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International audience; The next step for fusion energy after the ITER tokamak is the demonstration power plant DEMO. In this framework , system codes are used to address high-level key design issues for the DEMO pre-conceptual phase. They aim at capturing the interactions between the subsystems of a fusion reactor. SYCOMORE is a modular system code which includes physics and technology models coupled to an optimizer in order to explore a large design parameter space. In the present paper, trade-off studies focused on technology modules are reported including the influence of some design-driving assumptions on the reactor performances and size, starting from a European DEMO1-like design (more than 500 MW net electric power and 2 h burn duration). The increase of the mechanical stress limits in TF and CS magnets can help reducing the reactor size, slightly more when high temperature superconductors are used in the TF coil. The tritium breeding ratio can be improved to more than 1.10 by a moderate increase of the size, but the tritium burn-up ratio needs one additional meter of major radius for every percent increase. Divertor coolant options are also compared, showing some differences between helium, hot and cold water scenarios at various incident divertor heat fluxes.

Content: EU DEMO studies for pulsed (DEMO1) and steady-state (DEMO2) concepts are currently in the pre-conceptual phase [1]. DEMO1 aims at producing about 2GW of fusion power with a burn time of approximately 2 hours. Within EUROfusion Power Plant Physics and Technology department, DEMO scenario modelling is carried out as part of the validation of feasibility and performance of DEMO designs. One of the most challenging activities deals with numerical investigations of DEMO1 transient phases including ramp-up and ramp-down, which exhibit peculiar issues with respect to existing devices. Studies on ramp-up have been carried out to highlight the effects of different ramp-up options in terms of robustness of the access to the desired flattop scenario. A heating power during ramp-up, additional to the one required during flattop, appears to be necessary for plasma burn initiation and access to H-mode, with Paux,RU50MW depending on the uncertainties on L-H transition scaling. Current ramp-rate and heating power influence also plasma position controllability, and results are presented in terms of the achieved li(3) and bp. Additional power requirements and integration of different systems which are relevant for DEMO heating mix assessment are here discussed. Ramp-down phase in DEMO poses specific issues on vertical stability given the distance of control actuators from the plasma. Ramp-down trajectories with controllable plasma boundaries have been coupled to transport studies showing the necessity of additional ramp-down heating power to avoid radiative plasma collapses. Off-axis power deposition helps plasma controllability, together with a current ramp-rate100kA/s. Plasma radiation also dominates the H-L transition, which is investigated and appears to be a critical step in terms of plasma control. DEMO performance is strongly linked to the maximum plasma elongation, which has to be assessed comparing different ramp-down trajectories. [1] G. Federici et. al, Fus. Eng. Des. 89, 882-889 (2014)

During a tokamak discharge, several control modes may have to be run in sequence in order to perform the control of the different discharge phases. The transitions between these control modes are not always easy to handle because in most cases the coupling between the controlled plasma quantities is not taken into account in each control mode design process. This paper presents a new Multi-Inputs/Multi-Outputs (MIMO) controller applied on Tore Supra to control both plasma current and flux variations through the central solenoid voltage and the lower hybrid current drive (LHCD) system power. It deals with the transition from a loop voltage floating mode to a loop voltage control mode. The controller, synthesized and tuned using a model-based approach, has been validated in simulation before its successful implementation on Tore Supra experiments.

A methodology has been developed to consistently investigate, taking into account main reactor components, possible magnet solutions for a pulsed fusion reactor aiming at a large solenoid flux swing duration within the 2–3 h range. In a conceptual approach, investigations are carried out in the equatorial plane, taking into account the radial extension of the blanket-shielding zone, of the toroidal field magnet system inner leg and of the central solenoid for estimation of the pulsed swing. Design criteria are presented for the radial extension of the superconducting magnets, which is mostly driven by the structures (casings and conductor jacket). Typical available cable current densities are presented as a function of the magnetic field and of the temperature margin. The magnet design criteria have been integrated into SYCOMORE, a code for reactor modeling presently in development at CEA/IRFM in Cadarache, using the tools of the EFDA Integrated Tokamak Modeling task force. Possible solutions are investigated for a 2 GW fusion power reactor with different aspect ratios. The final adjustment of the DEMO pulsed reactor parameters will have to be consistently done, considering all reactor components, when the final goals of the machine will be completely clarified.

A 20 MW Lower Hybrid Current Drive system using an antenna based on the Passive-Active Multijunction (PAM) concept is envisaged on ITER. This paper gives an overview of the mechanical analysis, modeling and design carried out on two major elements of the antenna: the grill front face, and the RF feed-through or windows. The front face will have to withstand high heat and fast neutrons fluxes directly from the plasma. It will be actively cooled and present a beryllium coating upon ITER requirement. The RF window being a critical safety importance class component (SIC) because of its tritium confinement function, two of them will be put in series on each line to achieve a double barrier. A design of a water cooled 5 GHz CW RF “pillbox” window capable of sustaining 500 kW of transmitted power is proposed. Both studies allow to move forward, and focus on critical issues, such as manufacturing processes and R&D associated programs including tests of mock-ups.

This paper discusses the scientific and technical requirements for a Neutral Beam Injection system on the FAST tokamak and describes a preliminary conceptual design of a suitable injector. FAST is being proposed as a European experiment in support to the operations on ITER and to the design of DEMO. The specific mission of this device is an integrated approach to a number of outstanding burning plasmas physics and operational issues with an emphasis on the impact of fast particles on turbulent transport. Such scientific requirements set a series of technical challenges regarding the injector and the coupling of the injector to the FAST main chamber that are addressed in the paper. A preliminary conceptual design of the injector is proposed which attempts to meet the stated requirements.

The feedback control of fusion experiments in tokamak devices is entering a new area driven by the increase of control requirements for obtaining burning plasmas under safe operation conditions. A project aiming at setting up a flight simulator for the development of advanced controllers has started last year at CEA. This simulator will reuse most of the components of the European Integrated Tokamak Modelling (ITM) simulation platform. Thus, it will benefit from the development made by the task force and it will be able to offer a development platform for the new controllers of present day European tokamaks and future machines. This paper provides an overview of the architecture of the simulator. The functional specifications of the simulator have been defined and the needs in interface implementation are analysed as well.