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Motivated by the losses of runaway electrons during disruptive events in present and future tokamaks like the ITER, we investigate the transport in mixed phase space Hamiltonian systems. We are interested in the regime where the sticky regions, remaining after the disintegration of invariant transport barriers, form layers in the phase space separated by partial barriers to transport. The example we use is the standard map. We show that the survival probability from the internal region can be thought of as the resultant of a radial passage from layer to layer. We show that both exponential and power-law decays are present at the same time, but with independent effective time-scales. [ABSTRACT FROM AUTHOR]

De-aliasing is an essential procedure for eliminating the aliasing error in nonlinear simulations, such as nonlinear gyrokinetic turbulence simulations. An ideal approach to de-aliasing in the periodic dimension is the Fourier truncation. Finite difference low-pass filtering applied in the non-periodic direction strongly dampens aliasing modes. At the same time, it induces numerical dissipation in the region of the physically realistic solution. It is shown analytically that the long-wave dissipation coefficient is proportional to the ( N p − 3) power of the wavenumber under desirable constraints satisfying the highest order of accuracy, where N p is the number of filter points. Numerical results after applying the optimized low-pass filtering to the nonlinear gyrokinetic turbulence simulation suggest that the nine-point format preserves intact mesoscopic zonal structures in tokamak plasma, and is therefore suitable for long-time nonlinear turbulence simulations. [ABSTRACT FROM AUTHOR]

Using the SPIDER code, three series of calculations of plasma equilibrium in a tokamak are performed—with low, medium, and high pressure. In each series, the cross section is purely elliptical with elongation varying from K = 1 (circular plasma) to K = 2.4. For these configurations, the integrals S 1 ÷ S 3 defining the right-hand sides of the virial relations in Lao et al. [Nucl. Fusion 25, 1421 (1985)] are calculated. Their dependences on plasma parameters, elongation K of magnetic surfaces, and radial derivatives of K and Shafranov shift Δ are analyzed. The accuracy of analytical expressions for S 1 ÷ S 3 proposed in Pustovitov [Phys. Plasmas 29, 092507 (2022)] is assessed. [ABSTRACT FROM AUTHOR]

Following Greene et al. [Phys. Fluids 14, 671 (1971)] and Connor et al. [Phys. Plasmas 31, 577 (1988); Plasma Phys. Control. Fusion 34, 161 (1992); and Nucl. Fusion 33, 1533 (1993)], the Grad-Shafranov equation for an axisymmetric tokamak plasma equilibrium is solved via an expansion in the, supposedly small, inverse aspect-ratio of the plasma, ϵ. The displacements of equilibrium magnetic flux-surfaces due to plasma shaping are assumed to be O (ϵ) smaller than the minor radii of the surfaces, but no other restriction is placed on the nature of the shaping. The solution of the Grad-Shafranov equation is matched to a vacuum solution that extends to infinity, and consists of an expansion in toroidal functions. The external poloidal magnetic field generated by a finite set of discrete external poloidal magnetic field-coils is calculated, and incorporated into the toroidal function expansion. In this manner, the shape of a large aspect-ratio tokamak plasma is directly related to the currents flowing in the external poloidal field-coils. Finally, a pedestal in the plasma pressure, and the associated spike in the bootstrap current, are incorporated into the model. [ABSTRACT FROM AUTHOR]

Experimental advanced superconducting tokamak (EAST) fast control power supply (FCPS) is an essential device to realize balance control of plasma vertical displacement during controllable nuclear fusion process. The primary control purpose of EAST FCPS is to quickly output current to track reference current. Change of load inductance parameters due to plasma motion poses higher robustness requirements for fast tracking control of output current. Therefore, an improved gray prediction parameter identification sliding mode control method is proposed to achieve fast tracking and robust control of output current under changes in load inductance parameters. Parameter identification sliding mode control method applied to identify load inductance parameters in real‐time, accurate EAST FCPS output current tracking sliding mode control model is established. High‐order terms are added to discrete exponential convergence reaching law, and a new smooth saturation function is designed to replace traditional sign function to achieve chattering suppression and accelerate system convergence speed. Gray prediction used in current sampling to achieve output current trajectory advanced prediction to further accelerate output current response speed. Four times equal interval time sampling within a fixed switching period and new information priority period by period prediction is proposed to improving gray prediction, achieving digital control delay compensation, and improving prediction accuracy of output current trajectory at abrupt edge. Simulation and experimental verification show that the proposed improved gray prediction parameter identification sliding mode control method has good output current tracking control performance under changes in load parameters. Compared with PI control method currently used in engineering, the proposed control method performs better in output current fast response and overshoot suppression. [ABSTRACT FROM AUTHOR]

Earlier, a conventional grill launcher-based LHCD system has been developed to drive plasma current non-inductively in tokamaks at IPR (ADITYA and SST1) at a frequency (fo = ωo/2π) of 3.7 GHz, which has the ability to launch pure spectrum with high directivity. Grill having large number of sub-waveguides launches a sharp spectrum, which drive LH current more efficiently and therefore was preferred for carrying out plasma physics studies in a controlled manner. An over-dense plasma near the mouth of the launcher provides good coupling of LHWs with plasma and therefore the grill launcher needs to be placed close to the last close flux surface (LCFS). The grill launcher at IPR allows to launch LHWs having a parallel refractive index (N||) from 1.5 to 4.0 by varying the phase difference between adjacent sub-waveguides from 60° to 160°. For ADITYA tokamak, owing to space constraints, the grill launcher with eight sub-waveguides (arranged in two rows, each having four sub-waveguides) could be developed to launch of power up to 120 kW, and therefore, the topology of the ADITYA LHCD was not very complex. However, for the 1 MW LHCD system for SST1 tokamak, the grill launcher with sixty-four sub-waveguides (arranged in two rows, each having thirty-two sub-waveguides) was developed and the SST1 LHCD system, thus conceived, was huge and very complex offering tremendous engineering design challenges. For reactor-grade plasmas, where rf power of a few tens of MW would be launched in to the plasmas, a more compact and reactor-relevant LH launcher, often known as passive–active multijunction (PAM) is foreseen. The PAM launcher offers very good coupling with low reflection co-efficient (∼2–5%) when the plasma density near the mouth of the launcher is near cutoff densities for a given frequency (ωo = ωpe, the electron plasma frequency), and thus allowing it to be placed far away from the LCFS of the plasma, thereby minimizing deleterious effects on launcher due to hostile edge plasma conditions and makes this launcher relevant for reactor-grade plasmas. Further, the space behind the passive waveguide can be utilized for neutron shielding and efficient thermal management. On the contrary, it offers reduced directivity and the spectrum can be varied over a narrow range. Further, the network of transmission lines feeding this launcher is relatively simpler and offers less technological challenges. Thus, 250-kW PAM launcher (comprising of two modules)-based LHCD system for ADITYA-U tokamak has been designed and developed. The launched N|| may be varied over a narrow range (2.25 ± 0.375) by varying the phase difference between adjacent modules over 180° ± 90°. The physics requirements, design philosophy, engineering challenges/solutions, the key outcome and achievements as we embark on to PAM launcher are presented and discussed. Based on our experience of PAM launcher, for ADITYA-U tokamak, conceptual design of the PAM launcher for SST1 is also presented. [ABSTRACT FROM AUTHOR]

