RFX-mod2 is the last upgrade of the reversed field pinch machine operated at Consorzio RFX. A significant modification consists of the replacement of the first wall tiles, proposed as a key factor for the improvement of the gas density with the reduction of the hydrogen retention, and designed in coherence with the magnetic front-end modification that foresees the tiles supported by the existing MHD passive stabilising shell. The main choices in the design of the new first wall tiles are the polycrystalline graphite as bulk material, the use of the existing fixing keys with fastening bayonets, and the tile width that shall be less than the diameter of the larger port holes to allow remote handling operations for the maintenance of the first wall; the latter fix the number of tiles to 2016 as in the original configuration. The expected decrease of the plasma-wall interaction determined the first wall design with expected power densities up to 50 MW/m2 considering the deformation of the last magnetic surface in both reversed field pinch and tokamak configurations. At the other side, the need of sensor integration and shielding of the passive stabilising shell from the plasma addressed the tile thickness and surface extension. Local prominences have been modelled on tiles, based on visual inspections of actual surfaces after previous operations, in order to limit the plasma in regions far from openings and supporting structures. The tile resistant section has been increased coherently with all the interfaces and constraints, so decreasing the maximum stress at 3.5 MPa calculated from finite element analysis that simulates the operating condition. This low stress level together with a measurement of the experimental loads during next RFX-mod2 operations could qualify the use of extruded graphite for a further first wall change in the future. Indeed, extruded graphite is considered attractive given its high directional thermal diffusivity (about 50% better then polycrystalline graphite) to enhance the heat transmission and so improving the gas density control, and the low stress induced may allow this mechanically less performing grade of graphite.

Laser interferometer/polarimeter systems are used in magnetically confined fusion experiments for simultaneous measurements of the line-integrated electron density and of the current-induced magnetic field. In this work, we present the design of the interferometer/polarimeter system for the Divertor Tokamak Test facility (DTT) experiment, a new tokamak device dedicated to investigate alternative power exhaust solutions for the nuclear fusion DEMOnstration Power Station (DEMO). The optical design is based on the exploitation of a 7+7 chord scheme, which allows determining density and poloidal field, contributes to evaluate the plasma magnetic equilibrium and can provide the real time estimate of the q profile. Since the optical scheme is thought to be compatible with a possible Double Null divertor configuration, an equatorial port is used. To the aim of protecting the in-vessel optics, each chord employs a back reflecting mirror installed in the high field side inner wall close to the divertor, where some plasma-free space is available, and one retroreflector installed in the space behind the low field side outer first wall. With respect to polarimetric measurements and low effects of density gradients, the optimal laser source solution would be 100/50 ?m. With this setup, in low/medium density conditions, the longer wavelength will provide a good magnetic field measurement, while the shorter wavelength will allow vibration compensation for density measurements. In high-density regimes, the short wavelength alone can provide both magnetic field information from Faraday rotation and density measurements from the Cotton-Mouton effect. The two wavelengths are close enough to each other also to provide a good sharing of optical components. The main drawback of this solution is the availability of reliable laser sources at ? 50 ?m. In this regard, the QCL laser development in the THz domain seems promising, however commercially available CH3OH lasers at 118.8/96.5 ?m are a viable alternative.

