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AbstractA hard X-ray pinhole camera system has been recently built at the HL-2A tokamak to measure the evolution of space-time distribution of fast electrons in the energy range of 20 to 200 keV. The camera is mainly composed of a fan-shaped detector array, an observation window, a pinhole mechanism, and a data processing system. The detector array consists of 21 CdTe detectors that are arranged in a poloidal section. The camera views the plasma perpendicularly through an observation window mounted in a horizontal port on the equatorial plane. The data processing is implemented by a fast spectrometry based on field-programmable gate array technology. The time and space resolution of the camera can reach 2 to 16 ms and 2 cm, respectively. During the HL-2A experiment campaign in 2018, measurements of fast electrons produced by lower hybrid waves using the camera were successfully performed. The performance of the camera and the first experimental results with some discussions are presented in this paper.

AbstractTomography inversion has been used routinely for processing outputs of plasma radiation diagnostics. Various tomographic algorithms have been developed, with those based on Tikhonov regularization being among the fastest while still providing reliable results. This paper presents a further speed optimization of the minimum Fisher Tikhonov regularization algorithm based on reducing iteration cycles used during the calculation. Ten to twentyfold speed gain is achieved compared to the original implementation. Robustness of the new method is demonstrated using both artificially generated data sets and real data from the soft X-ray diagnostics at the COMPASS tokamak. The advantage gained by the optimization is investigated in particular with respect to the possibility of real-time control of the plasma position; the option of impurity control is also discussed.

AbstractA new hard X-ray (HXR) camera system has been planned to be developed for HL-2A tokamak (R0= 1.65 m, a = 0.4 m, Bt= 2.8 T, and Ip= 0.5 MA), which is dedicated to the tomography of fast electron bremsstrahlung emission in the energy range 10 to 200 keV. The camera system includes two independent HXR cameras, which are both located in the same poloidal plane. Each camera is made up of 30 detection chords and views the whole poloidal cross section of the plasma. The spatial and temporal resolutions of the camera are 2 to 3 cm and 1 to 2 ms, respectively. HXR detection is performed using cadmium telluride (CdTe) semiconductors. Both simulation and experimental results suggest that an Al foil with a 0.3-mm thickness is the best candidate for filtering the low-energy X-ray photons. Powerful inversion techniques are employed to obtain the local HXR profiles as functions of time and photon energy. The HXR camera system planned for HL-2A tokamak is presented in detail.

AbstractTo obtain a fast estimation of the total impurity density distribution in a poloidal cross section from soft X-ray (SXR) measurements during quasi-stationary phases, the possibility that ionization equilibrium may have little influence on the emissivity profile of Ni and Fe in the core region of tokamak plasmas is investigated. Preliminary and encouraging results that support this assumption under certain conditions are found. A simplified approach aimed at computing a satisfactory estimation of the total density of a unique and identified impurity directly from an absolutely calibrated SXR tomographic inversion is implemented. An example of application to a previously and independently performed transport simulation of a Ni injection in Tore Supra is then given.

AbstractThe real-time reconstruction of the plasma magnetic equilibrium in a tokamak is a key point to access high-performance regimes. Indeed, the shape of the plasma current density profile is a direct output of the reconstruction and has a leading effect for reaching a steady-state high-performance regime of operation. The challenge is thus to develop real-time methods and algorithms that reconstruct the magnetic equilibrium from the perspective of using these outputs for feedback control purposes. In this paper the validation of the JET real-time equilibrium reconstruction codes using both a Bayesian approach and a full equilibrium solver named Equinox will be detailed, the comparison being performed with the off-line equilibrium code EFIT (equilibrium fitting) or the real-time boundary reconstruction code XLOC (X-point local expansion). In this way a significant database, a methodology, and a strategy for the validation are presented. The validation of the results has been performed using a validated database of 130 JET discharges with a large variety of magnetic configurations. Internal measurements like polarimetry and motional Stark effect have been also used for the Equinox validation including some magnetohydrodynamic signatures for the assessment of the reconstructed safety profile and current density.

