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An upgrade of the RFX-mod experiment is presently in the final design phase, aimed at widening the explored operational scenarios both in RFP and Tokamak configuration. The main design driver for this machine upgrade is the enhancement of the ‘shell-plasma proximity’, which is expected to provide a significant improvement in the plasma magnetic confinement. The achievement of this aim implies a major change of the internal components of the machine such as the removal of the present vacuum vessel, transferring the function of vacuum barrier to the duly modified toroidal support structure, and the integration of a new in-vessel support system to sustain the conductive stabilizing shell and the whole first wall. The paper presents an overview of the design choices and the proposed implementations, assessed on the base of engineering analyses and results of experimental tests performed on mock-ups of the new components. The solutions conceived to fulfill vacuum and electrical requirements of the in-vessel components, to guarantee their reliability during normal and abnormal operating conditions events, and interface compatibility with existing components and torus assembly sequence are finally highlighted.

An upgrade of RFX-­mod toroidal machine assembly is under development, aimed at widening the explored operational scenarios both in RFP and Tokamak configuration. The main design driver for this machine upgrade, named RFX-mod2, is the enhancement of the ‘shell-plasma proximity’, which is expected to provide a significant improvement in the plasma magnetic confinement. In order to achieve this goal, the existing Passive Stabilizing Shell will be adapted in order to be enclosed in the new vacuum vessel. This reconfiguration of the internal components of the experiment requires the design of a new structure aimed at supporting both the shell and the plasma facing components. The paper describes the design aspects of the shell assembly, with particular focus on the thermo-mechanical FEM analyses performed to verify the design of the components at the expected operating conditions.

The results produced by experimental campaigns of RFX-mod shows the need to improve passive MHD control and minimise braking torque on plasma. These improvements require major mechanical changes on the present components of the machine including first wall, vacuum vessel, and toroidal support structure (TSS). The vacuum vessel will be removed and the first wall tiles will be directly supported by the passive stabilising shell, so increasing the poloidal cross section by 28 mm radially and bringing the passive stabilising shell as close as possible to the plasma boundary. This paper presents the mechanical design modifications of the torus assembly. Composite rings are designed to support the passive stabilising shell on the TSS. The vacuum boundary is shifted at the TSS by developing different joint solutions compatible with the stringent requirements of the present components: ceramic-metal brazed rings at the two poloidal joints, fully welded solution at the external equatorial joint, and resistive weld plate at the internal equatorial joint. The vacuum vessel portholes for pumping, fuelling, diagnostics, and sensor cable routing are redesigned to be integrated onto the TSS. The design is supported by thermo-mechanical analyses and verifications carried out applying load combinations corresponding to the expected load cases.

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.

The externally applied loop voltage, required to drive current carrying magnetic configurations, such as tokamaks or reversed field pinches (RFP), can induce currents between the plasma and the surrounding conducting components, such as limiters, divertor sections, stabilizing plates or enclosing stabilizing shells with insulated electrical gaps. An appropriate modeling of this problem can quantify these currents and give constraints for the design of new devices. In this paper a methodology based on lumped element circuit is used to evaluate this problem for the case of in vessel components of the RFX-mod2, which led to a redesign of the electrical connections of plasma facing components. In the new design some parts, having applied potential in the kV range, are still exposed to the weakly ionized scrape off plasma. The process of arc formation between them and their prevention had been experimentally investigated and possible arc suppression strategies identified.

The measurement of the plasma current in ITER plays an outstanding role as it is part of the machine protection and is a safety-relevant measurement: it will be used in relation with regulatory limits to show that the operation remains within the safe envelope defined in the ITER license. The Continuous External Rogowski (CER) is an inductive sensor designed for current measurements and located in the casing of 3 Toroidal Field Coils (TFCs). After the completion of the design of the CER, 4 prototypes of the sensor were manufactured and R&D activities were performed under a Grant with the European Domestic Agency (F4E-GRT-012). The work was carried out between 2010 and 2011 by the ITERMAG consortium comprising 3 laboratories: CRPP (Switzerland) as leader, CEA (France) and RFX (Italy). The R&D campaign on CER prototypes consisted in the measurement of about 100 parameters to characterize the CER in terms of electrical, thermal, mechanical and also of vacuum compatibility. From these results, electromagnetic modeling of the CER response was performed. It is demonstrated that the CER fulfills ITER requirements. However, the vacuum compatibility of the prototype has to be improved and solutions to cope with this issue are proposed. (C) 2013 Elsevier B.V. All rights reserved.

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

This paper deals with an R&D activity in progress in support of the design of the ITER NBI prototype (MITICA), aimed at the identification of the most appropriate design solutions and installation guidelines to keep under control and minimize the effects of ionizing radiation on the electronic and passive components related to control, diagnostics, and electrical insulation located inside the MITICA vessel and its concrete biological shield. The method followed to approach the problem is based on three steps: estimation of expected radiation levels map, identification of electronic and passive components database for MITICA diagnostics and finally the assessment of the ionizing radiation compatibility for the identified components. The ultimate aim of this activity is the definition, on the basis of this assessment, of a list of preventive or mitigating actions (such as relocation, shielding or preventive maintenance) for the most high-risk components.