Dispersion Interferometers (DI) present the fundamental advantage against conventional ones to be insensitive to mechanical vibration without requiring a second wavelength interferometer to measure path length variations. On the other side their optical setup requires a duplication of nearly all optical components for any additional channel making in this way quite complicate the realization of a multi-channel interferometer which is often required to translate the line integral measurements to local (or nearly local) measurements. To get the advantage of the dispersion interferometer without its drawback in this work we propose to join the dispersion technique to the bean scanning one, which has been already successfully implemented in the convention mid-infrared two-color interferometer [1]. To this aim we present a preliminary design of a DI scanning interferometer for the new Divertor Test Tokamak (DTT) presently in the design phase. The role of the DTT facility [2] is to help the development of a reliable solution for the power and particle exhaust in a reactor. To this aim, DTT has been designed to study a large suite of alternative divertor magnetic configurations in order to ensure acceptable conditions at the walls while maintaining sufficient core performance. In this contest measurements of plasma parameters in the divertor region is import though not often easy. To contribute to improve divertor measurements the proposed interferometer will measure the density along the divertor legs from the strike points up toward the X-point. The interferometer will use a CO2 laser (l=10.6 ?m) and a double pass optical scheme. Phase modulation method [3] is used to improve the resolution of the measurement and to extent measurement range above the 10-19 m-2 line integral limitation of the standard homodyne implementation. Both improvements are important in this application considering the expected high extension of density range in the DTT divertor region. Comparing to shorter wavelength, easily used in the DI interferometer, CO2 wavelength improves density resolution while providing good immunity to diffraction effect due to the expected high density gradient.

EAST has implemented the feedback control of the radiated power to protect divertor target plates from overheating in H-mode long pulse discharges [1]. Since now, by the real-time control system it has been obtained a radiative fraction up to 40%, and it was found the neon (Ne) gas one of the best choices as the additional radiator. In order to analyze the transfer process in scrape-off layer (SOL) and impurity behavior in the scenario of radiative control, the edge code SOLEDGE2D-EIRNE [2]. In this article, two typical upper-single null (USN) tungsten divertor discharges were modeled: an H-mode discharge in radiative feedback control phase (neon seeding) with Prad=0.8MW , and one without neon seeding with Prad<=0.5MWused as the reference pulse. The experimental data without neon seeding show a nearly uniform radiation emission distribution in the different regions (main plasma and divertor region). For the neon seeding phase, the change of the radiated emission also shows a uniform increment both in the main plasma and the divertor region. This kind of distribution can be caused by various factors: a lot of seeded light impurity may be transported into main plasma or to an increment of core accumulation of heavy impurities like tungsten. In the modeling situation, the low edge electron temperature is one of the possible reasons to allow more Ne particles into separatrix but results also suggested that heavy impurity might increase. However, whether the neon seeding causes an additional sputter of W still needs more study. Modeling results confirm that the additional neon gas injection provides the reduction of the divertor peak power fluxes, mitigates the power load on the divertor region, which is consistent with the diagnostic data from Langmiur probe.The Quasi-snowflake (QSF) discharge on EAST has been also simulated to assess the effectiveness of neon seeding for this configuration. Based on this simulation result of an existing non-seeded discharge, a prediction of the neon seeding phase with the same configuration is done to estimate the neon transport process under the upper QSF shape.

Clear evidences that, due to a strong outward impurity convection, impurity core penetration is prevented have been found in the RFX-mod RFP device. A comparable convection of the main gas has not been observed [1] so that a favorable situation with peaked or flat density profiles and hollow impurity profiles is produced. Analysis of impurity transport relies on best reconstruction of impurity emission pattern with a 1-dim Collisional-Radiative code in which the radial impurity flux is schematized as a sum of a convective and a diffusive term [2,3]. The diffusion coefficient D and the velocity V, which are input to the simulation are varied until the experimental emission is reproduced. While the steady-state impurity profile is determined by the ratio V/D (peaking factor) , the discrimination between D and V requires transient perturbative experiments. The experimental evidence of impurity outward convection in RFX-mod helical regimes occurring at high plasma current (I>1.2 MA) has been found in Li and C solid room temperature pellets experiments [4], Ne doped D2 cryogenic pellet injection, Ne gas puffing and Ni LBO experiments [5](W LBO didn't show accumulation effects too). Similar D and V have been found for all the considered impurity species, without strong dependence on mass/charge. RFX-mod is now being upgraded to RFX-mod2, aiming at reducing secondary tearing mode amplitude which affects the duration of the improved confinement Single Helicity states [6]. In order to perform more detailed analysis of the impurity transport inside the outward convection barrier, the impurity source should be further inside the plasma. With this aim, Ni-tracer encapsulated solid pellet (Ni-TESPEL) experiments are foreseen in the new device [7]. The available 1-dimensional, time dependent Ni Collisional Radiative code, used to reconstruct experimental Ni emissions in RFX-mod [ 4] has been upgraded in preparation of such experiments in RFX-mod2 including the possibility of a Ni source (boundary condition) inside the plasma, placed in a time dependent position. The solid pellet injector already used in RFX-mod to inject C and Li solid pellets, will be adapted to inject TESPEL in RFX-mod2 (0.7/0.9 mm polystyrene ball with Ni powder inside, injection velocity up to 200 m/s can be reached). In this contribution, the solid pellet injector will be described, simulations of the pellet ablation [8] for different scenarios of RFX-mod2 plasma will be presented, Ni ion density, line and continuum emission profiles predicted by the code will be described and discussed.

