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Concern about adequate tritium resources to start up deuterium-tritium (DT) fusion reactors in the near term is misplaced: Proposed demonstration reactors and pilot plants will not realistically be buildable for a great many decades. More disturbing, tritium fuel self-sufficiency of eventual DT reactors may be unattainable, because it is unlikely that tritium-breeding blankets can completely replenish burned-up fuel as well as unburned tritium that is lost in numerous reactor subsystems. The concept of doubling time is unworkable. The only viable fuel solution for sustained operation demands deuterium-based reactors where the tritium is concurrently generated and burned up in the fusioning plasma. [ABSTRACT FROM AUTHOR]

Centrifugal confinement fusion, a promising alternative to toroidal confinement devices like tokamaks and stellarators, leverages supersonic plasma rotation within a magnetic mirror configuration to achieve simplified coil design, compactness, and enhanced stability. This brief article explores the potential of centrifugal confinement fusion for propulsion applications. A previous concern regarding the escape of energetic ions, essential for propulsion, is addressed through test-particle simulations. The results indicate that the earlier estimate based on adiabatic invariance was overly pessimistic, and the underlying physics is clarified. [ABSTRACT FROM AUTHOR]

A molecular pump is a high vacuum acquisition piece of equipment that provides a clean vacuum environment for the Experimental Advanced Superconducting Tokamak (EAST) device. Its running state affects the smooth development of the EAST experiment. Because of fatigue degradation of internal components of the molecular pump, vacuum leakage may occur during long-term operation, causing secondary hazards to the device. In order to improve the accuracy of molecular pump fault prediction, based on the long short-term memory network (LSTM), the deep long short-term memory network (DE-LSTM) and the bidirectional long short-term memory network (Bi-LSTM) are combined. The deep bidirectional long short-term memory network (DE-Bi-LSTM) algorithm is proposed, and the piecewise linear degradation model is introduced to predict fault of the molecular pump. By collecting the vibration signals leaked in the atmosphere and running to the fault time series on the destructive test platform simulating molecular pump fault, data were extracted in the time domain. Finally, the obtained feature vector set was used as the input of the DE-Bi-LSTM algorithm through data standardization to train the model and realize the prediction of molecular pump fault. The experimental results show that the proposed method is optimal to LSTM, DE-LSTM, and Bi-LSTM in predicting performance. [ABSTRACT FROM AUTHOR]

Runaway electrons (REs) generated during disruptions pose a significant safety threat to tokamaks, as they can melt and damage the plasma-facing components (PFCs). Therefore, studying RE behavior is crucial for fusion devices. The interaction between REs and the first wall/PFCs results in the emission of high-energy X-rays, known as bremsstrahlung. To investigate RE behavior, it is necessary to quantitatively evaluate the emission of hard X-rays. A real-time hard X-ray spectrometer, utilizing a LaBr3 detector, has been successfully developed for studying REs on the HL-2M tokamak. This spectrometer has a counting rate capability reaching 3 MHz, with an energy resolution of 3.3% at 662 keV (137Cs). The time resolution for energy spectrums is as short as 1 ms. During the HL-2M discharge, observations were made on the hard X-ray energy spectrum, and by analyzing the spectrum within the energy range of 250 keV to 750 keV, the temperature of the corresponding REs was deduced. [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]

Parameter space for spherical tokamak reactors is explored quantitatively to elucidate the main constraints for spherical tokamak design choices. Using a constant plasma current Ip search constraint, a set of four Ip scenarios (5, 10, 15, and 20 MA) is first explored in a wide parameter space. Considering modest but gradually increasing auxiliary power, a set of four machine configurations (major radius $$R$$ R = 1.25, 1.75, 2.25, and 3.5 m) is explored next, optimizing the Ip and the bootstrap fraction. Constraints that narrow down the vast parameter space are elaborated along with critical assumptions, such as current drive efficiency, H-mode enhancement factor, nuclear shielding efficiency, and confinement scaling. Limits on the current density of the center post and how it affects the shielding are quantitatively indicated, thereby setting a lower limit on the aspect ratio. [ABSTRACT FROM AUTHOR]

The flight simulator predicts the dynamic behavior of a full plasma discharge (described in terms of one-dimensional profiles) by employing multiple control loops based on synthetic diagnostics, which could also emulate realistic sensor and actuator models. It serves as a valuable tool for designing and optimizing plasma scenarios, as well as for assessing the feasibility of controlling discharges. The Fenix flight simulator, originally developed for the ASDEX Upgrade, has been ported to EU-DEMO and is capable of modeling any tokamak. One of the essential elements in a flight simulator is the link between the co-simulated plasma physics and the control loops. This element is tightly coupled to the specifications of both the plasma model and the control algorithms to be implemented; but on the other hand, to ensure the portability and applicability of the flight simulator to different scenarios or devices, the coupling between plasma and control algorithms should be neutral to any concrete device and configuration. In addition, as a serial component of the control loop, data exchange takes place at every single step of the control simulation, therefore an efficient implementation is critical for the overall simulation performance. This paper summarizes the universal approach recently implemented in Fenix, which satisfies all the above requirements while remaining lightweight. [ABSTRACT FROM AUTHOR]

