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A project is proposed of the complex of diagnostics of the plasma-facing elements of the first wall and divertor of tokamak with reactor technologies (TRT). The main objectives of the complex are control of the integrity of plasma-facing elements, monitoring of the state of the first wall and the thermal flows onto the first wall and divertor targets as well as studies of the physical processes in the material of the first wall during its interaction with plasma. A review of expected effects of interaction of plasma with the plasma-facing elements is given; in the first place, with the tungsten facing of the divertor. Based on the analysis of the methods that were tested at existing tokamaks and methods that are being developed within the framework of the international ITER project, a complex of complementary diagnostics is proposed, which satisfies the TRT tasks. In the paper, a variant of implementation at TRT is discussed of frequency/amplitude modulated laser radar, two-wavelength digital holography, active (laser) and passive IR thermography, diagnostics of fuel build-up based on laser desorption and ablation, and the placement of witness samples. A variant is proposed of the placement of these diagnostics at TRT and their physical integration. [ABSTRACT FROM AUTHOR]

Following the analysis on linear spectra of shear Alfvén fluctuations excited by energetic particles (EPs) in the Divertor Tokamak Test (DTT) facility plasmas [Wang et al., Phys. Plasmas 25, 062509 (2018)], in this work, nonlinear dynamics of the corresponding mode saturation and the fluctuation induced EP transport is studied by hybrid magnetohydrodynamic-gyrokinetic simulations. For the reversed shear Alfvén eigenmode driven by magnetically trapped EP precession resonance in the central core region of DTT plasmas, the saturation is mainly due to radial decoupling of resonant trapped EPs. Consistent with the wave-EP resonance structure, EP transport occurs in a similar scale to the mode width. On the other hand, passing EP transport is analyzed in detail for the toroidal Alfvén eigenmode in the outer core region, with mode drive from both passing and trapped EPs. It is shown that passing EPs experience only weak redistributions in the weakly unstable case, and the transport extends to meso-scale diffusion in the strongly unstable case, due to orbit stochasticity induced by resonance overlap. Here, a weakly/strongly unstable regime is determined by the Chirikov condition for resonance overlap. This work then further illuminates rich and diverse nonlinear EP dynamics related to burning plasma studies and the capability of DTT to address these key physics. [ABSTRACT FROM AUTHOR]

Neutron/gamma discrimination techniques are widely applied in scintillator-based neutron diagnostics for present nuclear fusion tokamak experiments (e.g. JET – Neutron Camera and compact neutron spectrometer) and will also be necessary for neutron diagnostics of up-coming machines (e.g. ITER Radial Neutron Camera). Neutron/gamma discrimination in scintillators relies on the fact that the detectors output pulses have different shapes depending on the impinging particle; several discrimination techniques are described in literature such as the charge-integration, curve-fitting and pattern recognition [1]. This paper aims at describing a new technique for neutron/gamma discrimination in scintillators based on trapezoidal filtering, which targets Field Programmable Gate Array (FPGA) implementation due to its recursive nature. Furthermore its capability to restore the baseline of each detected pulse and the fact that the output signals are shorter than the correspondent incoming pulses, points this technique as a promising solution for applications in high count rate conditions. First results coming from the application of a real-time FPGA implementation of the trapezoidal filter to simulated neutron/gamma data including pile-up events and to real scintillator data will be presented. [ABSTRACT FROM AUTHOR]

In this paper, techniques for optimal input design are used to optimize the waveforms of perturbative experiments in modern fusion devices. The main focus of this paper is to find the modulation frequency for which the accuracy of the estimated diffusion coefficient is maximal. Mathematically, this problem can be formulated as an optimization problem in which the Fisher information matrix is maximized. First, this optimization problem is solved for a simplified diffusion model, while assuming a slab geometry and a semi-infinite domain. Later, the optimization is repeated under more general conditions such as a cylindrical geometry, finite domain, and simultaneous estimation of multiple transport coefficients. Based on the results of these optimizations, guidelines are offered to select the modulation frequency and to determine the optimality of the corresponding experiment. [ABSTRACT FROM AUTHOR]

We introduce a new near real-time solution for the tokamak resistive MHD Δ ′ matrix. By extending state transition matrix methods introduced in [Glasser et al., Phys. Plasmas 25(3), 032507 (2017)] and leveraging the asymptotic methods of [A. H. Glasser, Phys. Plasmas 23, 072505 (2016)], we have developed STRIDE—State Transition Rapid Integration with DCON (Asymptotic) Expansions—a code that solves for Δ ′ in <500 ms. The resistive MHD stability remains a foremost challenge in successful tokamak operation, and its numerically demanding analysis has received attention for many years. Our code substantially improves upon the speed and robustness of earlier Δ ′ calculation methods, affording solutions for previously intractable equilibria and helping enable the real-time control of ideal and resistive MHD tokamak stability. In this paper, we pedagogically review tearing stability analysis and motivate and define Δ ′ in the slab, cylindrical, and toroidal geometries. We also benchmark STRIDE against the calculations of [Nishimura et al., Phys. Plasmas 5, 4292–4299 (1998)] and Furth et al. [Phys. Fluids 16, 1054 (1973)] for Δ ′ in a cylindrical geometry, and the Δ ′ matrix calculations of [A. H. Glasser, Phys. Plasmas 23, 112506 (2016)] in the full toroidal geometry. [ABSTRACT FROM AUTHOR]

The fast linear (Born approximation) version of the X-mode Doppler reflectometry (DR) synthetic diagnostics is developed in the framework of the ELMFIRE global gyrokinetic modeling of the FT-2 tokamak ohmic discharge. The DR signal frequency spectra and the dependence of their frequency shift and shape on the probing antenna position are computed and shown to be similar to those measured in the high magnetic field side probing DR experiment at the FT-2 tokamak. The fluctuation poloidal velocities are determined using the DR experiment and synthetic diagnostics and shown to be within 15%. However, the computed and measured dependences of the DR signal power on the antenna position appear to be different presumably due to underestimation of the small-scale trapped electron mode turbulence component in the measurement region by the code. [ABSTRACT FROM AUTHOR]

