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Gabriel D. Parker, Zachariah Koyn, Tim Graening, Hsin Wang, Xiang (Frank) Chen, Patrick E. Albert, Xiao-Ying Yu
Oak Ridge National Laboratory, Energy Driven Technologies LLC, The Pennsylvania State University
Jin Whan Bae, Arpan Sircar, Katarzyna Borowiec, Vittorio Badalassi, Cami Collins
Oak Ridge National Laboratory
Accurate modeling of particle transport within fusion blankets is essential for predicting performance metrics such as heat deposition and the tritium breeding ratio (TBR). However, high-fidelity coupling of thermal fluids from computational fluid dynamics (CFD) to neutronics simulations often incurs significant computational costs due to the complexity of surface intersection calculations in Monte Carlo codes. This paper presents an accelerated multiphysics coupling method for neutronics that utilizes hierarchical agglomerative clustering to map complex material property distributions to a neutronics model. Implemented within the Fusion Reactor Design and Assessment (FREDA) framework, the method leverages existing Python packages to automate the creation of clustered geometries for OpenMC. The approach is demonstrated on a sector model of an ARC-class tokamak with an immersion molten salt blanket, and an simple geometry with varying isotopic concentrations. Results show that the clustering method significantly reduces computational burden without compromising fidelity, providing a foundation for agile iteration of neutronics simulations involving multiple coupled material properties.
Rishi Pillai, Debashish Sur, Severine Cambier, Adam Willoughby, Tracie Lowe, Youyang Zhao, Amanda Bye, Susana Reyes
Oak Ridge National Laboratory, XCIMER Energy Corporation
Gilson Ronchi, Robert S. Wilcox, Jonathan H. Yu, Andreas Holm, Filippo Scotti, Adam McLean, Morgan Shafer, Joshua Hicok, Tyler Elsey
Oak Ridge National Laboratory, General Atomics, Lawrence Livermore National Laboratory
J.D. Lore, F. Scotti, A. Marinoni, K.E. Thome, C. Lasnier, M. Shafer, D. Truong, H.Q. Wang
Lawrence Livermore National Laboratory, Oak Ridge National Laboratory, General Atomics, University of California San Diego
Jhovanna Garcia, Paul W. Humrickhouse, Steven J. Zinkle, Rinkle Juneja
Oak Ridge National Laboratory, University of Tennessee
Nicola Lonigro, Kevin Verhaegh, James R Harrison, Bruce Lipschultz, Christopher Bowman, Fabio Federici, Jack Flanagan, Daniel Greenhouse, David Moulton, Peter Ryan, et al.
Culham Centre for Fusion Energy, United Kingdom Atomic Energy Authority, University of York, Eindhoven University of Technology, ORNL
The first high-power (≥ 3 MW) H-mode experiments using a double-null X-point-target (XPT) divertor configuration have been performed on MAST-U. The XPT geometry is obtained by combining a large strike point radius, similar to the Super-X divertor (SXD), with an additional X-point near the separatrix in the baffled outer divertor chambers and leads to additional exhaust benefits over the SXD. The broader electron density profile near the secondary X-point leads to additional plasma-neutral interactions, evidenced by a broader hydrogenic emission profile, and resulting in larger power and ion sinks. The increase in plasma-neutral interactions also leads to lower target electron temperatures and heat fluxes. These benefits appear to extend to transients, and preliminary evidence of improved ELM buffering in the XPT is presented. These results showcase how multiple alternative divertor configuration strategies can be combined to improve momentum, power, and particle losses, which may be required for the challenging exhaust conditions of future reactors.
Atul Kumar, Dhyanjyoti Nath, Wouter Tierens, Jeremy D Lore, R S Wilcox, Gilson Ronchi, Morgan W Shafer, Aditya Y Joshi, Onkar Sahni, Mark Shephard, et al.