As a significant input for neutronics analysis of tokamak devices, the plasma neutron source is a bridge connecting fusion plasma physics and engineering design. A plasma neutron source model is established based on the EAST plasma configuration. The maximum neutron wall loading (NWL) is near the outboard midplane and is more than ~20% that of the inboard midplane. The investigations demonstrate that special care should be taken for radiation shielding and protection of the key components located on the outboard midplane and on the inboard midplane. The effect of the tritium accumulation on neutronics analysis should be carefully considered for EAST tokamak D-D plasma operations. The effect of density peaking (DP) on the neutronics analysis is also investigated. It is evident that the peak NWLs are all near the outboard midplane and that the poloidal distributions of the NWL are slightly different for these cases. With increasing DP, both the outboard and inboard peak NWLs decrease. However, the decrease in NWL is very small; NWL decreases only 11% when the neutron source peak increases about 1.5 times. [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]

Tokamaks are often considered to be a leading candidate for near-term, cost-effective fusion energy, but these devices are susceptible to sudden loss of confinement events called disruptions. The threat of disruptions has garnered serious attention in research for the next generation of burning plasma experiments, such as ITER, but has received little treatment in economic studies of magnetic fusion energy. In this paper, we present a model for quantifying the effect of disruptions on the cost of electricity produced by a tokamak power plant (TPP). We outline the various ways disruptions increase costs and decrease revenues, introduce metrics to quantify these effects, and add them to a levelized cost of electricity (LCOE) model. Additionally, we identify several rate-limiting repair steps and introduce a classification system of disruption types based on the time to return to operations. We demonstrate how the LCOE model can be used to find the cost of electricity and the requirements for disruption handling of a TPP, and we further highlight where future research can have a strong impact in neutralizing the "showstopping" potential of disruptions. [ABSTRACT FROM AUTHOR]

Nuclear fusion depends on tritium breeding and self-sufficiency. Tritium represents a hazard due to its radioactivity and migration properties. Because of these difficulties, ITER, the largest fusion experiment so far, relies on a conservative static procedure to monitor the tritium inventory. Future commercial fusion plants can avoid operation halts if a dynamic monitoring strategy proves itself valid. Tritium plant models have been developed for this kind of monitoring and analysis task, but sensor accuracy and reliability are an issue still to be addressed, and the path to dynamic monitoring remains unclear. The present work shows the modeling procedure of the Tokamak Exhaust Processing system in a commercial simulator, Aspen HYSYS, to reproduce the inventories, streams, process conditions, and compositions of this subsystem during operation. The model is verified in a steady-state scenario using data from the available literature. A demonstration of such a tritium plant subsystem shows meaningful value for several reasons. First, this process has not been modeled before in commercial dynamic simulators, which are typically used in the process industry. It will also allow new stakeholders to participate in future fusion-related projects. Second, it will play a key role in industry-like tritium process monitoring, in which the new model will act as a digital twin of the plant. Data-driven diagnostics can be fueled by model data, helping engineers to generate additional data that could otherwise be expensive to get directly from the plant. For these reasons, models will represent an essential part of a dynamic monitoring system, necessary for feasible fusion projects. [ABSTRACT FROM AUTHOR]

A micro-tearing mode has been observed in electron temperature fluctuations and magnetic fluctuations with the frequency range of 80–120 kHz in the pedestal of the Experimental Advanced Superconducting Tokamak (EAST) using Electron Cyclotron Emission Imaging (ECEI) and Mirnov magnetic coil measurement. It appears in the type-I ELMs period after the L-H transition. The measured poloidal phase velocity of the electron temperature fluctuations is ∼40 km/s (f = 80–120 kHz) in the electron diamagnetic direction in the laboratory frame with the poloidal wave number around kΘ∼0.15 /cm. The two-dimensional structure of MTMs is directly observed using two-dimensional Electron cyclotron emission imaging. The experiment confirmed that the electron temperature (ECEI) and radial electric field (Backward Doppler reflectometer) during MTM meet the pressure balance equation. Until now, the theory and simulation results showed that transport is in heat flux, small particle flux, while the experimental results of this paper indicate that the particle flux is small. [ABSTRACT FROM AUTHOR]

The Tokamak à Configuration variable (TCV) is equipped with two neutral beam injection (NBI) systems delivering up to 1.2 MW each for pulse durations of up to 2 s. The first system (NBI1), designed for an injection energy in the range of 25 to 30 keV has been operational since 2016. The existing concrete neutron shielding of the experimental hall proved insufficient for fully protecting human accessible areas, limiting the number of daily plasma pulses using NBI1. The recently commissioned second system (NBI2) is designed for injection synergies in the range 50 to 60 keV. Both systems are tangentially oriented in opposite directions in order to permit experiments with low or no net torque. Calculations with the TRANSP and ORBIS heating codes show that neutron rates from deuterium-deuterium fusion reactions may be as high as 1014 n/s, up to 10 times higher than with the lower energy beam only. This is due both to the ~five times larger beam-plasma neutron rates from the higher energy beam and to an exceptionally high contribution from beam-beam reactions between the opposing beams. The radiation protection policy at the Swiss Plasma Center is that all staff members be considered as members of the general public, limiting the daily personal dose to 4 µSv. This is also the maximum admissible daily dose in any publicly accessible zone, whether occupied or not. Currently, with only the lower energy beam, this limit can be attained in the control room adjacent to the device hall after only five NBI pulses out of a possible 30 daily pulses. To allow for exploitation of the two beams at full specifications, the source side of the existing barite concrete walls of the 15 × 20 × 8 m large TCV hall will be covered with 20-cm-thick polythene (PE) cladding and a ceiling made of 35-cm-thick PE will be added. The total mass of PE will be 200 tons. The usage of PE at this scale for neutron shielding is unprecedented at any fusion research facility. [ABSTRACT FROM AUTHOR]

