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This study examines the mechanisms underlying sawtooth suppression in the High Beta Tokamak-Extended Pulse (HBT-EP) device. It is observed that strong-intensity sawtooth activities correlate with reduced-amplitude magnetohydrodynamics (MHD) edge modes which are identified as $m/n = 3/1$ external kink modes, while sawtooth suppression correlates with larger and saturated edge mode amplitudes. To further investigate these correlations, the plasma–wall coupling was manipulated by adjusting the positions of the conducting walls in HBT-EP. It was found that strong sawtooth events occur when the normalized wall radius $b/a$ is within a critical value. This implies that the plasma–wall distance must be sufficiently small to ensure effective stabilization of the edge mode. Even slight differences in major radius result in significantly different discharge styles, categorized as ‘sawtoothing discharges’ and ‘sawtooth-suppressed discharges’ respectively. Through a series of mode structure analyses, we confirm the coexistence and coupling of the $m/n = 1/1$ helical core, $m/n = 2/1$ tearing mode, and $m/n = 3/1$ external kink mode during sawtooth-suppression, and that this coupling induces anomalous current broadening. Based on these findings, we conclude that sawtooth suppression in the HBT-EP tokamak is consistent with the process of magnetic flux pumping.

A recently developed reduced model of H-mode sustainment based on interchange-drift-Alfvén turbulence description in the vicinity of the separatrix matching experimental observations in ASDEX Upgrade has been extended to experiments with the unfavorable $\nabla B$ drift. The combination with the theory of the magnetic-shear-induced Reynolds stress offers a possibility to quantitatively explain the phenomena. The extension of the Reynolds stress estimate in the reduced model via the magnetic shear contribution is able to reproduce the strong asymmetry in the access conditions depending on the ion $\nabla B$ drift orientation in agreement with experimental observations. The Reynolds stress profile asymmetry predicted by the magnetic shear model is further extended by comparison with GRILLIX and GENE-X simulations matched with comparable experiments in realistic X-point geometry. The predictions of the radial electric field well depth and its difference between the favorable and unfavorable configurations at the same heating power from the extended model also show consistency with experimental measurements.

The comprehensive visual tool TRVIEW is presented, a software to read and fit experimental measurements and equilibrium reconstruction in the frame of nuclear fusion research. At the same time, TRVIEW provides a framework to setup physics workflows, especially those related to heating and transport. The TRVIEW package was originally created in order to provide file input to transport codes such as ASTRA and TRANSP. However, having a comprehensive set of python classes for the experimental plasma parameters, it allows direct coupling to several modules, such as the heating and current-drive codes RABBIT and TORBEAM and the equilibrium codes FEQIS and NEMEC. The direct coupling allows also direct visualisation of the input and output parameters and profiles in TRVIEW. The option of storing equilibrium files in the standard gEQDSK format is also enabled in TRVIEW. Special algorithms have been developed or adapted for TRVIEW: a class for fitting the plasma separatrix contour with Fourier moments, a recursive procedure for spline fits of the experimental kinetic profiles. Finally, TRVIEW includes the option to read, visualise and write integrated modelling and analysis system (IMAS) files with a comprehensive set of integrated data structures, covering most of the discharge and plasma parameters and several source profiles, calculated with the TORBEAM and RABBIT codes. This enables users to setup workflows for code validation and physics investigations using the world-wide established IMAS environment. TRVIEW can take the experimental input from the shotfile system of the ASDEX Upgrade tokamak or from IMAS files. Since the algorithms and the coupled codes are device-agnostic, TRVIEW is portable to any other fusion devices.

To address the needs for a fusion pilot plan design, DIII-D/EAST joint experiments on DIII-D have demonstrated high normalized beta β _N ∼ 4.2, toroidal beta β _T ∼ 3.3% with q _min > 2, q _95 ⩽ 8 sustained for more than six energy confinement times in high poloidal beta regime. The excellent energy confinement quality ( H _98y2 ∼ 1.8) is achieved with an internal transport barrier at high line-averaged Greenwald density fraction f _Gr > 0.9. The trapped gyro-Landau fluid (TGLF) modeling of the transport characteristics shows that the beam-driven rotation does not play an important role in the high confinement quality. The modeling also captures very well several transport features, giving us confidence in using integrated modeling to project these experimental results to future machines. The high-performance phase is terminated by fast-growing modes triggered near the n = 1 ideal-wall kink stability limit. New radio frequency (RF) capabilities for off-axis current drive could remove the residual ohmic current to achieve a fully non-inductive state, and improve the mode–wall coupling to increase the ideal-wall β _N limit, enabling sustainment of the fully non-inductive high performance plasma in stationary conditions.

Using the assumption of a weak normalized turbulent viscosity, usually valid in practice, the modulated-transport model (Stoltzfus-Dueck 2012 Phys. Plasmas 19 055908) is generalized to allow the turbulent transport coefficient to vary in an arbitrary way on radial and poloidal position. The new approach clarifies the physical interpretation of the earlier results and significantly simplifies the calculation, via a boundary-layer asymptotic method. Rigorous detailed appendices verify the result of the simple boundary-layer calculation, also demonstrating that it achieves the claimed order of accuracy and providing a concrete prediction for the strong plasma flows in the immediate vicinity of the last closed flux surface. The new formulas are used to predict plasma rotation at the core-edge boundary, in cases with and without externally applied torque. Dimensional formulas and extensive discussion are provided, to support experimental application of the new model.

The current response of a hot magnetized plasma to a radio-frequency wave is non-local, turning the electromagnetic wave equation into an integro-differential equation. Non-local physics gives rise to wave physics and absorption processes not observed in local media. Furthermore, non-local physics alters wave propagation and absorption properties of the plasma. In this work, an iterative method that accounts for parallel non-local effects in 2D axisymmetric tokamak plasmas is developed, implemented, and verified. The iterative method is based on the finite element method and Fourier decomposition, with the advantage that this numerical scheme can describe non-local effects while using a high-fidelity antenna and wall representation, as well as limiting memory usage. The proposed method is implemented in the existing full wave solver FEMIC and applied to a minority heating scenario in ITER to quantify how parallel non-local physics affect wave propagation and dissipation in the ion cyclotron range of frequencies (ICRF). The effects are then compared to a reduced local plane wave model, both verifying the physics implemented in the model, as well as estimating how well a local plane wave approximation performs in scenarios with high single pass damping. Finally, the new version of FEMIC is benchmarked against the ICRF code TORIC.