AI/ML-based data-driven methodologies are becoming increasingly effective in understanding and predicting plasma disruption in tokamaks by identifying critical signatures present in various diagnostic signals obtained from tokamaks. A high-performance ML-based disruption predictor requires large accurately labelled data. Until now, plasma shots from the ADITYA tokamak have primarily been classified (labelled) as disruptive or non-disruptive manually. Here, we present three computational techniques, namely the Sorted-array approach, the Interval comparison approach and the Threshold-Straight line method for automatic labelling of the ADITYA shots as disruptive or non-disruptive based on the plasma current dropdown time. Statistical analysis and comparison between automatic labelling and manual labelling indicate the promising potential of the proposed techniques. A correlation analysis is also conducted by incorporating plasma diagnostics such as Plasma current, Loop voltage, Bolometer, Mirnov, Hard X-ray, Soft–X-ray, Radiation from Hydrogen-alpha, ionised oxygen and ionised carbon. This comprehensive study offers valuable insights into diverse physical phenomena associated with disruptions. Furthermore, correlation analysis based on current quench time highlights the significance of different diagnostics in providing distinct signatures related to plasma disruption. The insights obtained from this work can play a pivotal role in advancing the development of data-driven disruption prediction systems for ADITYA tokamak. [ABSTRACT FROM AUTHOR]

The detachment cliff is a bifurcative transition to partial detachment recently discovered at the DIII-D tokamak (McLean et al 2015 J. Nucl. Mater. 463 533–6). This work presents a database analysis of target parameters in L-mode and H-mode discharges to search for a detachment cliff at ASDEX Upgrade (AUG). Most of the transitions from attached to partially detached divertor conditions observed in H- and L-mode discharges in AUG show bifurcative-like characteristics that are consistent with the properties of the detachment cliff if the B × ∇ B drift is directed towards the active X-point. In the operational space of power and density, the bifurcative transitions identified during an L-mode discharge occur at injected power and density higher than a threshold value ( P tot > 0.7 MW and n e > 1.6 × 10 19 m−3, respectively). Furthermore, the temperatures at which the transitions start are found to be insensitive to the injected impurity, the injected power and the value of the upstream density. Finally, the study of the evolution of the target parameters, of the intensity of the D α line and of specific manometers and bolometer lines of sights shows that the physical process underlying the detachment cliff and the self-sustained divertor oscillations (Heinrich 2020 Nucl. Fusion 60 076013) might be the same. [ABSTRACT FROM AUTHOR]

Toroidal Alfvén eigenmodes (TAEs) can transport fusion-born energetic particles out of the plasma volume, thereby decreasing plasma self-heating efficiency and possibly damaging reactor walls. Therefore, understanding TAE destabilization and identifying saturation mechanisms are crucial to achieving burning plasma. Here, a fully gyrokinetic study is employed. In the case studied, the primary drive mechanism is identified as the resonance between the magnetic drifts and the TAE, and this is seen to be disrupted by equilibrium flow shear, which can stabilize the mode by rotating it in the poloidal plane. It is found that zonal flows do not play a significant role in the saturation of these TAEs and that there are no saturation mechanisms present in the local gyrokinetic picture, which are able to saturate the mode at physically relevant transport levels in the case of TAE-only turbulence. Instead, we confirm that the global profile flattening of fast-ion density is the key saturation mechanism. The nonlinear excitation of TAEs traveling along the electron diamagnetic direction and its beating with the ion diamagnetic TAE, resulting in large amplitude oscillations that may help detect TAEs more easily in tokamaks, are also reported. [ABSTRACT FROM AUTHOR]

Disruption-generated runaway electron (RE) beams represent a potentially severe threat for tokamak plasma-facing components. Application of properly designed 3D fields can act as a mitigation mechanism, as recently investigated in ASDEX Upgrade (AUG) and COMPASS experiments and in the tokamak discharges of RFX-mod. In all of these devices, the dynamics of the disruption are affected by the application of magnetic perturbations (MPs), and the resulting RE beam current and lifetime are significantly reduced. These experiments show, in particular, that the strength of the observed effects strongly depends on the poloidal spectrum of the applied MPs, which has been reconstructed including the plasma response. This paper reports the main findings on RE mitigation from the previously mentioned three devices, highlighting the common physics behind them and their interpretation by using the guiding center code ORBIT. [ABSTRACT FROM AUTHOR]