Behaviors of runaway electrons have been investigated by injection of argon impurity gas during the low-density hydrogen discharges in the STOR-M tokamak. Argon gas is puffed using a fast electromagnetic gas puffing valve with various amounts of gas puffed at different injection times. It has been found that puffing of an appropriate amount of gas can suppress the already generated runaway electrons evidenced by the intensity reduction of the hard X-ray (HXR) radiation caused by the impact of runaway electrons on the STOR-M chamber wall. On the other hand, excessive argon gas puffing causes earlier termination of the discharge. [ABSTRACT FROM AUTHOR]

Core particle fueling, an essential task in the European demonstration fusion power plant EU-DEMO, relies on adequate pellet injection. However, pellets are fragile objects, and their delivery efficiency can hardly be assumed to be unity. Exploring kinetic control of the EU-DEMO1 scenario indicates that such missed-out pellets do cause a considerable problem for keeping a burning plasma. Missed-out pellets can cause a severe drop of plasma density that in turn results in a potential drastic loss of burn power. Efforts are under way at the ASDEX Upgrade (AUG) tokamak aiming to provide real-time monitoring of pellet arrival and announcement of missed-out cases to the control systems. To further optimize the controllers, system identification experiments have been performed to identify the dynamic response of the system to the actuators. [ABSTRACT FROM AUTHOR]

The HL-2M device is a new generation tokamak at the Southwestern Institute of Physics that is used to develop nuclear fusion research. To confine the plasma, the toroidal field (TF) coil of HL-2M is needed to output 140-kA direct current to generate a 2.2-T magnetic density flux. Therefore, the TF power supply system was developed. It is powered by two flywheel pulsed motor-generator sets, and a cophase counter parallel connection bridge rectifier was used to provide more current. According to the TF power supply system topology, the amplitude of the magnetic flux density of the TF coil can be tuned by adjusting the excitation system of the flywheel pulse generator; the control strategy and mathematical formula are given. Then, the TF power supply protection system is established, and protection strategies under different faults are designed. The control and protection performance of the TF power supply system has been successfully applied and verified in the first plasma discharge of the HL-2M experiment. [ABSTRACT FROM AUTHOR]

The vacuum vessel (VV) is one of the main components of the HL-2M tokamak device. It is a thin-shell, all-welded torus made of Inconel 625 material. Maintaining the welding quality and controlling welding distortion are the main manufacturing difficulties. Welding, forming, and assembly are the research focus. This technical note describes the fundamental manufacturing research and development (R&D) and the actual manufacturing procedure of the HL-2M VV. Three stages of R&D work are carried out. The weldability of Inconel 625 plate is studied. Partial and full-scale mock-ups are made. By optimizing the manufacturing process, the HL-2M VV was completed in 2019, and the measurement data meet the design requirements. HL-2M has good vacuum performance, the total leak rate reaches 6.0 × 10−8 Pa·m3/s at room temperature, and the ultimate vacuum pressure reaches 2.4 × 10−6 Pa after baking up to 100°C. [ABSTRACT FROM AUTHOR]

The well-known stability of the drift wave in a sheared slab geometry does not hold in the presence of dust particles. Due to the presence of dust particles, the magnetic shear damping is reduced drastically. As a result, collisionless drift modes become unstable under typical parameter regimes of tokamak. Consequently, drift wave must still be considered as an underlying dynamic of anomalous transport in tokamak edges, where dust particles are found to be abundant, the same physics is expected in the spacial environments as well. [ABSTRACT FROM AUTHOR]

Controlled tungsten (W) dust injection experiments have been performed in the STOR-M tokamak. Tokamak discharges were initiated after a known delay time with respect to the injection of W micro-spheres. The plasma parameters were detrimentally affected, proportional to an increased amount of dust within the plasma volume. High speed camera imaging of the injected dust revealed that the dust particles experience a strong force in the counter-plasma current direction. This force decreased for high dust number density due to the associated decline of plasma parameters. [ABSTRACT FROM AUTHOR]