AbstractIn this paper we propose a soft-X-ray method to characterize dust accumulation or layer formation on a given substrate. The method determines the differential absorption based on the X-ray lines emitted from the substrate by fluorescence as a result of film or powders deposited on the substrate surface. We have chosen to use molybdenum as the material for the substrate because it is used in present-day tokamaks and it is being considered as material for the first mirror. It also offers the advantage of having two strong lines, well separated in energy: the L-shell emissions centered at [approximately]2.3 keV and the Kαlines at [approximately]17.4 keV. The transparency of the layer can be then measured at 2.3 keV, provided the Kαline is unaffected. The feasibility of the proposed method was clearly demonstrated in laboratory experiments, providing estimations of the thicknesses that can be detected, for a number of relevant elements for fusion devices (Be, C, Fe, and W).

AbstractThe Tore Supra tokamak is devoted to long-duration, high-performance plasma research. The real-time measurements and control (RTMC) system has been developed to address the basic tokamak controls, such as plasma equilibrium and density control. Over the years, more and more sophisticated and demanding controls have been implemented, allowing the improvement of plasma performance and machine protection. This includes current profile control to enhance the performance and to avoid magnetohydrodynamic instabilities, infrared monitoring of plasma-facing components to prevent overheating, and disruption detection and mitigation techniques. Most of these improvements are relevant to the plasma operation in a full steady-state regime. This paper describes the present status of the Tore Supra RTMC system, detailing recent progress and highlighting the advantages of the various control schemes implemented so far.

AbstractRealizing high-power long-duration discharges puts specific constraints on diagnostics: Their front parts have to withstand important thermal loads, eventually requiring active cooling of critical parts, and drifts in measurements have to be avoided in order to supply reliable measurement during the whole discharge duration. Furthermore, the importance of diagnostics for missions other than physics understanding, such as machine operation or safety control, increases. The diagnostics system of Tore Supra consists of roughly 30 diagnostics, covering a large range of plasma parameters from the core to the edge. They have been designed for long-duration plasma discharges, which can last up to 1000 s. Their inner components have been dimensioned to endure continuous high-radiation fluxes, and most of them have been conceived to give a fair measurement all along the discharges.

AbstractLow-frequency (5- to 20-kHz) and high-frequency (40- to 200-kHz) modes are studied during radio-frequency heating experiments on the Tore Supra tokamak by means of correlation electron cyclotron emission. High-frequency modes are detected when the plasma is heated by ion cyclotron range of frequency waves in the minority D(H) heating scheme in combination with lower hybrid current drive (LHCD) producing a flat or slightly reversed q-profile. They are identified as Alfvén cascade modes. When this mode is triggered, fast ion losses (<20%) are detected from the neutron emission rate, and an additional heat load on plasma-facing components can be measured by an infrared camera when the fast ion energy is sufficiently large. Low-frequency modes are commonly triggered during LHCD experiments performed at low loop voltage. This mode can be observed with moderate lower hybrid power when the q-profile is monotonic or at higher power when the q-profile is flat in the core (r/a< 0.2) or reversed. It is identified, in most cases, as an electron fishbone-like mode. These modes can be stabilized by a slight modification of the q-profile provided by an increase of lower hybrid power or by a small addition of electron cyclotron current device.

Active feedback control for regulation of the safety factor (q) profile at the start of the high stored energy phase of an advanced tokamak discharge has been demonstrated in the DIII-D tokamak. The time evolution of the on-axis or minimum value of qis controlled during and just following the period of ramp-up of the plasma current using electron heating to modify the rate of relaxation of the current profile. In L-mode and H-mode discharges, feedback control of qis effective with the appropriate choice of either off-axis electron cyclotron heating or neutral beam heating as the actuator. The qprofile is calculated in real time from a complete equilibrium reconstruction fitted to external magnetic field and flux measurements and internal poloidal field measurements from the motional Stark effect diagnostic.