On the J-TEXT tokamak, the dynamics of edge magnetic topology during the opening of the edge magnetic islands induced by the external Resonant Magnetic Perturbation (RMP) are investigated. The edge island chain is pushed outward by increasing plasma toroidal current to intersect the poloidally and toroidally localized divertor plate, forming an open magnetic island in the Scrape-Off Layer (SOL). The location of the strike points on the divertor plate predicted by the HINT code is in agreement with the experimental observations. The influence of the magnetic topology on edge plasma profiles has also been investigated using Langmuir probes. The properties of edge ${T_{\text{e}}}$ , ${n_{\text{e}}}$ , ${P_{\text{e}}}$ and ${E_{\text{r}}}$ profiles are studied as function of three magnetic structures in ${q_{\text{a}}}$ -dependence experiments and two magnetic structures in RMP configuration-dependence experiments. Some common features are observed. Inside the edge closed and SOL remnant islands, flat ${P_{\text{e}}}$ but non-flat ${T_{\text{e}}}$ and ${n_{\text{e}}}$ profiles are detected. When transitioning from the edge closed island to the partially open island with remnant island, the local flat ${P_{\text{e}}}$ profile is shifted outward accompanied by a narrower flattened ${P_{\text{e}}}$ region. The steep slopes of ${T_{\text{e}}}$ , ${n_{\text{e}}}$ and ${P_{\text{e}}}$ are measured in the SOL regions characterized by a long connection length ${L_{\text{c}}}$ , and the longer ${L_{\text{c}}}$ , the steeper slopes. ${E_{\text{r}}}$ exhibits a negative well inside the remnant island with infinite ${L_{\text{c}}}$ and in the SOL regions where ${L_{\text{c}}}$ significantly exceeds than the electron mean free path ${\lambda _{\text{e}}}$ , but develops a positive value in the SOL regions where ${L_{\text{c}}}$ is relatively short.

The magnetic separatrix surface is designed to provide the final and critical confinement to the hot stationary-operation core plasma in modern tokamak reactors in the absence of an external magnetic perturbation (MP) or transient magneto-hydrodynamic perturbation, while diverting the exhaust heat to divertor plates. All the stationary operational boundary plasma studies and reactor designs have been performed under this assumption. However, there has been a long-standing suspicion that a stationary-operation tokamak plasma even without external MPs or edge localized modes (ELMs) activities may not have a stable closed separatrix surface, especially near the magnetic X-point. Here, the first gyrokinetic numerical observation is reported that the divertor separatrix surface, due to homoclinic tangles caused by intrinsic electromagnetic turbulence, is not a stable closed surface in a stationary operation phase even without MPs or ELMs. Unlike the MP- or ELM-driven homoclinic tangles that could cause deleterious effects to core confinement or divertor plates, it is found that the micro-turbulence driven homoclinic tangles could connect the divertor plasma to the pedestal plasma in a constructive way by broadening the divertor heat-exhaust footprint and weakening the pedestal slope to the ELM-safe direction. Micro-turbulent homoclinic tangles can open a new research direction in understanding and controlling these two most troublesome and non-locally connected edge-plasma issues in a tokamak fusion reactor.

Extended-MHD modeling of planned Runaway Electron Mitigation Coils (REMC) for SPARC and DIII-D is performed with the NIMROD code. A coil has been designed for each machine, with the two differing in shape and location, but both having n = 1 symmetry (with n the toroidal mode number). Compared to previous modeling efforts, three improvements are made to the simulations boundary conditions. First a resistive wall model is used in place of an ideal wall. Second, the ThinCurr code is used to compute the time-dependent 3D fields used as magnetic boundary conditions for the simulations. Third, the simulation boundary is moved from the first-wall location to the Vacuum Vessel (VV), which extends the boundary past the location of the internal REMC. To remove the 3D coil from the simulation domain, an equivalent set of 3D fields is calculated at the VV boundary that produce approximately the same field distribution at the last closed flux surface assuming vacuum between the two. Each of these three boundary condition improvements leads to an improvement in the predicted performance of the REMC for both machines. The resistive wall alone primarily effects the resonance of the coil with the plasma after the TQ, affecting the q-profile evolution in the SPARC modeling, and allowing the applied spectrum to be modified in response to the plasma in the DIII-D modeling. The movement of the simulation boundary has the most significant effect on the RE confinement overall, including in the early stages, particularly for a DIII-D inner wall limited equilibrium, where the RE loss fraction increases from 90% to $\gt$ 99%, with SPARC RE losses also occurring much earlier when the boundary is placed at the VV.

To sustain good lower hybrid wave (LHW)–plasma coupling for long-pulse plasma operation, for the first time, coupling feedback control is designed and realized in EAST using a proportion integration differentiation method by choosing the reflection coefficient (RC) of LHW power as the reference for gas-puffing feedback, and including one pulse test and multi-pulse experiments. Experiments show that such feedback control can work correctly and maintains good LHW–plasma coupling effectively for a long time, suggesting the possibility of feedback control application on LHW–plasma coupling in long-pulse plasma. Furthermore, during the feedback control process of multi-pulse supersonic molecular beam injection (SMBI), the stored energy changes from 29 kJ to 58 kJ, and the energy confinement factor ( H _89 ) increases from 0.98 to 1.45, implying a positive effect of coupling feedback on plasma performance. Experiments between SMBI puffing and the gas puffing system, fed by a piezoelectric valve near the antenna, are further investigated, showing that the response time of the RC with SMBI is faster than that by the piezoelectric valve. In addition, SMBI puffing on the electron-drift side of the LHW antenna is a little quicker than that on the ion-drift side. Studies suggest that such feedback control is effective for long-pulse LHW–plasma coupling, and the gas puffing by SMBI on the electron-drift side of the LHW antenna could offer an effective way to sustain good LHW coupling in steady-state operation in the future. Further optimization will be continued at a later date.

The effects of poloidally inhomogeneous heat sources are investigated through a gyrokinetic formula in collisionless toroidal plasmas. A gyrokinetic dispersion relation is newly derived under the assumption that equilibrium parallel heat flows are generated to remove the injected poloidally nonuniform heat source. The dispersion relation is numerically solved, considering both inboard and outboard heat source injections. In the case of the inboard source injection, both Stringer spin-up and geodesic acoustic mode (GAM) are excited. Conversely, outboard injection leads to the emergence of a heat source-driven GAM (referred to as Q-GAM), featuring a frequency around half that of the standard GAM. Various physical quantities of the Q-GAM, such as mode frequency and source threshold, are analyzed through parametric scans. The Q-GAM exhibits similarities with the energetic-particle-driven GAM (EGAM), particularly in its frequency range, and both belong to one of the strong Landau damped poles. Despite having distinct driving mechanisms and structural differences in parallel velocity and poloidal coordinates, the response function of the perturbed parallel pressure to the potential, mainly contributing to the destabilization of each mode around half of the GAM frequency, is derived to have a similar form for both the Q-GAM and EGAM cases.