The Reversed Field Pinch (RFP) configuration is often characterized by a wide spectrum of unstable tearing modes (TMs) involved in the generation and sustainment of the magnetic field in the plasma. This dynamo process is heavily influenced also by other processes or parameters like intrinsic plasma flow (no external torque sources are normally present), wall resistivity and interaction with external non axi-symmetric magnetic fields provided by active coils. The transition between fast-slow rotation branches under the application of magnetic feedback boundary conditions was studied in [1] for the case of RFX-mod, a medium size (a = 0.459 m, R0 = 2m) flexible toroidal magnetic confinement device. In particular, RFX-mod is equipped with an advanced and sophisticated feedback system realized by a grid of independently controlled 48 (toroidal) x4 (poloidal) active saddle coils. In this contribution the interaction of the m=1, n>9 tearing modes with the bulk plasma close to the fast-slow transition threshold is investigated with the help of a systematic set of experiments where the active feedback control is selectively switched on and off on specific modes during the discharge dynamics. The peculiar result found is that also in fast rotating discharges, although the radial component at the wall of high n tearing modes is almost zero, in the absence of active feedback control a gradual modification (growth) of the total mode amplitude is measured, leading potentially to wall locking of that mode and to the back transition to the slow rotation branch. Data showing the mode-mode and mode-wall interaction will be presented for several combinations of non-controlled TMs. The process is discussed also in function of the dynamics of the plasma flow, following the lines presented in [2], and showing the clear correlation between active control, MHD dynamics and plasma flow.

RFX-mod is a Reversed Field Pinch device that allowed performing experiments in regimes with a plasma current up to 2 MA, thanks to its MHD active control system. Experiments have shown that improved plasma performances are obtained when in the resonant part of the m = 1 spectrum one dominant tearing mode is much higher than the other secondary ones (quasi single helicity states). Tearing modes play a crucial role in determining energy and particle transport. Based on the present understanding of the interplay between passive conductive boundaries and tearing modes in an RFP, an upgrade of RFX-mod machine assembly has been designed, dubbed RFX-mod2, and it is now being implemented. The highly resistive Inconel vessel will be removed, graphite tiles will be attached to the copper stabilizing shell and the stainless steel support structure will be modified in order to be vacuum tight. In RFX-mod2, the shell-plasma proximity decreases from b/a = 1.11 to b/a = 1.04 and copper, instead of Inconel, will be the continuous conducting structure nearest to the plasma. MHD non-linear simulations show that secondary tearing modes amplitude and the edge bulging due to their phase locking will decrease; moreover the plasma current threshold for tearing modes wall locking will also significantly increase.

