Giuseppe Ramogida | ENEA, Italian National Agency for New Technologies, Energy and Sustainable Economic Development (original) (raw)

Papers by Giuseppe Ramogida

The vertical position and shape controller for Ignitor has been designed on the basis of the CREA... more The vertical position and shape controller for Ignitor has been designed on the basis of the CREATELlinearized plasma response model, which assumes an axisymmetric system and describes the electromagnetic interaction of the plasma with the surrounding structures by a small number of global parameters (i.e., βpol, li, Ip). In particular, the vertical stabilization system has been designed assuming that the vertical plasma centroid position can be estimated by a suitable linear combination of the available magnetic measurements. A possible partial failure of these magnetic diagnostics has already been taken into account, showing a good resilience to such events. However, in case of severe failures, it will be necessary to resort to a completely different (i.e. non-magnetic) measurement of the vertical position. As an example, we apply this method to the simulated signal of a double, soft X-ray spectrometer looking at the top and bottom of the plasma edge. The spatial and spectral features of these segnals seem, in many cases, sufficient to discriminate beween actual movements of the plasma column and changes in the plasma paramters. R. Albanese, F. Villone, Nucl. Fusion 38, 723 (1998) F. Bombarda, et al., 35th EPS Plasma Phys. Conf. P4.073 (2008)

Ignitor has adopted an ``extended limiter'' configuration to fill all the available volume with t... more Ignitor has adopted an ``extended limiter'' configuration to fill all the available volume with the plasma, and to keep the peak power on the wall to less than 2 MW/m^2 for the reference ignition scenario. To achieve this challenging result, the FW shape follows closely the plasma column and needs to be built with strict tolerances. Accurate predictions of the plasma conditions near the edge were important for the design process, but the FWL geometry presents unique and partially unexplored features that have prompted the development of new modeling tools[1] for the SOL of Ignitor. The new analysis now includes the effect of neutral atoms, obtained by coupling the plasma fluid code ASPOEL with the neutral solver EIRENE. Preliminary results confirm that a large fraction of the recycling neutrals is ionized in the SOL itself, before entering the main plasma. As a consequence, the plasma temperature in the SOL is reduced, limiting wall sputtering. Another configuration with BT13 T, Ip10 MA and double X-points just outside the FW is analyzed, to facilitate access to the H-regime. In this configuration, the incidence angle of the magnetic field onto the wall grows rapidly near the tangency point, which challenges the need to keep the peak power at a low level. [1]F. Subba, et al., J. Nucl. Mater. 363-365, 693 (2007). *Sponsored by ENEA.

Fusion Engineering and Design, 2009

The present work evaluates, using 3D finite element (FE) electromagnetic (EM) analyses, the poloi... more The present work evaluates, using 3D finite element (FE) electromagnetic (EM) analyses, the poloidal field coil (PFC) stray field reduction inside and outside the main ITER building due to the presence of ferromagnetic content in the concrete and other iron components outside the vessel (mainly the huge iron boxes of the NBI – neutral beam injector – and the iron doors at the end of the port corridors).To perform these analyses a 360° 3D EM model of the ITER building has been developed, named electromagnetic model of the building complex (EMMOBC), which includes the poloidal field coils, the plasma, a coarse model of the two heating & current drive (H&CD) NBIs, the coils of the NBI active magnetic field reduction system, and all the main building components that could include ferromagnetic materials. The plasma scenarios at the start of flat-top (SOF) and at the end of burning (EOB) have been considered. The effect on the stray field on the NBI due to the presence of the active (AMFRS) and passive (PMFRS) magnetic field reduction system of the near NBI and of the others iron component in the building has been evaluated, using EMMOBC that include the coarse model of the two NBIs.The coil currents of the AMFRS in the H&CD NBI have been optimized for the stray field coming from the SOF and EOB plasma scenarios at plasma current of 15 MA. The stray field at SOF and EOB, including the effects of the ferromagnetic iron content (outside the vessel), has been evaluated inside and outside the main ITER building using the EMMOBC. Finally the field perturbation produced on the plasma q = 2 surface has been evaluated.

