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A model has allowed us to perform the analysis of the tritium inventory and permeation in the international thermonuclear experimental reactor (ITER) breeding blanket under the hypothesis of steady state conditions. Li 2 ZrO 3 (reference) and Li 2 TiO 3 (alternative) have been studied as breeding materials. The total breeder inventory assessed is 7.64 g for the Li 2 ZrO 3 at reference temperature. The model has also been used for a parametric analysis of the tritium permeation. At reference temperature and purge helium velocity of 0.01 m/s, the HT partial pressure is ranging from 10 to 30 Pa in the breeder and 1.5×10 −3 Pa in the beryllium. At 0.1 m/s of purge helium velocity, the HT partial pressure is reduced of one order by magnitude in the breeder and becomes 5×10 −5 Pa in the beryllium. The tritium permeation into the coolant for the whole blanket is ranging from 100 to 250 mCi per day for purge helium velocity of 0.01 m/s. The analysis of the tritium inventory and permeation for the alternative Li 2 TiO 3 breeding material has been carried out too. The tritium inventory in the breeder is in the range from 6 to 375 g larger than in Li 2 ZrO 3 by about a factor 5; the tritium permeation into coolant is comparable to the Li 2 ZrO 3 one. This analysis provides indications on the influence of the operating parameters on the tritium control in the ITER breeding blanket; particularly the control of the tritium inventory by the temperature and the tritium permeation by the purge gas velocity.

A computer model has been developed to evaluate tritium permeation to coolant and in vanadium tubes inventory in Safety and Environmental Assessment of Fusion Power (SEAFP) blanket. The mean tritium partial pressure in gaseous breeder phase are in the range from 0.5 to 5 Pa for helium purge gas velocity from 0.1 to 0.4 m/s; in these conditions the tritium permeation to coolant changes from 32.8 to 16.4 g/day and the tritium inventory in vanadium tubes from 4000 to 2000 g. The H/T ratio involves a relevant tritium permeation variation: with 0.2 m/s helium purge gas velocity varying the H/T ratio from 100 to 50 the tritium permeation to coolant ranges from 23.2 to 32.7 g/day. The analysis shows that defects free thin permeation barriers (SiC and Al{sub 2}O{sub 3}) are very effective to making negligible the tritium permeation to coolant and the tritium inventory in tubes. 9 refs., 5 figs., 1 tab.

The analysis of the tritium inventory in Li 2 O, carried out for the Safety and Environmental Assessment of Fusion Power (SEAFP) helium-cooled ceramic blanket, is based on a diffusion and desorption tritium release model. Within the specific range of breeder temperatures taken into account, desorption was the dominant mechanism so it can be defined as the rate controlling step. At steady state, the model for the tritium inventory in the solid Li 2 O breeder is supported by a computer code for several operating conditions. At reference conditions of breeder temperatures, by varying the mean grain radius from 1 to 5 μm, a tritium inventory from 0.5 to 2.8 g can be obtained. A helium purge gas velocity from 0.1 to 0.4 m/s gives rise to gas pressure losses from 0.22 to 0.9 MPa, which could probably be reduced by increasing the pebble diameter to 1 mm. This breeder configuration seems to ensure reactor safety.

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.

One of the objectives assigned to a DEMO blanket will be to limit tritium leakage to the steam circuit by permeation through the steam generator tubing of the tritium unavoidably present in the coolant. The control of these losses will require, on one hand, to minimize the amount of tritium reaching the coolant under both nominal and accidental conditions, and, on the other hand, to have an efficient tritium recovery process and chemistry control for maintaining at a low level the concentration in reduced tritiated species in the coolant.

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.

Tritium diffusion and transport behaviour through porous γ-LiAlO2 solid breeder rods of a blanket for a Commercial Tokamak, has been studied for a typical reactor start-up from a “clean” breeder. The proposed tritium transport model accounts for grain diffusion, (using a simplified model), desorption from the surface and pore diffusion, and has been used for the evaluation of the overall characteristic time of the transient tritium release into the helium purge gas. Because of the strong dependence of diffusion and desorption coefficients on temperature, also a detailed thermal analysis has been performed. This study considers a bidimensional axisymmetrical geometry for the blanket and the calculations have been carried out using the general-purpose Finite Element program TRIO E.F.

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...

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.

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%.

Within the frame of the ITER conceptual design phase a Water Cooled Ceramic Blanket (WCCB) design has been developed by ENEA/ANSALDO, in close collaboration with the NET Team, for application as ITER Driver Blanket. Such a design has been selected as European design proposal for the ITER Driver Blanket and is now under comparison with the Japanese, United States and Soviet Union design proposals. An out of pile experiment is being performed to investigate the thermal-hydraulic and mechanical performance of a typical blanket module. Status of design and experimental activities is reported together with a summary of the main features and performances of the EC blanket, based on detailed analyses performed in several fields of investigation (Neutronics, Thermal-hydraulics, Thermomechanics, Tritium Dynamics, Safety, Industrial Feasibility, etc) during the past two years.

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.