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Daniel Suarez, Sergey Smolentsev, Anatoliy Vorobev
Alexandra V Dudkovskaia, Jeff Candy, Emily A Belli, Michail Savvas Anastopoulos Tzanis, Steven A.M. McNamara, Andrew Oakleigh Nelson, Yang Ren, Timothy Stoltzfus-Dueck, Carl Friedrich Benedikt Zimmermann
ST-E1 is Tokamak Energy's fusion power plant concept. A series of reference flat-top plasma operating points were developed in [S McNamara et al. Nucl. Fusion 66 (2026) 086008]. The present work focuses on the low-density baseline conceptual power plant (BCP) operating point of [S McNamara et al. Nucl. Fusion 66 (2026) 086008], selected as representative of ST-E1, and investigates its micro-stability, turbulent transport and transport-informed optimisation using first-principles spectral flux-tube electromagnetic gyrokinetic simulations. Linear gyrokinetic calculations, involving electrons, thermal ions, impurities, and thermal and fast helium ions, complemented by dynamic mode decomposition to identify subdominant and stable drift-wave eigenmodes, reveal two principal long-wavelength electromagnetic instability branches: hybrid h-ITG modes driven by thermal gradients and FI-KBMs, i.e., kinetic-ballooning-like modes strongly influenced by plasma beta and fast-ion population. At shorter wavelengths, electron-temperature-gradient modes become unstable towards plasma edge while remaining stable in the core. A comprehensive sensitivity analysis identifies principal equilibrium parameters governing these instabilities. Building on these linear predictions, nonlinear electromagnetic gyrokinetic simulations quantify turbulent transport of thermal plasma and fast helium ions. The reference operating point is found to exhibit saturated, rather than runaway, electromagnetic turbulence, establishing it as a physically meaningful baseline for further optimisation. Turbulent transport is governed primarily by combined effects of plasma beta, safety factor and fast-ion fraction, with intrinsic-rotation-driven ExB shear providing complementary suppression. Although fast ions provide free energy driving FI-KBM branch, increasing fast-ion fraction is found to reduce saturated turbulent transport across all species through nonlinear self-organisation. To explore the resulting multidimensional parameter space, a physics-informed interpolation framework constrained and validated by nonlinear gyrokinetic simulations is developed to enable identification of transport-favourable operating points without exhaustive nonlinear parameter scans. This framework yields a family of transport-favourable operating points for ST-E1 low-density BCP equilibrium and provides quantitative targets for future self-consistent equilibrium optimisation.
Choong-Seock Seock Chang, Seung-Hoe Ku, Xin Zhang, Trenton Brewer, Nicolas Lopez, Chris Marsden
Fish-scale divertor plates are becoming popular in the design of next-generation magnetic fusion reactors to make the edge of one divertor tile sits safely behind the shadowed profile of the preceding tile, hence to protect the leading edge from extreme heat deposition. Recently, ST40 tokamak has observed an extremely narrow and peaked divertor heat-load footprint near the separatrix strike point on the fish-scale plates, on top of the usual ion-drift-width scale footprint. This observation raised concern over the severely localized burn even when the well-known ion-drift-width scale burn issue can be resolved [X. Zhang et al., Nucl. Mater. Energy 41, 101772 (2024)]. In this report we demonstrate that the lost ion gyro-orbits to the tilted tile edges can lead to a significant positive-potential sheath at a practical magnetic field incidence angle that is much greater than the usual incidence angle for the ion gyro-sheath formation on flat strike-surface. This could make the kinetic electron heat-flow spilled over from the confined region down the divertor legs to gain a significant kinetic energy amplification and allow a highly localized heat-load peak under a low edge turbulence condition. Possible intrinsic mitigation mechanism is discussed. This issue should be considered carefully in the design of next generation magnetic fusion reactors.
Steven Stemmley, Daniel O'Dea, Braden Moore, Riley Trendler, Quentin Mooney, Emily Greene, David N. Ruzic
Lin Yang, Pierre-Clément Simon, Emre Yildirim, Jose Manuel Trueba Cutillas, Matthew Robinson, Masashi Shimada
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.
