The ITER project requires additional heating by two neutral beam injectors, each accelerating to 1 MV a 40 A beam of negative deuterium ions, to deliver to the plasma a power of about 17 MW for one ...hour. As these requirements have never been experimentally met, it was recognized as necessary to setup a test facility, PRIMA (Padova Research on ITER Megavolt Accelerator), in Italy, including a full-size negative ion source, SPIDER, and a prototype of the whole ITER injector, MITICA, aiming to develop the heating injectors to be installed in ITER. This realization is made with the main contribution of the European Union, through the Joint Undertaking for ITER (F4E), the ITER Organization and Consorzio RFX which hosts the Test Facility. The Japanese and the Indian ITER Domestic Agencies (JADA and INDA) participate in the PRIMA enterprise; European laboratories, such as IPP-Garching, KIT-Karlsruhe, CCFE-Culham, CEA-Cadarache and others are also cooperating. Presently, the assembly of SPIDER is on-going and the MITICA design is being completed. The paper gives a general overview of the test facility and of the status of development of the MITICA and SPIDER main components at this important stage of the overall development; then it focuses on the latest and most critical issues, regarding both physics and technology, describing the identified solutions.
Abstract
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.
An advanced technology has been developed and employed for the main circuit breakers (CB) of the quench protection circuits (QPC) of the superconducting coils of JT-60SA: it consists in a Hybrid ...mechanical-static CB (HCB) composed of a mechanical Bypass switch (BPS) for conducting the continuous current, in parallel to a static circuit breaker (SCB) based on integrated gate commutated thyristor (IGCT) for current interruption. It was the result of a R&D program carried out since 2006 to identify innovative solutions for the interruption of high dc current, able to improve the maintainability and availability of the CB. The HCB developed for the JT-60SA QPC is the first realization of a dc circuit breaker based on this design approach for interrupting current of some tens of kA with reapplied voltage of some kV. It also represents the first application of hybrid technology with IGCT for protection of superconducting magnets in fusion experiments. The paper aims at giving a comprehensive overview of the main R&D activities devoted to the development of this new technological approach; then, the key aspects of the design, manufacturing and testing of the QPCs for JT-60SA, successfully completed in Naka Site in summer 2015 are presented. Finally, the significance of this research is discussed and the possible future developments, in particular in view of DEMO fusion reactor, are outlined.
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 × 1013 fast neutron/cm2/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.
Outcomes from RFX-mod, the largest Reversed Field Pinch (RFP) experiment, have shown a substantial increase of the electron and ion temperatures versus the plasma current with a trend that allows ...hypothesizing significant D-T fusion processes in larger devices with higher plasma current. Such a device, in which the plasma is purely ohmically heated, could act as an efficient, robust and cheap fusion neutron source with a neutron rate to be used as the basis for a Fusion-Fission Hybrid Reactor (FFHR). The peculiar features of this neutron source can be summarized as: ohmically heated plasma, toroidal field winding rated for a low magnetic field (one-two orders of magnitude lower than in Tokamaks), very simple, robust and cheap construction, simple access for Remote Handling and easy maintenance. Almost continuous neutron production, as required in a hybrid reactor, is guaranteed by a continuously pulsed operation without the need of additional current drive systems.
The aim of the paper is to investigate the relationship between the machine size (major and minor device radius), the attainable stored volt-seconds and, consequently, the maximum plasma current and pulse duration. The analysis uses the experimental RFX-mod data (electron temperatures, plasma currents, loop voltages during flat top) to extrapolate the performances of larger RFP devices with significant neutron production from D-T reactions.
This study shows a realistic possibility, in an inductively operated RFP of larger size than RFX-mod, with plasma current 15–20 MA and temperature ≈ 10–15 keV, to reach the level of neutron generation of a fusion-fission hybrid reactor, leading to a neutron production rate in the range of 1019 n/s and a wall neutron load of 0.2 MW/m2.
The ITER full size plasma source device design Sonato, P.; Agostinetti, P.; Anaclerio, G. ...
Fusion engineering and design,
06/2009, Letnik:
84, Številka:
2
Journal Article, Conference Proceeding
Recenzirano
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/m
2 in D
2 (D
−) and 350
A/m
2 in H
2 (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
m
2.
The main design choices will be presented in the paper as well as an overview of the design of the main components and systems.
This paper describes the feasibility studies of a hybrid DC circuit breaker (CB) for quench protection of superconducting magnets rated for current up to some tens of kiloamperes (kA) and voltage of ...several kilovolts (kV). The proposed scheme is based on a mechanical switch paralleled to a fully controlled static CB; integrated gate commutated thyristors (IGCTs) were selected. In normal operation, the mechanical switch handles the continuous flow of DC current minimizing the on-state losses. When protection against quench is requested, the current is diverted into the static breaker that quickly interrupts the current, withstanding the reapplied voltage of several kV. The reliable turn-ON of several IGCTs in parallel with low voltage applied between anode and cathode is an issue to be assessed to confirm the feasibility of this design solution. The paper highlights that the components commonly used as snubber or protection circuit can guarantee the turn-ON of all the devices in parallel and that the IGCTs can be turned-ON with few volts applied; this latter aspect was proved by means of specific tests. Finally, a comparison between a static only and a hybrid CB is performed, highlighting the benefits introduced by the hybrid solution.
•FRESCO is a code for rapid evaluation of the cost of electricity of a fusion power plant.•Parameters of the basic machine and unitary costs of components derived from ITER.•Power production ...components and plant power balance are extrapolated from PPCS.•A special effort is made in the investigation of the pulsed operation scenarios.•Technical and economical FRESCO results are compared with those of two PPCS models.
FRESCO (Fusion REactor Simplified COsts) is a code based on simplified models of physics, engineering and economical aspects of a TOKAMAK-like pulsed or steady-state fusion power plant. The experience coming from various aspects of ITER design, including selection of materials and operating scenarios, is exploited as much as possible.
Energy production and plant power balance, including the recirculation requirements, are derived from two models of the PPCS European study, the helium cooled lithium/lead blanket model reactor (model AB) and the helium cooled ceramic one (model B). A detailed study of the availability of the power plant due, among others, to the replacement of plasma facing components, is also included in the code.
The economics of the fusion power plant is evaluated through the levelized cost approach. Costs of the basic components are scaled from the corresponding values of the ITER project, the ARIES studies and SCAN model. The costs of plant auxiliaries, including those of the magnetic and electric systems, tritium plants, instrumentation, buildings and thermal energy storage if any, are recovered from ITER values and from those of other power plants.
Finally, the PPCS models AB and B are simulated and the main results are reported in this paper.
The paper reports a comprehensive overview of the power supply (PS) systems of the neutral beam (NB) injectors of the main fusion experiments in view of introducing an alternative concept, with ...respect to the reference design, for the power supplies feeding the ion source of the international thermonuclear experimental reactor (ITER) NB injector. The overview is proposed to understand the evolution of the systems, analyze the motivations behind the design choices and highlight how they were often the results of a compromise between contrasting requirements. The approach adopted in the PS reference design of the ITER NB injector is described and the results, in particular in terms of feasibility of some devices and accessibility issues, are discussed. The second part of the paper is dedicated to the description of the alternative concept for the ion source power supply (ISPS) design; the solution proposed is discussed with the aim of evaluating and comparing the different implications with respect to the reference design.
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