ITER has diagnostics with machine protection, basic and advanced control, and physics roles. Several are distributed on the inner and outer periphery of the vacuum vessel. They have reduced maintainability compared to diagnostics in ports. They also endure some of the highest nuclear and EM loads of any diagnostic for the longest time. They include

This paper presents the results of a feasibility study aimed at the design of a mechanical support and electrical connector for in-vessel magnetic sensors compatible with remote handling maintenance. The work starts with the review of the reference design proposed in the ITER Design Description Documents and highlights the issues still open. A new design proposal is then developed taking into account the experience of similar applications developed for JET. The design proposal is supported by CAD modelling and thermal and electromagnetic analyses. In conclusion the developments necessary to qualify the proposal before considering it mature for production are outlined.

In the framework of the strategy for the development and the procurement of the NB systems for ITER, it has been decided to build in Padova a test facility, including two experimental devices: a full size plasma source with low voltage extraction and a full size NB injector at full beam power (1 MV). These two different devices will separately address the main scientific and technological issues of the 17 MW NB injector for ITER. In particular the full size plasma source of negative ions will address the ITER performance requirements in terms of current density and uniformity, limitation of the electron/ion ratio and stationary operation at full current with high reliability and constant performances for the whole operating time up to 1 h. The required negative ion current density to be extracted from the plasma source ranges from 290 A/m2 in D2 (D−) and 350 A/m2 in H2 (H−) and these values should be obtained at the lowest admissible neutral pressure in the plasma source volume, nominally at 0.3 Pa. The electron to ion ratio should be limited to less than 1 and the admissible ion inhomogeneity extracted from the grids should be better than 10% on the whole plasma cross-section having a surface exposed to the extraction grid of the order of 1 m2. The main design choices will be presented in the paper as well as an overview of the design of the main components and systems.

The failure mode and effects analysis (FMEA) is a widely used analytical technique that helps in identifying and reducing the risks of failure in a system, component or process. The application of a systematic method like the FMEA was deemed necessary and adequate to support the design process of the ITER NBI (neutral beam injector). The approach adopted was to develop a FMEA at a general “system level”, focusing the study on the main functions of the system and ensuring that all the interfaces and interactions are covered among the various subsystems. The FMEA was extended to the whole NBI system taking into account the present design status. The FMEA procedure will be then applied to the detailed design phase at the component level, in particular to identify (or define) the ITER Class of Risk. Several important failure modes were evidenced, and estimates of subsystems and components reliability are now available. FMEA procedure resulted essential to identify and confirm the diagnostic systems required for protection and control, and the outcome of this analysis will represent the baseline document for the design of the NBI and NBTF integrated protection system. In the paper, rationale and background of the FMEA for ITER NBI are presented, methods employed are described and most interesting results are reported and discussed.

A new system of magnetic probes was installed during the 2005 shutdown and was commissioned during the 2005/2006 restart phase of JET. The system is made of traditional pick-up coils and Hall sensors, all located outside the vessel. Direct field measurements from Hall sensors are used to correct the drift of the integrators of the pick-up coils signals. This feature will be crucial for future ITER-like devices, where long-lasting flat top phases are expected, in a high neutron yield and a high temperature environment. After a general overview of the system, the paper describes the major manufacturing and installation issues, including the construction of the supports and probes as well as the acceptance tests before and after installation. The functional commissioning of the system includes the integration of the new signals in the JET Data Acquisition System and the analysis of several discharges with and without plasma. The critical aspects of the assessment of the reliability of the signals are shown and commented on.

Experimental results and theoretical studies call for Reversed Field Experiment (RFX) machine and power supply improvements to allow studies that go beyond those of a conventional Reversed Field Pinch (RFP) with passively stabilized turbulent MHD dynamo. The new paths opened by recent results in RFX and other RFP machines are introduced; then the goals and the design lines of the technical modifications of RFX, mainly addressed to improve the first wall, the plasma magnetic boundaries and to increase the operational flexibility of the toroidal field circuit power supply, are reported.

This paper deals with the electromagnetic analyses carried out for the design of a new system developed to actively control the magnetic configuration of RFX. The system, which consists of a new conductive shell and a new set of saddle coils closely fitting the plasma region, is intended for replacing the present shell, keeping the existing vacuum vessel and principal coils of the machine unchanged. In the first part of the paper, the basic concepts representing the aims and the constraints of the system are described. Following this, the results of the electromagnetic analyses over the operating frequency range, performed by means of both analytical and finite element codes, are reported and discussed. The efficacy of the system is pointed out with the evaluation of the magnetic field generated by the new saddle coils and the reduction of field errors provided by the new shell with an overlapped gap. Finally, the calculation of the electrical parameters of the coils, taking into account the eddy current effect, is reported.