Nuclear Analysis and Optimization of the Molten-Salt Fusion Hybrid Reactor

摘要

An improved method of studying the neutronic characteristics of fusion hybrid reactor blankets has been developed. Two major improvements over previous analysis methods have been accomplished. The first of these improvements is the introduction of one-dimensional, homogenized-region blanket neutronic models in which resonance and spatial self-shielding effects are treated explicitly. The second improvement involves the application of an iterative gradient-ascent based optimization scheme. In this method, key blanket dimensions and concentrations are automatically varied in a search for a configuration which maximizes neutronic performance. The specific fusion hybrid blanket design to which these new methods of analysis are applied in an evolution of the U-233 producing molten-salt-in-tubes concept studied by Lawrence Livermore National Laboratory (LLNL). Optimistic analysis techniques initially predicted the fissile fuel production capacity of this blanket to be 6400 kg of U-233 per year when driven by a 3000 MW tandem mirror fusion driver. The improved and more realistic analysis techniques employed in this study predict that an optimized molten-salt blanket design will produce over 6700 kg of U-233 per year when driven by the same tandem mirror device. Finally, the techniques and data base developed in this study have been designed to be easily extended to the task of performing future, more extensive analysis. Such an analysis might involve the minimization of fuel costs in an entire fusion hybrid reactor complex. 20 refs., 10 figs., 14 tabs. (ERA citation 13:035435)