Injection of a hydrogen neutral beam (H) is studied into the boron-11 (11B) fusion tokamak plasma. The conclusions show that the most desirable injection energy of the hydrogen beam, E H , into pure 11B target plasma is 5 MeV for performing the H11B reaction with the maximum fusion energy multiplication factor ( ζ H 11 B ). This factor for E H > 5 M e V decreases with the increase in E H and increases with the electron temperature. A change in the thermal ion mixture had only a small effect on ζ H 11 B . ζ H 11 B decreases when the effect of radiation losses is considered through bremsstrahlung P B and synchrotron powers P S . The difference between the curves of P B and the fusion power P f is minimized in the ion temperature range 300–400 keV. A higher ion-to-electron temperature ratio is required to obtain higher P f P B . P f P S increases with an increase in plasma beta. Increasing the ratio of the fusion power to each radiation power can enhance the ζ H 11 B value. [ABSTRACT FROM AUTHOR]

An H-bridge power supply has been developed for different experiments on the STOR-M tokamak. Four insulated-gate bipolar transistors (IGBTs) are used to guide the current to the load from a capacitor bank discharge. The timing signal to the gates of the IGBTs is controlled by a field programmable gate array (FPGA). The power supply can be programmed to drive either a sinusoidal current waveform to an inductive load or a bipolar rectangular current waveform to a resistive load. The maximum frequency and current tested for the sinusoidal current waveform are 25 kHz and 2200 A, respectively. For bipolar current waveforms, the maximum current tested is 1500 A at 3 kHz. The sinusoidal current to a set of helical coils was used for the Resonant Magnetic Perturbation (RMP) experiments and the bipolar rectangular current waveform was used for both the electrode biasing experiments and the RMP experiments on the STOR-M tokamak. [ABSTRACT FROM AUTHOR]

This article presents the design and fabrication of 20 ± 0.3 dB directional coupler having 2 kW power handling capability in the frequency range of 140–200 MHz. The developed coupler has an important application in the prototype of the Ion Cyclotron Resonance Frequency (ICRF) system of Tokamak. The directional coupler used to couple the forward and reflected power for measurement and provide the feedback to the existing matching network responsible for protecting the high-power RF source. The developed coupler is compact in size, robust, and capable of handling 2 kW, continuous wave (CW) RF power. This paper also analyses the dielectric breakdown strength using electric field distribution inside the coupler, and the average power handling capability is determined using numerical analysis. The test results using Vector Network Analyser (VNA) are obtained as, the coupling flatness of 20 ± 0.3 in 140–200 MHz frequency range, return loss and isolation are −30 dB and −35 dB, respectively. The work presented in this paper is helpful for the development of high-power coupler in radio frequency (RF) range and may be applied in various applications such as RF Plasma, Radar, navigation, satellite, and research, etc. [ABSTRACT FROM AUTHOR]

A dust dispenser has been designed and characterized for future experiments in the STOR-M tokamak. Controlled dust injection is demonstrated in a test-bench device and characterized with different techniques. Upon activation of the dispenser, a plume of dust particles is dispensed. The particle size distribution is found using scanning electron emission images. A laser is used to illuminate the falling particles dispensed and a fast video camera is used to track the images of those particles. The evolution of the spatial distribution of the falling particles is reconstructed using those images based on a free-fall model. The total mass of the dispensed dust particles is directly measured using an analytic balance and is also inferred from the intensity of the scattered light recorded by the video camera. [ABSTRACT FROM AUTHOR]

This paper presents the design and simulation of a novel TE10 to TE30 mode converter operating at 3.7 GHz and is subjected to high RF power of 500 kW in Continuous Wave (CW) regime. The design incorporates a stepped impedance transformer as opposed to the conventional capacitive button to improve the return loss and power handling capacity of the converter. A rigorous RF, breakdown and multiphysics analysis including thermal, fluid dynamics and stress physics is performed on the mode converter to ascertain its viability and usefulness as a power dividing network in the Passive Active Multijunction launcher for Lower Hybrid Current Drive systems of tokamaks. The proposed mode converter exhibits a return loss of ∼50 dB, mode conversion efficiency of ∼97% and a peak E-field of 685 kV/m is obtained as compared to that of 4.2 MV/m obtained in the conventional mode converter. Thus, the power handling capacity of the proposed mode converter design is found to be 97.3% higher than the conventional mode converter. [ABSTRACT FROM AUTHOR]

Today, seven members—the European Union, the United States, Japan, China, Russia, South Korea, and India—have formed an organization and invested in what is planned to be the world’s first reactor-scale fusion machine. By 2027, the France-based, multibillion-euro experiment known as ITER aims to produce 10 times the energy it consumes. In this interview, Osamu Motojima, the project’s director general, discusses the skepticism surrounding fusion as a practical energy source, the staggering costs of a large-scale fusion project, and the likelihood that ITER will be the silver bullet that slays the world’s future energy woes. [ABSTRACT FROM PUBLISHER]

Vertical Disruption Events (VDE) in tokamaks, may cause significant melting along tile ridges of the vessel upper dump plates with release of metallic particulate and droplets as shown in and in references therein. Accurate studies of the motion of such debris released in the vessel are important for the inventory of tritiated composites in a reactor. Recently numerical codes have been developed to track this particulate of variable mass and charge, in interaction with the Scrape-Off layer (SOL) plasma. Here a contribution is given for a more consistent discussion of the additional acceleration of the dust due to asymmetric loss of mass in flight. A simplified description is constructed in analogy with the physics of comets. [ABSTRACT FROM AUTHOR]

