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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 blanket option featured by helium cooling and lithiated ceramics inside tubes is under development at CEA/ENEA within the framework of the European Test Blanket Program. A first conceptual design...

In order to reduce toxicity in elderly patients with newly diagnosed APL, since 3/1997 the Italian cooperative group GIMEMA evaluated an amended AIDA protocol for patients aged > 60 years, consisting of the same induction with ATRA and Idarubicin but only 1st consolidation course (Idarubicin + Cytarabine), followed by 2 years maintenance with ATRA alone. Up to now, 56 patients (25 males and 31 females, median age 66.2 years, 46 with PS 0-1 and 10 with PS 2) are fully evaluable. At onset, according to GIMEMA-PETHEMA risk score, 18 were low-risk(32.5%), 31 intermediate risk (55%) and 7 high risk (12.5%). After induction treatment, 54 patients (96.4%) achieved CR and 2 (3.6%) died from haemorrhage (1) and infection (1). ATRA syndrome was documented in 5 patients (9.3%): 13/56 patients (23.2%) showed during induction other toxicities (WHO 3 – 4) not related to ATRA. After CR achievement, 2 patients died in CR from haemorrhage (1) and infection (1) and 52 received the consolidation course: on the whole, during consolidation 4 patients (7.6%) had a toxicity WHO 3 – 4 and 2 of them (3.8%) died from haemorrhage (1) and infection (1). The remaining 50 patients started maintenance treatment: up to now, 12 patients (22.2%) relapsed, after a median time from morphological CR of 19 months (range 7 – 86). Overall survival (OS) was 76.1% and 73.3% at 3 and 5 years, respectively. Disease free survival (DFS) was 64.5% and 61.3% at 3 and 5 years, respectively. At the univariate analysis, PS =2 (p=0.0019), WBC count > 3 x 109 /l (p=0.018) and male gender (p=0.03) had a bad prognostic impact on DFS, while only PS=2 (p=0.05) did it on OS. Age, PLTS count, WBC count > 10 x 109 /l, and risk score did not affect both OS and DFS. At the multivariate analysis on DFS, only PS =2 retained prognostic significance (HR = 3.8). In conclusion, the amended GIMEMA protocol has shown to be effective and safe in elderly APL patients, as the rate of death in CR was reduced when compared with previous results in not amended GIMEMA LAP AIDA 0493: however, to face with a relapse rate > 20%, future strategies might be designed which exploit the use of more targeted approaches including combinations of ATRA, arsenic trioxide, and anti-CD33 monoclonal antibodies.

The proposed first wall (FW) for the NET/ITER (Next European Torus/International Thermonuclear Experimental Reactor) consists of a box structure, made of stainless steel, with an active cooling of the side facing the plasma by a flow of water through the tubes inserted into it. An innovative brazing process to improve the quality of the brazed joints between cooling tubes and plates has been developed and optimized within the framework of NET technological task PDT 1.2. The brazing technique is based on creep expansion of the tubes during the high-temperature holding time of the brazing treatment, which minimizes the gap that was initially present around the tubes. A flat, mockup FW panel with plate dimensions 622*216*33 mm and with eight cooling tubes was built using this technique. Thermal fatigue tests are scheduled to be carried out in the near future on this mockup. An automatic, computerized procedure for the nondestructive examination, of the brazed joints using ultrasonic probes has also been developed. >

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%.

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.

Empirical scaling laws for electron energy confinement in Ohmically heated tokamaks, for plasma regimes with Zeff ≈ 1, are shown to be compatible with dimensional constraints imposed by the collisional high-β model. New simple forms of the relevant transport coefficient which are consistent with the observed profiles are derived from the model.

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.

The feasibility of a hybrid configuration superconducting coils/copper coils for a next generation tokamak TF magnet has been investigated. On the basis of this hybrid solution, the conceptual design has been developed for a medium-high toroidal field tokamak reactor (HTMR). The results of this study show the possibility of designing a tokamak reactor with reduced size in comparison with other INTOR like devices, still gaining some margins in front of the uncertainties in the scaling laws for plasma physics parameters and retaining the presence of a blanket with a tritium breeding ratio of about 1.

A theoretical model of the wall-free stream interaction, in MHD generators with turbulent non-equilibrium plasma, due to the variation of the gas parameters between opposite electrodes, is presented. The model allows one to evaluate, as a consequence of the gasdynamic boundary layers, some key parameters affecting the bulk of the generator, once the distribution of the gas properties in the cross-section of the duct is given. These parameters are the direction of the average current in the bulk and the total electrode voltage drops.

Ionization instability in a stratified non-equilibrium plasma subjected to crossed E and B fields is investigated by means of a simple model without diffusion processes and boundary effects. An inhomogeneous infinite plasma, consisting of alternate one-dimensional layers of high and low electron temperature (T02>T01) lying in the plane perpendicular to the magnetic field is taken as the unperturbed state. It is shown that, in order that a periodic stratified plasma be stable on the whole, some conditions on the thicknesses of the layers must be fulfilled and at least one of the two sets of layers (e.g. that with electron temperature T01) must be intrinsically stable (in the sense that a uniform infinite plasma with electron temperature T01 must be stable with respect to the ionization instability).

As an application of the general theory described in a previous paper (1), the evolution of a weakly unstable, infinite plasma, with immobile ions, and without external fields, is studied for the so-called bump-in-tail case. It is shown that, in the quasi-linear approximation, the equations of motion of the electron distribution function and the spectrum of plasma waves are equivalent to the usual adiabatic formulation. The present approach yields the equation of motion for the plasma waves in the form of a diffusion equation in wave-number space.

In deriving the dispersion equation of ionization waves in a (low temperature) seeded MHD plasma, allowance is made for an electron current along the impressed magnetic field. This leads to a critical value of the Hall which depends on the angle ( alpha ) between current and magnetic field by a factor 1/sin alpha . The most unstable waves no longer travel at right angles relative to the magnetic field direction; their increment and direction are given in terms of alpha and magnetic field strength.

The evolution of a weakly turbulent, unstable, spatially uniform plasma is considered by means of a kinetic theory based on the Klimontovich equation for a single species of interacting particles. The usual expansion in powers of the amplitude of the electric field is used. This leads to a quasi-linear analysis in which the effects of the discreteness of matter are explicitly taken into account. A pair of coupled equations for the one-particle distribution function and the field fluctuations is deduced. In the special case of small growth rates the latter is transformed into an equation for the spectrum of the fluctuating energy density. The results are compared with those of previous treatments.

The complete dispersion relation is analytically solved by neglecting the electron pressure term in the Ohm's law together with the heat convection in the electron energy equation. Within this limit, it results that the threshold of ionization instability does not depend on the relaxation rates within a broad range of operating conditions for the MHD energy conversion method.