Comparison of thermal neutron distributions within shield materials obtained by experiments, SN and monte carlo code calculations

Abstract

Benchmark experiments of thermal neutron distributions within the shield materials, graphite pile and pure water, were performed by using 252Cf fission neutrons and gold foil activation detectors, and, to these results, the estimates obtained by using the discrete ordinate code ANISN and the Monte Carlo code MCNP5 with two different cross-section libraries, ENDF/B-VI and the Japanese new version of JENDL-3.3, were compared. The results revealed that the MCNP5 calculations with the two libraries closely agree with the experiments and that there are slight differences between the MCNP5 and the ANISN calculations. The differences are caused mainly by the overestimation of the thermal neutron absorption cross sections constructed in NJOY99. The ANISN calculations with the modified absorption cross sections reproduced the results of the MCNP5 fairly well.