ABSTRACT A new facility for fusion, the Fusion Advanced Studies Torus (FAST), has been proposed t... more ABSTRACT A new facility for fusion, the Fusion Advanced Studies Torus (FAST), has been proposed to prepare ITER scenarios and to investigate non linear dynamics of energetic particles, relevant for the understanding of burning plasmas behavior, using fast ions accelerated by heating and current drive systems [1]. This new facility is considered an important tool also for the successful development of the demonstration/prototype reactor (DEMO), because the DEMO scenarios can take valuable advantage by a preparatory activity on devices smaller than ITER with sufficient flexibility and capable plasma conditions, before to testing them on ITER itself. To keep the cost of this new facility low enough was chosen to use the copper as main conductor material for the toroidal and poloidal fields coils. So FAST is a compact (Ro = 1.82 m, a = 0.64 m, triangularity ? = 0.4) and cost effective machine consisting of 18 Toroidal Field Coils (TFC), 6 Central Solenoid (CS) coils, 6 External Poloidal coils (3 + 3), adiabatically heated during the plasma pulse and cooled down at cryogenic temperatures (30 K) by helium gas between two consecutive pulses [1], [2]. Then a careful consideration of thermal loads due to the high density currents circulating in the coils is mandatory to establish the limiting performances of this tokamak. Moreover FAST is a very demanding machine because of the large number of scenarios foreseen on it and this strong flexibility in operation capability doesn't allow to choose the more demanding scenario on a simple B2t criterion. In fact several scenarios are to be investigated to ascertain the more onerous conditions on the machine not only because of the non linear copper resistivity growth with the temperature but also due to its non linear variation with the applied magnetic fields. In this work the thermal analysis carried on the toroidal magnet are presented with particular regard to the role of the magneto-resistive effect and taking in account als- o the copper specific heat variation with temperature. The analysis are performed by F.E.M. (Finite Elements Method) models realized with a commercial multi-physics code (by COMSOL?) and comparing its results with the ones obtained using an axisymmetric integral code developed in ENEA.

Fusion creates more neutrons per energy released than fission or spallation, therefore DT fusion ... more Fusion creates more neutrons per energy released than fission or spallation, therefore DT fusion facilities have the potential to become the most intense sources of neutrons for material testing. An Ignitor-like device, that is a compact, high field, high density machine could be envisaged for this purpose making full use of the intense neutron flux that it can generate, without reaching ignition. The main features of this High Field Neutron Source Facility, which would have about 50% more volume than Ignitor, are illustrated and the R&D required in order to achieve relevant dpa quantities in the tested materials are discussed, in particular the adoption of superconducting magnet coils. Radiation damage evaluations have been performed by means of the ACAB code for some fusion-relevant materials, like pure iron, ASI316L, EUROFER, SiC/SiC, Mo, Graphite, V-15Cr-5Ti. Values ranging from 1.6 x10-26 to 2.4 x10-25 dpa per source neutron have been obtained. Some full-power months of operation are sufficient to obtain relevant radiation damage values in terms of dpa: the setup of a duty cycle for the device in order to obtain such operation times is the next required step to proceed with the evaluation. *Sponsored in part by ENEA of Italy and by the U.S. D.O.E.

A thermo-mechanical analysis of the Central Solenoid of Ignitor has been carried out using the AN... more A thermo-mechanical analysis of the Central Solenoid of Ignitor has been carried out using the ANSYS code based on a linear 3D Finite Element model. The adopted model takes into account the insulation layers and the epoxy resin fillings required by the coil design. The structural assessment considers, in particular, the transient conditions at both the beginning of the plasma current pulse (heating) and during the cooling phase of the solenoid that follows the end of the plasma current pulse. This transient which causes high shear stresses on the insulation material has led to the design of an optimized feeder connection. The stresses under the most critical conditions (start-up) are within the allowable values found by tests carried out by Ansaldo.

IGNITOR is a high field, high plasma current compact experiment designed to be first to reach and... more IGNITOR is a high field, high plasma current compact experiment designed to be first to reach and study ignited plasma conditions. The design is characterized by a high degree of flexibility obtained by mean of a higher number of poloidal coils and a ``large'' volume available to the plasma relative to the machine overall dimensions. The most advanced operation scenario (11 MA, 13 T) is based on one that involves the optimal filling of the plasma chamber (``extended First Wall configuration''). The double X-point plasma configuration (X- points on the plasma chamber) enables it to reach ignition with a relatively modest amount auxiliary heating and a sufficient magnetic safety factor in the H-regime. This scenario involves a plasma current of 9 MA with the 13 T maximum toroidal field. Other plasma scenarios with reduced performances are based on a 9 T toroidal field and involve plasma currents of 7 or 6 MA, in the extended First Wall or the double X-point configuration, respectively. The plasma start-up phase has been carefully studied and an optimal choice of the poloidal field coils currents has led to obtaining a relatively large area with a nearly null and flat magnetic field, without reducing the available maximum flux swing (up to 36 Wb) from the Poloidal Field coils system. ^*Sponsored in part by ENEA and the US D.O.E.