The use of spin polarized fuel could increase the deuterium-tritium (D-T) fusion cross section by a factor of 1.5 and, owing to alpha heating, increase the fusion power by an even larger factor. Issues associated with the use of polarized fuel in a reactor are identified. Theoretically, nuclei remain polarized in a hot fusion plasma. The similarity between the Lorentz force law and the Bloch equations suggests polarization can be preserved despite the rich electromagnetic spectrum present in a magnetic fusion device. The most important depolarization mechanisms can be tested in existing devices. The use of polarized deuterium and ³He in an experiment avoids the complexities of handling tritium, while encompassing the same nuclear reaction spin-physics, making it a useful proxy to study issues associated with full D-T implementation. ³He fuel with 65% polarization can be prepared by permeating optically-pumped ³He into a shell pellet. Dynamically polarized 7Li-D pellets can achieve 70% vector polarization for the deuterium. Cryogenically-frozen pellets can be injected into fusion facilities by special injectors that minimize depolarizing field gradients. Alternatively, polarized nuclei could be injected as a neutral beam. Once injected, the lifetime of the polarized fuel is monitored through measurements of escaping charged fusion products. Multiple experimental scenarios to measure the polarization lifetime in the DIII-D tokamak and other magnetic-confinement facilities are discussed, followed by outstanding issues that warrant further study. [ABSTRACT FROM AUTHOR]

This paper describes the upgradation of power supply systems for the improved performance of a small spherical tokamak (MT-I) with an aspect ratio of 1.67 in operation at the Pakistan Tokamak Plasma Research Institute (PTPRI), Islamabad. Three mutually independent power supplies are designed to generate current pulses of desired shape and magnitude in central solenoid (CS), toroidal field (TF) and vertical field (VF) coil systems. The main parts of a power supply are capacitor banks as energy storage devices, ignitrons and thyristors as fast switches for controlled power switching. A double capacitor bank topology with enhanced second capacitor bank voltage is found to be suitable for CS to improve the plasma startup conditions. Simulations are performed in MATLAB Simulink for confirmation of the designed power supply system. The Rogowski coil and current transformers are employed to measure plasma current and the coil’s current waveforms during the experiment in the presence of hydrogen as working gas. An ocean spectrometer and a photron high-speed camera (SA-8) are used for optical measurements and fast plasma imaging. An improvement in startup conditions is observed, enhancing the plasma current duration and light emission intensity recorded in the high-speed camera masked with a Hα filter. [ABSTRACT FROM AUTHOR]

Metallic joints within tokamak devices necessitate high interface hardness and superior bonding properties. However, conventional manufacturing techniques, specifically the hot isostatic pressing (HIP) diffusion joining process, encounter challenges, including the degradation of the SS316L/CuCrZr interface and CuCrZr hardness. To address this, we explore the potential of laser powder bed fusion (LPBF) technology. To assess its viability, we fabricated 54 SS316L/CuCrZr samples and systematically investigated the impact of varied process parameters on the microhardness and tensile strength of the dissimilar metal interfaces. Through comprehensive analysis, integrating scanning electron microscopy (SEM) imagery, we elucidated the mechanisms underlying mechanical property alterations. Notably, within a laser volumetric energy density range of 60 J/mm3 to 90 J/mm3, we achieved elevated interface hardness (around 150 HV) and commendable bonding quality. Comparative analysis against traditional methods revealed a substantial enhancement of 30% to 40% in interface hardness with additive manufacturing, effectively mitigating CuCrZr hardness degradation. [ABSTRACT FROM AUTHOR]

Biaxially prestressed large concrete structures of the confinement building in nuclear power plants (NPPs) should meet the safety requirement for the extension of the service time. Long‐term delayed strains of concrete are one of the key factors determining the safety factor in these structures. This article presents 30‐year long in situ measurement results of strain evolution of confinement buildings in four different NPPs. The delayed strains are predicted at a material level using the next‐generation Eurocode‐2, and the influence of temperature as proposed by the fib model code 2010, making use of delayed strain characteristics of the corresponding concrete from a previous study. We found that the default law given in Eurocode underestimates the delayed strain. However, with the possibility of adjusting the shrinkage and creep laws, the prediction results fit with a good accuracy for the in situ measurement. [ABSTRACT FROM AUTHOR]

In this work the phenomenon of the quasi-coherent mode of plasma fluctuations in purely ohmic plasmas of the T-10 tokamak is investigated. The plasma magnetic fluctuation spectra along with spectra of relative electron density oscillations and plasma potential fluctuation spectra were studied. The presence of both electrostatic and magnetic components of the quasi-coherent mode has been established in the core and as well at the edge regions of the plasma column. This fact allows one to consider the quasi-coherent mode as one of the types of electromagnetic turbulence of fusion plasma. [ABSTRACT FROM AUTHOR]

We develop a method to use the mixed third and second harmonic electron cyclotron emission (ECE) signal in the DIII-D tokamak to reconstruct the electron temperature profile of a rotating magnetic island. The third harmonic ECE is removed by extracting the rotating-island-associated fluctuations in the mixed signal, and the extracted fluctuation is combined with the equilibrium temperature obtained from other diagnostics after correcting for the third harmonic reabsorption. The accuracy of the reconstruction is studied by considering a DIII-D shot where an unmixed signal from an island is available on the low field side of the plasma and a mixed signal from the same island is available from the high field side. It is found that the reconstruction method successfully reproduces the island shape and temperature perturbation magnitude without the distortion caused by third harmonic ECE mixing. However, the radial location of the reconstructed island is somewhat displaced relative to the location of the q = 2 surface in the axisymmetric equilibrium reconstruction, resulting in a corresponding inaccuracy in the absolute temperature of the island. It is conjectured that this may arise from an inaccuracy of the reconstructed axisymmetric equilibrium in this region. [ABSTRACT FROM AUTHOR]

A numerical study of the diamagnetic drift effect on the nonlinear interaction between multi-helicity neoclassical tearing modes (NTMs) is carried out using a set of four-field equations including two-fluid effects. The results show that, in contrast to the single-fluid case, 5/3 NTM cannot be completely suppressed by 3/2 NTM with diamagnetic drift flow. Both modes exhibit oscillation and coexist in the saturated phase. To better understand the effect of the diamagnetic drift flow on multiple-helicity NTMs, the influence of typical relevant parameters is investigated. It is found that the average saturated magnetic island width increases with increasing bootstrap current fraction f b but decreases with the ion skin depth δ. In addition, as the ratio of parallel to perpendicular transport coefficients χ ∥/ χ ⊥ increases, the average saturated magnetic island widths of the 3/2 and 5/3 NTMs increase. The underlying mechanisms behind these observations are discussed in detail. [ABSTRACT FROM AUTHOR]

