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O. Sauter

Publications and source records attributed to O. Sauter.

16 recordsLinked to original sources

High-power TCV scenario for conventional and alternative divertor studies

Alternative divertor configurations (ADCs) must be evaluated under boundary plasma conditions approaching reactor-level values to be considered a reliable, physics-based solution for tokamak power exhaust. Most ADC experiments performed to date were at relatively low exhaust power. This work presents a high-power scenario on the TCV tokamak enabling the study of a wide variety of divertor magnetic shapes under an expanded SOL and power exhaust parameter space. The scenario is characterized by high power levels of electron cyclotron resonance heating ($2.5\,\text{MW}$ fully absorbed in a $\sim1\,\text{m}^{3}$ plasma) at high plasma current (edge safety factor $q_{95}\approx 2.5$), and low upstream separatrix densities ($n_{e,\text{u}}\approx1\times10^{19}\,\text{m}^{-3}$, Greenwald fraction $f_{\text{G}}\approx 0.1$). Stationary parallel heat fluxes up to $100\,\text{MW m}^{-2}$ are measured at the divertor target, an order of magnitude above previous TCV power exhaust studies. The obtained SOL collisionality and Lengyel detachment scaling metric lie within range of values expected in future reactors (SPARC, ITER, ARC).

physics.plasm-ph

The Negative Triangularity Tokamak Path for Fusion Pilot Plants: Experimental Progress and Future Prospects

This paper reviews the experimental progress of negative triangularity (NT), a tokamak configuration where the poloidal cross-section is a reversed-D shape compared to the conventional positive triangularity (PT) shape. NT is a promising reactor scenario that addresses the fundamental tension between performance, exhaust, and robustness. NT studies have accelerated globally across these three pillars over the past several years. While tokamak pilot plants are typically designed for the standard PT H-mode regime, this approach faces significant challenges in balancing high core performance with manageable heat and particle exhaust as well as reliable robustness. In contrast, NT plasmas have achieved H-mode-level confinement while remaining robustly free of the deleterious edge localized mode (ELM) instability. Regarding exhaust, NT offers a larger divertor wetted area on the outboard side and demonstrates compatibility with detachment and operation at high core radiation fraction without the constraints of the L-H power threshold, while also exhibiting low core impurity retention. NT operates with high reproducibility over a wide operating space, demonstrated by robust discharge-to-discharge consistency, and has access to plasmas with very high Greenwald fractions and/or low edge safety factors compared to PT H-mode plasmas. Further research is required to answer outstanding questions related to reactor confinement extrapolation, the optimal triangularity for a reactor, and core-edge integration. NT studies in existing and planned tokamaks are increasing, as is interest in possible reactor concepts. The unique physics and engineering advantages of NT offer a robust and simplified foundation for a viable fusion power plant.

physics.plasm-ph

The physics of ELM-free regimes in EUROfusion tokamaks

The development of operational scenarios without large Type-I ELMs is of utmost importance for the stable operation and longevity of future tokamaks. The EUROfusion tokamak exploitation program has therefore made the understanding of ELM-free regimes a major topic of exploration across all its contributing devices (ASDEX Upgrade, JET, MAST-Upgrade, TCV, and WEST). An integrated program to investigate a range of Type-I ELM-free regimes has been developed covering the enhanced D-alpha (EDA), magnetic perturbations (MP), negative triangularity (NT), quasi-continuous exhaust (QCE), quiescent H-mode (QH), the baseline small ELMs (SE), I-mode, and X-point radiator (XPR) regimes. This contribution focuses on the development and understanding of the NT and QCE regimes on ASDEX Upgrade, JET, and TCV. The importance of transport via ballooning modes in both regimes is highlighted, as well as the progress in developing access models based on ideal-MHD. In the case of the QCE, this can also be expressed as a minimum separatrix density, which corresponds well to experimentally measured separatrix densities. Particular focus is paid to the performance of the QCE in terms of the achieved pedestal top values, which, when appropriately normalised, do not differ significantly from ELMy H-mode plasmas. This, combined with the predicted minimum separatrix density for the 15 MA ITER baseline plasma, highlight the relevance of the QCE as a potential operational scenario for both ITER and future reactors.

physics.plasm-ph

Applications of a novel model-based real-time observer for electron density profile control experiments in TCV

