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Plasma vertical displacement control is essential for the stable operation of tokamak devices. The traditional plasma vertical displacement calculation method is not suitable for balancing speed and accuracy simultaneously, which is necessary for real-time feedback control. In this study, neural networks are used to rapidly detect vertical displacement recognition. Based on a fully connected neural network, the vertical displacement calculation model is trained and tested using magnetic data of approximately 2000 shots. To compare the effects of different inputs on vertical displacement calculation, different magnetic measurement diagnostic signals are used to train and test the model. Compared with a full magnetic measurement dataset, 39 magnetic measurement signals (38 magnetic probes and plasma current) show better accuracy with mean square error

Simultaneous control of transient heat load induced by large-amplitude edge-localized modes (ELMs) and steady-state heat load on divertor targets under metal wall environment is crucial for steady-state operation of future tokamak fusion reactors, such as ITER and the China Fusion Engineering Test Reactor (CFETR). In the recent experiments, sustained partial energy detachment without confinement degradation has been achieved in the Experimental Advanced Superconducting Tokamak (EAST) in high-performance grassy-ELM H-mode with q95 ~ 5.9 by a newly developed detachment feedback control scheme, in which we first used electron temperature (Tet) measured by divertor Langmuir probes to identify the onset of energy detachment, and then the system switched to the feedback control of total radiation power measured by absolute extreme ultraviolet (AXUV) system. Tet around the upper outer strike point was successfully maintained less than 8 eV with seeding of 80% Ne and 20% D2 mixture from upper outer divertor, and the total radiation power was maintained ~1.4 MW, around 52% of injected power. There was no significant decrease of the plasma stored energy and H98,y2 factor (~1) over the entire detachment feedback control process. These experiment results demonstrate good compatibility of the high-performance grassy-ELM regime with radiative divertor. In order to confirm the compatibility in a wider range, stable partial energy detachment in grassy-ELM H-mode with relatively lower q95 (~5.4) was also achieved in EAST through the newly developed integrated-feedback-control technique. The new detachment feedback control without confinement degradation in grassy-ELM H-mode provides a candidate mode for EAST long-pulse operation in the future with well control of ELM-induced transient and steady heat fluxes on the divertor target.

The compatibility of efficient divertor detachment with high-performance core plasma is vital to the development of magnetically controlled fusion energy. The joint research on the EAST and DIII-D tokamaks demonstrates successful integration of divertor detachment with excellent core plasma confinement quality, a milestone towards solving the critical plasma–wall-interaction (PWI) issue and core-edge integration for ITER and future reactors. In EAST, actively controlled partial detachment with T et,div ∼ 5 eV around the strike point and H 98 > 1 in different H-mode scenarios including the high β P H-mode scenario have been achieved with ITER-like tungsten divertor, by optimizing the detachment access condition and performing detailed experiments for core-edge integration. For active long-pulse detachment feedback control, a 30 s H-mode operation with detachment-control duration being 25 s has been successfully achieved in EAST. DIII-D has achieved actively controlled fully detached divertor with low plasma electron temperature (T et,div ⩽ 5 eV across the entire divertor target) and low particle flux (degree of detachment, DoD > 3), simultaneously with very high core performance (β N ∼3, β P > 2 and H 98 ∼ 1.5) in the high βP scenario being developed for ITER and future reactors. The high-β P high confinement scenario is characterized by an internal transport barrier (ITB) at large radius and a weak edge transport barrier (ETB, or pedestal), which are synergistically self-organized. Both the high-β P scenario and impurity seeding facilitate divertor detachment. The detachment access leads to the reduction of ETB, which facilitates the development of an even stronger ITB at large radius in the high β P scenario. Thus, this strong large radius ITB enables the core confinement improvement during detachment. These significant joint DIII-D and EAST advances on the compatibility of high confinement core and detached divertor show a great potential for achieving a high-performance core plasma suitable for long-pulse operation of fusion reactors with controllable steady-state PWIs.

