Shinsuke Tokunaga

A simulation study of large power handling in the divertor for a Demo reactor

Power exhaust for a 3?GW class fusion reactor with an ITER-sized plasma was investigated by enhancing the radiation loss from seeding impurity. The impurity transport and plasma detachment were simulated under the Demo divertor condition using an integrated divertor code SONIC, in which the impurity Monte-Carlo code, IMPMC, can handle most kinetic effects on the impurity ions in the original formula. The simulation results of impurity species from low Z (neon) to high Z (krypton) and divertor length with a plasma exhausted power of 500?MW and radiation loss of 460?MW, and a fixed core?edge boundary of 7???1019?m?3 were investigated at the first stage for the Demo divertor operation scenario and the geometry design. Results for the different seeding impurities showed that the total heat load, including the plasma transport and radiation , was reduced from 15?16?MW?m?2 (Ne and Ar) to 11?MW?m?2 for the higher Z (Kr), and extended over a wide area accompanied by increasing impurity recycling. The geometry effect of the long-leg divertor showed that full detachment was obtained, and the peak qtarget value was decreased to 12?MW?m?2, where neutral heat load became comparable to and due to smaller flux expansion. Fuel dilution was reduced but was still at a high level. These results showed that a divertor design with a long leg with higher Z seeding such as Ar and Kr is not fulfilled, but will be appropriate to obtain the divertor scenario for the Demo divertor. Finally, influences of ? and D? enhancement were seen significantly in the divertor, i.e. the radiation and density profiles became wider, leading to full detachment. Both qtarget near the separatrix and Te at the outer flux surfaces were decreased to a level for the conventional technology design. On the other hand, the problem of fuel dilution became worse. Extrapolation of the plasma transport coefficients to ITER and Demo, where density and temperature will be higher than ITER and edge-localized modes are mitigated, is a key issue for the divertor design.

Management Strategy for Radioactive Waste in the Fusion DEMO Reactor

We have considered a strategy for reducing the radioactive waste generated by the replacement of in-vessel components, such as blanket segments and divertor cassettes, for the fusion DEMO reactor. In the basic case, the main parameters of the DEMO reactor are a major radius of 8.2 m and a fusion power of 1.35 GW. Blanket segments and divertor cassettes should be replaced independently, as their lifetimes differ. A blanket segment comprises several blanket modules mounted to a back-plate. The total weight of an in-vessel component is estimated to be about 6,648 ton (1,575, 3,777, 372, and 924 ton of blanket module, back-plate, conducting shell, and divertor cassette, respectively). The lifetimes of a blanket segment and a divertor cassette are assumed to be 2.2 years and 0.6 years, respectively, and 52,487 tons of waste is generated over a plant life of 20 years. Therefore, there is a concern that the contamination-control area for radioactive waste may need to increase due to the amount of waste generated from every replacement. This paper proposes a management scenario to reduce radioactive waste. When feasible and relevant, back-plates of blanket segment and divertor cassette bodies (628 ton) should be reused. Using the three-dimensional neutron transportation code MCNP, the displacement per atom (DPA) of the SUS316LN back-plates is 0.2 DPA/year and that of the F82H cassette bodies is 0.6 DPA/year. Therefore, the reuse of back-plates and cassette bodies would be possible if re-welding points are arranged under neutron shielding. We found that radioactive waste could be reduced to 20 % when tritium breeding materials are recycled. Finally, we propose a design for the DEMO building that uses a hot cell and temporary storage.