Real-time control of tokamak plasmas encompasses sustaining a high-performance stationary state, avoiding disruptions, and managing ramp-up and ramp-down phases. Real-time estimation and control of electron density is fundamental for monitoring and controlling particle confinement, heating efficiency, exhaust conditions, impurity concentration, fusion power, and proximity to the density limit. Building on the integration of a multi-rate observer based on RAPDENS into the TCV control system, this study explores its application to density profile control for detachment studies, ECH, and NBH L-mode plasmas, and high-performance H-mode scenarios. TCV experiments demonstrate the observer's capability to support detachment studies in complex divertor geometries, controlling the line-averaged density within the last-closed flux surface while rejecting interferometer pick-up from Scrap-Off Layer density in the divertor. The estimated density profile enables local control of central density in ECH/NBH L-mode plasmas below cutoff conditions; heating-induced profile peaking modification is treated as a disturbance to the control task. Real-time estimation of time-varying transport coefficients, such as the pinch velocity-to-diffusivity ratio, improves model predictive capabilities, and the underlying turbulent transport is characterized via linear and non-linear gyrokinetic simulations with GENE. Simultaneous control of edge-normalized density and toroidal beta in H-mode plasmas is then demonstrated, yielding good confinement, scenario reproducibility, and a diagnostics-independent edge-density metric, while avoiding density limits and diagnostic faults propagation.

physics.plasm-ph

Non-dimensional confinement scaling in similar negative triangularity plasmas on the DIII-D and TCV tokamaks

Similarity experiments were performed on the DIII-D and TCV tokamaks to explore the scaling of energy confinement in negative triangularity plasmas using non-dimensional variables. Near up-down symmetric plasmas with large top-bottom averaged negative triangularity were created in a lower single null configuration, with the shape of the separatrix being closely matched between the two devices. The normalized energy confinement is found to weakly improve at increasing collisionality and, between the two devices, shows a machine size scaling behavior between Bohm and gyro-Bohm. Engineering scaling on a large DIII-D dataset is in agreement with the non-dimensional experiment.

physics.plasm-ph

Kinetic Equilibrium Prediction at TCV using RAPTOR and FBT

We present results from a new Kinetic-Equilibrium Prediction (KEP) workflow and shot preparation for full TCV discharges, by coupling predict-first RAPTOR transport simulations with FBT inverse equilibrium calculations. RAPTOR is a 1.5D transport code which has been extensively used for plasma shot optimization and real-time modeling. We show that rapid pre-shot simulations can be performed directly using information from the pulse schedule across a wide range of plasma shapes and scenarios, given an estimate of the confinement quality factor H98(y,2) and line-averaged density. The resulting p' and TT' profiles are then provided to the pre-shot equilibrium computation performed by FBT - a static free-boundary solver routinely used at TCV - achieving convergence between the two codes in a few iterations. Finally, we show that this coupling, when integrated into the TCV shot preparation, improves the evaluation of the coil currents needed to match the target plasma shape; in particular providing an accurate estimate of critical quantities such as the internal inductance $l_i$ and normalized pressure $\beta_N$, giving more realistic information to tokamak operators about the expected pulse behavior and enabling them to adjust the plan correspondingly.

physics.plasm-ph

Disentangling Core and Edge Mechanisms of the Density Limit in DIII-D Negative Triangularity Plasmas

The density limit is investigated in the DIII-D negative triangularity (NT) plasmas which lack a standard H-mode edge. We find the limit may not be a singular disruptive boundary but a multifaceted density saturation phenomenon governed by distinct core and edge transport mechanisms. Sustained, non-disruptive operation is achieved at densities up to 1.8 times the Greenwald limit ($n_\mathrm{G}$) until the termination of auxiliary heating. Systematic power scans reveal distinct power scalings for the core ($n_e \propto P_\mathrm{SOL}^{0.27\pm0.03}$) and edge ($n_e \propto P_\mathrm{SOL}^{0.42\pm0.04}$) density limits. The edge density saturation is triggered abruptly by the onset of a non-disruptive, high-field side radiative instability that clamps the edge density below $n_\mathrm{G}$. In contrast, the core density continues to rise until it saturates, a state characterized by substantially enhanced core turbulence. Core transport evolves from a diffusive to an intermittent, avalanche-like state, as indicated by heavy-tailed probability density functions (kurtosis $\approx 6$), elevated Hurst exponents, and a $1/f$-type power spectrum. These findings suggest that the density limit in the low-confinement regime is determined by a combination of edge radiative instabilities and core turbulent transport. This distinction provides separate targets for control strategies aimed at extending the operational space of future fusion devices.

physics.plasm-ph

Comparison of detachment in Ohmic plasmas with positive and negative triangularity

