In the development of magnetic fusion power plants, the tritium breeding function is effectively integrated in the blanket, which also serves as the main thermal power conversion system and an effective shield for the adjacent reactor components from neutrons and g-rays. An integrated tritium breeding blanket acts as a shield, as a heat exchanger, and as a breeding zone of tritium fuel, as pictured in Figure 2. The blanket has a primary or first wall that faces the plasma, and this component is in direct contact with the edge of the fusion plasma. The latter is typically designed to remove the plasma radiative power and part of the nuclear heating, either with or without a cooling circuit separate from the blanket system.

In order to obtain a closed D-T fuel cycle for a fusion power plant, it is mandatory that the tritium production rate is, at least effectively, equal to its consumption rate and accounts for decay and losses at scheduled or unscheduled plant outages; this prin­ciple is usually called ‘tritium self-sufficiency.’ These conditions will not be achieved in near-term fusion devices, where tritium resources available from fis­sion plants can be used and where production from a so-called ‘driver’ blanket is an additional or alterna­tive source for fuel supply.

Effective tritium production requires that the lith­ium compounds are located in such a way that the

maximum capture of D—T neutrons is obtained in the so-called tritium breeding blanket. As most fusion devices require partial use of the plasma­facing area for plasma heating, plasma diagnostics, plasma control, and fuel exhaust, the effective cap­ture of D—T neutrons for breeding tritium requires the use of neutron multipliers in the blanket. Net tritium breeding ratios (TBRs) foreseen for power plants should be about 1.05—1.1.

The breeder material used in blanket designs that have attractive thermal efficiency for magnetic con­finement power plants should conform to certain requirements. It should

1. breed tritium in a relatively small volume with a high production rate

2. release tritium in a manner that allows fast proces­sing into plasma fueling

3. possess physical and chemical stability at high temperature

4. display compatibility with adjacent structures and other blanket components

5. exhibit adequate irradiation behavior

6. not pose specific safety risks under off-normal and accidental conditions

7. have activation characteristics allowing recycling or treatment as low-active waste.

Lithium-based ceramics are recognized as attrac­tive tritium breeding materials for the first generation of fusion power plants, due to their inherent thermal stability and chemical inertness.

This chapter describes the development ofceramic breeder (CB) blankets and material production routes applied or investigated and summarizes the properties and R&D results for a number of lithium-based ceramic materials.1 In this chapter the chemical for­mulas are used, though the actual composition are very often non-stoichiometric, which is more evident when a larger fraction of lithium has been burned. Most of the work presented in this chapter is subject to rapid evolutions in local, national or international programs. The authors like to stress that any of the activities, e. g. those concerning ITER can be quite different in their evolution.