A study on the radiation shield for toroidal field (TF) coils of a tokamak type fusion device is reported. The study was performed to provide the design data base for the radiation shielding analysis for TF coils which can be commonly used for other systems, and to produce some universal recommendation about the neutron flux attenuation in the shield of a fusion reactor. Some simple estimation procedure instead of difficult and expensive neutron calculation can be carried out in this case on the basis of the fundamental knowledge on neutron behavior. The present studies are composed of the fundamentals required for shield estimation, the analysis of shield effectiveness, the analysis of the shielding performance of blankets, the analysis of radiation permeating through the inhomogeneous blanket and shield of a fusion reactor, and the analysis of the TF coil shield of the International Thermonuclear Experimental Reactor (ITER) on the basis of the results of the ITER conceptual design activities for three years. The methodological recommendation was developed for the ANISN and the DOT3 codes. (K.I.).
Zimin, Sergei 
- Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment