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The European technology program for the DEMO blanket is presently focused on two main blanket concepts: Helium Cooled Ceramic Blanket (HCCB) and Self Cooled or Water Cooled liquid Lithium-Lead Blanket (SC or WC-LiPbB). Both these concepts are affected by some critical feasibility issues, i.e.: need of beryllium (chemical toxicity, high cost) as a neutron multiplier in HCCB, corrosion and activation products in LiPbB concepts, as well as MHD effects specifically in SC-LiPbB, tritium permeation and marginal breeding capability in WC-LiPbB. On the basis of these considerations, use of pure liquid lithium was reconsidered as a possible alternative for DEMO blanket applications. Its properties, advantages and drawbacks have been reviewed and a conceptual blanket design has been developed. The results show that the Helium Cooled liquid Lithium Blanket (HCLiB) is an attractive concept from the thermohydraulic and neutronic points of view even though concerns remain with respect to the chemical reactivity of liquid lithium.

During the ITER conceptual design phase a reference driver blanket concept has been selected, based on the use of ceramic breeder (LiAlO2 or Li2O) with low-temperature (60–100 °C) water cooling and on-line tritium recovery via He purging gas. Since tritium release from breeder materials requires operation temperature > 400–450 °C, the combination of low-temperature cooling and tritium on-line recovery are conflicting design requirements, leading to the need of insulating thermal barriers between breeder and coolant which provides an increase of design complexity and feasibility concerns. To overcome this problem while maintaining the benefits of low temperature and pressure cooling a tritium batch recovery procedure has been preliminarly analysed: the tritium could be accumulated in the breeder material during normal low-temperature operation and then periodically recovered during scheduled reactor shutdown by heating the breeder via hot gas or steam and using decay heat. The results show that, assuming a maximum allowable tritium accumulation of about 500 gr in the breeder material, one reactor shutdown every 5 days of continuous burning (corresponding to about 40 normal operation days at 12% reactor availability) is required for tritium extraction. The tritium recovery phase is expected to take 0.5 to 1.5 days, depending on the specific ceramic breeder used in the blanket and the recovery temperature.

The Computer code, TRIDYN, has been developed to study the mass and energy transient of a PIN blanket concept under pulsed load (burn/off-burn). In the model the tritium release is controlled by means of two mechanisms: diffusion within the bulk and desorption at the surface of the grain. The model includes the thermal profile evaluations in the breeder and, by taking into account the thermal transient, evaluates the tritium inventory, permeation, and purge concentration evolution.

Two large size mock-up panels have been manufactured within a comprehensive ENEA-Ansaldo development program for the NET first wall. The panels, manufactured according to the fully brazed solution, will be part of a thermal cycling campaign to be started at ISPRA by the end of the year. The paper describes all the steps followed for the manufacturing and the non destructive examination, together with the analytical work.

In the ceramic blanket concept, a sweep gas is required to recover the “in situ” tritium produced. The tritium is removed mainly in an oxidate form. Two process options for the tritiated water reduction step have been developed in the present work. Both options have been studied under a NET contract. The reduction is achieved by means of a catalytic water gas shift reactor. A model has been developed to evaluate the behavior of all the process components. The hydrogen isotopes produced are removed by means of a Pd/Ag membrane separator, and the residual tritiated water is trapped in molecular sieves and recirculated to the shift reactor.

The advanced ENEA poloidal solid breeder-in-tube (BIT) blanket design for a DEMO fusion reactor is presented. Its main operating conditions and features are max. neutron fluence = 3×1022 n/cm2; max. beryllium vol. swelling = 4%; max. beryllium differential swelling = 3%; module poloidal length = 8.5 m; helium inlet/outlet temperature = 250/500°C; pumping power percentage >4%; max. structure temperature >550°C; ceramic breeder min/max. temperature 450/750°C; purge circuit separated from the coolant circuit; coaxial manifolding system; 3-D value of the t.b. r ≈ 1.20; and thermal cycle efficiency 35%.