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内容記述 |
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) were reviewed from the viewpoints of the plasma physics study and the divertor engineering design. Radiative cooling is a common approach for the power exhaust scenario. Requirements of the main plasma radiation fraction (fradmain = Pradmain/Pheat) and the plasma performance determine the divertor design concept. Different challenges, 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 (Praddiv) for larger Psep/Rp ? 25MWm-1 designs, will contribute to optimize the future DEMO designs. ITER-based divertor geometry with longer leg length (1.6-1.7 m) is a common baseline design, and power exhaust simulations with larger Psep =150-300 MW have been performed using integrated divertor codes. Geometry effects (ITER like one or more open one without baffle) on plasma detachment profile and the required divertor radiation fraction (fraddiv = Praddiv/Psep) were important key aspects of the studies. All simulations showed that the divertor plasma detachment were extended widely and that reduction in the peak qtarget ? 10 MWm-2 for the large fraddiv = 0.7-0.8, while the peak qtarget location and value were different in the partially detached divertor. Simulation results also demonstrated that diffusion coefficients of the heat and particle fluxes were critical parameters for DEMO divertor design, and that effects of various drifts on outboard-enhanced asymmetry of the heat flux, might imply the need for longer divertor leg to insure the detached divertor operation window.
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) and Cu-alloy pipe. Engineering design adequate under higher neutron irradiation condition was required. Therefore, inlet coolant temperature (Tcool) was increased and has so far large 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-level was exposed. Improved technologies of high heat flux components based on the ITER W-MB unit have been developed for EU-DEMO. Different coolant conditions (lower- and higher-Tcool) were provided for the Cu-alloy and reduced activation ferritic martensitic (RAFM) steel heat sink units, respectively, and the higher-Tcool coolant was also provided for the CB and supporting structures. Appropriate conditions for the higher-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 optimization of the divertor and DEMO design. |
Asakura-NME2023-DEMOdivertorReview.pdf (11.0 MB)