The efficient and safe operation of large fusion devices relies on plasma configuration inside the vacuum chamber. Due to EAST PF coils distribution, it is difficult to physical decouple plasma control parameters and poloidal field (PF) coils current. In this paper, we present a MIMO decoupling controller aimed to overcome the intrinsic limitations of simpler SISO controller, this new controller has been validated by means of the Tokamak Simulation Code (TSC) and implemented on the EAST tokamak. Preliminary results show the potential of this approach for the EAST plasma shape control systems. © 2018 Elsevier B.V.

A multi-function divertor feedback control system has been built on Experimental Advanced Superconducting Tokamak (EAST) to treat the divertor heat load issue. With the real-time data of the Langmuir probes and the impurity seeding, the divertor electron temperature (Te,div) is well controlled to achieve the partial detachment phase. The first trial of the CD4 seeding for the Te,div reduction has been achieved on EAST long-pulse discharge. In the seeding phase, the Te,div was maintained close to 5 eV, and the surface temperature of the target plate (Tsurface,div) had a reduction of about 150 oC. The plasma stored energy had a reduction in the control phase, so it is necessary to find a way to keep the good plasma confinement in the next step. The CD4 injection also mitigated the low hybrid wave coupling rate in some degree. The big volume of the CD4 injection lifted the Greenwald density fraction from ~0.4 to ~0.7, which made the SOL into high recycling state. Most of the injected carbon particles were in the high ionized state, and with the lower of the Te,div, the tungsten line emission was suppressed obviously.

A real-time disruption predictor using random forest was developed for high-density disruptions and used in the plasma control system (PCS) of the EAST tokamak for the first time. The disruption predictor via random forest (DPRF) ran in piggyback mode and was actively exploited in dedicated experiments during the 2019–2020 experimental campaign to test its real-time predictive capabilities in oncoming high-density disruptions. During dedicated experiments, the mitigation system was triggered by a preset alarm provided by DPRF and neon gas was injected into the plasma to successfully mitigate disruption damage. DPRF’s average computing time of ∼250 μs is also an extremely relevant result, considering that the algorithm provides not only the probability of an impending disruption, i.e. the disruptivity, but also the so-called feature contributions, i.e. explainability estimates to interpret in real time the drivers of the disruptivity. DPRF was trained with a dataset of disruptions in which the electron density reached at least 80% of the Greenwald density limit, using the zero-dimensional signal routinely available to the EAST PCS. Through offline analysis, an optimal warning threshold on the DPRF disruptivity signal was found, which allows for a successful alarm rate of 92% and a false alarm rate of 9.9%. By analyzing the false alarm causes, we find that a fraction (∼15%) of the misclassifications are due to sudden transitions of plasma confinement from H- to L-mode, which often occur during high-density discharges in EAST. By analyzing DPRF feature contributions, it emerges that the loop voltage signal is that main cause of such false alarms: plasma signals more apt to characterize the confinement back-transition should be included to avoid false alarms.

In this study, a full convolutional neural network is trained on a large database of experimental EAST data to classify disruptive discharges and distinguish them from non-disruptive discharges. The database contains 14 diagnostic parameters from the ∼104 discharges (disruptive and non-disruptive). The test set contains 417 disruptive discharges and 999 non-disruptive discharges, which are used to evaluate the performance of the model. The results reveal that the true positive (TP) rate is ∼ 0.827, while the false positive (FP) rate is ∼0.067. This indicates that 72 disruptive discharges and 67 non-disruptive discharges are misclassified in the test set. The FPs are investigated in detail and are found to emerge due to some subtle disturbances in the signals, which lead to misjudgment of the model. Therefore, hundreds of non-disruptive discharges from training set, containing time slices of small disturbances, are artificially added into the training database for retraining the model. The same test set is used to assess the performance of the improved model. The TP rate of the improved model increases up to 0.875, while its FP rate decreases to 0.061. Overall, the proposed data-driven predicted model exhibits immense potential for application in long pulse fusion devices such as ITER.