Oak Ridge National Laboratory, Rensselaer Polytechnic Institute, General Atomics, University of Wisconsin-Madison
The high-power helicon wave system in the DIII-D tokamak could potentially introduce new plasma--material interaction (PMI) challenges owing to rectified RF sheath potentials that develop near the antenna and surrounding plasma-facing components. We present the first application of the STRIPE (Simulated Transport of RF Impurity Production and Emission) framework to helicon-induced PMIs, extending previous STRIPE studies of ICRH antennas by incorporating net erosion, local re-deposition, and three-dimensional global impurity transport. The integrated workflow couples SOLPS, COMSOL, RustBCA, GITR, and GITRm to simulate carbon erosion, re-deposition, and impurity transport for two experimentally constrained DIII-D H-mode helicon operating scenarios with different antenna--plasma gaps, coupled RF powers, and edge plasma conditions. COMSOL predicts rectified RF sheath potentials of 1--5~kV localized near the lower portion of the antenna, where the magnetic field intersects the surface at grazing incidence. Carbon self-sputtering dominates the erosion source, whereas RF-accelerated D$^+$ ions contribute approximately 1\% of the total gross erosion. The smaller-gap operating scenario exhibits substantially stronger gross erosion, enhanced local re-deposition ($\sim$12\%), and a larger confined carbon inventory owing to increased plasma accessibility and broader RF sheath coverage. Comparison with available DIII-D measurements shows no distinct change in the global carbon signal that correlates with the helicon RF pulse, consistent with the simulations indicating that the helicon-generated carbon source remains small compared with the existing background carbon inventory under the present graphite-wall operating conditions. These results demonstrate the capability of STRIPE to integrate RF sheath modeling, plasma transport, surface interaction physics, and three-dimensional impurity transport for the interpretation of helicon-induced PMIs. The study further identifies the principal sources of modeling uncertainty, including grazing-angle RF sheath physics, slow-wave resolution, plasma-background extrapolation, and trace-impurity transport assumptions, providing a framework for future validation and model development.
Alexey Romanovich Knyazev, Alexandra L. Lachmann, Alan Geoffrey Goodman, Abdullah Hyder, Michael Czekanski, Donald Spong, Elizabeth J. Paul
Columbia University, Max Planck Institute for Plasma Physics, Cornell University, Oak Ridge National Laboratory
In this work, we investigate prompt ion drift orbit losses caused by shear-Alfvén waves (SAWs) in quasi-symmetric and quasi-isodynamic (QI) stellarators optimised for equilibrium confinement of energetic particles (EPs). We use the ideal reduced magnetohydrodynamic (MHD) model for SAW perturbations and study their impact on the collisionless EP drift dynamics. We present a semi-analytical model for resonance between the passing EP and SAW, generalised to arbitrary quasi-symmetric configurations including the quasi-poloidal case relevant to QI equilibria. Analysis reveals that an increase in the number of field periods upper N Subscript normal f normal p N f p $N_{\mathrm{fp}}$ suppresses stochasticity in quasi-helical (QH) and quasi-isodynamic, but not quasi-axisymmetric (QA) stellarators. We show that wave-induced transitions between passing and trapped orbits cause significant losses in QA and QH, but not in QI configurations. For the considered equilibria at scales relevant to fusion power plants, we numerically determine the SAW amplitudes needed to induce prompt loss of fusion-born alpha particles. Using the weighted Birkhoff averaging technique, we confirm that the onset of prompt losses across all orbit classes occurs with the onset of stochasticity in ion motion. This motivates extending the stochasticity-onset criterion beyond passing orbits in future work.
Jacobo Varela Rodríguez, Kenichi Nagaoka, Masato Matsuoka, Hideo Nuga, Ryosuke Seki, Kunihiro Ogawa, Kenji Tanaka, Y Todo, Hao Wang, Jialei Wang, et al.
