Accounting for Spectral History Effects with improved microscopic depletion in DYN3D code

Nodal diffusion codes such as DYN3D are used routinely for nuclear reactor simulations. These codes obtain homogenized few-group macroscopic reaction cross sections (XS) of coarse-mesh space elements (nodes) from XS-libraries, which are generated using lattice neutron transport codes. The libraries represent the dependence of homogenized XS on operational parameters such as fuel temperature, moderator density, moderator temperature, boron concentration and fuel burnup.

Typically, XS libraries are calculated using a branching procedure in which a 2D fuel assembly is first depleted to a certain burnup using core average operating conditions. Then, the branching calculations are performed at predetermined burnup points for all expected combinations of operating conditions. However, the local operating conditions (moderator density, fuel temperature, control rod presence etc.) in the core nodes may differ significantly from the core average conditions. Therefore, XS generation using a single assembly depletion calculation under core averaged conditions neglects the local variations of the spectrum history and may lead to errors in the XS. In order to account for the local spectrum history effects, a new XS correction method was recently developed and implemented in DYN3D. The method utilizes the local Pu-239 concentration as an indicator of spectral history. Pu-correction was verified in a wide range of spectral conditions. However, it is not able to reproduce fuel reactivity changes due to outage periods.

This paper presents a new hybrid method developed and implemented in DYN3D, which utilizes advantages of both, the micro-depletion correction and Pu-correction. The macroscopic XS are corrected using local concentrations of the most neutronically important nuclides, which are calculated by DYN3D using fast and accurate CRAM method. The isotopic microscopic cross sections and macroscopic transport and scattering XS are corrected applying Pu-correction methodology. General applicability of the proposed method is demonstrated on various fuel types and spectral condition, including BWR and PWR unit cells with UOX and MOX fuel.