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An understanding of the behaviour of the D or He fuel used in tokamak discharges is essential in analyses such as modelling edge and divertor transport and the erosion of the vessel walls. However, poor agreement is found between measurements made on the JET tokamak and collisional-radiative models used to predict the hydrogen-like D and He line intensities. The range of temperatures of the plasmas emitting the radiation is also limited, in contrast to that for many impurities for which a wide range is possible. This is particularly so for He II whose line intensities tend to have the same near-constant ratios in most pulses, suggesting that the emission originates in plasma regions with very similar electron temperatures. To gain understanding and to allow quantitative comparisons with theoretical models, extensive observations of the VUV Lyman series have been made, for all discharge scenarios run during three He campaigns. Those for He discharges in both JET ITER-like wall (JET-ILW) and JET C (JET-C) campaigns are presented here. He discharges have the advantage of fewer impurities resulting in less complex spectra than when D is used as the fuel. However, the characteristics of the observed discrepancies are similar in both species, allowing He to be used as a proxy for D in order to gain understanding of the discrepancy. In addition, the study of He avoids the complication of molecular species contributing to the level populations. Opacity effects are also expected to be less severe in He discharges. Nevertheless, so as to ensure that the measurements are not unduly affected by opacity, comparisons have also been made with emission from Balmer and Paschen series members. Measurements of both line intensities and their ratios are presented for all-pulse surveys and for individual pulses. In exceptional cases in which the He emission is intense a dependence on the He II line intensity is demonstrated. The discrepancy between these measurements and the theoretical models is illustrated. [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]

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 article considers different methods of divertor plasma diagnostics planned for using in the T‑15MD tokamak. Technical problems arising during operation of optical systems in the divertor zone are discussed, including degradation of in-vessel optical elements. The main attention is paid to the conceptual design of the "Passive Spectroscopy in Divertor" diagnostic system. The optical system, including in-vessel mirrors, and methods for protecting its components from the negative effects of plasma are described in detail. Using synthetic diagnostics and numerical simulation methods, the possibility of solving the problem of tomographic reconstruction of the two-dimensional profile of plasma radiation in the T-15MD divertor is demonstrated. Based on the results presented, it was concluded that passive spectroscopy can be used for obtaining data on plasma parameters in the divertor with good spatial resolution, which will make it possible to study the physics of processes and monitor the operation of the T-15MD divertor, including the operation in detachment regime. [ABSTRACT FROM AUTHOR]

The Hard X-Ray collimation system was installed at the T-15MD tokamak (major radius of R = 1.48 m, minor radius of a = 0.67 m, elongation of k = 1.7–1.9, triangularity of δ = 0.3–0.4, toroidal magnetic field of BT = 2 T, plasma current of up to Ip = 2 MA, and discharge duration of up to 30 s) to provide the possibility of measuring the spatial and temporal evolution of runaway electron beams. The diagnostic is based on LaBr3(Ce) scintillation detectors, witch provides energy spectrum measurements for energies form 0.3 to 10 MeV. Monte Carlo simulation of the shielding efficiency in the GEANT4 environment showed that the collimator provides 2° field of view (FOV). Horizontal and vertical FOV movement along the plasma cross-section is arranged by a rotating base and screw support elements. An algorithm for processing the signals from the scintillation detectors with a graphical user interface and data visualization is used for reconstruction of the spectra. The results of testing the detector in a laboratory setup and in preliminary experiments carried out with the T-15MD tokamak in which runaway electrons beams were generated are presented. [ABSTRACT FROM AUTHOR]

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

In this study, the effects of high-energy hydrogen ion irradiation, generated by the MTPF-2 type modern plasma focus device, on the surface and structural properties of graphite were investigated. High-energy protons produced by the plasma focus device were used to irradiate graphite samples in various shots. The radiation-induced changes on the graphite surface were examined using optical microscope and scanning electron microscopy. The modifications on the irradiated samples' surface were clearly observed in the scanning electron microscopyand optical microscope images. The microscopic images of the irradiated samples revealed point sputtering as well as pores on the sample surface, along with point melting, which are dependent on the ion fluence (number of shots). X-ray diffraction analysis was employed to study the structural changes in the graphite caused by the high-energy proton irradiation. The X-ray diffraction spectra of the irradiated samples showed shifts in the peak positions, changes in the peak intensities, and an increase in the crystallite size compared to the X-ray diffractionspectra of the reference sample. Additionally, a Faraday cup detector was used to characterize the hydrogen ion spectrum produced in the plasma focus device. The results indicated that the average energy of the generated ions was ~ 46 keV. [ABSTRACT FROM AUTHOR]

