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In this work, we explore the triangularity effects on turbulent transport employing global gyrokinetic simulations performed with the ORB5 code. Numerous experiments on the Tokamak á Configuration Variable (TCV) and, more recently, on the DIII-D machine, have demonstrated superior confinement properties in L-mode of negative triangularity (NT) over positive triangularity (PT) configuration. This presents a particularly attractive scenario, as L-mode operation eliminates or significantly mitigates the presence of hazardous edge-localized modes (ELMs). However, a full theoretical understanding of all these observations remains elusive. Specifically, questions remain about how NT improvements can extend to the core where triangularity is very small, and whether these improvements can scale to larger devices. This paper addresses these two questions. Our analysis is divided into two parts: we first demonstrate that the confinement enhancement in NT configurations arises from the interdependent edge-core dynamics, and then we present the results of a system size scan. Crucially, we find that the relative turbulent transport reduction of NT over PT appears not to be contingent on machine dimensions or fluctuation scales and is moreover robust with respect to variations in plasma profiles. This insight underscores the fundamental nature of the NT confinement advantage and paves the way for its potential application in future fusion devices, regardless of their size. [ABSTRACT FROM AUTHOR]

A new integrating model has been developed to couple tokamak edge multiscale magnetohydrodynamic (MHD) events and transport simulations, such as edge‐localized mode (ELM) cycles. As a proof of principle, we first start from a set of three‐field two‐fluid model equations, which includes the pressure, current, and vorticity. The equations are separated into the slowly evolving part of the axisymmetric component by taking a time average of the axisymmetric component. The time‐averaged fluxes, which are quadratic in fluctuating quantities, act as driven terms for the time‐averaged axisymmetric quantities that determine the plasma transport, and therefore the large‐scale evolution of the plasma profiles. Then the HL‐2A's ELM cycles are simulated using the model. Good agreements of ELM size and pedestal recovery time have been achieved for the solutions obtained from the coupled simulation compared with experiment. For one ELM cycle simulation, the coupled code can achieve a speedup of a factor of up to 30 over standalone code. [ABSTRACT FROM AUTHOR]

In its first phase of exploitation, JT-60SA will be equipped with an inertially cooled divertor, which can sustain heat loads of 10 MW/m[sup 2] on the targets for a few seconds, which is much shorter than the intended discharge duration. Therefore, in order to maximize the duration of discharges, it is crucial to develop operational scenarios with a high radiated fraction in the plasma edge region without unacceptably compromising the scenario performance. In this study, the core and edge conditions of unseeded and neon-seeded deuterium H-mode scenarios in JT-60SA were investigated using METIS and SOLEDGE3X-EIRENE codes. The aim was to determine whether, and under which operational conditions, it would be possible to achieve heat loads at the targets significantly lower than 10 MW/m[sup 2] and potentially establish a divertordetached regime while keeping favorable plasma core conditions. In first analysis, an investigation of the edge parameter space of unseeded scenarios was carried out. Simulations at an intermediate edge power of 15 MW indicate that, without seeded impurities, the heat loads at the targets are higher than 10 MW/m[sup 2] in attached cases, and achieving detachment is challenging, requiring upstream electron densities at least above 4 χ 10[sup 19] m[sup -3]. This points toward the need for impurity injection during the first period of exploitation of the machine. Therefore, neon seeding simulations were carried out, performing a seeding rate scan and an injected power scan while keeping the upstream electron density at the separatrix at 3 χ 10[sup 19] m[sup -3]. They show that at 15 MW of power injected into the edge plasma, the inner target is easily detached and presents low heat loads when neon is injected. However, at the outer target, the heat fluxes are not lowered below 10 MW/m[sup 2], even when the power losses in the edge plasma are equal to 50% of the power crossing the separatrix. Therefore, the tokamak will probably need to be operated in a deep detached regime in its first phase of exploitation for discharges longer than a few seconds. In the framework of core-edge integrated modeling, using METIS, the power radiated in the core was computed for the most interesting cases. [ABSTRACT FROM AUTHOR]

