The ITER blanket system removes the surface heat flux from the plasma and from bulk heating by the neutrons, reduces the activity in the vacuum vessel (W) structural material to the level allowable to ensure vessel reweldability for the ITER fluence goal and, in combination with the vacuum vessel, protects the superconducting coils and other ex-vessel components from excessive nuclear heating and radiation damage. The blanket system contributes with its eddy currents to the passive stabilization of the plasma motion. It minimizes the effects of electromagnetic loads on the VV due to plasma disruptions, and provides a well defined load path to the VV for net vertical and horizontal loads arising from vertical displacement events (VDE's). The system is designed to allow the possibility of replacing the shield with a breeding blanket, within the same dimensional, maintenance, and coolant constraints, to provide the tritium to meet the technical objectives of the Enhanced Performance Phase.

The basic blanket system concept as well as the arrangement and function of its components is essentially unchanged from that established in 19951. However, as discussed in this paper, the design of each component has progressed significantly as a result of the detail design and technical analysis efforts of the last two years. The main components of the blanket system are:

• A back plate: a structure comprising a double wall shell that supports the first wall/shield modules and routes the coolant water to them.

• First wall/shield modules: comprising a plasma facing first wall (FW) section, and a shielding (or later breeding) section. Primary wall and baffle modules are distinguishable by the function of their FW.

• Limiters: define the plasma boundary during plasma start-up and shutdown and are located in equatorial ports.

• Flexible connectors, electrical straps, and branch pipes: the remote handling compatible structural, electrical, and cooling connections between the modules and back plate.

• Filler shields: shielding permanently mounted to the back plate in the triangular gaps between FW/shield modules.

The system will use austenitic stainless steel 316L(N)-IG (ITER Grade) as the primary structural material cooled by water with inlet conditions of 3.8 MPa and 140°C. The plasma facing surface of the FW will be beryllium except the lower region of the baffles, where tungsten is used. The electrical straps and heat sink layer in the FW will be copper alloy. A titanium alloy is the prime candidate material for the flexible connectors.