Fabio Camilo de Souza, Kenneth G McClements, Alexander Philip Kofi Prokopyszyn, Antti Snicker, Alex Reyner-Vinolas, Javier Gonzalez Martin, Lucia Sanchis, Rafael Marqués Gómez
This paper presents simulations of alpha-particle transport in ITER driven by static 3D magnetic field perturbations in the high-performance 15 MA Q = 10 baseline scenario, specifically resonant magnetic perturbations (RMPs) arising from edge-localized mode (ELM) control coils, combined with toroidal field ripple (TFR) and effects from ferromagnetic materials. We employ the Lorentz-Orbit Code for Use in Stellarators and Tokamaks (LOCUST), which tracks fast-ion orbits under the Lorentz force and Monte Carlo collisions with the bulk plasma, taking into account the detailed geometry of ITER plasma-facing components (PFCs). LOCUST uses GPU cards to enable the high-resolution modelling required to accurately resolve power fluxes across surfaces with complex morphologies, including unprotected cooling pipes beneath the dome divertor, and the generation of reliable synthetic diagnostics for the ITER Fast Ion Loss Detector (FILD) to support its design. The simulations include a range of ELM control coil current profiles with toroidal mode number n=3. The results indicate that the total alpha-particle energy loss has a negligible impact on plasma performance, remaining below 1% of the alpha energy produced in D–T reactions. Furthermore, the power flux density on the divertor structures and the first wall remains well below design limits and is comparable to thermal and radiative loads. The simulated alpha flux on the FILD scintillator plate is well above the noise threshold and can be significantly higher than conservative estimates.
T. Brewer, X. Zhang, C. Marsden, A. Scarabosio, J. Bryant, P. F. Buxton, M. Fontana, T. Gray, N. Lopez, L. Martinelli, et al.
Very narrow heat flux profiles have been observed on ST40 with decay lengths in the near scrape-off-layer (SOL) of $λ_{q,near}<$~1~mm in H-mode plasmas. Measurements from divertor Langmuir probes are compared with an infrared (IR) thermography system for upper-single null diverted plasmas in L-mode and ELM-free H-mode. SOL current to the grounded divertor targets is measured from the Langmuir probes with profiles exhibiting a similar exponential decay to $λ_{q,near}$ suggesting the near SOL heat flux is related to enhanced electron current to the divertor. Inclusion of electron current contributions to the heat flux transmission coefficient is crucial in capturing similar narrow profiles to those from the IR thermography system. Ongoing upgrades to ST40 will enable more investigations on SOL power decay lengths on inboard and outboard targets in double and single null configurations.
M. Scarpari, X. Zhang, K. Borowiec, P.F. Buxton, G. Calabro, S. Carusotti, A. Ciula, V. Godhani, J.D. Lore, E.N.J. Maartensson, et al.
Sai Zhang, Adriaan A. Riet, Kamya Chandrasekhar, Nathan Hinton
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.
Mohamed Khalid Elsharif Mohamed, Alasdair Burchill, Liviu Maatescu, Emre Yildirim, Jonathan Naish, Matthew Robinson, Vandeep Godhani, Samara Levine, Sandeep Irukuvarghula, Chris Wilson, et al.
A comprehensive analysis of the thermal management and net-power performance is presented for the ST-E1 low aspect ratio tokamak fusion power plant. The study addresses the critical interplay between coolant system design, power cycle integration, and electrical loads, which significantly influence net-power projections and overall plant viability. Specifically, it quantifies the impact of coolant loop design, pressure drops, pumping power, and thermal storage on plant efficiency, and examines the conditions required to achieve the pilot-phase net power target of 300–500 MWe. The findings provide insight into the system-level trade-offs that govern fusion plant viability and inform pathways toward improved net-power performance.
J. Willis, K. Chandrasekhar, P. Cheema, J. England, V. Godhani, E. Guise, S.M. Levine, R. Pocock, A. Scott, A. Shone, et al.