Results from an extensive profile database analysis of JET density profiles in H-mode show that the density peaking factor ne0/ < ne> in JET H-modes increases as the effective collisionality drops from ~1 at mid-radius to below 0.1 as expected for ITER. Density peaking is also strongly correlated with the Greenwald number NG, the particle outward flux ? from the neutral beam source and Ti/Te. The correlations with li, q95, ?N, ?*, LTe, LTi, the toroidal Mach number and its shear are weak or insignificant. H- modes heated only by ICRH are, on average, only slightly less peaked than H-modes dominated by NBI, demonstrating that neutral beam fuelling can only explain a modest part (~20%) of the peaking. Scaling expressions involving ?eff, NG, R?/(ne?) and Ti/Tesuggest that ne0/ may exceed 1.5 in ITER, providing a boost of fusion power of more than 30% for fixed ? and average density with respect to the usual assumption of a flat density profile.

Active control of plasma profiles is an essential requirement for operating within plasma stability limits, for steady-state operation and for optimization of the plasma performance. In DIII-D, plasma profiles have been actively controlled using various actuators in the following manner: (a) real time closed loop control of the qprofile evolution using electron cyclotron heating and neutral beam injection as actuators; (b) active control of the density and pressure profiles in quiescent H-mode and quiescent double barrier plasmas using electron cyclotron current drive (ECCD) and pellet injection; (c) active control of the edge profiles to suppress edge localized modes using resonant magnetic perturbation with toroidal mode number n= 3, (d) real time control of the current density profile to suppress neoclassical tearing modes using localized deposition of co-ECCD.

Recent research on advanced tokamak in JET has focused on scenarios with both monotonic and reversed shear q-profiles having plasma parameters as relevant as possible for extrapolation to ITER. Wide internal transport barriers (ITBs), r/a~ 0.7, are formed at ITER relevant triangularity ? ~ 0.45 and moderate plasma current, IP= 1.5-2.5?MA, with ne/nG~ 60% when ELMs are moderated by Ne injection. At higher current (IP? 3.5?MA, ? ~ 0.25) wide ITBs sitting at r/a? 0.5, in the positive shear region, have been developed. Generally MHD events terminate these barriers otherwise limited in strength by power availability. ITBs with core density close to Greenwald value, Te~ Tiand low toroidal rotation (4 times lower than standard ITBs) are obtained in plasma target preformed by opportune timing of lower hybrid current drive (LHCD), pellet injection and a small amount of NBI power. Wide ITBs, r/a~ 0.6, of moderate strength, can be sustained without impurities accumulation for a time close to neoclassical resistive time in 3?T/1.8?MA discharges that exhibit reversed magnetic shear profiles and type-III ELMy edge. These discharges have been extended to the maximum duration allowed by JET subsystems (20?s) bringing to the record of injected energy in a JET discharge: E~ 330?MJ. Portability of ITB physics has been addressed through dedicated similarity experiments. The ITB is identified as a layer of reduced diffusivity studying the propagation of the heat wave generated by modulating the ICRF mode conversion (MC) electron heating. Impressive results, QDT~ 0.25, are obtained in these deuterium discharges with 3He minority when the MC layer is located in the core. The ion behaviour has been investigated in pure LHCD electron ITBs optimizing the 3He minority concentration for direct ion heating. Preliminary results of particle transport, studied via injection of a trace of tritium and an Ar-Ne mixture, will be presented.

The recent success in coupling lower hybrid (LH) waves in high performance plasmas at JET together with the first demonstration on FTU of the coupling capability of the new passive active multijunction launcher removed major concerns on the possibility of using LH on ITER. LH exhibits the highest experimental current drive (CD) efficiency at low plasma temperature thus making it the natural candidate for off-axis CD on ITER where current profile control will help in maintaining burning performance on a long-time scale. We review recent LH results: long internal transport barrier obtained in JET with current profile sustained and controlled by LH acting under real time feedback together with first LH control of flat q-profile in a hybrid regime with Te ∼ Ti. Minutes long fully non-inductive LH driven discharges on Tore Supra (TS). High CD efficiency with electron cyclotron in synergy with LH obtained in FTU and TS opening the possibility of interesting scenarii on ITER for MHD stabilization. Preliminary results of LH modelling for ITER are also reported. A brief overview of ITER LH system is reported together with some indication of new coming LH experiments, in particular KSTAR where CW klystrons at the foreseen ITER frequency of 5 GHz are being developed.

