CÁLCULO DE FLUJO NEUTRÓNICO Y FLUENCIA EN LA ENVOLVENTE DEL NÚCLEO Y LA VASIJA DE UN REACTOR BWR

Abstract

One of the main objectives related to the safety of any nuclear power plant, including the Laguna Verde Nuclear Power Plant (CNLV) is to ensure the structural integrity of reactor pressure vessel. To identify and quantify the damage caused by neutron irradiation in the vessel of any nuclear reactor, it is necessary to know both the neutron flux and the neutron fluence that the vessel has been receiving during its operating lifetime, and that the damage observed by mechanical testing are products of microstructural effects induced by neutron irradiation; therefore, it is important the study and prediction of the neutron flux in order to have a better understanding of the damage that these materials are receiving. The calculation here described uses the DORT code, which solves the neutron transport equation in discrete ordinates in two dimensions (x-y, r- and r-z), according to a regulatory guide, it should make an approximation of the neutron flux in three dimensions by the so called Synthesis Method. It is called in that way because it achieves a representation of 3D neutron flux combining or summarizing the fluxes calculated by DORT r-, r-z and r. This work presents the application of Synthesis Method, according to Regulatory Guide 1190 [1], to determine the 3D fluxes in internal BWR reactor using three different spatial meshes. The results for the neutron flux and fluence, using three different meshes in the directions r, θ and z were compared with results reported in the literature obtaining a difference not larger than 9.61%, neutron flux reached its maximum, 1.58E+12 n/cm2s, at a height H4 (239.07 cm) and angle 32.236o in the core shroud and 4.00E+09 n/cm2s at a height H4 and angle 35.27o in the inner wall of the reactor vessel, positions that are consistent to within ±10% over the ones reported in the literature.