VALIDATION OF NEUTRON PROPAGATION CALCULATIONS USING THE DORT AND DOTSYN CODES AND THE SPECIAL DOSIMETRY BENCHMARK EXPERIMENT AT THE FRENCHST LAURENT REACTOR

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 v alidating the calculation of neutron propagation in a power reactor, t he French national electric utility EDF installed numerous neutron dos imeters in special surveillance capsules and in three ex-vessel chains in the reactor pit. These dosimeters were irradiated for one cycle an d then removed, providing measured reaction rates at locations both in side 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 c apability of the DORT code was used and a special variable mesh genera tor was written for it, along with a graphical interface for visual ch ecks. A 100 energy group transport cross-section library derived from the master ENDF/B-IV and ENDF/B-VI libraries was used. The results sho w 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, C/M, of 0.94 and a small dispersion in the C/M distribution. The C/M values at the ex-vessel dosimeter locations averaged 1.12, which indicates that the transport cross-section libra ry slightly underestimates the attenuation of the fast neutrons passin g through the reactor vessel. These results are comparable to those fr om a similar analysis carried out independently based on the same St. Laurent dosimetry benchmark, using a Monte Carlo code which is reporte d elsewhere. (C) 1997 Elsevier Science S.A.