Nuclear Technology / Volume 184 / Number 1 / October 2013 / Pages 29-41
Technical Paper / Neutron Transport and Shielding
The Électricité de France nuclear park consists of 58 pressurized water nuclear reactors. To ensure their good performance and safety, ex-core neutron shielding studies are regularly performed. For example, neutron flux calculations in ex-core ionization chambers and pressure vessel neutron fluence studies are carried out. In the first case, ex-core ionization chambers are neutron detectors located in the reactor pit, around the reactor vessel. They are dedicated to reactor operation and core protection. In the second case, the calculation of the fast fluence (for energy >1 MeV) in the pressure vessel is used to determine its fracture toughness and integrity. To improve the fluence computations, new efficient parametric methods are under development. For these two problems, Monte Carlo transport codes such as TRIPOLI-4® allow us to perform simulations in realistic complex three-dimensional geometries and to produce reference results.
The aim of the present paper is to present together the theoretical background of our approach based on the continuous-energy Green's functions computation and storage to perform both vessel neutron fluence and ex-core chamber responses. The normalized source contribution or importance factor formalism using Green's functions computation is also described, with its associated statistical uncertainty calculation. Application examples to realistic nuclear plant configurations are given.