Alternative configurations for the DEMO divertor are being explored to achieve an improved mitigation of the heat and particle loading at the plasma-material interfaces. These configurations with a variable volume in the private flux region (PFR), which consequently influences the neutral behavior and hence the achieved divertor pumping efficiency. This paper studies, in terms of the pumping efficiency, a wide range of prominent proposed alternatives to the conventional, single-null divertor, namely the "double null" divertor, the "X divertor", the "Super-X divertor" and the "snowflake" divertor. The investigation of the impact of neutral gas dynamics on the particle exhaust for an alternative divertor under steady-state operation, will be performed using the numerical tool DIVGAS (Divertor Gas Simulator). The DIVGAS code is based on the Direct Simulation Monte Carlo (DSMC) method, in which the solution of the Boltzmann kinetic equation is reproduced by simulating the collisions and the ballistic flight of model particles, which statistically mimic the behavior of real molecules. In view of DEMO, the DIVGAS code will allow for identifying the design space of an optimum divertor, which features high pumping efficiency. In our workflow, the neutral flow field for each of the aforementioned alternative divertor configurations will be derived from plasma boundary conditions extracted from SOLPS fluid simulations, exactly at the interfaces between the Scrape-off layer (SOL) and the PFR. All assumed plasma scenarios are based on a highly dissipative divertor relying on a partially detached divertor operating regime, similar to ITER, or even on full detachment. Moreover, the position and the size of the pumping port as well as their influence on the neutral flow behavior in the PFR are analyzed and the advantages and disadvantages of each case are discussed.

Alternative configurations for the DEMO divertor are being explored to achieve an improved mitigation of the heat and particle loading at the plasma-material interfaces. These configurations with a variable volume in the private flux region (PFR), which consequently influences the neutral behavior and hence the achieved divertor pumping efficiency. This paper studies, in terms of the pumping efficiency, a wide range of prominent proposed alternatives to the conventional, single-null divertor, namely the "double null" divertor, the "X divertor", the "Super-X divertor" and the "snowflake" divertor. The investigation of the impact of neutral gas dynamics on the particle exhaust for an alternative divertor under steady-state operation, will be performed using the numerical tool DIVGAS (Divertor Gas Simulator). The DIVGAS code is based on the Direct Simulation Monte Carlo (DSMC) method, in which the solution of the Boltzmann kinetic equation is reproduced by simulating the collisions and the ballistic flight of model particles, which statistically mimic the behavior of real molecules. In view of DEMO, the DIVGAS code will allow for identifying the design space of an optimum divertor, which features high pumping efficiency. In our workflow, the neutral flow field for each of the aforementioned alternative divertor configurations will be derived from plasma boundary conditions extracted from SOLPS fluid simulations, exactly at the interfaces between the Scrape-off layer (SOL) and the PFR. All assumed plasma scenarios are based on a highly dissipative divertor relying on a partially detached divertor operating regime, similar to ITER, or even on full detachment. Moreover, the position and the size of the pumping port as well as their influence on the neutral flow behavior in the PFR are analyzed and the advantages and disadvantages of each case are discussed.

One of the main challenges, within the European Fusion Roadmap, in view of the construction of a demonstration plant (DEMO, the first nuclear fusion power plant able to provide power to the electricity grid around 2050), is the thermal power on the divertor. ITER plans to test the actual possibilities of a "standard" divertor operating in a plasma fully detached condition, i.e. no contact between plasma and first wall of the vessel. This solution could be unsuitable to be extrapolated to the operating conditions of DEMO and future reactors; then the problem of thermal loads on the divertor may remain particularly critical in the road to the realization of the reactor. For this reason, a specific project has been launched, aimed to define and design a Tokamak named "DTT (Divertor Tokamak Test)". This Tokamak has to carry out a number of scaled experiments, to be integrated with the specific physical condition expected and technological solutions included in DEMO. DTT should retain the possibility of testing different divertor magnetic configurations, including liquid metal divertor targets, and other possible solutions promising to face with the power exhaust problem. The construction has recently been approved by the Italian government. DTT will be a high field superconducting toroidal device (6 T) carrying plasma current up to 5.5 MA in pulses with length up to about 100 s and with 45 MW of additional heating power. The nominal cross section is elongated with a major radius R=2.11 m and a minor radius a=0.64m. The DTT parameters are selected so as to reproduce edge conditions as close as possible to those expected in DEMO (in terms of a set of dimensionless parameters characterizing the physics of Scrape Off Layer, SOL, and of the divertor region), while fully fitting (again, in terms of the dimensionless parameters) with DEMO bulk plasma performance. Maximum flexibility is guaranteed, within the limits of a given budget and a tight time schedule consistent with the needs of the European Road Map. This paper describes the status of the design activities of DTT. Emphasis is given on an integrated design approach, illustrate the rationale for the design choices, focusing on the main components, namely magnet system, plasma scenarios, vacuum vessel, in-vessel components, thermal shield, neutron shield, and additional heating system.

