The studies on the development of fusion–fission hybrid reactors (FFHR) have gained consensus in recent years as an intermediate step before fusion energy. This work proposes a possible approach to FFHRs based on the coupling of a Reversed Field Pinch fusion machine and a Molten Salt Subcritical fission test bed. The proposed test bed is characterized by the coexistence of a fast-neutron fission core and a dedicated thermal-neutron zone, allowing the performing of tritium breeding and actinides transmutation studies. The neutronic design solutions and the results obtained by the irradiation of FLiBe salt (inside the thermal-neutron zone) and of an actinide target (inside the core) are shown. The outcomes of the analysis reveal the potential of FFHR systems as breeding/burner systems. In particular, the results regarding tritium breeding are very encouraging as the system is demonstrated to be able to reach a very high Tritium Breeding Ratio.

The magnetic energy storage and transfer (MEST) system is a newly developed scheme to supply the superconducting (SC) coils in large-size tokamak and reversed field pinch (RFP) reactors to cope with the issues related to electrical power handling in these applications. It is a very promising alternative to face both the issues related to high active power peaks, required during plasma formation and control, and the huge reactive power demand, linked to the use of thyristor converters. The MEST system operating principle and control strategy supplying a single SC coil have already been outlined and studied in previous works. From this starting point, the MEST control system has been expanded to supply a complete set of SC poloidal field coils, considering the magnetic coupling and the energies associated with the magnets-plasma system, essential to control the overall system energy and achieve a certain degree of decoupling between the grid and the coils. This article presents and discusses the principles of controlling the currents of a complete set of poloidal field coils using the MEST system with efficient electrical power management. The application study to PILOT fusion-fission hybrid reactor (FFHR) is reported using analytical calculations and numerical simulations to assess the control operation.

• Upgrade of RFX-mod experiment proposed for studies of RFP and Tokamak physics. • High-performance polymers and surface coatings for in-vessel components insulation. • In-vessel components made from Metal Additive Manufacturing . • Effective collaboration between research institutions and industries. • Support of local authorities to scientific research and technological innovation. A substantial modification of the toroidal complex of the RFX experiment, named RFX-mod2, is currently under completion, involving the whole core system of the machine and in particular the vacuum vessel, the entire plasma facing components and a wide set of in-vessel diagnostic systems . The combination of challenging requirements, in terms of electrical insulation and vacuum compatibility of in-vessel components, and stringent geometrical constraints to comply with interfaces to existing machine components, in particular external coils and diagnostic systems, called for the adoption of novel technological solutions developed in collaboration with local industries in the framework of an industrial innovation project co-funded by Regione Veneto. The paper presents an overview of the technological solutions implemented and tested during the manufacturing phase of the new components of the RFX-mod2 machine complex.

The Magnetic Energy Storage and Transfer system (MEST) aims at improving the power handling in supplying the SuperConducting (SC) coils of fusion experiments. It is based on smart use of Superconducting Magnetic Energy Storage technology and allows the introduction of a certain degree of decoupling between the grid and the load. The MEST operation is based on the storage and transfer of the needed energy between the Load Coil (LC) and an additional storage inductor named Sink Coil (KC); the energy transfer is realized via capacitors switched by fully controllable semiconductors. This paper focuses on the control system of the MEST and proposes a control strategy, based on two decoupled control loops, with the aim both to guarantee the desired current on the load coil and the compensation of the system losses. A model of the MEST with the proposed control system has been developed to analyze the performance through numerical simulations. The control strategy and the simulation results will be presented and discussed in the paper.

Reconnection events in high current reversed field pinch plasmas are often associated to the partial or total loss of the helical magnetic topology. The electron temperature collapse during these phenomena is investigated in RFX-mod thanks to high time resolution soft-x-ray diagnostics; these data are used, together with magnetic energy reconstructions, for energy balance analysis. The paper shows that the energy released during reconnection events, similarly to astrophysical plasmas, might be involved in ion heating, the latter being estimated by the energy distribution function of neutral atoms, a rather interesting feature in a reactorial perspective. These issues will be further investigated in RFX-mod2 , an upgrade of the present device starting its operations from 2022, where the modified boundary conditions are expected to increase the helical states duration and reduce the frequency of reconnection events.