In recent years, negative triangularity (NT) has emerged as a potential high-confinement L-mode reactor solution. In this work, detachment is investigated using core density ramps in lower single null Ohmic L-mode plasmas across a wide range of upper, lower, and average triangularity (the mean of upper and lower triangularity: $δ$) in the TCV tokamak. It is universally found that detachment is more difficult to access for NT shaping. The outer divertor leg of discharges with $δ\approx -0.3$ could not be cooled to below $5~\mathrm{eV}$ through core density ramps alone. The behavior of the upstream plasma and geometrical divertor effects (e.g. a reduced connection length with negative lower triangularity) do not fully explain the challenges in detaching NT plasmas. Langmuir probe measurements of the target heat flux widths ($λ_q$) were constant to within 30% across an upper triangularity scan, while the spreading factor $S$ was lower by up to 50% for NT, indicating a generally lower integral Scrape-Off Layer width, $λ_{int}$. The line-averaged core density was typically higher for NT discharges for a given fuelling rate, possibly linked to higher particle confinement in NT. Conversely, the divertor neutral pressure and integrated particle fluxes to the targets were typically lower for the same line-averaged density, indicating that NT configurations may be closer to the sheath-limited regime than their PT counterparts, which may explain why NT is more challenging to detach.

physics.plasm-ph

Simultaneous access to high normalized current, pressure, density, and confinement in strongly-shaped diverted negative triangularity plasmas

Strongly-shaped diverted negative triangularity (NT) plasmas in the DIII-D tokamak demonstrate simultaneous access to high normalized current, pressure, density, and confinement. NT plasmas are shown to exist across an expansive parameter space compatible with high fusion power production, revealing surprisingly good core stability properties that compare favorably to conventional positive triangularity plasmas in DIII-D. Non-dimensionalizing the operating space, edge safety factors below 3, normalized betas above 3, Greenwald density fractions above 1, and high-confinement mode (H-mode) confinement qualities above 1 are simultaneously observed, all with a robustly stable edge free from deleterious edge-localized mode instabilities. Scaling of the confinement time with engineering parameters reveals at least a linear dependence on plasma current although with significant power degradation, both in excess of expected H-mode scalings. These results increase confidence that NT plasmas are a viable approach to realize fusion power and open directions for future detailed study.

physics.plasm-ph

Flexible, integrated modeling of tokamak stability, transport, equilibrium, and pedestal physics

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.

physics.plasm-ph

How accurate are flux-tube (local) gyrokinetic codes in modeling energetic particle effects on core turbulence?

Flux-tube gyrokinetic codes are widely used to simulate drift-wave turbulence in magnetic confinement devices. While a large number of studies show that flux-tube codes provide an excellent approximation for turbulent transport in medium-large devices, it still needs to be determined whether they are sufficient for modeling supra-thermal particle effects on core turbulence. This is called into question given the large temperature of energetic particles (EPs), which makes them hardly confined on a single flux-surface, but also due to the radially broad mode structure of energetic-particle-driven modes. The primary focus of this manuscript is to assess the range of validity of flux-tube codes in modeling fast ion effects by comparing radially global turbulence simulations with flux-tube results at different radial locations for realistic JET parameters using the gyrokinetic code GENE. To extend our study to a broad range of different plasma scenarios, this comparison is made for four different plasma regimes, which differ only by the profile of the ratio between the plasma kinetic and magnetic pressure. The latter is artificially rescaled to address the electrostatic limit and regimes with marginally stable, weakly unstable and strongly unstable fast ion modes. These energetic-particle-driven modes is identified as an AITG/KBAE via linear ORB5 and LIGKA simulations. It is found that the local flux-tube simulations can recover well the global results only in the electrostatic and marginally stable cases. When the AITG/KBAE becomes linearly unstable, the local approximation fails to correctly model the radially broad fast ion mode structure and the consequent global zonal patterns. According to this study, global turbulence simulations are likely required in regimes with linearly unstable AITG/KBAEs. In conditions with different fast ion-driven modes, these results might change.

physics.plasm-ph

First-principles density limit scaling in tokamaks based on edge turbulent transport and implications for ITER

A first-principles scaling law, based on turbulent transport considerations, and a multi-machine database of density limit discharges from the ASDEX Upgrade, JET and TCV tokamaks, show that the increase of the boundary turbulent transport with the plasma collisionality sets the maximum density achievable in tokamaks. This scaling law shows a strong dependence on the heating power, therefore predicting for ITER a significantly larger safety margin than the Greenwald empirical scaling (Greenwald et al, Nucl. Fusion, 28(12), 1988) in case of unintentional H-L transition.