One of the main challenges, within the European Fusion Roadmap, in view of the construction of a demonstration plant (DEMO, the first nuclear fusion power plant able to provide power to the electricity grid around 2050), is the thermal power on the divertor. ITER plans to test the actual possibilities of a "standard" divertor operating in a plasma fully detached condition, i.e. no contact between plasma and first wall of the vessel. This solution could be unsuitable to be extrapolated to the operating conditions of DEMO and future reactors; then the problem of thermal loads on the divertor may remain particularly critical in the road to the realization of the reactor. For this reason, a specific project has been launched, aimed to define and design a Tokamak named "DTT (Divertor Tokamak Test)". This Tokamak has to carry out a number of scaled experiments, to be integrated with the specific physical condition expected and technological solutions included in DEMO. DTT should retain the possibility of testing different divertor magnetic configurations, including liquid metal divertor targets, and other possible solutions promising to face with the power exhaust problem. The construction has recently been approved by the Italian government. DTT will be a high field superconducting toroidal device (6 T) carrying plasma current up to 5.5 MA in pulses with length up to about 100 s and with 45 MW of additional heating power. The nominal cross section is elongated with a major radius R=2.11 m and a minor radius a=0.64m. The DTT parameters are selected so as to reproduce edge conditions as close as possible to those expected in DEMO (in terms of a set of dimensionless parameters characterizing the physics of Scrape Off Layer, SOL, and of the divertor region), while fully fitting (again, in terms of the dimensionless parameters) with DEMO bulk plasma performance. Maximum flexibility is guaranteed, within the limits of a given budget and a tight time schedule consistent with the needs of the European Road Map. This paper describes the status of the design activities of DTT. Emphasis is given on an integrated design approach, illustrate the rationale for the design choices, focusing on the main components, namely magnet system, plasma scenarios, vacuum vessel, in-vessel components, thermal shield, neutron shield, and additional heating system.

DTT (Divertor Tokamak Test) is an Italian scientific project in the field of nuclear fusion aimed at creating a European facility for the testing of power exhaust and helium removal strategies alternative to the one implemented in ITER, the international experiment in fusion that will prove the feasibility of fusion energy on a large scale from 2025. In the European Fusion Roadmap to the realisation of fusion energy within 2050, one of the main mission is indeed the study of the power released by the plasma core to the so-called divertor, a keycomponent of any present tokamak machine on which the thermal power is directed. This will be tested also in ITER on a "standard" divertor operating in a plasma fully detached condition, i.e. no contact between plasma and first wall of the vessel. Nonetheless, the baseline strategy implemented in ITER could not be extrapolated to DEMO and future power plants, then the problem of thermal loads on the divertor may remain particularly critical in the road to the realization of a reactor. For this reason, a facility like DTT, where a number of scaled experiments, fully integrated with the expected physical parameters and engineering solutions to be used in DEMO, could be tested, is of paramount importance. Indeed, DTT has been designed so far to retain the possibility of testing different divertor magnetic configurations, including liquid metal divertor targets, and other possible promising solutions to face the power exhaust problem. The construction, recently approved by the Italian government, is rapidly approaching on the base of a completely renovated design with respect to the original proposal in order to account also for double-null configurations (up-down symmetric) and to keep costs within the original budget. DTT will be a high field superconducting toroidal device (6 T) carrying plasma current up to 5.5 MA in pulses with length up to about 100s and with 45 MW of additional heating power. The nominal cross section is elongated with a major radius R=2.11 m and a minor radius a=0.64m. This paper describes the status of the design activities of DTT. In particular the integrated design approach adopted, the rationale for the design choices, the development of the integrated management system are illustrated. Also a look on the main components, namely magnet system, plasma scenarios, vacuum vessel, invessel components, thermal shield, neutron shield, and additional heating system will be taken.