In order to develop a new alternative magnetic diverted (AMD) discharge on EAST, it is necessary to define a series of target configuration parameters on plasma control system (PCS), especially this new configuration requires a strong effort on the poloidal field (PF) coil currents. By combining the static equilibrium configuration optimization and flux consuming estimation for PF coils current, multiple basic scenarios with exactly the same target configuration were designed for different stages on plasma current flat-top phase. The high PF coil currents requirement was optimized by the specific equilibrium design procedure to keep them away from the limitation. In this paper, we present the fast scenario design details, together with the experiments results.

During 2016–2018 experimental campaigns, the plasma magnetic control architecture of the EAST tokamak was revised in order to achieve improved performances, with the final aim of feedback control of alternative divertor configurations (i.e. with multiple X-points). This paper reports on the results obtained with the Multi-Input–Multi-Output (MIMO) plasma shape controller tested during the last experimental campaign, which prompted a considerable improvement of the control performances. Simulation results are also discussed.

For long pulse or steady-state advanced plasma discharges, it is necessary to keep plasma parameters such as plasma shape, loop voltage, plasma pressure in a steady state way. In particular, the heat load on the divertor target must be effectively reduced. By using Low Hybrid Wave (LHW) for the current drive, loop voltage has been feedback controlled while plasma current was controlled by the PF coil current. The control of the plasma pressure has been demonstrated by using LHW. Heat load reduction has been done by radiation and advanced shape configuration. By using impurity seeding with gas puff for the feedforward and Super-Sonic Molecular Beam Injection (SMBI) to feedback the total radiation, radiation can be effectively controlled with slight influence to the core confinement. The quasi-snowflake (QSF) discharge in H-mode has been achieved. It is verified again in H-mode operation, heat load can be effectively reduced under the QSF shape. All these new control algorithms give rise to more assistances to the EAST long pulse operation. © 2018 Elsevier B.V.

This paper deals with the model-based approach that has been adopted to design the plasma magnetic control at the EAST tokamak. Such a design approach, which is based on a linear model for the response of the plasma and of the surrounding coils, has been successfully applied in 2016 to design an alternative solution for the plasma vertical stabilization, the plasma centroid position control, and the controller of the currents in the poloidal field circuits. After an introduction to the linear model for the plasma/circuits response, the proposed control algorithms are presented together with the preliminary results obtained during the 2016 experimental campaign at EAST. A brief description of the ongoing and future design activities is also given.

Recently, the first ever 100 s long, steady-state H-mode discharge with good control of impurities, core and edge MHD stabilities, and heat exhaust was demonstrated in the Experimental Advanced Superconducting Tokamak (EAST) using the ITER-like (International Tokamak Experimental Reactor) tungsten upper divertor. Using both radio frequency (RF) power and neutral beam injection (NBI) heating, EAST has demonstrated fully non-inductive scenarios with an extension of fusion performance at high density and low rotation: β P ∼ 2.5, β N ∼ 2.0, H98,y2 ∼ 1.2, bootstrap current fraction fBS ∼50% at q95 ∼ 6.8. With pure RF power heating, plasmas have been maintained for up to 21 s (over 40 times the current relaxation time) with zero loop voltage and small edge localized modes (ELMs) at high density (ne/nGW ∼ 0.6–0.8), β P ∼ 2.0, β N ∼ 1.6, and ƒBS ∼47%. Experimental investigations show how plasma current profiles, turbulent transport and radiation properties self-consistently evolve toward fusion relevant steady state conditions. Modeling and physics experiments have confirmed the synergistic effects between electron cyclotron heating (ECH) and low hybrid wave (LHW), where ECH enhances the heating and current drive from LHW injection, enabling fully non-inductive operation at higher density. Small/no ELMs facilitate the RF power coupling in the H-mode phase and reduce divertor erosion. A low tungsten concentration was observed at high β P with a hollow profile in the core. Reduction of the peak divertor heat flux with f rad of up to 40% was compatible with the high β P scenario by using active radiation feedback control. With features such as dominant electron heating, zero/low NBI torque and an ITER-like tungsten divertor, fully non-inductive high-performance experiments on EAST offer unique contributions towards the succesful operation of ITER and CFETR (the Chinese Fusion Engineering Testing Reactor).

