Journal of Nuclear Fuel Cycle and Waste Technology

In this study, ORIGEN-2 one-group cross-section libraries were developed for prism-type VHTR whole-core models to enable accurate and efficient depletion and decay heat calculations. These libraries were generated through continuous-energy Monte Carlo depletion simulations using the McCARD code. Two computational approaches were compared. The Core Depletion (CD) approach - previously developed and used as a reference - performs whole-core depletion with McCARD, followed by decay heat calculation using ORIGEN-2. In contrast, the Point Depletion (PD) approach applies McCARD-generated one-group cross-sections within ORIGEN-2 for both depletion and decay heat analysis. Among the library-generation methods evaluated, the batch-wise library - constructed via localized tallying from whole-core depletion - yielded nuclide inventories and decay heat results in close agreement with those of the CD approach. Its improved accuracy stems from its ability to capture the soft neutron spectrum induced by large graphite reflectors and the detailed burnup history - features not properly represented in single-block models or in whole-core depletion with unified tallying. Notably, the batch-wise method enabled precise prediction of long-term actinide decay heat and achieved high accuracy in estimating major source-term nuclide inventories, with relative errors remaining below 5% in most cases when compared to the reference solution, offering a validated and efficient framework for VHTR core analysis, safety evaluation, and spent fuel management.

Keywords

VHTR, ORIGEN-2, One-group cross-section, Point depletion, Core depletion, Decay heat