This paper describes the methodology and results of Monte Carlo neutron transport calculations performed to determine the neutron fluxes required for assessments of radioactive inventories of Magnox power plant. Use has been made of existing validated Monte Carlo models of the plant, which have been extended to include details beyond the pressure vessels to the outer surface of the biological shield.

This paper describes the calculation of the attenuation of neutron dose through the steel pressure vessels of Magnox power plant operated by BNFL, Magnox Generation Division. These data are required for the assessment of the structural integrity of the plant. Detailed three-dimensional neutron transport models of the reactors, pressure vessels and surrounding bioshields have been developed using the AEA Technology Monte-Carlo code MCBEND. These have been used to calculate neutron fluxes through the pressure vessel in both sub-core and side-core regions of pressure vessels, the two bounding extremes of the incident neutron spectra. The effect of postulated cracks in the RPV is shown not to influence the predicted neutron doses. A comparison with neutron fluxes measured from through wall samples removed from a decommissioned reactor show the calculation route to be accurate.