TABLE C.1 Parameters for Burning Plasmas in the International Thermonuclear Experimental Reactor (ITER)



Major radius

6.2 m

Minor radius

2.0 m

Magnetic field

5.3 T

Plasma current

15 MA

Fusion power

500 MW

Q (fusion power/power in)


Burn time

≥400 s

Wall loading

0.57 MW/m2

Plasma volume

837 m3

Heating/current drive power

73 MW


SOURCE: Information obtained from the ITER Web site, Accessed September 1, 2003.

the technology that will be required for a fusion power plant. A cutaway figure of the device is shown in Figure 1.1 in Chapter 1 of this report, and the ITER operating parameters are summarized in Table C.1. ITER is a $5 billion device that utilizes reactor-relevant fusion technologies, including superconducting magnets and techniques for control of the plasma profiles, to create self-heated plasmas.

The ITER project has benefited greatly from the expertise and scrutiny of fusion-plasma researchers throughout the world. The present design is the result of a decade of effort. This work included one major redesign that reduced the anticipated cost by a factor of 2 by reducing the size and eliminating some of the capability to test fusion power components and technologies. The engineering design of ITER is well developed, and prototypes for many of the systems have been built. ITER has been designed to accommodate a range of heating and current drive technologies and to have the most complete set of plasma diagnostics of the three proposed burning plasma experiments. It will facilitate studies of plasmas for pulse lengths much longer than the plasma current redistribution time, which will enable studies of steady-state operation. The long pulse capability, the range and flexibility of heating and current drive technologies, and the extensive diagnostic set provide the capability to explore and evaluate advanced, steady-state operating regimes. The present ITER design would demonstrate the integrated operation of some of the important technologies for fusion power. It also has the capability to test some of the key nuclear components necessary for a fusion power plant, such as tritium breeding blanket modules required to close the deuterium-tritium (D-T) fuel cycle.

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