China Fusion Engineering Test Reactor (CFETR) vacuum vessel (VV), which is welded from forming shells and ribs, is a toroidal double shell structure device. However, there are manufacturing errors in the forming and welding processes that will lead to accumulative deviation of the VV dimensions and affect the VV contour. In order to reduce and decentralize errors, the surfaces of welded parts are measured and then fit with least-squares method. According to the fitting result, new benchmarks are reset in the form of controlling points for the next assembly. This paper is about the application of contour fitting method in manufacturing of the first 1/32 VV sector of the CFETR 1/8 VV mock-up. [ABSTRACT FROM AUTHOR]

First wall panels (FWPs), which adjoin the inner wall of the blanket modules in the vacuum vessel (VV) of a Tokamak device, are in structures of multilayer bounded together with a solid welding technique in order to perform its heat exchange, VV protection, and neutron breeding functions. The quality of the welding joint between layers is the key factor for FWP integrity. In order to conduct online inspection of the delamination defect in the FWPs, a nondestructive testing (NDT) method capable to detect delamination defect without accessing into the VV is required. In this paper, the feasibility of the laser infrared thermography (LIRT) testing NDT method was investigated experimentally for this purpose. To clarify its detectability under practical VV environment, inspections of several inspection modes were conducted based on the practical structure of FWP and VV of the EAST Tokamak device, i.e., modes of different distances and angles of FWPs toward the LIRT transducers. In practice, an LIRT testing system was established and several double-layered plate specimens with different artificial delamination defects were inspected under the selected testing conditions. Through thermography signal reconstruction, an image processing algorithm was proposed and adopted to enhance the defect detectability. From the results of different inspection modes, it was found that the angle factor may worsen the inspection precision and reduce the detectability for delamination defects in case of big defect depth-to-width ratio, even though the LIRT method is still applicable for inspection of relative large defects in FWP. Finally, the detectability in different inspection modes was clarified, which proved the feasibility of LIRT for FWP online inspection. [ABSTRACT FROM AUTHOR]

This paper clearly demonstrates the architecture of time synchronization network of experimental advanced superconducting tokamak (EAST) poloidal field (PF) power supply control system and analyzes characteristics of the IEEE 1588 in detail at the same time. IEEE 1588 has unique advantages for this paper to utilize as protocol in order to improve the accuracy of the time synchronization network (TSN). Consequently, we design a time synchronization system based on PCI eXtensions for Instrumentation-6683H. This paper utilizes one master node to synchronize all slave nodes on the time network in each second. The result shows that the synchronization accuracy can be achieved within 50 ns. According to results of experiments, this system can perfectly meet the requirements of EAST PF power supply control system. [ABSTRACT FROM AUTHOR]

The Divertor Tokamak Test (DTT) facility is proposed for studying power exhaust solutions as well as integrated physics and technology aspects for the demonstration power plant. To illuminate the richness of new novel plasma physics that can be explored in this device, linear stability properties and shear Alfvén fluctuation spectra of a typical DTT reference scenario are investigated by self-consistent hybrid magnetohydrodynamic-gyrokinetic simulations. The DTT core plasmas can be divided into two regions, characterized by reverse shear Alfvén eigenmode in the central core and by toroidal Alfvén eigenmode in the outer core region. The non-perturbative effect of energetic particles (EPs) and the wave-EP resonance condition as well as power transfer is analyzed in great detail, demonstrating the peculiar role played by EPs in multi-scale dynamics. The most unstable mode numbers of dominant Alfvénic fluctuations are shown to be of the order of 10, consistent with the typical orbit widths of the EPs normalized to the plasma minor radius and the DTT target design. [ABSTRACT FROM AUTHOR]

A space-resolved extreme ultraviolet (EUV) spectrometer has been used to observe the spatial distribution and temporal behavior of core impurity emissions at the experimental advanced superconducting tokamak (EAST), evaluating the effects of resonant magnetic perturbations (RMPs) on impurity concentration. The vertical profiles of impurity emissions from the 0 ≤ Z ≤ 450 mm range results in spectra in the wavelength range of 30 Å $\le \lambda \le 500\,{\mathring {\mathrm {A}}}$. Line emissions of impurity carbon, lithium, and iron were observed for C VI (33.7 Å, 2p–1s), Li III (135 Å, 2p–1s), and Fe XXIII (132.9 Å, 1s22s2p–1s22s2), respectively, whereas the unresolved transition array (UTA) of tungsten in the wavelength range of $\lambda = 48$ to 70 Å is monitored to study tungsten behavior with RMP. It is clearly observed that the intensities of C VI, Li III, Fe XXIII, and W-UTA emissions are reduced by the application of RMP fields, during stages of edge-localized modes (ELMs). In addition, the reduction in Fe and W emissions with $n = 1$ and 2 (n: toroidal mode number) RMP is more significant than that of carbon and lithium emissions, i.e., ~50% to 70% for Fe XXIII and W-UTA, and ~30% for C VI and Li III. Considering that the changes in $n_{e}$ and $T_{e}$ cannot account for the impurity emissions mitigation, the EUV measurements may indicate that RMP can significantly and selectively decrease the impurity accumulation in the core region with slight effects on the main plasma. Spatial distributions of impurity emissions are also measured for different impurity species using the EUV spectrometer, confirming the space-resolved diagnostic’s feasibility to study RMP effects on core impurity concentration. [ABSTRACT FROM AUTHOR]

Experimental advanced superconducting tokamak (EAST) is a D-shaped full superconducting tokamak device. Plasma-facing components (PFCs) of EAST lower divertor are most likely to be damaged during the operation, which become fail easily and have to be replaced and repaired frequently. To evaluate whether the divertor is able to reach the design objective, an assessment of technical risks by means of reliability, availability, maintainability, and inspectability (RAMI) for PFCs of EAST lower divertor has to be performed. A functional analysis is carried out in a bottom–up approach, which is described using the IDEFØ methodology. Based on the functional analysis, a failure mode, effect, and criticality analysis is performed to evaluate potential causes of failures and their effects. Criticality charts highlight the risk levels of different failure modes with regard to the probability of their occurrence and the impact on the operation, and actions are taken to mitigate the major risks. According to reliability block diagrams, the availability and the reliability of functions are calculated under stipulated operating condition. The RAMI analysis results provide the reference for the coming upgrade of EAST lower divertor and will be further optimized as the upgrade of the lower divertor. [ABSTRACT FROM AUTHOR]