The recent success in coupling lower hybrid (LH) waves in high performance plasmas at JET together with the first demonstration on FTU of the coupling capability of the new passive active multijunction launcher removed major concerns on the possibility of using LH on ITER. LH exhibits the highest experimental current drive (CD) efficiency at low plasma temperature thus making it the natural candidate for off-axis CD on ITER where current profile control will help in maintaining burning performance on a long-time scale. We review recent LH results: long internal transport barrier obtained in JET with current profile sustained and controlled by LH acting under real time feedback together with first LH control of flat q-profile in a hybrid regime with Te~ Ti. Minutes long fully non-inductive LH driven discharges on Tore Supra (TS). High CD efficiency with electron cyclotron in synergy with LH obtained in FTU and TS opening the possibility of interesting scenarii on ITER for MHD stabilization. Preliminary results of LH modelling for ITER are also reported. A brief overview of ITER LH system is reported together with some indication of new coming LH experiments, in particular KSTAR where CW klystrons at the foreseen ITER frequency of 5?GHz are being developed.

To shed some light on the development of the fast m/n = 1/1 precursor to the sawtooth crash and its influence on plasma transport properties in the vicinity of the q = 1 surface, series of dedicated experiments have been conducted on the Tore Supra and TEXTOR tokamaks. It has been concluded that, before a crash, the hot core gets displaced with respect to the magnetic axis, drifts outwards by as much as 8–10 cm and may change its shape. Observation of the magnetic reconnection process has been made by means of electron cyclotron emission diagnostics. The heat pulse is seen far outside the inversion radius. The colder plasma develops a magnetic island on the former magnetic axis, after the hot core expulsion. Different kinds of behaviour of the m = 1 precursor before the crash, with respect to the displacement of the hot core and the duration of the oscillating phase, have been observed. An ideal kink model alone cannot be used for explanation; therefore, resistive effects play an important role in the mode development. Possible mechanisms that lead an m = 1 mode to such behaviour, and their links to the change in the central q-profile, are discussed. Results have been discussed in the light of various theoretical models of the sawtooth.

To shed some light on the development of the fast m/n= 1/1 precursor to the sawtooth crash and its influence on plasma transport properties in the vicinity of the q= 1 surface, series of dedicated experiments have been conducted on the Tore Supra and TEXTOR tokamaks. It has been concluded that, before a crash, the hot core gets displaced with respect to the magnetic axis, drifts outwards by as much as 8-10?cm and may change its shape. Observation of the magnetic reconnection process has been made by means of electron cyclotron emission diagnostics. The heat pulse is seen far outside the inversion radius. The colder plasma develops a magnetic island on the former magnetic axis, after the hot core expulsion. Different kinds of behaviour of the m= 1 precursor before the crash, with respect to the displacement of the hot core and the duration of the oscillating phase, have been observed. An ideal kink model alone cannot be used for explanation; therefore, resistive effects play an important role in the mode development. Possible mechanisms that lead an m= 1 mode to such behaviour, and their links to the change in the central q-profile, are discussed. Results have been discussed in the light of various theoretical models of the sawtooth.

Real-time control of many plasma parameters will be an essential aspect in the development of reliable high performance operation of next step tokamaks. The main prerequisites for any feedback scheme are the precise real-time determination of the quantities to be controlled, requiring top quality and highly reliable diagnostics, and the availability of robust control algorithms.

This paper describes a new technique which has been implemented on the JET tokamak to investigate integrated real-time control of several plasma profiles simultaneously (such as current, temperature and pressure) and reports the results of the first experimental tests. The profiles are handled through their projection on a suitable basis of functions according to the Galerkin scheme. Their response to three actuators (heating and current drive powers injected in the plasma) is linearized in an experimentally deduced multi-input multi-output model. The singular value decomposition of this model operator allows us to design a distributed-parameter real-time controller which maximizes the steady state decoupling of the multiple feedback loops. It enables us to control several coupled profiles simultaneously, with some degree of fuzziness to let the plasma evolve towards an accessible non-linear state which is the closest to the requested one, despite a limited number of actuators.