The results are presented of an experimental activity performed in the RFX-mod device aimed at characterizing plasma dynamics in the so-called Ultralow-q (ULq) magnetic configuration, which corresponds to edge safety factor values below 1. The role of the edge safety factor in determining plasma dynamics is studied. In particular, a characterization of MHD activity is performed. The results of dedicated non-linear 3D visco-resistive MHD simulations are in good quantitative agreement with the experimental observations. In particular, the predicted tendency for ULq plasmas to be characterized by magnetic spectra dominated by a single mode (either a kink or a double resonant internal mode) is confirmed by experiment. Magnetic reconnection plays a relevant role in determining the dynamics of the magnetic topology. Both almost quiescent and largely fluctuating plasmas are observed with a strong sensitivity on the edge safety factor. The main MHD properties of the ULq are compared to those of RFP and tokamak discharges, also produced in the RFX-mod device. MHD modes exhibit toroidal rotation at a frequency depending on mode amplitude. Differently from what encountered in RFP plasmas at comparable current levels, no wall locking is detected.

The EU DEMO Plant Electrical System (PES) main scopes are to supply all the plant electrical loads and to deliver to the Power Transmission Grid (PTG) the net electrical power generated. The studies on the PES during the Pre Concept Design (PCD) Phase were mainly addressed to understand the possible issues, related to the special features both of the power generated, with respect to a power plant of the same size, and of the power to be supplied to the electrical loads. For this purpose, the approach was to start the design of the different PES components adopting technologies already utilized in fusion experiments and in Nuclear Power Plants (NPP) to verify their applicability and identify possible limits when scaled to the DEMO size and applied to the specific pulsed operating conditions. This work is not completed, however several issues have been already identified related to the pulsed operation of the turbine generator, the large amount of recirculation power, the very high peaks of active power required for the plasma formation and control, the huge reactive power demand, if thyristor converter technology was adopted to supply the superconducting coils, etc.. The paper gives an overview on the features and scope of the PES and its subsystems, on the main achievements during the Pre Concept Design (PCD) Phase, on the challenges for the development of the conceptual design in the next framework program and on the plan to face them.

The reversed field pinch (RFP) could be a valid basis for the 14.1-MeV neutron source as required in fusion–fission hybrid reactors (FFHRs). The use of superconductors in an RFP reactor, although mandatory in the magnetizing/equilibrium coils, could represent an issue due to their limitations in the allowed magnetic field derivative, which is quite high during the initial plasma current phase of the RFP operation. A new concept is proposed based on a double-coil system: the first made with superconductors and the other with conventional copper coils. The basic principle is first introduced with the help of a simplified model that allows understanding the double-coil system operation. A feasibility study applying this solution has been carried out in the design of the poloidal magnetic system of a pilot RFP ( $R$ = 4 m and $a\,\,=0.8$ m) as a fusion core for an FFHR. This article presents the preliminary results and, in particular, the possibility of obtaining high initial loop voltage for a fast rise of the RFP plasma current with a magnetic field derivative within the superconductor limits.

RFX-mod is an experimental fusion device, which contributes to plasma physics studies both in reversed field pinch (RFP) and tokamak configurations. Its high flexibility, due to an active magnetohydrodynamic (MHD) control system and the modular coils power supply (CPS) system, allowed operating RFX-mod in a wide range of experimental conditions with a plasma current up to 2 MA. Experiments with such high plasma current allowed the study on new promising confinement regimes, dominated by a self-organization process with the generation of a helical structure in the plasma core. Presently, RFX-mod is under a significant upgrade (RFX-mod2) to extend the operational scenarios increasing the proximity between the conductive shell and the plasma. This main modification of the magnetic front-end, together with other main improvements, is expected to increase the performance of the machine in both magnetic configurations. In the frame of the studies to exploit the new potential of RFX-mod2 achieving a higher plasma current and longer flat-top duration, a solution based on additional magnetic energy storage has already been proposed. The study of RFP physics at higher plasma current could be crucial to confirm positive trends, such as electron temperature and persistence of quasi single helicity states with the plasma current, and to explore and achieve improved confinement states. This article presents an alternative reconfiguration of the poloidal power supply system of RFX-mod2 based on a combined resistor–capacitor energy transfer system. This system allows to store energy in capacitor banks during the first phase of the plasma current ramp-up and release it to the plasma when the magnetizing current changes polarity, driving the plasma current over 2.5 MA. The proposed upgrade does not involve radical poloidal CPS modifications, maintaining the present converters number and ratings and remaining within the power limits of the main power transformers (300 MVA).