An effective fusion reactor maintenance scheme enables safe operations and short downtimes. This in turn leads to high availability, which is critical to the commercial viability of a power-producing plant. In tokamak-based fusion power plants, the chosen maintenance approach has a significant impact on the spatial design of the tokamak, as well as the surrounding infrastructure, and therefore needs to be considered from the outset. Tokamak Energy has developed a pre-concept design of a fusion power plant, ST-E1. This work describes the major drivers and constraints that have been considered, presents the tokamak architecture and chosen maintenance regime, and discusses how this enables the plant’s two-phased approach to demonstrating commercial operations. It also shows the implications for the design of other systems areas, in particular the machine structural arrangement and bioshield and hot cell layout. The reactor core segmentation and removal scheme replaces entire toroidal segments radially through a large vacuum port, along a single axis only. The result is a change-tolerant machine and plant layout that can accommodate the evolving designs of the tokamak.
E. Yildirim, J. Naish, J. Trueba, S. Levine, A. Kumar, V.B. Gohani, M.K.E. Mohamed, A. Vorobev, L. Yang, P.C. Simon, et al.
An integrated tritium fuel-cycle architecture has been developed for ST-E1, a low-aspect-ratio, high-power fusion power plant. The approach couples a helium-cooled liquid-lithium breeder blanket with a dedicated circulation loop providing tritium extraction, heat removal, and inventory control within a unified system boundary. Downstream of the blanket, the tritium processing system comprises of extraction, purification, isotope separation, and storage subsystems sized to accommodate the plant’s dynamic tritium production and consumption rates. Inventory modelling is used to assess candidate extraction schemes, evaluate tritium residence and holdup, and quantify sensitivities associated with protium control and intermediate inventories. The integrated assessment indicates that the combined blanket-processing architecture can satisfy startup and steady-state tritium requirements within credible operating margins, with a low overall inventory and acceptable doubling time. The results identify tritium extraction sensitivity and fuelling efficiency as the dominant drivers of inventory performance, rather than marginal improvements in breeding ratio. Overall, the analysis demonstrates the feasibility of a self-consistent fuel cycle for ST-E1 and highlights the critical dependencies between blanket inventory management, processing throughput, and overall plant performance.
N.A. Lopez, A.F.P. McAdam, N. Bertelli, S. Shiraiwa, M. Ono, Y. Takase, X. Zhang, M. Borscz, J. Stirling, A. Alieva, et al.
This work describes the physics basis for the proposed auxiliary heating and current drive system on the ST-E1 fusion power plant. The ST-E1 flattop plasma considered here is fully non-inductive with a bootstrap fraction of 0.9 and the remaining current driven by EC waves. Using the recently published physics-based optimization method for EC launchers (Lopez et al 2025 Plasma Phys. Control. Fusion 67 055012), we show that the target flattop ECCD can be achieved with a net efficiency of 52 kA MW −1 using fundamental O-mode (O1) with frequency range 160–200 GHz launched from the low-field side top half of the vacuum vessel (LFS top-launch). From considering two candidate rampup scenarios, we conclude that LFS top-launch O1 ECCD can be equally effective during the early stages of plasma operation, although poloidal steering might be needed. X-mode waves injected from the LFS midplane are also shown to be effective for rampup even when T e < 1 keV. We also present modeling results for the pre-conceptual design of an ICRH system proposed for ST-E1. Using TORIC, we find that an ICRH system aiming for 42–48 MHz and toroidal mode number n ϕ ∼ 10 robustly achieves dominant ion damping via Helium-3 minority heating transitioning to second-harmonic Tritium heating. We then show that such waves can be efficiently generated by a 5-strap traveling-wave antenna (TWA) using the Petra-M code. The TWA has a 40–45 MHz passband within which ∼ 60 % of the power entering the TWA is coupled to the plasma with the remaining ∼ 40 % of the power being transmitted through the TWA and possibly recirculated; the power reflected back into the transmission lines is negligible. This passband structure persists even when the evanescent distance is increased by a factor of two, or when the magnetic-field angle is increased by 30 ∘ , demonstrating inherent load resilience that will be crucial for effective ICRH on ST-E1.
X. Zhang, N.A. Lopez, M. Borscz, J. Kang, Y. Takase, M. Scarpari, C. Marsden, M. Ono, S.A.M. McNamara, E.N.J. Maartensson, et al.
