In the current design of the ITER divertor32-34 for the start-up phase, tungsten has been selected as armor for the divertor dome and the upper part of the divertor vertical targets. In addition, due to exces­sive co-deposition of tritium in CFC raising regu­latory concerns related to tritium inventory limits, a full tungsten divertor will be installed before the D-T phase of operation.32

The PFC design for ITER consists of bulk W bonded to an actively pressurized water-cooled Cu alloy heat sink. Here W has no primary structural function. However, due to the operating conditions listed in Table 1, the PFMs face large mechanical loads particularly at the interface to the heat sink material during cyclic steady state heat loads (see Section 4.17.4.2) and at the plasma-loaded surface during transient thermal events (see Section 4.17.4.1). Furthermore, the material response to these loads is influenced by the material damage or degradation due to neutron irradiation (see Table 1, Sections

4.17.4.3.3 and 4.17.4.3.4).

Along with thermally induced loads, the interac­tion of the PFM with the plasma, that is, the hydro­gen isotopes D and T as well as the fusion product He, is of importance (see Section 4.17.4.4) because they have an influence on material erosion and near­surface material degradation.

The further development of the ITER design led to four conceptual designs for the DEMO divertor.25,35 These designs include either water (inlet 140 °C/outlet 170 °C) or, due to the higher achievable efficiency, more probably He-cooling (inlet 540 °C/outlet 700 °C). In all cases bulk W is foreseen as the armor material that will have to face peak steady state heat loads of 15 MW m-2 in case of the water-cooled design and 10MWm-2 for the He-cooled designs. In contrast to ITER, off-normal events such as disruptions have to be avoided completely and transient thermal events during nor­mal operation, for example, ELMs, have to be miti­gated below the damage threshold of the material (see Section 4.17.4.1). This may be particularly important considering the expected neutron damage that will amount up to 40-60 dpa during the planned operation of the fusion reactor35 leading to a signifi­cant amount of transmutation products.36 However, the main limiting factor is expected to be the materi­al’s erosion leading to a maximum lifetime of 2 years for the divertor armor.35

In comparison to tokamaks, calculations for a device such as ARIES-CS predict steady state heat loads between 5 and 18MWm-2.30,37 Similar to DEMO, a He-cooled W divertor is anticipated with a maximum heat removal capability of 10 MW m~2. The design limits for neutron irradiation at the shield of ARIES-CS are up to 200 dpa at 40 fpy (full power years).38 The component lifetime limits are similarly dictated by the material’s expected erosion.

Finally, tungsten or more specifically tungsten coatings find their application also in the dry wall concept for inertial fusion devices, for example, the National Ignition Facility (NIF). In future, inertial confinement devices, thermal loads will occur only in the form of transient thermal loads (P = 0.1 MJ m~2, t = 1-3 ms, f= 5-15 Hz, Tbase > 500 °C).31 These are similar to those expected during ELMs and almost identical to those occurring in an X-ray anode39 and, therefore, affect a thin surface layer only.