L-mode negative triangularity (NT) operation is a promising alternative to the positive triangularity (PT) H-mode as a high-confinement edge localised mode-free operational regime. In this work, two TCV Ohmic L-mode core density ramps with opposite triangularity δ ≃ ± 0.3 are investigated using SOLPS-ITER modelling. This numerical study aims to investigate the power exhaust differences between NT and PT focusing, in particular, on the geometrical effect of triangularity. To disentangle the latter from differences related to cross-field transport, anomalous diffusivities for particle ( D n A N ) and energy ( χ e / i A N ) transport are fixed to the same values in PT and NT. The simulation results clearly show dissimilar transport and accumulation of neutral particles in the scrape-off layer for the two configurations. This gives rise to different ionization sources in the edge and divertor regions and produces differences in the poloidal and cross-field fluxes, ultimately leading to different power and particle divertor fluxes in the two configurations. Simulations recover the experimental feature of a hotter and attached outer target ( T e , OSP NT ≳ 5 eV ) in the NT scenario compared to the PT counterpart. [ABSTRACT FROM AUTHOR]

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 > 99%, with SPARC RE losses also occurring much earlier when the boundary is placed at the VV. [ABSTRACT FROM AUTHOR]

The centrifugal force effects from toroidal rotation in improved confinement plasmas are analyzed on high- Z impurities in tokamaks. Tungsten (W) transport simulations are performed using the impurity transport code developed in the integrated code TASK. The geometric factors PA and PB are introduced into the neoclassical transport coefficients to include the effects of the toroidal rotation, which come from poloidal asymmetry in the high- Z impurity profiles. Inward neoclassical particle pinch driven by the main ion density gradient is enhanced by the poloidal asymmetry to be the dominant mechanism for W accumulation in the plasma central region. Simulations with experimental plasma profiles show good agreement with the experimental results and first-principle simulation results in the H-mode. In the hybrid mode and advanced mode, the impurity accumulation is enhanced in the internal transport barrier (ITB) regions. The condition to suppress impurity accumulation is investigated by calculating dependencies on the toroidal rotation velocity and ITB position. The neoclassical transport is sufficiently small with the prospected ITER condition of the Mach number of main ions Mi ∼ 0.1. The impurity transport inside the ITB is strongly influenced by competition between the density peaking effect and the temperature screening effect, and the present simulations show suppression of the impurity accumulation with the outer ITB position to improve the plasma performance, due to the relatively larger temperature gradient of the main ion. [ABSTRACT FROM AUTHOR]

Mock-up of the system for remote monitoring of the accumulation of hydrogen isotopes in the walls of the Globus-M2 tokamak was assembled and tested. The measurements were performed using the LIA-QMS laser diagnostics (laser-induced ablation with registration using quadrupole mass-spectrometry). The data were obtained on the content of hydrogen isotopes in deposits appearing after exposing tungsten collectors to the loads in the volume of the Globus M tokamak. After testing the diagnostics under laboratory conditions, it was mounted at the Globus-M2 facility. In-situ measurements of the content of hydrogen isotopes in the graphite tiles of the tokamak divertor were performed. The possibility of combining the L-IA‑QMS diagnostics with the LIBS (laser-induced emission spectroscopy) diagnostics has been confirmed, in order to obtain information on the composition of the ablated material. In addition, the LIBS method was used for obtaining the deuterium/protium isotopic ratio during measurements in the Globus-M2 facility. [ABSTRACT FROM AUTHOR]

The exponential growth (avalanching) of runaway electrons (REs) during a tokamak disruption continues to be a large uncertainty in RE modeling. The present work investigates the impact of tokamak geometry on the efficiency of the avalanche mechanism across a broad range of disruption scenarios. It is found that the parameter ν * , crit , describing the collisionality at the critical energy to run away, delineates how toroidal geometry impacts RE formation. In particular, utilizing a reduced but self-consistent description of plasma power balance, it is shown that for a high-density deuterium-dominated plasma, ν * , crit is robustly less than one, resulting in a substantial decrease in the efficiency of the RE avalanche compared to predictions from slab geometry. In contrast, for plasmas containing a substantial quantity of neon or argon, ν * , crit ≳ 1 , no reduction of the avalanche is observed due to toroidal geometry. This sharp contrast in the impact of low- vs high-Z material results primarily from the relatively strong radiative cooling from high-Z impurities, enabling the plasma to be radiatively pinned at low temperatures and thus large electric fields, even for modest quantities of high-Z material. [ABSTRACT FROM AUTHOR]

This paper highlights the most important results achieved at the spherical tokamak Globus-M2 with a high magnetic field. This paper also covers the most important topics of fusion research: thermal energy confinement in regimes with neutral beam injection, toroidal Alfvén eigenmode and correspondent fast ions confinement issues, L-H transition, turbulence suppression and edge-localized modes' behavior, experimental and theoretical study of regimes with nitrogen seeding that allow to significantly reduce thermal loads on the divertor plates, and experiments and simulations of lower hybrid current drive. The research results provide the basis for the next step toward a fusion neutron source—the development of the Globus-3 spherical tokamak. [ABSTRACT FROM AUTHOR]

The evolution of plasma profiles and radial electric fields in the edge of tokamaks during a transition to an improved confinement regime is explored in an unfavorable configuration based on self-consistent, flux-driven simulations of reduced Braginskii models with ion pressure dynamics. The edge plasma response to the heating power is explored by varying the heat source strength. The energy transfer from turbulent to mean flows is induced by the large-amplitude fluctuations as the power input becomes sufficiently strong. It is found that ion pressure fluctuations play an important role in the generation of the Reynolds power. Consequently, the plasma spontaneously forms radial electric fields localized at the edge, which reduces the radial correlation and amplitudes of edge fluctuations. An edge temperature pedestal also forms while the density profiles remain nearly unchanged, featuring an I-mode-like regime. [ABSTRACT FROM AUTHOR]