The objective of this study is twofold: firstly, to demonstrate the consistency between the anomalous transport results produced by updated Multi-Mode Model (MMM) version 9.0.4 and those obtained through gyrokinetic simulations; and secondly, to showcase MMM’s ability to predict electron and ion temperature profiles in low aspect ratio, high beta NSTX discharges. MMM encompasses a range of transport mechanisms driven by electron and ion temperature gradients, trapped electrons, kinetic ballooning, peeling, microtearing, and drift resistive inertial ballooning modes. These modes within MMM are being verified through corresponding gyrokinetic results. The modes that potentially contribute to ion thermal transport are stable in MMM, aligning with both experimental data and findings from linear CGYRO simulations. The isotope effects on these modes are also studied and higher mass is found to be stabilizing, consistent with the experimental trend. The electron thermal power across the flux surface is computed within MMM and compared to experimental measurements and nonlinear CGYRO simulation results. Specifically, the electron temperature gradient modes (ETGM) within MMM account for 2.0 MW of thermal power, consistent with experimental findings. It is noteworthy that the ETGM model requires approximately 5.0 ms of computation time on a standard desktop, while nonlinear CGYRO simulations necessitate 8.0 h on 8 K cores. MMM proves to be highly computationally efficient, a crucial attribute for various applications, including real-time control, tokamak scenario optimization, and uncertainty quantification of experimental data.

The ADITYA upgrade (ADITYA-U), a medium-sized $\left( {{R_0} = 75{\text{ cm}},\,\,a = 25{\text{ cm}}} \right)$ conventional tokamak facility in India, has been consistently producing experiments findings by using circular and shaped-plasmas. Recognizing the plasma parameters aligning closely with the design parameters of circular limited plasmas, ADITYA-U shifted its focus toward exploring the operational regime for experimentation on saw-tooth and MHD phenomena. Moreover, ADITYA-U has made consistent advancements toward conducting preliminary plasma shaping experiments through the activation of top and bottom divertor coils utilizing hydrogen as well as deuterium fuels. Confinement is improved by a factor of ∼1.5 in ${D_2}$ plasmas when compared to H _2 plasmas of ADITYA-U. Further, ADITYA-U operations emphasize preventing disruptions and runaway electrons (REs) to ensure safe operations for future fusion devices. Significant suppression of REs has been achieved in ADITYA-U with the application of pulsed localized vertical magnetic field (LVF) perturbation, thereby establishing the technique’s independence from the tokamak device. The successful RE mitigation requires a critical threshold of LVF pulse magnitude, which is approximately 1% of the toroidal magnetic field, and a minimum duration of ∼5 ms. Apart from this, several novel findings have been achieved in the ADITYA-U experiments, including the modification of sawtooth duration through gas-puff, the emergence of MHD-induced geodesic acoustic mode-like oscillations, the propagation of fast heat pulses induced by MHD activity, the control of RE dynamics through Gas-puffs, the propagation of pinch-driven cold-pulses, the transport and core accumulations of argon impurities, the mass dependency of plasma toroidal rotation and the detection of ‘RICE’ scaling, as well as the characterization of edge plasma using wall conditioning methods, such as glow discharge cleaning using a combination of Ar -H _2 mixture, localized wall cleaning by electron cyclotron resonant plasma, and the development of machine learning-based disruption predictions, will be discussed in this paper.

A baseline scenario of deuterium–tritium (D–T) plasma with peripheral high-field-side fuelling pellets has been produced in JET in order to mimic the situation in ITER. The isotope mix ratio is controlled in order to target the value of 50%–50% by a combination of tritium gas puffing and deuterium pellet injection. Multiple factors controlling the fuelling efficiency of individual pellets are analysed, with the following findings: (1) prompt particle losses due to pellet-triggered edge-localised modes (ELMs) are detected, (2) the plasmoid drift velocity might be smaller than that predicted by simulation, (3) post-pellet particle loss is controlled by transient phases with ELMs.The overall pellet particle flux normalised to the heat flux is similar to that in previous pellet fuelling experiments in AUG and JET.

The RFX-mod2 installation is planned to be completed by 2024 and the start of operations is expected in 2025. The high flexibility of the machine (already tested in the previous RFX-mod experiment) allows operation in Reversed Field Pinch and tokamak configuration as well as ultra-low q pulses. In this work we present predictive analysis on transport, performances and plasma control in RFX-mod2 in view of the first experimental campaigns.

Experimental studies of the dynamics of shear flow and turbulence spreading at the edge of tokamak plasmas are reported. Scans of line-averaged density and plasma current are carried out while approaching the Greenwald density limit on the J-TEXT tokamak. In all scans, when the Greenwald fraction ${f_{\text{G}}} = {{\bar n} \mathord{\left/ \right. } {{n_{\text{G}}}}} = {{\bar n} \mathord{\left/ \right. } {\left( {{{{I_{\text{p}}}} \mathord{\left/ \right. } {\pi {a^2}}}} \right)}}$ increases, a common feature of enhanced turbulence spreading and edge cooling is found. The result suggests that turbulence spreading is a good indicator of edge cooling, indeed better than turbulent particle transport is. The normalized turbulence spreading power increases significantly when the normalized ${\boldsymbol{E}} \times {\boldsymbol{B}}$ shearing rate decreases. This indicates that turbulence spreading becomes prominent when the shearing rate is weaker than the turbulence scattering rate. The asymmetry between positive/negative (blobs/holes) spreading events, turbulence spreading power and shear flow are discussed. These results elucidate the important effects of interaction between shear flow and turbulence spreading on plasma edge cooling.

