FLASH predictions of the MB-2 steamline break tests
- Westinghouse Electric Corp., West Mifflin, PA (United States)
If a main streamline 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 of 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 steamline bread (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 Electric Corporation 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 MS-2 SLB data base is accurate and of sufficient detail, to provide a valuable basis for amking comparisons relative to code predictions. Results obtained using the FLASH transient safety analysis code were found to be in excellent agreement with the data.
- OSTI ID:
- 6672514
- Report Number(s):
- CONF-921102-; CODEN: TANSAO
- Journal Information:
- Transactions of the American Nuclear Society; (United States), Vol. 66; Conference: Joint American Nuclear Society (ANS)/European Nuclear Society (ENS) international meeting on fifty years of controlled nuclear chain reaction: past, present, and future, Chicago, IL (United States), 15-20 Nov 1992; ISSN 0003-018X
- Country of Publication:
- United States
- Language:
- English
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