The radiative rates and charge states of atoms in a high-temperature ( keV) plasma are of great interest for the research of the controlled fusion. In the current paper, new compact polynomial fits describing the coronal radiative cooling rates and the mean impurity charge are obtained for 11 elements: helium, lithium, beryllium, carbon, nitrogen, oxygen, neon, argon, krypton, xeon and tungsten. In the modern tokamak experiments, these elements are the most relevant impurity spices. In contrast to the previous works, the new fits are based on the most recent and complete atomic database Atomic Data and Analysis Structure. The manuscript presents coefficients for least-squares polynomial fits and compares them with the earlier known approximating formulae. These results can be extremely useful for system codes of fusion power plants, since they provide quick estimation of the radiation losses in the plasma. [ABSTRACT FROM AUTHOR]

The simple map and the symmetric quartic map are the generic representation of the open-unbounded and closed-compact magnetic topologies for single-null divertor tokamaks, respectively. The map parameter represents the magnetic asymmetries. The two maps are used to calculate the homoclinic tangles in the two topologies in divertor tokamaks. It is found that as the magnetic asymmetries become larger, the homoclinic tangle of the simple map becomes extremely elongated radially as opposed to the homoclinic tangle of the symmetric quartic map which becomes wider poloidally. [ABSTRACT FROM AUTHOR]

The stability of a magnetic island in a toroidal magnetic confinement device depends on various factors besides the usual tearing-mode stability parameter , determined by the local current profile.The presence of a sheared flow in the vicinity of the rational surface that supports the island is one of the factors that affects its stability since it can give rise to a polarization current around the island position. The contribution of the polarization current to the stability has been computed for a tokamak geometry. Here, we consider the case of magnetic islands with a shear flow in a stellarator which has a non-axisymmetric magnetic geometry. The main difference is a contribution to the polarization current from the toroidal electrostatic oscillation. A correction due to the global toroidal magnetic geometry is also present. It is found that the regime where the stability is affected corresponds to the large island width relative to the ion gyroradius. Thus, the contribution is relevant for low-temperature regimes. In that case, the polarization current is destabilizing for frequencies larger than the ion diamagnetic frequency. Our results imply that the sheared flow can produce a growth of the magnetic island in a cold plasma but it can become narrower as the temperature rises. [ABSTRACT FROM AUTHOR]

Edge plasma properties have been modified in the Saskatchewan Torus-Modified tokamak by means of resonant magnetic perturbations (RMP). It has been found that the radial profiles of ion saturation current and floating potential in the edge region can be modified by an externally applied static (m = 2, n = 1) RMP field. An increase in the pedestal plasma density (n) and more negative electric field (Er) have been observed in the plasma edge region. It is believed that the RMP field altered the plasma transport in the edge and scrape-off layer regions, leading to a higher density pedestal and a potential drop in some cases. During the enhanced confinement phase, it is possible to identify a region where intermittent transport events, the so-called blobs, are created and the holes of lower density left behind. [ABSTRACT FROM AUTHOR]

Extensive studies are being carried out to develop a divertor (Divertor: a device in a thermonuclear reactor, intended for receiving of particle flows and radiation from the periphery of the plasma column. It is the most heavily loaded part of the reactor chamber surfaces with energy fluxes.) for tokamaks with a container for supplying liquid metal to the reception surface, with the construction of actual structures with high reliability and long service life. In this case, it is necessary to develop methods for joining dissimilar materials. In the transition section, molybdenum and steel (12Cr18Ni10Ti) pipes are joined by brazing. In this work, the method of producing a strong, heat resisting and hermetic joint between molybdenum and 12Cr18Ni10Ti corrosion-resisting steel is developed. Brazing is carried out using a paste based on CTEMET-1301 brazing alloy. The method of assembling the components in order to produce optimum gaps in brazing resulting in the homogeneous brazed joint without the formation of a brittle eutectic interlayer is developed. The brazed specimens were tested for heat resistance, leak tightness and service life at the given load. All the specimens withstood the tests. The mechanical properties of the brazed joints were determined by measuring the microhardness of the region of the welded joint and by shear tests. The results show that the strength characteristics depend strongly on the size of the gap between the brazed components. For the specimens assembled with the optimum gap the ultimatum shear strength was 500 ± 50 MPa. [ABSTRACT FROM PUBLISHER]

Electron cyclotron emission imaging (ECEI), as an advanced diagnostic, provides two-dimensional (2D) images of electron temperature fluctuation in the core plasma region. The ECEI system on Experimental Advanced Superconducting Tokamak (EAST) tokamak consists of 384 channels by 24 (vertical) × 16 (horizontal). The system nearly covers the wholesurface in cross-section of EAST. Here, we present a quantitative method to estimate the sawtooth collapse time. All of the electron temperature fluctuations inside thesurface are summed up, and the collapse time is estimated as the time lag between the maximum and minimum values of the total fluctuation during one sawtooth cycle. Using this method, statistical analysis of sawtooth behaviors is studied in the EAST Ohmic heating plasma. Relationships of heat flux, sawtooth period and collapse time are also presented. [ABSTRACT FROM PUBLISHER]