A diamagnetic diagnostic system is designed, prepared and installed at Aditya Upgrade Tokamak. Also, the raw data is acquired in a specified acquisition system with suitable front-end electronics. The entire system consists of separate loops and is first calibrated in the presence of all coils, which conduct electricity during the plasma discharge. Then, the change in the toroidal magnetic field due to the plasma is measured using the diamagnetic system, and hence, several plasma parameters including poloidal beta, stored energy, etc. are estimated. The measurements were then repeated during various tests of the tokamak plasma, and several interesting facts were observed. The present paper covers the entire concept of designing magnetic diagnostics – from basic considerations to final data interpretation. [ABSTRACT FROM AUTHOR]

Identification of physical properties of materials is very important because they are in general unknown. Furthermore, their direct experimental measurement could be costly and inaccurate. In such a situation, a cheap and efficient alternative is to mathematically formulate an inverse, but difficult, problem that can be solved, in general, numerically; the challenge being that the problem is, in general, nonlinear and ill-posed. In this paper, the reconstruction of a lower-order unknown time-dependent coefficient in a Cahn–Hilliard-type fourth-order equation from an additional integral observation, which has application to characterizing the nonlinear saturation of the collisional trapped-ion mode in a tokamak, is investigated. The local existence and uniqueness of the solution to such inverse problem is established by utilizing the Rothe method. Moreover, the continuous dependence of the unknown coefficient upon the measured data is derived. Next, the Tikhonov regularization method is applied to recover the unknown coefficient from noisy measurements. The stability estimate of the minimizer is derived by investigating an auxiliary linear fourth-order inverse source problem. Henceforth, the variational source condition can be verified. Then the convergence rate is obtained under such source condition. [ABSTRACT FROM AUTHOR]

We present the design and laboratory tests for a new Q-band frequency tunable Doppler backscattering (DBS) system suitable for probing poloidal wavenumber kñ = 6–8 cm−1 density fluctuations and their flow velocities in the pedestal and scape-off layer (SOL) of the DIII-D tokamak. This system will provide new measurements in the increasingly important and under-diagnosed far pedestal and SOL plasma regions. These results are important for experimental transport studies and necessary for the validation of transport models, both of which are important to fusion energy research. The use of a single tunable frequency reduces the complexity and potential failure points as compared to a multichannel system. This new system utilizes a 33–50 GHz tunable source and will be integrated into the current V-band DBS in DIII-D using a broadband Q- and V-band multiplexer. A full-scale mockup of the quasi-optical system was used to test and optimize the performance. These tests include beam profile measurements at different distances (and angles) from a paraboloidal focusing and steering mirror. The measurements cover the full frequency range 33–75 GHz of the integrated/combined Q–V band DBS system and target a large radial coverage of the low-field side of the plasma from ρ = 1.1 to ρ = 0.5, where ρ is the normalized flux surface radial coordinate. [ABSTRACT FROM AUTHOR]

Recent experiments at the ASDEX Upgrade tokamak have provided the first ever measurements from the imaging heavy-ion beam probe. In this work, we show that the developed simulation framework can reproduce qualitatively the measurement's observed shape and position. Quantitatively, we demonstrate that the model reproduces, within the experimental uncertainties, the observed signal levels. A detailed explanation of the synthetic model is presented, along with the calibration of the optical setup that reproduces the measurements. [ABSTRACT FROM AUTHOR]

The Laser Blow-Off (LBO) impurity injection system is a crucial tool for studying impurity transport and plasma behavior. Conducting proactive impurity transport research is challenging on experimental advanced superconducting tokamak (EAST) due to the uncontrollable generation of impurity sources; therefore, it is necessary to develop a laser blow-off impurity injection system for injecting controlled trace impurity particles. This study presents the design and test results of an LBO system for the EAST. The system aims to provide precise and repeatable control over the timing and quantity of impurity injection. The system primarily consists of a laser source, two mirrors, a moveable focusing lens, a target material, and a vacuum system. The movement of the focusing lens is achieved by a three-dimensional displacement system. The operation of the system is completed by a remote control system. With the accurate control system, the laser spot diameter is adjustable, allowing for modification of impurity injection quantity. The test results demonstrate that the system can rapidly detect external trigger signals and ensure precise timing for the impurity injection. Furthermore, this system can also quickly change the focal point of the laser spot, addressing the requirements for impurity injections during the experiments with less than 0.4 mm position error for laser spot focusing. Test results have shown that the aluminum film material can be peeled off by the LBO system when the laser energy exceeds 650 mJ and the smallest ablation spot is about 1 mm. This study is of significant importance for conducting plasma impurity transport research on the EAST. [ABSTRACT FROM AUTHOR]

