19 Search Results

The most efficient and promising operational regime for the International Thermonuclear Experimental Reactor tokamak is the high-confinement mode. In this regime, however, periodic relaxations of the plasma edge can occur. These edge-localized modes pose a threat to the integrity of the fusion device. Here we reveal the strong impact of energetic ions on the spatio-temporal structure of edge-localized modes in tokamaks using nonlinear hybrid kinetic–magnetohydrodynamic simulations. A resonant interaction between the fast ions at the plasma edge and the electromagnetic perturbations from the edge-localized mode leads to an energy and momentum exchange. Energetic ions modify, for example, the amplitude, frequency spectrum and crash timing of edge-localized modes. The simulations reproduce some observations that feature abrupt and large edge-localized mode crashes. The results indicate that, in the International Thermonuclear Experimental Reactor, a strong interaction between the fusion-born alpha particles and ions from neutral beam injection, a main heating and fast particle source, is expected with predicted edge-localized mode perturbations. This work advances the understanding of the physics underlying edge-localized mode crashes in the presence of energetic particles and highlights the importance of including energetic ion kinetic effects in the optimization of edge-localized mode control techniques and regimes that are free of such modes.

Abstract 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 $\delta \approx -0.3$ 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.

Building on prior analysis of ASDEX Upgrade (AUG) experiments (Henderson et al 2023 Nucl. Fusion63 086024), this study compares simple analytical formula predictions for divertor detachment onset and reattachment timescales in JET experiments. Detachment onset primarily scales with divertor neutral pressure, impurity concentration, power directed to the targets, machine size, and integral perpendicular power decay length. JET experiments, focusing on seeding mixtures of Ne and Ar, align with the detachment onset predictions. Radiation efficiencies among the impurities show good agreement with the model predictions, contrasting with AUG observations which suggested higher efficiency for Ar and lower efficiency for Ne. The time taken to re-ionise the neutral volume in front of the outer target in fully detached divertor conditions was measured following both abrupt increases in injected neutral beam power and, separately, cutting of the impurity gas flow. Re-ionisation of the neutrals occurs within approximately 1 s on JET, which aligns with the simple model prediction derived from AUG data. While the AUG results are not new, their comparison with the JET results enhances understanding, reinforcing confidence in using simple models to predict future reactor scenarios.

In 2021 JET exploited its unique capabilities to operate with T and D–T fuel with an ITER-like Be/W wall (JET-ILW). This second major JET D–T campaign (DTE2), after DTE1 in 1997, represented the culmination of a series of JET enhancements—new fusion diagnostics, new T injection capabilities, refurbishment of the T plant, increased auxiliary heating, in-vessel calibration of 14 MeV neutron yield monitors—as well as significant advances in plasma theory and modelling in the fusion community. DTE2 was complemented by a sequence of isotope physics campaigns encompassing operation in pure tritium at high T-NBI power. Carefully conducted for safe operation with tritium, the new T and D–T experiments used 1 kg of T (vs 100 g in DTE1), yielding the most fusion reactor relevant D–T plasmas to date and expanding our understanding of isotopes and D–T mixture physics. Furthermore, since the JET T and DTE2 campaigns occurred almost 25 years after the last major D–T tokamak experiment, it was also a strategic goal of the European fusion programme to refresh operational experience of a nuclear tokamak to prepare staff for ITER operation. The key physics results of the JET T and DTE2 experiments, carried out within the EUROfusion JET1 work package, are reported in this paper. Progress in the technological exploitation of JET D–T operations, development and validation of nuclear codes, neutronic tools and techniques for ITER operations carried out by EUROfusion (started within the Horizon 2020 Framework Programme and continuing under the Horizon Europe FP) are reported in (Litaudon et al Nucl. Fusion accepted), while JET experience on T and D–T operations is presented in (King et al Nucl. Fusion submitted).