This paper studies a generalized Euler problem of inhomogeneous cantilever columns with an end linked to rotational spring and another free end. This paper has two-fold aims. One is to give a characteristic equation to exactly determine buckling loads when axial inhomogeneity is exponential gradient, and the other is to provide an approximate explicit expression for buckling loads. For the former, we make substitution of variables and transform an associated problem to a Bessel equation. Under appropriate boundary conditions, an exact characteristic equation for buckling loads is derived. For the latter, we reduce it to an integral equation, and then apply the moment method to obtain an approximate expression for buckling load. By comparing approximate results with the exact ones, the approximation critical loads provide satisfactory accuracy for a large range of gradient index and rotational spring stiffness. The dependence of buckling loads on the gradient index and spring stiffness is examined. The buckling loads and shapes for an exponential graded column are presented graphically for typical rotational spring stiffness. The derived exact solution can be taken as a benchmark solution to examine the accuracy of other numerical approaches and is of benefit to optimum design of non-homogeneous columns in engineering.

A simulation of a disruption in the experimental advanced superconducting tokamak (EAST) is carried out with the CarMa0NL code and the results are successfully compared with the experimental data. The fundamental role of three-dimensional (3D) features of conducting structures surrounding the plasma is demonstrated and is fully accounted for in simulations. The results demonstrate that the currents measured by Rogowski coils placed around the supports of in-vessel plasma facing components (PFCs) are largely due to eddy currents induced by the plasma motion and current decay, with a relatively small contribution coming from the currents directly injected from the plasma into the structures (halo currents).

Recent Experimental Advanced Superconducting Tokamak (EAST) experiments have successfully demonstrated a long-pulse steady-state scenario with improved plasma performance through integrated operation since the last IAEA FEC in 2016. A discharge with a duration over 100 s using pure radio frequency (RF) power heating and current drive has been obtained with the required characteristics for future long-pulse tokamak reactors such as good energy confinement quality (H98y2 ~ 1.1) with electron internal transport barrier inside ρ < 0.4, small ELMs (frequency ~100–200 Hz), and good control of impurity and heat exhaust with the tungsten divertor. The optimization of X-point, plasma shape, the outer gap and local gas puffing near the low hybrid wave (LHW) antenna were integrated with global parameters of B T and line-averaged electron density for higher current drive efficiency of LHW and on-axis deposition of electron cyclotron heating in the long-pulse operation. More recently, a high β P RF-only discharge (β P ~ 1.9 and β N ~ 1.5, /n GW ~ 0.80, f bs ~ 45% at q 95 ~ 6.8) was successfully maintained over 24 s with improved hardware capabilities, demonstrating performance levels needed for the China Fusion Engineering Test Reactor steady-state operation. A higher energy confinement is observed at higher β P and with favorable toroidal field direction. Towards the next goal (≥400 s long-pulse H-mode operations with ~50% bootstrap current fraction) on EAST, an integrated control of the current density profile, pressure profile and radiated divertor will be addressed in the near future.

The Chinese Fusion Engineering Testing Reactor (CFETR), complementing the ITER facility, is aiming to demonstrate fusion energy production up to 200 MW initially and to eventually reach DEMO relevant power level 1 GW, to manifest a high duty factor of 0.3–0.5, and to pursue tritium self-sufficiency with tritium breeding ratio (TBR) >1. The key challenge to meet the missions of the CFETR is to run the machine in steady state (or long pulse) and high duty factor. By using a multi-dimensional code suite with physics-based models, self-consistent steady-state and hybrid mode scenarios for CFETR have been developed under a high magnetic field up to 6.5 T. The negative-ion neutral beam injection together with high frequency electron cyclotron wave and lower hybrid wave (and/or fast wave) are proposed to be used to drive the current. Subsequently the engineering design of CFETR including the magnet system, vacuum system, tritium breeding blanket, divertor, remote handling and maintenance system will be introduced. Some research and development (R&D) activities are also introduced in this paper.

