Abstract

This article is concerned with application of finite element model updating to a part of a secondary feed water pipeline of a VVER type nuclear power plant. A mathematical model of the pipeline was produced with finite element method. The pipeline was subjected to modal testing in three different conditions. The results of the modal testing were then compared with those predicted with the model. This comparison revealed discrepancies between these two sets of results. It is a commonly known fact that incorrect values of parameters in a model, among other things, do cause these discrepancies. In order to improve the correlation between the measured and the predicted results and hence improve the reliability of the model, the model was subjected to finite element model updating. In this updating process, incorrect values of parameters in a model are adjusted, so that a desired level of correlation between measured and predicted results is achieved. As a result, the correlation between the measured and the predicted results was improved. The bulk of the improvement was achieved by adjusting the values of those parameters that were associated with boundary conditions of the model.

Abstract

The European Commission fourth framework programme project ‘Assessment of passive safety injection systems of advanced light water reactors’ involved experiments on the PACTEL test facility and computer simulations of selected experiments. The experiments focused on the performance of passive safety injection systems (PSIS) of advanced light water reactors (ALWRs) in small break loss-of-coolant accident (SBLOCA) conditions. The PSIS consisted of a core make-up tank (CMT) and two pipelines. A pressure balancing line (PBL) connected the CMT to one cold leg. The injection line (IL) connected it to the downcomer. The project involved 15 experiments in three series. The experiments provided information about condensation and heat transfer processes in the CMT, thermal stratification of water in the CMT, and natural circulation flow through the PSIS lines. The project included validation of three thermal-hydraulic computer codes (APROS, CATHARE and RELAP5). The analyses showed the codes are capable of simulating the overall behaviour of the transients. The codes predicted accurately the core heatup, which occurred when the primary coolant inventory was reduced so much that the core top became free of water. The detailed analyses of the calculation results showed that some models in the codes still need improvements. Especially, further development of models for thermal stratification, condensation and natural circulation flow with small driving forces would be necessary for accurate simulation of the phenomena in the PSIS.