Following successful campaigns on the compact high field spherical tokamak ST40, resistive gold foil bolometers have now been installed to measure the radiated power profile. Positioned on the midplane, two bolometer cameras offer perspectives of the horizontal and vertical planes, while the third camera, situated above the midplane, provides a vertical view of the top divertor plates, including the X-point and the strike points. These cameras use a commercial off-the-shelf data acquisition platform with modules specifically designed for bolometers in a Wheatstone bridge configuration, providing both phase sensitive detection and in situ calibration. This paper presents an overview of the design and functionality of all three cameras and the commissioning of the horizontal plane camera. The commissioning of the horizontal camera shows data from a range of plasma pulses, presenting a positive correlation between the relative profiles measured from the bolometer camera and an absolute extreme ultraviolet diode camera viewing along similar lines of sight. [ABSTRACT FROM AUTHOR]

First proton production rates from the d(d,p)t reaction in the Mega Amp Spherical Tokamak Upgrade (MAST-U) are measured. The data were taken during the MAST-U experimental campaign with an upgraded version of the proton detector (PD) previously used in MAST. The new detector array consists of three collimated silicon surface barrier detectors with a depletion depth of 300 μm and a collimated 120 μm thick diamond detector, mounted on the MAST-U reciprocating probe arm. This array measures the energies of unconfined energetic 3 MeV protons and 1 MeV tritons mainly produced by beam-thermal DD reactions during neutral beam injection heating. Diamond detectors have the potential to be uniquely suited to detect charged fusion products as they promise to be much more radiation resistant and much less sensitive to temperature variations compared to silicon-based detectors. Using silicon and diamond-based detectors simultaneously allowed us to directly compare the performance of these two detector types. PD particle rates measured during different plasma scenarios are presented and compared to neutron rates measured using the neutron camera upgrade and TRANSP predictions. [ABSTRACT FROM AUTHOR]

The SPARC tokamak will be equipped with a hard X-ray (HXR) monitor system capable of measuring the bremsstrahlung emission from runaway electrons with photon energies in excess of about 100 keV. This diagnostic will detect the formation of runaway electron beams during plasma start-up and inform the plasma control system to terminate the discharge early to protect the machine. In this work, we present a 0D estimate of the HXR emission in SPARC during plasma start-up. Then we discuss the characterization of a prototype of the HXR monitor. The detector mounts a 1 × 1-in.2 LaBr3 inorganic scintillator coupled with a photomultiplier tube and has been tested with γ-ray sources to find its dynamic range. Finally, two possible modes of operation for spectroscopic and current mode measurements on SPARC are proposed. [ABSTRACT FROM AUTHOR]

A new 14 MeV neutron spectrometer utilizing the magnetic proton recoil (MPR) technique is under development for the SPARC tokamak. This instrument measures neutrons by converting them into protons, whose momenta are subsequently analyzed using a series of magnets before detection by an array of scintillators known as the hodoscope. In this work, we explore various solutions for the hodoscope detectors through laboratory tests with radioactive sources and simulations. We present findings on light collection and pulse shape discrimination based on detector types, as well as optimal solutions for photo-detectors studying the differences between SiPM and PMT. Our results also led to the determination of a better optimized design for the hodoscope detectors, consisting of a 0.7 cm width and a 13 cm length EJ276D scintillation rod. [ABSTRACT FROM AUTHOR]

The two electron cyclotron emission imaging (ECEI) systems installed at adjacent ports (G and H) on the KSTAR tokamak incorporate large-aperture mm-wave optics, broadband electronics, and high speed digitization (up to 1 MSa/s) for 2D and quasi-3D visualization of MHD-scale fluid dynamics. Recently, the ECEI systems have been proved to be capable of visualization of smaller scale fluctuations albeit with a limited spatiotemporal resolution and even capable of measurement of ion cyclotron harmonic waves by direct high-speed sampling of the ECE IF signals. A four-channel prototype subsystem with a higher sampling rate up to 16 GS/s has been integrated into the G-port ECEI system, enabling the measurement of plasma waves in the GHz range in the form of modulated ECE signals and characterization of high-frequency turbulence during the evolution of pedestal. To achieve higher toroidal resolution in the turbulence measurement, the H-port ECEI system is now being upgraded to have a toroidally dual detector array of 2(toroidal) × 12(vertical) × 8(radial) channel configuration and a high-speed subsystem of 2(toroidal) × 4 channel configuration. The new mm-wave optics has been designed via beam propagation simulation, and the measured performance of the fabricated lens indicates a toroidal resolution of 8–10 cm depending on the focus position and zoom factor, allowing for the measurement of parallel wavenumber up to k‖ ∼ 0.8 cm−1. [ABSTRACT FROM AUTHOR]

