The design progress in a compact low aspect ratio (low A) DEMO reactor, ‘SlimCS’, and its design issues are reported. The design study focused mainly on the torus configuration including the blanket, divertor, materials and maintenance scheme. For continuity with the Japanese ITER-TBM, the blanket is based on a water-cooled solid breeder blanket. For vertical stability of the elongated plasma and high beta access, the blanket is segmented into replaceable and permanent blankets and a sector-wide conducting shell is arranged inbetween these blankets. A numerical calculation indicates that fuel self-sufficiency can be satisfied when the blanket interior is ideally fabricated. An allowable heat load to the divertor plate should be 8 MW m−2 or lower, which can be a critical constraint for determining a handling power of DEMO.

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Power exhaust to the divertor and the conceptual design have been investigated for a steady-state DEMO in Japan with 1.5 GW-level fusion power and the major radius of 8.5 m, where the plasma parameters were revised appropriate for the impurity seeding scenario. A system code survey for the Ar impurity seeding suggested the volume-averaged density, impurity concentration and exhaust power from the main plasma of   =  205–285 MW. The divertor plasma simulation (SONIC) was performed in the divertor leg length of 1.6 m with the fixed exhaust power to the edge of   =  250 MW and the total radiation fraction at the edge, SOL and divertor (  =  0.8), as a first step to investigate appropriate design of the divertor size and geometry. At the outer target, partial detachment was produced near the strike-point, and the peak heat load () at the attached region was reduced to ~5 MW m−2 with appropriate fuel and impurity puff rates. At the inner divertor target, full detachment of ion flux was produced and the peak was less than 10 MW m−2 mostly due to the surface-recombination. These results showed a power exhaust scenario and the divertor design concept. An integrated design of the water-cooling heat sink for the long leg divertor was proposed. Cu-ally (CuCrZr) cooling pipe was applicable as the heat sink to handle the high heat flux near the strike-point, where displacements per atom rate was estimated to be 0.5–1.5 per year by neutronics calculation. An arrangement of the coolant rooting for Cu-alloy and Reduced Activation Ferritic Martensitic (RAFM) steel (F82H) pipes in a divertor cassette was investigated, and the heat transport analysis of the W-monoblock and Cu-alloy pipe under the peak of 10 MWm−2 and nuclear heating was performed. The maximum temperatures on the W-surface and Cu-alloy pipe were 1021 and 331 °C. Heat flux of 16 MW m−2 was distributed in the major part of the coolant pipe. These results were acceptable for the plasma facing and structural materials.

A new system proposed for the generation of radioisotopes with accelerator neutrons by deuterons (GRAND) is described by mainly discussing the production of 99 Mo used for nuclear medicine diagnosis. A prototype facility of this system consists of a cyclotron to produce intense accelerator neutrons from the nat C(d,n) reaction with 40 MeV 2 mA deuteron beams, and a sublimation system to separate 99m Tc from an irradiated 100 MoO 3 sample. About 8.1 TBq/week of 99 Mo is produced by repeating irradiation on an enriched 100 Mo sample (251 g) with accelerator neutrons for two days three times. It meets about 10% of the 99 Mo demand in Japan. The characteristic feature of the system lies in its capability to reliably produce a wide variety of high-quality, carrier-free, carrier-added radioisotopes with a minimum level of radioactive waste without using uranium. The system is compact in size, and easy to operate; therefore it could be used worldwide to produce radioisotopes for medical, research, and industrial applications.

At JAEA, a test blanket module (TBM) with a water-cooled solid breeder is being developed. This paper presents recent achievements of research activities for the TBM, particularly addressing the pebble bed of the tritium breeder materials and tritium behaviour. For the breeder material, the chemical stability of Li2TiO3 was improved using Li2O additives. To analyse the pebble bed behaviour, thermomechanical properties of the Li2TiO3 pebble bed were assessed experimentally. To verify the pebble bed's nuclear properties, the activation foil method was proposed and a preliminary experiment was conducted. To reduce the tritium permeation, the chemical densified coating method was developed and the coating was attached to F82H steel. For tritium behaviour, the tritium recovery system was modified in consideration of the design change of the TBM.