Extension of Hybrid Micro‐Depletion Model for Decay Heat Calculation in DYN3D Code
26th Symposium of AER on VVER Reactor Physics and Reactor Safety (2016, Helsinki, Finland)
Reactor dynamics and safety analysis
Abstract
OPTIMIZATION OF MICROSCOPIC DEPLETION METHODOLOGY IN DYN3D CODE FOR REACTIVITY CALCULATIONSY. Bilodid1, D. Kotlyar21Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstraße 400, 01328 Dresden, Germany Tel.: +49 351 260 2020Fax: +49 351 260 3299 y.bilodid@hzdr.de2Department of EngineeringUniversity of Cambridge, CB2 1PZ Cambridge, United KingdomABSTRACTNodal diffusion codes such as DYN3D 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. Typically, fuel depletion is simulated by a lattice code in 2D single assembly model using core average operating conditions (moderator density, fuel temperature, control rod presence etc.).However, the local operating conditions in the core nodes may differ significantly from the core average values. XS generated using a depletion calculation under core averaged conditions neglect the local variations of the spectrum history and should be corrected. In order to account for the local spectrum history effects a number of methods, including microscopic depletion and various formulations of spectral indexes was developed and implemented in nodal codes.Recently a new hybrid method was developed and implemented in DYN3D, which combines the generalised micro-depletion correction with Pu-history indicator. The detailed nuclide content (over 1000 nuclides) is calculated by DYN3D and utilized to correct macroscopic XS, while Pu-correction is applied to the isotopic microscopic cross sections. This general approach not only accurately accounts for fuel spectral history but also provides detailed isotopic content for spent fuel inventory, radiotoxicity and decay heat applications. On the other hand, tracking of the detailed nuclide content is computationally expensive in full core burnup simulation and excessive for spectral history accounting.This paper describes an optimisation of a hybrid micro-depletion method for practical reactor simulations. The nuclide inventory estimated for each node is limited to nuclides which influence fuel reactivity. The transmutation matrix solved by DYN3D was simplified to about 200 nuclides using HELIOS 2 isotopic cross section library. HELIOS 2 was also used to generate homogenised macro- and microscopic cross section for DYN3D in demonstrated test cases. The optimised hybrid micro-depletion method was verified on various spectral history effects against HELIOS 2 reference.26th Symposium of AER on VVER Reactor Physics and Reactor Safety 40 10 – 14 October 2016, Helsinki, Finland