physics.plasm-ph

Nonlocal effects in negative triangularity TCV plasmas

Global gradient driven GENE gyrokinetic simulations are used to investigate TCV plasmas with negative triangularity. Considering a limited L-mode plasma, corresponding to an experimental triangularity scan, numerical results are able to reproduce the actual transport level over a major fraction of the plasma minor radius for a plasma with $δ_{\rm LCFS}=-0.3$ and its equivalent with standard positive triangularity $δ$. For the same heat flux, a larger electron temperature gradient is sustained by $δ<0$, in turn resulting in an improved electron energy confinement. Consistently with the experiments, a reduction of the electron density fluctuations is also seen. Local flux-tube simulations are used to gauge the magnitude of nonlocal effects. Surprisingly, very little differences are found between local and global approaches for $δ>0$, while local results yield a strong overestimation of the heat fluxes when $δ<0$. Despite the high sensitivity of the turbulence level with respect to the input parameters, global effects appear to play a crucial role in the negative triangularity plasma and must be retained to reconcile simulations and experiments. Finally, a general stabilizing effect of negative triangularity, reducing fluxes and fluctuations by a factor dependent on the actual profiles, is recovered.

physics.plasm-ph

Integrated real-time supervisory management for off-normal-event handling and feedback control of tokamak plasmas

For long-pulse tokamaks, one of the main challenges in control strategy is to simultaneously reach multiple control objectives and to robustly handle in real-time (RT) unexpected events (off-normal-events -- ONEs) with a limited set of actuators. We have developed in our previous work a generic architecture of the plasma control system (PCS) including a supervisor and an actuator manager to deal with these issues. We present in this paper recent developments of real-time decision-making by the supervisor to switch between different control scenarios (normal, backup, shutdown, disruption mitigation, etc.) during the discharge, based on off-normal-event states. We first standardize the evaluation of ONEs and thereby simplify significantly the supervisor decision logic, as well as facilitate the modifications and extensions of ONE states in the future. The whole PCS has been implemented on the TCV tokamak, applied to disruption avoidance with density limit experiments, demonstrating the excellent capabilities of the new RT integrated strategy.

eess.SY

The impact of anisotropy on ITER scenarios

We report on the impact of anisotropy to tokamak plasma configuration and stability. Our focus is on analysis of the impact of anisotropy on ITER pre-fusion power operation 5~MA, $B=1.8$~T ICRH scenarios. To model ITER scenarios remapping tools are developed to distinguish the impact of pressure anisotropy from the change in magnetic geometry caused by an anisotropy-modified current profile. The remappings iterate the anisotropy-modified current profile to produce the same $q$ profile with matched thermal energy. The analysis is a step toward equilibria that are kinetically self-consistent for a prescribed scenario. We find characteristic detachment of flux surfaces from pressure surfaces, and an outboard (inboard) shift of peak density for $T_{\parallel}>T_\perp$ ( $T_{\parallel}<T_\perp$). Differences in the poloidal current profile are evident, albeit not as pronounced as for the spherical tokamak. We find that the incompressional continuum is largely unchanged in the presence of anisotropy, and the mode structure of gap modes is largely unchanged. The compressional branch however exhibits significant differences in the continuum. We report on the implication of these modifications.

physics.plasm-ph

Power requirements for electron cyclotron current drive and ion cyclotron resonance heating for sawtooth control in ITER

13MW of electron cyclotron current drive (ECCD) power deposited inside the q = 1 surface is likely to reduce the sawtooth period in ITER baseline scenario below the level empirically predicted to trigger neo-classical tearing modes (NTMs). However, since the ECCD control scheme is solely predicated upon changing the local magnetic shear, it is prudent to plan to use a complementary scheme which directly decreases the potential energy of the kink mode in order to reduce the sawtooth period. In the event that the natural sawtooth period is longer than expected, due to enhanced alpha particle stabilisation for instance, this ancillary sawtooth control can be provided from > 10MW of ion cyclotron resonance heating (ICRH) power with a resonance just inside the q = 1 surface. Both ECCD and ICRH control schemes would benefit greatly from active feedback of the deposition with respect to the rational surface. If the q = 1 surface can be maintained closer to the magnetic axis, the efficacy of ECCD and ICRH schemes significantly increases, the negative effect on the fusion gain is reduced, and off-axis negative-ion neutral beam injection (NNBI) can also be considered for sawtooth control. Consequently, schemes to reduce the q = 1 radius are highly desirable, such as early heating to delay the current penetration and, of course, active sawtooth destabilisation to mediate small frequent sawteeth and retain a small q = 1 radius.

physics.plasm-ph