2019

Häkkinen, Silja

A simplified benchmark based on #2670 ISTC PIE has been calculated. The benchmark describes the irradiation history of eight samples cut out of four fuel pins of a VVER-440 fuel assembly irradiated to the average burnup of 38.5 MWd/kgU. The purpose of the work was to validate Serpent’s burnup calculation methods for burnup credit applications and to repeat earlier calculations of the same benchmark because of an error in the fuel definition.<br/><br/>Nuclide concentrations [kg/Uinit] for nuclides recommended in actinide and fission product burnup credit evaluation were calculated and compared to measurement values. The calculations were performed using three different cross section libraries ENDF/B-VII.1, JEFF-3.2 and JENDL-4.0. The different libraries yielded rather consistent results except for Ag-109 with JEFF-3.2 for which the results differed 10 % from the results with the other libraries. The<br/>calculations corresponded rather well to the measurement results for U-238, U-236, most plutonium isotopes and Am-241. Large differences (&gt; 100 %) were observed in Ag-109, Eu-151, Eu-153 and Gd-155. For the other nuclides, differences were mostly of the order of 1-10 %, but larger differences were observed in some nuclides like U-234, Np-237, Pu-238 and samarium isotopes. Some of the differences can be explained by measurement uncertainty. Some differences may also arise from the way the sample specific power values have been determined in the benchmark specifications.<br/><br/>Some sensitivity calculations were also performed using different burnup related modelling options and algorithms. Differences to reference calculations were rather small. The use of substep method and TTA method had a notable effect of approximately 1-2 % to the nuclide concentrations. Spatial discretization had a small effect of 0,5 - 1,5 % for some actinides and Gd-155.