Bifurcation Analysis of Nuclear-Coupled Thermal Hydraulics of BWR Using BIFDD

Stability and bifurcation analyses of nuclear-coupled thermal hydraulic instability in BWR has been performed using a semi-analytical method. The BWR model used in this study consists of three parts: neutron kinetics, fuel rod heat conduction and single and two-phase heated channel thermal hydraulics. Point reactor model is currently being used for neutron kinetics and will be extended in the future to higher order lambda or omega-mode. In the heat conduction part, a piecewise quadratic approximation to radial temperature distribution in fuel pellet and cladding is assumed. ODEs for the expansion coefficients of the quadratic spatial profiles are developed by applying variational principle. Similar to the heat conduction model, the spatial enthalpy distribution in the single phase region and steam quality in the two-phase region in the BWR core are approximated by quadratic polynomials. Two-phase flow is modeled using the homogeneous equilibrium model. A bifurcation analysis code, BIFDD, is then used to perform the analysis for the stability boundary (SB) and the nature of Poincare´-Andronov-Hopf bifurcation (PAH-B). Results in control-rod-induced-reactivity—inlet-subcooling-number space show that both super or sub-critical bifurcation can occur along the SB—the subcritical bifurcation occurs for very small or very large subcooling number values; super-critical PAH-B occurs for intermediate values of subcooling number.