VENUS-2 MOX-fuelled reactor dosimetry benchmark calculations at VTT

VTT, the Technical Research Centre of Finland, has participated in to the on-going international blind benchmark on 3-D VENUS-2 MOX-fuelled reactor dosimetry calculations with the deterministic discrete-ordinate code TORT and the stochastic Monte Carlo code MCNP. Calculations have also been performed by in-house code MultiTrans, which is a deterministic 3-D radiation transport code under development. For both deterministic transport codes, the BUGLE-96 cross-section library was used. MCNP cross-sections were taken from the endf60 library based on ENDF/B-VI. For calculation of ⁵⁸Ni(n,p), ¹¹⁵In(n,n’), ¹⁰³Rh(n,n’), ⁶⁴Zn(n,p), ²³⁷Np(n,f), and ²⁷Al(n,α) responses, the IRDF-90 version 2 dosimetry cross-section library was used. With MCNP, this data was directly utilised in the SAND-II scheme, but for deterministic codes the data was condensed into the 47 BUGLE neutron groups. Corresponding fission flux values have been calculated for all the VENUS-2 detector positions. The comparison of the calculated fission fluxes between TORT and MCNP shows rather good agreement: 75 % of the values agree within 10 %. The maximum difference between TORT and MCNP results is 26 % for fission flux values of the ²⁷Al(n,α) reaction inside the water gap. The MultiTrans results, on the other hand, show very large discrepancies inside the neutron pad when compared to the other codes, with 32 % maximum difference between MultiTrans and TORT, and 40 % between MultiTrans and MCNP. The disagreement inside the neutron pad was to some extent anticipated due to more approximative radiation transport method used in MultiTrans. The measured values are at the moment not yet open to the participants, and a comparative analysis between the calculated and measured values will remain as a future work.