We present the design and operation of a suite of Gas Puff Imaging (GPI) diagnostic systems installed on the Tokamak à Configuration Variable (TCV) for the study of turbulence in the plasma edge and Scrape-Off-Layer (SOL). These systems provide the unique ability to simultaneously collect poloidal 2D images of plasma dynamics at the outboard midplane, around the X-point, in both the High-Field Side (HFS) and Low-Field Side (LFS) SOL, and in the divertor region. We describe and characterize an innovative control system for deuterium and helium gas injection, which is becoming the default standard for the other gas injections at TCV. Extensive pre-design studies and the different detection systems are presented, including an array of avalanche photodiodes and a high-speed CMOS camera. First results with spatial and time resolutions of up to [Formula: see text] mm and 0.5  µs, respectively, are described, and future upgrades of the GPI diagnostics for TCV are discussed.

In situ, two-dimensional (2D) Langmuir probe measurements across a large part of the TCV outer divertor are reported in L-mode discharges with and without divertor baffles. This provides detailed insights into time averaged profiles, particle fluxes, and fluctuation behavior in different divertor regimes. The presence of the baffles is shown to substantially increase the divertor neutral pressure for a given upstream density and to facilitate the access to detachment, an effect that increases with plasma current. The detailed, 2D probe measurements allow for a divertor particle balance, including ion flux contributions from parallel flows and E × B drifts. The poloidal flux contribution from the latter is often comparable or even larger than the former, and the divertor parallel flow direction reverses in some conditions, pointing away from the target. In most conditions, the integrated particle flux at the outer target can be predominantly ascribed to ionization along the outer divertor leg, consistent with a closed-box approximation of the divertor. The exception is a strongly detached divertor, achieved here only with baffles, where the total poloidal ion flux even decreases towards the outer target, indicative of significant plasma recombination. The most striking observation from relative density fluctuation measurements along the outer divertor leg is the transition from poloidally uniform fluctuation levels in attached conditions to fluctuations strongly peaking near the X-point when approaching detachment.

Multi-spectral imaging of helium atomic emission (HeMSI) has been used to create 2D poloidal maps of $$Τ$$_e and $$\mathcal{n}$$_e in TCV's divertor. To achieve these measurements, TCV's MANTIS multispectral cameras (Perek et al 2019 Rev. Sci. Instrum.90 123514) simultaneously imaged four He I lines (two singlet and two triplet) and a He II line (468 nm) from passively present He and He⁺. The images, which were absolutely calibrated and covered the whole divertor region, were inverted through the assumption of toroidal symmetry to create emissivity profiles and, consequently, line-ratio profiles. A collisional-radiative model (CRM) was applied to the line-ratio profiles to produce 2D poloidal maps of $$Τ$$_e and $$\mathcal{n}$$_e. The collisional-radiative modeling was accomplished with the Goto helium CRM code (Zholobenko et al 2018 Nucl. Fusion58 126006, Zholobenko et al 2018 Technical Report, Goto 2003 J. Quant. Spectrosc. Radiat. Transfer76 331–44) which accounts for electron-impact excitation (EIE) and deexcitation, and electron–ion recombination (EIR) with $$H$$e⁺. The HeMSI $$Τ$$_e and $$\mathcal{n}$$_e measurements were compared with co-local Thomson scattering measurements. The two sets of measurements exhibited good agreement for ionizing plasmas: ($5 eV$ ≤ $$T$$_e ≤ $60eV$, and $$2$$ X $10$¹⁸$$m$$^-3 ≤ $$\mathcal{n}$$_e ≤ $$3$$ X $10$¹⁹$$m$$^-3) in the case of majority helium plasmas, and ($10 eV$ ≤ $$T$$_e ≤ $40eV$, $$2$$ X $10$¹⁸$$m$$^-3 ≤ $$\mathcal{n}$$_e ≤ $$3$$ X $10$¹⁹$$m$$^-3) in the case of majority deuterium plasmas. However, there were instances where HeMSI measurements diverged from Thomson scattering. When $$T$$_e ≤ $10 eV$ in majority deuterium plasmas, HeMSI deduced inaccurately high values of $$Τ$$_e. This disagreement cannot be rectified within the CRM's EIE and EIR framework. Second, on sporadic occasions within the private flux region, HeMSI produced erroneously high measurements of $$\mathcal{n}$$_e. Multi-spectral imaging of Helium emission has been demonstrated to produce accurate 2D poloidal maps of $$Τ$$_e and $$\mathcal{n}$$_e within the divertor of a tokamak for plasma conditions relevant to contemporary divertor studies.