Abnormal (deviating from the target) variations in the plasma vertical position Z and current $I_\textrm{p}$ (such as vertical displacements, transient $I_\textrm{p}$ ‘spikes’ and quenches) constitute common elements of a disruption, a phenomenon that is to be mitigated, or ultimately avoided in future reactor-relevant tokamaks. While these abnormalities are generally recognized cross-shot and cross-device, details in terms of appearance (or not) and order of these abnormalities in disruption event chains (DEC) are bound to the plasma state at the time of the chain initiation. Detection of these abnormalities is thus indicative not only of the onset of the plasma collapse itself but also of the disruption driving cause that is promoted at a particular plasma state. Here, the occurrence of disruptions, explored via the detection of an $I_\textrm{p}$ quench, and the analysis of DEC constituted by $I_\textrm{p}$ and Z abnormalities is reported for in total seven full device-year pairs of operation of three machines (4, 2 and 1 years of KSTAR, MAST-U and NSTX-U operation, respectively) using the DECAF code expanded tools and capabilities. It is shown that the disruption occurrence depends not only on the details of the plasma state but also on (device-dependent) technical elements of the shot exit scenario. Year-to-year changes in the main disruption causes and a reduction in the disruptivity rate, bound by device and operation upgrades, are reported. Particular trigger instances of DEC (and the full chains when applicable) are shown to occupy different parts of the operation space diagrams, in accordance with prior expectations. Plasma elongation is identified as an important factor influencing details of the chains and its role will be further explored.

The vertical motion and shrinking of the cold plasma column after a tokamak disruption leads to a natural decrease in the edge safety factor when most of the current is carried by runaway electrons (REs). Reaching a low edge safety factor can potentially cause a strong plasma instability. We present magnetohydrodynamic simulations of the termination of a post-disruption plateau-phase RE beam in ITER when the edge safety factor falls close to two. Growth of instabilities is observed to result in stochastization of the magnetic field and a prompt loss of REs. As RE impact must be mitigated in ITER, the effect of parameters that influence the final termination have been assessed. Higher background plasma resistivity is seen to cause larger mode magnitudes and stronger stochastization, leading to less remnant REs after the termination event. Lower ion-densities also project a qualitatively similar behavior although weaker in effect. Using computations from a wall collision model, the ensuing load distribution on the first-wall is also presented.

To achieve their goals, future thermonuclear reactors such as ITER and DEMO are expected to operate plasmas with a high magnetic field, triangularity and confinement. To address the corresponding challenges, the concept of the high-field ( $B_{\textrm{T}} \unicode{x2A7D}$ 5 T), high-current ( $I_{\textrm{P}} \unicode{x2A7D}$ 2 MA) COMPASS Upgrade tokamak was established, and the device is currently being constructed in Prague, Czech Republic. This contribution provides an overview of the priority physics topics for the future physics programme of COMPASS Upgrade, namely: (i) characterisation of alternative confinement modes, (ii) a power exhaust including liquid metals, (iii) operation with a hot first wall and (iv) the influence of plasma shape on pedestal stability and confinement. The main scenarios are presented, as predicted by METIS and FIESTA codes. Pedestal pressure and density are estimated using EPED, multi-machine semi-empirical scaling and a neutral penetration model. Access to detachment is estimated using a detachment qualifier.

Effects of the kinetic thermal ions (KTIs) on ideal infernal modes and resistive infernal modes have been investigated by using magnetohydrodynamic (MHD) simulation without KTIs and kinetic-MHD hybrid simulation with KTIs. For the ideal infernal modes, the pressure profile is significantly flattened at the saturated state for both the models with and without the KTIs. As the beta value decreases, the ideal infernal modes are stabilized while the resistive infernal modes are still unstable. For the resistive infernal modes, while the saturated pressure profile is significantly flattened in the MHD simulation without KTIs, the pressure profile is not flattened at the saturated state in the kinetic-MHD hybrid simulation with KTIs. The suppression of the saturation level by the effects of the KTIs results from the phase mismatch between the radial velocity and perturbed pressure mode structures. This indicates that KTIs play an essential role for the suppression of pressure profile flattening due to slowly growing resistive MHD instabilities.

Full remote scientific operation of the DIII-D National Fusion Facility is now possible through significant advances in the computer science hardware and software infrastructure made over the last decade. Capabilities around information visualization, data movement, and communication have all been enhanced. The level of capability deployed to remotely operate DIII-D required an infrastructure advancement over what had previously been achieved in the fusion community. The large quantity of real-time data that is automatically displayed on DIII-D’s control room screens can now be visualized by remote participants via web-based applications. New audio/video solutions using the VoIP and instant messaging application Discord have been implemented to mimic the dynamic and ad-hoc scientific conversations that are critical in successfully operating an experimental campaign. Discord’s ability for a user to rapidly move between audio channels, text with images, and share screens is a significant enhancement over traditional videoconferencing tools. In addition, multiple combinations of broadcast audio are made available via a web-based application to allow remote participants to simultaneously listen to general announcements/sounds while conducting their own specific conversations. Secure methodologies have been put into place to allow remote control of hardware including DIII-D’s plasma control system application. Secure methods also included the ability of the on-site team to closely coordinate their work with remote team members which has been enhanced through extensions to the wireless network and the use of tablet computers for audio/video/screen sharing. However, no amount of software can fully replace the need for ‘hands on hardware.’ This infrastructure was severely stress tested during the COVID-19 pandemic where occupancy of the DIII-D control room was restricted. Operational efficiency during the pandemic, measured in discharges per hour, remained high (3.8 ± 0.8) compared to values obtained pre-pandemic (3.7 ± 0.8).

The RFX-mod2 device, the upgraded version of the previous RFX-mod with a modified magnetic boundary, is presently under realization and will start to be operated in 2025. Significant upgrades of the diagnostic capabilities have been proposed and are under development. These include a largely increased number of in-vessel magnetic and electrostatic sensors, a new fast reciprocating manipulator for the exploration of the edge plasma in a wide range of experimental conditions, the improved Thomson scattering and soft x-ray diagnostics system for a detailed determination of the behavior of the electron temperature profile, new dedicated systems for the space and time resolved analysis of x-ray spectra and neutron rate, a reflectometric diagnostic for real-time determination of plasma position, two diagnostics devoted to the imaging of light impurities and influxes behavior along with arrays of halo current sensors. These diagnostic upgrades will be accompanied by a significant effort to improve the control of the electron density and of the impurity influxes by means of proper treatment of plasma facing components with in-vessel fixed electrodes distributed over the first wall. The described advancements will allow a deeper understanding of physics phenomena in the wide variety of magnetic configurations, including the tokamak, the reversed-field pinch and the Ultra-low q, which can be produced in RFX-mod2 thanks to its flexibility and unique MHD control capabilities.

