The power exhaust concept and an appropriate divertor design are common critical issues for tokamak DEMO
design activities which have been carried out in Europe, Japan, China, Korea and the USA. Conventional divertor
concepts and power exhaust studies for recent DEMO designs (Pfusion = 1 – 2 GW, Rp = 7 – 9 m) are reviewed
from the viewpoints of the plasma physics issues and the divertor engineering design. Radiative cooling is a
common approach for the power fusion scenario. Requirements on the main plasma radiation fraction (frad
main =
Prad
main/Pheat) and the plasma performance constrain the divertor design concept. Different challenges contribute to
optimizing the future DEMO designs: for example, (i) increasing the main plasma radiation fraction for ITER level Psep/Rp designs and simplifying the divertor geometry, and (ii) extending ITER divertor geometry with
increasing divertor radiation (Prad
div) for larger Psep/Rp ≥ 25MWm− 1 designs. Power exhaust simulations with large
Psep = 150 – 300 MW have been performed using integrated divertor codes considering an ITER-based divertor
geometry with longer leg length (1.6 – 1.7 m), as in a common baseline design. Geometry effects (ITER like
geometry or more open one without baffle) on the plasma detachment profile and the required divertor radiation
fraction (frad
div = Prad
div/Psep) were key aspects of these studies. All simulations showed that the divertor plasma
detachment were extended widely across the target plate with a reduction in the peak heat load of qtarget ≤ 10
MWm− 2 for the large frad
div = 0.7 – 0.8, while the peak qtarget location and value were noticeably different in the
partially detached divertor. Simulation results also demonstrated that radial diffusion coefficients of the heat and
particle fluxes were critical parameters for DEMO divertor design, and that effects of plasma drifts on outboard enhanced asymmetry of the heat flux, suggested the need for longer divertor leg to ensure the existence of a
detached divertor operation with qtarget ≤ 10 MWm− 2
.
Integrated design of the water cooled divertor target, cassette body (CB) and cooling pipe routing has been
developed for each DEMO concept, based on the ITER-like tungsten monoblock (W-MB) with Cu-alloy cooling
pipes. Engineering design adequate under higher neutron irradiation condition was required. Therefore, inlet
coolant temperature (Tcool) was increased. In current designs, it still shows a large potential variation between
70 ◦C and 200 ◦C. The influence of thermal softening on the Cu-alloy (CuCrZr) pipe was fostered near the strike point when the high qtarget of ~10 MWm− 2 was studied. Improved technologies for high heat flux components
based on the ITER W-MB unit have been developed for EU-DEMO. Different coolant conditions (low- and high Tcool) were provided for Cu-alloy and reduced activation ferritic martensitic (RAFM) steel heat sink units,
respectively. The high-Tcool coolant was also considered for the CB and supporting structures. Appropriate
conditions for the high-Tcool coolant, i.e. 180 ◦C/ 5 MPa (EU-DEMO) and 290 ◦C/ 15 MPa (JA-DEMO, CFETR and
K-DEMO), will be determined in the future optimizations of the divertor and DEMO design