JT-60SA is a superconducting tokamak in the framework of the broader approach agreement between Europe and Japan. For this international project among its various procurements, the Italian National Agency for new technology energy and sustainable economic development (ENEA) is providing four ac/dc converters for the central solenoid (CS) superconducting magnets, two ac/dc converters for the equilibrium field (EF) superconducting magnets, and two ac/dc converters for the fast plasma position control coils (FPPCC). Furthermore, ENEA is also providing six converter transformers of which four are of the dry type for FPPCC power supplies (PSs) and two of oil type for CS PSs. CS/EF PSs feed poloidal superconducting magnets of JT-60SA, whose purpose is the generation of strong magnetic fields able to induce and confine the plasma current in a prefixed geometrical configuration. The control system of CS/EF PSs has to be able to reproduce in real-time a reference current or voltage profile, known as “scenario” and these are particularly critical in these systems for the high currents required and for the high magnetic energy stored in the superconducting magnets. All performed type tests, carried out in accordance to the IEC60146 Standards, pointed toward a good dynamic behavior of the control system both for forward and for reverse rectifier bridges, assuring the smooth transitions between different operating modes: single mode, circulating current mode and dual mode. This is the focus of this paper. [ABSTRACT FROM AUTHOR]

Sufficient pumping speed and good pumping performance should be guaranteed for the HL-2M tokamak. In order to provide sufficient pumping capacity, two in-vessel cryopumps will be installed at outer floor and ceiling, respectively. Pumping performance of HL-2M in-vessel cryopump has been estimated by direct simulation Monte Carlo program and test particle Monte Carlo program under free molecular flow condition. The predicted pumping speeds of each bare cryopump are 51.29 m3/s for H2, 38.04 m3/s for D2, and 24.94 m3/s for He. An advanced divertor system will be installed, located on the floor of the HL-2M vessel. When the conductance of the divertor is taken into account, the pumping speeds of the floor ring are 20.11 m3/s for H2, 15.14 m3/s for D2, and 12.50 m3/s for He. This shows that the effective pumping speed of the cryopump is greatly affected by the conductance of the divertor. A parametric study has been carried out, allowing the sticking coefficients of some different gases to vary within a +/−0.05 range with respect to the corresponding reference values. The results indicate that the He pumping speeds may be significantly affected by a small change of the sticking coefficient, while negligible effects are predicted to occur, as far as H2 and D2 are concerned. The results of the dynamic evaluation of the pumping process indicate that the cryopump has a quick time response. [ABSTRACT FROM AUTHOR]

A significant engineering upgrade of the DIII-D neutral beam system has recently completed, providing ability to smoothly and independently vary the beam voltage and current. This enables greater control of the injected beam power, torque, and plasma instability drive for fusion experiments. Modifications to the high-voltage equipment and the plasma control system were made and tested over the last year to allow beam energy to vary by as much as 20 kV over a 0.5-s period anywhere within the 45–85 kV operating range of the beams. The beam current can be made to track the voltage (keeping the perveance constant), or the current and voltage can be varied independently (scanning the perveance). The smooth variation of beam energy avoids the extremely perturbative effects of pulsewidth modulation, the only tool previously used for regulating the injected neutral beam power. With independent control of voltage and current, the beam ion velocity space can be tailored to facilitate new experiments that explore, for example, the detrimental effects of Alfvén eigenmodes. These modes are driven by energetic particles and can possibly be avoided in steady-state scenarios by timely variation of the beam energy while maintaining constant input power. A description of the modifications made to the power supply and beam controls will be presented, as well as some initial physics results employing the new variable beam energy system. [ABSTRACT FROM AUTHOR]

The China fusion engineering test reactor (CFETR) could output 50–200 MW fusion power for demonstrating power generation. The overall structure of fusion power plant in concept design is similar to a pressurized-water reactor (PWR), except a doubly fed induction generator with flywheel (DFIGF) has been proposed to replace the traditional synchronous generator used in a PWR. The heat source of fusion plant is not constant in long-pulse operation mode of tokamak. Hence, it will lead to frequent inrush in synchronous generator and power grid. Although a high-power electric heating pressure regulator can mitigate the inrush, it consumes a lot of energy. The proposed DFIGF has good power regulation ability and large energy storage capacity, which suits the application when instability of heat source exists. The simulation model has been established and dynamic response analysis has demonstrated that DFIGF can ride-through the burning interval smoothly and provide a flexible connection with power grid. [ABSTRACT FROM AUTHOR]

Two blackbody sources permanently located within the international thermonuclear experimental reactor (ITER) diagnostic shield module at equatorial port 9 will operate in conjunction with two remotely retractable mirrors to generate and direct blackbody radiation to calibrate the radial and oblique views of the ITER electron cyclotron emission diagnostic. The main calibration requirements include a high-emissivity surface heated to 700 °C or higher, the ability to perform the calibration in high vacuum and in the presence of a magnetic field, and 5000-h operational lifetime over 20 years. Major constraints include a heater current limit of 40 A, no direct fluid cooling to mitigate failure risks, small size to avoid compromising neutron shielding, and adequate structural support to mitigate vibration loads. This paper describes the preliminary design of a hot calibration source supported by analysis and test results of a calibration source test prototype developed to address the calibration source requirements and its constraints. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

A large distributed data acquisition system based on PXI/PXIe technology has been implemented on EAST tokamak. Now the whole data system has more than 60 data acquisition units and more than 3000 raw channels, and the maximum data throughput is about 5GBytes/s. Most of the acquired data are continuously stored into MDSplus Wan et al. database, moreover the calculated and analyzed data are also saved into MDSplus. In case of long-term discharge mode, the raw data of one shot is more than 1TBytes, and the total data of one campaign is more than 100TBytes. MDSplus is helpful to store tokamak experiment discharge data, and the scientific data are saved into database by shot and tree, and it is easy to know the signal information of one shot or tree. However, it is difficult to get the general metadata information of all the shots and trees. Even if we can read the information by some scripts, but it is not so convenient and the data retrieval speed is slow. So we have planned to design a data statistics and retrieval system for EAST MDSplus data system, which can provide all the metadata information of all the trees such as signal name, sampling rate, data size, data available or not, and so on, basing on these metadata the system can give detailed statistics and reports automatically by some pre-defined model, which can help users to grasp the status and schema of all the data. The system has been developed based on apache/mysql/php framework and the PCIe SSD storage is adopted to promote the database access speed. The design details will be given in the paper. [ABSTRACT FROM AUTHOR]