The neural network model, MISHKA-NN is developed to mitigate the computational burden associated with the linear ideal magnetohydrodynamic (MHD) stability analysis of the pedestal based on the peeling–ballooning (P–B) model. By utilizing both 1D plasma profiles (current density, pressure gradient, and safety factor) and 0D parameters (plasma geometry, total current, and toroidal mode number), the model predicts linear growth rate of edge-localized ideal MHD instability in a given equilibrium state. By enabling the prediction of each instability within a second, the model reduces the time required for plotting a pedestal P–B stability diagram ( $\,j-\alpha$ diagram) from approximately 100 CPU hours to a few CPU minutes. Notably, even with the utilization of parametric pressure and current profiles and plasma boundary shapes for the training dataset, the model shows a satisfactory level of performance in benchmarking the $j-\alpha$ diagram for the reconstructed equilibrium from a KSTAR tokamak experiment. We anticipate the model to serve as a versatile alternative to 2D linear MHD stability codes, alleviating numerical costs.

Subcritical energetic electrons (SEEs) produced by the runaway electron (RE) avalanche source at energies below the runaway threshold are found to be the primary contributor to surface heating of plasma-facing components (PFCs) during final loss events. This finding is supported by theoretical analysis, computational modeling with the Kinetic Orbit Runaway electrons Code (KORC), and qualitative agreement with DIII-D experimental observations. The avalanche source generates significantly more secondary electrons below the runaway threshold, which thermalize rapidly when well-confined. However, during a final loss event, the RE beam impacts the first wall, and SEEs are deconfined before they can thermalize. Additionally, because the energy deposition length decreases faster than energy, the deposited energy density, and thus the maximum PFC surface temperature change, is larger for SEEs than REs. KORC simulations employ an analytic first wall to model particle deconfinement onto a non-axisymmetric wall composed of individual tiles. PFC surface heating is calculated using a 1D model extended to include an energy-dependent deposition length scale. Simulations of DIII-D qualitatively agree with infrared (IR) imaging only when SEEs from the avalanche source are included. These results demonstrate that SEEs are the dominant contributor to PFC surface heating and indicate that the avalanche source plays a critical role in the PFC damage caused during final loss events. The prominence of SEEs also has important implications for interpreting IR imaging, one of the primary diagnostics for RE-wall interaction diagnosis, despite REs dominating the energy and current density. This result improves predictions of wall damage due to post-disruption REs to estimate material lifetime and design RE mitigation systems for ITER and future reactors.

It was found previously that the negative triangularity (NT) configuration is more MHD-unstable for low n modes than the positive triangularity (PT) case, although the situation is reversed for intermediate n modes and the NT configuration becomes more stable for intermediate n modes ( $n = 3 - 10$ ) (Zheng et al 2021 Nucl. Fusion 61 116014). Here, n is the toroidal mode number. In this work, we extend the studies to include the rotation effects, as well as the diamagnetic drift effects, to see how the resistive wall modes (RWMs) in the NT configuration are affected as compared with the PT configuration. This is particularly motivated by noting that the wall interface with the plasma is quite different between the NT and PT configurations. It affects the plasma rotation and diamagnetic drift effects on the low n RWM. We consider the DIII-D-NT-experiment equilibrium reconstructed by the EFIT code. Based on the equilibrium g-file, the extended equilibria are constructed with the VMEC code by varying the beta values while keeping the pressure and poloidal current flux profiles basically unchanged. The bootstrap current contribution to the equilibria is taken into account with the Sauter formula. The MHD stability is then computed using the AEGIS code with the rotation and diamagnetic drift effects taken into account. We found that, although the NT configuration is less stable for n = 1 MHD modes, the rotation and diamagnetic drift stabilization effects on RWMs are more effective in the NT configuration than in the PT one. Note that even in the PT case, the stabilization of RWMs by the rotation and kinetic effects is critical. Because the low-n RWMs in the regular NT case are more unstable, the rotation and diamagnetic drift stabilization effects found in this research are important for the NT tokamak concept.

The local stability analysis of plasma waves driven by runaway electrons (REs) has been performed considering hot plasma Maxwellian background, with electron and ion temperatures of the order of 1 keV. It is shown that hot plasma waves, namely electron plasma waves (EPWs) and ion Bernstein waves (IBWs) can be driven unstable by RE at their coalescence frequency via Cherenkov resonance by RE with energy distribution peaked at about 8 MeV. A skew-normal distribution is used as a model of the RE energy distribution. The EPW and IBW couples of waves occur between any successive ion-cyclotron harmonics frequencies nf _ci , above the lower hybrid resonance. At their confluence, the perpendicular group velocity vanishes and significant RF emissions are expected. The frequency gap between two successive confluences is ∼ f _ci . Groups of RF line emissions, separated by almost constant frequency gap ∼ f _ci /2 are detected during various quiescent runaway plasma discharges in the FTU tokamak. The analysis of a specific discharge suggests that the frequencies of the line emissions observed and the frequencies occurring at the EPW-IBW confluences are in reasonable agreement. A possible explanation of the line emissions with ∼ f _ci /2 gap in terms of nonlinear mode coupling is proposed.

The experimental and theoretical analysis were focused on experiments conducted to assess the effect of plasma isotopes, protium (H), deuterium (D), and tritium (T) on ion cyclotron resonance heating (ICRH) related plasma wall interactions. Comparison of L-mode discharges with N = 1 ^3 He and N = 1 H minority ICRH heating scenarios were done for different isotopes. For the selected pulses, the behaviour of high-Z, mid-Z and low-Z intrinsic impurity and radiated power behaviour was investigated based on data from VUV, visible spectroscopy, and bolometry diagnostic at Joint European Torus. It was found that for N = 1 ^3 He scenario during radiofrequency antennas operation, core W, Ni content, Be source and the radiated power are higher for π /2 in comparison to dipole antenna phasing. Lowest core Ni, W content and radiated power is clearly observed for H plasmas in comparison to D and T, where for this ICRH scenario behaviour was similar. However, lower Be photon flux is observed for T in comparison to D plasmas. Be sputtering by He particles is responsible for such an effect. Additionally, several computer simulations were conducted using the COREDIV code. The difference in the electron temperature was due to the difference in the isotope masses. Increased temperature in the central plasma in the case of T plasmas leads to higher radiation in the central plasma in comparison to H plasmas. As a result, the power across separatrix is lower and the temperature on the divertor plate decreases with the increase of the isotope mass. At these temperatures on the divertor plate, W is not sputtered by the main plasma ions H, D and T and by He. For the N = 1 H ICRH scenario clear difference between D and T plasma was observed with higher metallic impurity content for T plasma in comparison to D. Impurity content in the plasmas is found to be sensitive to the power balance between the antenna straps. Its minimum is observed for the maximum of P _cen / P _tot .