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]

The DIII-D tokamak has elucidated crucial physics and developed projectable solutions for ITER and fusion power plants in the key areas of core performance, boundary heat and particle transport, and integrated scenario operation, with closing the core-edge integration knowledge gap being the overarching mission. New experimental validation of high-fidelity, multi-channel, non-linear gyrokinetic turbulent transport models for ITER provides strong confidence it will achieve Q ⩾ 10 operation. Experiments identify options for easing H-mode access in hydrogen, and give new insight into the isotopic dependence of transport and confinement. Analysis of 2,1 islands in unoptimized low-torque IBS demonstration discharges suggests their onset time occurs randomly in the constant β phase, most often triggered by non-linear 3-wave coupling, thus identifying an NTM seeding mechanism to avoid. Pure deuterium SPI for disruption mitigation is shown to provide favorable slow cooling, but poor core assimilation, suggesting paths for improved SPI on ITER. At the boundary, measured neutral density and ionization source fluxes are strongly poloidally asymmetric, implying a 2D treatment is needed to model pedestal fuelling. Detailed measurements of pedestal and SOL quantities and impurity charge state radiation in detached divertors has validated edge fluid modelling and new self-consistent 'pedestal-to-divertor' integrated modeling that can be used to optimize reactors. New feedback adaptive ELM control minimizes confinement reduction, and RMP ELM suppression with sustained high core performance was obtained for the first time with the outer strike point in a W-coated, compact and unpumped small-angle slot divertor. Advances have been made in integrated operational scenarios for ITER and power plants. Wide pedestal intrinsically ELM-free QH-modes are produced with more reactor-relevant conditions, Low torque IBS with W-equivalent radiators can exhibit predator-prey oscillations in T e and radiation which need control. High- β P scenarios with q min > 2, q 95–7.9, β N > 4, β T–3.3% and H 98y2 > 1.5 are sustained with high density ( n ¯ = 7E19 m−3, f G–1) for 6 τ E, improving confidence in steady-state tokamak reactors. Diverted NT plasmas achieve high core performance with a non-ELMing edge, offering a possible highly attractive core-edge integration solution for reactors. [ABSTRACT FROM AUTHOR]

The objectives of NSTX-U research are to reinforce the advantages of STs while addressing the challenges. To extend confinement physics of low- A, high beta plasmas to lower collisionality levels, understanding of the transport mechanisms that set confinement performance and pedestal profiles is being advanced through gyrokinetic simulations, reduced model development, and comparison to NSTX experiment, as well as improved simulation of RF heating. To develop stable non-inductive scenarios needed for steady-state operation, various performance-limiting modes of instability were studied, including MHD, tearing modes, and energetic particle instabilities. Predictive tools were developed, covering disruptions, runaway electrons, equilibrium reconstruction, and control tools. To develop power and particle handling techniques to optimize plasma exhaust in high performance scenarios, innovative lithium-based solutions are being developed to handle the very high heat flux levels that the increased heating power and compact geometry of NSTX-U will produce, and will be seen in future STs. Predictive capabilities accounting for plasma phenomena, like edge harmonic oscillations, ELMs, and blobs, are being tested and improved. In these ways, NSTX-U researchers are advancing the physics understanding of ST plasmas to maximize the benefit that will be gained from further NSTX-U experiments and to increase confidence in projections to future devices. [ABSTRACT FROM AUTHOR]