We report our recent investigation on a passive mitigation strategy of the mirror deposition in magnetically confined fusion (MCF) devices, employing a newly designed duct system with baffles at KSTAR tokamak. Our mitigation strategy of the first mirror deposition is to suppress the deposition of impurity species onto the first mirror by confining inert gas such as helium in the duct with baffles located in front of the first mirror. To achieve this final goal, feasibility of this technique at a real tokamak was investigated with regard to the effect on the plasma condition in this paper. To assess the effect of the helium gas flow on the plasma condition, 5 sccm amount of helium gas was injected into the duct for about 70% of shots in the year 2016. This quantity of helium gas was found to be insignificant in terms of the effect on plasma performance. To estimate the deposition quantity on the sample during plasma operation in-situ, quartz crystal microbalances (QCMs) were also installed for thickness detection in real time at KSTAR. The net deposition rates of these samples at KSTAR were about 0.6–40 ng/h cm 2 (2.7 × 10 −3 –0.18 nm/h (graphite)) depending on the kinds of operations such as plasma shots, glow discharge cleaning, and baking of the tokamak first wall. We found that the most detrimental condition with regard to the mirror deposition rate at KSTAR is the glow discharge wall cleaning (GDC) in the baking condition. [ABSTRACT FROM AUTHOR]

Chinese Fusion Engineering Testing Reactor (CFETR) is a superconducting magnet Tokamak, which has the equivalent scale with complementary function to International Thermonuclear Experimental Reactor (ITER). The vacuum vessel (VV) which has a double-layer structure, Cooling water circulating through the double-layer structure will remove the heat generated during operation. The VV will provide a high-vacuum environment for the plasma, improve radiation shielding and plasma stability and provide support for in-vessel components. The CFETR VV is composed of 16 sectors, the angle of each sector is 22.5°. The Research and Development (R&D) of the key technologies to the VV manufacture have been carried out a few years ago by Institute of Plasma Physics Chinese Academy of Science (ASIPP), including Narrow-Gap welding, cutting and non-destructive testing, Poloidal Segment (PS) and sector assembly technologies, etc. ASIPP is constructing a 1/8 sector real size VV mock-up now. The manufacture of the PS for first 1/32 sector VV mock-up has been completed in 2015, the PS will be assembled into a whole this year. This paper will describe the study of the assembly technology for the CFETR 1/32 Vacuum Vessel mock-up. [ABSTRACT FROM AUTHOR]

Since 2006 lithium as an advanced plasma facing material has been tested on the Frascati Tokamak Upgrade (FTU). Lithium in the liquid phase acts both as plasma facing component, i.e. limiter, and plays also a role in plasma operation because by depositing a lithium film on the walls (lithization) oxygen is gettered. As in all deposition processes, even for the lithization, the presence of impurities in plasma phase strongly affects the properties of the deposited film. During the 2008 campaigns of FTU it was observed a strong release of carbon dioxide (during disruptions), resulting in successive serious difficulty of operation. In order to find the possible reactions occurred, we have analyzed the surface of two tiles of the toroidal limiter close to the Liquid Lithium Limiter (LLL). The presence of molybdenum oxides and carbides suggested that the surface temperatures could have exceeded 1000 K, likely during disruptions. lithium oxides and hydroxides have been found on the tiles and in the dust collected in the vessel, confirming the presence of LiO and LiOH and a not negligible concentration of Li 2 CO 3 especially at the LLL location. On the basis of the above results, we propose here a simple rationale, based on a two reactions mechanism, which can explain the formation of Li 2 CO 3 and its subsequent decomposition during disruption with release of CO 2 in the vessel. Admitting surface temperatures above 1000 K during a disruption, relatively high partial pressures of CO 2 are also predicted by the equilibrium constant for Li 2 CO 3 decomposition. [ABSTRACT FROM AUTHOR]

The superconducting tokamak JT-60SA is currently being assembled at the QST laboratories in Naka (Japan). Within the European contribution in the framework of the Broder Approach, Spain is responsible for providing JT-60SA cryostat. It is a stainless steel vacuum vessel which encloses the tokamak providing the vacuum environment. Due to functional purposes, the cryostat was divided into three large assemblies: the Cryostat Base, the Cryostat Vessel Body Cylindrical Section and the Top Lid. The second is an envelope made from SS304 that will be assembled by mechanical connection between the individual sectors. As part of the manufacturing process, dimensional inspections are carried out by laser tracker to check the tolerances of the pieces. Due to the high mechanical flexibility of the sectors, the way to support the pieces resulted very critical for the inspections as it was predicted by FEA carried out. The paper summarizes the measurement procedure for the dimensional inspections as well as the pre-assembly procedure of the whole component. [ABSTRACT FROM AUTHOR]

As part of the conceptual design studies for a European DEMO, different tokamak geometries are being considered. As identified in the EFDA Roadmap to the realisation of Fusion Energy, 2013: “ The integration of the Remote Maintenance system within the DEMO plant is an essential task within the DEMO CDA phase. This will involve establishing requirements, functions and interfaces with many other systems to ensure that plant availability and maintainability are considered from the outset. ” In order to fulfil this integration requirement different geometries of DEMO have been assessed for their impact on remote maintenance (RM). The aspect ratio and number of TF coils have been identified as the pivotal variables, driving the tokamak geometry, with significant effects on remote maintenance both in terms of technical feasibility and speed of operation. Tokamak geometries with aspect ratios from 2.6 to 4.0 and 16 or 18 TF coils have been compared. The results of this evaluation show that higher aspect ratios and lower numbers of TF coils are beneficial to RM both in terms of technical feasibility and speed of operations. To deliver a maintainable DEMO, efforts must be made to maximize its aspect ratio and minimize the number of TF coils. [ABSTRACT FROM AUTHOR]