This paper introduces an innovative method for reconstructing 2D magnetic flux contours and plasma parameters of dynamically moving tokamak plasmas. While conventional methods like EFIT, based on the Grad–Shafranov equation, are suitable for plasma equilibria with a single magnetic axis, our approach utilizes the MHD equations and shows promise for tokamak plasmas in motion or containing multiple magnetic axes, which may not strictly adhere to plasma equilibria. By utilizing limited edge magnetic probe measurements, our developed model successfully reconstructs the time evolution of two merging plasma toroids in the TS-6 experiment. A comparison with direct 2D magnetic probe measurements in a low β regime reveals a reconstruction error of approximately 3%.

This study presents an analytical investigation of energetic particle-induced geodesic acoustic modes (EGAMs) within a gyro-kinetic model, incorporating finite-orbit-width (FOW) effects up to the second order. The inclusion of second-order FOW effects introduces two distinct types of energetic particle-wave resonances, occurring at $\omega = \omega_t^h$ and $\omega = 2\omega_t^h$ , respectively, where $\omega_t^h$ denotes the transit frequency of energetic particles (EPs). It is found that two unstable EGAM branches coexist: a low frequency branch (LFB) characterized by $0 \lt \omega_\textrm{LFB} \lt \omega_{t,\textrm{max}}^h$ , and a high frequency branch (HFB) marked by $\omega_{t,\textrm{max}}^h \lt \omega_\textrm{HFB} \lt 2\omega_{t,\textrm{max}}^h$ . The instability of LFB primarily arises from the resonance $\omega = \omega_t^h$ , mainly introduced by first-order FOW effects. As a result, the instability of LFB always exists regardless of the presence or absence of second-order FOW effects, and is barely modified by these effects. In contrast, the instability of HFB is exclusively attributed to the resonance $\omega = 2\omega_t^h$ induced by second-order FOW effects. Consequently, the HFB exhibits instability in the presence of these effects.

In the JET DTE2 campaign a new method was successfully tested to detect the heating of bulk electrons by α-particles, using the dynamic response of the electron temperature T _e to the modulation of ion cyclotron resonance heating (ICRH). A fundamental deuterium (D) ICRH scheme was applied to a tritium-rich hybrid plasma with D-neutral beam injection (NBI). The modulation of the ion temperature T _i and of the ICRH accelerated deuterons leads to modulated α -heating with a large delay with respect to other modulated electron heating terms. A significant phase delay of ∼40° is measured between central T _e and T _i , which can only be explained by α -particle heating. Integrated modelling using different models for ICRH absorption and ICRH/NBI interaction reproduces the effect qualitatively. Best agreement with experiment is obtained with the European Transport Solver/Heating and Current Drive workflow.

DIII-D experiments demonstrate that high pressure, broad current profile equilibria can be accessed in the high poloidal beta regime by optimizing the MHD mode control poloidal spectrum. A novel, variable spectrum (VS) magnetic feedback scheme implemented using the DIII-D internal non-axisymmetric coils (I-coils) facilitated access to reduced internal inductance $\ell_\mathrm{i}$ operation above the no-wall beta limit compared with both no feedback and fixed spectrum feedback. In addition, the VS feedback helped avoid beta collapses caused by marginally unstable resistive wall mode activity. The lower and upper I-coil rows were configured in two independent feedback loops, allowing the feedback field’s poloidal spectrum to vary and track changes in the plasma mode structure as the edge safety factor q _95 varied from 11 to 6 during the discharges. The q _95 dependence of the measured phase difference between the lower and upper I-coil rows during VS feedback is qualitatively compatible with ideal MHD simulations of the least-stable plasma kink mode and with plasma response simulations that included kinetic modifications to ideal MHD. The VS feedback approach is a straightforward way to improve resilience to variations in mode structure that occur as plasma parameters change. The demonstrated expansion of the operating space to lower $\ell_\mathrm{i}$ is expected to improve the coupling of the plasma kink mode to external fields and beneficial wall eddy currents, and is compatible with high bootstrap fraction operation.

A long standing shortfall in the predicted L-mode edge energy transport by reduced quasi-linear models of gyrokinetic turbulent transport has been resolved. The improved model TGLF-SAT2 has higher fidelity to gyrokinetic simulations of the electron-scale contribution to the electron energy transport and the ion-scale flux surface shape dependence of energy transport. The success of TGLF-SAT2 in predicting the L-mode and Ohmic edge profiles is critical to whole pulse simulation and opens the door to prediction of the H-mode power threshold.

The effect of impurity on the electrostatic microturbulence in ADITYA-U tokamak is assessed using global gyrokinetic simulations. The realistic geometry and experimental profiles of the ADITYA-U are used, before and after argon gas seeding, to perform the simulations. Before the impurity seeding, the simulations show the existence of the trapped electron mode (TEM) instability in three distinct regions on the radial-poloidal plane. The mode is identified by its linear eigenmode structure and its characteristic propagation in the electron diamagnetic direction. The simulations with Ar ^1+ impurity ions in the outer-core region show a significant reduction in the turbulence and transport due to a reduction in the linear instability drive, with respect to the case without impurity. A decrease in particle and heat transport in the outer-core region modifies the plasma density profile measured after the impurity seeding. It, thus, results in the stabilization of the TEM instability in the core region. Due to the reduced turbulence activity, the electron and ion temperatures in the central region increase by about 10%.

