Methods, Methodologies and Formulas for Simplified Neutronics Analyses of Fusion Reactors.

Abstract

The six articles on the simplified neutronics analyses of a fusion reactor published in other journals by the author in 1991-1994 are reviewed. Although the most results were derived under the limitation of materials and models considered and the requried accuracy for the designs, the description may be permissible from the fusion reactor design-oriented point of view. The four simplified blanket engineering methods which have been developed to evaluate the tritium generation in boronized shields, the lithium and beryllium burn-up in blankets and the bismuth and the polonium build-ups in LiPb-bearing blankets, are briefly summarized. The sets of differential equations for the densities of relevant isotopes have been solved using simplified equations. The four shielding methodologies which have been developed to evaluate the fast (E>0.1 MeV) neutron flux, the dose rate, the displacement damage rate and the helium generation rate in in-vessel components of a fusion reactor are also introduced briefly. The abovelisted neutronics responses with the distance from the first wall are described by the exponentially decreasing functions. Attenuation lengths λ required to characterize the attenuation along the thickness of in-vessel composition are obtained for different compositions. The methodology to estimate the total nuclear heating in Toroidal Field (TF) coils is discussed as well. The simplified analytical formulas are also reviewed for 14-MeV and fast neutron fluxes at the exit from a radiation shield with a straight-through diagnostic channel at some distance from the channel orifice. The formulas are derived by fitting the results of the two-dimensional transport calculations. The Simon-Clifford and the Shin formulas are employed in the proposed streaming formulas to treat the 14-MeV and the fast neutron fluxes, respectively, inside the diagnostic channel.