CREST is a conceptual tokamak reactor design with high β plasma, high thermal efficiency, competitive cost and water-cooled ferritic steel components. Some of its parameters are similar to those of the ITER advanced mode plasma. In this manuscript, the specific issues and analysis on damage to TF coils of CREST were carried out based on the three-dimensional model of the CREST with the widely used code MCNP/4C and the IAEA latest released FENDL/2.1 data library. Damage to some specific regions of the TF coils near large openings and at the inboard mid-plane are calculated and analyzed. Parameters such as the distributions of nuclear heat density, fast neutron flux, dose rate to the epoxy insulator, and peak displacement dose to Cu conductor for the TF coil near these regions were calculated and analyzed. The shield thicknesses at these regions are optimized.
ASJC Scopus subject areas
- Nuclear and High Energy Physics
- General Materials Science
- Nuclear Energy and Engineering