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Depending on the final decision on the operation time of ITER (International Thermonuclear Experimental Reactor), the Driver Blanket might become a basic component of the machine with the main function of producing a significant fraction (close to 0.8) of the tritium required for the ITER operation, the remaining fraction being available from external supplies. The Driver Blanket is not required to provide reactor relevant performance in terms of tritium self-sufficiency. However, reactor relevant reliability and safety are mandatory requirements for this component in order not to significantly affect the overall plant availability and to allow the ITER experimental program to be safely and successfully carried out. Within the framework of the ITER Conceptual Design Activities (CDA, 1988–1990), a conceptual design of the ITER Driver Blanket has been carried out by ENEA Fusion Dept., in collaboration with ANSALDO S.p.A. and SRS S.r.l., and in close consultation with the NET Team and CFFTP (Canadian Fusion Fuels Technology Project). Such a design has been selected as EC (European Community) reference design for the ITER Driver Blanket. The status of the design at the end of CDA is reported in the present paper.

The DEMO specifications defined so as to ensure the consistency of the various blanket conceptual design studies performed within the framework of the European Test Blanket Program me have been recently updated. A very first attempt has been made to adapt the European Ceramic Breeder Inside-Tube DEMO blanket to these new specifications. Two solutions have been investigated. The first would ensure tritium self-sufficiency of the plant with a large safety margin. The other one, which fully preserves the design simplicity and reliability of the initial design, appears to be somewhat marginal from the tritium breeding capability point of view, but to offer good improvement prospects.

A conceptual study of a helium-cooled solid breeder blanket for a tokamak reactor is presented. Tritium breeding capability together with system reliability are taken as the main design criteria. The blanket consists of tubular poloidal modules made of a central bundle of ceramic rods (γLiAlO 2 ) with a coaxial distribution of the inlet/outlet coolant flow (He) surrounded by a multiplier material (Be) in the form of bored bricks. The Be to γLiAlO 2 volume ratio is 4/1. The He inlet and outlet branches are cooling Be and γLiAlO 2 , respectively. A purge He flow running through small central holes of the ceramic rods is derived from the main flow. Under the typical conditions of a tokamak reactor (neutron wall load = 2 MW/m 2 ), a full coverage tritium breeding ratio of 1.47 is achieved for the following design and operating parameters: outlet He temperature = 570 °C; inlet He temperature = 250 °C; total extracted power = 2700 MW; He pumping power percentage = 2%; minimum/maximum γLiAlO 2 temperature = 400/900 °C; maximum structural temperature = 475 °C; and maximum Be temperature = 525 °C.

In collaboration with ANSALDO, ENEA is performing a design and manufacturing feasibility study for the first wall of NET during the physics operation phase. The main design specifications are the average neutron wall load = 1 MW/m 2 , peak surface heat flux = 0.4 MW/m 2 , total number of bum pluses = 1×10 4 , average burn pulse duration = 100 s, structure material = ss 316L S.A., coolant = H 2 O (He) at 50 °C. The ENEA-ANSALDO (reference design is based on the use of flat plates coupled by nicrobrazing to water poloidal cooling tubes. The final aim of the work is manufacturing a representative NET first wall box segment (mockup of indicative dimensions 0.65×0.25×0.15 m 3 ) suitable for testing in the 190 kW Thermal Fatigue Test Facility at JRC-Ispra. In this paper we report on the related design work and the results of the studied performed: technological development (manufacturing test activity, brazed joints characterization tests, mockup design and manufacturing), thermomechanical analysis, plasma-wall interaction studies (tests on protective carbon composite material, theory), and tritium transport analysis.

Stability of thermal equilibria of cylindrical plasmas are studied in the case where ohmic heating is balanced by thermal diffusion, for a wide class of heat conductivities depending both upon the temperature and the azimuthal magnetic field.

Choking effects due to Lorentz force in a gaseous MHD generator with segmented electrodes are analysed by means of the quasi-one-dimensional approximation. As an illustrative example, the flow of a perfect gas with constant electrical conductivity through a constant area and constant load factor duct is considered in detail. In the case of such a duct fed by an isentropic converging nozzle, maximum or choking flow and critical back pressure are evaluated in terms of electrical parameters, magnetic field strength and duct length, for fixed supply conditions. This allows one to acquire some quantitative grasp of choking and its relevant effects on the practical operation of an MHD generator.

The interaction of gas flow and electric current in the bulk of a subsonic linear generator for closed cycle is studied by means of a steady quasi-one-dimensional analysis. The characteristic of this interaction is the strong dependence of the gas electrical conductivity upon the flow Mach number, which can lead to rapid changes of electrical and gas properties along the channel. The influence of ionization turbulence on the electrical conductivity of the gas is taken into account by means of two different models based on the laminar electron energy balance (model I) and on the approximate average electron energy balance (model II), respectively. The theoretical results pertaining to the transverse current along the channel are compared with the experimental data obtained with the Frascati MHD facility. A good agreement is found with calculations based on model I, while model II leads to a substantial overestimation of the current and, hence, of the generator power output and electrical efficiency.