A ITER-like vertical stabilization (VS) algorithm has been successfully deployed and commissioned at EAST. The proposed algorithm decouples the VS from the plasma shape control, while the algorithms previously implemented to stabilize the EAST plasma exhibit a strong coupling with plasma shape control system. As a consequence, the VS algorithms previously implemented at EAST prevent the deployment of advanced multi-input-multioutput (MIMO) plasma shape control schemes, such as the ones proposed in Albanese et al 2016 (Proc. 2016 IEEE Multi-Conf. System Control (Buenos Aires, Argentina) pp 611-6) and Kolemen et al (2015 J. Nucl. Mater. 463 1186). Indeed, such MIMO controllers rely on the decoupling with the VS system. The proposed ITER-like stabilizes the plasma column (i.e. it controls to zero the plasma vertical speed) on the fastest possible time scale, while leaves the control of the plasma vertical position to the plasma shape controller. Thanks to this frequency separation approach, the plasma shape controller can than be designed starting from the stabilized system, without explicitly taking the VS into account. In this paper we present the implementation details of the adopted solution for the EAST vertical stabilization, together with the results obtained during the 2016 experimental campaign.

This declaration, signed by an interdisciplinary task force of 234 experts from 83 different countries with different backgrounds, highlights the threat posed by antimicrobial resistance and the need for appropriate use of antibiotic agents and antifungal agents in hospitals worldwide especially focusing on surgical infections. As such, it is our intent to raise awareness among healthcare workers and improve antimicrobial prescribing. To facilitate its dissemination, the declaration was translated in different languages. © Copyright 2017, Mary Ann Liebert, Inc.

In this paper we discuss the problem of plasma vertical stabilization at the EAST tokamak. By exploiting a plasma/circuit linearized model, we show that the plant cannot be strongly stabilized by using the in-vessel coils and a single-input-single-output controller that feeds back only the plasma vertical speed (z)over dot(p) (i.e. without integral action on (z)over dot(p)). Moreover, a stable multi-input-single-output controller that stabilizes the plant without the need of feeding back the plasma vertical position is presented. The proposed solution permits to achieve stabilization of the EAST plant without coupling the vertical stabilization system with the plasma shape controller. Such decoupling is a key requirement to enable advanced design approaches for plasma shape controller.

KT'K is a new reversed field pinch (RFP) device, and the Central Control System (CCS) for KTX is developed to integrate, harmonize and supervise all of the control subsystems. The new control system includes graphical user interface (GUI), discharge process control module, timing system, safety & interlock system, gas injection module and shot information server. The details about the central control system architecture and components will be described in this article.

The Experimental Advanced Superconductive Tokamak (EAST) Device began operation in 2006. EAST's inner structure is very complicated and contains a lot of subsystems which have a variety of different functions. In order to facilitate the understanding of the device and experimental information and promote the development of the experiment, the virtual EAST system has established an EAST virtual reality scene in which the user can roam and access to information by interacting with the system. However, experiment-related parameter information, diagnostic information and magnetic measurement information are displayed in the form of charts, figures, tables and two dimensional graphics. In order to express the experimental results directly, three-dimensional data visualization results are created using computer graphics technology. Data visualization is the process of visualizing data based on the characteristics of the data, the selection of the appropriate data structure and the proper sequence of visual pipeline. We use the visualization toolkit (VTK) to realize the data visualization in VEAST system and give the detailed steps of data visualization of plasma column, electron cyclotron emission diagnostic and plasma magnetic field. Besides, the general format is defined for the users to organize their data so that they can visualize their data in our system.

