Development of an Advanced PWR Reactor Internals System Finite Element Model for Flow-Induced Vibration Analyses

Abstract

The reactor vessel internals (RVI) are located within the reactor vessel, which is part of the reactor coolant system (RCS) loop in Westinghouse nuclear plant designs. Historically, the coolant passing through the RCS loop has been highly turbulent and has generated significant turbulence-induced excitation (TE) for the RVI. In an effort to analytically quantify the response of RVI structures due to TE for new designs, a methodology is employed which combines both first principle concepts as well as operational experience. As part of this process, TE-induced forcing functions are developed based upon the numerous flow fields around the components of interest. After the TE-induced forcing functions are developed, they are applied to a system finite element model (SFEM), in a transient dynamic finite element analysis to capture dynamic system-level interactions. To benchmark the response of the numerical model, both narrowband and broadband model responses are compared to empirical data extrapolated from model-scale flow-induced vibration test results. This comparison shows a strong agreement between the empirical data and the SFEM, validating that the dynamic response and system interactions of RVI structures due to TE can be accurately characterized through numerical simulation of the system.

Efficient global sensitivity analysis for flow-induced vibration of a nuclear reactor assembly using Kriging surrogates

“…Correspondingly, the vibration analysis program establishes expected measurements (predictions) and associated acceptance criteria, which are based on the material fatigue life. Extensive verification and empirical validation of numerical models, for both forcing functions (e.g., turbulence and acoustic phenomena) and structural response, are employed to develop the predictions (Palamara, et al, 2015).…”