In recent years, neutron attenuation measurements have demonstrated that neutron absorber panels with discrete boron carbide particles are less effective in neutron absorption than equivalent homogenous absorber panels. The penalty observed in these attenuation measurements with thermal neutrons has been assumed to apply to the criticality analysis of fuel storage cells, without regard to the very significant difference in the neutron spectrum involved. Furthermore, the attenuation measurements are made with a collimated beam of thermalized neutrons impinging perpendicularly on the absorber panel, whereas in fuel storage cells, the neutrons are nearly isotropic and impinge on the panel at all angles. In an effort to understand the phenomena involved and to more accurately determine the effect on criticality safety analyses, calculations were made with the MCNP code, which is capable of explicitly describing discrete B4C particles of various mean size. Results of these calculations, described herein, show that neutron attenuation measurements are only weakly related to criticality analysis and that the penalty seen in attenuation measurements does not apply and should not be directly applied in criticality analyses.