A data acquisition and control system has been realized for a Multi-Energy Soft X-Ray (ME-SXR) diagnostic, which has been built for EAST (Experimental Advanced Superconducting Tokamak) electron temperature profile in the edge plasma region. The system has two functions: long-pulse data acquisition and electronics' automatic gain-control. It provides 96 high-speed channels with optional sampling rates from 250 KSPS to 1 MSPS, which can long-pulse acquires signals of ME-SXR. In the meanwhile, it transfers the data to the MDSplus server in real-time and users can access to the data during the discharge. The signals from the diode array detectors are in the microampere range, which should be amplified by the electronics with the transimpedance in 8 stages. In the past, electronics' gain settings were totally man-control. In order to realize auto-control, it makes data de-noising and calculates appropriate gains, according to the pulse situation and ME-SXR's reference signals in the last shot. The goal of the system is to ensure long-pulse data acquisition and electronics' automatic gain-control. The system has been demonstrated in the 2015 EAST campaign. The details are presented in the paper.

A two-dimensional orthogonal thermal-mechanical finite element model by Deform2D finite element analysis software is established in the article. By the adaptive meshing technique, not only cutting process but also the effect on the process of aluminum alloy Al6061-T6 processing as friction coefficient changing is simulated. The simulation shows that the friction coefficient has significant effect on the cutting temperature and cutting force, and the effect is nonlinear. With the increasing of the friction coefficient, the cutting temperature and cutting force will both increase. The impact the friction coefficient has on the surface residual stress is much smaller than the impact on the cutting temperature and cutting force.

The DIII-D digital plasma control system (D3PCS) has been in use since 1993 controlling DIII-D plasmas. Control research and D3PCS development at DIII-D has revealed aspects of advanced tokamak control which can help inform the ITER design. The D3PCS has also been adapted for use at several fusion devices worldwide, which has allowed the DIII-D team to obtain a significant understanding of common requirements for plasma control on multiple fusion devices, along with substantial experience in the alternative computing, data acquisition, and networking technologies presently available. We describe some of what has been learned and highlight the relevance of these lessons for ITER Control, Data Access and Communication (CODAC). Several capabilities of the D3PCS are described in the context of ITER requirements. This description touches on multiple subsystems described in the CODAC conceptual design. We also discuss features of D3PCS architecture that are appropriate for today's experimental devices but may not be appropriate for ITER, which requires a much more comprehensive system for ensuring device safety. Even for these applications, the knowledge gained in implementing methods to aid in ensuring device safety on present devices provides useful guidance for an eventual ITER solution.

The DIII-D digital plasma control system (PCS) has been adapted and used at a number of tokamaks around the world over the past several years. Implementations of the DIII-D PCS at KSTAR and EAST in particular have made use of, and improved upon, the methodology for off-normal event/fault detection and response that was originally incorporated into the DIII-D PCS. This work discusses the lessons learned implementing and installing codes for detecting and responding to off-normal events in both plasma discharges and tokamak systems. In particular, we discuss the implementation details for codes at DIII-D, EAST, and KSTAR, and discuss how those experiences are being incorporated into a proposed finite state machine architecture in support of ITER research at DIII-D.

The current DIII-D plasma control system (PCS) has evolved through several iterations into a robust platform that has been adopted at several fusion devices around the world. Each installation, as well as each new upgrade at DIII-D, has presented new challenges. Each of these challenges has provided an additional opportunity to expand our understanding of the requirements, alternative operational methods, and differing real-time implementations for Tokamak plasma control. This paper presents a brief historical overview of PCS hardware evolutions and describes some of the design, structure, and techniques that have allowed the PCS to be a productive component at many fusion facilities. It will also discuss some of the major differences between the individual PCS installations and bring to light some of the major challenges that were overcome during integration. The lessons learned from these experiences provide general solutions and can inform control system designs for other next-generation devices. We also describe some limitations of the PCS relative to identified present and future needs at DIII-D and other devices, and discuss planned upgrades to the PCS to address these needs.