National Institutes of Natural Science, University of Texas at Austin, Oak Ridge National Laboratory, Nagoya University, Universidad Carlos III de Madrid
Burst events can severely deteriorate the performance of a future stellarator fusion reactor by reducing the plasma heating efficiency and confinement. Bursting activity is observed in the Large Helical Device (LHD) operation scenarios strongly heated by neutral beam injectors (NBI). This study analyzes the performance degradation in the LHD discharge $189008$ linked to the destabilization of burst events. On that aim, the stability of Alfv'en Eigenmodes (AE) and energetic-ion-driven resistive interchange modes (EIC) is analyzed. Linear and nonlinear simulations using the gyro-fluid code FAR3d reproduce the multiple Toroidal AEs (TAEs) measured in the frequency range of $40-80$ kHz leading to the destabilization of an MHD burst (induced by passing energetic particles injected by the tangential NBI), as well as the EIC destabilized around $12$ kHz that causes the EIC burst (caused by helically trapped energetic particles injected by the perpendicular NBI). This study suggests burst events can be induced by the nonlinear interaction between passing (tangential NBI) and helically trapped (perpendicular NBI) energetic particles (EP) populations, creating a feedback effect that enhances the EP drive. The MHD burst causes $30\%$ of passing EP losses and the EIC burst up to $90\%$ of trapped EP losses, EP confinement degradation that may partially explain the decrease of the device performance. The analysis also indicates the EP transport induced during the burst events is ballistic linked to avalanche-like processes, leading to the intense EP fluxes. In addition, the bursting activity affects the thermal plasma confinement observed as a decrease of the thermal electron density, reproduced in the simulations as a decay of the thermal plasma pressure. The simulations output shows a reasonable agreement with different diagnostic measurements, including the enhanced magnetic perturbations detected by the Mirnov coils once the perpendicular NBI operation and bursting activity begin, the intensity and radial location of the radial electric field and shear flows from charge exchange spectroscopy diagnostic (CXS) data as well as the instabilities frequency range and dominant modes numbers calculated from magnetics data. The main conclusion of the analysis is that, the nonlinear feedback between different EP populations can enhance the EP drive leading to the onset of bursting events and the degradation of burning plasmas performance, particularly in scenarios with a large amount of EPs generated by different heating mechanisms that may destabilize the alpha particle population.
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.
Tokamak Energy Ltd, University of Tuscia, Oak Ridge National Laboratory
Scott Greenwood, Katarzyna Borowiec, Wesley Williams, Cory Ball, Susana Reyes, Amanda Bye
Oak Ridge National Laboratory, Xcimer Energy
Hassan Al Khawaldeh, Andres Pajares, Sai Tej Paruchuri, Vincent Graber, Tariq Rafiq, Eugenio Schuster, June-Woo Juhn, T Pederson, Daisuke Shiraki, Francesca Turco
Lehigh University, General Atomics, Korea Institute of Fusion Energy, Oak Ridge National Laboratory, Columbia University
Precise regulation of plasma density is crucial for achieving and maintaining fusion-relevant conditions in reactor-grade tokamaks. Tokamak reactors will utilize both gas puffing and pellet injection as standard fueling mechanisms. However, the coordinated use of gas puffing and pellet injection within feedback control frameworks is largely unexplored, underscoring the necessity to develop and validate dedicated strategies on existing machines. These strategies must address the various challenges associated with both actuators, including actuation delays, unknown fueling efficiencies, and the coexistence of continuous-time and discrete-time dynamics. To overcome these challenges, which surpass the capabilities of traditional empirically tuned Proportional-Integral-Derivative (PID) control, an indirect adaptive control algorithm is proposed in this work for the regulation of the line-averaged electron density through coordinated gas puffing and pellet injection. After initial validation in simulations with a multiple-reservoir global particle model, the controller was successfully implemented and tested on the DIII-D tokamak. Experimental results demonstrate robust density tracking under reactor-relevant scenarios, showcasing the controller's ability to seamlessly coordinate actuators while handling disturbances and evolving actuator constraints. This work provides a critical step forward in the control of future fusion reactors.