The short exciton diffusion length in organic semiconductors results in a strong dependence of the conversion efficiency of organic photovoltaic (OPV) cells on the morphology of the donor-acceptor bulk-heterojunction blend. Strong light–matter coupling provides a way to circumvent this dependence by combining the favorable properties of light and matter via the formation of hybrid exciton–polaritons. By strongly coupling excitons in P3HT-C60 OPV cells to Fabry–Perot optical cavity modes, exciton-polaritons are formed with increased propagation lengths. We exploit these exciton–polaritons to enhance the internal quantum efficiency of the cells, determined from the external quantum efficiency and the absorptance. Additionally, we find a consistent decrease in the Urbach energy for the strongly coupled cells, which indicates the reduction of energetic disorder due to the delocalization of exciton–polaritons in the optical cavity. [ABSTRACT FROM AUTHOR]

The unusual graphic representation of time series based on Symmetrized Dot Pattern (SDP) helps capturing subtle dynamics in the analyzed signals, otherwise difficult to identify when applying traditional techniques. SDP is creating features and forming a global percept easily readable and recognizable for a human observer. Thanks to this method, local correlations of the signals of any sampled data series can be visualized. This work describes the application of SDP to measurements of tokamak plasma radiation, namely the soft X-ray line-integrated brightness on WEST, where it was thus possible to analyze different phases of the discharge and in particular to identify sawtooth oscillations. In the future, the SDP method could be used to monitor the plasma state and to warn against the appearance of undesirable plasma behavior. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

Nuclear fusion is understood as an energy reaction that does not emit greenhouse gases, and it has been considered as a long-term source of low-carbon electricity that is favourable to curtail rapid climate change. Fusion offers a pathway to resolve energy security and the unequal distribution of energy resources since seawater is its ultimate fuel source and a few grams of fuel can generate mega kilowatts of power. The development and testing of new materials and technologies are unceasing to achieve the net fusion energy through national and international collaboration as well as private partnerships. The ever-growing number of research works report various designs and magnet-based fusion devices, such as stellarators, lasers, and tokamaks. This article provides an overview on the utilization of nuclear energy as a clean energy source, as well as the strategies and progress towards establishing successful commercial fusion energy to the grid and transition to a reliable clean energy source. The overview focuses on the fusion nuclear development in five major countries, UK, US, China, Japan, and Russia. Identified technical and financial challenges are also described at the end of this article. The International Thermonuclear Experimental Reactor (ITER) has been an international reference program for fusion energy development and most developed countries with nuclear development capacity are aiming to complete their in-house fusion energy facilities in parallel to ITER. Many fusion programs are finishing the conceptual design and shifting into the phase of engineering design for the planned DEMO fusion facilities. The significant challenges were identified from the perspective of device efficiency and robustness, sustainable funding, and facility maintenance and safety, which must be addressed diligently to realize fusion energy as alternative clean energy that mitigates climate change and supports the goals of energy security. [ABSTRACT FROM AUTHOR]

The path of tokamak fusion and International thermonuclear experimental reactor (ITER) is maintaining high-performance plasma to produce sufficient fusion power. This effort is hindered by the transient energy burst arising from the instabilities at the boundary of plasmas. Conventional 3D magnetic perturbations used to suppress these instabilities often degrade fusion performance and increase the risk of other instabilities. This study presents an innovative 3D field optimization approach that leverages machine learning and real-time adaptability to overcome these challenges. Implemented in the DIII-D and KSTAR tokamaks, this method has consistently achieved reactor-relevant core confinement and the highest fusion performance without triggering damaging bursts. This is enabled by advances in the physics understanding of self-organized transport in the plasma edge and machine learning techniques to optimize the 3D field spectrum. The success of automated, real-time adaptive control of such complex systems paves the way for maximizing fusion efficiency in ITER and beyond while minimizing damage to device components. Damaging energy bursts in a tokamak are a major obstacle to achieving stable high-fusion performance. Here, the authors demonstrate the use of adaptive and machine-learning control to optimize the 3D magnetic field to prevent edge bursts and maximize fusion performance in two different fusion devices, DIII-D and KSTAR. [ABSTRACT FROM AUTHOR]

Tokamak reactors’ performance is inherently tied to the precise control of plasma shape, position, and current while staying within the operational constraints, specifically managing the control signals and power supply voltages. Furthermore, many tokamak models exhibit strong coupling between control variables, necessitating the use of Multiple-Input Multiple-Output (MIMO) decoupling controllers. The primary control objectives include achieving high tracking and decoupling performance within these operational constraints, along with a focus on robustness. Fractional Order Proportional-Integral-Derivative (FOPID) controllers offer an advantage due to their additional degrees of freedom, which contribute to improved performance, robustness, and flexibility compared to conventional control methods. In light of these advantages, we have designed Optimal FOPID, and also two Integer Order optimal PID (IOPID), controllers for plasma current and horizontal position control in the IR-T1 tokamak, which were optimized using Particle Swarm Optimization to minimize objective functions. Our results have shown that the Integral of Time-weighted Absolute Error (ITAE) criterion exhibits the best tracking and robustness behavior in this context. Examining an actual experimental discharge in this tokamak as a test case, validate that the OFOPID control scheme is effective at maintaining stability when faced with disturbances and fast variations in the plasma parameters. [ABSTRACT FROM AUTHOR]

Tokamak is an important device for controlled nuclear fusion research. The plasma electronic density control system (PEDCS) is an important system for controlling the Tokamak discharge process, which should be of high stability, rapidity, and accuracy. Gas seeding systems are widely used in many Tokamak devices to achieve plasma electronic density control. According to the mechanism model analysis for the plasma electronic density object, an adapted single neuron proportion integration differentiation (PID) control algorithm with the radial basis function (RBF) neural network tuning is studied. The principle and the implementation of the intelligent control algorithm are described in detail in this paper. The intelligent controller enables the system to optimize the PID parameters online according to the density state in the discharge process. The experimental results show that the adapted algorithm achieves a good control effect and also improves the control performance. The proposed method provides a useful reference for Tokamak devices and other similar control systems. [ABSTRACT FROM AUTHOR]