ST-E1 is a low aspect ratio fusion power plant being designed by Tokamak Energy targeting 1.5 GW of fusion power. Characterization of the ST-E1 flat-top scenario is described elsewhere McNamara et al (2026 Nucl. Fusion 66 086008); here we focus on addressing the question of how to ramp-up the ST-E1 plasma from an initial state following breakdown and flux-surface formation to the target flat-top state. Being low-aspect ratio, the available solenoid flux of ST-E1 is limited. Therefore, particular consideration is placed on developing ramp-up scenarios that predominantly use inductive flux provided by external vertical field coils. Through time-dependent modeling with METIS, we show that this is possible when the ramp-up is performed at relatively high plasma density: although auxiliary current drive efficiency is reduced, this is significantly outweighed by (1) higher electron-ion collisional equilibration, (2) higher fusion power once ions become sufficiently hot, (3) higher poloidal beta for increased vertical-field flux, and (4) potentially favorable exhaust compatibilities. Ultimately, we show the target ST-E1 flat-top performance can be reached after a ramp-up period lasting 150 s using less than 40 Vs of solenoid flux (with vertical field providing ∼ 90 Vs of flux). The sensitivity to model assumptions are presented, with the general observation that deleterious effects can be mitigated through minor alterations of the auxiliary power temporal waveform and/or total auxiliary power level. The impact of a solenoid and the auxiliary power mix (electron cyclotron heating only versus electron and ion cyclotron heating) on the ST-E1 ramp-up success are also discussed in appendices. On this latter topic, we show that the effect of direct-ion heating during ramp-up is obscured by the uncertainty in the pedestal dynamics, identifying a clear line of future work required to make a definite decision on the ST-E1 auxiliary power mix.
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.
M. Robinson, A. Scarabosio, E. Vekshina, J.H. Nichols, J.D. Lore, K. Borowiec, J. Varje, S.M. Levine, M. Scarpari, E.N.J. Maartensson, et al.
Power exhaust challenges and potential solutions for a 5 m major radius, low-aspect ratio burning tokamak have been explored. 1D edge plasma models have been used to screen for access to detachment using short and long outer divertor legs in double and single null configurations, using Ar as the primary impurity and assuming tungsten plasma-facing components (PFCs). These show that detachment access can be accessed for all but the most conservative assumptions on scrape-off layer (SOL) width and power, but that trade-offs will be required between magnet engineering and the size of the acceptable window of as-yet uncertain plasma parameters. SOLPS-ITER was used to further model selected plasma scenarios, confirming that Ar seeding can be used to achieve dissipative divertor scenarios with peak deposited heat fluxes below 15 MWm − 2 . Initial scoping of first wall loads and positioning of limiters has been carried out, showing the feasibility of protecting the breeding blanket wall during steady state without impeding tritium breeding. Initial PFC technology selection is also presented, identifying this as a critical area where further work is needed to find an attractive solution for helium-cooled PFCs that can handle high heat fluxes without excessive power requirements. Key questions and trade-offs for concept development have been identified, including: how to achieve high radiation for reduction of SOL power without core performance degradation; whether power exhaust can be well-controlled in a double null plasma; mechanical design and materials challenges of high-heat flux PFCs; and control of material erosion, redeposition and tritium retention.
J. Willis, S.A.M. McNamara, E.N.J. Maartensson, J. Astbury, E. Yildirim, N. Hinton, C.L. Wilson, X. Zhang, J.W. Berkery, L. Baylor, et al.
