Photon and Neutron Spectra From Spent Fuel Bundles for CANDU 6 Reactors

Abstract

This paper describes a methodology for calculating photon (gamma) dose rate spectra from fission-product decay gammas and neutron dose rate spectra generated from a 37- element natural uranium spent-fuel bundle in air using the MCNP-4C and ORIGEN-S codes. The 37-element fuel bundles are used in all CANDU stations, except Pickering A and B stations, which use 28-element natural uranium fuel bundles. The gamma dose rate from spent fuel is very large at discharge and decays with time, with the gamma spectra becoming softer as short-lived, high-energy gamma-emitting fission products decay quickly with time. The neutrons are also produced from spontaneous fission and by (α,n) reactions. The neutron dose rate is insignificant with respect to the gamma dose rate, but it decays very slowly with time because of the extremely long half-lives of actinides. The spectra of gamma and neutron radiation released from the spent fuel bundle are hardened because of the absorption of low-energy gammas in the fuel bundle, whereas the spectra of neutrons released are softened because of scatter events in the fuel bundle. The neutron spectra are much harder and few thermal neutrons are generated if the spent fuel is in air. Gamma detection instruments such as bundle counters (e.g., one for each spent-fuel port at a CANDU 6 reactor) count the movement of spent fuel bundles between the reactor and the spent-fuel storage bays. The bundle counters are located close to the spent-fuel ports in the spent-fuel discharge room. Using Monte Carlo codes such as MCNP, the method that is used to calculate the dose rates from spent-fuel bundles can be expanded and applied to shielding modifications for the bundle counters, if desired, or it could be applied to designing a different system that detects neutrons emitted from spent fuel in air.