The architecture of the Spherical Tokamak for Energy Production (STEP) has been developed to enable a hybrid maintenance approach using ports in the vacuum vessel for a limited list of tasks that must be performed shortly after shutdown, and larger openings to simplify and speed up major refits. Robotic handling systems in zero-human entry facilities will prevent workers from being exposed to the most hazardous environments. While the approach is largely grounded in existing technologies, the scale and environment of STEP will require significant technology development. Notably, programmes have been established to develop service connections and in-vessel robotic technologies. The engineering integration of the maintenance strategy into the tokamak remains a priority, as does ongoing work to simplify and reduce the cost of the buildings required to facilitate maintenance. Remountable magnet joints are critical to ensuring life-limited magnet components can be replaced during the STEP lifetime and realizing the STEP maintenance strategy. It is a high-risk endeavour owing to the low technology maturity of the potential solutions and owing to the tough and intertwined technical challenges and constraints imposed by both the fundamental physics and the STEP requirements and architecture. An integrated design approach has been taken to balance many competing factors and integrate with interfacing systems, and a multi-faceted technology development programme has been established to address technical risk and to inform, verify and validate the STEP remountable magnet design. This article is part of the theme issue 'Delivering Fusion Energy – The Spherical Tokamak for Energy Production (STEP)'. [ABSTRACT FROM AUTHOR]

This theme issue collects together papers summarising the conceptual design of the Spherical Tokamak for Energy Production (STEP). In 2019, the UK government funded the first design stages of a prototype fusion powerplant based on a compact toroidal geometry, called STEP. The primary technical aims of STEP are to produce net energy, to be self-sufficient in tritium fuel and to demonstrate a maintenance regime that would extrapolate to appropriate availability for commercial powerplants. After 5 years and over 1000 person-years of detailed scientific and engineering conceptual design, this theme issue acts as a compendium of the current design basis for STEP, noting that this is a snapshot in time and that the design will continue to evolve. This article is part of the theme issue 'Delivering Fusion Energy – The Spherical Tokamak for Energy Production (STEP)'. [ABSTRACT FROM AUTHOR]

The Spherical Tokamak for Energy Production (STEP) programme aims to deliver a first-of-a-kind fusion prototype powerplant (SPP). The SPP plasma places extreme heat, particle and structural loads onto the plasma-facing components (PFCs) of the divertor, limiters and inboard and outboard sections of the first wall. The PFCs must manage the heat and particle loads and wider powerplant requirements relating to safety, net power generation, tritium breeding and plant availability. To enable STEP PFC concepts to be identified that satisfy these wide-ranging requirements, an iterative design ('Decide & Iterate') methodology has been used to synchronize a prioritized set of decisions, within the fast-paced, iterative, whole plant concept design schedule. This paper details the 'Decide and Iterate' methodology and explains how it has enabled the identification of the SPP PFC concepts. These include innovative PFC solutions such as a helium-cooled discrete and panel limiter design to increase tritium breeding while providing sufficient coverage and enabling individual limiter replacement; the integration of the outboard first wall with the breeding zone to enhance fuel self-sufficiency and power generation; and the use of heavy water (D2O) within the inboard first wall and divertor PFCs to increase tritium breeding within the outboard breeding zone. This article is part of the theme issue 'Delivering Fusion Energy – The Spherical Tokamak for Energy Production (STEP)'. [ABSTRACT FROM AUTHOR]

The Spherical Tokamak for Energy Production (STEP) programme hypothesizes that a compact machine offers a route to reduced capital cost that directly tackles the barrier to entry of this potentially transformative technology. History has shown that with an unsolved, complex and highly interdependent design challenge, there is a need to balance exploration of the problem with progress. Almost all complex systems arise from the evolutionary improvement of simpler systems which is an approach the programme has adopted by working through a virtual natural selection of design families towards a single concept consistent with the initiating hypothesis. Issues are uncovered and solved more rapidly this way because the effort is focused on an end. In this current phase, STEP has had to be an agile fast-moving programme to work with what emerges as well as what was planned, to sit with uncertainty and to embrace self-organizing principles. The complex decision-making and compromises in emerging trades have led to a concept respectful of the tight aspect ratio hypothesis which carefully balances cost, performance and deliverability. It remains a high-risk and high-reward programme, but the character of the challenge is better understood building confidence and enhancing capability to advance the evolving design further. This article is part of the theme issue 'Delivering Fusion Energy – The Spherical Tokamak for Energy Production (STEP)'. [ABSTRACT FROM AUTHOR]