The solution of the problem of heat exhaust has been pointed out as one of the main challenge towards the realization of magnetic confinement fusion. In the last years, two concepts have been proposed in alternative to the conventional divertor solution adopted for ITER: modification of the magnetic topology in the divertor region and liquid metal as plasma facing component. The role of the Divertor Tokamak Test facility (DTT) in the power exhaust implementation strategy is discussed. The evolution of the project, since the original proposal in 2015 to the present design, is shown. The DTT facility is well integrated in the European strategy and the final decision on the divertor configuration will be made, within 2022-23, on the basis of the indication of the Power Exhaust Group constituted by the EUROfusion Consortium. Finally, the main milestones and the timeline of the project are illustrated.

This paper presents the engineering aspects of the design review of the Italian DTT (Divertor Tokamak Test facility), illustrating the rationale for the design choices and focusing on the main differences with respect to the original proposal.

A gas baffle will soon be inserted in the vessel of the tokamak à configuration variable (TCV) [Reimerdes, Nucl. Mat. and Energy 2017]. This upgrade aims at achieving more favorable conditions for the onset of detachment. In this work, we simulate the effect of the baffle on the TCV edge plasma with the SolEdge2D-EIRENE code [Bufferand, Nucl. Fusion 2015], which couples a fluid plasma model to a kinetic model for neutrals and impurities. Firstly, a specific TCV shot with a baffle-compatible shape is simulated. This comparison allows to tune perpendicular transport coefficients in order to match upstream experimental profiles, and results in a good agreement with the experiment at the targets. The same simulation is then carried out including the baffle. The neutral compression ratio, namely the ratio between divertor and upstream neutral pressure, is shown to improve by a factor of order 10, resulting in bigger power and momentum losses in the divertor plasma. Next, we perform a scan in upstream density to access different divertor regimes, revealing that the neutral compression increases as we approach detachment. Finally, in view of possible future optimizations, the level of baffle closure is varied in the simulations and the feedback on plasma properties is discussed.

The effect of magnetic geometry on scrape-off layer (SOL) transport and detachment behaviour is investigated on the TCV tokamak with the goal of assessing the potential of alternative divertor geometries and for the validation of theoretical models. L-mode experiments reveal that increasing connection length and hence divertor volume by either increasing poloidal flux expansion or divertor leg length have different effects on the boundary plasma. In attached conditions, the SOL heat flux width q inferred from target infrared thermography measurements is weakly dependent on poloidal flux expansion but increases approximately with the square root of the divertor leg length. The divertor spreading factor S shows no clear trend with leg length but decreases with flux expansion. TOKAM3X turbulence simulations of the leg length scan are in qualitative agreement with the experiment and can explain observations by a strongly asymmetric (ballooning) transport at and below the X-point. Evidence for increased transport in the region of low poloidal field is obtained in the Snowflake minus geometry. The presence of an additional X-point in the low-field side SOL increases the effective SOL width by approximately a factor two. Increasing flux expansion and leg length both result in enhanced divertor radiation levels, with the effect being much larger in the latter case. This behaviour, together with the observed trend in q, is consistent with a substantial drop in the density threshold for divertor detachment with increasing leg length and a weak variation with flux expansion. Novel spectroscopic techniques reveal that the drop in target ion current and access to detachment is caused by a reduction of the divertor ionization source due to power starvation, while volume recombination is only a small contributor. This interpretation is confirmed by SOLPS modelling. TCV alternative divertor studies are being extended to neutral beam heated H-mode plasmas. The H-mode power threshold is found to vary weakly between standard, X-, and Super- X geometries. In all cases, ELMy H-mode is obtained at intermediate current, while the discharges are ELM-free at high current. Signs of detachment have so far only been observed in the latter case. Ongoing experiments further investigate H-mode detachment in these plasmas and will be extended to Snowflake configurations.