RFX-mod is an experimental fusion device built to explore and study plasma physics, especially in Reversed Field Pinch (RFP) configuration. RFX-mod allowed performing experiments with a plasma current up to 2 MA and obtaining stationary self-organized helical plasmas. Considering the results achieved so far, a significant upgrade of the device (RFX-mod2) is presently being implemented. This paper presents the feasibility study of a significant increase of the plasma current and flat-top duration via a limited upgrade of the RFX power supply system, without overstressing the poloidal windings or overcoming the power limits of the main step-down transformers (300 MVA) and maintaining the present set of thyristor converters units. The application of this concept to RFX-mod2 will be studied starting from the RFX-mod configuration, focusing on the needed modification of the present circuits and the desired improved performance. The increase of the maximum plasma current and flat-top duration, under the current and mechanical limits of the coils, can be reached using additional magnetic energy storage and a complete revision of the configuration and operating principle of the power supply system of the poloidal field coils. With this new operational concept, during the charging phase, the magnetizing winding is charged along with an additional inductor, which acts as an energy storage reservoir. The stored energy, in the order of 100 MJ, is transferred to the magnetizing winding, and thus to the plasma, via transfer resistors in the last phase of the plasma current ramp-up, to reach values of plasma current up to 2.6 MA.

The requirements of ITER neutral beam injectors (1 MeV, 40 A negative deuterium ion current for 1 h) have never been simultaneously attained; therefore, a dedicated Neutral Beam Test Facility (NBTF) was set up at Consorzio RFX (Padova, Italy). The NBTF includes two experiments: SPIDER (Source for the Production of Ions of Deuterium Extracted from Rf plasma), the full-scale prototype of the source of ITER injectors, with a 100 keV accelerator, to investigate and optimize the properties of the ion source; and MITICA, the full-scale prototype of the entire injector, devoted to the issues related to the accelerator, including voltage holding at low gas pressure. The present paper gives an account of the status of the procurements, of the timeline, and of the voltage holding tests and experiments for MITICA. As for SPIDER, the first year of operation is described, regarding the solution of some issues connected with the radiofrequency power, the source operation, and the characterization of the first negative ion beam.

A new Magnetic Energy Storage and Transfer (MEST) system, which can improve the power handling in fusion experiments, has been recently conceived. It is particularly suitable to feed the DEMO Central Solenoid (CS), in principle without the need for resistive switching networks (SNUs), but can be applied to supply the Poloidal Field (PF) coils too. The operating principle of this system, described for one central solenoid circuit, is to pre-charge an additional Superconducting Magnetic Energy Storage (SMES) coil at least up to twice the maximum energy expected in the load (CS) and to transfer the energy from one to the other and viceversa via switched-capacitor. With this approach, the energy is exchanged between the load and the storage system, thus flattening the active power profile to be required from the ac side and substantially nullify the reactive power absorbed. p In this paper, the application of this concept to the European DEMO is studied, starting from the present circuit configuration and from the current and voltage scenario under consideration for the plasma breakdown and ramp-up. A first tentative rating of the system components is reported, discussing also the future R&D steps to explore the industrial feasibility of such a scheme.