The performance of the integrated system for vertical stability, shape and plasma current control... more The performance of the integrated system for vertical stability, shape and plasma current control for the Ignitor machine has been assessed by means of the CREATELlinearized model of plasma responseootnotetextR. Albanese, F. Villone, Nucl. Fusion 38, 723 (1998) against a set of disturbances for the reference 11 MA limiter configuration and the 9 MA Double Null configuration. A new design, based on the methodology of the eXtreme Shape Controller (XSC) at JET, has been tested : by using all the shape control circuits with the exception of those used to control the vertical stability is possible to control up to four independent linear combinations of the 36 plasma-wall gaps. The results point out a substantial improvement in shape recovery, especially in the presence of a disturbance in li. The new shape controller can also automatically generate, via feedback control, new plasma shapes in the proximity of a given equilibrium configuration. The XSC ToolsootnotetextG. Ambrosino, R. Albanese et al., Fus. Eng.& Des. 74, 521 (2005) have been adapted and extended to develop linearized Ignitor models including 2D eddy currents and to solve inverse linearized plasma equilibria.

A fabrication procedure for a typical Plasma Chamber (PC) sector has been developed to cover all ... more A fabrication procedure for a typical Plasma Chamber (PC) sector has been developed to cover all the manufacturing phases, from the raw materials specification (including metallurgical processes) to the machining operations, acceptance procedures and vacuum tests. Basically, the sector is made of shaped elements (forged or rolled) welded together using special fixtures and then machined to achieve the final dimensional accuracy. An upgraded design of the plasma chamber's vertical support that can withstand the estimated electromagnetic loads (Eddy and Halo current plus horizontal net force resulting from the worst plasma disruption scenario VDE, Vertical Displacement Event) has been completed. The maintenance of the radial support can take place hands-on with a direct access from outside the cryostat. With the present design, vacuum tightness is achieved by welding conducted with automatic welding heads. On the outer surface of the PC a dedicated duct system, filled by helium gas, is included to cool down the PC to room temperature when needed.

Nuclear Fusion, 2011

Several experiments aimed at optimizing plasma pre-ionization using electron cyclotron (EC) waves... more Several experiments aimed at optimizing plasma pre-ionization using electron cyclotron (EC) waves have been carried out on many tokamaks in recent years as the basis of a multi-machine comparison study made to define the best operation scenarios for ITER, where the plasma breakdown will have to be achieved with a toroidal electric field of only 0.3 V m-1. The FTU (Frascati Tokamak Upgrade, R = 0.935 m, a = 0.3 m) contribution to this study is the main subject of this work. A reduction in electric field, as can be obtained with pre-ionization by ECH, can lower the transformer flux consumption in the start-up phase leading to a longer plasma current flat top. This point is of particular interest in the conceptual design of the steady-state scenario of the proposed FAST tokamak and has also been addressed. In the FTU experiment the scan in pre-filling pressure has evidenced the capability of EC power to increase, by a factor 4, the range of working pressure useful for plasma start-up. Varying the breakdown a minimum electric field of 0.41 V m-1 has been found with 0.8 MW of EC in perpendicular injection. A scan in magnetic field has evidenced that plasma start-up is likely insensitive to alignment between EC resonance and null position. A total transformer flux saving of 22% has been found acting on plasma resistivity (by increasing electron temperature) and on the plasma starting point (for an internal inductance reduction).

Fusion Engineering and Design, 2005

A new reference 11 MA operational scenario for Ignitor has been developed in order to reduce elec... more A new reference 11 MA operational scenario for Ignitor has been developed in order to reduce electromagnetic loads and power supply requests, and to avoid the use of Dispersion Strengthened Copper in some of the PF coils. The analysis and a relevant simulation of a typical fast vertical disruption have been carried out with the MAXFEA code, obtaining values within the engineering constraints during the whole operating scenario. A new approach for mitigating the EM loads on the plasma chamber has also been investigated, based on the use of copper toroidal layers added to the plasma chamber in order to simulate the effects of a plasma chamber of varying thickness in the outer regions of its vertical cross section. This appears to be quite effective not only in increasing the time constant of the plasma displacement but also in reducing the vertical force and its combined effect with hoop force on the vessel.

