Kazuo Hoshino

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.

Critical design factors for sector transport maintenance in DEMO

This paper mainly focuses on a sector transport maintenance scheme from the aspects of high plant availability. In this study, three different maintenance schemes are considered based on (1) the number of maintenance ports and (2) the insertion direction. The design study clarifies critical design factors and key engineering issues on the maintenance scheme: (1) how to support an enormous overturning force of the toroidal field coils in the large open port for sector transport and (2) define the transferring mechanism of sectors in the vacuum vessel. On reviewing these assessment factors, the sector transport using a limited number of horizontal maintenance ports is found to be a more realistic maintenance scheme. In addition, evaluating maintenance scenarios under high decay heat is proposed for the first time. The key design factors are the cool-down time in the reactor and the cooling method in the maintenance scheme to keep components under operational temperature. Based on one-dimensional heat conduction analysis, after one month cool-down time, each sector of SlimCS could be transported to the hot cell facility by gas cooling.

Estimation of Tritium Permeation Rate to Cooling Water in Fusion DEMO Condition

Abstract The approximate estimation of tritium permeation rate under the acceptable assumption from a safety point of view is surely useful to progress the design activities for a fusion DEMO reactor. Tritium permeation rates in the blanket and the divertor were estimated by the simplified evaluation model under the recent DEMO conditions in the water-cooled blanket with solid breeder as a first step. Plasma driven permeation rates in tungsten wall were calculated by applying Doyle & Brice model and gas driven permeation rates in F82H were calculated for hydrogen-tritium two-component system. In the representative recent DEMO condition, the following tritium permeation rates were obtained, 1.8 g/day in the blanket first wall, 2.3 g/day in the blanket tritium breeding region and 1.6 g/day in the divertor. Total tritium permeation rate into the cooling water was estimated to be 5.7 g/day.

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.

Self-consistent treatment of the sheath boundary conditions by introducing anisotropic ion temperatures and virtual divertor model

One-dimensional SOL-divertor plasma fluid simulation code which considers anisotropy of ion temperature has been developed so as to deal with sheath theory self-consistently. In our fluid modeling, explicit use of boundary condition for Mach number M at divertor plate, e.g., M = 1 , becomes unnecessary. In order to deal with the Bohm condition and the sheath heat transmission factors at divertor plate self-consistently, we introduced a virtual divertor (VD) model which sets an artificial region beyond divertor plates and artificial sinks for particle, momentum and energy there to model the effects of the sheath region in front of the divertor plate. Validity of our fluid model with VD model is confirmed by showing that simulation results agree well with those from a kinetic code regarding the Bohm condition, ion temperature anisotropy and supersonic flow. We also show that the strength of artificial sinks in VD region does not affect profiles in plasma region at least in the steady state and that sheath heat transmission factors can be adjusted to theoretical values by VD model. Validity of viscous flux is also investigated.