Thermal-hydraulic analysis of the bypass flow in boiling water reactors. Final report

This report covers research into the thermal-hydraulic characteristics of the by-pass flow in BWR's performed at MIT. The objective was to develop a better understanding of bypass region heat sources, flows, and void distributions. The term bypass in this report refers to those flow paths in the reactor between the assemblies and external to the channel walls. The heat deposited in the bypass region was divided into three sources: (1) conduction across the channel walls, (2) energy generated in the control rod by neutron absorption, and (3) energy deposited directly into the coolant by neutron moderation and gamma attenuation. The conduction source was modeled using standard heat transfer equations. The control rod heating was modeled using results from calculations using the computer code QUANDRY. The direct deposition energy was modeled using data from KENO neutronics calculations. A model was produced for estimating the bypass flowrates through statistical regression analysis of results calculated by the computer code FIBWR (Flow in Boiling Water Reactors). The resulting correlations were functions of percent total core power, percent total core flow, assembly geometry type and assembly power. The above models were coded into the computer code COBRA IIIc/MIT-2 and calculations were performed to determine the magnitude and distribution of voiding in the bypass flows