Research On Debris-Loading Head-Loss Tests For Fuel Assembly During Loss Of Coolant Accidents

During the long term core cooling (LTCC) period of the postulated loss-of-coolant accident (LOCA) in PWR, debris may generate in the containment due to dynamic propagation of the pressure wave followed by the ensuing steam/water-jet impingement. A fraction of this fragmented materials will be transported to the emergency core cooling system (ECCS) sump by the break-induced steam/water flows and the containment sprays. Most of the debris can be collected by the recirculation sump screen, however, it still leaves the possibility that some fragments might get access to the primary system through the bypass and thereafter challenge the core flow path and cause the head-loss across the fuel assemblies. It can be indicated from relevant researches that such phenomenon is critical for the LTCC of the nuclear power plant. The U.S. Nuclear Regulation Commission (NRC) issues a regulation that the commercial nuclear power plants should maintain long-term cooling of the reactor core after LOCA scenario. The National Nuclear Safety Administration (NNSA) issues a notification about the reform and renovation of the sump screen. NNSA also regulates that all the operating plants should verify the recycling function of the ECCS can meet the standards and regulations when there are impurities being entrained in coolant.Based on the background stated above, the necessity of studying the phenomenon is fully expounded firstly in this thesis, and the experimental parameters are listed according to the characteristic of LTCC for Qinshan power plant. A large thermal-hydraulic test loop should be constructed to support this research. One key step for the construction is to determine the ranges of the sensors. Therefore, numerical simulation has been carried out by Computational Fluid Dynamics (CFD), and some empirical formulas have been investigated in this paper to estimate the measuring range of the sensors. It can also lay the theoretical foundation for the comparison to the experimental data. In addition, this paper describes details about the test loop and the experimental cases. Bench mark test, sensitivity test and limiting test has been carried out, and the quantitative pressure data are obtained and rational analysis is provided.The proceeding of the project referred in this paper will provide reliable and quantitative data for Qinshan nuclear power plant downstream effect analysis and safety analysis program. It can also provide strongly support for the efficient operation of ECCS and further provide a guarantee for the plant to pass the nuclear safety review.