A Validation of Neutron Fluxes Through the CANDU® 6 End Shield

Abstract

Worldwide in the radiation shielding community, the two-dimensional (2-D) discrete-ordinates code DORT [1] remains one of the codes used to calculate the transport of neutrons and photons through thick shields. In the Canadian nuclear industry, DORT continues to be used to calculate the neutron-flux, photon-flux, heating, activation, dose-rate, and damage profiles across the different types of CANDU reactor end shields and radial shields. Including its predecessor DOT-IV that evolved to what is now DORT, this 2-D discrete-ordinates code has been used in the design of many CANDU reactors, i.e., Pickering Nuclear Generating Stations (NGS), Bruce NGS, Darlington NGS, CANDU 6 stations, and the CANDU 3, CANDU 9, and the new ACR-1000 designs. Hitherto the validation of DORT applications for CANDU plants relied on limited CANDU station measurements and foreign benchmark experiments. The DORT code is often used to calculate particle fluxes across the reactor shields. Validation exercises that have been performed to date have yet to address the application of the code to calculating deep penetration through the primary shields of a CANDU reactor. This paper provides some validation results of neutron fluxes through a CANDU 6 end shield to bridge the gap in the validation studies.