As a major in-vessel component of a tokamak-type fusion reactor, the divertor is mainly in charge of removal of particles and partial power exhaust via scrape-off layer. The target plate of the divertor is directly exposed to non-uniform high heat flux on the surface by particle bombardment and radiation. In the case of ITER and a DEMO reactor, the peak surface heat flux is expected to reach up to 10 MW/m 2 during normal operation and 20 MW/m 2 during slow transient events like loss of plasma detachment. This paper reports the results of a preliminary code-based design study and fabrication technology verification test which were conducted for developing an ITER-like divertor target design for the DEMO divertor. The structural failure evaluation against the ratchetting and fatigue criteria of the ITER SDC-IC showed that the design with reduced dimensions would allow sufficient design margin (reserve factor) for three distinct thermal loading cases. The first trial of mock-up fabrication using a new joining furnace at ENEA was successfully completed. The ultrasonic inspection test made before and after the cyclic HHF tests at GLADIS facility demonstrated high quality of fabrication and robust design concept. [ABSTRACT FROM AUTHOR]

ASDEX Upgrade (AUG) is the only tokamak in Europe to have low activation ferritic steel as the inner vessel wall facing component. Together with the massive tungsten tiles in the lower divertor, AUG is the tokamak with the closest DEMO wall. The project is a first step towards the extensive use of ferritic steel in future fusion reactors. For example, the test blanket module of ITER is planned to have a ferritic steel wall and thick tungsten tiles as a plasma facing component. The ‘ad hoc’ ferritic steel built with low activation capability is known as Eurofer. As the low activation property is not a requirement for AUG, the material selected for the project is the martensitic steel P92 which is the most similar material to Eurofer from a magnetic point of view. The purpose of the project is to improve understanding of the magnetic perturbation of the ferritic steel both on the plasma and magnetic probes, evaluating and controlling these effects. Additionally, the effect of the additional forces on the supporting structure has been addressed. Bearing this in mind, in 2013 a step wise program has been started and part of the W coated graphite tiles in the region of the inner column were replaced by steel tiles [1]. The first campaign did not suffer any particular issue related to the new material. According to the calculations, the plasma was almost unperturbed, thanks also to the toroidal symmetry of the tiles inside the vessel, and the magnetic probe measurements were properly corrected [2]. Inspection of the machine pointed out some hardware problems. The graphite tiles adjacent to the steel tiles were damaged. The graphite tiles had broken edges in 5 from 64 positions and notches in many others. The coating of the graphite and steel tiles, made of tungsten and TiO respectively, was damaged. At first glance it was clear that the steel tiles were moving but it was definitely unexpected. In understanding the process, the location of the damage was the crucial hint. In fact all failures were located at the boundary between 2 vacuum vessel octants. To justify this failure mode inside the vessel, a hypothesis (about current flowing in the heat shield supporting structure) was made and FEM analyses were carried out in this direction. With extreme caution, in 2015 just one additional row of steel tiles was added together with diagnostics that confirmed the hypothesis. Now that a clear understanding of the problem has been reached, the project to add further rows of steel tiles can be continued. For the next campaign it is planned to replace all the tiles in the middle region of the heat shield together with stiffening and modification of the supporting structure. In this paper the learning process from the damage of the tiles and its causes, from the FEM analysis results to the data diagnostics will be reported. The future plans for steel tiles in AUG will be discussed. [ABSTRACT FROM AUTHOR]

The Tore Supra tokamak is being transformed in an x-point divertor fusion device in the frame of the WEST (W-for tungsten-Environment in Steady-state Tokamak) project, launched in support to the TER tungsten divertor strategy. The WEST project aims to test W monoblock Plasma Facing Units (PFU) under long plasma discharge (up to 1000 s), with thermal loads of the same magnitude as those expected for ITER. Therefore the divertor is a key component of the WEST project, and so is its support structure, which has to handle strong mechanical loads. The WEST upper and lower divertor are made of 12 30° sectors, each one composed of 38 PFU that can be made of tungsten, CuCrZr or graphite. A generic 316L stainless steel 30° conic support plate is used to hold the 38 PFU together, regardless of their material. The PFUs are fixed on the support plate thanks to 152 Xm19 stainless steel fixing elements (4 per PFU), and in each of this fixing element an Aluminium-Nickel-Bronze alloy (Al-Ni-Br) pin is engaged in a slotted hole, in order to allow thermal expansion in the length direction of the PFU. The support plate is fixed on the divertor coil casing thanks to 10 M10 screws. Mechanicals loads which act on the PFUs are transmitted to the support plate through the fixing elements. These loads are due to Vertical Displacement Event (VDE), disruptions and thermal expansion of the PFU. First the different load cases, PFU configurations and scenario are presented. Then an ANSYS plastic mechanical simulation is performed in order to validate the number of cycles of the support plate for each scenario: 30,000 cycles in steady-state and 3000 cycles in VDE. Finally reactions forces from the previous ANSYS simulation are used in order to calculate the stress in the M10 screws. [ABSTRACT FROM AUTHOR]

The Helium Cooled Pebble Bed (HCPB) Breeding Blanket (BB) is one of the 4 BB concepts being investigated in the EU for their possible implementation in DEMO. During 2014 the former “beer-box” BB concept based on the ITER’s HCPB Test Blanket Module suffered several design changes so as to meet the different counteracting nuclear, thermohydraulic and thermomechanical requirements. These studies evidenced that the concept is too rigid to meet the tight TBR requirements imposed for the EU DEMO (i.e. TBR ≥ 1.10). Additionally, the complex manifold system with unbalanced helium mass flow in each of the 2 parallel cooling loops made the concept thermohydraulically complex. However, parametric studies during 2015 revealed that the HCPB concept have potential for a better nuclear performance, as well as margin for a significant simplification of the cooling internals by redefining the cooling plates and the architecture of the blanket, building a symmetric flow scheme. This paper describes the new HCPB concept based on an integrated FW with the breeding zone thermohydraulics and helium manifold systems. The former complex manifold backplates have been compacted and integrated in the cooling plates, releasing ≈300 mm of radial space that can be used now to increase breeder zone, the neutron shielding, to reinforce the Back Supporting Structure (BSS) or basically to reduce the reactor size. Detailed neutronic analyses have yielded a TBR of ∼1.20 for the baseline design. Initial analyses show a correct thermohydraulic behavior. Preliminary thermomechanical analyses also indicate that the design can potentially withstand an in-box LOCA at 9 MPa at a level C according to the RCC-MRx code. Future consolidation activities are described, which shall lead to a concept meeting the BB requirements. [ABSTRACT FROM AUTHOR]

