Investigation of Reflood Quenching Phenomena in LWR Type SMRs

Abstract

With recent developments and emergence of the light water reactor (LWR) type Small Module Reactor (SMR), there has been an increasing demand for evaluation of different thermal-hydraulic characteristics of these type of reactors. Considering the commonly adapted integral design feature in LWR type SMR designs, the thermal-hydraulic characteristics may be significantly different than conventional LWR designs, especially Pressurized Water Reactor (PWR) type SMRs. Additionally, it is well known that due to the inherently safe design principle, the establishment of Natural Circulation (NC) is one of the most critical safety features for all advanced reactors, including most of the LWR type SMRs. Experimental tests have shown that one of the greatest challenges to the establishment of NC under accident conditions is the downward quenching phenomena near the exit of the fuel core as well as the potential countercurrent flow limiting (CCFL) phenomena that might develop between rising void in the boiling hot channels and the downward traveling subcooled water in the cold surrounding channels (such as low power or guide tube or instrumentation channels). Integral SMRs, in particular, have an increased probability of these phenomena occurring due to the high upper plenum to reactor core length ratio. The large volume of unheated coolant in the upper plenum will be one of the key driving forces of the downward quenching phenomena and is one of the major differences between SMR and traditional PWR designs. The highly non-uniform power distributions, both axially and radially, tend to help promote such localized downward reflood quenching phenomena as well. This paper will provide clarification and verification of the reflooding phenomena with CFD modeling and recent experimental observations, while also exploring the impacts and potential risks related to the reflood quenching phenomena in the integral designed LWR type SMRs. To further investigate these impacts, first the difference between the experimental phenomena, in particular the ones observed during conventional PWR rod bundle CHF tests using uniform heaters, and the actual reactor design characteristics must be explored, particularly with respects to measurement uncertainty, thermal-hydraulic instability, and operation safety. Second, the role of mixing vane spacer grid is assessed with relation to the exit quenching phenomena. Different spacer grid designs are examined with respect to their functions in promoting or diminishing the downward quenching phenomena as well as their roles in the establishment of natural circulation. In addition, the impacts of downward reflood quenching on CHF (Critical Heat Flux) level, the location of CHF, and the potential of leading to flow instability are also examined.