrogate. We apply this method to optimize configurations targeting quasi-axisymmetry, quasi-helical symmetry, quasi-isodynamicity, and a combination of omnigenity with piecewise omnigenity. The QSS-optimized configurations exhibit smoother plasma boundaries and winding surfaces, lower normal-field reconstruction errors, and reduced coil complexity, while preserving favourable magnetic-symmetry and transport properties. QSS provides a practical proof-of-principle strategy for co-optimizing plasma physics and coil engineering in stellarator design.