Tokamak plasmas with high fusion performances are prone to magnetohydrodynamic (MHD) low m,n instabilities which may affect the energy and particle confinement time and possibly lead to disruptive plasma termination. In presently operating tokamaks high space resolution (∼2 cm) and high time resolution (0.01–0.1 ms) electron cyclotron emission (ECE) diagnostics are embedded in the control loop finalized to MHD control, often in synergy with pick-up coils sensitive to the magnetic fluctuations. Microwave diagnostics have plasma facing components (antennas) that are electrically passive, have metal body and are mechanically fixed. Such characteristics of robustness and reliability are a promising feature but the diagnostic performances in term of achievable space resolution and profile coverage for prompt detection of instabilities in the high temperature and density conditions of the DEMO reactor need to be assessed. The results of such assessment based on the EU-DEMO1-2015 scenario show that extraordinary mode ECE in 2nd harmonics seems to have enough space resolution in the region interested by 3/2 and 2/1 NTMs. [ABSTRACT FROM AUTHOR]

DEMO will be a burning fusion device, most likely a tokamak type and thus, the plasma needs to be re-fuelled. A project was launched in the European DEMO Programme to develop concepts for the matter injection. This approach and the actual status of the project are presented. In a first step, requirements for actuators on plasma core density were elaborated. Related modelling activities indicated that only sufficiently deep fuel deposition can achieve target operational parameters. Hence, suitable techniques had to be identified and evaluated with respect to their availability and capability. Several techniques for pellet injection have been benchmarked in view of the defined requirements. Finally, cryogenic pellet injection was chosen as the main actuator on plasma core density. From further modelling activities, assuming for the pellet mass the ITER reference value, required launching speeds were derived, with respect to different injection geometries. Gas puffing and the respective technical system are necessary for pre-fill, ramp-up and plasma confinement enhancement. The ITER gas injection system (GIS) is assessed in view of suitability for DEMO. The piping system and the manifold concept can be adopted. The Gas Valve Box (GVB) is considered not to be an optimum solution for DEMO. Instead of this GVB, a pressure based RUN/VENT flow regulation and injection system is proposed in order to meet DEMO requirements. The principle of this system is described as well as some considerations about injection locations; further orienting gas flux numbers are provided. [ABSTRACT FROM AUTHOR]

To carry out research related to electron cyclotron waves, 6 MW ECRH system including four 105 GHz/1 MW/2s and two 140 GHz/1 MW/3s units will be developed on the HL–2M tokamak being built in the first stage. Dual-frequency transmission lines with same components for the 105 GHz and 140 GHz systems are designed to make the fabrication easier. The corrugated waveguides are used to ensure the bandwidth of the transmission systems. In addition, the parameters of polarizers are specially considered to obtain the required polarizations at the two different frequencies. The transmission efficiency is analyzed with the multimode propagation model, and also the polarization characteristics of polarizers are calculated for the two different frequencies. The results indicate that both the transmission efficiency and the polarization characteristics can meet the requirements of ECRH system on HL–2M. [ABSTRACT FROM AUTHOR]

The system of diagnostics, data acquisition and control foreseen on the Divertor Test Tokamak experiment (DTT) is presented. Conceived in an integrated way, the control system meets the specifications of a fusion experiment devoted to the study of the power exhaust problem in view of DEMO. Diagnostics and feedback control are particularly functional to the need of maintaining the plasma close to equilibrium in situations prone to instabilities where the plasma wall interaction is optimized. Strongly oriented to the exploration of control methods suitable for DEMO, DTT will specifically experiment on physics and engineering model based control systems. Control and data flow schemes are inspired by those of ITER. [ABSTRACT FROM AUTHOR]

Materials exposed to plasma may undergo various forms of surface modifications. Among the important phenomena for tungsten - as the prime candidate plasma-facing material for fusion devices – is a formation of helium-induced fibreform nanostructure, so-called tungsten fuzz. In this paper, we report direct observations of the interaction of the pre-prepared fuzzy tungsten surfaces with ELMy H-mode plasmas in the COMPASS tokamak as well as consequent ex-situ morphological analyses, with a particular focus on arcing as a potential erosion mechanism. Arcing events are documented from high-speed camera imaging. The sample surfaces are examined by scanning electron microscopy. Arc traces were observed on all samples, while their number was dependent on the sample position and orientation. Inside the arc traces, localized melting and densification of the original fuzz was observed, resulting in thickness reduction. The modified structure still retained some porosity and did not extend into the bulk. [ABSTRACT FROM AUTHOR]

Quantitative comparisons are presented between gyrokinetic simulations and experimental values of the carbon impurity peaking factor in a database of JET H-modes during the carbon wall era. These plasmas feature strong NBI heating and hence high values of toroidal rotation and corresponding gradient. Furthermore, the carbon profiles present particularly interesting shapes for fusion devices, i.e., hollow in the core and peaked near the edge. Dependencies of the experimental carbon peaking factor (R=LnC ) on plasma parameters are investigated via multilinear regressions. A marked correlation between R=LnC and the normalised toroidal rotation gradient is observed in the core, which suggests an important role of the rotation in establishing hollow carbon profiles. The carbon peaking factor is then computed with the gyrokinetic code GKW, using a quasi-linear approach, supported by a few non-linear simulations. The comparison of the quasi-linear predictions to the experimental values at mid-radius reveals two main regimes. At low normalised collisionality, v*, and Te=Ti < 1, the gyrokinetic simulations quantitatively recover experimental carbon density profiles, provided that rotodiffusion is taken into account. In contrast, at higher v* and Te=Ti > 1, the very hollow experimental carbon density profiles are never predicted by the simulations and the carbon density peaking is systematically over estimated. This points to a possible missing ingredient in this regime. [ABSTRACT FROM AUTHOR]