Tokamaks are a very promising option to exploit nuclear fusion as a programmable and safe energy source. A very critical issue for the practical use of tokamaks consists of the power flow required to initiate and sustain the fusion process, in particular in the poloidal field coils. This flow can be managed by introducing a DC energy storage based on supercapacitors. Because such storage may be the most expensive and largest part of the poloidal power supply system, an excessive size would cancel its potential advantages. This paper presents innovative strategies to optimize the DC storage in poloidal power supply systems. The proposed solution involves the sharing of the DC storage between different coil circuits. The study is supported by novel analytical formulas and by a circuital model developed for this application. The obtained results show that this method and the related algorithms can noticeably reduce the overall size of the storage and the power exchange with the grid, providing a practical contribution toward the feasibility and the effectiveness of nuclear fusion systems. [ABSTRACT FROM AUTHOR]

Introduction: This work explores the use of eXplainable artificial intelligence (XAI) to analyze a convolutional neural network (CNN) trained for disruption prediction in tokamak devices and fed with inputs composed of different physical quantities. Methods: This work focuses on a reduced dataset containing disruptions that follow patterns which are distinguishable based on their impact on the electron temperature profile. Our objective is to demonstrate that the CNN, without explicit training for these specific mechanisms, has implicitly learned to differentiate between these two disruption paths. With this purpose, two XAI algorithms have been implemented: occlusion and saliency maps. Results: The main outcome of this paper comes from the temperature profile analysis, which evaluates whether the CNN prioritizes the outer and inner regions. Discussion: The result of this investigation reveals a consistent shift in the CNN's output sensitivity depending on whether the inner or outer part of the temperature profile is perturbed, reflecting the underlying physical phenomena occurring in the plasma. [ABSTRACT FROM AUTHOR]

Ion cyclotron resonance heating is considered as one of the methods of additional heating of plasma and production of the non-inductive current in the T-15MD tokamak. To transfer the maximum power to the plasma, it is needed to know impedance of an antenna–plasma system, to match it with impedance of an RF power generator and its transmission line. The work is devoted to the development of a code for the calculation of antenna impedance of the ICR heating system of plasma in toroidal magnetic traps. To find impedance of the antenna–plasma system in the simplified geometry of antenna consisting of conductive plates, the wave equation is solved in the "cold" plasma approximation, and the spectrum of the RF power emitted by antenna is calculated. The dependences of the impedance of the antenna–plasma system on distances between antenna and the Faraday screen and between the Faraday screen and the plasma are obtained for the geometry of the T-15MD tokamak. Two-dimensional distribution of electric field of a wave in the plasma is obtained. [ABSTRACT FROM AUTHOR]

Experimental results from several tokamaks suggest a strong impact of divertor density regimes on turbulent transport in the edge plasma. Reciprocally, the change in transverse transport and SOL width affects the access to density regimes, making it a fundamental topic for heat exhaust issue. Such phenomenology is highly nonlinear and can only be approached quantitatively using numerical simulations treating turbulence and neutrals recycling physics self‐consistently. In this study, the SOLEDGE3X edge multi‐fluid code is used to investigate the mutual interaction between edge plasma turbulence and neutrals recycling. A fluid neutrals model based on the assumption of a charge‐exchange‐dominated plasma‐neutral interaction has been implemented. Two simulations in circular geometry are compared: one without neutrals, where the particle flux is driven by a constant in flux from the core region, and the other one with neutrals recycling included in which the particle input to the system is self‐consistently injected by a gas puff from the midplane. The presence of the neutrals triggers three types of perturbations on the plasma: a local and non‐axisymmetric one driven by the gas puff, a global perturbation affecting both profiles and turbulence properties in the whole domain, and a local one in the vicinity of the limiter where recycling occurs. The largest effect of the inclusion of self‐consistent neutrals recycling is a large‐scale reorganization of the plasma profiles and turbulence properties due to the dissociation of particle and energy fluxes. These effects are expected to be more important at higher densities regimes or in diverted configuration. [ABSTRACT FROM AUTHOR]

