A two-dimensional discrete-ordinates analysis of the CANDU 9 end-shield penetrations

The two-dimensional discrete-ordinates radiation transport code DORT has been used to calculate neutron and gamma fluxes and corresponding dose rates through the end-shield outlet penetration of a central fuel channel in a CANDU 9 480/NU reactor. The 67 group coupled neutron-gamma library cross sections used in the analysis are based on primary ENDF/B-IV data, but with significant updates from new ENDF/B-VI data for many materials common to CANDU applications. Calculations were made for normal steady -state full power operation and after reactor shutdown. The neutron and gamma dose rates were calculated throughout the end shield including the plane at the outer surface of the fuelling machine tube sheet and outside the end-fittings during reactor operation. (DM)