The 2D simulation of the reference plasma disruption in Ignitor (a fast current quench following ... more The 2D simulation of the reference plasma disruption in Ignitor (a fast current quench following a Vertical Displacement Event - VDE), has been performed using the MAXFEA code. The resulting excitation loads have been used in a detailed 3D model to evaluate the electromagnetic loads in the first wall and its supports. The model of the most loaded region of the first wall has been performed using a zooming procedure that allows the replacement of the out-of-model plasma, poloidal coils and passive structures with current filaments surrounding the modeled region. To identify the most stressed tiles under the effect of the electromagnetic loads due to the VDE, a preliminary structural analysis has been carried out using a simplified model. Then a more detailed model has been performed for these tiles, with the aim of evaluating the stress on the screw and the pre-load needed to avoid detachment between the tile and the tile-carrier. The thermal loads on the tiles are evaluated independently and the implications of possible other disruption scenarios are discussed.

A detailed 3D finite element model has been developed in order to evaluate the electromagnetic lo... more A detailed 3D finite element model has been developed in order to evaluate the electromagnetic loads on the (mechanical) carriers of the tiles that constitute the First Wall of Ignitor during a reference Vertical Disruption Event. A thermo- structural analisys of the most stressed tile carrier with a cycled load has been completed. The study employed a non-linear ANSYS Code. The results show a temperature increase up to 341^ oC for a single step of 4 sec. The stresses and deformations on the component which has undergone a cycled load are within the limits of the allowable values. The design layout of the First Wall has been finalized, taking into account all requirements of the IGNITOR Machine. The electrical diagnostics placed inside the plasma chamber have been included in the tile carrier design. The First Wall has been tailored with special consideration for the Faraday Shield facing the ports through which ICRH is injected. *Sponsored in part by ENEA of Italy and by the U.S. DOE.

The IGNITOR vertical position and shape controller has been designed on the basis of the CREATE-L... more The IGNITOR vertical position and shape controller has been designed on the basis of the CREATE-L linearized plasma response model, taking into account the engineering constraints of the machine and the features of the burning plasma regimes to be obtained. Special care has been devoted to the design of a robust control system, that can operate even when a degradation of the performance of the electro-magnetic diagnostics may occur. The coupling between the vertical position control and the plasma shape control has been analyzed, in order to allow the plasma vertical position to be stabilized also in the case where a shape disturbance is provoked by a change of the main plasma parameters. Simulations of the control system response have been carried out using realistic models of the electrical power supply system. The non-linear computation of equilibrium flux maps before and after the perturbation shows that the system is able to recover from all the assumed disturbances with this control scheme. In addition, the control of the plasma current and of the separatrix of the double-null plasma configuration is being studied.^*Sponsored in part by ENEA and the US D.O.E.

FAST (Fusion Advanced Studies Torus) is a proposal for a Satellite Facility which can contribute ... more FAST (Fusion Advanced Studies Torus) is a proposal for a Satellite Facility which can contribute the rapid exploitation of ITER and prepare ITER and DEMO regimes of operation, as well as exploiting innovative DEMO technology. FAST operates with high performance H-Mode (BT up to 8.5 T; IP up to 8 MA) as well as Advanced Tokamak operation (IP=3 MA), and full non inductive current scenario (IP=2 MA) [1]. The project is based on a dominant 30 MW of ICRH, 6 MW of LH and 4 MW of ECRH. Helium gas at 30K is used for cooling the resistive copper magnets, which dimensions have been determined to limit the coil temperature at the end of the longest pulses (pulse length up to 170 s). The peak current density in the reference H-mode is about 45 MA/m2. The Finite Element Method (FEM) has been employed to analyse the stresses of the load assembly structure [2] using the ANSYS code. Due to the full structural cooperation, the Toroidal Field Coils, Central Solenoid and mechanical structure, have been modelled as a whole, using mechanical and thermal smeared properties. Structural analyses were performed at the most significant times of the operating scenario. The FW consists of a bundle of coaxial tubes armoured with 4 mm plasma-sprayed tungsten. The conceptual design of the FW, emphasising the more relevant aspects of the mechanical solution, is reported. The expected high power flux (18 MW/m2) in the divertor target plates [3] impose the use of monoblock W tiles actively cooled. An optimization of the divertor configuration, as a result of upgraded plasma edge modelling, is presented. The engineering aspects and the principal features of the divertor revised design are discussed.