The programme to design plasma scenarios for the Spherical Tokamak for Energy Production (STEP), a reactor concept aiming at net electricity production, seeks to exploit the inherent advantages of the spherical tokamak (ST) while making conservative assumptions about plasma performance. This approach is motivated by the large gap between present-day STs and future burning plasmas based on this concept. It is concluded that plasma exhaust in such a device is most likely to be manageable in a double null (DN) configuration, and that high core performance is favoured by positive triangularity (PT) plasmas with an elevated central safety factor. Based on a full technical and physics assessment of external heating and current drive (CD) systems, it was decided that the external CD is provided most effectively by microwaves. Operation with active resistive wall mode (RWM) stabilization as well as high elongation is needed for the most compact solution. The gap between existing devices and STEP is most pronounced in the area of core transport, owing to high normalized plasma pressure in the latter which changes qualitatively the nature of the turbulence controlling transport. Plugging this gap will require dedicated experiments, particularly on high-performance STs, and the development of reduced models that faithfully represent turbulent transport at high normalized pressure. Plasma scenarios in STEP will also need to be such that edge localized modes (ELMs) either do not occur or are small enough to be compatible with material lifetime limits. The high current needed for a power plant-relevant plasma leads to the unavoidable generation of high runaway electron beam current during a disruption, where novel mitigation techniques may be needed. This article is part of the theme issue 'Delivering Fusion Energy – The Spherical Tokamak for Energy Production (STEP)'. [ABSTRACT FROM AUTHOR]

The Spherical Tokamak for Energy Production (STEP) environment will include magnetic, thermal, mechanical and environmental loads far greater than those seen in the Joint European Torus campaigns of the past decade or currently contemplated for ITER. Greater still are the neutron peak dose rates of 10−6 displacements per atom, per second, which in-vessel materials in STEP are anticipated to be exposed to. Reduced activation and high-fluence resilience therefore dominate the materials strategy to support the STEP Programme. The latter covers the full life cycle from downselected compositions and new microstructural developments to irradiation-informed modelling and end-of-life strategies. This article discusses how the materials downselection is oriented in plant power trade-off space, outlines the development of an advanced ferritic-martensitic structural steel, describes the 'Design by Fundamentals' mesoscale modelling approach and reports some of the waste mitigation routes intended to make STEP operations as sustainable as possible. This article is part of the theme issue 'Delivering Fusion Energy – The Spherical Tokamak for Energy Production (STEP)'. [ABSTRACT FROM AUTHOR]

The design, delivery and operation of a large-scale infrastructure project are challenging at best. For the Spherical Tokamak for Energy Production (STEP) prototype powerplant (SPP), the challenges increased dramatically. In addition to being a large-scale infrastructure project, it is a cutting edge, first of a kind (FOAK) technology demonstrator. The design teams are working in a volatile, uncertain, complex and ambiguous environment, where technology is constantly emerging, maturing and changing. STEP will be unlike any power plant ever built and requires the development of new technologies and capabilities, but also a novel approach to planning and maturing the design. By taking a holistic view of the engineering life cycle from the start, the programme will be better positioned to achieve an SPP that is fit for purpose and can be used to show a path to ultimate commercial viability for subsequent power plants. This paper will review the key challenges in maturing a FOAK fusion power plant and look in depth at how the STEP team are maturing the required capabilities and planning to ensure successful delivery of the SPP. This article is part of the theme issue 'Delivering Fusion Energy – The Spherical Tokamak for Energy Production (STEP)'. [ABSTRACT FROM AUTHOR]

