Validation Data and Model Development for Fuel Assembly Response to Seismic Loads

Earthquakes and Loss of Coolant Accidents (LOCA) can damage nuclear power plants cores, which is a major safety concern and set safety margins. Of both types of commercial LWR operational in the U.S., PWR cores are especially sensitive to seismic transients. During an earthquake, the fuel assembly spacer grids can collide, which may in turn result in buckling of guide tubes - preventing control rods insertion- and in bending of the fuel rods - creating contacts between cladding and fuel pellets, and hot spots. The main objective of the present work is to provide comprehensive data characterizing the dynamics of the fluid and the structure in PWR fuel assemblies under seismic loads. The data will enable validating the US DOE SHARP virtual reactor simulation suite on this important multi-physics problem. To achieve this objective an international 'contributory' approach will be used: a series of experiments will be conducted in parallel at the George Washington University (GW) and at the Commissériat à l'Energie Atomique (CEA) in France. The data obtained in these experiments will be used to develop and validate computational models of various levels of fidelity at GW, Argonne National Lab (ANL), and the CEA. Under simulated seismic and operating plant conditions, GW will develop new instruments and conduct highresolution, spatio-temporal channel velocity and fuel rod displacement measurements in a 6x6 fuel rod bundle.The CEA experiments on the other hand, will focus on the interaction of multiple reduced fuel assemblies (i.e.2x2 matrix) at lower resolution. At ANL and GW, the fuel bundle data will be used to develop and validate high-fidelity multi-physics numerical tools for fully-coupled fluid-structure models as implemented in the ANL Reactor Integrated Performance and Safety Code (IPSC) component of the U.S. Department of Energy's Nuclear Energy Advanced Modeling and Simulation Program (NEAMS). Successful completion of the project will greatly benefit safety of existing and future nuclear reactors by: i) introducing new experimental techniques to the field of flow induced vibrations, and providing an extensive database of high-dimensional experimental data suitable for validating computational approaches of various levels of fidelity, preserved in NE-KAMS; ii) gaining new insights of the multi-scale, multi-physics phenomena governing the response of fuel assemblies to seismic loading; iii) developing predictive tools on core response to fast transients encountered during earthquakes. This will permit refining safety margins on seismic loads in current and future PWR. These developments will also be beneficial to nuclear fuel reliability, such as grid-to-rod fretting failure (GTRF), which is responsible for nearly 3/4 of PWRs fuel rods failure and is critical to power uprates and burnup increase; this aspect will also benefit CASL and EPRI.