FLASH predictions of the MB-2 steam line break tests

If a main steam line from a pressurized water reactor (PWR) steam generator were to rupture, the effect would be a depressurization of the secondary side and a consequential overcooling transient on the primary side. Analyses must accurately predict the effects of the rapid cooldown of the reactor vessel coolant on positive nuclear-kinetic reactivity feedback to the core plus thermal shock to the reactor vessel and other primary system components. Many early studies of the steam line break (SLB) transient made extremely conservative assumptions to maximize the primary to secondary heat transfer which in turn maximized the reactor vessel cooldown rate. Among the more significant of these assumptions was that flow from the break was pure steam and that the tube bundle remained covered until the secondary mass inventory was significantly reduced. The Model F commercial PWR steam generator testing performed in the Model Boiler No. 2 (MB-2) facility located at the Westinghouse Engineering Test Facility in Tampa, Florida provided data to better qualify the actual variation in these key parameters. A conclusion of this analysis is that the MB-2 steam line break data base is accurate and of sufficient detail to provide a valuable basis for making comparisons relative to code predictions. Results obtained using the FLASH transient safety analysis code were found to be in excellent agreement with the data.