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We discuss the MHD stability of the H-mode pedestal region in two edge localized mode (ELM)-suppressed H-mode regimes on DIII-D, the quiescent (Q)H-mode and resonant magnetic perturbation (RMP) H-mode. The QH-mode is obtained at low density with most of the neutral beam power injected counter to the plasma current, and is characterized by a continuous, low toroidal mode number (n), edge harmonic oscillation (EHO). QH-mode is observed to transition back to ELMing H-mode as the rotational shear is decreased by increasing the co-current neutral beam power fraction or by increasing the plasma to conducting wall distance, consistent with a model for the EHO as a peeling instability with added drive from rotational shear, which saturates through self transport of momentum or damping of rotation through drag of the mode on the conducting wall. In RMP H-mode, an n = 3 coil provides the non-axisymmetric perturbation. Two regimes of RMP ELM suppression are observed. At low collisionality with large resonant field perturbation, the pedestal pressure gradient and current density are reduced below the peeling-ballooning mode stability limit; while at higher collisionality with large non-resonant perturbation there is little change in the pedestal parameters and type II ELMs are enhanced.

In order to inform core performance projections and divertor design, the baseline SPARC tokamak plasma discharge is evaluated for its expected H-mode access, pedestal pressure and edge-localized mode (ELM) characteristics. A clear window for H-mode access is predicted for full field DT plasmas, with the available 25 MW of design auxiliary power. Additional alpha heating is likely needed for H-mode sustainment. Pressure pedestal predictions in the developed H-mode are surveyed using the EPED model. The projected SPARC pedestal would be limited dominantly by peeling modes and may achieve pressures in excess of 0.3 MPa at a density of approximately 3 × 1020 m−3. High pedestal pressure is partially enabled by strong equilibrium shaping, which has been increased as part of recent design iterations. Edge-localized modes (ELMs) with >1 MJ of energy are projected, and approaches for reducing the ELM size, and thus the peak energy fluence to divertor surfaces, are under consideration. The high pedestal predicted for SPARC provides ample margin to satisfy its high fusion gain (Q) mission, so that even if ELM mitigation techniques result in a 2× reduction of the pedestal pressure, Q > 2 is still predicted.

The SPARC tokamak is a critical next step towards commercial fusion energy. SPARC is designed as a high-field ($B_0 = 12.2$ T), compact ($R_0 = 1.85$ m, $a = 0.57$ m), superconducting, D-T tokamak with the goal of producing fusion gain $Q>2$ from a magnetically confined fusion plasma for the first time. Currently under design, SPARC will continue the high-field path of the Alcator series of tokamaks, utilizing new magnets based on rare earth barium copper oxide high-temperature superconductors to achieve high performance in a compact device. The goal of $Q>2$ is achievable with conservative physics assumptions ($H_{98,y2} = 0.7$) and, with the nominal assumption of $H_{98,y2} = 1$, SPARC is projected to attain $Q \\approx 11$ and $P_{\\textrm {fusion}} \\approx 140$ MW. SPARC will therefore constitute a unique platform for burning plasma physics research with high density ($\\langle n_{e} \\rangle \\approx 3 \\times 10^{20}\\ \\textrm {m}^{-3}$), high temperature ($\\langle T_e \\rangle \\approx 7$ keV) and high power density ($P_{\\textrm {fusion}}/V_{\\textrm {plasma}} \\approx 7\\ \\textrm {MW}\\,\\textrm {m}^{-3}$) relevant to fusion power plants. SPARC's place in the path to commercial fusion energy, its parameters and the current status of SPARC design work are presented. This work also describes the basis for global performance projections and summarizes some of the physics analysis that is presented in greater detail in the companion articles of this collection.

In order to inform core performance projections and divertor design, the baseline SPARC tokamak plasma discharge is evaluated for its expected H-mode access, pedestal pressure and edge-localized mode (ELM) characteristics. A clear window for H-mode access is predicted for full field DT plasmas, with the available 25 MW of design auxiliary power. Additional alpha heating is likely needed for H-mode sustainment. Pressure pedestal predictions in the developed H-mode are surveyed using the EPED model. The projected SPARC pedestal would be limited dominantly by peeling modes and may achieve pressures in excess of 0.3 MPa at a density of approximately 3 × 1020 m-3. High pedestal pressure is partially enabled by strong equilibrium shaping, which has been increased as part of recent design iterations. Edge-localized modes (ELMs) with >1 MJ of energy are projected, and approaches for reducing the ELM size, and thus the peak energy fluence to divertor surfaces, are under consideration. The high pedestal predicted for SPARC provides ample margin to satisfy its high fusion gain (Q) mission, so that even if ELM mitigation techniques result in a 2x reduction of the pedestal pressure, Q > 2 is still predicted.

