A Two-Dimensional Discrete-Ordinates Analysis of the CANDU 9 End-Shield Penetrations

Abstract

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 ENDFA3-IV data, but with significant updates from new ENDF/B-VI data for many materials common to CANDU applications. Three different fuel-channel configurations were analyzed: i. the reference design of fuel-channel components; ii. a "new" fuel-channel design, where the rolled joint area of the end-fittings was longer than the reference design by 102 mm; a corresponding 102-mm length was removed from the body of the shield plug. and the liner tube was shortened by 102 mm; and iii. same as case (ii) above, but the steel of the shield plug was replaced by heavy-water coolant, 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. The fluxes and dose rates were compared for these three configurations to assess the effects on dose rates in the fuelling-machine vaults. Twenty-four hours after reactor shutdown, the dose rates outside the end-fittings, that is, in the fuelling machine vaults, have contributions from i. decay gamma rays from the 60~aoct ivation product formed from cobalt impurities in the stainless steel of the fuelling-machine side tube sheet and other end-shield components; ii. fission-product decay gamma rays from irradiated fuel in the reactor core; iii. prompt gamma rays from the low fission power of the core (because of fissions from delayed neutrons and photoneutrons inside the core); and iv. activation gamma rays from corrosion and fission products deposited in the end-fittings and feeder pipes. The neutron flux distribution through the end shield during normal operation was used to calculate the resulting 60Co activity of the fuelling-machne tube sheet. the shield plug, the end-fittings, the closure plug, etc. This 60Co activity was used as a source distribution in a separate DORT analysis to calculate the 60Co gamma dose rates at the outer surface of the end-fittings. The fission-product decay gamma spectrum from an ORIGEN-S calculation was incorporated as a fixed source in the fuel for the DORT transport calculation to determine the fission-product decay gamma dose rates at 24 h after shutdown. The dose rates that are due to the low fission power of the core, i-e., neutronic power of the core (-2.0 x 10^-6 FP at 24 h after shutdown), were also calculated by prorating the full-power dose rates with this neutronic power fraction. The total dose rate from these three sources, i.e., sources (i) to (iii), were summed and compared among the three proposed fuel-channel configurations. The shutdown dose rates outside the end-fittings from source (iv) were not evaluated in this paper. However, the dose rates from this source have been measured to be as high as 2 to 5 mSvh at CANDU 6 plants.