Izvestiya vuzov. Yadernaya Energetika

The peer-reviewed scientific and technology journal. ISSN: 0204-3327

Calculation analysis of experiments on passing neutrons through the reflector layers at the BFSs critical assemblies for complementing the lead-cooled fast reactor verification database

9/16/2020 2020 - #03 Modelling processes at nuclear facilities

Zherdev G.M. Andrianova O.N. Borovskaya J.V. Zhirnov A.P. Teplukhina E.S.

DOI: https://doi.org/10.26583/npe.2020.3.13

UDC: 621.039.51.17

The article presents the results of efforts on complementing the lead-cooled fast reactor verification database by analyzing and revising previously performed calculation and experimental studies on passing neutrons through the layers of the steel reflector carried out in different years on the IPPE BFS critical facilities. In particular, the considered data can be used for the assessment of uncertainties in the energy release in the lead-cooled fast reactor steel reflector. It was analyzed that experiments at the BFS-66 assembly on modeling neutron and photon fluxes in fast reactor shielding compositions, as well as experiments at the BFS-64 and BFS-80-2 assembles on modeling the neutrons and gamma rays transport through the layers from various materials of fast reactor reflectors. The information and data presented in previously published materials describing these experiments has been analyzed and supplemented with relevant data necessary for the preparation of detailed calculation models for precision neutronic codes. Based on the updated and refined data, detailed, precise neutronic calculation models with a detailed specification of the heterogeneous structure of the BFS facilities and experimental devices have been developed and calculations have been carried out confirming their operability. The calculations were performed using Monte Carlo neutronic codes (MCU-BR, MCNP, MMK-RF, MMK-ROCOCO) with the BNAB-RF, MDBBR50 and ROSFOND nuclear data libraries of the main neutronic characteristics measured at the BFS-66, -64, -80-2 critical assemblies. The developed calculation models of the considered integral experiments can be used to justify the designs of lead-cooled fast neutron reactors, to verify the neutronic codes and nuclear data, to evaluate uncertainties in reactor characteristics associated with nuclear data.


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integral experiments BFS Monte Carlo codes ROSFOND ABBN MCU-BR MMK-RF MMK-ROCOCO