Effective control of impurities and precise regulation of the fueling gas are supreme prerequisites for optimal operation in any fusion device. Conventional wall-conditioning methods fall short of achieving optimal wall conditioning. Conventional wall-conditioning methods, such as vessel baking and H _2 /(D _2 )-fueled glow discharge cleaning (GDC), are generally required to remove wall-absorbed impurities in bulk after vessel venting. The excess amount of hydrogen, injected during H _2 GDC, can be reduced by helium (He)-fueled GDC. However, He removal from the vessel is more challenging due to its low molecular mass, very low condensation temperature, and inert characteristics. In ADITYA-U, optimal wall conditioning cannot be achieved using H _2 followed by He-fueled GDC when applied for extended periods spanning hours or days. A GDC with a mixture of argon and hydrogen (Ar–H _2 ) is introduced in the ADITYA-U tokamak to obtain better wall conditioning than H _2 followed by He GDC. In Ar–H _2 GDC, long-lived ArH ^+ ions are formed in sufficient numbers and accelerated toward the vessel wall with high momentum. This results in the breaking of high energy bonds of impurities with the wall/plasma facing components, which is not possible by H ^+ , H _2 ^+, H _3 ^+ ions in H _2 GDC due to their lower momentum. An optimal blend ratio of Ar to H _2 is established at 15%–20% for the mixture. This composition ensures that the introduction of high- Z Ar does not adversely affect tokamak plasma operations. The C- and O-containing impurities are reduced beyond the limit of the prolonged operation of H _2 GDC. Relative low pressures of dominant impurities such as CO, CH _4 , and H _2 O are obtained due to the Ar–H _2 GDC compared to routinely operated H _2 GDC. A comparison study of H _2 GDC and the developed Ar–H _2 GDC is performed in terms of wall conditioning and tokamak plasma operation. The encouraging results of the Ar–H _2 GDC are obtained in both wall cleaning and tokamak operation scenarios in the midsize tokamak ADITYA-U. This development and application of Ar–H _2 GDC are beneficial for large-sized fusion devices, leading to improved impurity reduction, reduced operational fuel consumption (H _2 /D _2 /He), and enhanced control over fuel recycling/extraction.

Strongly-shaped diverted negative triangularity (NT) plasmas in the DIII-D tokamak demonstrate simultaneous access to high normalized density, current, pressure, 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 key parameters, expanded operating spaces featuring 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 observed, even simultaneously, and 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.

The operational space and global performance of plasmas with edge-localized modes (ELMs) suppressed by resonant magnetic perturbations (RMPs) are surveyed by comparing AUG, DIII-D, EAST, and KSTAR stationary operating points. RMP-ELM suppression is achieved over a range of plasma currents, toroidal fields, and RMP toroidal mode numbers. Consistent operational windows in edge safety factor are found across devices, while windows in plasma shaping parameters are distinct. Accessed pedestal parameters reveal a quantitatively similar pedestal-top density limit for RMP-ELM suppression in all devices of just over $3\times 10^{19}$ m ^−3 . This is surprising given the wide variance of many engineering parameters and edge collisionalities, and poses a challenge to extrapolation of the regime. Wide ranges in input power, confinement time, and stored energy are observed, with the achieved triple product found to scale like the product of current, field, and radius. Observed energy confinement scaling with engineering parameters for RMP-ELM suppressed plasmas are presented and compared with expectations from established H and L-mode scalings, including treatment of uncertainty analysis. Different scaling exponents for individual engineering parameters are found as compared to the established scalings. However, extrapolation to next-step tokamaks ITER and SPARC find overall consistency within uncertainties with the established scalings, finding no obvious performance penalty when extrapolating from the assembled multi-device RMP-ELM suppressed database. Overall this work identifies common physics for RMP-ELM suppression and highlights the need to pursue this no-ELM regime at higher magnetic field and different plasma physical size.

The recent Federici et al (2024 Nucl. Fusion 64 036025) article makes the argument that higher magnetic fields cannot reduce the size and cost of a tokamak-based fusion power plant due to: structural considerations of the toroidal field (TF) coils, the required thickness for neutron shielding and the blanket, and challenges with heat exhaust in the divertor. This conclusion is based on a series of assumptions that are design decisions made by the authors, not fundamental limits on physics or engineering. This Comment demonstrates that the conclusions of Federici et al are invalid if one makes different design choices and that its results are therefore not broadly generalizable.

Shattered pellet injection (SPI) is a promising method for controlling plasma disruptions in tokamaks. In this study, we present numerical modelling of the fragmentation of cryogenic deuterium pellets within the context of SPI, using the peridynamic (PD) theory. A dedicated in-house code has been developed, leveraging the meshfree method and GPU parallelization. The mechanical properties of cryogenic solid deuterium are obtained from available literature, and calibrated based on the shatter threshold along with the remaining solid mass fraction after shatter. The results from the bond-based PD successfully reproduce the main experimental results reported in the literature, both qualitatively and quantitatively.

In a future fusion reactor, the main fuelling method will likely rely on the injection of solid hydrogen pellets. Current predictions assume that this goal can be achieved, since being based on a technology which is already largely developed. However, this belief is founded on modelling tools that are usually aligned to the observation made in existing devices and then extrapolated to reactor conditions. This approach needs a sound consideration of its intrinsic restrictions and any observed feature not reproduced by the utilised codes should be applied to check their validation and possibly contribute to their refinement. One specific feature still lacking an explanation of a reasonable and self-consistent mechanism in the current models is the appearance of a phenomenon called striations, which are high frequency variations in the radiation emitted during the pellet ablation process. In order to provide a sound and reliable database for further considerations, a dedicated analysis of this effect has been performed on the mid-size tokamak ASDEX Upgrade. Therefore, such cases have been selected with the relevant signal recorded with sufficient temporal resolution during experiments covering a wide variation of plasma and pellet parameters which are regarded to be potentially influential on the striation pattern. In addition, it was ensured that for any specific case the observed behaviour was reproducible for several individual ablation events under identical conditions. In all cases considered, the observed radiation-intensity variations appear with a typical pattern showing a broad peak of frequencies in the range 50 – 150 kHz. This characteristic unveils a notable resilience against any parameter variation. This new collection of data can now act as firm basis to corroborate future modelling code-validation efforts. In addition, the analysis method can provide a relatively simple way of reviewing future modelling predictions.

We report an experimental examination of non-axisymmetric (3D) magnetic fields for the control of Alfvén eigenmodes (AEs) in KSTAR. Application of the phase-sweeping n = 1 3D magnetic field identifies the effective 3D field phase and threshold amplitude for suppression of toroidal AEs. Such observations indicate that at least two conditions on the 3D field phase and amplitude should be satisfied for the AE suppression. The phase window of AE suppression is largely resonant and thereby overlapped with that of mode locking, while the threshold of mode locking is slightly higher than that of AE suppression, which implies a narrow 3D configuration window for AE suppression. Numerical analyses on the AE stability and fast ion phase-space transport suggest that the key mechanism of the AE suppression is the reduction of the AE drive through redistribution of fast ion phase-space distribution by strong resonant interactions of the fast ions with the 3D magnetic field.