The Spherical Tokamak for Energy Production (STEP) programme is an ambitious but challenging endeavour to design and deliver a prototype fusion power plant. It is a rapid, fast-moving programme, designing a first of a kind device in a Volatile, Uncertain, Complex and Ambiguous (VUCA) environment, and digital tools play a pivotal role in managing and navigating this space. Digital helps manage the complexity and sheer volume of information. Advanced modelling and simulation techniques provide a platform for designers to explore various scenarios and iteratively refine designs, providing insights into the intricate interplay of requirements, constraints and design factors across physics, technology and engineering domains and aiding informed decision-making amidst uncertainties. It also provides a means of building confidence in the new scientific, technological and engineering solutions, given that a full-scale-integrated precursor test is not feasible, almost by definition. The digital strategy for STEP is built around a vision of a digital twin of the whole plant. This will evolve from the current digital shadow formed by system architecting codes and complex workflows and will be underpinned by developing capabilities in plasma, materials and engineering simulation, data management, advanced control, industrial cybersecurity, regulation, digital technologies and related digital disciplines. These capabilities will help address the key challenges of managing the complexity and quantity of information, improving the reliability and robustness of the current digital shadow and developing an understanding of its validity and performance. This article is part of the theme issue 'Delivering Fusion Energy – The Spherical Tokamak for Energy Production (STEP)'. [ABSTRACT FROM AUTHOR]

The Spherical Tokamak for Energy Production (STEP) requires high-field magnet designs and has therefore adopted the REBCO-based high-temperature superconductor (HTS) as its current carrier. The HTS enables the toroidal field (TF) coils to be remountable, which unlocks STEP's vertical maintenance approach; however, remountable joints, approximately 18 GJ of stored energy and limited space down the centre of a spherical tokamak, make the TF coils the most challenging. STEP has pursued a passive approach to TF coil quench protection in order to limit coil terminal voltage. Initial results suggest that a solution may rely on tuning internal coil resistance coupled with actively powered heaters. The pre-conceptual inter-coil structure demonstrates acceptable stresses and deflections under steady-state operating conditions and preliminary fault scenarios, and loads are distributed to limit the tensile force on the TF centre rod. Finally, the HTS must operate reliably in a high radiation environment and endure high neutron fluences, ensuring commercially relevant magnet lifetimes. Initial experiments indicate that instantaneous gamma irradiation of HTS has no negative impact on current carrying capacity. Experimental programmes are underway to cold irradiate HTS to fusion-relevant fluences and to develop a method of assuring tape irradiation tolerance using oxygen ions as an analogue for neutrons. This article is part of the theme issue 'Delivering Fusion Energy – The Spherical Tokamak for Energy Production (STEP)'. [ABSTRACT FROM AUTHOR]

The UK's fusion energy approach has developed over the past 5 years to include government policy initiatives and a range of public sector investments designed to be delivered in partnership with the private sector. These have aimed to create an environment that stimulates innovation and investment to deliver economic as well as scientific and environmental benefits throughout the lifetime of the public sector fusion energy programme. The Spherical Tokamak for Energy Production acts as a focus and anchor for both public and private sector efforts to develop fusion energy, developing the supply chain and potential for Intellectual Property development and export opportunities well ahead of the anticipated STEP completion date of 2040. This is maximized by the UK's approach to a holistic research and innovation programme backed up by a regulatory and skills programme. This article is part of the theme issue 'Delivering Fusion Energy – The Spherical Tokamak for Energy Production (STEP)'. [ABSTRACT FROM AUTHOR]

Success of the UK's Spherical Tokamak for Energy Production (STEP) programme requires a robust plasma control system. This system has to guide the plasma from initiation to the burning phase, maintain it there, produce the desired fusion power for the desired duration and then terminate the plasma safely. This has to be done in a challenging environment with limited sensors and without overloading plasma-facing components. The plasma parameters and the operational regime in the STEP prototype will be very different from tokamaks, which are presently in operation. During fusion burn, the plasma regime in STEP will be self-organizing, adding further complications to the plasma control system design. This article describes the work to date on the design of individual controllers for plasma shape and position, magneto hydrodynamic instabilities, heat load and fusion power. Having studied 'normal' operation, the article discusses the philosophy of how the system will handle exceptions, when things do not go exactly as planned. This article is part of the theme issue 'Delivering Fusion Energy – The Spherical Tokamak for Energy Production (STEP)'. [ABSTRACT FROM AUTHOR]

