Recent Publications

Yesterday

Nuclear Fusion

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.

Sep 7

Nuclear Fusion

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.

Sep 3

Aug 31

Aug 26

Plasma Physics and Controlled Fusion

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.

Aug 20

Plasma Physics and Controlled Fusion

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

Plasma Physics and Controlled FusionAug 20, 2026Plasma & ConfinementMagnets & SuperconductorsAI, Modeling & Simulation

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.

Aug 19

Nuclear Fusion

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.

Aug 17

Aug 4

Nuclear Fusion

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.

arXiv (physics.plasm-ph)

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.

Jul 24

Jul 14

arXiv (physics.plasm-ph)

Optimized finite-$β$ tokamak-stellarator hybrid configurations achieved by planar dipole-field coils

Yihui Liang, Hengqian Liu, Guodong Yu, Zhenyu Zhou, Caoxiang Zhu, Yao Zhou

Tokamak--stellarator hybrids seek to combine tokamak-like compactness and confinement with stellarator-like externally generated rotational transform and steady-state operation. In this work, we build on the recent tokamak--stellarator hybrid study using planar dipole-field coils (PDCs) [Yu et al., arXiv:2605.03599], in which the fixed-position, programmable coils on an axisymmetric winding surface generate flexible three-dimensional shaping fields. Using single-stage free-boundary optimization of coil currents and plasma-equilibrium parameters, we construct vacuum and finite-$β$ configurations. The vacuum cases show controllable external transform and magnetic well. The finite-$β$ cases accommodate various density, temperature, and pressure profiles, producing quasi-axisymmetric (QA) equilibria with self-consistent bootstrap current, favorable Mercier stability, and reduced demand for external current drive. Re-optimization enables $β$ ramp-up and access to different field-period QA branches with moderate coil-current changes. At large rotational transform, a toroidally omnigenous (TO)-like configuration exhibits more favorable infinite-$n$ ideal-ballooning behavior than a QA reference with matched profiles, even though ballooning stability is not directly optimized for. These results demonstrate that PDCs provide a flexible platform for achieving optimized finite-$β$ hybrid configurations.

Jul 10

Nuclear Fusion

Integrated physics and magnet design for the ST-E1 fusion power plant

E.N.J. Maartensson, N. Welch, M. Scarpari, F. Malinowski, M. Borscz, M. Bristow, P.F. Buxton, J. Kang, M. Kruip, A.V. Langtry, et al.

ST-E1 is Tokamak Energy’s commercially competitive fusion power plant design featuring a lifetime high-temperature superconducting (HTS) magnet cage, which has completed its pre-concept design stage. A central challenge at the pre-concept stage is the need to iteratively and consistently integrate the development of the magnet cage, core plasma physics, and power-exhaust systems, while avoiding serial design loops and late discovery of infeasible operating scenarios. Accurate representation of inductive current drive is particularly critical, as it tightly couples magnetic equilibria, flux swing, and HTS operational limits across these systems. To address this challenge, a novel integrated workflow has been developed to enable rapid, early-stage iteration during the pre-concept design phase. A key novelty is the deliberate de-coupling of magnetic equilibria generation from coil optimisation and detailed engineering, allowing a broad range of time-dependent operating scenarios to be explored without repeated high-fidelity coil redesign. The workflow combines Metis, a fast 1.5D fixed-boundary modelling tool, with FreeGS, a free-boundary equilibrium solver, to generate flat-top and time-dependent ramp-up scenarios together with the associated flux-swing traces and magnetic equilibria. These outputs are passed to SCOPE, an in-house tool that optimises poloidal field and central solenoid coils based on physics requirements while respecting HTS operational limits. The resulting coil designs are then refined within a dedicated coil-engineering workflow performing higher-fidelity structural, electromagnetic, and HTS analyses. The integrated workflow has been applied to evaluate a wide range of ramp-up, ramp-down, and flat-top scenarios, enabling consistent down-selection of technically achievable operating points across all phases of ST-E1operation. The resulting set of coils and operating scenarios collectively satisfy the full system requirements of the ST-E1 pre-concept design.

Nuclear Fusion

Nuclear technology considerations and neutronics for the ST-E1 fusion power plant

Kamya Chandrasekhar, Christopher Lister Wilson, Samara Michelle Levine, Sandeep Irukuvarghula, Vandeep Bharatbhai Godhani, Jonathan David Naish, Ariel Aníbal Márquez, Michael J Loughlin

This paper presents Tokamak Energy’s pre-concept design of ST-E1, a low aspect ratio tokamak fusion reactor. Efficacy of radiation shielding in protecting the high temperature superconducting magnets systems is a major driver dictating tokamak size, plant lifetime, and balance of plant. It is crucial that neutronics and material considerations are incorporated into plant design from the earliest design stages and comprehensively taken into account across all areas through the systems engineering framework. At Tokamak Energy, a dedicated workflow for plant-scale neutronics simulations has been developed using G4Tokamak, an in-house application based on the Geant4 toolkit. This workflow enabled faster and more efficient design iterations, and its results were verified through benchmarking against MCNP. A four-part shielding architecture, comprising of the centre column, vacuum vessel, outboard horizontal and vertical elements, was designed to balance neutronics performance, thermal management, structural demands, and cost. The shielding system, at this pre-concept stage, demonstrates the feasibility of limiting the neutron flux to the magnet systems to support a full power life of 30 years, and limiting the nuclear heating of the cold mass within cryogenic cooling capacity. Activation studies highlight maintenance and long-term waste management challenges, with non-reduced activation ferritic martensitic steels and plasma facing components showing the highest activation. Collectively, these results establish a comprehensive neutronics and materials framework, baseline core shielding design for ST-E1, and demonstrate its readiness for progression to detailed concept development.

Jul 8

Jul 7

arXiv (physics.plasm-ph)

Towards joint optimization of stellarator coils and support structures

Lanke Fu, Alan A. Kaptanoglu

The support structure is an integral part of the design of nuclear fusion reactors, especially 3D stellarator devices. A practical reactor's coils and support structures must have three competing qualities: an accurate magnetic field for good confinement, sufficient rigidity to protect the brittle high-temperature superconductor (HTS) from damage, and a simple geometry for low-cost construction. In existing devices, the coil geometry is often optimized without knowledge of the support structures' design and the coils' true stress and deformation. The support structures are then placed by hand through repeated finite element analyses (FEA) until engineering requirements are met. This makes the structural design of stellarator coil systems lengthy and labor-intensive. Using new developments in differentiable structural mechanics, we present coil-fem, an open-source software tool that integrates support differentiable FEA into the stellarator coil optimization loop. It enables the integrated optimization of coil geometry and support clamp locations to simultaneously reduce magnetic field errors and stresses in the coil body. We also present the first combined coil-support optimization in the stellarator literature. Using a penalty term based on coil-fem, we produced a coil set with 2.4x lower RMS von Mises stress and similar field error compared to an unoptimized baseline.

Jul 1

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