Sophia Guizzo

This work presents the compact experimental negative triangularity (NT) reactor (CENTAUR), a low overnight cost, high-field tokamak, breakeven reactor design, achieving a predicted total fusion power of 40 MW 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 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 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 4 C 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 $2 billion US determined using a custom costing model that predicts a total overnight cost of 1.6 B ± 0.2 B US.

Utilizing the TokaMaker code, this manuscript presents a scheme for determining the optimal locations of poloidal field coils in tokamak configurations, aimed at reducing the currents required to reproduce target equilibrium in future devices. The definition of the coil position space is described, as well as the cost function is to be minimized, which incorporates the constraints of the problem. The MANTA reactor design, featuring strong negative triangularity shaping, is considered as a test example, where preliminary results are obtained with a reduced number of degrees of freedom in order to highlight the procedure for finding local minima. Application of the optimization procedure, guided by a study of the electric current regularization parameter, results in new suggestions for the placement of each poloidal field (PF) coil. This leads to a reduction in the total current flowing through the PF coils by 37% while accurately reproducing the target plasma boundary compared to a baseline case.

In liquid breeding blanket concepts for fusion reactors, including liquid metals and molten salts, the circulating breeder may interact directly with plasma dynamics through electromagnetic effects. As blanket designs mature and pilot-plant studies progress, simple models are needed to capture this interaction with sufficient fidelity. To address this need, we present a new axisymmetric simulation framework for coupled plasma-blanket systems. The blanket regions are treated with the time-dependent visco-resistive incompressible magneto-hydrodynamic (MHD) equations. For the plasma region, the ideal MHD equilibrium equation is solved at each timestep allowing the plasma to be simulated on the liquid blanket flow timescale. Flexible geometry definitions and support for anisotropic conductivity enable exploration of a wide range of candidate blanket configurations, including poloidal ducts and toroidally isolated blanket segments. We demonstrate the framework’s capability by simulating a vertical displacement event (VDE) within a toroidally conducting PbLi blanket, showing how the fluid treatment of the blanket alters predicted plasma evolution timescales and structural forces compared to static models. Simulations of both VDEs and midplane current quenches in toroidally conducting blankets produce peak liquid metal velocities greater than 1 m/s. These transient flows are on the same order as the steady-state background flows required for self-cooled concepts, and an order of magnitude greater than those in externally cooled designs. In toroidally insulating configurations and molten salt concepts, significantly less flow is induced, with peak fluid velocities of several cm/s. These findings underscore the importance of coupling blanket and plasma dynamics to accurately assess the viability of conducting liquid blanket designs and set engineering requirements.

A technique, volumetric power optimization, is presented for enhancing the power output of magnetic confinement fusion devices. Applied to a tokamak, this approach involves shifting the burning plasma region to a larger plasma volume while introducing minimal perturbations to the plasma boundary shape. This edge perturbation—squareness—is analogous to pinching and stretching the edge boundary. Stability calculations confirm that this edge alteration is compatible with maintaining plasma stability. This optimization method for optimizing fusion power output could improve the performance of magnetic confinement fusion power plants. Published by the American Physical Society 2025

A plasma scenario with negative triangularity (NT) shaping is achieved on MAST-U for the first time. While edge localized modes (ELMs) are eventually suppressed as the triangularity is decreased below δ ≲ − 0.06 , an extended period of H-mode operation with Type-III ELMs is sustained at less negative δ even through access to the second stability region for ideal ballooning modes is closed. This documents a qualitative difference from the ELM-free access conditions documented in NT scenarios on conventional aspect ratio machines. The electron temperature at the pedestal top drops across the transition to ELM-free operation, but a steady rise in core temperature as δ is decreased allows for similar normalized β in the ELM-free NT and H-mode positive triangularity shapes.

Negative triangularity (NT) tokamak configurations may be more susceptible to magneto-hydrodynamic instability, posing challenges for recent reactor designs centered around their favorable properties, such as improved confinement and operation free of edge-localized modes. In this work, we assess the vertical stability of plasmas with NT shaping and develop potential reactor solutions. When coupled with a conformal wall, NT equilibria are confirmed to be less vertically stable than equivalent positive triangularity (PT) configurations. Unlike PT, their vertical stability is degraded at higher poloidal beta. Furthermore, improvements in vertical stability at low aspect ratio do not translate to the NT geometry. NT equilibria are stabilized in PT vacuum vessels due to the increased proximity of the plasma and the wall on the outboard side, but this scenario is found to be undesirable due to reduced vertical gaps which give less spatial margin for control recovery. Instead, we demonstrate that informed positioning of passively conducting plates can lead to improved vertical stability in NT configurations on par with stability metrics expected in PT scenarios. An optimal setup for passive plates in highly elongated NT devices is presented, where plates on the outboard side of the device reduce vertical instability growth rates to 16% of their baseline value. For lower target elongations, integration of passive stabilizers with divertor concepts can lead to significant improvements in vertical stability. Plates on the inboard side of the device are also uniquely enabled in NT geometries, providing opportunity for spatial separation of vertical stability coils and passive stabilizers.