In the experimental advanced superconducting tokamak (EAST), a novel ion cyclotron range of frequency (ICRF) antenna-based diagnostic system is designed to measure ion cyclotron emission (ICE) driven by high-energy ions. The diagnostic system includes ICRF antenna straps, a three-tune impedance matching system, a coaxial switching system, a direct current block, and a data acquisition and storage system. Using the coaxial switching system, the ICRF antenna can be switched from the heating mode to the coupling mode between two discharges. In the 2023 EAST experiment campaign, core ICE was observed using the ICRF antenna-based diagnostic system during neutron beam injection heating, and the obtained results agreed well with the signal detected by the previous high-frequency B-dot probe-based diagnostic system. [ABSTRACT FROM AUTHOR]

As the plasma beta (β) increases in high-performance tokamaks, electromagnetic turbulence becomes more significant, potentially constraining their operational range. To investigate this turbulence, a cross-polarization scattering (CPS) diagnostic system is being developed on the HL-3 tokamak for simultaneous measurements of density and magnetic fluctuations. In this work, a quasi-optical system has been designed and analyzed for the Q-band CPS diagnostic. The system includes a lens group for beam waist size optimization, a rotatable wire-grid polarizer for polarization adjustment, and a reflector group for measurement range regulation and system response enhancement. Laboratory tests demonstrated a beam radius of order 4 cm at the target measurement location (near the plasma pedestal), cross-polarization isolation exceeding 30 dB, and poloidal and toroidal angle adjustment ranges of ±40° and ±15°, respectively. These results verify the system's feasibility through laboratory evaluations. The quasi-optical system has been installed on the HL-3 tokamak during the 2023 experimental campaign to support the development of CPS diagnostics. [ABSTRACT FROM AUTHOR]

The MAST-U Super-X divertor provides the opportunity to study fusion plasma exhaust under novel conditions. However, in order to study these conditions, advanced diagnostics are required. Following the development of the MAST-U Multi-Wavelength Imaging (MWI) diagnostic, we present the installation of a multi-delay coherence imaging spectroscopy (CIS) system within the MAST-U MWI, along with modifications made to the MWI for effective operation. This diagnostic will measure either carbon ion flow velocities and temperatures or electron densities through Dγ emission. We have extended previously developed techniques for wavelength calibration to account for errors due to the misalignment of interferometer components. In addition, we have developed a comprehensive calibration procedure to account for the temperature dependence of the instrument's delays by fitting to a linearly modified version of the delay equation presented by Veiras et al. [Appl. Opt. 49(15), 2769 (2010)]. Together, these procedures reduce the cost and hardware complexity of implementing CIS instruments when compared to those that use in situ or tunable laser calibration systems, as calibrations can be generated to good accuracy using previously measured data. [ABSTRACT FROM AUTHOR]

Divertor detachment is a promising method to solve the power exhaust problem in tokamak devices or even in future magnetic fusion reactors. In this work, a detachment experiment (HL-2A shot #38008) with mixed gas seeding (60% nitrogen and 40% deuterium) is simulated using the SD1D module in BOUT++. In the process from attachment to detachment, the target electron temperature and the target ion saturation current in simulations are found to be consistent with the experimental results measured by Langmuir probes on the target plate. In order to understand the underlying detachment mechanism on HL-2A, this work analyses the role of different particle species in the cases with different seeding rates. It shows that the plasma density varies little and the density of neutrals ( D and D 2 ) slightly decreases with the increase of nitrogen seeding rate, such that plasma–neutral interactions cannot effectively reduce plasma energy and plasma momentum in the divertor. The case with a high seeding rate predicts that increasing seeding rate cannot reach a target temperature lower than 2.5 eV , which is the required temperature for strong plasma–molecule interactions. Thus, the plasma–molecule interactions may not be important in the divertor during nitrogen seeding. This work also studies the parallel forces including (1) force due to the parallel electric field, (2) friction forces, (3) ion- and electron-thermal forces, and (4) collisional reactions (e.g. charge exchange recombination and ionisation). It is found that the friction force between nitrogen ions and other particle species (primarily D + ) is the dominant force towards the target, while ion- and electron-thermal forces are the dominant force pushing nitrogen ions back to upstream. Parallel forces determine the parallel distribution of nitrogen impurities, and therefore decide the region of nitrogen radiation. [ABSTRACT FROM AUTHOR]

High-frequency electromagnetic fluctuations and edge-coherent modes (ECMs) have been detected by Doppler reflectometry (DR) in the pedestal region of the Experimental Advanced Superconducting Tokamak. The frequency of the electromagnetic fluctuations was proportional to the Alfvén velocity and localized in the ellipticity-induced Alfvén eigenmodes (EAEs) gap in the Alfvén continuum. Therefore, the fluctuations are presumed to be EAE. The DR measurements of poloidal flow fluctuations show that EAEs and ECMs have amplitude envelopes that are localized near the bottom of the radial electric field well. Simultaneous EAE and ECM activity generates sidebands at f E A E ± f E C M , the strengths of the upper and lower sidebands are unequal, with the upper sideband being stronger. This is consistent with the theoretical prediction of Chen et al. Compared with ECM-only discharges, the inclusion of EAEs leads to an increase in D α intensity, thus indicating enhanced outward particle transport. [ABSTRACT FROM AUTHOR]

A first-of-its-kind, inductively driven micro-particle (Pellet) accelerator and injector have been developed and operated successfully in ADITYA-U circular plasma operations, which may ably address the critical need for a suitable disruption control mechanism in ITER and future tokamak. The device combines the principles of electromagnetic induction, pulse power technology, impact, and fracture dynamics. It is designed to operate in a variety of environments, including atmospheric pressure and ultra-high vacuum. It can also accommodate a wide range of pellet quantities, sizes, and materials and can adjust the pellets' velocities over a coarse and fine range. The device has a modular design such that the maximum velocity can be increased by increasing the number of modules. A cluster of lithium titanate/carbonate (Li2TiO3/Li2CO3) impurity particles with variable particle sizes, weighing ∼50–200 mg are injected with velocities of the order of ∼200 m s−1 during the current plateau in ADITYA-U tokamak. This leads to a complete collapse of the plasma current within ∼5–6 ms of triggering the injector. The current quench time is dependent on the amount of impurity injected as well as the compound, with Li2TiO3 injection causing a faster current quench than Li2CO3 injection, as more power is radiated in the case of Li2TiO3. The increase in radiation due to the macro-particle injection starts in the plasma core, while the soft x-ray emission indicates that the entire plasma core collapses at once. [ABSTRACT FROM AUTHOR]

