Effects of Parallel Channel Interactions, Steam Flow, Liquid Subcool and Channel Heat Addition on Nuclear Reactor Reflood Transients

Abstract

Tests were performed to examine the effects of parallel channel interactions, steam flow, liquid subcool and channel heat addition on the delivery of liquid from the upper plenum into the channels and lower plenum of Boiling Water Nuclear Power Reactors during reflood transients. Early liquid delivery into the channels, following a loss of Coolant Accident, will help prevent overheating and melt down of the reactor fuel bundles. The tests were performed at the General Electric Nuclear Energy Division Laboratory, California. The channels consisted of two 5.22m long *25.4mm long*23.6mm i.d. stainless steel tubes, with unequal orificing at the bottom, and equal orificing at the top. Provisions were made for electrical resistance heating of 3.5m of each tube, and for visual observation of flows through the tubes. Test fluids were steam and saturated or subcooled water. Subcools ranged from 3.3 deg C to 37.2 deg C, and system pressures varied from near atmospheric to a little over 1.7 bar. Test section heat fluxes were between 2.58 and 13.95 KW/m2. It was observed that channel heat additions tended to make each tube behave independently of the other. As a result of subcool and vapour condensation, vapour supply into the lower plenum increased liquid delivery into the channels, and decreased the system rewet and reflood times when the subcool was in excess of about 20 deg C. Parallel channel interactions were observed to produce co-current downflow in the less restricted tube, with counter-current flow existing in the more restricted tube. This is desirable. When conditions permitted, the interactions gave rise to the classical "steam bound" flow configuration - (i.e. water hold up in the upper plenum due to top orifice Counter Current Flow Limitation, partial filling of the more restricted channels, a partially full lower plenum, and pure vapour flow in the less restricted channel). This configuration is undesirable for thermal recovery of a reactor following a loss of coolant accident.