J. Varela, J. Garcia, D. Zarzoso, S. Mazzi, Y. Kazakov, Z. Stancar, M. Baruzzo, J. Ongena, D.A. Spong, L. Garcia, et al.
United Kingdom Atomic Energy Authority, University of Texas at Austin, Universidad Carlos III de Madrid, CEA, Oak Ridge National Laboratory
The present study is dedicated to analyze the extensive fast ion (FI) induced perturbations in ion cyclotron resonance heating (ICRH) dominated plasmas in JET deuterium–tritium (D–T) discharges, the closest experiment to reactor-like operation. A set of nonlinear simulations are performed with the gyro-fluid FAR3d code to reproduce the FI induced activity and energetic particle (EP) transport observed in the discharge 99896 by multiple EP populations. Well validated simulations with JET D–T diagnostics show that Toroidal Alfvén eigenmodes (TAEs) and fish-bone are destabilized by ICRH accelerated passing D and trapped H ions, respectively. Multiple EP simulations show important nonlinear couplings between EP species, leading to a strong effect on the alpha particle confinement even in the alpha particles do not destabilize EP modes. Simulations including trapped H and alpha particles indicate fish-bones may induce alpha particle losses up to 3 % . The situation is different with respect to TAEs, which have a smaller impact on the alpha particle transport, leading to negligible losses. In addition, the generation of shear flows during the saturation phase of TAE and fish-bones is observed in the simulations. Shear flows induced by TAEs in the inner-middle plasma region are almost two times larger compared to the shear flows caused by the fish-bone in the inner plasma region. Shear flow generation by AE/fish-bones may have an important impact in the thermal plasma and alpha particle confinement of devices as ITER and JT60SA. Consequently, nonlinear effects in multiple EP populations must be considered in the analysis of EP transport. It is shown that this is a fundamental result that has to be considered to be properly evaluate fusion power in reactor relevant plasmas.
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.
Tokamak Energy Ltd, Savannah River National Laboratory (SRNL), Oak Ridge National Laboratory (ORNL), Idaho National Laboratory (INL), The Pennsylvania State University
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.
Kamya Chandrasekhar, Christopher Lister Wilson, Samara Michelle Levine, Sandeep Irukuvarghula, Vandeep Bharatbhai Godhani, Jonathan David Naish, Ariel Aníbal Márquez, Michael J Loughlin
Tokamak Energy Ltd, Oak Ridge National Laboratory
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.
Tokamak Energy Ltd, Oak Ridge National Laboratory, Fondazione LINKS, Princeton Plasma Physics Laboratory
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.
Tokamak Energy Ltd, Princeton Plasma Physics Laboratory, Oak Ridge National Laboratory (ORNL), Idaho National Laboratory (INL), Savannah River National Laboratory (SRNL)
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.
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.
Tokamak Energy Ltd, Princeton Plasma Physics Laboratory, Columbia University, Oak Ridge National Laboratory, Marathon Fusion
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.
Alessandra Tonel, Lidia Piron, Paolo Zanca, David Terranova, Ephrem Delabie, Krassimir K. K Kirov, Morten Lennholm, Sheena Menmuir, Piero Martin
Acciaierie Venete SpA, University of Padova, UKAEA, CCFE, Oak Ridge
The interplay between externally induced locked modes and the plasma toroidal rotation is studied in JET L-mode plasmas, through the analysis of shots in which the non-disruptive compass scan technique [Piron L. et al. Nuclear Fusion 64 (2024) 066029] was performed. In these experiments, an external magnetic field perturbation was applied up to the onset of a locked mode, with toroidal mode number n = 1, at constant plasma density. In Ohmic plasmas, during the execution of the compass scan, it is observed that the toroidal rotation brakes in a wide range of the core. In the case of Neutral Beam Injection (NBI)-heated shots, the rotation braking is significantly reduced. The experimental analyses have been interpreted by the RFXlocking code adapted to the JET tokamak [Zanca P. et al. Nuclear Fusion 55 (2015) 043020], which also allowed to investigate the roles of the electromagnetic force and of the neoclassical toroidal viscosity, in the rotation braking phenomenon. Plasma rotation has been analyzed also during a density variation in the absence of external magnetic field perturbations, highlighting that a profile transition occurs in the core.
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