Abstract
Reactor pressure vessel flow field characterization is an important part of the thermo-hydraulic design of nuclear power plants, which is related to the design of the reactor internal structure, the core inlet flow distribution and the operational safety of nuclear power plants; this paper takes the Siemens KONVOI pressurized water reactor as the research object, and carries out the modeling and large basin simulation calculations for the upper plenum, the lower plenum, the outlet section, the inlet section and the core of the RPV, and adopts the porous medium model for fuel assembly level simulation of the core; Siemens KONVOI reactor-type core operation data and ROCOM experimental data are selected for numerical validation of this study. The salt tracer distribution and cross-mixing were simulated in the form of solving additional scalar equations. The results show that the coolant flow pattern is consistent with the experimental results of the ROCOM unit. Comparison with the calculation results of the subchannel procedure on this basis verifies the accuracy of the thermal parameters of the core and ensures the validation of the established CFD calculation model.
In this paper, the reasonableness of the calculation results of the low plenum and the core is verified by comparing with the existing data, and a polyhedral mesh division scheme suitable for RPV calculations is given to analyse the coolant flow cross-mixing characteristics under different operating conditions, and the calculation results show that the low plenum of the RPV can cross-mix the coolant and distribute the flow rate efficiently, and the flow rate of the external core inlet channel is relatively large, and the coolant flow in the vertical projection of the perforated drum (flow rate distribution device) is relatively large. The flow in the core inlet channel at the vertical projection of the perforated drum (flow distribution device) is relatively small, and the flow distribution in the central area of the core inlet channel is relatively uniform; finally, this paper investigates the transient response characteristics of the flow in the pressure vessel under the accidental condition, and the thermal-hydraulic characteristics of the KONVOI pressurised water reactor under the three-loop operation condition are investigated. The flow distribution characteristics at the core inlet and outlet under asymmetric loop flow conditions are summarised, which provides certain reference value for the optimal design and safety analysis of the reactor.