Climate change and rapidly rising energy demand, driven in part by artificial intelligence and data-centre growth, create an urgent need for stable, low-carbon, and abundant power. Fusion is a promising long-term solution, yet its commercialisation faces a fundamental paradox in today’s investment environment: pilot plants are essential to de-risk physics, engineering, and operations, but their limited lifetime energy output and high upfront costs make them difficult to finance. This paper presents Tokamak Energy’s response: ST-E1, a pre-concept design for a low-aspect-ratio tokamak power plant engineered specifically to overcome this challenge. ST-E1 is designed from the outset for phased operation—pilot and commercial phases, with an upgrade phase in between—with emphasis on commercial viability, maintainability, nuclear engineering, modularity, and upgradability. A key design principle is the deliberate separation of long-lived assets, such as the magnet cage and vacuum vessel, from replaceable in-vessel systems. This provides an attractive and credible investment approach to generate operational data and de-risk key technologies while preserving most capital-intensive assets for later commercial phases. The architecture supports continuous optimisation toward high net electric power (targeting 800–1000 MW net electric), a normalised capital expenditure of $ 12–14k/kW of net electric power, and high availability (targeting > 80 % ). A tokamak core with a 5 m major radius, aspect ratio of 2.3, and on-plasma axis toroidal field of 5.25 T was selected to meet these objectives. This paper summarises the ST-E1 design philosophy, principal features, and development methodology. It introduces a Focus Collection of 11 papers detailing the pre-concept design of the entire tokamak and corresponding plant.
C.L. Wilson, J. Astbury, M.J. Ginsberg, N. Hinton, E.N.J. Maartensson, S.A.M. McNamara, J. Willis, E. Yildirim
To support the pre-concept design of Tokamak Energy’s (TEs) ST-E1 fusion power plant, we developed a new systems code, PyTok. PyTok codifies plasma physics and engineering constraints within a pure Python, object-oriented framework designed for rapid design iteration and exploration of the power-plant parameter space. The code base and its human- and machine-readable data structures enable coupling to external physics codes and data-analysis tools. Informed by TEs design philosophy and techno-economic analyses, the code was used to identify an initial reference design point for ST-E1. A commercially competitive fusion power plant is achievable with a tokamak power core of major radius 5.0 m, aspect ratio 1.9 ⩽ A ⩽ 2.3 and toroidal field 3.25 ⩽ B T ⩽ 5.25 T. The selected design point serves both as a pilot plant and as a commercially competitive reactor using the same permanent power core hardware, enabled by an in-vessel component upgrade phase. The commercial phase targets steady-state operation and a normalised capital expenditure of $12 000–15 000 per kW e , corresponding to a fusion power of 2.0–2.5 GW and a net electric power of 800–1000 MW.
S.A.M. McNamara, S. Abouelazayem, A. Alieva, M.S. Anastopoulos Tzanis, J.W. Berkery, A. Dnestrovskii, A.V. Dudkovskaia, F. Janky, J. Kang, N.A. Lopez, et al.
As part of the U.S. Department of Energy’s Milestone-Based Fusion Energy Development Program, Tokamak Energy has completed the pre-concept design of the ST–E1 fusion power plant. ST–E1 is envisaged to operate in two phases: a pilot plant phase, targeting sustained net power production of 300 − 500 MWe for a duration > 1 hr , followed by a commercial power plant phase targeting steady-state operations and a normalised overnight capital cost of ⩽ 12 000 $ / kWe . The design process adopted was highly iterative, integrating all major plant systems and progressing in a phased fidelity approach. At the pre-conceptual stage, the emphasis has been on exploring the design space, identifying the main system-level trade-offs, and making the key decisions that define the overall plant concept, rather than optimising a single operating point. This paper, part of a focused collection detailing the ST–E1 pre-concept design, addresses the development of a series of reference flat-top plasma operating points for the pilot plant phase. A modelling workflow was established to develop and assess candidate plasma design points and explore key dependencies. The workflow includes integrated core plasma modelling, magnetohydrodynamic (MHD) stability assessment, equilibrium generation, scrape-off-layer and exhaust modelling, heating & current drive design and optimisation, and turbulent transport modelling. Using this framework, the impact of several key parameters on the flat-top operating space was investigated, including the density limit, core radiation fraction and divertor power loading, level of external heating and curent drive power and assumed pedestal characteristics. The MHD stability, controllability and micro-stability characteristics of these plasmas were also analysed. These investigations informed the definition of a set of fully non-inductive, flat-top reference operating points that satisfy the high-level ST–E1 mission, including a low and high density case, a case that is stable to resistive wall modes and a case with reduced divertor power loading.
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