In studies of impurity transport in quasi-stationary hot plasma, the initial kinetic equation and the diffusive-convective transport model take into account ionization and recombination as "sources and sinks" of particles. Due to the incompatible representation of the radial dynamics and charge kinetics of impurity charge states, this approach and the results obtained appear to be out of system. The basis for their systematic criticism is the ideas of the theory of random processes proposed by M.A. Leontovich in 1935 as a theoretical alternative to the gas-kinetic equation. In this case, the charge-radial transport of an impurity in a quasi-stationary plasma is defined as a syncretic vector random Markov process of charge state transport. Its coupling (ergodicity) in a two-dimensional Markov system excludes "sources and sinks" from it in principle, and the relaxation convergence is directed to the formation of equilibrium invariant density profiles. The impurity equilibrium and density profiles are specified by a system of invariant functions that provide analysis of any types of density profiles observed in experiments. Modeling of radial profiles of helium, boron and carbon impurities allows us to find variants of their transformation from accumulation in the center to concentration near the plasma edge, transport coefficients and systematic connection with plasma parameters. [ABSTRACT FROM AUTHOR]

The results of the first gyrokinetic calculations of anomalous heat fluxes in the T-10 tokamak plasma obtained for typical conditions of a discharge no. 71 568 with ohmic heating are presented. The calculations have been performed at the Kurchatov Institute Supercomputer Center. The experimentally measured electron density and temperature profiles, ion temperature profiles with a large gradient leading to the so-called ion temperature gradient (ITG) turbulence, and also the profiles of carbon and oxygen impurity densities measured using the charge exchange recombination spectroscopy (CXRS) active diagnostics are used as input data. The "experimental" electron and ion heat fluxes are estimated from the heat balance condition using the ASTRA transport code. The analytical dependence of heat fluxes on the effective plasma charge is presented. Gyrokinetic calculations of anomalous electron and ion heat fluxes are performed for the T-10 tokamak for the first time. The well-known gyrokinetic GENE code is used in the so-called linear and nonlinear approximation with fixed density and temperature gradients taking into account the influence of carbon and oxygen impurities. A linear dependence of heat fluxes on the effective plasma charge is found, and the sensitivity of the results to input parameter errors is investigated. The results of gyrokinetic calculations for the T-10 tokamak are compared with the results obtained for facilities with similar input parameters. A comparison is made of gyrokinetic calculations of heat fluxes performed using the GENE code with the results of calculations by the CONTRA-T code, intended for the self-consistent simulation of low-frequency turbulence and transport processes in tokamaks with a large aspect ratio. Good agreement obtained in the work between the results of transport calculations using the ASTRA, GENE, and CONTRA-T codes based on various transport models for the ohmic discharge of the T-10 tokamak with a circular cross section, provides grounds for the further simulation of transport processes in plasma with additional heating and a more complex cross section shape of the plasma column. [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]

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]

Low-frequency drift-wave instabilities play a crucial role in the radial transport of present-day tokamaks, and trapped electron collisions can significantly influence these instabilities. In this paper, the effects of trapped electron collisions on these instabilities are investigated based on linear gyro-kinetic simulations. The basic numerical techniques including dispersion relation integral method and orthogonal basis function expansion are presented in detail with necessary benchmark work. The results demonstrate that in medium gradients, the increase of trapped electron proportions promotes the growth rate and radial heat transport largely for quasi-linear trapped electron modes (TEMs) and ion temperature gradient (ITG) modes. Moreover, trapped electron collisions have strong stabilizing effects, especially for TEMs driven by electron temperature gradients. Two distinctive branches, namely Mode #1 and #2, are investigated in steep gradients. Both behave in a varied instability nature during different ranges of the normalized wave vector k ^ θ . Mode #1 mainly induces radial heat transport during k ^ θ < 0.5 and is significantly suppressed by the collisions. Mode #2 mainly induces radial heat transport during 0.4 < k ^ θ < 0.8 , and is largely enhanced by the collisions. When the collisionality is large enough, Mode #2 has stronger transport capacity than the other. Mode #2 at the medium wave vector, known as dissipative TEM, may provide the mechanism of the edge coherent mode observed in EAST H-mode plasmas, wherein collisionality plays an important role in the mode excitation. [ABSTRACT FROM AUTHOR]

In this research, we develop a data-driven disruption predictor based on Bayesian deep probabilistic learning, capable of predicting disruptions and modeling uncertainty in KSTAR. Unlike conventional neural networks within a frequentist approach, Bayesian neural networks can quantify the uncertainty associated with their predictions, thereby enhancing the precision of disruption prediction by mitigating false alarm rates through uncertainty thresholding. Leveraging 0D plasma parameters from EFIT and diagnostic data, a temporal convolutional network adept at handling multi-time scale data was utilized. The proposed framework demonstrates proficiency in predicting disruptions, substantiating its effectiveness through successful applications to KSTAR experimental data. [ABSTRACT FROM AUTHOR]

Predictive design modeling of a Dissipation-Focused Divertor for future operation in DIII-D reveals that increasing the poloidal distance of the pump duct entrance from the target surface along the low-field side divertor baffle increases neutral compression and modifies the spatial distribution of power dissipation. With a divertor pump located mid-leg between the target and the X-point, SOLPS-ITER boundary plasma simulations without drifts predict the formation of a dense neutral cloud near the target with > 30x higher neutral compression in detachment, a more stable detachment front located further from the target, and ∼25% lower outer midplane separatrix density required for detachment onset, compared to a pump located in the scrape-off layer at the target surface. Up to 19 MW of power flowing into the divertors is modeled using the following two numerical implementations for particle pumping: a specified fraction of particles incident on variable wall sections of the plasma grid is removed from the computational domain (so-called albedo pumping), and a pump duct is modeled which includes dynamics of kinetic neutrals in the duct. The simulations show that the detachment front is located between the divertor target and the X-point and is relatively stable near the pump entrance, without a strong dependence on gas puff rate or injected power. The mid-leg pump design spatially separates the two primary functions of a divertor (power handling and particle exhaust), with the majority of power dissipation occurring near the target plate and particle exhaust taking place further upstream. The benefit of enhanced dissipation using mid-leg pumping comes at the cost of a higher outer midplane separatrix density for a given amount of particle injection.

The technique based on quadrupole mass spectrometry of the material ejected by laser-induced ablation – LIA-QMS is proposed as a tool to measure H/D/T content in carbon co-deposits in the first wall and divertor regions of tokamak reactors. Tungsten tiles exposed in Globus-M2 tokamak were taken for validation experiments. Measurement accuracy was determined by comparison with the results obtained by laser-induced desorption – LID-QMS and conventional thermal desorption spectroscopy (TDS). Total amount of deuterium in hydrocarbon deposits was found to be 2.8 × 1017 D/cm2, when measured by – LIA-QMS, 3.0 × 1017 D/cm2 for LID-QMS and 2.9 × 1017 D/cm2 for TDS. The main uncertainty of D surface concentration are caused by inhomogeneity of the deposit thickness over the sample area or by presence of hydrocarbon species in the released gas.

Fusion power plants will require detachment to mitigate sputtering and keep divertor heat fluxes at tolerable levels. Controlling detachment on these devices may require the use of real-time scrape-off-layer modeling to complement the limited set of available diagnostics. In this work, we use the configurable Hermes-3 edge modeling framework to perform time-dependent, fixed-fraction-impurity 1D detachment simulations. Although currently far from real-time, these simulations are used to investigate time-dependent effects and the minimum physics set required for control-relevant modeling. We show that these simulations reproduce the expected rollover of the target ion flux — a typical characteristic of detachment onset. We also perform scans of the input heat flux and impurity concentration and show that the steady-state results closely match the scalings predicted by the 0D time-independent Lengyel–Goedheer model. This allows us to indirectly compare to SOLPS simulations, which find a similar scaling but a lower value for the impurity concentration required for detachment for given upstream conditions. We use this result to suggest a series of improvements for the Hermes simulations, and finally show simulations demonstrating the impact of time-dependence.

Plasma-wall interaction (PWI) is a critical concern in tokamaks because of its significant impact on the lifetimes of plasma-facing materials (PFMs), fuel retention, and plasma performance. In 2020, the Experimental Advanced Superconducting Tokamak (EAST) PFMs were upgraded to predominantly metallic walls. Following the 2021 experimental campaign, the deposition distribution and fuel retention on the surfaces of the PFMs along both the poloidal and toroidal directions were analyzed. The poloidal tests commenced at the high-field side (HFS), proceeded to the lower divertor, then to the low-field side (LFS), and finally to the upper divertor. The toroidal tests were performed at the midplane of the HFS, beginning from port A and ending at port P. The distributions of the deposits in the poloidal and toroidal directions were clearly asymmetrical. Mo and Fe particles sputtered from the first wall and inner stainless steel (SS) components were prone to deposition in the lower divertor region, as evidenced by the fact that the elemental content in the far-SOL region of the inner divertor exhibited Mo and Fe peaks, as did both the near- and far-SOL regions of the outer divertor. In addition, quick re-deposition of W and Fe was observed, as demonstrated by the fact that their contents near the erosion sources were higher than those farther away along the toroidal first wall on the HFS. This tendency was stronger for W than for Fe. Further analysis indicated that the deposits consisted of Li, C, O, W, Mo, Fe, Cu, and Ni. The Li and Fe contents were much higher than those of other metal impurities, with peak values of 8.17 μg/mm2 and 7.78 μg/mm2, respectively. The Li content decreased along the HFS first wall from 8.17 μg/mm2 at P-2 to 1.82 μg/mm2 at P-22, and then increased to 2.72 μg/mm2 at P-32 in the divertor, while the Fe content was higher around the top side and the midplane of the HFS. The deposited Li originated from routine wall conditioning and existed primarily in the form of Li2CO3. Additionally, other metal impurities deposited on the Li2CO3 surfaces exhibited various irregular shapes, often appearing as aggregated and recrystallized small particles. Furthermore, significant deuterium (D) retention on the order of 1021 atoms/m2 was measured on all the analyzed SiC-coated graphite tiles on the HFS. The D content at location P-10 at the midplane was the highest along the poloidal direction of the HFS because it was directly facing the NBI beam.

Real-time monitoring and control of thermal loads on tungsten plasma-facing components is mandatory for ITER to avoid their overheating during plasma discharges. For the Start of Research Operations (SRO) phase, dedicated First Wall Heat Load Control (FWHLC) functions are included in the ITER Plasma Control System (PCS) to prevent the development of excessive plasma heat loads on the inertially cooled first wall (FW). In this work, we propose a FWHLC design with a model-based approach, which features a simplified thermal model for the FW armour. This control-oriented model simulates the thermal response of ITER FW to slow changes in plasma equilibrium and energy, which during ITER operation will be monitored by the real-time wide angle infrared camera system. This allows computationally efficient runs for rapid controller performance assessment but can also assist operation scenario design by pre-empting potential heat load issues. FWHLC is designed to react to measured and predicted FW temperatures overcoming predesigned limits with real-time variations of the plasma shape or auxiliary power references, aiming to reduce thermal loads before a premature plasma termination is required to protect the FW. The new scheme is tested in closed loop simulations, where perturbations to nominal ITER low current scenarios are introduced in the wall thermal model to trigger the response of FWHLC, supporting the effectiveness of the proposed policy.

The estimation of the heat flux density distribution profiles in tokamak devices is a very important research topic for edge plasma physics purposes and also to ensure the safety of the machine. In the radial direction, the heat flux exhibits an exponential decay that could be captured by thermal sensors distributed in the plasma facing components. Radially distributed thermal sensors based on Fiber Bragg grating technology have been embedded in the WEST lower divertor to study the heat flux deposition profiles during plasma operation. The comparison between embedded measurements and a 3D finite element model shows a small decay length (5 – 10 mm) on top of a wider heat flux with a decay length around 30 to 50 mm. A tool using neural network has been developed in order to predict the values of the different parameters describing the deposited heat flux from embedded temperature measurements in steady state. A large span of deposited heat fluxes with maximum heat flux ranging from 1 to 9 MW/m2 and decay length from 5 to 50 mm were characterized using this tool over a database of more than 250 experimental L-mode pulses performed in WEST in attached divertor configuration. The comparison of the predicted heat flux parameters values with macroscopic plasma parameters have revealed the appearance of the narrow component with the increase of the divertor power load (Pdiv) with a threshold dependant of the plasma current (IP).

Global power balance calculations in steady state H mode plasmas varying the distance between separatrices (drsep) and the divertor configuration have been performed in MAST Upgrade. As drsep becomes more negative, more of the power crossing the separatrix goes to the lower divertor. The inner divertor receives a higher fraction of the power exhaust in a Super-X divertor plasma compared to a conventional divertor plasma at similar negative drsep, which is a concern for high power devices employing alternative divertor configurations for power exhaust handling. Global power accounting suggests >30% of the input power is unaccounted for with the power loss channels quantified in this work. Charge exchange and orbit losses from the NBI could account for a large fraction of unaccounted power but it is not possible to precisely determine this without further diagnostic calibration.

