Authors & Affiliations
Doronkov D.V., Dmitriev S.M., Demkina T.D., Dobrov A.A., Doronkova D.S., Pronin A.N., Ryazanov A.V., Solntsev D.N.
Nizhny Novgorod State Technical University n.a. R.E. Alekseev
Abstract
The article presents the results of experimental and computational studies of the coolant flow in the fuel bundle of the mixed core of a WWER-type reactor. The aim of the work is to study the redistribution of the transverse and axial flow velocity of the coolant during the flow of grids of different heights of various designs with significantly different hydraulic resistance coefficients. To achieve this goal, a series of experiments was carried out on an aerodynamic research stand on scale models of various fragments of a fuel bundle of a WWER-type reactor mixed core and computational modeling of the cross-flow coolant formation between adjacent TVSA-T of different designs using the thermal-hydraulic calculation code KORSAR. The picture of the coolant flow is represented by axial velocity distribution cartograms, as well as transverse velocity distribution graphs of the flow between adjacent TVSA-Ts of various designs. An analysis of the results of experimental studies made it possible to identify the main regularities in the formation of the flow in the WWER-type reactor mixed core. A high degree of agreement between the results of computational and experimental modeling of cross-flow coolant flows was obtained, which confirms the possibility of using the KORSAR thermal-hydraulic code in substantiating the reliability and safety of mixed cores operation. The resulting database of experimental data can be used for local validation of modern domestic and foreign CFD programs, as well as programs for core thermal-hydraulic calculation.
Keywords
nuclear reactor, core, fuel assembly, fuel
rod, spacer grid, mixing grid, combined spacer grid, coolant hydrodynamics,
cross flows
Article Text (PDF, in Russian)
UDC 621.039