This paper briefly outlines the current status of CODAC system of the EAST tokamak. Integrated central system has been working for synchronizing all the subsystems and responsible for the integral safety protection by interlock. Data acquisition system has been established for the discharge with duration up to 100 seconds. Various data visualization tools such as EASTVIEWER to view the flux surfaces, RTSCOPE to view plasma boundary and CCD image in real-time, WEBSCOPE to view the diagnostic data via a web browser and EASTVOD to view and search plasma discharge image at different time, haven developed for EAST operation. Plasma control system is inherited from DIII-D, and has been matured to a stage to feedback control the EAST plasma shape, density and current in real time.

The Experimental Advanced Superconducting Tokamak (EAST) was the first shaped tokamak of mega-ampere scale to achieve plasma utilizing a fully superconducting poloidal field coil system, and it is ...

This presentation will describe the DIII-D collaborations with various tokamak experiments throughout the world which have adapted custom versions of the DIII-D plasma control system (PCS) software for their own use. Originally developed by General Atomics for use on the DIII-D tokamak, the PCS has been successfully installed and used for the NSTX experiment in Princeton, the MAST experiment in Culham UK, the EAST experiment in China, and the Pegasus experiment in the University of Wisconsin. In addition to these sites, a version of the PCS is currently being developed for use by the KSTAR tokamak in Korea. A well-defined and robust PCS software infrastructure has been developed to provide a common foundation for implementing the real-time data acquisition and feedback control codes. The PCS infrastructure provides a flexible framework that has allowed the PCS to be easily adapted to fulfill the unique needs of each site. The software has also demonstrated great flexibility in allowing for different computing, data acquisition and real-time networking hardware to be used. A description of the current PCS software architecture will be given along with experiences in developing and supporting the various PCS installations throughout the world.

Evidence is shown of the capability of non-axisymmetrical conducting structures in the Experimental Advanced Superconducting Tokamak (EAST) to guarantee the passive stabilization of the n = 0 MHD unstable mode. Suitable numerical modeling of the experiments allows a clear interpretation of the phenomenon. This demonstration and the availability of computational tools able to describe the effect of 3D conductors will have a huge impact on the design of future fusion devices, in which the conducting structures closest to plasma will be highly segmented.

A simple 0D analytical model for the prediction of the quench point has been presented by Anghel [Anghel A. A simple model for the prediction of quench point in cable-in-conduit conductors. Cryogenics 2003;43:225–32]. The model offered a straightforward approach to estimate the quench behavior of cable-in-conduit-conductor (CICC) and provided a clear view of the relationships between conductor parameters and its quench properties. However, this 0D model did not address the current-sharing effect of copper. In this paper, it was considered by approximation that superconductor and copper form two resistors in parallel, the redistribution of current due to the copper was taken into account. The result of the new model was compared with that of a previous model. The 0D model predicts the quench with a higher accuracy.

The experiments of Cable-in conduit conductor (CICC) short samples with high proportion of segregated copper strands have been carried out in SULTAN facility last September. These experiments aimed to investigate transient stability and AC losses of CICC conductor coated with different resistive barriers (Pb-30Sn-2Sb or Ni plating on strands) and to check the design of PF and TF CICC for HT-7U magnets. The resistive barriers' influences on the stability and AC losses of CICC are evaluated. These experimental results are used for the choice of HT-7U TF and PF CICC design. (C) 2002 Elsevier Science Ltd. All rights reserved.

The active feedback control of radiated power to prevent divertor target plates overheating during long-pulse operation has been developed and implemented on EAST. The radiation control algorithm, with impurity seeding via a supersonic molecular beam injection (SMBI) system, has shown great success in both reliability and stability. By seeding a sequence of short neon (Ne) impurity pulses with the SMBI from the outer mid-plane, the radiated power of the bulk plasma can be well controlled, and the duration of radiative control (feedforward and feedback) is 4.5 s during a discharge of 10 s. Reliable control of the total radiated power of bulk plasma has been successfully achieved in long-pulse upper single null (USN) discharges with a tungsten divertor. The achieved control range of is 20%–30% in L-mode regimes and 18%–36% in H-mode regimes. The temperature of the divertor target plates was maintained at a low level during the radiative control phase. The peak particle flux on the divertor target was decreased by feedforward Ne injection in the L-mode discharges, while the Ne pulses from the SMBI had no influence on the peak particle flux because of the very small injecting volume. It is shown that although the radiated power increased, no serious reduction of plasma-stored energy or confinement was observed during the control phase. The success of the radiation control algorithm and current experiments in radiated power control represents a significant advance for steady-state divertor radiation and heat flux control on EAST for near-future long-pulse operation.

