One-Group Cross-Section Generation for Monte Carlo Burnup Codes: Multigroup Method Extension and Verification


One-Group Cross-Section Generation for Monte Carlo Burnup Codes: Multigroup Method Extension and Verification

Kotlyar, D.; Fridman, E.; Shwageraus, E.

Allowing Monte Carlo (MC) codes to perform fuel cycle calculations requires coupling to a point depletion solver. In order to perform depletion calculations, one-group (1-g) cross sections must be provided in advance. This paper focuses on generating accurate 1-g cross section values using Monte Carlo transport codes.
The proposed method is an alternative to the conventional direct reaction rate tally approach, which requires substantial computational effort. The method presented here is based on the multi-group (MG) approach, in which pre-generated MG sets are collapsed with MC calculated flux. In our previous studies, we showed that generating accurate 1-g cross sections requires their tabulation against the background cross section (σ0) to account for the self-shielding effect in unresolved resonance energy range. However, in previous studies, the model used for calculation of σ0 was simplified by relying on user specified Bell and Dancoff factors.
This work demonstrates that 1-g values calculated under the previous simplified model assumptions may not always agree with the directly tallied values. More specifically, the assumption is not universally applicable to the analysis of wide spectrum of reactor systems and may be inaccurate when the number of energy groups is reduced (i.e. from tens of thousands to hundreds of groups). Therefore, the original background cross section model was extended by implicitly accounting for the Dancoff and Bell factors. The method developed here reconstructs the correct value of σ0 by utilizing statistical data generated within the MC transport calculation by default. The proposed method was implemented in BGCore code system. The 1-g cross section values generated by BGCore were compared with those tallied directly from the MCNP code. Very good agreement in the 1-g cross values was observed. The method does not carry any additional computational burden and it is universally applicable to the analysis of thermal as well as fast reactor systems. Adopting this MG methodology, which accounts for self-shielding, allows generating highly accurate cross sections even for significantly reduced number of energy groups (hundreds vs. tens of thousands). This reduction considerably improves the computational efficiency which makes feasible the analysis of large scale reactor problems.

Keywords: Coupled Monte Carlo codes; BGCore; Multi group; One-group cross sections; Background cross section

Permalink: https://www.hzdr.de/publications/Publ-20182
Publ.-Id: 20182