(An) assessment of two-phase critical flow models and data for applications in the safety analysis of nuclear power plants

Abstract

Six existing two-phase critical flow models have been examined for their applicability to Safety Analysis of PWR Nuclear Power Systems. They were tested analyst 20 types of experimental critical data obtained from the open literature for two-phase and subcooled stagnation conditions. The data represented several geometries and had the diverse property ranges. The two-phase critical flow models were evaluated with the available experimental data which were subdivided into a number of subsets based on the upstream (or stagnation) conditions. And a new simple correlation for two-phase critical flow discharge coefficient, which has two independent variables for subcooled stagnation conditions, has been developed in the present work by stepwise regression technique. The new correlation is again tested for its accuracy by comparing with experimental data. Results of the comparison show that the agreement between the predictions of new correlation and the experimental data is relatively good for pipes and nozzles with vertical upward flow and the correlation seems to be superior to all other models in subcooled upstream stagnation conditions, in particular for the operating conditions of PWRs.