Abstract
The development and deployment of gas-cooled reactors (GCRs) require flexible modelling and simulation tools for a detailed understanding of the system behavior in normal operation and accidental scenarios. Amongst the identified safety-relevant scenarios for the gas-cooled reactors, outlet plenum flow distribution during normal operations was ranked to be of high importance with a low knowledge level in the phenomenon identification and ranking table (PIRT). The present paper contributes to the development of accurate simulation models using stand-alone and coupled codes for the GCR lower plenum region, which are needed to address nuclear reactor safety issues early enough so that delays in licensing and facility construction can be minimized. For the first time in this paper, an assessment of the predictive capabilities of coupled system-thermalhydraulics and CFD codes against heat transfer measurements in the lower plenum of a prismatic block GCR is reported. The lower plenum mixing experiments of the integral high temperature test facility at the Oregon State University (OSU HTTF) were simulated in the stand-alone and coupled mode to predict the lower plenum mixing. The coupled-code analysis used a 3D CFD code STAR-CCM + and a 1D thermalhydraulics code RELAP5-3D to account for system-level feedback on the flow and heat transfer for lower plenum mixing occurring during the normal operation at the reduced power and/or mass flow rate. For one of the simulated OSU test conditions, the existing models in STAR-CCM + and RELAP5-3D predicted the measured core temperatures reasonably well at two different elevations, suggesting that the available models are adequate for such an analysis. Through this study, a better understanding of the suitability of the models in the CFD and system thermalhydraulics (TH) codes is gained for more accurate prediction of flow and heat transfer in GCRs in anticipation of the eventual need to perform safety and licensing analyses.
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