Fusion-fission hybrid reactors are seen as a possible, mid-term, CO2-free energy source. Starting from the progress in understanding the reversed field pinch plasma confinement, a pilot neutron source with a configuration of R = 4 m and a = 0.8 m and a plasma current of up to 14 mega-ampere (MA) is proposed. A staged experimental approach, with increased complexity and investment, was identified to tackle the existing issues related to scientific and technological aspects and to test the D-T operation at reduced fusion power (P_fus approximate to 30 MW, Q approximate to 0.4, continuous pulsed operation).

To reach fusion conditions and control plasma configuration in ITER, a suitable combination of additional heating and current drive systems is necessary. Among them, two Neutral Beam Injectors (NBI) will provide 33 MW hydrogen/deuterium particles electrostatically accelerated to 1 MeV; efficient gas-cell neutralisation at such beam energy requires negative ions, obtained by caesium-catalysed surface conversion of atoms inside the ion source. As ITER NBI requirements have never been simultaneously attained, a Neutral Beam Test Facility (NBTF) was set up at Consorzio RFX (Italy), including two experiments. MITICA is the full-scale NBI prototype with 1 MeV particle energy. SPIDER, with 100 keV particle energy, aims at testing and optimising the full-scale ion source: extracted beam uniformity, negative ion current density (for one hour) and beam optics (beam divergence <7 mrad; beam aiming direction within 2 mrad). This paper outlines the worldwide effort towards the ITER NBI realisation: the main results of the ELISE facility (IPP-Garching, Germany), equipped with a half-size source, are described along with the status of MITICA; specific issues are investigated by small specific facilities and by joint experiments at QST and NIFS (Japan). The SPIDER experiment, just come into operation, will profit from strong modelling activities, to simulate and interpret experimental scenarios, and from advanced diagnostic instruments, providing thorough plasma and beam characterisation. Finally, the results of the first experiments in SPIDER are presented, aimed at a preliminary source plasma characterisation by plasma light detectors and plasma spectroscopy.

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.

A new magnetic energy storage scheme is studied for improving the power handling in fusion experiments: it can be applied both to tokamak or RFP experiments to supply the poloidal superconducting coils and can efficiently support the operation of the Central Solenoid (CS), without the need for resistive switching networks, thus with the advantage of energy dissipation avoidance. The basic idea, presented with reference to the CS circuit, is to provide an additional Superconducting Magnetic Energy Storage coil (TC), pre-charged along with the CS one to the same current value and then to transfer step by step the energy from one to the other and vice versa during the different phases of the plasma pulse, via switched capacitor. The principle is applied to the case of a Fusion-Fission Hybrid Reactor, based on Reversed Field Pinch configuration operated exploiting the flux double swing.

The Reversed Field Pinch (RFP) configuration looks to be an attractive option for fusion-fission hybrid reactors: the toroidal magnetic systems would be made of copper coils instead of more expensive superconductive magnets; fusion conditions could be reached by ohmic heating only, therefore additional heating systems would not be required; the fission blanket could be located in the most external part of the torus thus facilitating maintenance operations. The paper aims at assessing the potentialities, such as fuel fertilization and/or nuclear waste transmutation and electricity production of a hybrid reactor with a RFP fusion core (R = 6 m, a = 1) whose conceptual design and plasma performances are based on RFX-mod, the largest RFP experiment currently in operation. Fusion conditions can be reached by heating a D-T plasma up to 9.6 keV by ohmic heating, generated by a 20 MA plasma current induced and sustained by flux swing only. The neutron flux (2.1 x 10(13) fast neutron/cm(2)/s) is used to breed tritium in both the inner and outer blanket sections and induce fission reactions in dedicated areas in the external blanket section where Pu + MA (60%)-Zr (40%) rods are located. Both neutronic and safety analyses corroborate the viability of a FFH reactor with a RFP core.

In June 2018 the SPIDER device, which is the full-size prototype of the negative ion source and extractor for the ITER Heating Neutral Beam has entered the first operation phase at the Neutral Beam Test Facility (NBTE) in Padova, Italy. This paper describes the present status of the device, the experimental plans and the results obtained during the first experimental campaign of the radio-frequency driven plasma source.

A number of modifications 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 Reversed Field Pinch 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.