High pressure noble gas injection is a promising technique to mitigate the effect of disruptions in tokamaks. In this paper, results of mitigation experiments with low-Z massive gas injection (helium) on the EAST tokamak are reported. A fast valve has been developed and successfully implemented on EAST, with valve response time ≤150 μs, capable of injecting up to particles, corresponding to 300 times the plasma inventory. Different amounts of helium gas were injected into stable plasmas in the preliminary experiments. It is seen that a small amount of helium gas () can not terminate a discharge, but can trigger MHD activity. Injection of 40 times the plasma inventory impurity () can effectively radiate away part of the thermal energy and make the electron density increase rapidly. The mitigation result is that the current quench time and vertical displacement can both be reduced significantly, without resulting in significantly higher loop voltage. This also reduces the risk of runaway electron generation. As the amount of injected impurity gas increases, the gas penetration time decreases slowly and asymptotes to (~7 ms). In addition, the impurity gas jet has also been injected into VDEs, which are more challenging to mitigate that stable plasmas.

The motivation for developing fusion-fission hybrid reactors is discussed in the context of electricity power requirements by 2050 in China. A detailed conceptual design of the Fusion Experimental Breeder (FEB) was developed from 1986-1995. The FEB has a subignited tokamak fusion core with a major radius of 4.0 m, a fusion power of 145 MW, and a fusion energy gain Q of 3. Based on this, an engineering outline design study of the FEB, FEB-E, has been performed. This design study is a transition from conceptual to engineering design in this research. The main results beyond that given in the detailed conceptual design are included in this paper, namely, the design studies of the blanket, divertor, test blanket, and tritium and environment issues. In-depth analyses have been performed to support the design. Studies of related advanced concepts such as the waste transmutation blanket concept and the spherical tokamak core concept are also presented.

An efficient technique for building plasma equilibria with preset or constrained q profiles by solving the nonlinear Grad–Shafranov equation iteratively is proposed. The technique uses a set of dynamic virtual poloidal field probes in the plasma core region. Numerical examples are shown for monotonic q profiles and negative central magnetic shear equilibria which are obtained with this technique. As a proof of principle, the technique is applied on the EAST tokamak, to constrain the plasma equilibrium reconstruction with the experimental q = 1 surface information from the ECE diagnostic in addition to the external magnetic measurements.

A low aspect ratio tokamak transmutation system is proposed as an alternative application of fusion energy on the basis of a review of previous studies. This system includes: (1) a low aspect ratio tokamak as fusion neutron driver, (2) a radioactivity-clean nuclear power system as blanket, and (3) a novel concept of liquid metal centre conductor post as part of the toroidal field coils. In the conceptual design, a driver of 100 MW fusion power under 1 MW/m2 neutron wall loading can transmute the amount of high level waste (including minor actinides and fission products) produced by ten standard pressurized water reactors of 1 GW electrical power output. Meanwhile, the system can produce tritium on a self-sustaining basis and an output of about 2 GW of electrical energy. After 30 years of operation, the biological hazard potential level of the whole system will decrease by two orders of magnitude.

The fully exposed center conductor post (CCP) in a spherical tokamak will receive strong neutron irradiation. The analytical results for a CCP in a spherical tokamak reactor are given considering the irradiation effects such as radiation damage, transmutation, nuclear and resistive heat removal, induced radioactivity and blanket tritium breeding ratio, etc. These results are compared with those for the first wall of conventional tokamaks, assuming the same technical requirements.