Trajectories of magnetic field lines are a 1½ degree of freedom Hamiltonian system. The unperturbed separatrix and the perturbed separatrix in a divertor tokamak are fundamentally different. Magnetic asymmetries cause the separatrix to form extremely complicated structures known as homoclinic tangles. After each toroidal circuit, the perturbed separatrix manifolds meet in a fixed poloidal plane and intersect to form homoclinic tangle in order to preserve the topological invariants. This tangle becomes extremely complicated as the magnetic field lines take more and more toroidal turns. This effect is most pronounced near the X-point. The homoclinic tangles of the DIII-D tokamak separatrix from the magnetic perturbation representing the peeling-ballooning modes are studied. The homoclinic tangles can have important implications for the edge physics in divertor tokamaks. [ABSTRACT FROM PUBLISHER]

It is very important to calculate the equilibrium magnetic field configuration exactly and to estimate the parameters of the device needed for the equilibrium and stability in the design and operate the low-aspect-ratio tokamak, one of the axis-symmetrical torus devices which is widely used in the research of nuclear fusion using plasma. In the previous researches, there were the theoretical and numerical interpretation methods for the high-aspect-ratio tokamak plasma. But for the low-aspect-ratio tokamak plasma, numerical methods such as the finite element method and iteration method are usually used. In this research, we interpreted and examined the features of magnetohydrodynamic equilibrium of the low-aspect-ratio tokamak plasma in combination with the Green function method and finite element method which has high numerical value accuracy. [ABSTRACT FROM PUBLISHER]

The power threshold for low (L) to high (H) confinement mode transition achieved by radio-frequency heating and molybdenum first wall with lithium coating has been experimentally investigated on the EAST tokamak for two sets of divertor geometries and materials: tungsten/carbon divertor and full carbon divertor. For both sets of divertors, the power threshold was found to decrease with gradual accumulation of the lithium wall coating, suggesting the important role played by the low Z impurities and/or the edge neutral density on the L–H power threshold. When operating in the upper single null configuration, with the ion grad-Bdrift direction away from the primary X-point, a lower normalized power threshold is observed in EAST with the tungsten/carbon divertor, compared to the carbon divertor after intensive lithium wall coating. A newly installed cryopump increasing the pumping efficiency also plays an important part in the observed lower threshold. In addition, the H-mode in the Quasi-Snowflake divertor configuration has been obtained on EAST, exhibiting higher L–H power threshold compared to the lower single null configuration with similarIP/BTpairs. [ABSTRACT FROM PUBLISHER]

A thermo-magneto-hydrodynamics port-Hamiltonian model is derived for the plasmas in tokamaks. Electromagnetic field and material domain balance equations are expressed in covariant forms, together with the magneto-hydrodynamics interconnection structure connecting them together. The balance equations for the entropy, mass and momentum, as well as closure equations in the material domain, are derived from the Boltzmann equation (kinetic theory). The Gibbs–Duhem equation is used to compute the irreversible entropy source term and to define the interdomainℜ-field of the model. All derived interdomain couplings in the material domain are represented using Dirac and Stokes–Dirac structures and the resistivityℜ-field structure. The complete model is summarized in a Bond Graph. [ABSTRACT FROM PUBLISHER]

Nuclear fusion offers the potential for being a near limitless energy source by fusing together deuterium and tritium nuclei to form helium inside a plasma burning at 100 million K. However, scientific and engineering challenges remain. This paper describes how such a plasma can be confined on Earth, and discusses the similarities and differences with fusion in stars. It focuses on the magnetic confinement technique and, in particular, the method used in a tokamak. The confinement achieved in the equilibrium state is reviewed and it is shown how the confinement can be too good, leading to explosive instabilities at the plasma edge called edge localised modes (ELMs). It is shown how the impact of ELMs can be minimised by the application of magnetic perturbations and discusses the physics behind the penetration of these perturbations into what is ideally a perfect conducting plasma. [ABSTRACT FROM AUTHOR]

The biggest challenge of designing and building tokamak electron cyclotron resonance heating (ECRH) pulse step modulation (PSM) power supply is satisfying its required output voltage rising time to be less than 100 µs while suppressing the voltage overshoot to be no more than 1%. To fulfill the two requirements, a new control strategy with startup time in microsecond range is proposed in this paper, and a new control system to realize the control strategy is introduced. The control system was built and tested on 60 kV/50 A ECRH power supply. The experimental results indicate that the control system can restrain the overshoot effectively, increase response speed, and obviously improve the dynamic characteristics of the PSM power supply system. Thus, the proposed control system helps the PSM power supply to meet the design specifications. [ABSTRACT FROM PUBLISHER]

