ITER operation, nominally expected to last 21 years overall, is divided into five main phases.

Integrated Commissioning

This phase completes the construction of ITER by ensuring all systems operate together and includes the preparation of the machine to attain the first hydrogen plasma.

Hydrogen Phase

This phase allows full commissioning of the tokamak system in a non-nuclear environment without depending on fully-remote handling.

Major aspects of the full DT discharge scenario can be checked, including:

• plasma current initiation,
• current ramp-up,
• formation of a divertor configuration,
• current ramp-down,
• control of and loads due to disruptions or vertical displacement events,
• access to H-mode and adequacy of heating installed,
• operational limits on density, beta, safety factor,
• requirements/capability of steady state operation.

The peak heat flux onto the divertor target will be of the same order of magnitude as for the full DT phase.

Some important issues cannot be fully tested in this phase. These include:

• evaporation of the divertor target surface expected during a disruption,
• effects of neutron irradiation of the in-vessel materials,
• alpha-particle heating of the plasma.

Although there are no neutrons in this phase, test blanket module electromagnetic and hydraulic tests can take place and give very useful information further iterated system designs to be installed for DT operation.

The actual length of this phase depends on the merit of the ongoing experimentation wiith regard to the later DT operation, in particular the abiloty to achieve good H-mode confinement with a suitably high plasma density.

Deuterium Phase with low T

In this phase, neutrons will be produced, and tritium will be produced from DD reactions. Part of this tritium will then be burnt in DT reactions. Although the fusion power is low, the activation level inside the vacuum vessel will not allow human access after a few deuterium discharges with powerful heating. However, the capacity of the heat transfer system (except for the divertor and heating devices) could initially be minimal, and demand for the tritium processing system would be very small. Since tritium already exists in the plasma, addition of a small amount of tritium from an external source will not significantly change the activation level of the machine. So later in the phase integrated DT commissioning can take place, with short pulses at high fusion power.

The major achievements would be as follows:

• replacement of H by clean D plasma;
• confirmation of L-H threshold power and confinement scalings;
• establishment of a reference plasma (current, heating power, density, detached/semi-detached divertor, ELMy H-mode, etc.);
• particle control (fuel/ash/impurity/fuelling/pumping);
• steady-state operation with full heating power;
• finalisation of nuclear commissioning with a limited amount of tritium;
• demonstration of high fusion power comparable to the nominal value for the full DT burn, for a short time.

Some information can be provided in this phase on test blanket neutronics behaviour, allowing optimisation of the designs for later DT operation.

Low Duty DT Phase

During this phase the fusion power and burn pulse length will be gradually increased until the inductive operational goal is reached. Non-inductive, steady-state operation will also be developed. Test blanket modules will begin to accumulate results in a situation resembling their operating environment, allowing fine tuning of the designs, and a reference mode of operation for that testing will be established.

High Duty DT Phase

This phase will try to improve overall performance, emphasise testing of components and materials with higher neutron fluences, aim for high availability, and further improved modes of plasma operation. The implementation and length of this phase will be depend on the results from the preceding three phases and assessment of the merits and priorities of programmatic proposals.

Whether and when to incorporate tritium breeding during this phase will be decided on the basis of the availability of tritium from external sources, the results of breeder blanket testing, and experience with plasma and machine performance. Such a decision would lead to a non-operating period of about 2 years while the blanket system is installed in the outboard plasma region, as provided for in the design and initial installation, and the opportunity would undoubtedly also be taken to upgrade ancillary equipment at that time.