Fusion Engineering and Design, 2011

The fusion advanced studies torus (FAST) has been proposed as a flexible and cost effective machi... more The fusion advanced studies torus (FAST) has been proposed as a flexible and cost effective machine that is able to support the development of ITER and DEMO operating scenarios exploiting some innovative technology solutions and to investigate the physics of high-performance plasmas in a dimensionless parameter range close to ITER. The FAST magnet consists of 18 coils, spaced by 20° in the toroidal angle, each made up of 14 copper plates, suitably arranged in order to realise 3 turns in the radial direction, with 89.2 kA per coil (in the H-mode plasma scenario 6.5 MA at 7.5 T). The finite number and toroidal extension of the toroidal field coils (TFCs) cause a periodic variation of the toroidal field from its nominal value called toroidal field ripple (TFR). An active ripple reduction system has been comprehensively investigated, by using proper 3D finite elements models, to provide an efficient and flexible system able to minimize the TFR in the region of interest. An optimization study of position and size of the coils required to reduce to an acceptable level for the operations the maximum ripple on the plasma (well below 0.3%), feeding them with currents sustainable during the whole scenario (∼1/10 of the current flowing in TFCs), is presented in this paper.

The performance of the control system for the position and shape of the elongated, tight aspect r... more The performance of the control system for the position and shape of the elongated, tight aspect ratio plasma column of Ignitor has been analyzed using the CREATELlinearized MHD deformable plasma response model. The possible failure of the relevant electromagnetic diagnostics has been taken into account by considering the feasibility of vertical control by other means, employing X-ray emission and thermography to evaluate displacements of the center of the plasma column and deformations of its outer surface interacting with the first wall. A realistic description of the power supplies has been introduced in the simulation scheme, thus allowing the optimization of the PID (Proportional-Integral- Derivative) controller. Both a voltage and a current loop control scheme have been analyzed: the first has been found to be only marginally better than the second one. The problem of controlling the shape of the plasma cross section has been dealt with by considering shape deformations induced by varying one of the plasma macroscopic parameters (eg., Ip, βpol, li) by a few percent.^*Sponsored in part by ENEA of Italy and by the U.S. DOE. R. Albanese, F. Villone Nucl. Fusion 38 723 (1998).

The structural analysis of the IGNITOR machine Load Assembly has been completed taking into accou... more The structural analysis of the IGNITOR machine Load Assembly has been completed taking into account the friction coefficients at the interfaces between its main components. A Finite Element ANSYS model was used to analyze the non-linear mechanical behavior of the structure. The calculation shows stresses within the allowable limits at the operating temperature. Interlaminar shear stresses values on the insulators of the toroidal field coils have been validated by the results of tests performed by Ansaldo. The non-linear analysis takes into account both the in-plane and the out-of-plane loads. Under normal operating conditions the assumed friction coefficient on the wedging surfaces is adequate to assure the structural stability of the Load Assembly. Furthermore, once unloaded, the structure comes back without any permanent deformation. The safety factors of the average shear stresses against the insulation shear rupture strength at the beginning of Ignitor life is always greater than 3, while at the end of life this is reduced to about 2 because of the degradation of mechanical properties due to the neutron dose. Keys of proper dimensions between the 30^o extension of the C-clamps modules have been adopted to assure structural stability. *Sponsored in part by ENEA of Italy and by the U.S. DOE.

The Bitter-type Toroidal Field Coils (TFC) adopted for Ignitor consist of plates that are cooled ... more The Bitter-type Toroidal Field Coils (TFC) adopted for Ignitor consist of plates that are cooled down to 30 K by Helium gas. Copper OFHC has been selected for these plates, allowing for an Electron Beam (EB) welding solution of the cooling channels. Kabel Metal set up the welding parameters and qualified the process to achieve full joint penetration with acceptable metallurgical structure. The qualification covers both the welding of the cooling channels and the inlet/outlet tube made on two full size samples. A metallographic examination and vacuum and pressure tests have been preformed to validate the basic suitability of the EB welding process. *Sponsored in part by ENEA of Italy and by the U.S. DOE.

Fusion Engineering and Design, 2009

Nuclear Fusion, 2011

Fusion Engineering and Design, 2005

Fusion Engineering and Design, 2011