A horizontal water storage tank was analyzed for seismic shaking at the ITER Tokamak Complex in France. The objectives were to (a) estimate the seismic forces in the tank;(b) calculate the sloshing response of the tank; (c) determine if baffles are needed to control sloshing; and (d) evaluate the possibility of using a single fixed support in the longitudinal direction to allow free thermal expansion of the tank. An approximate conservative analysis predicted very high sloshing waves and seismic forces in the tank. The fluid-structure interaction in the tank showed that only about 28% of the liquid moves with the tank wall and generates seismic forces in the longitudinal direction. The remaining 72% of the liquid sloshes near the free surface and does not generate significant seismic forces. The sloshing wave is nat high enough in the longitudinal direction as the fundamental sloshing mode is not excited because of its very low natural frequency. Hence, baffles are not needed to control sloshing. The seismic force is low enough for a single fixed support to resist the entire seismic force in the longitudinal direction. [ABSTRACT FROM AUTHOR]

In this contribution, we have presented two techniques for the determination of plasma equilibrium position in IR-T1 tokamak: relaxation and optical methods. An analysis method of tokamak plasma equilibrium by a relaxation method with a specified magnetic axis is presented. The degrees of freedom due to designated positions of the magnetic axis are possible by using poloidal field coil currents. Stable steady-state tokamak plasma equilibria are calculated along with the magnetohydrodynamic potential energy. The plasma generates a plasma current which partially or fully cancels the magnetic field from the poloidal field coils. For low-temperature plasmas, the plasma current distribution is centrally peaked; for high-temperature plasmas, the plasma current has a hole. A centrally peaked current distribution in a low-temperature plasma is evolved into a current distribution with a hole by increasing the plasma pressure by Ohmic heating, radio frequency heating, or by neutral beam injection heating. In the second technique, an image-processing technique was used for the output signal of the charge coupled device camera and plasma emission intensity profile and then the plasma position was obtained. Results are compared and discussed. [ABSTRACT FROM AUTHOR]

The conceptual design of the subcritical advanced burner reactor (SABR), a 3000-MW(thermal) annular, modular sodium pool-type fast reactor, fueled by metallic transuranic (TRU) fuel processed from discharged light water reactor fuel and driven by a tokamak D-T fusion neutron source based on ITER physics and technology, has been substantially upgraded. Several issues related to the integration of fission and fusion technologies have been addressed, e.g., refueling a modular sodium pool reactor located within the magnetic coil configuration of a tokamak, achieving long-burn quasi-steady-state plasma operation, access for heating and current drive power transmission to a toroidal plasma surrounded by a sodium pool fast reactor, suppression of magnetohydro-dynamic effects in a liquid metal coolant flowing in a magnetic field, tritium self-sufficiency in a TRU transmutation reactor, shielding the superconducting magnets from fusion and fission neutrons, etc. A design concept for a SABR that could be deployed within 25 years, based on the IFRIPRISM metal-fuel, sodium pool fast reactor technology and on the ITER fusion physics and technology, is presented. This design concept can be used for realistic fuel cycle, dynamic safety, and other performance analyses of a SABR. [ABSTRACT FROM AUTHOR]

Intrinsic rotation has been observed in lower hybrid current-driven (LHCD) H-mode plasmas with type-III edge-localized modes (ELMs) on Experimental Advanced Superconducting Tokamak (EAST), and it is found that the edge toroidal rotation accelerated before the onset of the ELM burst. Magnetic perturbation analysis shows there is a perturbation amplitude growth below 30 kHz corresponding to the edge rotation acceleration. Using the filament model, the neoclassical toroidal viscosity (NTV) code shows there is a co-current NTV torque at the edge, which may be responsible for the edge rotation acceleration. For maximum displacement ∼1 cm and toroidal mode numbern=15, the calculated torque density is ∼0.44 N/m2, comparable with the average edge toroidal angular momentum change rate ∼1.24 N/m2. Here, the 1 cm maximum magnetic surface displacement estimated from the experimental observation corresponds to a maximum magnetic perturbation ∼ 10−3–10−2 T, in accordance with magnetic perturbation measurements during ELMs. By varyingnfrom 10 to 20, the magnitude of the edge NTV torque density is mainly ∼0.1–1 N/m2. This significant co-current torque indicates that the NTV theory may be important in rotation problems during ELMs in H-mode plasmas. To better illuminate the problem, magnetic surface deformation obtained from other codes is desired for a more accurate calculation. [ABSTRACT FROM AUTHOR]

A hot limiter biasing system with a simplified fast switch circuit was designed, constructed, and installed on the IR-T1 tokamak, and then the negative voltage applied to a hot limiter inserted inside the tokamak fixed limiter and the plasma current, poloidal, and radial components of the magnetic fields, loop voltage, diamagnetic flux, and the ion saturation currents in the absence and presence of the biased limiter were measured. Results of measurements of biasing effects on the plasma equilibrium behavior and edge plasma rotation are compared and discussed. [ABSTRACT FROM PUBLISHER]

The high-energy current of runaway electrons during a major disruption in tokamak reactors can cause serious damage to the first wall of the reactors and reduce their life time. Therefore, finding a method to minimize runaway electron generation during a major disruption is much needed. Tokamak limiter biasing is one of the methods that can be used for controlling the radial electric field and can induce a transition to an improved confinement state. This paper attempts to examine the effect of limiter biasing on the generation of runaway electrons during a major disruption. To do so, a horizontal biased limiter placed on the tokamak was used. Main parameters such as plasma current, loop voltage, emitted hard X-ray intensity, magnetohydrodynamic (MHD) oscillation andHαradiation and spectrum of hard X-rays, in the presence and absence of negative and positive limiter biasing, were measured. The results show that the application of limiter biasing during a major disruption can reduce runaway electron generation. [ABSTRACT FROM PUBLISHER]