Orbit following Monte Carlo code (ASCOT) calculations of neutral beam (NB) induced torque, current and heating have been performed for a selected set of plasma scenarios to optimise the neutral beam injection system for the proposed high magnetic field spherical tokamak ST40. It is found that there are strong variations especially in the current drive efficiency and in the toroidal torque depending on the NB alignment and injection energy. The optimal alignments as well as the optimal injection energy depend both on the scenario and whether the current drive or the toroidal torque is to be maximised. Due to the relatively small calculated current drive efficiency (0.05–0.2 MA/MW) we find that for the studied scenarios a relatively central low field side (LFS) deposition below midplane provides the best overall performance still at a tolerable shine through power. [ABSTRACT FROM AUTHOR]

A control-oriented, two-timescale, linear, dynamic, response model of the rotational transform ι profile and the normalized beta β N is proposed based on experimental data from the DIII-D tokamak. Dedicated system-identification experiments without feedback control have been carried out to generate data for the development of this model. The data-driven dynamic model, which is both device-specific and scenario-specific, represents the response of the ι profile and β N to the electric field due to induction as well as to the heating and current drive (H&CD) systems during the flat-top phase of an H-mode discharge in DIII-D. The control goal is to use both induction and the H&CD systems to locally regulate the plasma ι profile and β N around particular target values close to the reference state used for system identification. A singular value decomposition (SVD) of the plasma model at steady state is carried out to decouple the system and identify the most relevant control channels. A mixed-sensitivity robust control design problem is formulated based on the dynamic model to synthesize a stabilizing feedback controller without input constraints that minimizes the reference tracking error and rejects external disturbances with minimal control energy. The feedback controller is then augmented with an anti-windup compensator, which keeps the given controller well-behaved in the presence of magnitude constraints in the actuators and leaves the nominal closed-loop system unmodified when no saturation is present. The proposed controller represents one of the first feedback profile controllers integrating magnetic and kinetic variables ever implemented and experimentally tested in DIII-D. The preliminary experimental results presented in this work, although limited in number and constrained by actuator problems and design limitations, as it will be reported, show good progress towards routine current profile control in DIII-D and leave valuable lessons for further advancements in the field. [ABSTRACT FROM AUTHOR]

The high power loads impinging on the first wall and particularly the divertor of fusion reactors is a decisive factor to the success of nuclear fusion. An alternative to solid plasma facing components is the use of liquid metals such as lithium or tin due to the regenerative properties of the liquid surface. Another suitable candidate is the eutectic lithium tin alloy (30 at.% Li) which is suggested to display beneficial properties of both its constituent elements. The application of these materials as liquid metal plasma facing components depends on several factors such as their affinity to retain hydrogenic isotopes and the discharge performance degradation induced by the enhanced impurity contamination, among others. An experimental setup has been developed to produce and expose samples to ISTTOK plasmas on both liquid and solid states. Samples of Li-Sn alloy were exposed at ISTTOK to deuterium plasmas. Post-mortem analysis of the samples was performed by means of ion beam diagnostics. To quantify the fuel retention on the samples the nuclear reaction analysis (NRA) technique was applied. Complementary, Rutherford backscattering spectrometry (RBS) was used for determination material composition, particularly of impurities, on the samples. Regardless of the high sensitivity of these techniques no deuterium was detected in the samples. Emission of the Li-I 670.7 nm line indicates that there was interaction of the plasma with the samples. Alternative reasons for the low retention of this material are discussed. Lithium segregation to the surface of the sample was observed. [ABSTRACT FROM AUTHOR]

Analytical solutions of electromagnetism driven Grad-Shafranov equation (GSE) can be used for theoretical studies of plasma equilibrium, transport and Magnetohydrodynamic stability of tokamaks. In this research, an extended analytic solution to the Grad-Shafranov equation was presented. The solution describes standard tokamaks configurations. It allows the simulation magnetic surfaces of plasmas with elongation and triangularity, with an independent choice of pressure and plasma current. We have determined equilibrium magnetic surfaces in IR-T1 tokamak. With applying these solutions to IR-T1 tokamak, a small, air core, low beta and large aspect ratio tokamak with a circular cross section, we have calculated poloidal magnetic flux. Due to the generality and high accuracy of the first exact solution for all of the magnetic configurations of interest, the result of this solution for IR-T1 tokamak is acceptable equilibrium compare to the second one for this tokamak. These analytical solutions can be used as an input of feedback control system. [ABSTRACT FROM AUTHOR]

The role of the bootstrap and polarization currents for the stability of neoclassical tearing modes is investigated employing both a drift kinetic and a gyrokinetic approach. The adiabatic response of the ions around the island separatrix implies, for island widths below or around the ion thermal banana width, density flattening for islands rotating at the ion diamagnetic frequency, while for islands rotating at the electron diamagnetic frequency the density is unperturbed and the only contribution to the neoclassical drive arises from electron temperature flattening. As for the polarization current, the full inclusion of finite orbit width effects in the calculation of the potential developing in a rotating island leads to a smoothing of the discontinuous derivatives exhibited by the analytic potential on which the polarization term used in the modeling is based. This leads to a reduction of the polarization-current contribution with respect to the analytic estimate, in line with other studies. Other contributions to the perpendicular ion current, related to the response of the particles around the island separatrix, are found to compete or even dominate the polarization-current term for realistic island rotation frequencies. [ABSTRACT FROM AUTHOR]