The extended-MHD NIMROD code [C.R. Sovinec and J.R. King, J. Comput. Phys. 229, 5803 (2010)] is verified against the ideal-MHD ELITE code [H.R. Wilson et al. Phys. Plasmas 9, 1277 (2002)] on a diverted tokamak discharge. When the NIMROD model complexity is increased incrementally, resistive and first-order finite-Larmour radius effects are destabilizing and stabilizing, respectively. The full result is compared to local analytic calculations which are found to overpredict both the resistive destabilization and drift stabilization in comparison to the NIMROD computations.

We report on the characterization of five third chromosome mutations with strong effects on the formation of the eggshell or chorion. Three mutations, defining two loci, result in substantially reduced follicle cell-specific amplification of the major chorion structural genes and, hence, in underproduction of the corresponding mRNAs and proteins. The other two mutations, though displaying structural chorion abnormalities, appear to have no significant effect on amplification and to express normally the major chorion structural genes. The possible nature of these mutations is discussed.

Extended-MHD modeling of DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] quiescent H-mode (QH-mode) discharges with nonlinear NIMROD [C. R. Sovinec et al., J. Comput. Phys. 195, 355 (2004)] simulations saturates into a turbulent state but does not saturate when the steady-state flow inferred from measurements is not included. This is consistent with the experimental observations of the quiescent regime on DIII-D. The simulation with flow develops into a saturated turbulent state where the n=1 and 2 toroidal modes become dominant through an inverse cascade. Each mode in the range of n=1-5 is dominant at a different time. Consistent with experimental observations during QH-mode, the simulated state leads to large particle transport relative to the thermal transport. Analysis shows that the amplitude and phase of the density and temperature perturbations differ resulting in greater fluctuation-induced convective particle transport relative to the convective thermal transport. Comparison to magnetic-coil measurements shows that rotation frequencies differ between the simulation and experiment, which indicates that more sophisticated extended-MHD two-fluid modeling is required.

The extended-MHD NIMROD code [C.R. Sovinec and J.R. King, J. Comput. Phys. 229, 5803 (2010)] models broadband-MHD activity from a reconstruction of a quiescent H-mode shot on the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)]. Computations with the reconstructed toroidal and poloidal ion flows exhibit low-n perturbations (n=1-5) that grow and saturate into a turbulent-like MHD state. The workflow used to project the reconstructed state onto the NIMROD basis functions re-solves the Grad-Shafranov equation and extrapolates profiles to include scrape-off-layer currents. Evaluation of the transport from the turbulent-like MHD state leads to a relaxation of the density and temperature profiles.

A new modular code called BOUT++ is presented, which simulates 3D fluid equations in curvilinear coordinates. Although aimed at simulating Edge Localised Modes (ELMs) in tokamak X-point geometry, the code is able to simulate a wide range of fluid models (magnetised and unmagnetised) involving an arbitrary number of scalar and vector fields, in a wide range of geometries. Time evolution is fully implicit, and 3rd-order WENO schemes are implemented. Benchmarks are presented for linear and non-linear problems (the Orszag-Tang vortex) showing good agreement. Performance of the code is tested by scaling with problem size and processor number, showing efficient scaling to thousands of processors. Linear initial-value simulations of ELMs using reduced ideal MHD are presented, and the results compared to the ELITE linear MHD eigenvalue code. The resulting mode-structures and growth-rate are found to be in good agreement (BOUT++ = 0.245, ELITE = 0.239). To our knowledge, this is the first time dissipationless, initial-value simulations of ELMs have been successfully demonstrated.

The STEP (Stability, Transport, Equilibrium, and Pedestal) integrated-modeling tool has been developed in OMFIT to predict stable, tokamak equilibria self-consistently with core-transport and pedestal calculations. STEP couples theory-based codes to integrate a variety of physics, including MHD stability, transport, equilibrium, pedestal formation, and current-drive, heating, and fueling. The input/output of each code is interfaced with a centralized ITER-IMAS data structure, allowing codes to be run in any order and enabling open-loop, feedback, and optimization workflows. This paradigm simplifies the integration of new codes, making STEP highly extensible. STEP has been verified against a published benchmark of six different integrated models. Core-pedestal calculations with STEP have been successfully validated against individual DIII-D H-mode discharges and across more than 500 discharges of the \\(H_{98,y2}\\) database, with a mean error in confinement time from experiment less than 19%. STEP has also reproduced results in less conventional DIII-D scenarios, including negative-central-shear and negative-triangularity plasmas. Predictive STEP modeling has been used to assess performance in several tokamak reactors. Simulations of a high-field, large-aspect-ratio reactor show significantly lower fusion power than predicted by a zero-dimensional study, demonstrating the limitations of scaling-law extrapolations. STEP predictions have found promising EXCITE scenarios, including a high-pressure, 80%-bootstrap-fraction plasma. ITER modeling with STEP has shown that pellet fueling enhances fusion gain in both the baseline and advanced-inductive scenarios. Finally, STEP predictions for the SPARC baseline scenario are in good agreement with published results from the physics basis.