Validation of neutron propagation calculations using the DORT and DOTSYN codes and the special dosimetry benchmark experiment at the French St. Laurent reactor Original Research Article

The St. Laurent B1 unit is a 900 MWe PWR that was loaded with MOX fuel at the beginning of its 8th cycle. In order to provide the means of validating the calculation of neutron propagation in a power reactor, the French national electric utility EDF installed numerous neutron dosimeters in special surveillance capsules and in three ex-vessel chains in the reactor pit. These dosimeters were irradiated for one cycle and then removed, providing measured reaction rates at locations both inside and outside the reactor vessel. This paper describes the analysis performed by Framatome using the DORT + DOTSYN 3-D synthesis modules to obtain calculated dosimeter responses at each instrument location. Several 2-D and 1-D models were developed, paying particular attention to the geometrical representation of the reactor: the variable mesh capability of the DORT code was used and a special variable mesh generator was written for it, along with a graphical interface for visual checks. A 100 energy group transport cross-section library derived from the master ENDF/B-IV and ENDF/B-VI libraries was used. The results show very good agreement between the calculated and measured fast neutron reaction rates at the in-vessel locations, with an average calculated (C) to measured (M) ratio, CM, of 0.94 and a small dispersion in the CM distribution. The CM values at the ex-vessel dosimeter locations averaged 1.12, which indicates that the transport cross-section library slightly underestimates the attenuation of the fast neutrons passing through the reactor vessel. These results are comparable to those from a similar analysis carried out independently based on the same St. Laurent dosimetry benchmark, using a Monte Carlo code which is reported elsewhere.