Calculations of Anisotropy Factors and Dose Equivalents for Unmoderated 252Cf Sources

The effects upon dose equivalent quantities of encapsulating 252 Cf fission neutron sources were investigated with the Monte Carlo code MCNP. The SR-Cf-100 encapsulation was modelled and the resulting anisotropy and dose equivalent quantities compared with reference values. A SR-Cf-100 source was also modelled inside an aluminium capsule to simulate the one in use at the Pacific Northwest Laboratory. These encapsulations, although resulting in noticeable anisotropy effects, do not require a correction to be made to existing reference dosimetric quantities based on a Maxwellian distribution source spectrum