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Activation analyses play a vital role in nuclear reactor design. Activation analyses, along with nuclear analyses, provide important information for nuclear safety and maintenance strategies. Activation analyses also help in the selection of materials for a nuclear reactor, by providing the radioactivity and dose rate levels after irradiation. This information is important to help define maintenance activity for different parts of the reactor, and to plan decommissioning and radioactive waste disposal strategies. The study of activation analyses of candidate structural materials for near-term fusion reactors or ITER is equally essential, due to the presence of a highenergy neutron environment which makes decisive demands on material selection. This study comprises two parts; in the first part the activation characteristics, in a fusion radiation environment, of several elements which are widely present in structural materials, are studied. It reveals that the presence of a few specific elements in a material can diminish its feasibility for use in the nuclear environment. The second part of the study concentrates on activation analyses of candidate structural materials for near-term fusion reactors and their comparison in fusion radiation conditions. The structural materials selected for this study, i.e. India-specific Reduced Activation Ferritic−Martensitic steel (IN-RAFMS), P91-grade steel, stainless steel 316LN ITER-grade (SS-316LN-IG), stainless steel 316L and stainless steel 304, are candidates for use in ITER either in vessel components or test blanket systems. Tungsten is also included in this study because of its use for ITER plasma-facing components. The study is carried out using the reference parameters of the ITER fusion reactor. The activation characteristics of the materials are assessed considering the irradiation at an ITER equatorial port. The presence of elements like Nb, Mo, Co and Ta in a structural material enhance the activity level as well as the dose level, which has an impact on design considerations. IN-RAFMS was shown to be a more effective low-activation material than SS-316LN-IG.

Summary Helium cooled dual breeder (HCDB) blanket concept is designed for Indian DEMO fusion reactor and it is made of two tritium breeder materials PbLi and Li2TiO3. It has helium as a coolant and the India specific RAFMS as a structural material. High‐pressure helium first cools the plasma facing first‐wall and after that, it will extract heat from PbLi and ceramic breeder. Since PbLi is not used as a coolant, it therefore circulates with a low flow rate. It will overcome the corrosion and MHD issues associated with high temperature and high flow rate of PbLi. The idea behind the concept is to make a design which can be made using the existing blanket materials, extract high‐grade heat from the reactor and also enhance the availability. In HCDB blanket the role of neutron multiplier is done by PbLi, an alternative to the beryllium and it eliminates the issue associated with high toxic beryllium handling. It can be a potential tritium breeding blanket concept along with lead lithium cooled ceramic breeder (LLCB) and helium cooled ceramic breeder (HCCB) for near term Indian demonstration nuclear fusion power plant. In order to realize the HCDB conceptual design, preliminary estimations of tritium production, nuclear heat density have been carried out. The thermal behaviors of the HCDB blanket in Indian DEMO conditions have been also assessed and reported here. The assessment establishes the proof of HCDB blanket concept and supports it to be a good alternate blanket candidate for the Indian DEMO. The paper describes the HCDB concept along with analysis to verify the tritium self‐sufficiency and materials temperature limits.