Fast neutron flux and fluence outside the fissile zone of PWR fuel assembly calculation methodology using MCNP Monte Carlo code and science package

As one of the major world fuel vendors, Framatome ANP is deeply involved in the continuous evolution of PWR fuel design. In this framework, fast neutron flux (and/or fluence) is an important parameter to evaluate the mechanical behaviour of a fuel assembly material inside and outside the fissile zone. In the fissile zone, the neutron flux can be calculated by Framatome ANP's SCIENCE code package. But this is not the case outside the fissile zone. This paper presents a calculation method developed by the neutronics department of the Framatome ANP Fuel Sector. For a given fuel cycle, this method is able to provide the best estimate fast neutron flux and fluence profiles outside the fissile zone (assembly top and bottom) for all fuel assemblies. The reference three-dimensional Monte Carlo code MCNP is used to calculate the fast neutron flux attenuation using a heterogeneous model describing in detail the fuel assembly and surrounding components. Results from a set of sensitivity analyses, also completed by MCNP, constitute a data base for further application. Framatome ANP 's SCIENCE core design package is used to give best estimate information for the studied cycle: power level, fission neutron source, and moderator density of each assembly. A Man/Machine Interface procedure, via the COPILOTE management system of the SCIENCE code package, is developed to interface the MCNP results and the SCIENCE reactor cycle model to easily obtain the fast neutron flux and fluence. After a brief presentation of MCNP and SCIENCE, the calculation method and the man/machine interface procedure are described in detail. The analysis of calculation results, sensitivity and uncertainty are also presented. (authors)