Plasma differential rotation is found to be capable of preventing disruptive neoclassical tearing modes (NTMs) seeded by nonlinear three-wave coupling. As tearing modes degrade confinement and can lead to disruptions, stabilization strategies are crucial to the successful operation of future devices. In ITER-relevant scenarios on DIII-D, rotationally coupled $m/n$ = 1/1 and 3/2 modes have been observed to drive 2/1 islands through three-wave coupling. The frequency of the driven 2/1 mode is set by matching conditions and the frequencies of the driving modes. When the driven mode frequency matches the local plasma rotation frequency, e.g. at low differential rotation, the driven 2/1 island can grow into a disruptive NTM. Using neutral beam torque as an actuator to scan the differential rotation, these experiments demonstrate that a sufficiently large frequency mismatch prevents destabilization of disruptive 2/1 NTMs by three-wave coupling. This work indicates that differential rotation can be used as an actuator to prevent NTMs seeded by three-wave coupling.

The heat flux width ( λ _q ) is a key parameter determining the heat load at divertor targets. In recent years, drifts have been found to play a remarkable role in the edge plasma transport, while its influence on λ _q has not been well understood. Investigations of the influence of drifts on λ _q , systematic simulations using the SOLPS-ITER code are performed in this work. The statistics of the simulation results show that the drift under favorable/unfavorable B _T tends to increase the λ _q in the outer/inner side and decrease the λ _q in the other side, which is consistent with the experiment observations. At the upstream and the target, the mechanisms of the influence of the drifts on λ _q are different. The upstream λ _q ( λ _q _,u ) is directly affected by the drift-induced convective heat flux, while λ _q at the target ( λ _q _,t ) is indirectly influenced through heat conduction (in the high-recycling regime) and the sheath (in the detached regime) due to the change of plasma parameters there. Furthermore, the synergetic effect of geometry and drift under favorable B _T leads to an anomalously large λ _q _,t in the inner side at high density.

Since the first plasma realized in 2020, a series of key systems on HL-3 (known as HL-2M before) tokamak have been equipped/upgraded, including in-vessel components (the first wall, lower divertor, and toroidal cryogenic/water-cooling/baking/glow discharge systems, etc.), auxiliary heating system of 11 MW, and 28 diagnostic systems (to measure the plasma density, electron temperature, radiation, magnetic field, etc.). Magnet field systems were commissioned firstly for divertor plasma discharges. During the 2nd experimental campaign of HL-3 tokamak, several great progresses have been achieved. Firstly, the successful operation with plasma current larger than 1 MA was achieved under a divertor configuration. Secondly, the advanced divertor concept with two distinct snowflake configurations was realized. It is found that the distribution of ion saturation current and heat flux on bottom plate becomes wide due to magnetic surface expansion, demonstrating the advantage of such configuration in the heat flux mitigation. In addition, using the combination of NBI, ECRH and LHCD, the standard sawtoothing high confinement mode of megampere plasma was firstly accessed on the HL-3. The successful commissioning of HL-3 is beneficial for the initial operation of ITER.

This work presents the first analysis of the disruptive locked mode (LM) triggered by the dynamics of a confinement change. It shows that, under certain conditions, the LM threshold during the transient is significantly lower than expected from steady states. We investigate the sensitivity to a controlled n = 1 error field (EF) activated prior to the L-H transition in the COMPASS tokamak, at q _95 ∼ 3, β _N ∼ 1, and using EF coils on the high-field side of the vessel. A threshold for EF penetration subsequent to the L-H transition is identified, which shows no significant trend with density or applied torque, and is an apparent consequence of the reduced intrinsic rotation of the 2/1 mode during this transient phase. This finding challenges the assumption made in theoretical and empirical works that natural mode rotation can be predicted by global plasma parameters and urges against using any parametric EF penetration scaling derived from steady-state experiments to define the EF correction strategy in the entire discharge. Furthermore, even at EFs below the identified penetration threshold, disruptive locking of sawtooth-seeded 2/1 tearing modes is observed after about 30% of L-H transitions without external torque.

This paper deals with resistive wall tearing mode (RWTM) disruptions. RWTMs are closely related to resistive wall modes. RWTMs are tearing modes whose linear and nonlinear behavior is strongly dependent on the resistive wall outside the plasma. The consequence for ITER, is that the thermal quench timescale could be much longer than previously conjectured. Active feedback stabilization is another possible way to mitigate or prevent RWTM disruptions. Simulations of RWTM disruptions are reviewed for DIII-D and MST. MST has a longer resistive wall time than ITER, and disruptions are not observed experimentally when MST is operated as a standard tokamak. Simulations indicate that the RWTM disruption time scale is longer than the experimental shot time. Edge cooling causes contraction of the current profile, which can destabilize RWTMs. The equilibria studied here have the q = 2 rational surface close to the edge of the plasma, and low current density between the q = 2 surface and the wall. A sequence of low edge current model equilibria has major disruptions only for a resistive, not ideal, wall, and edge $q \unicode{x2A7D} 3.4.$ This is consistent with regimes of tokamak disruptivity, suggesting that tokamak disruptions caused by edge cooling at low edge q could be RWTMs.

To achieve optimal operation of the polarimetry-based FOCS, the light polarisation state at the input of the sensing fibre part must be close to a linear one. In the case of a FOCS deployed on a tokamak, the Joint European Torus (JET) in the present work, the long fibre optics link between the laser source and the sensing fibre modifies the polarisation in an unpredictable way, making it unclear which source polarisation state is to be set. A method for performing the necessary polarisation adjustment in a systematic way is proposed based on the FOCS analysis. The method requires performing data acquisition at two different input polarisations. Based on these measurements, the optimal laser source polarisation can be found. The method was experimentally verified using laboratory set-up and then successfully demonstrated with the FOCS installed at JET.

The article discusses the development of steady-state burning plasma in magnetic confinement fusion energy. It highlights the progress made in increasing the fusion triple product on tokamak devices, but notes that larger devices have not been built since the International Thermonuclear Experimental Reactor (ITER). However, advancements in physics and technology on tokamaks, such as the development of high-temperature superconductors, have shown promise for achieving breakthroughs in fusion energy. The article also discusses the potential benefits and challenges of strong magnetic fields in facilitating access to the burning plasma regime. Overall, the article suggests that the fusion field is entering a new stage of DT fusion with new opportunities and challenges. [Extracted from the article]

A quasilinear equation which allows describing evolution of the electron distribution function and generation of non-inductive currents by helicons is obtained. It is shown that in the analyzed case the Fokker–Planck equation can be approximated by a one-dimensional equation in the longitudinal electron velocity space with a diffusion coefficient proportional to the helicon power absorbed by electrons due to Landau damping. [ABSTRACT FROM AUTHOR]