Nuclear coupled flow instability study for natural circulation BWR startup transient

Natural circulation Boiling Water Reactor (BWR) startup transient was investigated in Purdue University Multidimensional Test Assembly (PUMA) facility based on a natural circulation BWR design. Strategy and results of the experiments, which consider the effects of void-reactivity and fuel heat conduction time constant, are discussed. Total reactivity is treated to be composed of two components: external reactivity due to control rod motion and void-reactivity. A detailed analysis for heat conduction problem is performed to derive dimensionless groups. Based on area-averaged heat conduction equations for pellet and clad regions, Fourier and Biot numbers are derived to simulate wall heat flux response. Power transient, which has been used for startup transient investigation without void-reactivity feedback is used to derive the control rod reactivity. Twelve conductivity probes are used to measure local void fraction inside core at three axial locations. The local void-fraction data is used to calculate volume average void fraction, which is used to calculate the voil-reactivity. A real-time Point Kinetic Model solver is implemented to PUMA heater power control program to determine power transient during startup. The results demonstrate that the inclusion of void-reactivity feedback worsen the scenario for startup instabilities and may cause large amplitude neutron flux oscillations. (author)