Buoyancy-driven flow excursions in fuel assemblies

Technical Report · Fri Sep 01 04:00:00 EDT 1995

DOI:https://doi.org/10.2172/107010· OSTI ID:107010

Laurinat, J. E.; Paul, P. K.; Menna, J. D. ^[1]

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1. Westinghouse Savannah River Company, Aiken, SC (United States)

A power limit criterion was developed for a postulated Loss of Pumping Accident (LOPA) in one of the recently shut down heavy water production reactors at the Savannah River Site. These reactors were cooled by recirculating heavy water moderator downward through channels in cylindrical fuel tubes. Powers were limited to safeguard against a flow excursion in one of more of these parallel channels. During-full-power operation, limits safeguarded against a boiling flow excursion. At low flow rates, during the addition of emergency cooling water, buoyant forces reverse the flow in one of the coolant channels before boiling occurs. As power increased beyond the point of flow reversal, the maximum wall temperature approaches the fluid saturation temperature, and a thermal excursion occurs. The power limit criterion for low flow rates was the onset of flow reversal. To determine conditions for flow reversal, tests were performed in a mock-up of a fuel assembly that contained two electrically heated concentric tubes surrounded by three flow channels. These tests were modeled using a finite difference thermal-hydraulic code. According to code calculations, flow reversed in the outer flow channel before the maximum wall temperature reached the local fluid saturation temperature. Thermal excursions occurred when the maximum wall temperature approximately equaled the saturation temperature. For a postulated LOPA, the flow reversal criterion for emergency cooling water addition was more limiting than the boiling excursion criterion for full power operation. This criterion limited powers to 37% of the limiting power for previous long-term reactor operations.

• Research Organization: Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Technology; American Nuclear Society, La Grange Park, IL (United States); American Inst. of Chemical Engineers, New York, NY (United States); American Society of Mechanical Engineers, New York, NY (United States); Canadian Nuclear Society, Toronto, ON (Canada); Japan Society of Multiphase Flow, Kyoto (Japan)
• OSTI ID: 107010
• Report Number(s): NUREG/CP--0142-Vol.1; CONF-950904--Vol.1; ON: TI95017077
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
DIAGRAMS
EXCURSIONS
EXPERIMENTAL DATA
FLUID FLOW
FUEL ASSEMBLIES
HEAVY WATER
HYDRAULICS
LOSS OF FLOW
PRODUCTION REACTORS
REACTOR SAFETY
SAFETY ANALYSIS
SAVANNAH RIVER PLANT