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]

[Introduction] The superconducting tokamak serves as the foundation for steady-state operation. Timely and reliable quench detection is the key to ensuring the secure operation of superconducting magnets, which has the highest level of safety in fusion device operation. [Method] In this paper, an overview of the principles and implementation approaches of quench voltage detection for fusion magnets was provided. It briefly summarized the research progress of quench detection and the quench detection schemes adopted by mainstream tokamak devices and introduced the basic principles and processes of quench voltage detection. Taking the experimental advanced superconducting tokamak (EAST) as an example, this paper analyzed the interference sources and coupling mechanisms of quench voltage detection, proposed a two-stage decoupling compensation scheme, and established an optimization analysis model for primary compensation and a dynamic compensation mechanism for plasma-coupling interferences. [Result] The experimental results indicate that the compensation scheme can achieve a high noise suppression ratio greater than 99.9%, which can effectively improve the signal-to-noise ratio (SNR) and reliability of quench detection. [Conclusion] The established optimization analysis model for primary compensation and the dynamic compensation mechanism for plasma-coupling interferences can be extended to other tokamak devices, thereby providing necessary accumulation for the design of quench detection and the secure operation of magnets in future fusion reactors.

[Introduction] Fusion energy has the characteristics of large energy released through reactions, safe and reliable operation, abundant fuel sources, and low environmental pollution. Therefore, it is expected to become a commercial energy source that can be supplied in the market on a large scale, providing stable energy output and power supply in the future. To popularize the development path of magnetic confined fusion energy in China, the article reviews the discovery and realization path of fusion energy. [Method] The article provided an overview of the early research and development process of magnetic confined fusion energy in China by the method of literature review. The paper took the development of magnetic confined fusion energy as an example to provide a preliminary overview of the construction of typical research devices in China, such as tokamak, stellarator, spherical tokamak, reversed field pinch, magnetic mirror field, linear plasma, and dipole magnetic field devices. [Result] Based on the construction and researches of these devices, China has cultivated a group of scientific and technological talents in the field of magnetic confined fusion research, achieving significant progress. In addition, the article provides an overview of international cooperation in fusion energy research, as well as the International Thermonuclear Experimental Reactor (ITER) project that China is participating in construction. [Conclusion] Although the current research on fusion energy still needs to overcome significant challenges from various aspects, such as fusion plasma physics, fusion reactor materials, and tritium self-sustaining technology, it is believed that the magnetic confined fusion energy in China will turn from a blueprint into a reality in the future thanks to the urgent need for energy structure transformation and the strong support for fusion researches in China.

[Introduction] Plasma disruption poses a significant threat to the tokamak nuclear device during its running and can cause damage to the device. Such damage can be reduced by adopting the disruption mitigation system, which has an action time highly dependent on the real-time running plasma disruption prediction system for predicting the plasma disruption moment. The deep-learning-based neural network has been used to train plasma disruption prediction models, and the real-time running of the deep-learning-based disruption prediction models requires a huge amount of real-time data from multiple diagnostics. [Method] The article proposed a design scheme for a real-time data acquisition system. The real-time data acquisition and transmission system was designed based on the modular structure and divided into the multiple channels acquisition module, ADC converting control and data reading module, data grouping and packing module and data transmission network module. The data transmission network module was developed on the hardware UDP network stack running on the FPGA at a speed of 10 G. This hardware UDP network stack featured a deterministic data transmission process, enabling a very low transmission latency of the system. [Result] The real-time data acquisition system has a sampling rate reaching 2 MSa/s, a data throughput rate exceeding 9.3 Gb/s, and a data transmission latency of less than 10 μs. [Conclusion] This data acquisition system facilitates the fast transmission of diagnostic data streams to disruption prediction models. The high sampling rates enable the system to perform real-time transmission of one-dimensional diagnostics such as radiation and electron temperature, improving the temporal resolution of data. The high data throughput rate can increase the transmission volume of diagnostic data, and the low data transmission latency can reduce the time required for disruption prediction models to obtain diagnostics data.

[Introduction] In recent years, as the installed capacity of the fusion device has been increasing, the power supply system of the device requires high-amplitude and long-pulse power output, which causes a strong power impact on the grid, and even power supply disconnection in severe cases to pose a serious threat to the safety of the fusion device. Additionally, the high-amplitude pulses also result in a lot of redundancy in the power supply system design. [Method] To solve these problems, this paper proposed a novel fusion power supply topology with energy storage, that is, the power supply system was designed with energy storage to mitigate the impact of pulse power on the grid. During the operation of the power supply, the grid provided steady-state power while the energy storage device delivered pulse power, effectively reducing the cost of the power supply. [Result] According to the experience in industrial power supply and distribution, this paper presents the calculation formulas for the selection of key components in different topologies and conducts a cost analysis and comparison for different power supply topologies based on simulation data. [Conclusion] The experimental results show that the new power supply topology not only reduces the power impact on the grid and improves the energy utilization of the power supply system, but also reduces transformer capacity, thus it is cost-efficient.