The European roadmap for fusion energy has identified plasma exhaust as a major challenge towards the realization of magnetic confinement fusion. To mitigate the risk that the baseline scenario with a single null divertor (SND) and a high radiation fraction adopted for ITER will not extrapolate to a DEMO reactor, the EUROfusion consortium is assessing potential benefits and engineering challenges of alternative divertor configurations. A range of alternative configurations that could be readily adopted in a DEMO design have been identified. They include the X divertor (XD), the Super-X divertor (SXD) and the Snowflake divertor (SFD). The flux flaring towards the divertor target of the XD is found to be limited by the minimum grazing angle at the target. The characteristic increase of the target radius in the SXD is a trade-off with the increased TF coil volume, but ultimately limited by forces onto coils. Engineering constraints also limit XD and SXD characteristics to the outer divertor leg with a solution for the inner leg requiring up-down symmetric configurations. Boundary models with varying degrees of complexity have been used to predict the beneficial effect of the alternative configurations on exhaust performance. Desired effects are an easier access to detachment, reluctance of the detachment front to move along the divertor leg and an increase of the divertor radiation without excessive core confinement degradation. Based on the extended 2-point model and achievable geometric variations the SOL radiation required for the onset of detachment decreases in the SXD and SFD with the tolerable residual power 9p1 ´ fradq being 30-40% larger than in the SND. Additional improvements are expected from the ability to increase frad without adverse effects on the core performance and through SOL broadening as postulated for the SFD. A systematic study of the alternative configurations and the SND reference using the divertor transport code TECXY confirms that the SFD detaches at a lower frad, but also shows that the potential gain is modest. The main expected advantage of the XD and similarly of the SXD is an increased reluctance of the detachment front to move towards the X-point. To that end the detachment dynamics are assessed with the SOLPS and SOLEDGE2D-EIRENE codes, which use more sophisticated models of the target geometry and neutral particles.

A conceptual design of an upgrade for RFX-mod was described at the past IAEA-FEC conference and the device is presently being modified. Further analysis of the RFX-mod database, concerning m=1 Tearing Modes locking threshold and reconnection events characterization are presented. Estimates of nonlinear behavior of Tearing Modes in RFX-mod2 together with detailed electromagnetic modeling aiming at optimizing error fields for the upgraded device have been performed in order to consolidate the design.

A number of modification s and enhancements are underway on the RFX - mod Reversed field Pinch (RFP) device . The main scientific goals motivating the modifications are the improvement of the confinement in the R eversed F i e l d P inch configuration and the investigation of a broad spectrum of plasma physics topics , through the exploitation of the multi configuration capability of the device, which can be operated as a RFP, ultra low q, circular and shaped tokamak . This paper describes the major challenges tackled to design technical solutions able to achieve this scientific mission.

The main mission of the Divertor Test Tokamak (DTT) is to explore viable solutions to the power exhaust issues in a fusion reactor. The ultimate goal will be to qualify and control in various divertor configurations DEMO relevant heat flux densities to the wall while preserving the integrity of both the plasma facing components and the plasma performance. Experiments will involve tailored magnetic topologies, highly radiative regimes and advanced materials. In this contribution, we describe the package of diagnostic systems that will be deployed to allow DTT accomplishing its tasks. 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. Focusing on the divertor diagnostics, a particular effort will consist in obtaining space resolved measurements of density, temperature, impurity densities and ionization front that can be compared with the 2D results of model simulations. This will be possible using mainly optical diagnostics based on filtered cameras and spectrometers deployed in imaging mode, complemented by local measurements obtained by a space resolved Thomson scattering, Langmuir probes and gas puffing imaging systems. Diagnostics of the main plasma will assure a full qualification of the core and of the pedestal in terms of thermal contents, equilibrium, fast particles densities, impurities, reactivity levels and turbulence. Strongly oriented to the exploration of control methods suitable for DEMO, DTT will also address the study of control systems based on physics and engineering models, which in DEMO are expected to take over the role of the diagnostics deemed to be incompatible with the harsh environment of a fusion reactor.