Scenarios of steady-state, fully non-inductive current in Tore Supra are predicted using a package of simulation codes (CRONOS). The plasma equilibrium and transport are consistently calculated with the deposition of power. The achievement of high injected energy discharges up to 1 GJ is shown. Two main scenarios are considered: a low density regime with 90&percent; non-inductive current driven by lower hybrid waves—lower hybrid current drive (LHCD)—and a high density regime combining LHCD and ion cyclotron resonance heating with a bootstrap current fraction up to 25&percent;. The predictive simulations of existing discharges are also reported.

In ASDEX Upgrade and JET, the ion temperature profiles can be described by R/LTi which exhibits only little variations, both locally, when comparing different discharges, and radially over a wide range of the poloidal cross-section. Considering a change of the local ion heat flux of more than a factor of two, this behaviour indicates some degree of profile stiffness. In JET, covering a large ion temperature range from 1 to 25 keV, the normalized ion temperature gradient, R/LTi, shows a dependence on the electron to ion temperature ratio or toroidal rotational shear. In particular, in hot ion plasmas, produced predominantly by neutral beam heating at low densities, in which large Ti/Te is coupled to strong toroidal rotation, the effect of the two quantities cannot be distinguished. Both in ASDEX Upgrade and JET, plasmas with internal transport barriers (ITBs), including the PEP mode in JET, are characterized by a significant increase of R/LTi above the value of L- and H-mode plasmas. In agreement with previous ASDEX Upgrade results, no increase of the ion heat transport in reversed magnetic shear ITB plasmas is found in JET when raising the electron heating. Evidence is presented that magnetic shear directly influences R/LTi, namely decreasing the ion heat transport when going from weakly positive to negative magnetic shear.

A large variety of JET discharges with internal transport barriers (ITBs) has been analysed in order to determine the main features which characterize turbulence stabilization at the barrier. It is found that the location of barriers is well correlated with regions where the E×B flow shearing rate exceeds the linear growth rate of the ion temperature gradient mode instability (γ_ηi

This paper summarizes the recent work on JET in the three areas of heating, current drive and energetic particles. The achievements have extended the possibilities of JET, have a direct connection to ITER operation and provide new and interesting physics. Toroidal rotation profiles of plasmas heated far off axis with little or no refuelling or momentum input are hollow with only small differences on whether the power deposition is located on the low field side or on the high field side. With LH current drive the magnetic shear was varied from slightly positive to negative. The improved coupling (through the use of plasma shaping and CD₄) allowed up to 3.4 MW of P_LH in internal transport barrier (ITB) plasmas with more than 15 MW of combined NBI and ICRF heating. The q-profile with negative magnetic shear and the ITB could be maintained for the duration of the high heating pulse (8 s). Fast ions have been produced in JET with ICRF to simulate alpha particles: by using third harmonic ⁴He heating, beam injected ⁴He at 120 kV were accelerated to energies above 2 MeV, taking advantage of the unique capability of JET to use NBI with ⁴He and to confine MeV class ions. ICRF heating was used to replicate the dynamics of alpha heating and the control of an equivalent Q&equal;10 `burn' was simulated.

In this paper, the algorithms to determine the safety factor q and density profiles in real time are presented. They have been designed to be implemented in JET, in the framework of the real time enhancements for the control of the internal transport barriers. In this method, only the signals of the magnetic diagnostics and of the interferometer polarimeter are used. Instead of solving the equilibrium equation, which at the moment does not seem fully compatible with real time requirements, the topology of the magnetic surfaces is determined on the basis of the external magnetic measurements. Then the density and poloidal field profiles are found inverting the interferometric and polarimetric measurements, respectively. Once the poloidal field has been obtained, the q profile is inferred. The density profiles obtained have been checked against the JET LIDAR diagnostic, showing very good agreement. On the other hand, the q profiles have been compared with the results of several versions of the equilibrium code EFIT, proving that the algorithms developed can identify with good precision both monotonic and reversed shear profiles. An optimized version of the code has also been written and it has been verified that the computing time is compatible with the requirements of JET real time system. The present algorithm is being implemented and is used during JET experimental campaigns.

