Measurements of the Neutron Spectra from Materials Used in Fusion Reactors and Calculations Using the ENDF/B-III and -IV Neutron Libraries

Proposed fusion reactor blanket designs bring into focus a large number of problems dealing with the interaction of 14-MeV neutrons with different materials. Carbon, oxygen, aluminum, titanium, and iron are among the materials used in the blanket. To have confidence in fusion reactor blanket calculations, a necessary prerequisite is that the transport code correctly describes the interaction of 14-MeV neutrons with the materials of the blanket. Spherical assemblies of the above materials ranging from 1 to 5 mean-free-paths in thickness have been bombarded with a centered nominal 14-MeV neutron source. The emitted neutron energy spectra were measured using time-of-flight techniques (3-nsec full-width-at-half-maximum system resolution) in a geometry where the flight path (7 to 10 m) is long compared to the dimensions of the spherical targets. The spectra have been calculated with the Monte Carlo neutron transport code TART using the ENDF/B-III and -IV neutron libraries and compared with measurements.