The Spherical Tokamak for Energy Production (STEP) prototype powerplant (SPP) will be a first-of-a-kind powerplant—its prime objective is to export electrical power, to the national power transmission system ('grid'), above 100 MWe. As part of a wider issue, addressing the STEP concept design, this article seeks to explore how electrical power will be generated from a spherical tokamak heat source. Accordingly, the following key functions of the SPP power infrastructure are reviewed. Cooling the tokamak: cooling the tokamak while extracting useful thermal energy. Generating power: conversion of thermal energy to electrical energy (power generation). Managing energy: management of the site-wide distribution, storage and energy export. In each of these areas, the design scope, challenges and solution spaces have been discussed. This has shaped the design of the SPP power infrastructure, which in turn has ensured a powerplant design focused on operability and performance. Furthermore, it has been demonstrated that the SPP will achieve its prime objective in generating net power, which is enabled by a unique power infrastructure. Confidence in the ability to generate net power will be refined as the design matures. Finally, this article recommends key opportunities that STEP could use to improve power generation and reduce the parasitic load of the SPP. This article is part of the theme issue 'Delivering Fusion Energy – The Spherical Tokamak for Energy Production (STEP)'. [ABSTRACT FROM AUTHOR]

This article describes why Spherical Tokamak for Energy Production (STEP) has been launched, what it aims to achieve (benefits) and, principally, how the whole programme will be delivered (strategy). The article draws on the work of major project delivery and organization design (OD) and applies this to the context of STEP, which is dominated by significant uncertainty in all dimensions (technical, financial, commercial and programmatic), where there is embryonic delivery capability, but where there are also global-scale opportunities. This leads to an approach based on securing and organizing the correct capability from both public and private sectors to work in a collaborative arrangement with a single purpose and, critically, in an operating model designed to manage uncertainty and emerging risks and to exploit opportunities. Placing adaptability at the core of the OD, particularly the ability to deliver emergent strategy through guided empowerment in pursuit of an ambitious aim, is a further development beyond much of the current thinking in major projects. The article concludes with an appendix that translates that programme approach into principles for managing the engineering design work. This article is part of the theme issue 'Delivering Fusion Energy – The Spherical Tokamak for Energy Production (STEP)'. [ABSTRACT FROM AUTHOR]

The European DEMO (EU-DEMO) reactor studies within EUROfusion aim to develop a fusion power plant concept. The large tokamak device needs an auxiliary heating power which, at the present stage, is provided by the Electron Cyclotron (EC) heating system with up to 130 MW foreseen to reach different regions of plasma for heating, suppression of instabilities and the possibility to support ramp-up and ramp-down phases. The present conceptual design of the system is based on 2 MW coaxial-cavity gyrotron sources, a transmission line (TL) using both circular corrugated waveguides and quasi-optical evacuated multi-beam TLs, and mirror antennas located in the Equatorial Port. In order to create a modular system, the sources are grouped in 'clusters', whose powers are combined in the quasi-optical TL, up to the tokamak building, where they are split and routed as single waveguides. In the launcher, they are combined together again on the launching mirrors, to save space for the apertures in the Breeding Blanket. The present EC heating system has a certain flexibility to adapt to changing design guidelines. The development status of the system is presented. [ABSTRACT FROM AUTHOR]

The work reported in this paper addresses two aspects. In the first part, numerical simulations are conducted to examine the impact of magnetic equilibrium shaping (elongation and triangularity), on both conventional Ion Temperature Gradient (ITG) modes and Short Wavelength ITG modes. This analysis is performed considering the experimental profiles and parameters of the ADITYA-U tokamak, employing the nonlinear global gyrokinetic Particle-in-Cell code ORB5. The linear and nonlinear collisionless electrostatic simulation of these modes are carried out with kinetic ions and adiabatic electrons. From the linear results, it has been observed that the magnetic equilibrium shaping slighty reduced the growth rates and widened the spectrum, and the maxima of growth rate curve is shifted to higher toroidal wave number. We find that the heat flux is reduced by a significant ≃ 35 % for the true circular magnetohydrodynamic magnetic equilibrium as compared to ad hoc concentric circular equilibrium reported in Singh et al (2023 Nucl. Fusion 63 086029). A further ≃ 10 % reduction in the heat flux is seen with magnetic equilibrium shaping. In the second part, linear collisionless electrostatic simulation studies of ITG coupled with fully kinetic electrons (both trapped and passing electrons are treated kinetically) using a drift-kinetic ordering is performed. It can be seen from the linear results that, in presence of kinetic electrons, the growth rate and real frequency of the ITG mode are increased significantly for ADITYA-U parameters and a mode propagating in the electron diamagnetic direction is identified at high toroidal wavenumbers. [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]