Internal transport barriers (ITBs) can be produced in JET by the application of strong additional heating during the current rise phase of the plasma discharge. Using up to 3 MW of lower hybrid power to tailor the q-profile prior to the main heating phase, a large variety of q-profiles ranging from low positive to strong negative central shear have been obtained during the current rise (0.4 MA s⁻¹). With negative central magnetic shear s&equal;(r/q)(&ddiff;r/&ddiff;q), the analysis of ITB triggering reveals a correlation between the formation of the ITB and q_&min; reaching an integer value (q&equal;2 or q&equal;3). This observation is confirmed by the analysis of the Alfvèn cascades. The minimum power required to access regimes with ITBs is probably related to the transport and magnetohydrodynamic properties of integer magnetic surfaces. Laser ablation and shallow pellet injection have also been attempted in recent JET ITB triggering experiments. This article was scheduled to appear in issue 7 of Plasma Phys. Control. Fusion. To access this special issue on advanced Tokamak research in EFDA-JET please follow this link: http://stacks.iop.org/0741-3335/44/7

The ASDEX Upgrade advanced tokamak scenario with central q close to 1 has been reproduced on JET. For almost identical q profiles, the comparative analysis does show similar features like the fishbone activity and the current profile evolution. In JET, transport analyses indicates that an internal transport barrier (ITB) has been produced. Gradient length criterions based on the ion temperature gradient turbulence stabilization are used to characterize the ITBs in both devices. The trigger of ITBs is associated with rational surfaces in both devices although the underlying physics for this triggering seems different. This experiment has the prospect to get closer to identity experiments between the two tokamaks.

In JET discharges where lower hybrid heating and current drive (LHCD) is applied early during the current ramp, a region of the plasma with zero current density is formed near the axis. At the boundary of this region the current density is large and B_θ increases rapidly over a small distance. In the central region the safety factor, q, is effectively infinite, but this falls steeply in the boundary region. Outside the boundary region q reaches a minimum, where the magnetic shear s≡r/q (dq/dr) becomes zero. The formation of this region of zero current is dependent on both the heating and the current drive effects of the LHCD. When LHCD is switched off the current profile begins to relax towards the resistive peaked current distribution of fully inductive tokamak operation. If LHCD is not used in the current rise then these current profiles are not established. Although the physical mechanism exists to drive the central plasma current below zero, in most cases it appears to be prevented from going negative. At least one MHD mechanism has been identified which could be responsible for this. The presence of the zero central current is closely linked to the periodic relaxation events seen in these discharges. In these discharges, internal transport barriers have been observed with additional heating powers substantially below the values required to obtain barriers in monotonic q profile cases.

We present the results of recent experiments related to real-time control of internal transport barriers (ITBs) in JET. Using a simple criterion to characterize the ITB existence, location and strength, we have successfully controlled for the first time the radial electron temperature profile within the ITB. The dimensionless variable used in the real-time algorithm - ratio of the ion gyro-radius to the local gradient scale length of the electron temperature - is a measure of the normalized electron temperature gradient and characterizes satisfactorily the main ITB features with a relatively low computational cost. We show several examples of control of this variable in various experimental conditions of toroidal field and plasma current, using different heating systems as control actuators. We also present a double-loop feedback scheme where both the global neutron rate from D-D reactions and the ITB strength are controlled simultaneously. In this case the ITB is sustained in a fully non-inductive current drive regime during several seconds. With the proposed control method, disruptions are avoided by holding the plasma performance at a prescribed target and this opens the route towards stationary operation of tokamak plasmas with ITBs. Initial results suggest that the additional control of the current profile is an important issue for achieving steady-state operation, in particular in the triggering and the sustainment of the ITB.

Quasi-steady operation has been achieved at JET in the high-confinement regime with internal transport barriers (ITBs). The ITB has been maintained up to 11 s. This duration, much larger than the energy confinement time, is already approaching a current resistive time. The high-performance phase is limited only by plant constraints. The radial profiles of the thermal electron and ion pressures have steep gradients typically at mid-plasma radius. A large fraction of non-inductive current (above 80%) is sustained throughout the high-performance phase with a poloidal beta exceeding unity. The safety factor profile plays an important role in sustaining the ITB characteristics. In this regime where the self-generated bootstrap current (up to 1.0 MA) represents 50% of the total current, the resistive evolution of the non-monotonic q-profile is slowed down by using off-axis lower-hybrid current drive.