T-11M lithium program is focused on a solution of technological issues of a steady-state tokamak with liquid lithium plasma facing components (PFC). Lithium, collected by the chamber wall of such tokamak is able to capture a considerable amount of tritium, which is unacceptable. In order to restrict the level of lithium deposited on the chamber wall and captured tritium it was suggested early to use a cryogenic target technique. Such target placed in the plasma of glow discharge (GDH, He or Ar) during the tokamak conditioning can play the role of collector of lithium and tritium atoms which were sputtered by GD bombardment of the wall. The collected lithium and tritium can be evacuated mechanically together with target from tokamak chamber through vacuum lock without venting. Cryogenic target, cooled by liquid nitrogen (LN), was installed in the T-11M and tested in different modes of wall conditioning and tokamak operations. The maximum speed of the lithium collection during GDH was 3.5 mg/h, that corresponds “to contamination” of wall by lithium during approximately 200 regular shots of T-11M which are equivalent to two-week regular operations. It was established that considerable part of lithium was collected in ionized state. On this basis it can be suggested the creation in tokamak chamber an equivalent ionic pump for extraction both lithium and tritium from chamber without venting during regular tokamak operation. [ABSTRACT FROM AUTHOR]

Dust production and its transport into the core plasma is an important issue for magnetic confinement fusion. Dust grains are charged by various processes, such as the collection of plasma particles and electron emissions, and their charge influences the dynamics of the dust. This paper presents the results of calculations of the surface potential of dust grains in a Maxwellian plasma. Our calculations include the charging balance of a secondary electron emission (SEE) from the dust. The numerical model that we have used accounts for the influence of backscattered electrons and takes into account the effects of grain size, material, and it is also able to handle both spherical and non-spherical grains. We discuss the role of the SEE under tokamak conditions and show that the SEE is a leading process for the grains crossing the scrape-off layer from the edge to core plasma. The results of our calculations are relevant for materials related to fusion experiments in ITER. [ABSTRACT FROM AUTHOR]

Energy and pitch angle resolved measurements of escaping neutral beam ions (E ≈ 80 keV) have been made during DIII-D L-mode discharges with applied, slowly rotating, n = 2 magnetic perturbations. Data from separate scintillator detectors (FILDs) near and well below the plasma midplane show fast-ion losses correlated with the internal coil (I-coil) fields. The dominant fast-ion loss signals are observed to decay within one poloidal transit time after beam turn-off indicating they are primarily prompt loss orbits. Also, during application of the rotating I-coil fields, outboard midplane edge density and bremsstrahlung emission profiles exhibit a radial displacement of up to δR ≈ 1 cm. Beam deposition and full orbit modeling of these losses using M3D-C1 calculations of the perturbed kinetic profiles and fields reproduce many features of the measured losses. In particular, the predicted phase of the modulated loss signal with respect to the I-coil currents is in close agreement with FILD measurements as is the relative amplitudes of the modulated losses for the co and counter-current beam used in the experiment. These simulations show modifications to the beam ion birth profile and subsequent prompt loss due to changes in the edge density; however, the dominant factor causing modulation of the losses to the fast-ion loss detectors is the perturbed magnetic field (δB/B ≈ 10−3 in the plasma). Calculations indicate total prompt loss to the DIII-D wall can increase with application of the n = 2 perturbation by up to 7% for co-current injected beams and 3% for counter-current injected beams depending on phase of the perturbation relative to the injected beam. [ABSTRACT FROM AUTHOR]

A simple dynamical cascade model for the evolution of free energy is considered in the context of gyrokinetic formalism. It is noted that the dynamics of free energy, that characterize plasma micro-turbulence in magnetic fusion devices, exhibit a predator–prey character. Various key features of predator–prey dynamics such as the time delay between turbulence and large-scale flow structures, or the intermittency of the dynamics are identified in the quasi-steady-state phase of the nonlinear gyrokinetic simulations. A novel prediction on the ratio of turbulence amplitudes in different parts of the wavenumber domain that follows from this simple predator–prey model is compared to a set of nonlinear simulation results and is observed to hold quite well in a large range of physical parameters. Detailed validation of the predator–prey hypothesis using nonlinear gyrokinetics provides a very important input for the effort to apprehend plasma micro-turbulence, since the predator–prey idea can be used as a very effective intuitive tool for understanding and designing efficient transport models. [ABSTRACT FROM AUTHOR]

A semianalytic expression for the edge plasma angular toroidal rotation frequency that includes the neoclassical toroidal viscosity braking influence is obtained. Based on the model presented in a previous paper [I.G. Miron, Contrib. Plasma Phys. 53, 214 (2013)], the less destabilizing error field spectrum is found in order to minimize the nonlinear effect of the NTV on the toroidal rotation of the edge of the plasma. (© 2013 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim) [ABSTRACT FROM AUTHOR]

The paper reports on experimental studies of electron beams in the ISTTOK tokamak, those were performed by means of an improved four-channel detector. The Cherenkov-type detector measuring head was equipped with four radiators made of two types of alumina-nitrate (AlN) poly-crystals: machinable and translucent ones, both of 10 mm in diameter and 2.5 mm in thickness. The movable support that enabled the whole detectors to be placed inside the tokamak vacuum chamber, at chosen positions along the ISTTOK minor radius. Since the electron energy distribution is one of the most important characteristics of tokamak plasmas, the main aim of the study was to perform estimations of an energy spectrum of the recorded electrons. For this purpose the radiators were coated with molybdenum (Mo) layers of different thickness. The technique based on the use of Cherenkov-type detectors enabled the detection of fast electrons (of energy above 66 keV) and determination of their spatial and temporal characteristics in the ISTTOK experiment. Measurements of hard X-rays (HXR), which were emitted during ISTTOK discharges, have also been performed. Particular attention was paid to the correlation measurements of HXR pulses with run-away electron beams. (© 2013 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim) [ABSTRACT FROM AUTHOR]

In parallel to the direct contribution to the procurement phase of ITER and Broader Approach, CEA has initiated research & development programmes, accompanied by experiments together with a significant modelling effort, aimed at ensuring robust operation, plasma performance, as well as mitigating the risks of the procurement phase. This overview reports the latest progress in both fusion science and technology including many areas, namely the mitigation of superconducting magnet quenches, disruption-generated runaway electrons, edge-localized modes (ELMs), the development of imaging surveillance, and heating and current drive systems for steady-state operation. The WEST (W Environment for Steady-state Tokamaks) project, turning Tore Supra into an actively cooled W-divertor platform open to the ITER partners and industries, is presented. [ABSTRACT FROM AUTHOR]