Radial electric fields up to ~4 kV m^–1 are observed in the boundary between the private flux region (PFR) and the scrape-off layer (SOL) driving E × B drifts between the inner and outer targets at speeds up to 2.8 km s^–1 in the Tokamak à configuration variable divertor. The resulting E × B fluxes, located in a narrow region ($$\Delta {\rho _\Psi } < 0.012$$ in normalized radius or $$\Delta $$^R – R_sep <4 mm mapped to the outer midplane) are equivalent to around 20% of the total heat and particle flux to the divertor targets (inner + outer). At the peak E_r, the E × B poloidal transport is equivalent to parallel flows with M_∥ ~ 3. In the snowflake divertor with a second X-point in the outer SOL, the drifts in the PFR-SOL boundary were equivalent to around 30% of the total heat and particle flux to the divertor targets and cover a region ~50% wider than in the single null ($$\Delta {\rho _\Psi }$$ ~ 0.018, $$\Delta $$R – R_sep ~ 6 mm). Furthermore, the location of the PFR-SOL boundary drift shifts radially in the E_∥ × B direction when reversing the toroidal field direction. Peaks in density and electron pressure have been identified near the primary X-point along with large gradients in density, temperature, and potential, the latter resulting in a local electric field ~2.7 kV m^–1 which drives a drift (1.9 km s^–1) upwards towards the closed flux surfaces. Floating potential (V_f) magnitudes up to 75 V (~2 kT_e) were measured, indicating that V_f and parallel currents should not be neglected when estimating plasma potential.

A new gas puff imaging diagnostic has been installed on the TCV tokamak, providing two-dimensional insights into scrape-off-layer (SOL) turbulence dynamics above, at and below the magnetic X-point. A detailed study in L-mode, attached, lower single-null discharges shows that statistical properties have little poloidal variations, while vast differences are present in the 2D behaviour of intermittent filaments. Strongly elongated filaments, just above the X-point and in the divertor far-SOL, show a good consistency in shape and dynamics with field-line tracing from filaments at the outboard midplane, highlighting their connection. In the near-SOL of the outer divertor leg, short-lived, high frequency and more circular (diameter ~15 sound Larmor radii) filaments are observed. These divertor-localised filaments appear born radially at the position of maximum density and display a radially outward motion with velocity ≈400 m s^-1 that is comparable to radial velocities of upstream-connected filaments. Conversely, in these discharges ($$\vec{B} × ∇B$$ pointing away from the divertor), these divertor filaments’ poloidal velocities differ strongly from those of upstream-connected filaments. The importance of divertor-localised filaments upon radial transport and profile broadening is explored using filament statistics and in situ kinetic profile measurements along the divertor leg. This suggests that these filaments contribute significantly to electron density profile broadening in the divertor.

One of our grand challenges towards fusion energy is the achievement of a high-performance plasma core coupled to a boundary solution. The high confinement mode (H-mode) provides such a high-performance fusion core due to the build-up of an edge transport barrier leading to a pedestal. However, it usually features type-I edge localized modes (ELMs) which pose a threat for long-duration plasma operation in future fusion devices as they induce large energy fluences onto the plasma facing components and typically are projected to damage the first wall. For future fusion devices, the integration of a stationary no-ELM regime with a power exhaust solution is indispensable. Several no-ELM and small-ELM regimes have extended their operational space in the past years, with the ultimate goal of providing an alternative core–edge solution to ITER and EU-DEMO. Prominent no-ELM or small-ELM alternatives include the I-mode, QH-mode, EDA H-mode, quasi-continuous exhaust (QCE) and ‘grassy’ ELM regimes, X-point radiator scenarios and negative triangularity L-mode. The state-of-the-art, including access conditions and main signatures, of these alternative regimes is reviewed. Many of these regimes partly match the operational space of ITER and EU-DEMO, however, knowledge gaps remain. Besides compatibility with divertor detachment and a radiative mantle, these include extrapolations to high Q operations, low core collisionality, high Greenwald fractions, impurity transport, amongst others. The knowledge gaps and possible strategies to close these gaps to show their applicability to ITER and EU-DEMO are discussed.