[en] Conceptual design of blanket and divertor for a force free helical reactor (FFHR) is presented. The DEMO relevant FFHR is heliotron type helical reactor having super conducting helical and poloidal coils based on the Large Helical Device (LHD) which is now under construction in NIFS. The main features of FFHR are (1) force free configuration of helical coils to reduce the total structural mass required for magnetic energy, (2) plasma operation with no current disruptions, (3) steady-state operation with a small fraction of the recirculating power. For the goal of a self-ignited D-T reactor of 3 GW thermal output, the design parameters for FFHR are investigated under the LHD scaling for energy confinement and density limit. In particular, to satisfy the reactor lifetime of 30 years, the engineering issues in FFHR are discussed by focusing on the following aspects; neutron wall loading as low as about 1.5 MW/m², tritium breeding with nuclear heating, cooling with liquid coolant and operation temperature in the blanket, selection of structural materials with low activation, radiation shielding to achieve more that 5 order reduction at super conducting coils, configuration of high heat flux components in divertor regions, helium ash removal with the efficiency around 30%, and safety

[en] The design of blankets for NET is reviewed. Three categories of blankets are discussed: non breeding shield blankets for the physics phase, tritium producing driver blankets to operate the NET machine for extended periods of time in the technology test phase, and DEMO relevant blankets, which shall demonstrate their performance as test modules in NET. This separation of functions has only recently been determined as have been the definitions for DEMO load conditions. The evolution of the different blanket concepts is described with emphasis on design characteristics and supporting RD. particular ITER aspects are complemented. (author). 36 refs.; 16 figs.; 3 tabs

[en] This paper recalls the main features of the TAURO blanket, a self-cooled Pb-17Li concept using SiC/SiC composites as structural material, developed for FPR. The objective of this design activity is to compare the characteristics of present-day industrial SiC-SiC composites with those required for a fusion power reactor blanket (FPR) and to evaluate the main needs of further R and D. The performed analyses indicated that the TAURO blanket would need the availability of SiC/SiC composites approximately 10 mm thick with a thermal conductivity through the thickness of approximately 15 Wm^-1K^-1 at 1000 C and a low electrical conductivity. A preliminary MHD analysis has indicated that the electrical conductivity should not be greater than 500 Ω^-1m^-1. Irradiation effects should be included in these figures. Under these conditions, the calculated pressure drop due to the high Pb-17Li velocity (approximately 1 m s^-1) is much lower then 0.1 MPa. The characteristics and data base of the recently developed 3D-SiC/SiC composite, Cerasep ^trademark N3-1, are reported and discussed in relation to the identified blanket design requirements. The progress on joining techniques is briefly reported. For the time being, the best results have been obtained using Si-based brazing systems initially developed for SiC ceramics and whose major issue is the higher porosity of the SiC/SiC composites. (orig.)

[en] Short communication

[en] China focuses on the development of the liquid LiPb blanket due to its attractive performances. A series of LiPb breeder blanket concepts were proposed for different applications of fusion in China. This paper gives an overview of the LiPb breeder blanket concepts and relevant R and D progress in China.

[en] Published in summary form only

[en] Short communication

[en] The testing stages in NET for the performance assessment of the various breeding blanket concepts developed at the present time in Europe for DEMO (LiPb and ceramic blankets) and the requirements upon NET to perform these tests are reviewed. Typical locations available in NET for blanket testing are the central outboard segments and the horizontal ports of in-vessel sectors. These test positions will be connectable with external test loops. The number of test loops (helium, water, liquid metal) will be such that each major class of blankets can be tested in NET. The test positions, the boundary conditions and the external test loops are identified and the requirements for test blankets are summarized (author). 6

[en] Published in summary form only

[en] A preliminary design of a 5500 MW/sub th/ modular stellarator power reactor, UWTOR-M, is presented. Discrete twisted coils are used in an l.3 configuration with maintainability as a prime consideration. The natural stellarator divertor is used for impurity control in conjunction with innovative high performance divertor targets. A unique blanket design is proposed which minimizes the overall tritium inventory in the reactor. Finally, a scheme for maintaining the first wall/blanket and other reactor components is discussed. 17 refs