Assessment of RAMONA-3B methodology with oscillatory flow tests

Abstract

The instability event at the LaSalle County Plant (GE BWR-5) imposed a new challenge on the computer codes available for reactor transient analysis. While the codes were originally designed to predict non-oscillatory transients, the new requirement on the code is to model limit cycle oscillations with large amplitudes, where feed-back effects from the core and the balance of plant, and the nonlinear effects are significant. Two of the United States Nuclear Regulatory Commission's (USNRC) computer codes, namely RAMONA-3B/MODO [1] and HIPA-BWR of Engineering Plant Analyzer [2] were expected, and are shown in part in this paper, to meet the above demands. The RAMONA-3B/MOD1 has now been upgraded from the RAMONA-3B/MODO. It has a three dimensional neutron kinetics model, coupled to multi-channel nonequilibrium drift-flux formulation, and an explicit integration scheme for the thermal hydraulics. The accuracy of the thermohydraulics in the RAMONA-3B code was assessed for the new application by modelling oscillatory transients in the FR1GG test facilty. Nodalization studies showed that twenty-four axial nodes are sufficient for a converged solution; calculations with twelve axial nodes produce, in comparison to the 24-node calculation, the deviation of 4.4% in the peak gain of the power to flow transfer function. The code predicted correctly the effects of power and inlet subcooling on the transfer function gain and the system resonance frequency. For the six available tests modeled, the code-predicted peak gain differs from the experimentally obtained gain on the average by +7%, with the standard deviation of ±30%. The uncertainty in the experimental data lies between −11% and +12%. The difference between predicted and measured frequency at the peak gain on the average is −6%, with the standard deviation of ±14%.