To separate the complexities in plasma physics and geometric effects, compact formulas for local geometric properties of a tokamak equilibrium are presented in this paper. They are written in a form similar to the Frenet formulas. All of the geometric quantities are expressed in terms of curvature and torsion of the three spatial curves for the moving local frame of reference, i.e., local orthogonal vector basis. In this representation, the local magnetic shear and the normalized parallel current are just the differences between two torsions of the vector basis. All of the geometric properties are coordinate invariants and form a prime set of quantities for describing tokamak plasma equilibrium. This prime set can be evaluated in both flux coordinates with closed flux surfaces and cylindrical coordinates including areas with open field lines, which may allow the extension of some analysis on the open field lines outside the last closed surface. Fundamental differential operators for stability and transport studies can be expressed explicitly in terms of these geometric properties. It can also be used to simplify analytic studies. [ABSTRACT FROM AUTHOR]

Plasma was heated at the second harmonic of electron cyclotron resonance (ECR) in the L-2M stellarator and the T-10 tokamak. The concept of equivalent tokamak and stellarator discharges was extended to the case of both full and partial absorption of EC power. Comparison of experimental electron temperature profiles with profiles calculated using the canonical profiles transport model allows us to estimate the efficiency of ECR heating in the L-2M discharges without suprathermal electrons, which distort the distribution function, preventing reliable measurements of temperature. The dependence of the ECR heating efficiency on the plasma density was obtained, describing experiments on the L-2M and TJ-II stellarators, and on the T-10 tokamak. The energy characteristics (the stored energy and the confinement time) for L-2M discharges were calculated. Predictions for ECR heating in the T-15MD tokamak are considered. The features of solving the ill-posed transport problem for the L-2M are discussed. [ABSTRACT FROM AUTHOR]

The distinctive features of the T-15MD tokamak microwave interferometer for measuring linearly integrated electron density, as well as the system for processing and recording its signals, are presented. The phase stability of microwave interferometer signals is analyzed. The results of measurements with a microwave interferometer during the first experimental campaign of the T-15MD tokamak are presented. The phase values were extracted by using an analog phase meter and post-processing of digitized microwave interferometer signals: an intermediate frequency signal and a local oscillator signal. It is shown that the results of the plasma density calculation by these two methods coincide. [ABSTRACT FROM AUTHOR]

The high-confinement mode in tokamaks (H-mode) is characterized by high pressure gradients at plasma edge, which results in the appearance of edge localized modes (ELMs). They are studied at the Globus-M2 spherical tokamak too, where edge localized modes are observed mainly in regimes with neutral beam injection. One of the ways for studying ELMs is the use of the Doppler backscattering (DBS) diagnostics installed at Globus-M2. It makes possible to estimate the amplitude of plasma density fluctuations and measure the radial electric field Er. In this work, the effect of edge localized modes on the Er field is studied in the radial range 0.4 < ρ < 1.1. It is shown that during ELMs the electric field increases in the entire measurement range. This indicates that ELMs affect the inner plasma regions as well. This is not consistent with the general ideas concerning the peripheral localization of ELMs, but is confirmed experimentally not only at Globus-M2. In addition, the results for the regime with ELMs are compared with those for the regime with limit cycle oscillations (LCOs) and it is shown that during LCOs such effect is not observed. [ABSTRACT FROM AUTHOR]

Prompt redeposition is a process in which a neutral atom sputtered from a divertor target or the first wall of a tokamak returns on the surface right after ionization, before experiencing any collisions with the background plasma. This paper presents analytical solutions of kinetic equations for sputtered neutrals and resulting ions. Using obtained distribution functions, the redeposition coefficient and average impact angle are derived under assumptions of thin Debye sheath. Our expression for the prompt redeposition coefficient explicitly depends on the angular and energy distributions of sputtered atoms and, therefore, is more universal than often used Fussmann equation valid only for cosine distribution. The effect of the angular distributions of sputtered particles on the prompt redeposition efficiency and parameters of redeposited ions is analyzed. [ABSTRACT FROM AUTHOR]

The influence of collisions on the growth rate of trapped-electron modes (TEMs) in core plasmas is assessed through both analytical linear gyrokinetics and linear gyrokinetic simulations. Both methods are applied to the magnetic geometry of the DIII-D tokamak, as well as the Helically Symmetric eXperiment (HSX) and Wendelstein 7-X (W7-X) stellarators, in the absence of temperature gradients. Here we analytically investigate the influence of collisions on the TEM eigenmode frequency by a perturbative approach in the response of trapped particles to the mode, using an energy-dependent Krook operator to model collisions. Although the resulting growth rates exceed perturbative thresholds, they reveal important qualitative dependencies: a geometry-dependent stabilization rate occurs for all wavenumbers at high collisionality, while at low collisionality, a geometry-sensitive mixture of collisionless, resonantly driven, and collisionally destabilized modes is found. Additionally, linear gyrokinetic simulations have been performed with a rigorous pitch-angle scattering operator for the same geometries. In the case of DIII-D and large wavenumber modes in HSX, the trends predicted by analytical theory are reproduced. Dissimilarities are, however, obtained in W7-X geometry and for low wavenumber modes in HSX, which are shown to be due to a collision-induced transition to the Universal Instability as the dominant instability at marginal collisionality. [ABSTRACT FROM AUTHOR]

