The Use of a Small Accelerator as a Source of 14-MeV Neutrons for Shielding Studies

It is important in calculating complex shields such as those proposed for the fusion reactors to ascertain that the neutron cross-section data sets used in the calculations are as accurate as possible and that the calculational methods used to transport the neutrons are as reliable as practical. To assure that both these criteria are met, a project at the Oak Ridge National Laboratory (ORNL) is being conducted in which a small accelerator is used to provide 14-MeV neutrons via the T(d, n)4He reaction and an NE-213 detector is used to measure the neutron and gamma-ray pulse-height spectra of the radiations transported through and/or created in very thick laminated shields of stainless steel (type 304) and borated polyethylene. To produce the neutron flux required, the targets are made by depositing about 4 mg/cm2 of TiT onto a 1.27-cm circular area of a O.254-cm thick copper disk. The NE-213 detector is operated in standard, state-of-the-art electronic circuits. A surface-barrier alpha counter and a small NE-213 detector are located permanently at a distance of about 150 cm from the target to monitor the reaction rate in the target. The pulse-height data are unfolded to produce energy spectra by using the computer program FERD. These results are then compared almost immediately with spectra obtained using two-dimensional radiation transport methods incorporating 53-neutron, 21-gamma-ray energy-group cross section data derived from the VITAMIN-C data set (ENDF/B). Laminated stainless-steel and borated polyethylene shields having thicknesses up to 412 g/cm2 have been measured.