Plasma disruption in tokamak experiments is a challenging issue that causes damage to the device. Reliable prediction methods are needed, but the lack of full understanding of plasma disruption limits the effectiveness of physics-driven methods. Data-driven methods based on supervised learning are commonly used, and they rely on labelled training data. However, manual labelling of disruption precursors is a time-consuming and challenging task, as some precursors are difficult to accurately identify. The mainstream labelling methods assume that the precursor onset occurs at a fixed time before disruption, which leads to mislabeled samples and suboptimal prediction performance. In this paper, we present disruption prediction methods based on anomaly detection to address these issues, demonstrating good prediction performance on J-TEXT and EAST. By evaluating precursor onset times using different anomaly detection algorithms, it is found that labelling methods can be improved since the onset times of different shots are not necessarily the same. The study optimizes precursor labelling using the onset times inferred by the anomaly detection predictor and test the optimized labels on supervised learning disruption predictors. The results on J-TEXT and EAST show that the models trained on the optimized labels outperform those trained on fixed onset time labels.

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]

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]

We present an acceleration method for sequences of large-scale linear systems, such as the ones arising from the numerical solution of time-dependent partial differential equations coupled with algebraic constraints. We discuss different approaches to leverage the subspace containing the history of solutions computed at previous time steps in order to generate a good initial guess for the iterative solver. In particular, we propose a novel combination of reduced-order projection with randomized linear algebra techniques, which drastically reduces the number of iterations needed for convergence. We analyze the accuracy of the initial guess produced by the reduced-order projection when the coefficients of the linear system depend analytically on time. Extending extrapolation results by Demanet and Townsend to a vector-valued setting, we show that the accuracy improves rapidly as the size of the history increases, a theoretical result confirmed by our numerical observations. In particular, we apply the developed method to the simulation of plasma turbulence in the boundary of a fusion device, showing that the time needed for solving the linear systems is significantly reduced. [ABSTRACT FROM AUTHOR]

The accumulation of heavy impurities like tungsten in the plasma core of fusion devices can cause significant radiative power losses or even lead to a disruption. It is therefore crucial to monitor the tungsten impurity concentration. In this paper, we follow the integrated data analysis approach using Bayesian probability theory to jointly estimate tungsten concentration profiles and kinetic profiles from soft X-ray, interferometry and electron cyclotron emission measurements. As the full Bayesian inference using Markov chain Monte Carlo sampling is time-consuming, we also discuss emulation of the inference process using neural networks, with a view to real-time implementation. [ABSTRACT FROM AUTHOR]

In fusion, the validation of turbulent transport models is undertaken with the goals of making basic physics discoveries as well as for development of new predictive models to improve the operation and enhance the performance of existing and future fusion reactors. A fusion industry is just beginning to emerge globally. Like fission, validation in fusion energy research is a vibrant research area, but unlike fusion, a fission industry exists. The fission power industry motivates validation efforts, often performed at universities with small-scale experiments and advanced models and simulations developed in-house. Because fission research spans basic physics and applications, and addresses near-term and long-term industry interests, validation is thriving. This perspective article describes the validation of turbulent transport models in both fusion research and fission research, draws parallels between the validation methods and techniques used in two areas of the fields, and presents an outlook for thriving university fusion and fission research programs underpinned by a virtual cycle of basic and applied research that supports industry needs as well as tackling intellectual grand challenges. [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]

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]

Copyright of Southern Energy Construction is the property of Southern Energy Construction Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Southern Energy Construction is the property of Southern Energy Construction Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Southern Energy Construction is the property of Southern Energy Construction Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Southern Energy Construction is the property of Southern Energy Construction Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

The possibility of carrying out measurements of plasma parameters in a tokamak with reactor technologies (TRT) by means of the technique based on the Stark effect by resolving the spectrum of the split lines of Balmer series emitted by fast hydrogen atoms injected into plasma is analyzed. The code containing the models of emission applicable for the high-temperature tokamak plasma, along with the library of functions for the ray-tracing simulation of geometric optics, was used. Simulation of spectra of active neutral beam emission and plasma emission, both as a result of charge exchange on beam atoms and passive one, taking into account reflections from the plasma facing vessel elements, allowed determining parameters of the diagnostic injector and relative position of the light-collection systems and heating injectors for which the useful and background spectra could be separated. The signal gathered by the detector is simulated. The shape of the visual angle along the line of sight, optical properties of the lens materials, the instrumental function of the spectral device, the sensor characteristics, and statistical noise of the signal are taken into consideration. Based on the obtained results, a preliminary concept of the motional Stark effect (MSE) diagnostic for the tokamak with reactor technologies is proposed. [ABSTRACT FROM AUTHOR]

