Abstract Basic feature of the advanced nuclear reactors is that they aim to enhance performance in terms of economy and safety of their operation. Disadvantage of many designs is that they exist only as concepts supported by calculations. However, as the non-traditional solution often rely on the combination of the materials not heavily involved in the reactor technology, question about the validity of nuclear data for these materials may arise. For example, Fluoride cooled High temperature Reactor (FHR) uses granular fuel dispersed in graphite and LiF-BeF2 coolant. In this work, neutronic calculations were performed using Serpent 2 code, employing three widely used nuclear data evaluations: ENDF/B-VII.0, ENDF/B-VIII.0, and ENDF/B-VIII.1 and compared with the reference solutions. It was demonstrated that the choice of the nuclear data library can significantly influence the results of calculations meaning that the neutronic description of the structural materials still plays role for the conceptual design of the FHR reactor, where the sensitivity of the multiplication coefficient is slightly shifted to epithermal neutron region.
Benchetrit et al. (Wed,) studied this question.
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