Limiters play a number of roles in the tokamak operation. It serves primarily to protect the wall from the plasma when there are disruptions, runaway electrons, or other instabilities and also the limiters localize the plasma–surface interaction. In this research, we presented the first results of movable limiter experiments and its effects on the tokamak plasma confinement. For this purpose, we designed, constructed, and installed a movable localized poloidal limiter, and then measured the effects of limiter position on the time intervals of plasma parameters such as plasma density, temperature, and energy confinement time. The results of effects of the movable limiter experiments on plasma confinement. [ABSTRACT FROM PUBLISHER]

Internal components of magnetic confinement fusion machines are subjected to significant heat fluxes. A large part of this power is directed towards plasma facing components. Even if these components are designed to receive about, surface temperature and heat flux measurements are important issues to guarantee safe plasma operations. In JET tokamak, few embedded thermocouples (TC) located 1 cm below the tile surface are used to measure the bulk temperatures of the Carbon Fiber tiles (coated with about 20m of tungsten with the ITER-like wall). We propose here to use an inverse thermal calculation based on Thermal Quadrupole method to locally deduce the deposited heat flux. The calculation requires the location of the peak and the normalized 1D-shape of the heat flux deposited on the target. [ABSTRACT FROM AUTHOR]

Magnetic perturbations in tokamaks lead to the formation of magnetic islands, chaotic field lines, and the destruction of flux surfaces. Controlling or reducing transport along chaotic field lines is a key challenge in magnetically confined fusion plasmas. A local control method was proposed by Chandre et al. [Nucl. Fusion46, 33–45 (2006)] to build barriers to magnetic field line diffusion by addition of a small second-order control term localized in the phase space to the field line Hamiltonian. Formation and existence of such magnetic barriers in Ohmically heated tokamaks (OHT), ASDEX UG and piecewise analytic DIII-D [Luxon, J.L.; Davis, L.E.,Fusion Technol.8, 441 (1985)] plasma equilibria was predicted by the authors [Ali, H.; Punjabi, A.,Plasma Phys. Control. Fusion49, 1565–1582 (2007)]. Very recently, this prediction for the DIII-D has been corroborated [Volpe, F.A., et al.,Nucl. Fusion52, 054017 (2012)] by field-line tracing calculations, using experimentally constrained Equilibrium Fit (EFIT) [Lao, et al.,Nucl. Fusion25, 1611 (1985)] DIII-D equilibria perturbed to include the vacuum field from the internal coils utilized in the experiments. This second-order approach is applied to the DIII-D tokamak to build noble irrational magnetic barriers inside the chaos created by the locked resonant magnetic perturbations (RMPs) (m, n)=(3, 1)+(4, 1), withmandnthe poloidal and toroidal mode numbers of the Fourier expansion of the magnetic perturbation with amplitude δ. A piecewise, analytic, accurate, axisymmetric generating function for the trajectories of magnetic field lines in the DIII-D is constructed in magnetic coordinates from the experimental EFIT Grad-Shafranov solver [Lao, L, et al.,Fusion Sci. Technol.48, 968 (2005)] for the shot 115,467 at 3000 ms in the DIII-D. A symplectic mathematical map is used to integrate field lines in the DIII-D. A numerical algorithm [Ali, H., et al.,Radiat. Eff. Def. Solids Inc. Plasma Sc. Plasma Tech.165, 83 (2010)] based on continued fraction decomposition of the rotational transform labeling the barriers for selecting and identifying the strongest noble irrational barrier is used. The results are compared and contrasted with our previous results on the ASDEX UG. About six times stronger a barrier can be built in the DIII-D than in the ASDEX UG. High magnetic shear near the separatrix in the DIII-D is inferred as the possible cause of this. Implications of this for the DIII-D and the International Thermonuclear Experimental Reactor (ITER) are discussed. [ABSTRACT FROM AUTHOR]

In this paper, an emissive-biased limiter (EBL) was designed and fabricated then the magneto hydrodynamic activity was investigated based on Mirnov oscillations and hard X-ray spectroscopy through the tokamak plasma biasing. The EBL is positioned at r/a=0.92, and the biased voltage, which is varied from−250 to 250 V, applied between the head of the emissive limiter and vacuum chamber. Furthermore, the effects of the biased limiter for both negative and positive applied voltages are measured, and the results are compared with cold-biased limiter. As the results of IR-T1 tokamak suggest, in emissive negative polarity, the duration of plasma current is increased, compared with no bias, cold positive and emissive positive polarities. The amplitude of Mirnov fluctuations in emissive negative polarity is larger and more regular, compared with emissive positive polarity. The amplitude of the hard X-ray fluctuation in emissive negative polarity is very low compared with cold negative, cold positive and emissive positive polarities which leads to minimum energy loss. [ABSTRACT FROM AUTHOR]