The paper presents a concept of the ECE diagnostic for the TRT facility and estimates the achievable measurement parameters in the baseline scenario. The target spectral region for the diagnostic corresponds to the first harmonic of the ECR frequency in ordinary polarization (O1) and the second harmonic in extraordinary polarization (X2). It is proposed to carry out measurements from the low-field side along two lines of sight: radial and toroidally oblique. The accessible spectral region in terms of the normalized radial coordinate is approximately estimated as –0.9 to 0.9 and –0.1 to 0.9. It is proposed to shape the input wave beam by means of a quasi-optical focusing system that provides a transverse size of the resolved region of approximately 3–5 cm for O1 and 1.2–3 cm for X2. For measurements, it is proposed to use Fourier transform spectrometers with a time resolution of about 10 ms and multichannel heterodyne receivers with a time resolution of about 1 μs. The minimum radial size of the resolved region is estimated to be 3–5 cm for O1 and 2–4 cm for X2, depending on the coordinate. [ABSTRACT FROM AUTHOR]

Calculations of the tungsten sputtering yields (the divertor material in the ITER tokamak) by He, Be, N, and O—impurity atoms in the plasma—were carried out at a collision energy of 0.010–100 keV using the Monte Carlo method. To calculate the trajectory of the incident particle, pair potentials obtained within the framework of density functional theory were used. These potentials were corrected for the parameters of the potential well obtained from spectroscopic measurements. The target consisted of tungsten randomly oriented crystals the size of one lattice constant. Next, the trajectories of the recoil particles were calculated using many-particle potentials calculated using density functional theory. Thermal vibrations of target atoms were taken into account. The vibration amplitude was taken to be 0.05 Å, which corresponded to room temperature. The strong dependence of the results on the shape of the surface potential barrier is shown, and the results are presented for two limiting cases of the surface state: a flat surface, when a planar surface potential barrier is realized, and a surface consisting of cones, when a spherical potential barrier occurs. In the experiment, the surface has some roughness, which depends on the experimental conditions. It is shown that the experimental results lie between the limiting cases we considered. Information was obtained on the average energy of sputtered atoms and angular distributions necessary for calculating the entry of impurities into the tokamak plasma. [ABSTRACT FROM AUTHOR]

This paper describes a new practical numerical model for the calculation of lateral electromagnetic (EM) loads in tokamaks during asymmetric vertical displacement events (AVDEs). The model combines key features of two recently reported trial models while avoiding their drawbacks. Their common basic feature is the superposition of two patterns of halo current: one perfectly symmetric and another perfectly anti-symmetric. This model combines the following features that have not been combined before (a) a helically distorted halo layer wrapping around core plasma, and (b) halo-to-wall interception belts slipping along plasma-facing walls. This combination almost doubles the lateral net forces. An AVDE creates significant lateral net moments. Being relatively modest at VDEs, the lateral moments become a dominant component of EM loads at AVDEs. The model carefully tracks the balance of net EM loads (zero total for the tokamak), as a necessary condition for the consequent numerical simulation of the tokamak’s dynamic response. This balance is needed as well for the development of tokamak monitoring algorithms and simulators. In order to decouple from the current uncertainties in the interpretation and simulation of AVDE physics, the model does not simulate AVDE evolution but uses it as an input assumption based on the existing interpretation and simulation of AVDE physics. This means the model is to be used in a manner of parametric study, at widely varied input assumptions on AVDE evolution and severity. Parametric results will fill a library of ready-for-use waveforms of asymmetric EM loads (distributed and total) at tokamak structures and coils, so that the physics community may point to specific variants for subsequent engineering analysis. This article presents the first practical contribution to this AVDE library.

Tokamaks are a very promising option to exploit nuclear fusion as a programmable and safe energy source. A very critical issue for the practical use of tokamaks consists of the power flow required to initiate and sustain the fusion process, in particular in the poloidal field coils. This flow can be managed by introducing a DC energy storage based on supercapacitors. Because such storage may be the most expensive and largest part of the poloidal power supply system, an excessive size would cancel its potential advantages. This paper presents innovative strategies to optimize the DC storage in poloidal power supply systems. The proposed solution involves the sharing of the DC storage between different coil circuits. The study is supported by novel analytical formulas and by a circuital model developed for this application. The obtained results show that this method and the related algorithms can noticeably reduce the overall size of the storage and the power exchange with the grid, providing a practical contribution toward the feasibility and the effectiveness of nuclear fusion systems.