Structural integrity assessment of secondary support structure and reactor pressure vessel following drop impact of core barrel assembly

Abstract This work presents a study on hypothetical drop accident of the core barrel assembly in a pressurized water reactor, in which structural dynamic responses of the secondary support structure and the reactor pressure vessel during impact are investigated numerically. The analysis results are conservative enough by considering the drop impact accident in an air environment. And the tensile test was conducted for the Inconel 690 material of secondary support structure in order to input real material property in simulations. During the impact, the peak impact force and impact velocity of the falling core barrel assembly are respectively 19,881 kN and 0.627 m/s. With respect to stress distributions, peak magnitudes of effective stress and maximum principal stress are both concentrated at the top edge of connection location between the secondary support structure and the reactor pressure vessel. For both of these components, their effective stresses exceed the corresponding yield strengths while the maximum principal stresses are less than their ultimate tensile strengths, indicating that they deform plastically in this drop impact accident. The impact energy absorbed by the reactor pressure vessel is much larger than that by the secondary support structure. With respect to a single support key, the first and second peak strain energy are concentrated in the connection location and contact location respectively while the free end is almost unaffected. And three concerned through-thickness paths are defined for stress linearization in which membrane stress and bending stress are separated. Stress results meet the design criteria of the ASME Boiler and Pressure Vessel Code with respect to stress limits, indicating that both of the secondary support structure and the reactor pressure vessel maintain their structural integrity following drop impact of the core barrel assembly. This investigation makes a contribution to the safety analysis of the pressurized water reactor.