A modified code system for transport of accelerator produced neutrons through ordinary concrete

Abstract

This paper reports the development of a computational system to estimate the production and subsequent transport of neutrons produced from the interactions of 50 and 100 MeV protons, 150 MeV carbons and 500 MeV lithium with thick tantalum targets. The energy-angle distributions of these neutrons are calculated by a hybrid nuclear reaction model code, ALICE91, using Kalbach systematics and modified to estimate neutron yield distributions from thick target interactions. Transmission of these neutrons is subsequently treated with the Anisotropic Source-Flux Iteration Technique (ASFIT) in the framework of a coupled neutron-gamma transport. The estimated transmitted dose through ordinary concrete is compared with the results obtained using empirical formulations as recommended by IAEA.