Experimental and Analytical Modeling of Natural Circulation and Forced Circulation BWRs: Thermal-Hydraulic, Core-Wide, and Regional Stability Phenomena

Experimental and Analytical Modeling of Natural Circulation and Forced Circulation BWRs: Thermal-Hydraulic, Core-Wide, and Regional Stability Phenomena

Furuya, M.

Van der Hagen, T.H.J.J. (promotor)

Applied Sciences

2006-04-24

Currently, 434 nuclear power plants are in operation worldwide. 21% of them are known as Boiling Water Reactors (BWRs). These BWRs have pumps that cool their reactor cores (the forced circulation BWRs). In the design of new BWRs, ways to cool the core by a natural circulation flow, without pumps, also called natural circulation BWRs, are being considered. A possible disadvantage of natural circulation BWRs might be their susceptibility to instabilities, which could then lead to both flow and power oscillations. We distinguish between pure thermal-hydraulic stability - where the fission power is assumed to be constant - and coupled thermalhydraulic-neutronic stability - where the two-phase mixture in the core influences the fission chain reaction. The thermal-hydraulic stability of a prototypical natural circulation BWR (the ESBWR) has been investigated with the SIRIUS-N facility, which behaves, due to proper scaling, similarly to the ESBWR in its thermal-hydraulics. As a result of experiments, two distinct responses are found in the relation of the oscillation period to the time that liquid passes in the chimney, indicating that the driving mechanisms of the instabilities are different for low and high pressures. At low pressures (0.1 - 0.5 MPa), the observed instability was shown to be caused by flashing induced density wave oscillations. At high pressures (1.0 - 7.2 MPa), the observed instability was shown to be caused by density waves known as Type-I oscillations, since the void fractions in the chimney inlet and exit are out of phase, and the instability occurs at low quality. In order to simulate coupled thermalhydraulic-neutronic stability, void fractions in reactor core sections of the thermal-hydraulic loop were measured, and used for a realtime simulation of the modal point kinetics of reactor neutronics. Experiments are conducted for both a natural circulation BWR (the ESBWR) and a forced circulation BWR (the ABWR) configurations. These stability experiments showed that the ESBWR and the ABWR have a significant stability margin to the thermalhydraulic-neutronic instability.

boiling water reactor
natural circulation
regional stability
dynamics
time-series analysis
thermalhydraulic-neutrnoic coupling

http://resolver.tudelft.nl/uuid:9122b230-3bde-4e3e-877e-ab7e02b97120

IOS

1-58603-605-X

Institutional Repository

doctoral thesis

(c) 2006 M. Furuya