Accumulation of heavy impurities in the tokamak core is detrimental for its performance and can lead to disruption of the plasma. In small to medium-sized tokamaks the effective neoclassical transport in the pedestal is typically oriented radially inward. In larger tokamaks—e.g., ITER—where the temperature gradient is higher and density gradients are lower due to the need to operate in a radiative divertor regime, the neoclassical transport is predicted to be outwards. The models are derived for axisymmetric quasi-steady-state plasmas. Applied 3D magnetic fields, i.e., Resonant Magnetic Perturbation (RMPs) as they are used to suppress Edge Localized Modes (ELMs), have experimentally been observed in AUG to enhance the outflow of heavy impurities in the pedestal. There is no model that can predict neoclassical heavy impurity transport in these ergodized 3D magnetic fields self-consistently. In this contribution, we present our kinetic tungsten transport simulation for an ASDEX Upgrade plasma with applied RMPs. Our model based on Hoelzl et al. [Nucl. Fusion 61, 065001 (2021)], van Vugt et al. [Phys. Plasmas 26, 042508 (2019)], and Korving et al. [Phys. Plasmas 30, 042509 (2023)] utilizes a full-orbit pusher, ionization, recombination, effective line, and continuum radiation and neoclassical collisions with the background plasma. The effective collisional radiative rates are from the OpenADAS database, the neoclassical collision operator uses the framework of Homma et al. [J. Comput. Phys. 250, 206–223 (2013)] and Homma et al. [Nucl. Fusion 56, 036009 (2016)]. We show that the adopted collision operator produces neoclassical transport within a satisfactory degree of accuracy. A sufficiently high RMP current causes an increase in tungsten diffusion in the pedestal by a factor of 2. We compare the average radial transport between axisymmetric and 3D RMP scenarios in the pedestal region. RMPs enhance the pedestal permeability for impurities, which results in enhanced transport. In addition to the enhanced transport, some of W is found to be trapped in 3D potential wells in the scrape-off layer. Due to the lack of suitable diagnostics for W in the pedestal, we investigate and suggest that argon can be an adequate substitute in experiments for model validation and further understanding impurity transport in scenarios with applied 3D magnetic fields. With the newly developed neutral model [Korving et al., Phys. Plasmas 30, 042509 (2023)], we can combine the interaction in the divertor with the 3D RMPs to model the tungsten transport from the divertor toward the core of the plasma. [ABSTRACT FROM AUTHOR]

Accurately predicting redeposition is vital for high-Z plasma-facing component (PFC) survivability in magnetic confinement fusion. In this study, we categorize high-Z redeposition into three mechanisms: geometric-driven (prompt), sheath-driven (local), and scrape-off-layer-driven (far) redeposition. To investigate these mechanisms, we employ Monte Carlo transport codes to simulate azimuthally symmetric tungsten source erosion and redeposition in a tokamak. By iteratively analyzing critical parameters, we evaluate redeposition scaling for each mechanism. Specifically, we investigate the impact of magnetic-field-to-PFC pitch angle assumptions on PFC losses into the scrape-off layer. Our findings reveal significant pitch angle sensitivity due to an asymmetric prompt vs local redeposition trade-off. These results enhance our understanding of redeposition phenomena in fusion plasma environments. [ABSTRACT FROM AUTHOR]

The saturation of the low-threshold parametric decay instability of an ordinary wave during electron cyclotron resonance heating in the edge transport barrier of a tokamak due to the stochastic damping of a daughter two-dimensionally localized oblique Langmuir wave has been considered. It has been found that the saturation of instability in modern devices occurs at a relatively low level and does not affect the energy balance during plasma heating. It has been shown that the efficiency of nonlinear pump under expected conditions of microwave power injection in the International Thermonuclear Experimental Reactor will exceed the maximum efficiency of stochastic damping, which will result in the breaking of amplitude-dependent saturation and can significantly modify the power deposition profile. [ABSTRACT FROM AUTHOR]

During the 2019 C4 experimental campaign of the WEST (W Environment in Steady-state Tokamak) (France), the infrared diagnostic produced more than seven terabytes of uncompressed video data. Constraints on the computer infrastructure required for storage, backup, and especially offline access to infrared videos made the use of a compression algorithm mandatory. This paper proposes an innovative method to compress infrared videos with controlled temperature precision. This compression method is based on a controlled averaging of the video that maximizes the compression potential of standard lossless video codecs such as H264/AVC or HEVC. The combination of the loss introduction algorithm and the H264/AVC lossless video codec obtains the best compression ratio in the range of 8 to 41 with a maximum temperature error of 2 °C. This method also outperforms the JPEG-LS algorithm in terms of compression ratio and image quality for the same temperature precision. [ABSTRACT FROM AUTHOR]

This work addresses discrepancies in the alignment of the H-mode pedestal profiles of the electron and ion properties in the DIII-D tokamak as measured by Thomson Scattering (TS) and Charge Exchange Recombination Spectroscopy (CER) diagnostics. While the alignment of these profiles is key for accurate studies of tokamak physics and plasma confinement, misalignments can occur due to inaccuracies, such as in magnetic equilibrium reconstructions required to map measurements in different poloidal and toroidal locations. Both FIDASIM, an established simulation package, and a simplified collisional radiative model are used to simulate neutral beam state densities and neutral beam emission. Simulated neutral beam emissions are calculated based on shifted TS profiles and compared to beam emission measurements from the Main Ion CER system to determine the best shift for aligning TS with CER. This analysis is performed on various DIII-D discharges. [ABSTRACT FROM AUTHOR]

Using a stereo camera system, a new diagnostic for the safety factor of the core plasma based on the pellet ablation trail is applied on the Experimental Advanced Superconducting Tokamak (EAST). In EAST discharge No. 128 874, a shattered pellet injection system is applied to inject a shattered neon pellet into the EAST. Since the strong magnetic field in tokamaks binds the ablated pellet material, the orientation of the pellet ablation trail is the same as the local magnetic field direction. Thus, from the three-dimensional reconstruction result of the pellet ablation trail, the local safety factor q can be obtained. The motional Stark effect (MSE) diagnostic is applied to determine the safety factor q profile in this shot. The determined safety factor q results for this new diagnostic are in quantitative agreement with those from the MSE diagnostic with the mean relative difference of only 6.8%, confirming the effectiveness of this new diagnostic of the safety factor. [ABSTRACT FROM AUTHOR]